Three Mile Island
Level 2 Containment Event Tree Analysis
PRA Notebook
TMI-PRA-015.2 REVISION 0
2004 PRA Model Rev.2
(TM1042)
April 2007
RISK MANAGEMENT TEAM
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Revisions:
REV. DESCRIPTION PREPARER/DATE REVIEWER/DATE APPROVER/DATE
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TABLE OF CONTENTS
Section Paqe
1.0 CONTAINMENT EVENT TREE ................................................................... 1-1
2.0 INTRODUCTION.......................................................................................... 2-1
2.1 CONTAINMENT EVENT TREE DESCRIPTION..................................... 2-1
3.0 CONTAINMENT EVENT TREE TOP EVENTS............................................ 3-1
4.0 TOP EVENT DECISION TREE MODELS.................................................... 4-1
4.1 INTRODUCTION .................................................................................... 4-1
4.2 CONTAINMENT BYPASS DECISION TREE ......................................... 4-3
4.3 EARLY CONTAINMENT FAILURE DECISION TREE............................ 4-3
4.3.1...... Direct Containment Heating...................................................... 4-4
4.3.2...... Rapid Steam Generation .......................................................... 4-4
4.3.3...... Combustible Gas Burns ............................................................ 4-5
4.3.4...... Direct Corium Contact............................................................... 4-6
4.3.5...... Missiles in Containment ............................................................ 4-7
4.4 LATE CONTAINMENT FAILURE DECISION TREE............................... 4-8
4.4.1...... Late Steam Overpressurization ................................................ 4-8
4.4.2...... Late Combustible Gas Burn...................................................... 4-9
4.4.3...... Late Non-Condensable Gas Overpressurization....................... 4-9
4.5 EX-VESSEL FISSION PRODUCT RELEASE DECISION TREE............ 4-9
4.6 BASEMAT MELT-THROUGH DECISION TREE .................................. 4-10
4.7 FISSION PRODUCT REVAPORIZATION DECISION TREE ............... 4-11
4.8 FISSION PRODUCT SCRUBBING DECISION TREE.......................... 4-12
5.0 CONTAINMENT EVENT TREE QUANTIFICATION .................................... 5-1
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5.1 INTRODUCTION .................................................................................... 5-1
5.2 ANALYSIS PERFORMED ...................................................................... 5-8
5.2.1...... Containment Capacity............................................................... 5-8
5.2.2...... Combustible Gas Burns ............................................................ 5-9
5.2.3...... Reactor Cavity Geometry........................................................ 5-10
5.2.4...... MAAP Model ........................................................................... 5-16
5.2.5...... Containment Base Pressure ................................................... 5-17
5.3 DECISION TREE BASIC EVENT QUANTIFICATION .......................... 5-18
5.4 REMOVAL OF ILLOGICAL CUTSETS VIA RECOVERY RULES....... 5-133
6.0 SOURCE TERM CALCULATIONS AND RELEASE CATEGORY
DEFINITIONS .............................................................................................................. 6-1
6.1 INTRODUCTION .................................................................................... 6-1
6.2 MAAP COMPUTER MODEL .................................................................. 6-1
6.2.1...... MAAP NODALIZATION ............................................................ 6-1
6.2.2...... SAFETY SYSTEMS MODELED IN MAAP................................ 6-3
6.3 RELEASE CATEGORY PARAMETER ANALYSIS................................. 6-3
6.4 RELEASE CATEGORY DEFINITIONS................................................... 6-4
6.4.1...... RELEASE CATEGORY DISCUSSION ..................................... 6-5
6.5 FINAL BINNING OF RELEASE CATEGORIES.................................... 6-46
7.0 CONTAINMENT EVENT TREE SOLUTION ................................................ 7-1
7.1 TREATMENT OF ILLOGICAL CUTSETS............................................... 7-4
8.0 REFERENCES............................................................................................. 8-1
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Table of Figures
Section Paqe
Figure 3-1 Nodal Logic for CET Events B and NOTB................................................... 3-2
Figure 3-2 Nodal Logic for CET Events C and NOTC ................................................. 3-3
Figure 5-2 Logic for Event EFFDEPRESS and NOEFFDEPRESS ........................... 5-36
Figure 5-3 Logic for Event NOSPARK....................................................................... 5-63
Figure 5-4 Logic for Event SPARK............................................................................ 5-64
Figure 5-5 Logic for Event NOSPARKAFT................................................................ 5-66
Figure 5-6 Logic for Event SPARKAFT ..................................................................... 5-67
Figure 5-7 Logic for Event NOSPARKLT .................................................................. 5-69
Figure 5-8 Logic for Event SPARKLT........................................................................ 5-70
Figure 5-9 Logic for Events PDSFANS And NOPDSFANS....................................... 5-75
Figure 5-10 Logic for Events PDSINJECCS and NOPDSINJECCS.......................... 5-77
Figure 5-11 Logic for Events PDSINJSP and NOPDSINJSP .................................... 5-78
Figure 5-12 Logic for Event PDSLOW....................................................................... 5-80
Figure 5-13 Logic for Event NOPDSLOW ................................................................. 5-81
Figure 5-14 Logic for Event PDSNOISL .................................................................... 5-83
Figure 5-15 Logic for Event NOPDSNOISL............................................................... 5-84
Figure 5-16 Logic for Event PDSNOSGTR ............................................................... 5-86
Figure 5-17 Logic for Event NOPDSNOSGTR .......................................................... 5-87
Figure 5-18 Logic for Events PDSPRESSH and NOPDSPRESSH........................... 5-89
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Figure 5-19 Logic for Events PDSPZRPORV and NOPDSPZRPORV...................... 5-90
Figure 5-20 Logic for Events PDSRCEQSG and NOPDSRCEQSG ......................... 5-92
Figure 5-21 Logic for Events PDSRCPWR and NOPDSRCPWR .............................. 5-93
Figure 5-22 Logic for Events PDSSGADV and NOPDSSGADV ................................ 5-95
Figure 5-23 Logic for Events PDSSPRAY and NOPDSSPRAY ................................. 5-96
Figure 5-24 Logic for Event PDSSSHR...................................................................... 5-98
Figure 5-25 Logic for Event NOPDSSSHR ................................................................ 5-99
Figure 5-26 Logic for Events PZPORVCONF and NOPZPORVCONF .................... 5-104
Figure 5-27 Logic for Events RECOVFANSLT and NORECOVFANSLT ................. 5-108
Figure 5-28 Logic for Events RECOVSPLT and NORECOVSPLT........................... 5-112
Figure 5-29 Logic for Event RELLOC....................................................................... 5-115
Figure 5-30 Logic for Event NORELLOC ................................................................. 5-116
Figure 5-31 Logic for Events SEQPRESSH and NOSEQPRESSH.......................... 5-121
Figure 5-32 Logic for Events SEQPRESSL and NOSEQPRESSL........................... 5-123
Figure 5-33 Logic for Events SGREL and NOSGREL.............................................. 5-125
Figure 5-34 Recovery Rules Logic to Exclude Illogical Cutsets Regarding the
Reactor Building Spray System......................................................................... 5-135
Figure 7-1 Model Logic Used to Exclude Non-Realistic Cutsets Associated with
Reactor Building Spray.......................................................................................... 7-4
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Table of Tables
Table Paqe
Table 3-1 Containment Event Tree Top Events ........................................................... 3-8
Table 5 aSSIGNMENT OF NUMERICAL VALUES TO VERBAL DESCRIPTORS ..... 5-2
Table 5-2 TMI-1 CET Basic Event Descriptions ........................................................... 5-2
Table 6-1 Representative Sequence Descriptions for Source Term Groups.............. 6-48
Table 6-2 Summary of Representative MAAP Sequences for TMI-1 Source
Terms.................................................................................................................. 6-50
Table 6-3 TMI-1 Source Term Summary.................................................................... 6-52
Table 6-4 Release Category 1.01 Containment Bypass, Outside the Auxiliary
Building, without Ex-Vessel Release of Fission Products, with Fission
Product Scrubbing............................................................................................... 6-57
Table 6-5 Release Category 1.02 Containment Bypass, Outside the Auxiliary
Building, without Ex-Vessel Release of Fission Products, without Fission
Product Scrubbing............................................................................................... 6-57
Table 6-6 Release Category 2.01 Containment Bypass, to the Auxiliary
Building,without Ex-Vessel Release of Fission Products, with Fission Product
Scrubbing ............................................................................................................ 6-58
Table 6-7 Release Category 2.02 Containment Bypass, to the Auxiliary Building,
without Ex-Vessel Release of Fission Products, without Fission Product
Scrubbing ............................................................................................................ 6-59
Table 6-8 Release Category 2.03 Containment Bypass, to the Auxiliary Building,
with Ex-Vessel Release of Fission Products, with Fission Product Scrubbing .... 6-60
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Table 6-9 Release Category 2.04 Containment Bypass, to the Auxiliary Building,
with Ex-Vessel Release of Fission Products, without Fission Product
Scrubbing ............................................................................................................ 6-60
Table 6-10 Release Category 3.01 Large Isolation Failure, to the Auxiliary
Building, without Ex-Vessel Release of Fission Products, with Fission
Product Scrubbing............................................................................................... 6-62
Table 6-11 Release Category 3.02 Large Isolation Failure, to the Auxiliary
Building, without Ex-Vessel Release of Fission Products, without Fission
Product Scrubbing............................................................................................... 6-63
Table 6-12 Release Category 3.03 Large Isolation Failure, to the Auxiliary
Building, with Ex-Vessel Release of Fission Products, with Fission Product
Scrubbing ............................................................................................................ 6-64
Table 6-13 Release Category 3.04 Large Isolation Failure, to the Auxiliary
Building, with Ex-Vessel Release of Fission Products, without Fission
Product Scrubbing............................................................................................... 6-65
Table 6-14 Release Category 3.05 Large Isolation Failure, Outside the Auxiliary
Building, without Ex-Vessel Release of Fission Products.................................... 6-66
Table 6-15 Release Category 3.06 Large Isolation Failure, Outside the Auxiliary
Building, with Ex-Vessel Release of Fission Products......................................... 6-67
Table 6-16 Release Category 4.01 Small Isolation Failure, to the Auxiliary
Building, without Ex-Vessel Release of Fission Products, with Fission
Product Scrubbing............................................................................................... 6-68
Table 6-17 Release Category 4.02 Small Isolation Failure, to the Auxiliary
Building, without Ex-Vessel Release of Fission Products, without Fission
Product Scrubbing............................................................................................... 6-69
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Table 6-18 Release Category 4.03 Small Isolation Failure, to the Auxiliary
Building, with Ex-Vessel Release of Fission Products, with Fission Product
Scrubbing ............................................................................................................ 6-70
Table 6-19 Release Category 4.04 Small Isolation Failure, to the Auxiliary
Building, with Ex-Vessel Release of Fission Products, without Fission
Product Scrubbing............................................................................................... 6-71
Table 6-20 Release Category 4.05 Small Isolation Failure, to the Environment,
without Ex-Vessel Release of Fission Products, with Fission Product
Scrubbing ............................................................................................................ 6-72
Table 6-21 Release Category 4.06 Small Isolation Failure, to the Environment,
without Ex-Vessel Release of Fission Products, without Fission Product
Scrubbing ............................................................................................................ 6-73
Table 6-22 Release Category 4.07 Small Isolation Failure, to the Environment,
with Ex-Vessel Release of Fission Products, without Fission Product
Scrubbing ............................................................................................................ 6-74
Table 6-23 Release Category 4.08 Small Isolation Failure, to the Environment,
with Ex-Vessel Release of Fission Products, without Fission Product
Scrubbing ............................................................................................................ 6-75
Table 6-24 Release Category 5.01 Early Containment Failure, without Ex-Vessel
Fission Product Release ..................................................................................... 6-76
Table 6-25 Release Category 5.02 Early Containment Failure, with Ex-Vessel
Fission Product Release ..................................................................................... 6-77
Table 6-26 Release Category 6.01 Late Overpressurization, with Catastrophic
Containment Failure, without Ex-Vessel Fission Product Release, without
Revaporization, with Fission Product Scrubbing ................................................. 6-78
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Table 6-27 Release Category 6.02 Late Overpressurization, with Catastrophic
Containment Failure, without Ex-Vessel Fission Product Release, without
Revaporization, without Fission Product Scrubbing ............................................ 6-79
Table 6-28 Release Category 6.03 Late Overpressurization, with Catastrophic
Containment Failure, without Ex-Vessel Fission Product Release, with
Revaporization, with Fission Product Scrubbing ................................................. 6-80
Table 6-29 Release Category 6.04 Late Overpressurization, with Catastrophic
Containment Failure, without Ex-Vessel Fission Product Release, with
Revaporization, without Fission Product Scrubbing ............................................ 6-81
Table 6-30 Release Category 6.05 Late Overpressurization, with Catastrophic
Containment Failure, with Ex-Vessel Release of Fission Products, without
Revaporization, with Fission Product Scrubbing ................................................. 6-82
Table 6-31 Release Category 6.06 Late Overpressurization, with Catastrophic
Containment Failure, with Ex-Vessel Release of Fission Products, without
Revaporization, without Fission Product Scrubbing ............................................ 6-83
Table 6-32 Release Category 6.07 Late Overpressurization, with Catastrophic
Containment Failure, with Ex-Vessel Release of Fission Products, with
Revaporization, with Fission Product Scrubbing ................................................. 6-84
Table 6-33 Release Category 6.08 Late Overpressurization, with Catastrophic
Containment Failure, with Ex-Vessel Release of Fission Products, with
Revaporization, without Fission Product Scrubbing ............................................ 6-85
Table 6-34 Release Category 7.01 Late Overpressurization, with Benign
Containment Failure, without Ex-Vessel Fission Product Release, with
Fission Product Scrubbing .................................................................................. 6-86
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Table 6-35 Release Category 7.02 Late Overpressurization, with Benign
Containment Failure, without Ex-Vessel Fission Product Release, without
Fission Product Scrubbing .................................................................................. 6-87
Table 6-36 Release Category 7.03 Late Overpressurization, with Benign
Containment Failure, with Ex-Vessel Release of Fission Products, with
Fission Product Scrubbing .................................................................................. 6-88
Table 6-37 Release Category 7.04 Late Overpressurization, with Benign
Containment Failure, with Ex-Vessel Release of Fission Products, without
Fission Product Scrubbing .................................................................................. 6-89
Table 6-38 Release Category 8.01 Containment Failure from Basemat Melt-
Through, with Ex-Vessel Release of Fission Products ........................................ 6-90
Table 6-39 Release Category 9.01 No Containment Failure, without Ex-Vessel
Fission Product Release, with Fission Product Scrubbing .................................. 6-91
Table 6-40 Release Category 9.02 No Containment Failure, without Ex-Vessel
Fission Product Release, without Fission Product Scrubbing ............................. 6-92
Table 6-41 Release Category 9.03 No Containment Failure, with Ex-Vessel
Fission Product Release, with Fission Product Scrubbing .................................. 6-93
Table 6-42 Release Category 9.04 No Containment Failure, with Ex-Vessel
Fission Product Release, without Fission Product Scrubbing ............................. 6-94
Table 7-1 Release Category Frequencies (Items Listed in Bold are Contributors
to LERF)................................................................................................................ 7-1
Table 7-2 Truncation Limit Comparison for Certain Release Categories ..................... 7-3
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1.0 CONTAINMENT EVENT TREE
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2.0 INTRODUCTION
The purpose of the Containment Event Tree (CET) is to quantify containment failure
modes and radionuclide releases. Any phenomena that have a significant effect on the
radionuclide release fractions or the timing, energy, and duration of the release are
included in the tree as a top (header) event. The core damage sequences were
categorized into Plant Damage States (PDSs), as determined in the Level 1 analysis in
TMI PRA Notebook TMI-PRA-015 (Reference 24). These core damage sequences are
treated as initiating events for the CET. The paths that the PDSs can take through the
event tree depend on how they affect the various events modeled. Because the path
taken at each top event is based on probabilities and system fault tree evaluations,
each PDS will appear at more than one CET end point with varying frequency. Thus,
each end point can have more than one PDS state contributing to its total frequency.
The methodology for the CET solution, the CET quantification, and source term
development were based on the TMI IPE Level 2 analysis of 1993, which was originally
based on the Oconee PRA Level 2 analysis. Oconee and TMI-1 designs were
compared to identify any significant differences in plant characteristics. Then, the
Oconee CET model and its quantification were modified to reflect these differences, as
well as develop a plant specific model for TMI-1.
2.1 CONTAINMENT EVENT TREE DESCRIPTION
Containment event trees have become so complex that the CETs can not be easily
represented and are difficult to understand by anyone other than the consequence
analyst. The approach used for the TMI-1 analysis (as with the previous Oconee
analysis) relies on converting the large and complex CET into a combination of a small
event tree and large decision trees.
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In developing the TMI-1 small CET, the only questions included are those that have an
effect on the release timing, energy, location, or fission product fractions. When
completed, each CET end state represented a separate release category. The CET
release category results are presented in Section 7.0.
After the event tree was developed, decision trees using both success and failure logic
were developed to determine the probability of the appropriate top (header) event in the
CET. This approach was used to avoid the use of NOT gates for sequence success
logic, which tended to make the model more complicated and difficult to quantify.
The CET developed for TMI-1 consists of 11 nodal top events that were modeled via the
use of Boolean logic, for both success and failure of each branch. The following section
defines and describes the CET top (header) events and their associated decision trees.
The top (header) events are summarized in Table 3-1, the event tree is shown in Figure
2-1, and the decision tree models are provided in Appendix A. A cross-reference is
provided for each decision tree to facilitate following decision tree logic. Each cross-
reference lists the basic events, gates, and their respective location in the associated
decision tree. Decision tree basic event descriptions are given in Table 5-2.
To make use of the CET, the important characteristics of the plant's containment must
be identified. Three of the more important features that must be considered are the
containment ultimate strength capacity, the concrete type, and the reactor cavity
arrangement.
The ultimate capacity of containment provides the basis for establishing containment
failure probability and failure modes given various accident progression scenarios. TMI-
1, like Oconee, is a Babcock & Wilcox PWR with vertical straight-tube (once-through)
steam generators that produce superheated steam at constant pressure. The reactor
and the nuclear steam supply system are contained within a Reactor Building that is a
post-tensioned reinforced-concrete cylinder and dome. The interior of the surface of the
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building is lined with a one-quarter inch thick welded steel plate to ensure a high degree
of leak tightness.
Generally, TMI-1 and Oconee can be placed into the category of PWR large dry
containments, because of their high mean failure pressure, overall containment volume,
and open lower containment configuration.
The type of concrete affects the type and properties of gases released during concrete
attack. TMI-1's concrete contains a limestone aggregate, which can result in significant
non-condensable gas production during concrete ablation.
The reactor cavity geometry affects how (or if) water can reach the cavity during a core
damage sequence. TMI-1's reactor cavity and the region around the reactor vessel are
very similar to Oconee's arrangement. The reactor cavity geometry is discussed in
more detail in Section 5.2.3. The cavity arrangement is important when considering the
following phenomena:
• Ex-vessel debris bed coolability
• Potential for direct containment heating
• Ex-vessel steam generation
• Ex-vessel hydrogen or combustible gas production
• Ex-vessel fission product release
• Hydrogen or combustible gas recombination
• Long-term containment overpressurization
• Basemat melt-through
• Potential for debris-liner contact
• Sources of water and pathways to the lower reactor cavity
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Figure 2-1
Containment Event Tree
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3.0 CONTAINMENT EVENT TREE TOP EVENTS
In this section, the CET top events are defined and described. The CET top events are
summarized in Table 3-1.
A: Containment Bypass
Does the release of radionuclides take place within the containment?
Success for this event means that containment is available as a barrier to fission
product release. Failure means containment is not available as a barrier to fission
product release. The types of accidents that bypass the containment are steam
generator tube ruptures (as an initiating event or an induced event) and interfacing-
systems LOCA. This top event is further developed using a decision tree model.
Section 4.2 discusses the containment bypass decision tree model.
B: Containment Isolation
Does the containment isolate such that: 1) a leakage rate sufficient to cause a
substantial increase in radionuclide release to the environment does not occur,
and 2) containment pressure response is not significantly affected?
Success for this event means that containment isolation performs its function so that
containment becomes a barrier against flow of radionuclides to the environment.
Failure means containment integrity is lost and a path is available for radionuclides to
reach the environment. This event is concerned with the time at the beginning of the
accident sequence (i.e., when isolation occurs) before radionuclides are released to the
containment atmosphere. Containment isolation can be determined directly from the
top events contained in the system fault tree models described in Reference 25. Figure
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3-1 shows how system top events are linked to the nodal logic of this portion of the
CET.
Figure 3-1 Nodal Logic for CET Events B and NOTB
C: Isolation Failure Size
Is the isolation failure equivalent to a small hole size in containment?
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Success for this event means that the isolation failure is small, i.e., system top event
SMALL-ISO (see Figure 3-1). For the Oconee and TMI-1 analysis, a small isolation
failure is defined as a six-inch equivalent diameter hole. Isolation failures of this type
allow some time for holdup inside containment where natural removal mechanisms
(e.g., plateout) will reduce radionuclide concentrations. Failure of this event implies that
the isolation failure is not small, i.e., system top event LARGE-ISO (Figure 3-1), and
allows little or no holdup in containment.
Both small and large isolation failures preclude late overpressurization. All other
containment overpressure sequences (hydrogen burns, direct containment heating, etc.)
are prevented only by large isolation failures. The size of the isolation failure can be
determined directly from the top events SMALL-ISO and LARGE-ISO, which were shown
above in Figure 3-1. See Figure 3-2 for the logic depicting events C (small isolation failure)
and NOTC (large isolation failure).
Figure 3-2 Nodal Logic for CET Events C and NOTC
D: Auxiliary Building Release
Does the fission product release pass through the Auxiliary Building?
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Success for this event means that the fission product release will pass through the
Auxiliary Building. This release path is the result of an interfacing-system LOCA or an
isolation failure to the Auxiliary Building. Failure for this event means that the fission
product release does not pass through the Auxiliary Building. A release path that
bypasses the Auxiliary Building is a pathway directly to the environment.
As shown in Figure 2-1, this top event is applicable only if containment is not isolated or
is bypassed. Determination of success or failure depends on the type of isolation
failure, where the fission products are released, and the PDS. For example, a SGTR
would be a failure, while most interfacing systems LOCAs would be a success.
E: Early Containment Failure
Does the containment remain intact until long after reactor vessel failure (i.e., a
time period which allow sufficient time for fission product settling)?
Success for this event means that containment remains intact long after reactor vessel
failure. Failure for this event means that containment has failed prior to or within the
time required for fission product settling and decay of short-lived isotopes. This time
period is typically defined as five hours after reactor vessel failure.
This top event is further developed using a decision tree model. Section 4.3 discusses
the early containment failure decision tree model including the phenomena associated
with early containment failure.
F: Late Containment Failure
Does the containment remain intact throughout the entire core melt sequence?
Event success means that the containment remains intact throughout the entire core
melt sequence. Releases to the environment after this point, if any, are due to normal
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containment leakage or basemat melt-through. Failure of this event means that
containment fails late in the core melt sequence due to an overpressurization event.
The nodal top events for failure (LATE) and for success (NOLATE) are further
developed using decision tree models. Logic for debris bed cooling, cavity geometry, as
well as prior hydrogen burns, is taken directly from the early containment decision tree
logic for node E. Section 4.4 discusses the late containment failure nodal top events,
including the phenomena associated with late containment failure.
G: Benign Containment Failure
Is late containment failure benign?
Success for this event means that a late overpressurization results in a benign
containment failure, i.e., leak-before-break. This failure mode is described as a series
of small cracks that develop in the containment structure such that further pressurization
does not occur. Failure of this event means that a late overpressurization results in a
catastrophic containment failure, which would cause containment to depressurize
rapidly. This is strictly a function of the containment type, and is quantified identically
for all PDSs (see Section 5.3).
H: Ex-Vessel Release Of Fission Products
Is a coolable debris bed established outside the reactor vessel so that significant ex-
vessel fission product releases do not occur?
Success for this event means that a coolable debris bed is established in the reactor
cavity or the containment, preventing an ex-vessel release. Failure means that a
coolable debris bed is not established, allowing the corium to attack the concrete
(producing non-condensable gases) and resulting in an ex-vessel release. The ex-
vessel release involves a significant amount of tellurium and other fission products.
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The nodal success and failure logic is further developed using a decision tree model.
Logic for debris bed cooling and cavity geometry is taken directly from the early
containment decision tree logic for node E. Section 4.5 discusses the ex-vessel fission
product release decision tree model.
I: Containment Basemat Failure
Is a coolable debris bed established in the reactor cavity to prevent containment
failure from basemat melt-through?
Success for this event means that the debris bed in the cavity is cooled, and concrete
ablation is stopped. Failure means that the debris bed is not cooled and ablates
concrete until the basemat is failed.
This top event is further developed using a decision tree model. Logic for debris bed
cooling and cavity geometry is taken directly from the early containment decision tree
logic for node E. Section 4.6 discusses the basemat melt-through decision tree model.
J: Revaporization Release
Is a revaporization release of volatile fission products at or near the time of containment
failure prevented?
Success for this event means that large amounts of volatile fission products have not
revaporized and are not available for release when containment overpressurizes.
Failure means that volatile fission products that were deposited in the RCS have
revaporized and are available to be released in large amounts when containment fails.
Revaporization is only considered for late catastrophic containment failures. Early
containment failures release fission products at or shortly after reactor vessel failure
resulting in high release fractions. The effects of revaporization, if any, would not be
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seen for this failure mode. Late containment failures, however, provide time for
radionuclide removal from the atmosphere by various methods. As a result, release
fractions at containment failure are lower so that revaporization of fission products will
have a larger impact. Revaporization is not considered for benign failures of
containment since the pressure remains high due to the slow depressurization of
containment. Since the pressure remains high in containment, revaporization is unlikely
to occur.
This top event is further developed using a decision tree model. Section 4.7 discusses
the revaporization decision tree model.
K: Fission Product Scrubbing
Are fission product removal mechanisms available to reduce the amount of
radionuclides released to the environment?
Success for this event means that the fission products are scrubbed by some method
prior to release to the environment. These mechanisms include:
• Containment scrubbing (e.g., sprays)
• Auxiliary Building scrubbing (e.g., plateout)
• Steam Generator scrubbing (e.g., water pool release)
Failure for this event means the fission products are not scrubbed prior to release to the
environment by any method.
This top event is further developed using a decision tree model. Section 4.8 discusses
the fission product scrubbing decision tree model.
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TABLE 3-1 CONTAINMENT EVENT TREE TOP EVENTS
EVENT NODE/STATE
DESCRIPTION
A Containment Bypass
Success Containment is available as a barrier to fission product release
Failure Containment is not available as a barrier to fission product release (SGTR, ISLOCA)
B Containment Isolation
Success Containment is isolated
Failure Containment is not isolated
C Large Isolation Failure
Success Isolation failure is small
Failure Isolation failure is large
D Auxiliary Building Release
Success Fission product release is through the Auxiliary Building
Failure Fission product release does not go through the Auxiliary Building
E Early Containment Failure
Success Early containment failure does not take place
Failure Early containment failure does occur
F Late Containment Failure
Success Late containment failure does not take place
Failure Late containment failure does occur
G Benign Containment Failure
Success Containment failure is benign, i.e., leak before break
Failure Containment failure is catastrophic
H Ex-Vessel Release of Fission Products
Success Ex-vessel release is prevented
Failure Ex-vessel release is not prevented
I Containment Basemat Failure
Success Containment failure from basemat melt-through is prevented
Failure Containment failure from basemat melt-through occurs
J Revaporization Release
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TABLE 3-1 CONTAINMENT EVENT TREE TOP EVENTS
EVENT NODE/STATE
DESCRIPTION
Success Revaporization release does not take place
Failure Revaporization release does occur
K Fission Product Scrubbing
Success Fission products are scrubbed in containment, steam generator, or Auxiliary Building
Failure Fission products are not scrubbed
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4.0 TOP EVENT DECISION TREE MODELS
4.1 INTRODUCTION
The CET is the primary tool for analyzing containment phenomena that could lead to a
release of fission products to the environment. As the number of containment
phenomena increases, the CET must grow to account for each of these phenomena.
Some reports, NUREG/CR-4551 [Reference 1], have described a CET with as many as
50 top events to cover the necessary detail. A CET of this size may indeed consider the
details, but it also leads to a large, unmanageable event tree that, in most cases, cannot
be easily represented or comprehended. The CET, described in this report, presents a
methodology in which a combination of event tree and decision tree modeling is used.
This approach captures the necessary detail, but keeps the CET at a manageable size.
The decision trees represent the basic events that can lead to a particular containment
phenomenon (early containment failure, ex-vessel fission product release, etc.). The
CET basic event description used for TMI-1's decision trees are given in Table 3-1. The
CET represents top events and containment phenomena that lead to a particular
release category. The benefit of this methodology is that a decision tree can be
expanded to include more detail (i.e., with additional logic and basic events) without
causing an expansion of the CET. In turn, if a CET top event needs to be added or
deleted, only the CET structure is affected. It should be noted that deletion of a CET top
event will cause the corresponding decision tree to be deleted, and an additional top
event may require a new decision tree to be developed. Also, note that while decision
trees may be added and deleted as the CET top events are modified, dependencies
exist within the decision trees for different top events, requiring the entire CET event
sequence to be quantified by linking decision trees.
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An advantage of this decision tree model is the ability to represent the effects of one
event on another. One way in which this is incorporated into the decision tree is
through the use of both success and failure logic models to avoid, or minimize, the use
of NOT gates. For example, to represent the probability of burning hydrogen after
reactor vessel failure, the logic must include the event in which hydrogen was not
burned prior to reactor vessel failure. This event will include the complementary
Boolean logic and basic events to represent the absence of hydrogen burns prior to
reactor vessel failure.
Another method of event interaction is the use of duplicate logic. For example,
hydrogen burns after reactor vessel failure can be caused by a spark inside
containment. However, since direct containment heating (DCH) could also occur at this
time, there is a possibility of the hot molten corium, which is dispersed into the
containment atmosphere, causing the hydrogen burn.
As described in Section 3.0, seven top events that consist of logic for each success and
failure branch for each node were developed using decision tree models:
A: Containment Bypass
E: Early Containment Failure
F: Late Containment Failure
H: Ex-Vessel Release of Fission Products
I: Containment Basemat Failure
J: Revaporization Release
K: Fission Product Scrubbing
The decision trees allow the consequence analyst to describe containment phenomena
logically through a series of logic models. Instead of quantifying the occurrence of the
phenomena, the quantification takes place by questioning the PDS, containment
safeguards/isolation state (CSS/CIS), and the applicable basic events. The following
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sections describe the decision trees and the associated phenomena used in the TMI-1
CET.
4.2 CONTAINMENT BYPASS DECISION TREE
The decision tree develops a probability of preventing containment bypass for a given
core melt sequence. The two mechanisms for bypassing the containment are
Interfacing-systems LOCA (ISLOCA) and a steam generator tube rupture (SGTR).
Therefore, containment bypass may be directly indicated because the PDS is an
ISLOCA or SGTR. The decision tree also develops the possibility of inducing an SGTR
during the accident progression.
Induced SGTRs could occur due to creep rupture of the steam generator tubes. Hot
gases generated during core degradation may raise the tube temperatures sufficiently
that the primary to secondary pressure difference could rupture the tubes. Creep
rupture can be prevented by minimizing the tube temperature, keeping heat transfer to
steam generator tubes low, or lowering the pressure differential across the tubes.
Appendix A contains the decision trees for containment bypass. The basic events
included in these decision trees and their probabilities are discussed in Section 5.3.
4.3 EARLY CONTAINMENT FAILURE DECISION TREE
There are several phenomena that can cause early containment failure:
• Direct containment heating (DCH)
• Rapid steam generation (RSG)
• Hydrogen burn prior to reactor vessel failure
• Hydrogen burn at reactor vessel failure
• Combustible gas burn early after reactor vessel failure
• Direct contact of corium with the containment wall
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• Reactor shield plug missile
Decision trees describe each of these events in terms of basic events and combines
them logically to develop a probability of early containment failure for a given core melt
sequence. Appendix A contains the decision trees for early containment failure
embedded as PDF files. The basic events included in these decision trees and their
probabilities are discussed in Section 5.3. The following sections will describe the
phenomena associated with early containment failure.
4.3.1 Direct Containment Heating
This phenomenon is important for sequences in which a core melt is initiated while the
RCS is at a high pressure. It has been hypothesized that the molten corium can be
ejected, under high pressure, from the reactor vessel and be dispersed into the
containment atmosphere as finely fragmented particles. Airborne particulate debris
could then rapidly release chemical (oxidation of metallic constituents) and thermal
energy directly to the containment atmosphere.
The gates used in the decision trees for success and failure are as follows:
NODCH: Containment failure from direct containment heating is prevented
DCH: Containment failure from direct containment heating
4.3.2 Rapid Steam Generation
This phenomenon is important for sequences in which water is present in the
containment at reactor vessel failure. Interaction of the molten corium with the water
pool can cause large amounts of steam to be generated quickly leading to a
containment overpressurization. Methods by which water can reach the reactor cavity
are discussed in Section 5.2.3.
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The gates used in the decision trees for success and failure are as follows:
NORSG: Containment failure from rapid steam generation is prevented
RSG: Containment failure from rapid steam generation
4.3.3 Combustible Gas Burns
Hydrogen or combustible gas burns are another method by which high pressure can be
generated to fail containment. Three burn events are considered in the early
containment failure decision tree. The first is a hydrogen burn that occurs prior to
reactor vessel failure, the second is a hydrogen burn that occurs immediately after
reactor vessel failure, and the third is a hydrogen or combustible gas burn that occurs
shortly after reactor vessel failure. The combustible gas burns prior to and at reactor
vessel failure are primarily hydrogen burns since there is no production of other
combustible gases, such as carbon monoxide, due to concrete attack. Combustible gas
burns at the later times assumes no hydrogen or combustible gas burns at an early time
(i.e., only one combustible gas burn is assumed to occur). For a combustible gas burn
to occur, a spark (or ignition source) must be present, and the containment must not be
inerted, such as with steam.
The generation of carbon monoxide, a combustible gas, due to corium-concrete
interaction is also considered. Since TMI-1's concrete contains a limestone aggregate,
non-condensable gases (e.g., carbon monoxide, carbon dioxide, and hydrogen) are
generated during concrete ablation. MAAP adds the concentration of carbon monoxide
to the concentration of hydrogen in containment (i.e., treating all combustible gases the
same), effectively increasing the total concentration of hydrogen in containment
available for a burn.
The gates used in the decision trees for success and failure of hydrogen burns prior to
reactor vessel failure are as follows:
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H2PRI: Containment failure from H2 burns before reactor vessel failure is prevented
NOH2PRI:
Containment failure from H2 burns before reactor vessel failure
The gates used in the decision trees for success and failure of hydrogen burns
immediately after reactor vessel failure are as follows:
H2AT: Containment failure from H2 burns at reactor vessel failure is prevented
NOH2AT:
Containment failure from H2 burns at reactor vessel failure
The gates used in the decision trees for success and failure of combustible gas burns
shortly after reactor vessel failure are as follows:
H2AFTER: Containment failure from combustible gas burns after reactor vessel failure is prevented
NOH2AFTER:
Containment failure from combustible gas burns after reactor vessel failure
4.3.4 Direct Corium Contact
This phenomenon is important for sequences in which RCS pressure is high at reactor
vessel failure. If the reactor cavity geometry allows sufficient corium dispersal, it may
be possible for the corium to come into direct contact with the walls of containment.
Direct corium contact is highly dependent on reactor cavity geometry (see Section
5.2.3).
The gates used in the decision trees for success and failure are as follows:
NOCONTACT: Containment failure from direct contact of corium is prevented
CONTACT: Containment failure from direct contact of corium
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4.3.5 Missiles in Containment
Missile generation mechanisms in containment, which could lead to early release of a
significant fraction of the core fission product inventory due to a containment failure, are:
• Alpha mode failure
• Reactor vessel becoming a rocket
• Pressure-generated missiles
The alpha mode failure of containment is postulated to result from in-vessel steam
explosions. However, a review of IDCOR Technical Report 85.2 [Reference 19]
indicates the probability of this event as remote. The probability that the reactor vessel
becomes a rocket and impinges into the containment vessel is equally as unlikely. For
the reactor vessel to become a projectile within containment, the reactor vessel
anchorage must fail. Based on NUREG/CR-4551 [Reference 1], the probability of this
event is also considered remote. Basic events, NORVROCKET and NOALPHA, are
discussed in Section 5.3.
Pressure-generated missiles, such as the reactor shield plugs identified at TMI-1, are
also important for sequences in which RCS pressure is high at reactor vessel failure. A
feature of the TMI-1 containment is the use of shield plugs (steel canisters containing
sand) around the reactor vessel upper head. The Oconee containment has analogous
concrete shield plugs. These shield plugs are held in place by their own weight, so that
if a significant differential pressure is present across the plugs, they may become
airborne. If ejected with sufficient velocity, the shield plugs might damage the
containment liner or other equipment (e.g., Reactor Building Cooling Units). Analysis
for Oconee has shown that following a high pressure reactor vessel failure, the reactor
shield plugs could become missiles that have the potential for striking the containment
walls. However, based on the analysis of shield plugs at Oconee, the potential plug
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trajectory is insufficient to reach the containment wall. Basic event MISSLE is
discussed in Section 5.3.
The gates used in the decision trees for success and failure are as follows:
NOMISSLE: Containment failure from missiles is prevented
MISSLE: Containment failure from missiles
4.4 LATE CONTAINMENT FAILURE DECISION TREE
There are several phenomena that can cause late containment failure:
• Late steam overpressurization
• Late combustible gas burn
• Late non-condensable gas overpressurization
Appendix A contains the decision trees for late containment failure embedded as PDF
files. The basic events included in these decision trees and their probabilities are
discussed in Section 5.3. The following sections will describe the phenomena
associated with late containment failure.
4.4.1 Late Steam Overpressurization
Late overpressurization is the result of sequences in which molten corium is ejected into
the containment (without failing the containment early) and continues to boil off water (if
available) until the ultimate strength of the containment is reached (if fans are
unavailable).
The gates used in the decision trees for success and failure are as follows:
NOSTEAM: No containment failure from steam generation
STEAM: Containment failure from steam generation
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4.4.2 Late Combustible Gas Burn
Hydrogen or other combustible gas (e.g., carbon monoxide) burns late in the sequence
after reactor vessel failure can cause a late containment failure. The gates used in the
decision trees for success and failure are as follows:
NOH2LATE: No Containment failure from late combustible gases
H2LATE: Containment failure from late combustible gases
4.4.3 Late Non-Condensable Gas Overpressurization
Interactions of the molten corium with the concrete can potentially produce significant
amount of non-condensable gases. The amount of non-condensable gases produced
due to concrete corium interaction is dependent on the material composition of the
concrete. Generally, the higher the limestone content, the more non-condensable
gases are produced. TMI-1's containment is constructed of concrete with limestone
aggregate or fill. Therefore, overpressurization due to non-condensable gases was
modeled in the TMI-1 late containment failure decision tree.
The production of non-condensable gases can lead to late containment
overpressurization. Significant non-condensable gas generation can only occur if the
reactor cavity is dry and corium-concrete interaction is allowed to occur.
The gates used in the decision trees for success and failure are as follows:
NOGASES: No containment failure from non-condensable gases
GASES: Containment failure from non-condensable gases
4.5 EX-VESSEL FISSION PRODUCT RELEASE DECISION TREE
Release of fission products ex-vessel occurs when core-concrete interaction takes
place. It is therefore necessary to provide water to the containment in order to cool the
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debris bed. Since the RCS inventory alone is not sufficient to obtain a cooled debris
bed, water must be provided by some other means (e.g., BWST). Since a static source
of water will only temporarily suspend concrete attack, a continuous source of water is
required to terminate the core-concrete interaction.
Another consideration, however, is whether the water (in whatever quantity) will be able
to reach the debris bed and penetrate sufficiently to provide long-term cooling.
Therefore, it is necessary to analyze the geometry of the cavity to determine if water
can reach the debris bed. The reactor cavity geometry is discussed in Section 5.2.3.
Once present, the water may still be ineffective for cooling if a crust forms between the
molten core and the water. If a crust forms and prevents water penetration, the water
pool may still be effective in scrubbing some of the fission products that are released.
Gaseous fission products, which sparge through cracks in the crust into the overlying
pool, may be scrubbed prior to release to the containment atmosphere. Mechanisms
for fission product scrubbing by an overlying water pool are discussed in further detail in
Section 4.8.
Appendix A contains the decision trees for ex-vessel fission product release embedded
as PDF files. The basic events included in these decision trees and their probabilities
are discussed in Section 5.3.
4.6 BASEMAT MELT-THROUGH DECISION TREE
The phenomenon of basemat melt-through is also heavily reliant on large amounts of
water being present in the containment. As with ex-vessel fission product releases, the
availability of water and its ability to cool the debris bed will determine if the corium will
erode the concrete basemat.
In order to fail containment through the basemat, a large amount (more than six feet) of
concrete must be eroded. If the corium pool spreads out over a large enough area,
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then there may not be enough corium in one location to sufficiently heat the concrete
and melt through the basemat. Also, a large thin layer of corium is much easier to cool
with a water pool.
Appendix A contains the decision trees for basemat melt-through embedded as PDF
files. The basic events included in these decision trees and their probabilities are
discussed in Section 5.3.
4.7 FISSION PRODUCT REVAPORIZATION DECISION TREE
Fission product revaporization is an important phenomenon due to the added amount of
fission products available for release at containment failure. This phenomenon is
governed by the ability of the fission products to plateout in the RCS and any
mechanism available to prevent their revaporization. Some of the mechanisms for
preventing revaporization are: total heat loss from the RCS, chemical form of the fission
products, or availability of other heat sinks to "capture" released radionuclides. As
discussed in Section 4.8, the solubility of fission products, vapor nucleation,
condensation, and particle coagulation are dependent on temperature gradients
between the fission product aerosols and potential heat sinks, NUREG/CR-4727
[References 9 and 17].
Revaporization is dependent on the temperature and pressure in containment and the
RCS. Revaporization typically occurs at high temperatures and low pressures
[Reference 19]. For instance, the availability of secondary side heat removal prevents
revaporization since it provides a heat sink for plateout within the cooler steam
generators. Temperatures are generally low in the local area where the plateout
occurs. Similarly, for a benign failure of containment, the containment pressure remains
high due the slow depressurization of containment; therefore, revaporization is unlikely
to occur, whereas revaporization is likely to occur given a catastrophic failure of
containment. During a catastrophic failure of containment, rapid depressurization
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occurs precipitating the revaporization of fission products in containment. Post-accident
releases are assessed until asymptotic behavior is observed, thus accounting fully for
any revaporization that occurs.
Appendix A contains the decision trees for fission product revaporization embedded as
PDF files. The basic events included in these decision trees and their probabilities are
discussed in Section 5.3.
4.8 FISSION PRODUCT SCRUBBING DECISION TREE
The phenomenon of fission product scrubbing is important in reducing the amount of
fission products that are released to the environment. Several mechanisms exist to
provide scrubbing both inside and outside containment.
Inside the containment, the main removal mechanism is the containment sprays. The
degree of scrubbing credit depends on containment spray water chemistry, as well as
droplet size. In general, large droplets are less effective for fission product removal
than are small droplets. Even for small isolation failures providing a release path
directly to the environment, some credit is taken for scrubbing due to the sprays.
Scrubbing can also take place within the steam generator during a tube rupture, or in
the Auxiliary Building for interfacing-systems LOCAs and containment isolation failures.
Fission products in the form of I2 and CsI have a high affinity for water due to their
solubility in water. Therefore, a substantial amount of I2 and CsI would be retained in
the primary system or containment sump water. The solubility of these fission products
is dependent on pH and length of time in contact with water [Reference 9]. The
alkalinity enhances the scrubbing effectiveness of the containment sprays [Reference
18]. NUREG/CR-4727 [Reference 17] describes mechanisms affecting the scrubbing
effectiveness of overlying water pools. The three primary mechanisms are: 1) diffusion
of aerosol particles to bubble walls, 2) sedimentation of particles within bubbles, 3)
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impact of aerosol particles on bubble walls due to their inability to follow the bubble
stream. MAAP utilizes these and other principles to calculate the effectiveness of
fission product scrubbing by overlying water pools in the reactor cavity and by water in
the steam generator [Reference 4]. The SUPRA code, which is a subroutine to MAAP,
utilizes a lookup scheme to determine the decontamination factors (DFs) for various
isotopes based on parameters such as location of break and depth of overlying pool.
The SUPRA code is limited by the data provided in the lookup tables. For instance,
SUPRA does not accurately model shallow pools. Since SUPRA tends to overestimate
the scrubbing DFs and is unable to accurately model shallow pools, a conservative
scrubbing factor of five is utilized to envelope all expected conditions.
Appendix A contains the decision trees for fission product scrubbing embedded as PDF
files. The decision trees FPSCRUBBED and FPNOSCRUBBED represent the logic
used for those accident scenarios that do not include bypass of containment. The trees
FPSCRUBBED2 and FPNOSCRUBBED2 contain the logic applicable only to
containment bypass scenarios, which excludes those scrubbing mechanisms applicable
to non-bypass scenarios. The basic events included in these decision trees and their
probabilities are discussed in Section 5.3.
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5.0 CONTAINMENT EVENT TREE QUANTIFICATION
5.1 INTRODUCTION
In this section, probabilities are assigned to basic events within the decision trees that
are used in the containment event tree. Probabilities are assigned based on the
resulting PDS for each core damage sequence. For some events, the probability will be
the same for all PDSs (i.e., independent of the PDS); in many cases, the assigned
probabilities are conditional on the events that are accounted for in the PDS (e.g.,
availability of sprays, availability of power, etc.). These probabilities are based on
information gathered from containment phenomenology codes such as MAAP, hand
calculations, previous studies, and other literature.
Based on this information, combined with an understanding of the characteristics for
each PDS, an estimate of the likelihood for each event is developed. Many basic
events are characterized with a verbal descriptor, with Table 5-1 then being used to
assign a numerical value. The basic event descriptions are provided in Table 5-2.
Quantification is also dependent on the PDS designation. The PDS is composed of a
core melt bin (CMB), a containment safeguards state (CSS), and a containment
isolation state (CIS). The PDSs are described with two designators. The first
designator (Table 3-1 of Reference 24) is a number designator from 1 to 19 to describe
the CMB. The CSS and CIS are combined in the second designator, which is a letter
designator from A to R (Table 3-13 of Reference 24). Thus, for example, a PDS with a
small LOCA initiating event, with an injection failure, no safeguards available, and with a
small isolation failure, would be designated 7L. Any deviation from the above scheme
would be noted in the description of the event.
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TABLE 5 ASSIGNMENT OF NUMERICAL VALUES
TO VERBAL DESCRIPTORS VERBAL DESCRIPTORS LIKELIHOOD
Certain 1.0
Almost Certain 0.99 / 0.999
Likely 0.9
Indeterminate 0.5
Unlikely 0.1 / 0.01
Remotely Possible 0.001
Impossible 0.0
TABLE 5-2 TMI-1 CET BASIC EVENT DESCRIPTIONS
BASIC EVENT NAME
DESCRIPTION
AFTSTREN1 Likelihood That Cont. Can Handle Comb. Gas Burn Press. W/ High Base Pressure
AFTSTREN2 Likelihood That Cont. Can Handle Comb. Gas Burn Press. W/ Low Base Pressure
ALPHA Alpha Mode Failure of Containment Exists
ATSTREN1 Likelihood That Cont Can Handle H2 Burn Press. W/ High Base Press.
ATSTREN2 Likelihood That Cont Can Handle H2 Burn Press. W/ Low Base Press.
AUXSPRAYS Likelihood That Scrubbing Capability of Fission Products Exists
AUXWATER Fission Product Releases Are Under Water in the Aux. Bldg.
CBREL Likelihood That FPs Are Not Released to Containment Instead of the Environment
CHEMICAL Chemical Form of Fission Products Has High Vaporization Temp
CTMT-F-BENIGN CONTAINMENT LEAK BEFORE BREAK
CTMT-F-NOTBENIGN PROBABILITY THAT CONTAINMENT FAILURE IS NOT BENIGN
CWLIMITHPME Plant Configuration and Layout Limits Material Reaching Cont. Wall with HPME
CWLIMITLPME Plant Configuration and Layout Limits Material Reaching Cont. Wall with LPME
CWNOLIMITHPME Plant Config and Layout Does Not Limit Material Reaching Cont. Wall With HPM
CWNOLIMITLPME Plant Config and Layout Does Not Limit Material Reaching Cont. Wall With LPM
DCHFANSEFF Likelihood That Reactor Building Fans Can Handle DCH Pressure Spike
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TABLE 5-2 TMI-1 CET BASIC EVENT DESCRIPTIONS
BASIC EVENT NAME
DESCRIPTION
DCHFANSNOEFF Likelihood That Reactor Building Fans Cannot Handle DCH Pressure Spike
DCHFRAG Insufficient Fragmentation to Create Significant Pressure
DCHSTREN1 Likelihood That Cont. Strength Can Handle DCH Press Spike W/ High Base Press
DCHSTREN2 Likelihood That Cont. Strength Can Handle DCH Press Spike W/ Low Base Press
DRYEFF Likelihood That Recombination Can Deplete Comb. Gas Given a Dry Cavity
DRYEFFLT Likelihood That Recombination Depletes Comb. Gas With a Dry Cavity Late
EFFDEPRESS_0 OPERATOR OPENS PORV AFTER FAILING TO INITIATE HPI COOLING
EFFDEPRESS_0-C OPERATOR FAILS TO OPEN PORV AFTER FAILING TO INITIATE HPI COOLING
EFFDEPRESS_99 OPERATOR MANUALLY OPENS PORV
EFFDEPRESS_99-C OPERATOR FAILS TO OPEN PORV
EFFFORCEHT Likelihood That Forced Circulation Heat Transfer is Low
EFFNATHT Likelihood That Natural Circulation Heat Transfer is Low
EFFNPMP Conf That Primary Sys Failure Precedes S/G Tube Failure W/ RCPs Off
EFFPMP Conf That Primary Sys Failure Precedes S/G Tube Failure W/ RCPs On
EQUALFANSAF Likelihood Fans Survive Containment Environment Early After RV Failure
EQUALFANSLT Likelihood That RB Fans Survive Containment Environment to Prevent LCF
EQUALFANSPRI Likelihood RB Fans Do Survive Containment Environment At Or Prior To RV Failure
EXSCRUBEFF Likelihood That Overlying Water Pool Will Scrub FPs Released From Corium
FASTHTRATE Heat Transfer Rate From Corium To Water Pool is Fast
FREEZELOW Likelihood Corium Does Freeze On Lower Containment or Cavity Floor
GEOMFREEZE Cavity Geometry Allows Enough Corium to Disperse For Freezing
GEOMH2 Cavity Geometry Does Retain All Corium
H2SRCAFTER Concrete Attack Produces Insufficient Combustible Gas After RV Failure
HEATIML Prob. that Failure of the Primary System Occurs Due to Heating
HEATLOSS Heat Losses From Primary System Are Very Large
HPCMEFF Likelihood That Retention Is Low for a High Pressure Core Melt
IISL Likelihood of Induced IS-LOCA
INERTAF Containment Has Low Base Pressure Early After RV Failure Without Steam Inerting
INERTLT Sequence Late After RV Failure Has Low Base Pressure From Gas Generation
LOWCONCBURN Random Low Concentration Burns Prevent Significant Accumulation of Comb. Gas
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TABLE 5-2 TMI-1 CET BASIC EVENT DESCRIPTIONS
BASIC EVENT NAME
DESCRIPTION
LPCMEFF Likelihood That Retention Is Not Low for a Low Pressure Core Melt
MELT Likelihood That Water Pool in Cavity Will Not Stop Concrete Attack
MISSLELIKE Likelihood That Cont Failure Is Not Prevented Given a Pressure Generated Missile
NCGASES Likelihood That Containment Cannot Handle Pressure from Non-Condensable Gases
NCGASHIGH Likelihood That Non Condensable Gas Production is High Given a Dry Cavity
NCONBYOPS Likelihood That Operators Start the Reactor Coolant Pumps
NOAFTSTREN1 Likelihood That Cont Cannot Handle Comb. Gas Burn Press. W/ High Base Pressure
NOAFTSTREN2 Likelihood That Cont Cannot Handle Comb. Gas Burn Press. W/ Low Base Pressure
NOALPHA No Alpha Mode Failure of Containment
NOATSTREN1 Likelihood That Cont Cannot Handle H2 Burn Press. W/ High Base Pressure
NOATSTREN2 Likelihood That Cont Cannot Handle H2 Burn Press. W/ Low Base Pressure
NOAUXSPRAYS Likelihood That Scrubbing Capability of Fission Products Does Not Exist
NOAUXWATER Fission Product Releases Are Not Under Water in the Aux. Bldg.
NOCBREL Likelihood That FPs Are Released to Containment Instead of the Environment
NOCHEMICAL Chemical Form of Fission Products Does Not Have High Vaporization Temperature
NODCHFRAG Sufficient Fragmentation to Create Significant Pressure
NODCHSTREN1 Likelihood That Cont. Strength Cannot Handle DCH Press Spike W/ High Base Press
NODCHSTREN2 Likelihood That Cont. Strength Cannot Handle DCH Press Spike W/ Low Base Press
NODRYEFF Likelihood That Recombination Cannot Deplete Comb. Gas Given a Dry Cavity
NODRYEFFLT Likelihood That Recombination Cannot Deplete Comb. Gas With a Dry Cavity Late
NOEFFFORCEHT Likelihood That Forced Circulation Heat Transfer is High
NOEFFNPMP Conf That Primary Sys Failure Does Not Precede S/G Tube Failure W/ RCPs Off
NOEFFPMP Conf That Primary Sys Failure Does Not Precede S/G Tube Failure W/ RCPs On
NOEQUALFANSAF Likelihood Fans Do Not Survive Containment Environment Early After RV Failure
NOEQUALFANSLT Likelihood That RB Fans Do Not Survive Containment Environment to Prevent LCF
NOEQUALFANSPRI Likelihood RB Fans Do Not Survive Containment Environment At Or Prior To RV Fail
NOEXSCRUBEFF Likelihood That Overlying Water Pool Will Not Scrub FPs Released From Corium
NOFREEZELOW Likelihood Corium Does Not Freeze On Lower Containment or Cavity Floor
NOGEOMFREEZE Cavity Geometry Does Not Allow Enough Corium to Disperse For Freezing
NOGEOMH2 Cavity Geometry Does Not Retain All Corium
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TABLE 5-2 TMI-1 CET BASIC EVENT DESCRIPTIONS
BASIC EVENT NAME
DESCRIPTION
NOH2SRCAFTER Concrete Attack Produces Sufficient Combustible Gas After RV Failure
NOHEATIML Prob. that Failure of the Primary System Does Not Occur Due to Heating
NOHEATLOSS Heat Losses From Primary System Are Not Large
NOHPCMEFF Likelihood That Retention is Not Low for a High Pressure Core Melt
NOIISL CONFIDENCE THAT AN INDUCED ISLOCA IS PREVENTED
NOINERTAF Containment Has High Base Pressure Early After RV Failure Without Steam Inerting
NOINERTLT Sequence Late After RV Failure Has High Base Pressure From Gas Generation
NOLOWCONCBURN No Random Low Concentration Burns Prevent Significant Accumulation of Comb. Gas
NOLPCMEFF Likelihood That Retention is Low for a Low Pressure Core Melt
NOMELT Likelihood That Water Pool in Cavity Will Stop Concrete Attack
NOMISSLELIKE Likelihood That Cont Failure is Prevented Given a Pressure Generated Missile
NONCGASES Likelihood That Containment Can Handle Pressure From Non-Condensable Gases
NONCGASHIGH Likelihood That Non Condensable Gas Production is Not High Given a Dry Cavity
NONCONBYOPS Likelihood That Operators Do Not Start the Reactor Coolant Pumps
NOOPSDEPRESS Likelihood That Operators Do Not Depressurize Steam Generators
NOOTHERSCRUB Likelihood That There is No FP Scrubbing By Other Systems Not in Aux. Bldg.
NOOTHERWATER Water Does Not Fill Cavity From Plant Specific Sources And Paths
NOOXIDIZED In-Vessel H2 Prod. Sufficient to Cause H2 Burns
NOPDSLOW_5 HIGH PRESSURE AT CORE MELT IS INDETERMINATE
NOPLATEOUT Likelihood That Plateout Will Not Scrub Fission Products
NOPRISTREN1 Likelihood That Cont. Cannot Handle H2 Burn Press W/ High Base Pressure
NOPRISTREN2 Likelihood That Cont. Cannot Handle H2 Burn Press W/ Low Base Pressure
NOPRVHPCONF PROBABILITY THAT PORV DOES NOT PREVENT HPME
NOPZRSAFETY Prob. that Pressurizer Safety Valves Do Not Stick Open During Core Damage
NORECACPRI Power Is Not Recovered to the RCPs Prior to RV Failure
NORECOFFSITEPWR OFFSITE POWER NOT RECOVERED WITHIN 24 HOURS
NORECOVFANSAFT Reactor Building Fans Are Not Recovered Early After RV Failure
NORECOVFANSPRI Reactor Building Fans Are Not Recovered At or Prior to RV Failure
NORECOVRV Recovery of Core Cooling Does Not Prevent Reactor Vessel Failure
NORECOVSPAFT Containment Sprays Are Not Recovered Early After RV Failure
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TABLE 5-2 TMI-1 CET BASIC EVENT DESCRIPTIONS
BASIC EVENT NAME
DESCRIPTION
NORECOVSPPRI Containment Sprays Are Not Recovered Prior to RV Failure
NORECOVSSHR Prob. that Secondary Side Heat Removal is Not Recovered Prior to RV Failure
NORSGFANSEFF Likelihood That Reactor Building Fans Cannot Handle Rapid Steam Production
NORSGSTREN1 Likelihood That Cont Strength Cannot Handle RSG Press. Spike W/ High Base Press
NORSGSTREN2 Likelihood That Cont Strength Cannot Handle RSG Press. Spike W/ Low Base Press.
NORVROCKET No RV Rocket
NOSPARKAFT_01 PROB THAT SPARK IS UNAVAILABLE EARLY AFTER RV FAILURE WITH RB SPRAY
NOSPARKAFT_9 PROB THAT SPARK IS UNAVAILABLE EARLY AFTER RV FAILURE WITHOUT RB SPRAY
NOSPARKAT Random Spark is Unavailable at RV Failure
NOSPARKLT-NOP RANDOM SPARK UNAVAILABLE WITHOUT OFFSITE POWER
NOSPARKLT-OP RANDOM SPARK UNAVAILABLE WITH OFFSITE POWER
NOSPARK_01 PROB THAT SPARK IS NOT AVAILABLE BEFORE RV FAILURE WITH RB SPRAY
NOSPARK_9 PROB THAT SPARK IS NOT AVAILABLE BEFORE RV FAILURE WITHOUT RB SPRAY
NOSPRAYEFFLT Likelihood That Spray Will Not Scrub FPs Prior to Release to Environment
NOSTREN1H2 Likelihood That Cont Cannot Handle Comb. Gas Burn Press. W/ High Base Pressure
NOSTREN2H2 Likelihood That Cont Cannot Handle Comb. Gas Burn Press. W/ Low Base Pressure
NOWATEREFF Likelihood That Water in S/G Will Not Scrub Fission Products
OPSDEPRESS Likelihood That Operators Depressurize Steam Generators
OTHERSCRUB Likelihood That There is FP Scrubbing By Other Systems Not in Aux Bldg.
OTHERWATER Water Does Fill Cavity From Plant Specific Sources And Paths
OXIDIZED In-Vessel H2 Prod. Not Sufficient to Cause H2 Burns
PDSLOW_5 LOW PRESSURE AT CORE MELT IS INDETERMINATE
PLATEOUT Likelihood That Plateout Will Scrub Fission Products
POHPO1_FF--HVBOA OPERATOR FAILS TO OPEN PORV
PRISTREN1 Likelihood That Cont. Can Handle H2 Burn Press W/ High Base Press.
PRISTREN2 Likelihood That Cont. Can Handle H2 Burn Press W/ Low Base Press.
PRVHPCONF PROBABILITY THAT PORV CAN PREVENT HPME
PZPORVCONF_0 Prob. That Operators Open PORV After Failing to Init HPI Cooling
PZPORVCONF_0-C PROB THAT OPERATORS FAIL TO OPEN PORV AFTER FAILING TO INIT HPI COOLING
PZPORVCONF_99 PROB THAT OPERATORS OPEN PORV
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TABLE 5-2 TMI-1 CET BASIC EVENT DESCRIPTIONS
BASIC EVENT NAME
DESCRIPTION
PZPORVCONF_99-C PROBABILITY THAT OPERATORS FAIL TO MANUALLY OPEN PORV
PZRNOPORVDEP Likelihood That Pressurizer PORV(s) Cannot Depress Primary System to S/G Press
PZRPORVDEP Likelihood That Pressurizer PORV(s) Can Depress Primary System to S/G Press
PZRSAFETY Prob. that Pressurizer Safety Valves Stick Open During Core Damage
RBSPRAY RB SPRAY SYSTEM IS AVAILABLE
RECACPRI Power Is Recovered to the RCPs Prior to RV Failure
RECFANSLT AVAILABILITY OF RB FANS WITHOUT POWER DEPENDENCY
RECOFFSITEPWR OFFSITE POWER RECOVERED WITHIN 24 HOURS
RECOVFANSAFT Reactor Building Fans Are Recovered Early After RV Failure
RECOVFANSPRI Reactor Building Fans Are Recovered At or Prior to RV Failure
RECOVRV Recovery of Core Cooling Does Prevent Reactor Vessel Failure
RECOVSPAFT Containment Sprays Are Recovered Early After RV Failure
RECOVSPPRI Containment Sprays Are Recovered Prior to RV Failure
RECOVSSHR Prob. that Secondary Side Heat Removal is Recovered Prior to RV Failure
RECSPRAYLT AVAILABILITY OF CONTAINMENT SPRAYS WITHOUT POWER DEPENDENCY
RSGFANSEFF Likelihood That Reactor Building Fans Can Handle Rapid Steam Production
RSGSTREN1 Likelihood Strength Can Handle RSG Event and Base Pressure is High
RSGSTREN2 Likelihood Strength Can Handle RSG Event and Base Pressure is Low
RVROCKET RV Rocket
SLOWHTRATE Heat Transfer Rate From Corium to Water Pool is Slow
SPARKAFT_1 PROB THAT SPARK IS AVAILABLE EARLY AFTER RV FAILURE WITHOUT RB SPRAY
SPARKAFT_99 PROB THAT SPARK IS AVAILABLE EARLY AFTER RV FAILURE WITH RB SPRAY
SPARKAT Random Spark is Available at RV Failure
SPARKLT-NOP RANDOM SPARK AVAILABLE WITHOUT OFFSITE POWER
SPARKLT-OP RANDOM SPARK AVAILABLE WITH OFFSITE POWER
SPARK_1 PROB THAT SPARK IS AVAILABLE BEFORE RV FAILURE WITHOUT RB SPRAY
SPARK_99 PROB THAT SPARK IS AVAILABLE BEFORE RV FAILURE WITH RB SPRAY
SPRAYEFF Likelihood That Sprays Will Scrub FPs for Small Containment Failure
SPRAYEFFLT Likelihood That Sprays Will Scrub FPs Prior to Release to Environment
SPRAYNOEFF Likelihood That Sprays Will Not Scrub FPs for Small Containment Failure
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TABLE 5-2 TMI-1 CET BASIC EVENT DESCRIPTIONS
BASIC EVENT NAME
DESCRIPTION
SSHRNOREVAP Likelihood That SSHR Will Not Prevent Revaporization
SSHRNORVPREC Secondary Side Heat Removal is Not Recovered Prior to Revaporization
SSHRREVAP Likelihood That SSHR Will Prevent Revaporization
SSHRRVPREC Secondary Side Heat Removal is Recovered Prior to Revaporization
SSHRSGTCOOL Likelihood That SSHR Will Keep Tubes Cool
SSHRSGTNOCOOL Likelihood That SSHR Will Not Keep Tubes Cool
STREN1H2 Likelihood That Cont Can Handle Comb. Gas Burn Press. W/ High Base Pressure
STREN2H2 Likelihood That Cont Can Handle Comb. Gas Burn Press. W/ Low Base Pressure
UNEFFNATHT Likelihood That Natural Circulation Heat Transfer is High
WALLNOSURVIV Containment Wall Does Not Survive Contact With Corium
WALLSURVIV Containment Wall Survives Contact With Corium
WATEREFF Likelihood That Water in S/G Will Scrub Fission Products
5.2 ANALYSIS PERFORMED 5.2.1 Containment Capacity
The ultimate capacity of containment to withstand increasing pressure loadings is
perhaps the most important risk-related feature of a nuclear power plant, because a
failure by overpressurization could result in the release of radionuclides to the
environment.
Appendix B compares TMI-1's containment structure with Oconee's containment
structure, and evaluates the ultimate capacity of the TMI-1 containment relative to
Oconee's. TMI-1's containment structure is similar in design to Oconee's containment
structure; it is this similarity that supports the comparative evaluation quantitatively
and/or qualitatively. The comparison indicated that the pressure capacity of the TMI-1
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containment structure is between 136.8 psig and 146.9 psig. The 136.8 psig is based
upon using the minimum material properties for the post stressing tendons. The 146.9
psig assumes the TMI-1 material strength is at least as much as Oconee's and only the
tendon steel area and geometry are used to predict the relative capacity of the TMI-1
hoop tendons. Details of the calculation and assumptions used can be found in
Appendix B. The containment capacity analysis for Oconee is found in Appendix C.
This analysis contains the containment failure probability distributions used in the
Oconee PRA. Based on the comparison in Appendix B, the Oconee curves are used
directly in this analysis to evaluate the TMI-1 containment capacity and probability of
failure.
5.2.2 Combustible Gas Burns
Predicting the resulting pressure rise from hydrogen or other combustible gas burns is
an important part of a Level 2 containment analysis. The resulting pressure rise from
combustible gas burns is a function of the base pressure, the concentration of hydrogen
and other combustibles, the percentage of steam inerting (concentration of oxygen
available), and the degree of containment mixing. All these items can vary from
sequence to sequence. To limit the extent of analysis, bounding assumptions were
made on the various parameters. For instance, many current PRAs assume adiabatic
burns of all the combustible gas available in the containment atmosphere. By using the
adiabatic burn curve, a multiplier can be determined using the ratio of the maximum
pressure after combustible gas burn and the initial pressure in containment. A bounding
pressure rise for a given hydrogen concentration can then be calculated by multiplying
the base pressure by the multiplier.
Hydrogen gas concentrations in containment occurring early during accident
progressions depend heavily on the degree of oxidation that takes place in the vessel
during core heatup. Bounding hydrogen concentrations were obtained from high
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pressure cycling relief valve core melt sequences, followed by a hot leg or surge line
failure prior to reactor vessel failure, which will expel hydrogen into containment.
Analysis of this type of scenario was used to bound conditions on other scenarios.
Hydrogen and other combustible gases are produced from concrete during corium-
concrete interaction given a dry reactor cavity.
5.2.3 Reactor Cavity Geometry
Information on the reactor cavity geometry is required for the quantification of many of
the CET basic events. The following sections describe the reactor cavity geometry and
the pathways by which water can enter the cavity. The reactor cavity geometry is
shown in Figure 5-1.
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Figure 5-1
TMI-1 Reactor Cavity Geometry
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5.2.3.1 Lower Reactor Cavity Penetrations Description
There are three approximately 12 inch diameter pipes that enter the cavity at floor level
(elevation 279+6) from the reactor compartment system. Two reactor compartment fan
units (AH-E-2A & AH-E-2B), located at elevation 308+0, blow air down into the cavity
beneath the reactor. The air rises up around the reactor vessel thermal shield, through
holes in the vessel skirt, and up through the reactor vessel cavity, exiting around the
sand-filled shield plugs and the raised seal plate, and around the reactor vessel
nozzles. The purpose is to keep the concrete cool around the reactor cavity walls,
nozzles, and seal ring (less than 150°F).
There are two 2 inch diameter pipes that drain into the lower cavity from above. They
drain from the flat annular surface around the reactor vessel skirt anchors on the
concrete reactor vessel foundation. The two drains are 180° apart. One enters the
cavity next to the door about 18 inches up the wall and the other enters through the roof
of the instrumentation trench just under the lip of the opening.
There is a reactor cavity flood line (6 inch diameter pipe) from the fuel transfer canal
entering the reactor compartment at elevation 298+0. The reactor cavity flood line has
a flange, located at the pool floor, which is removed during operation to leave the line
open. (During refueling, the flange would isolate the drain.)
The instrumentation trench is sealed off with 5,000 psia concrete fill. A temporary
opening in the roof of the cavity entrance tunnel (in the widened area just inside the
heavy steel door) that provided access to the vessel skirt/anchor area during
construction is also permanently closed with concrete fill.
There is a single 2 inch diameter floor drain in the cavity to the reactor building sump.
There is a manual valve (with a reach rod) in the line that is located near the sump that
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is normally closed to prevent any sand (from the sand plugs) reaching the reactor
building sump and creating a potential hazard to pumps (e.g., containment spray and
decay heat removal) operating in recirculation mode.
There is a 2 feet 6 inches x 2 feet entrance tunnel to the cavity. The tunnel leaves the
cavity at floor level (elevation 279+6), goes approximately 3 feet 6 inches and then
widens slightly, goes another 3 feet 2 inches and then narrows again where it rises 1
feet 8 inches to elevation 181+2, goes another 4 feet and ends at elevation 181+2 in the
Reactor Building outside the primary shield wall. There is a heavy steel door in the
entrance tunnel, about 6 feet 8 inches from the cavity wall (12 feet 8 inches from reactor
centerline), where the entrance tunnel widens slightly. The bottom of the door is raised
1 foot 8 inches up from the cavity floor where the tunnel rises to elevation 281+2 of the
reactor building basement. A rope is used to pull the door closed, and the door has a
concealed latching device designed so that the door will withstand a cavity pressure of
at least 160 psid. There is also a thin steel mesh door 4 feet away from the heavy steel
door (16 feet 8 inches from reactor centerline) where the tunnel exits to the reactor
building at elevation 281+2.
5.2.3.2 Reactor Building Spray Flow into Lower Cavity
Water can reach the lower cavity from above via the fuel transfer pool. Water can travel
down around the vessel and into the cavity via the same path that the air travels up from
the Reactor Compartment fans. That path includes:
• Seal plates that are jacked up during operation (elevation 321+0)
• Spacers around the shield plugs (located just below the seal ring) leave at least a 1 inch space circumferentially around the reactor vessel (at least 5 square feet flow area)
• The reactor vessel skirt (elevation 290+0) has twelve 9½ inch diameter holes (5.6 square feet flow area)
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• The thermal shield supports provide about a 5 inch clearance under the lower edge of the shield and a 4 inch space around the bottom plate between the plate and lower cavity wall (12.6 square feet flow area)
There is also a 6 inch diameter open reactor cavity flood line from the fuel transfer canal
to the reactor compartment, entering at elevation 298+0. Water from the flood line will
flow down the reactor compartment and into the lower cavity through the holes in the
reactor vessel skirt and around the thermal shield.
The 2 inch diameter drains from the vessel skirt/anchor area to the cavity are not
significant for filling the cavity since the flow area through the reactor vessel skirt is
much greater (5.6 square feet) than the two 2 inch diameter pipes.
The lower cavity and reactor compartment is designed to hold water up to the level of
the reactor nozzles (elevation 312+0). This volume amounts to about 49,800 gallons in
the reactor compartment and lower cavity, and an additional 18,600 gallons in the fuel
transfer canal below elevation 312+0. The lower cavity access door is watertight and
designed to withstand blowdown forces of at least 160 psid from the inside; the door is
assumed to remain intact after reactor vessel failure. If the door fails, the corium will be
in a coolable geometry as long as a sufficient flow rate of water is maintained through
the reactor compartment.
Some of the water from the reactor building sprays will find its way to the lower cavity
via the fuel transfer pool. Some of the spray flow that lands in the pool will flow directly
into the reactor compartment around the seal ring, and the remainder will spill into the
deep end of the pool (the gate is assumed to be open) and flow into the cavity via the
reactor cavity flood line. The flood line is assumed not to clog with corium because it
enters the reactor compartment at a high elevation (elevation 298+0, approximately
half-way up the vessel). Even if the flood line clogs, the water will enter the cavity via
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the seal ring. The deep end of the fuel transfer pool also has a drain (4 inch diameter)
to the reactor building sump with a normally closed valve (SF-V-31).
Assuming uniform dispersal of RB sprays over the whole RB area (approximately
12,000 square feet), about 14% of the spray flow will fall into the fuel transfer pool
(about 1,700 square feet), including the shallow and deep ends, and then into the
reactor compartment. For example, if the total spray flow is 3,000 gpm, then about 420
gpm will flow into the reactor compartment.
5.2.3.3 Corium Dispersal and Coolability
Corium dispersal is dependent on the plant configuration and obstacles in the reactor
cavity that could prevent or impede the dispersal of corium into a coolable geometry in
the lower levels of containment. The coolability of the corium is dependent on the
thickness of the corium pool and the availability of a continuous source of cooling water
(e.g., BWST). Literature suggests that if the corium bed is less than 10 inches thick,
then it is considered coolable. However, a corium bed greater than 20 inches thick is
not considered coolable.
The TMI-1 lower cavity configuration is relatively confining. It has a heavy, normally
closed, access door that can withstand at least 160 psid, as well as an instrument
trench sealed off with concrete fill (5,000 psia). In addition, pathways for corium
dispersal up around the reactor vessel and out through the biological shield wall
penetrations (i.e., hot leg and cold leg) exist. However, these pathways are torturous
and the corium is likely to remain in the reactor cavity unless the access door fails.
During HPME for which the pressure in the reactor cavity exceeds the design pressure
of 160 psid, the access door may fail allowing corium to disperse out into containment.
The likelihood that the access door fails due to the pressure developed in the reactor
cavity is considered in the quantification of various basic events discussed in Section
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5.3. However, pressure developed in the reactor cavity can be relieved via pathways
identified around the reactor vessel, thus preventing a failure of the access door in most
cases. If the corium can sufficiently disperse from the reactor cavity into lower
containment following a HPME, then the corium pool is expected to freeze in lower
containment. However, in most cases, spreading is only expected to occur after the
access door has been melted through. This dispersal mechanism is expected to result
in freezing of the corium without an overlying pool of water. The corium that remains in
the reactor cavity is expected to be greater than 20 inches thick due to the height of the
curb; therefore, the effectiveness of an overlying pool of water to cool the debris bed is
indeterminate. This uncertainty is reflected in the quantification of the basic event
NOMELT discussed in Section 5.3.
However, if the corium pool cannot disperse, then a continuous source of cooling water
(i.e., the containment sprays available in both the injection and the recirculation modes)
is required to quench the corium and terminate concrete attack. A static volume of
water will only temporarily suspend concrete attack, since the water boils away and
gives rise to a potential increase in pressure inside containment. Technical Report
15.2B [Reference 22] indicates that a continuous source of coolant to the corium bed
would terminate corium-concrete attack based on MAAP models. In the quantification
of the basic event NOMELT, the effectiveness of the overlying pool is coupled with the
corium thickness and coolability of the debris bed.
5.2.4 MAAP Model
When the MAAP code was originally used to investigate containment phenomenological
issues, the Oconee model was used. To limit the number of new MAAP runs for TMI-1,
Oconee input parameters were used when the system and containment features at
Oconee and TMI-1 were similar, or the issue was strictly phenomenological. In these
cases the original Oconee model was benchmarked against the TMI-1 model.
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For containment characteristics that were plant specific (e.g., concrete composition),
MAAP runs were made with a TMI-1 model to investigate the sensitivity of that
parameter (e.g., concrete composition had a significant impact on the generation of
non-condensable gases).
5.2.5 Containment Base Pressure
Base pressure in containment is impacted by the presence of inerting gases in
containment. These inerting gases primarily include steam and carbon dioxide. The
containment is considered inerted when the concentration of steam is greater than 50%.
Steam and carbon dioxide are treated the same in MAAP. Inerting impacts not only the
base pressure, but also the likelihood that combustible gases will burn in containment.
This is discussed in further detail in Section 5.3.
Steam inerting can occur prior to, early after, or late after reactor vessel failure. The
concentration, and therefore the base pressure in containment, can be reduced by the
operation of fan coolers to condense the steam in the containment atmosphere.
Inerting due to the production of non-condensable gases is primarily a concern late after
reactor vessel failure. Non-condensable gases are produced due to corium-concrete
interaction. Fan coolers have no impact on the concentration of non-condensable
gases in containment. Only the suspension or termination of the production of non-
condensables is effective. The provision of a static or continuous flow of water is
effective for suspending or terminating the production of non-condensable gases,
respectively. High and low base pressures are defined as above or below 40 psia,
respectively. Prior to concrete attack and resulting significant production of non-
condensable gases, a high base pressure in containment corresponds to greater than a
50% concentration of steam in containment. Therefore, steam inerting is assumed for
high base pressure scenarios prior to concrete attack.
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5.3 DECISION TREE BASIC EVENT QUANTIFICATION
AFTSTREN1 (NOAFTSTREN1)
Confidence that Containment Can (Cannot) Handle Combustible Gas Burn Pressure
with High Base Pressure Early After Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that containment will remain intact
given a combustible gas burn early after reactor vessel failure with high containment
base pressure. The quantification of this event reflects only the phenomenon of a
combustible gas burn (hydrogen and carbon monoxide).
Combustible gas concentrations in containment could fall within the range of 12% to
17%. Burns at these concentrations could result in a pressure spike to approximately 4
times the initial or base pressure, according to the EPRI large-scale hydrogen
experiments [Reference 6]. The resulting peak pressure could reach or exceed
containment capacity.
A base pressure in the range of 40 psia was used since the addition of more steam
would inhibit combustible gas burning. Based on MAAP runs, this base pressure
corresponds to steam concentrations of approximately 30-40%. In general,
containment is considered inerted when the steam concentrations are greater than
50%.
The ratio of peak pressure, after the combustible gas burn, to the initial or base
pressure is a function of the concentration of combustible gas in containment. The
value of 4 selected for this ratio is conservative and is based on pessimistic
assumptions of combustible gas generation.
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Quantification:
This basic event is quantified based on the probability of containment failure resulting
from a combustible gas burn with a high base pressure. Using a base pressure of 40
psia and a peak-to-base pressure ratio of 4, the peak containment pressure would be
approximately 160 psia (145.3 psig). This corresponds to a containment failure
probability of about 0.5 because it is near the median of the containment failure
probability curve used in the Oconee PRA [Reference 16]. This assumes that the TMI-1
containment ultimate strength (median value 146.9 psig) is similar to the Oconee
containment ultimate strength (median value 144 psig), based on the report "TMI-1
Containment Capacity" dated April 15, 1992 [Appendix B] and assuming TMI-1 tendon
material strength of at least as much as Oconee.
Therefore, both the success and failure events are designated as "indeterminate" for all
PDSs.
AFTSTREN1 = 0.5
NOAFTSTREN1 = 0.5
AFTSTREN2 (NOAFTSTREN2)
Confidence that Containment Can (Cannot) Handle Combustible Gas Burn Pressure
with Low Base Pressure Early After Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that containment will remain intact
given a combustible gas burn early after reactor vessel failure with low containment
base pressure. Note that the maximum pressure in containment with the containment
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safeguards operating establishes the base pressure. The quantification of this event
reflects only the phenomenon of a combustible gas burn.
For example, at Oconee a base pressure in the range of 20-25 psia could be obtained
with containment safeguards operating; similar pressures are expected at TMI-1. From
MAAP runs, this base pressure corresponds to steam concentrations of approximately
25%. A conservative ratio of approximately 4 can be used to calculate the pressure rise
resulting from a combustible gas burn.
Quantification:
The basic event is quantified based on the probability of containment failure resulting
from the combustible gas burn with a low base pressure. The containment failure
probability curve in the Oconee PRA [Reference 16] was used, assuming that the TMI-1
containment ultimate strength (median value 146.9 psig) is similar to the Oconee
containment ultimate strength (median value 144 psig), based on the report "TMI-1
Containment Capacity" dated April 15, 1992 [Appendix B].
For typical combustible gas concentrations with low base pressures, the probability of
containment failure is quite low. Therefore, the event for success is quantified as
"almost certain" for all PDSs, and the complementary event for failure is quantified as
"remotely possible".
AFTSTREN2 = 0.999
NOAFTSTREN2 = 0.001
ATSTREN1 (NOATSTREN1)
Confidence that Containment Can (Cannot) Handle Hydrogen Burn Pressure with High
Base Pressure at Reactor Vessel Failure
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Description:
This basic event represents the analyst's confidence that containment will remain intact
given a hydrogen burn immediately following reactor vessel failure with high
containment base pressure. The quantification of this event reflects only the
phenomenon of a hydrogen burn.
Quantification:
The quantification for this event is similar to that for AFTSTREN1. The basic events for
both success and failure are designated as "indeterminate" for all PDSs.
ATSTREN1 = 0.5
NOATSTREN1 = 0.5
ATSTREN2 (NOATSTREN2)
Confidence that Containment Can (Cannot) Handle Hydrogen Burn Pressure with Low
Base Pressure at Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that containment will remain intact
given a hydrogen burn immediately following reactor vessel failure with low containment
base pressure. The quantification of this event reflects only the phenomenon of a
hydrogen burn.
At reactor vessel failure a spike in containment pressure normally occurs. Thus the
pressure over which hydrogen burns would be applied is slightly higher than for burns
prior to or after reactor vessel failure.
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Quantification:
The quantification for this event is similar to that for AFTSTREN2. This basic event is
quantified as "almost certain" for all PDSs, and the complementary event for failure is
quantified as "remotely possible".
ATSTREN2 = 0.999
NOATSTREN2 = 0.001
AUXSPRAYS (NOAUXSPRAYS)
Confidence that Scrubbing Capability of Fission Products Exists (Does Not Exist) in the
Auxiliary Building
Description:
This basic event represents the analyst's confidence that the capability for scrubbing of
fission products exists in the Auxiliary Building. The availability of a system, such as a
fire protection system, in the Auxiliary Building and its effectiveness for scrubbing of
fission products is sequence dependent (e.g., the location of the ISLOCA).
Quantification:
At TMI-1, Auxiliary Building spray is available from the fire protection pumps, which
have dedicated diesels and a high-elevation tank in case of pump failure. The
sprinklers are heat activated. However, fire protection spray droplets are larger than
containment spray droplets; in addition, location of sprinkler heads relative to potential
ISLOCA locations is unknown.
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No credit will be taken for the scrubbing effect of Auxiliary Building sprays, therefore this
basic event is quantified as "impossible" for all PDSs, and its complementary event
quantified as "certain".
AUXSPRAYS = 0.0
NOAUXSPRAYS = 1.0
AUXWATER (NOAUXWATER)
Fission Product Releases Are (Are Not) Underwater in the Auxiliary Building
Description:
This basic event determines if the release of fission products to the Auxiliary Building
occurs underwater. This information is used to determine whether an overlying water
pool scrubs the fission products.
Interfacing-systems LOCA and isolation failures to the Auxiliary Building are the PDSs
that could have an underwater release. This event depends upon the location of the
break.
Quantification:
This basic event is quantified as "impossible" for all PDSs because water in the Auxiliary
Building will tend to flow to a lower elevation (where heat exchangers are located) rather
than collecting in likely ISLOCA break locations. The complementary term is quantified
as "certain".
AUXWATER = 0.0
NOAUXWATER = 1.0
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B (NOTB)
Containment Is (Is Not) Isolated
Description:
This basic event determines whether the containment is isolated or not based on fault
tree logic for the various containment isolation systems. This logic was developed as
part of the logic for the Level 1 to Level 2 "bridge" event tree [References 24 and 25].
The nodal logic for this event was described above in Section 3.0.
Quantified:
Based on the decision tree logic, this basic event is successful if the logic for top event
NOTB is satisfied, and considered failed if otherwise (top event B).
C (NOTC)
Isolation Failure Is Small (Large)
Description:
This basic event determines the isolation failure hole size. Given that containment is
not isolated, information is needed to determine whether the hole size is large or small
in order to properly assess the fission product release fractions.
Quantification:
The nodal logic is linked to the containment isolation system model, such that a system
fault tree directly determines the size of the hole. The success branch for this node
represents a small isolation failure and the failed branch a large isolation failure.
Section 3 above discusses the nodal logic for this event.
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CHEMICAL (NOCHEMICAL)
Chemical Form of Fission Products Has (Does not Have) High Vaporization
Temperature
Description:
This basic event is used to determine if the chemical form of the deposited fission
products will allow revaporization to occur. The isotopes of greatest concern are Iodine
and Cesium.
Quantification:
The chemical form of the fission products Iodine and Cesium is still being analyzed,
[Reference 19] along with the interactions of these fission products with large quantities
of boric acid and metal oxide surfaces within high temperature environments.
Experiments have been performed, [Reference 19], with the assumption that the
chemical forms will be CsI and CsOH, and have shown that at high temperatures
(1000°C) revaporization of these fission products can occur. These temperatures can
be obtained in certain accident sequences, mainly station blackout scenarios. Using the
assumption that CsI and CsOH are the dominant chemical forms, revaporization will not
be prevented for high temperature environments.
This basic event representing fission products having a high vaporization temperature is
quantified as "unlikely" for all core damage sequences. The complementary event is
quantified as almost "certain".
CHEMICAL = 0.01
NOCHEMICAL = 0.99
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CWLIMITHPME (CWNOLIMITHPME)
Plant Configuration and Layout Limits (Does Not Limit) Material Reaching the
Containment Wall due to High Pressure Melt Ejection
Description:
This basic event represents the probability that corium is prevented from reaching the
containment wall after a high pressure melt ejection (HPME). HPME involves an
ejection of molten fuel out of the reactor vessel into the reactor cavity. The probability of
the corium dispersing and reaching the containment wall is dependent upon the plant
configuration, obstacles in the reactor cavity that could prevent or impede the dispersal
of the corium, as well as any pathways out of the reactor cavity to the containment wall.
Quantification:
The TMI-1 cavity configuration has been reviewed and potential corium pathways
examined. As discussed in Section 5.2.3, the TMI-1 lower cavity configuration is
relatively confining and has a heavy, normally closed, access door that can withstand
an internal pressure of at least 160 psid. Pathways for corium dispersal up around the
reactor vessel and out through the biological shield wall penetrations (i.e., hot leg and
cold leg) were identified to exist. However, these pathways are torturous. Thus, this
basic event is quantified as "likely" for all PDSs, and its complementary event as
"unlikely".
CWLIMITHPME = 0.9
CWNOLIMITHPME = 0.1
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CWLIMITLPME (CWNOLIMITLPME)
Plant Configuration and Layout Limits Material Reaching the Containment Wall due to
Low Pressure Melt Ejection
Description:
This basic event represents the probability that corium is prevented from reaching the
containment wall after a low pressure melt ejection (LPME). LPME involves an ejection
of molten fuel out of the reactor vessel into the lower reactor cavity. The probability of
the corium dispersing and reaching the containment wall is dependent upon the plant
configuration, obstacles in the reactor cavity that could prevent or impede the dispersal
of the corium, as well as any pathways out of the lower reactor cavity to the containment
wall.
Quantification:
The TMI-1 cavity configuration has been reviewed and potential corium pathways
examined. As discussed in Section 5.2.3, the TMI-1 cavity configuration is relatively
confining and has a heavy, normally closed, access door that can withstand an internal
pressure of at least 160 psid. Therefore this basic event is quantified as "almost
certain" for all PDSs, and its complementary event as "remotely possible".
CWLIMITLPME = 0.99
CWNOLIMITLPME = 0.01
D (NOTD)
Release Is (Is Not) Through Auxiliary Building
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Description:
This basic event determines whether the fission product release will pass through the
Auxiliary Building. The Auxiliary Building is available to filter releases from an ISLOCA
and containment isolation failures. Most ISLOCAs are released through the Auxiliary
Building.
Quantification:
This basic event is quantified as "certain" for releases through the Auxiliary Building and
"impossible" if containment is isolated or the release is not through the Auxiliary
Building. The following combination of CMB and CSS/CIS conditions determine
success or failure of this node.
D = 1.0 For PDSs with CMB 1 through 14 that have CSS/CIS G through R, and all PDS 19 sequences
NOTD = 0.0 For all remaining PDSs
DCHFANSEFF (DCHFANSNOEFF)
Confidence that Reactor Building Fans Can (Cannot) Handle DCH Pressure Spike
Description:
This basic event represents the analyst's confidence that Reactor Building (RB) fans are
an effective heat sink for a DCH event.
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Quantification:
The heat transfer to the RB fans is considered to be too slow to be effective to minimize
the pressure spike. Therefore, this basic event is quantified as "unlikely" for all PDSs,
and its complementary event as being "almost certain".
DCHFANSEFF = 0.01
DCHFANSNOEFF = 0.99
DCHFRAG (NODCHFRAG)
Insufficient (Sufficient) Fragmentation to Create Significant Pressure
Description:
This basic event represents the analyst's confidence that the dispersed corium will not
fragment sufficiently to cause an efficient direct containment heating (DCH) event.
Review of DCH documents has shown that debris particle size can have a large effect
on the phenomenon of DCH. The DCH event, and subsequent pressure rise in
containment, is highly dependent on the ability of the corium particles to quickly transfer
their energy to the surrounding environment. IDCOR Technical Report 85.2 [Reference
19] states that any obstacles can entrain or redirect the dispersed corium and can
cause the particles to conglomerate together and, therefore, reduce the efficiency of the
heat transfer. Generally, the more torturous the path, the more likely that the corium will
not disperse.
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Quantification:
At TMI-1, there appears to be sufficient obstacles and torturous pathways such that
insufficient fragmentation is "likely" for all PDSs, and its complementary event as being
"unlikely".
DCHFRAG = 0.9
NODCHFRAG = 0.1
DCHSTREN1 (NODCHSTREN1)
Confidence that Containment Strength Can (Cannot) Handle DCH Pressure Spike with
High Base Pressure
Description:
This basic event represents the analyst's confidence that containment will remain intact
if a DCH event occurs at a time when the containment base pressure is high.
Analysis of containment response, using MAAP, to DCH events has shown that
pressure spikes can be quite high (60 to 80 psid). These pressures depend on the
fraction of core material participating. With a high base pressure (40 psia to 100 psia),
this could result in pressures of 100 to 180 psia.
However, the MAAP DCH model is limited since it transfers heat only to the lower
compartment. Therefore, it is possible that higher pressures could be reached if the
heat could be transferred to the entire containment volume. In addition, MAAP does not
oxidize the remaining zircaloy, which would be present in the ejected melt.
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Quantification:
The quantification of this basic event is based on the probability of containment failure
resulting from pressure spikes due to a DCH event at a high base pressure, and on the
analyst's judgment. The containment failure probability curve in the Oconee PRA
[Reference 16] was used, assuming that the TMI-1 containment ultimate strength
(median value 146.9 psig) is similar to the Oconee containment ultimate strength
(median value 144 psig), based on the report "TMI-1 Containment Capacity" dated April
15, 1992 [Appendix B]. This basic event is quantified as "unlikely" for all PDSs, and its
complementary event as "almost certain".
DCHSTREN1 = 0.01
NODCHSTREN1 = 0.99
DCHSTREN2 (NODCHSTREN2)
Confidence that Containment Strength Can (Cannot) Handle DCH Pressure Spike with
Low Base Pressure
Description:
This basic event represents the analyst's confidence that containment will remain intact
if a DCH event occurs at a time when the containment base pressure is low.
The general discussion from DCHSTREN1 applies for this basic event. For example, at
Oconee with a low base pressure of 20-25 psia, peak pressures on the order of 80 to
105 psia would be expected.
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Quantification:
At Oconee peak pressures of 80-105 psia, correspond to a containment failure
probability of less than .01. Due to the similarities between TMI and Oconee's
containment design, this basic event is quantified as "almost certain" for all PDSs, and
its complementary event as "unlikely".
DCHSTREN2 = 0.99
NODCHSTREN2 = 0.01
DRYEFF (NODRYEFF)
Confidence that Recombination Can (Cannot) Deplete Combustible Gas with a Dry
Cavity
Description:
This basic event represents the analyst's confidence that recombination will deplete the
available combustible gas given a dry cavity. This phenomenon will help to reduce the
amount of combustible gas available in containment and thus reduce the ability to
generate global burns.
Natural circulation induced flow through the reactor cavity, in conjunction with high
reactor cavity temperatures, can result in significant recombination of the combustible
gases, which are produced by core-concrete interaction, with available oxygen.
However, recombination may be limited by the rate at which oxygen is supplied to the
reactor cavity region and may be precluded by the high steam concentrations. Two
additional factors would tend to diminish or negate the effect of recombination: 1) the
combustible gases would compete with hot steel structures for the available oxygen,
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and 2) the steam produced by recombination could react with hot steel structures
producing combustible gas.
Quantification:
It is the analyst's judgment that for dry cavity cases, recombination is likely to occur.
However, it is uncertain as to the condition of the convective flow paths. That is,
obstacles may impede or block natural circulation of the corium in the reactor cavity and
reduce the effectiveness of the recombination of combustible gases in the reactor
cavity.
This basic event is quantified as "indeterminate" for all PDSs, with the complementary
event also being "indeterminate".
DRYEFF = 0.5
NODRYEFF = 0.5
DRYEFFLT (NODRYEFFLT)
Confidence that Recombination Can (Cannot) Deplete Combustible Gas with a Dry
Cavity Late After Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that recombination will deplete the
available combustible gas given a dry cavity late in the sequence. This phenomenon
will help to reduce the amount of combustible gas available in containment and thus
reduce the ability to generate global burns.
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Quantified:
This event is quantified in an identical manner to DRYEFF. This basic event is
quantified as "indeterminate" for all PDSs, with the complementary event also being
"indeterminate".
DRYEFFLT = 0.5
NO DRYEFFLT = 0.5
EFFDEPRESS (NOEFFDEPRESS)
Confidence that Operators Open (Do Not Open) the PORV Prior to Steam Generator
Tube Failure
Description:
This event represents the analyst's confidence that the operators will open the
pressurizer PORV to depressurize the primary system prior to creep rupture of the
steam generator tubes.
Creep rupture is a concern if the pressure difference across the tubes is high
accompanied by high tube temperatures. If the reactor coolant pumps (RCPs) are
started, it is expected that mixing of the hot core gases with gases in the upper plenum
and hot legs will rapidly bring tube temperatures to the point where creep rupture is a
concern. If the PORV is opened prior to RCP start, or time is available following RCP
start for the operators to open the PORV, tube rupture may be avoided. (The
effectiveness of the PORV in this situation is included elsewhere, see PZRPORVDEP.)
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Quantification:
Quantification depends on emergency procedures. Although the procedures do not
specifically discuss induced steam generator tube rupture as a reason for equalizing
tube delta-P, the emergency operating procedure (for superheated conditions) does
instruct operators to depressurize the primary system (open the PORV) when RCPs are
started. Since the above actions are proceduralized, this event can be quantified as
"almost certain" for most PDSs.
The exception, however, is where the PDS indicates the operators have already failed
to go on HPI cooling, in which case the analysis will not give credit for another
opportunity. For those PDSs that involve operator failure to open the PORV, such as to
initiate HPI cooling, it was determined that the conditional probability to later open the
PORV would be 0.0. Therefore, the resulting probability of EFFDEPRESS is based on
logic to account for those core damage sequences involving operator failure to initiate
HPI cooling. The success and failure logic for this event is depicted below in Figure 5-2.
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Figure 5-2 Logic for Event EFFDEPRESS and NOEFFDEPRESS
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EFFFORCEHT (NOEFFFORCEHT)
Confidence that Forced Circulation Heat Transfer is Low
Description:
This basic event represents the analyst's confidence that the heat transfer rate from the
reactor coolant system (RCS) gases to the steam generator tubes is low when the
RCPs are running under inadequate core cooling conditions.
Quantification:
Operation of the RCPs is expected to quickly raise tube temperatures to near the RCS
gas temperature. Therefore, this basic event is quantified as "unlikely" for all PDSs, and
its complementary event as "almost certain".
EFFFORCEHT = 0.01
NOEFFFORCEHT = 0.99
EFFNATHT (UNEFFNATHT)
Confidence that Natural Circulation Heat Transfer is Low (High)
Description:
This basic event represents the analyst's confidence that the heatup of the steam
generator tubes is low with the RCPs not running.
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Quantification:
Heating of the hot legs and the surge line are expected to be more significant during
natural circulation. This basic event is quantified as "likely" for all PDSs, and its
complementary event as "unlikely".
EFFNATHT = 0.9
UNEFFNATHT = 0.1
EFFNPMP (NOEFFNPMP)
Confidence that RCS Failure Precedes (Does Not Precede) Steam Generator Tube
Failures with RCPs Off
Description:
This basic event represents the analyst's confidence that either a hot leg or the surge
line would fail prior to tube failure with the RCPs off.
Quantification:
The proximity of the hot leg to the core exposes it to higher temperatures than exist at
the steam generator. With a cycling relief valve hot gases are pulled regularly through
the surge line. These conditions increase the likelihood that a hot leg or the surge line
would fail first. This view is consistent with that expressed in NUREG/CR-4551
[Reference 1].
This basic event is quantified as "likely" for all PDSs, and its complementary event as
"unlikely".
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EFFNPMP = 0.9
NOEFFNPMP = 0.1
EFFPMP (NOEFFPMP)
Confidence that RCS Failure Precedes (Does Not Precede) Steam Generator Tube
Failure with RCPs Running
Description:
This basic event represents the analyst's confidence that a primary system failure would
occur prior to a steam generator tube failure with the RCPs running.
Quantification:
A well-mixed system would expose all components to nearly the same gas temperature. The low mass of a steam generator tube will allow it to heat up more quickly than other more massive components. The tubes become more likely to fail first. Therefore, this basic event is quantified as "unlikely" for all PDSs, and its complementary event as "likely".
EFFPMP = 0.1
NOEFFPMP = 0.9
EQUALFANSAF (NOEQUALFANSAF)
Confidence that RB Fans Survive (Do Not Survive) Containment Environment Early
After Reactor Vessel Failure
Description:
This basic event represents the confidence that RB fans can survive the severe
containment environment after reactor vessel failure. The high radiation and or the high
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temperature steam environment within containment may impact performance of RB
fans.
Quantification:
No obvious differences were observed between the equipment qualification of the RB
fans for TMI-1 and Oconee. The Oconee RB fans are expected to remain functional
through the full range of post core melt conditions. Therefore, this basic event is
quantified as "certain" for all PDSs, and its complementary event as "impossible".
EQUALFANSAF = 1.0
NOEQUALFANSAF = 0.0
EQUALFANSLT (NOEQUALFANSLT)
Confidence that RB Fans Survive (Do Not Survive) Containment Environment Late After
Reactor Vessel Failure
Description:
This basic event represents the confidence that RB fans can survive the severe
containment environment late after reactor vessel failure. The high radiation and or the
high temperature steam environment within containment may impact performance of RB
fans.
Quantification:
No obvious differences were observed between the equipment qualification of the RB
fans for TMI-1 and Oconee. The Oconee RB fans are expected to remain functional
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through the full range of post core melt conditions. Therefore, this basic event is
quantified as "certain" for all PDSs, and its complementary event as "impossible".
EQUALFANSLT = 1.0
NOEQUALFANSLT = 0.0
EQUALFANSPRI (NOEQUALFANSPRI)
Confidence that RB Fans Survive (Do Not Survive) Containment Environment at or Prior
to Reactor Vessel Failure
Description:
This basic event represents the confidence that RB fans can survive the severe
containment environment at or prior to reactor vessel failure. The high radiation and or
the high temperature steam environment within containment may impact performance of
RB fans.
Quantification:
No obvious differences were observed between the equipment qualification of the RB
fans for TMI-1 and Oconee. The Oconee RB Fans are expected to remain functional
through the full range of post core melt conditions. Therefore, this basic event is
quantified as "certain" for all PDSs, and its complementary event as "impossible".
EQUALFANSPRI = 1.0
NOEQUALFANSPRI = 0.0
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EXSCRUBEFF (NOEXSCRUBEFF)
Confidence that Overlying Water Pool Will (Will Not) Scrub Fission Products Released
from Corium
Description:
This basic event represents the analyst's confidence that a water pool covering the
debris bed will effectively scrub the ex-vessel release of fission products. This event is
important for reducing the amount of certain fission products prior to release to the
environment (e.g., tellurium).
Quantification:
Review of the available literature, NUREG/CR-1989 [Reference 9], describing
experiments in this area has shown that water pools are very effective in scrubbing
fission products. As the core interacts with the concrete, fission products and other
gases are released. The fission products tend to deposit on these gas bubbles and
begin to migrate to the top of the water pool. Collapse of these gas bubbles within the
water pool will prevent the continued migration of the fission products.
However, experiments discussed in NUREG/CR-3024 [Reference 8] have also shown
that in many cases the core-concrete interaction is so violent that the gas bubbles are
not collapsed within the water pool. This would result in a reduced effectiveness of the
water.
In general, water pools are still considered effective during release of these fission
products. This basic event is quantified as "likely" for all PDSs, and its complementary
event as "unlikely":
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EXSCRUBEFF = 0.9
NOEXSCRUBEFF = 0.1
FREEZELOW (NOFREEZELOW)
Confidence that Corium Will (Will Not) Freeze on Lower Containment and Cavity Floor
Description:
This basic event represents the analyst's confidence that if corium reaches the lower
containment, it will freeze prior to starting core-concrete interaction. This event is
important for determining if ex-vessel hydrogen and other combustible gases are
produced by the core-concrete interaction. It is also important in determining the
probability of failing containment through the basemat.
Quantification:
Analysis of containment using MAAP has shown that if corium reaches the lower
containment, it will spread out over a large area. In this case, freezing of the corium is
likely and no core-concrete interaction is expected to take place. Thus, this basic event
is quantified as "almost certain" for all PDSs, and its complementary event as "unlikely".
FREEZELOW = 0.99
NOFREEZELOW = 0.01
G (NOTG)
Benign (Non-Benign) Containment Failure
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Description:
This basic event represents the analyst's confidence that if containment fails due to a
late overpressurization, the failure will be benign. A benign failure is defined as a failure
of the containment structure that does not lead to a rapid blowdown of the containment
atmosphere, i.e., leak-before-break. Instead, the containment structure relieves
pressure enough so that a continued pressure rise does not occur.
The importance of benign containment failures is that fission product releases tend to
be very long in duration and very low in energy. Both of these factors help to reduce the
release fractions to the environment.
Quantification:
Based on industry experiments and NUREG-1150 [Reference 21], steel containments
have shown that, when pressurized to the failure point, catastrophic failure is likely.
However, for concrete containments, such as that at TMI-1, localized yielding/cracking
of the concrete occurs before catastrophic failure, and containment failure would likely
be benign. Therefore, this basic event is quantified as "likely" for all PDSs, and its
complementary event as "unlikely".
Top event G is an equivalency gate with basic event CTMT-F-BENIGN as its input and
NOTG (Failure of node G) is an equivalency gate with CTMT-F-NOTBENIGN as its
input.
CTMT-F-BENIGN = 0.9
CTMT-F-NOTBENIGN = 0.1
GEOMFREEZE (NOGEOMFREEZE)
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Cavity Geometry Allows Enough Corium to Disperse For Freezing
Description:
Freezing of the corium in the cavity and the lower containment depends on the ability to
spread the corium out over a large area. This basic event represents the confidence
that the cavity geometry allows sufficient dispersal of corium to allow freezing to take
place. This event affects direct containment heating, ex-vessel combustible gas
production, as well as steam production in containment.
Quantification:
The TMI-1 containment drawings were reviewed to identify possible dispersal pathways
out of the reactor cavity. It was discovered that TMI-1 and Oconee's reactor cavity
geometry are similar. That is, most lower levels of the reactor cavity are isolated from
lower containment by either a steel plate or ten feet of concrete grout. Dispersal of
corium from high pressure melt ejection scenarios via these pathways would be very
unlikely.
However, pathways for corium dispersal up around the reactor vessel and out through
the biological shield wall penetrations (i.e., hot leg and cold leg) were identified to exist.
In addition, a pathway would be available if the reactor shield plugs were ejected.
However, these pathways are viewed as torturous.
Therefore, based on the reactor cavity geometry and pathways for corium dispersal, this
event is quantified as "indeterminate" for all PDSs, including its complementary event as
well.
GEOMFREEZE = 0.5
NOGEOMFREEZE = 0.5.
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GEOMH2 (NOGEOMH2)
Cavity Geometry Retains (Does Not Retain) All Corium
Description:
This basic event represents the analyst's confidence that the cavity retains so much of
the ejected material during HPME that insufficient core debris reaches the lower
containment where combustible gases might be ignited. Since hydrogen ignites easily,
very little material would be required.
Quantification:
Due to the geometry of the reactor cavity and the pathways available for the ejection of
corium out of the reactor cavity into lower containment, this basic event is quantified as
"remotely possible" for all PDSs, and its complementary event as "almost certain".
GEOMH2 = 0.001
NOGEOMH2 = 0.999
H2SRCAFTER (NOH2SRCAFTER)
Insufficient (Sufficient) Combustible Gas is Produced After Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that concrete attack by the core
debris produces too little combustible gas to result in a global burn.
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Quantification:
Water released from the concrete would be expected to react with any unoxidized
zirconium in the debris or with rebar and release hydrogen gas. The availability of
zirconium and steel is expected to result in significant hydrogen production. Therefore,
this basic event is quantified as "unlikely" for all PDSs, and its complementary event as
"likely".
H2SRCAFTER = 0.1
NOH2SRCAFTER = 0.9
HEATIML (NOHEATIML)
Failure of the Primary System Occurs (Does Not Occur) due to Heating
Description:
This basic event represents the analyst's confidence that failure of the hot leg or surge
line depressurizes the RCS prior to vessel failure. Such failures might occur due to the
high temperature that these components are expected to see during core uncovery.
Quantification:
When the RCS is at the relief valve setpoint, natural circulation flows transport hot
gases from the core into the hot legs. The high temperatures that result, including high
system pressure, may rupture the affected piping. The decision tree logic appropriately
addresses both the dependency of the hot leg or surge line failures on high temperature
conditions, as well as high pressure conditions prior to vessel failure. Analyst judgment
is used to assess the probability for failure of the RCS, as well as the probability for
failure preceding failure of the reactor vessel.
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This basic event is quantified as "likely" for all PDSs, and its complementary event as
"unlikely".
HEATIML = 0.9
NOHEATIML = 0.1
HEATLOSS (NOHEATLOSS)
Heat Losses from Primary System Are (Are Not) Very Large
Description:
This basic event determines if the heat losses from the RCS are large enough to
prevent significant revaporization of the fission products in the RCS.
During a core melt accident, fission products that leave the core boundaries could
plateout on the "colder" sections of the RCS. If the fission products collect in large
quantities, significant localized heat buildup can occur. If this heat can be radiated to
the containment atmosphere, revaporization of these deposited fission products may be
prevented or reduced.
Quantification:
Quantification of this event is dependent on the capability of fission products to plateout
on colder sections of the RCS. At TMI-1 most areas of the RCS are covered by
insulation. However, the capability of the insulation to remain intact and perform its
intended function in a severe environment has not been fully investigated. There is also
a possibility of plateout on the RCS where there is no insulation. This basic event is
quantified as "indeterminate" for all PDSs, and its complementary event also as
"indeterminate".
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HEATLOSS = 0.5
NOHEATLOSS = 0.5
HPCMEFF (NOHPCMEFF)
Confidence that Retention Is (Is Not) Low For a High Pressure Core Melt
Description:
This basic event represents the analyst's confidence that if the accident is a high
pressure core melt, retention of fission products in the RCS will be small. This event is
important in determining if a revaporization event will occur late in the accident
sequence; studies have shown revaporization events late in the sequence occur when
RCS retention is high.
Fission products released during core melt can plateout in various sections of the RCS.
This ability to plateout is highly dependent on the condition of the RCS. For example, if
the accident sequence is a large
LOCA, any fission products released during core melt will probably be transported
immediately to containment. If, however, the accident sequence has a "bottled" RCS
(e.g., a cycling relief valve scenario), then the released fission products will have more
time to migrate around the RCS and plateout.
Quantification:
Sensitivity studies run with the MAAP code have shown that for high pressure core melt
sequences, such as small LOCAs and cycling relief valves, the retention of fission
products is high. In general, high pressure core melt sequences retain fission products
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to a high degree. Therefore, this basic event is quantified as "remotely possible" for all
PDSs, and its complementary event as "almost certain".
HPCMEFF = 0.001
NOHPCMEFF = 0.999
LOWCONCBURN (NOLOWCONCBURN)
Random (No Random) Low Concentration Burns Prevent Significant Accumulation of
Combustible Gas
Description:
This basic event represents the probability that there is a "good" burn or a low
concentration burn of combustible gas in containment. Random low concentration
burns deplete the combustible gas in the containment atmosphere thereby preventing a
significant accumulation of combustible gas in containment.
Quantification:
In general, little or no credit is given for low concentration burning of combustible gas in
containments where combustible gas igniters are not available. This basic event is
quantified as "remotely possible" for all PDSs, and its complementary event as "almost
certain".
LOWCONCBURN = 0.001
NOLOWCONCBURN = 0.999
LPCMEFF (NOLPCMEFF)
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Confidence that Retention Is Not Low (Is Low) for a Low Pressure Core Melt
Description:
This basic event represents the analyst's confidence that, if an accident sequence is a
low pressure core melt, retention of fission products in the RCS will be small. This
event is important in determining if a revaporization event will occur late in the accident
sequence.
Fission products released during core melt can plateout in various sections of the RCS.
This ability to plateout is highly dependent on the condition of the RCS. For example, if
the accident sequence is a large LOCA, any fission products released during core melt
will probably be transported immediately to the containment.
Quantification:
Sensitivity studies run with the MAAP code have shown that for low pressure core melt
sequences, such as large and medium LOCAs, the retention of fission products will be
low. Therefore, this basic event is quantified as "unlikely" for all PDSs, and its
complementary event as "likely".
LPCMEFF = 0.1
NOLPCMEFF = 0.9
MISSLELIKE (NOMISSLELIKE)
Confidence that Containment Failure Is Not (Is) Prevented Given a Pressure Generated
Missile
Description:
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This basic event represents the analyst's confidence that pressure generated missiles
within containment will not fail containment.
Quantification:
At TMI-1, the reactor vessel cavity is capped by reactor shield plugs that consist of steel
canisters containing sand. If there were pressurization of the lower reactor cavity, these
shield plugs could become projectiles and possibly result in a containment failure.
However, based on analysis of shield plugs at Oconee, the potential plug trajectory is
insufficient to reach the containment walls. Therefore, this basic event is quantified as
"remotely possible" for all PDSs, and its complementary event as "almost certain".
MISSLELIKE = 0.001
NOMISSLELIKE = 0.999
NCGASHIGH (NONCGASHIGH)
Confidence that Non-Condensable Gas Production Is (Is Not) High Given a Dry Cavity
Description:
Core-concrete interaction can potentially produce significant amounts of non-
condensable gases depending on the chemical content of the concrete. These non-
condensable gases can then be a potential contributor to late containment failures. This
basic event represents the confidence that the chemical form of the concrete produces
high amounts of non-condensable gases.
Quantification:
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TMI-1's concrete contains calcium, thus there is a potential for a significant amount of
non-condensable gas production. Therefore, this basic event is quantified as "likely" for
all PDSs, and its complementary event as "unlikely".
NCGASHIGH = 0.9
NONCGASHIGH = 0.1
NCGASES (NONCGASES)
Confidence that Containment Cannot (Can) Handle Pressure from Non-Condensable
Gases
Description:
This basic event represents the confidence that containment can handle the pressures
associated with the long term buildup of non-condensable gases. This event is paired
with high non-condensable gas production (event NCGASHIGH).
Quantification:
A sensitivity analysis was performed, using MAAP, to determine the effect of varying the
concrete composition on survivability of containment. A late containment overpressure
was projected due to long term non-condensable gas buildup in containment, given a
dry cavity and no corium dispersal after reactor vessel failure. If unabated, the steadily
increasing pressure will eventually fail containment. Therefore, this basic event is
quantified as "almost certain" for all PDSs, and its complementary event as "unlikely".
NCGASES = 0.99
NONCGASES = 0.01
NCONBYOPS (NONCONBYOPS)
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Confidence that Operators Start (Do Not Start) the Reactor Coolant Pumps
Description:
This basic event represents the analyst's confidence that the operators will start the
reactor coolant pumps. The operation of the pumps is an important contributor to the
potential for an induced steam generator tube rupture.
Quantification:
The quantification of this basic event is based on the analyst's discussions with the
human reliability analyst and station personnel and a review of emergency operating
procedures. The emergency operating procedure (for superheated conditions) instructs
operators to start the reactor coolant pumps. (There is no consideration in the
procedure regarding status of secondary side heat removal as a criterion for pump
start.) Hence, the operators are very likely to start the reactor coolant pumps. This
basic event is quantified as "almost certain" for all PDSs, and its complementary event
as "unlikely".
NCONBYOPS = 0.99
NONCONBYOPS = 0.01
NOALPHA (ALPHA)
Alpha Mode Failure of Containment Does Not (Does) Exist
Description:
This basic event represents the analyst's confidence that there will be no alpha mode
failure of containment. An alpha mode failure is defined as a steam explosion within the
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reactor vessel that results in the reactor vessel head itself becoming a projectile
resulting in a catastrophic failure of the containment vessel.
Quantification:
Quantification of this event is based on the probability of the reactor vessel not
becoming a projectile due to a steam explosion within the reactor vessel head. For a
steam explosion to occur, small droplets of water must fall onto the molten core in the
reactor vessel to maximize the potential for heat transfer between the corium and the
water in order to precipitate a steam explosion within the reactor vessel. Studies have
shown that the likelihood that these conditions would exist within the reactor vessel and
result in the reactor vessel head becoming a projectile is remote [Reference 19].
This basic event is quantified as "almost certain" for all PDSs, and its complementary
event as "remotely possible".
NOALPHA = 0.999
ALPHA = 0.001
NOCBREL (CBREL)
Confidence that Fission Products Are (Are Not) Released to Containment Instead of the
Environment
Description:
This basic event represents the analyst's confidence that for an induced steam
generator tube rupture, most of the fission products get released to the containment
rather than directly to the environment.
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Quantification:
Since tube ruptures are most likely to occur when RCS pressure exceeds the secondary
safety valve setpoint, the releases that bypass containment are expected to be
significant. However, if vessel failure occurs soon after the tube rupture and the safety
valve reseats the releases would be minimized. Quantification of this event is based on
judgment whether the reactor vessel will fail just after steam generator tube rupture
(SGTR) and minimize the releases to the environment. A significant release occurs at
the time of the tube failure.
This basic event is quantified as "unlikely" for all PDSs, and its complementary event as
"almost certain".
NOCBREL = 0.01
CBREL = 0.99
NOIISL (IISL)
Confidence that an Induced ISLOCA Is (Is Not) Prevented
Description:
No mechanism for inducing an ISLOCA is being considered.
Quantification:
This basic event is quantified as "certain" for all PDSs, and its complementary event as
"impossible".
NOIISL = 1.0
IISL = 0.0
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NOINERTAF (INERTAF)
Containment has High (Low) Base Pressure Early After Reactor Vessel Failure without
Inerting
Description:
This basic event represents the confidence that a high base pressure exists within
containment early after reactor vessel failure, without inerting the atmosphere with
respect to combustible gas burns.
Quantification:
Generally, steam inerting would be expected, because it is very unlikely that a high
containment pressure could be achieved in the time classified as "early after reactor
vessel failure" (about five hours) without steam generation.
However, TMI-1's concrete contains calcium; therefore, there is a potential for a
significant amount of non-condensable gases (i.e., carbon monoxide and carbon
dioxide) to be produced which could contribute to containment base pressure. Some
literature has shown that if there is a concentration of greater than 50% carbon dioxide
or steam in containment, containment is considered inerted [Reference 9 and 10].
MAAP adds the concentration of carbon dioxide to that of steam to calculate a total
concentration of inerting gases in containment.
For example, under some conditions with high production of non-condensable gases, it
may be possible to achieve high pressures within containment without completely
inerting the atmosphere. However, due to the limited time in the "early after" phase, this
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basic event is quantified as "unlikely" for all PDSs, and its complementary event as
"likely".
NOINERTAF = 0.1
INERTAF = 0.9
NOINERTLT (INERTLT)
Sequence Late After Reactor Vessel Failure has High (Low) Base Pressure from Gas
Generation without Inerting
Description:
This basic event represents the confidence that a high base pressure exists within
containment late in the accident sequence, without inerting the atmosphere with respect
to combustible gas burns. Generally, high base pressures are associated with steam
production with no safeguards systems available. However, under some conditions, for
example with high production of non-condensable gases, it may be possible to achieve
relatively high pressures within containment without completely inerting the atmosphere.
Quantification:
The quantification of this event is similar to that of NOINERTAF; however, it is more
likely that a high containment pressure would be achieved because more time is
available to generate the gas pressure. Therefore, this basic event is quantified as
"likely" for all PDSs, and its complementary event as "unlikely".
NOINERTLT = 0.9
INERTLT = 0.1
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NOMELT (MELT)
Confidence that Water Pool in Cavity Will (Will Not) Stop Concrete Attack
Description:
This basic event represents the analyst's confidence that concrete attack can be
prevented or terminated by the presence of water in the cavity.
Quantification:
The quantification of this event reflects the analyst's judgment on debris bed coolability
based on a literature review, plant structural review, and analyses using MAAP.
Quantification of this basic event is also based on the reactor cavity geometry and the
ability of the corium to spread. In addition, the porosity of the concrete can have an
impact on debris coolability. For example, at Oconee, the containment basemat is
constructed of a basalt concrete. Some experimental evidence indicates that the crusts
formed from basalt concrete are more porous than the crusts of other concrete types
[Reference 8]. This situation is a positive factor in assessing debris bed coolability.
TMI-1 containment is constructed of limestone concrete. This type of concrete is less
porous than basaltic concrete; therefore, less water is expected to penetrate the crust to
stop the concrete attack.
A sensitivity study using MAAP has shown that the core debris is coolable. The debris
is coolable even when pessimistic values for various heat transfer parameters are
selected. The IDCOR methodology for evaluating core-concrete interaction is
embodied in the DECOMP subroutine of MAAP. Benchmarking of the DECOMP model
shows that the IDCOR approach is consistent with experimental observations, both for
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experiments with and without water. These efforts have provided confidence in the
DECOMP treatment of concrete ablation and debris coolability.
It is recognized that uncertainties still exist in the modeling of debris cooling, and that
some experiments have shown continued concrete ablation in the presence of an
overlying water pool. These factors prevent assigning a very large number to this basic
event.
This basic event is quantified as "indeterminate" for all PDSs, including the
complementary event as well.
NOMELT = 0.5
MELT = 0.5
NORVROCKET (RVROCKET)
No Reactor Vessel Rocket (Reactor Vessel Rocket)
Description:
This basic event represents the analyst's confidence that the reactor vessel does not
become a rocket or projectile within containment. This would result from the reactor
vessel anchorage failing, thus allowing the reactor vessel to penetrate the containment
vessel.
Quantification:
The probability of this failure mode is based on NUREG/CR-4551 [Reference 1]. This
basic event is quantified as "almost certain" for all PDSs, and its complementary event
as "remotely possible".
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NORVROCKET = 0.999
RVROCKET = 0.001
NOSPARK (SPARK)
Random Spark Is Not Available (Is Available) Before Reactor Vessel Failure
Description:
This basic event addresses the availability of some random, not a design feature,
ignition source, prior to reactor vessel failure. The existence of random ignition sources
can have an impact on the timing and magnitude of hydrogen burns. (The likelihood of
the ignition is also related to the concentration of the hydrogen, which is included in
other CET events, but is expected to be low prior to reactor vessel failure.)
Quantification:
NUREG/CR-4551 [Reference 1] was reviewed for insights into random spark sources
when power is not available. It is probable that sparks will not exist when power is not
available.
To determine the existence of a spark source, simplifying assumptions were made to
infer the status of power from the core damage sequence. RB fan power was not
considered a significant spark source because the AC induction motors are sparkless.
Therefore, the availability of RB spray pumps was used to characterize power
availability. Consequently, the logic for NOSPARK (SPARK) was determined by
accounting for the unavailability (availability) of the reactor building spray system and
the conditional probability for the absence (presence) of a spark. The conditional
probability that a spark does (does not) exist when the RB spray system is not available
was assumed to be "unlikely" ("likely"). For the case in which a spark does (does not)
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exist when the RB spray system is available was assumed to be "almost certain"
("unlikely").
The conditional basic events are listed below and the decision tree logic for events
NOSPARK and SPARK are depicted below in Figures 5-3 and 5-4. Event RBSPRAY
represents the availability of RB sprays and was determined by taking the complement
of the system unavailability estimate. Therefore, since the unavailability of the RB spray
system was on the order of 1E-3, basic event RBSPRAY was deemed to be "almost
certain."
NOSPARK_9 = 0.9
NOSPARK_01 = 0.01
SPARK_1 = 0.1
SPARK_99 = 0.99
RBSPRAY = 0.999
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Figure 5-3 Logic for Event NOSPARK
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Figure 5-4 Logic for Event SPARK
NOSPARKAFT (SPARKAFT)
A Random Spark Is Unavailable (Is Available) Early After Vessel Failure
Description:
This basic event addresses the availability of some random, not a design feature,
ignition source early after the reactor vessel failure (reactor vessel failure + 5 hours).
The existence of random ignition sources can have an impact on the timing and
magnitude of combustible gas burns. (The likelihood of the ignition is also related to the
combustible gas concentration, which is included in other CET events. Higher
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combustible gas concentrations after reactor vessel failure increase the likelihood that a
deflagration results, when an ignition source is available.)
Quantification:
NUREG/CR-4551 [Reference 1] was reviewed for insights into random spark sources
when power is not available. It is probable that sparks will not exist when power is not
available.
To determine the existence of a spark source, simplifying assumptions were made to
infer the status of power from the core damage sequence. Power recovery early after
reactor vessel failure was assumed not to be a contributor (all power recovery was
included in the "late" time frame). RB fan power was not considered a significant spark
source because the AC induction motors are sparkless. Therefore, the availability of
RB spray pumps was used to characterize power availability. Consequently, the logic
for NOSPARKAFT (SPARKAFT) was determined by accounting for the unavailability
(availability) of the reactor building spray system and the conditional probability for the
absence (presence) of a spark. The conditional probability that a spark does (does not)
exist when the RB spray system is not available was assumed to be "unlikely" ("likely").
For the case in which a spark does (does not) exist when the RB spray system is
available was assumed to be "almost certain" ("unlikely").
The conditional basic events are listed below and the decision tree logic for events
NOSPARKAFT and SPARKAFT are depicted below in Figures 5-5 and 5-6. Event
RBSPRAY represents the availability of RB sprays and was determined by taking the
complement of the system unavailability estimate. Therefore, since the unavailability of
the RB spray system was on the order of 1E-3, basic event RBSPRAY was deemed to
be "almost certain."
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NOSPARKAFT_9 = 0.9
NOSPARKAFT_01 = 0.01
SPARKAFT_1 = 0.1
SPARKAFT_99 = 0.99
RBSPRAY = 0.999
Figure 5-5 Logic for Event NOSPARKAFT
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Figure 5-6 Logic for Event SPARKAFT
NOSPARKAT (SPARKAT)
Random Spark Is Unavailable (Is Available) at Reactor Vessel Failure
Description:
This basic event addresses the availability of some random, not a design feature,
ignition source immediately after the reactor vessel failure. The existence of random
ignition sources can have an impact on the timing and magnitude of hydrogen burns.
Quantification:
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The reactor cavity itself is expected to be inerted thus preventing the molten core from
acting as an ignition source. The short time frame for the spark to occur makes this
event "likely" for all PDSs, and its complementary event "unlikely".
NOSPARKAT = 0.9
SPARKAT = 0.1
NOSPARKLT (SPARKLT)
Random Spark Is Unavailable (Is Available) Late After Reactor Vessel Failure
Description:
This basic event addresses the availability of some random, not a design feature,
ignition source late after the reactor vessel failure. The existence of random ignition
sources can have an impact on the timing and magnitude of combustible gas burns.
(The likelihood of the ignition is also related to the combustible gas concentration, which
is included in other CET events. When an ignition source is available, higher
combustible gas concentrations late after reactor vessel failure increase the likelihood
that a deflagration results.)
Quantification:
This basic event is quantified as "unlikely" whenever there is some power availability in
containment. Offsite power recovery is based on a previous analysis [Reference 31]
and accounts for power restoration within a 24 hour time frame. The basic event for
unavailability of a random spark is quantified as "unlikely" for all PDSs when offsite
power has been recovered (NOSPARKLT-OP) and "almost certain" when offsite power
is unavailable (NOSPARKLT-NOP). The complementary event in which a spark is
available when offsite power is present (SPARKLT-OP) is quantified as "almost certain"
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and "unlikely" when offsite power is unavailable. The event probabilities used for
recovery and non-recovery of offsite power within 24 hours are listed below. Figures 5-
7 and 5-8 depict the logic used to represent the status of this event.
RECOFFSITEPWR = 0.964
NORECOFFSITEPWR = 0.036
NOSPARKLT-OP = 0.01
NOSPARKLT-NOP = 0.99
SPARKLT-OP = 0.99
SPARKLT-NOP = 0.01
Figure 5-7 Logic for Event NOSPARKLT
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Figure 5-8 Logic for Event SPARKLT
OPSDEPRESS (NOOPSDEPRESS)
Confidence that Operators Will (Will Not) Depressurize Steam Generators
Description:
This basic event represents the analyst's confidence that the operators will depressurize
the steam generators.
The effect of depressurizing the steam generators is to lower the RCS pressure. This
could allow injection of water by low pressure systems, accumulator discharge, and
reduced potential for DCH.
Quantification:
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Quantification is based on emergency procedures. Although the procedures do not
specifically discuss avoidance of high pressure melt ejection (HPME) as a reason for
secondary-side depressurization, the emergency operating procedure (for superheated
conditions) does instruct operators to depressurize the secondary system (open the
atmospheric dump valves, ADVs). Since the above actions are proceduralized, this
basic event can be quantified as "almost certain", with its complementary event being
"unlikely".
OPSDEPRESS = 0.99
NOOPSDEPRESS = 0.01
OTHERSCRUB (NOOTHERSCRUB)
Confidence That There Is (Is No) Fission Product Scrubbing by Other Systems Not in
Auxiliary Building
Description:
This basic event represents the analyst's confidence that scrubbing is effective on
releases outside containment that are not SGTRs or released to the Auxiliary Building.
Those isolation failures that result in releases that pass through ventilation system filters
are examples. These filters may clog, but some filtering is expected.
Quantification:
No credit is taken for fission product scrubbing in these cases. This basic event is
quantified as "impossible" for all PDSs, and its complementary event as "certain".
OTHERSCRUB = 0.0
NOOTHERSCRUB = 1.0
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OTHERWATER (NOOTHERWATER)
Water Does (Does Not) Fill Cavity from Plant Specific Sources and Paths
Description:
This basic event represents the probability that the reactor cavity will be filled with water
early after containment failure by sources other than engineered safeguards systems.
This includes, for example, refueling water systems.
Quantification:
Quantification of this basic event depends on the reactor cavity geometry, the sources
of water, and pathways to the reactor cavity. At this time, no credit is being taken for
"creative" water sources to flood the reactor cavity. Therefore, this basic event is
quantified as "impossible" for all PDSs, and its complementary event as "certain".
OTHERWATER = 0.0
NOOTHERWATER = 1.0
OXIDIZED (NOOXIDIZED)
In-Vessel Hydrogen Production Insufficient (Sufficient) to Cause Hydrogen Burns
Description:
This basic event represents the analyst's confidence that the hydrogen produced in-
vessel from zircaloy oxidation will be insufficient to cause a global burn and challenge
containment. The in-vessel hydrogen production will determine the potential for burns
prior to and at vessel failure.
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Quantification:
At Oconee, the containment volume is approximately 1.8 million cubic feet, which would
require nearly 50% of the available zircaloy to be oxidized and released to containment
to achieve global burn conditions that would challenge containment. Similar fractions
for TMI-1 would be expected.
Sensitivity studies performed with MAAP have shown that without core blockage, the
typical fraction of zircaloy oxidation is 30-40%. Additionally, if two-sided oxidation of the
clad material is allowed, along with no core blockage, the fraction is 40-50%. Therefore,
it is possible to oxidize enough zircaloy in-vessel to cause a global burn, provided that
steam availability is not blocked by molten material and very large surface areas of the
fuel cladding are exposed.
Since pessimistic assumptions are needed to produce the required hydrogen
concentration, this basic event is quantified as "almost certain" for all PDSs, and its
complementary event as "unlikely".
OXIDIZED = 0.99
NOOXIDIZED = 0.01
PDSFANS (NOPDSFANS)
PDS Indicates that Reactor Building Fans Are Available (Unavailable) at or Prior to
Reactor Vessel Failure
Description:
This basic event determines the availability of the RB fans at or prior to a reactor vessel
failure. The reactor building fans provide for long-term pressure control.
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Quantification:
Quantification of this basic event is sequence specific and is determined directly from
the fault tree logic for unavailability of the reactor building fans (top event CF). The
decision tree logic for PDSFANS (NOPDSFANS) is depicted below in Figure 5-9:
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Figure 5-9 Logic for Events PDSFANS And NOPDSFANS
PDSINJECCS (NOPDSINJECCS)
PDS Indicates that the BWST Has (Does Not Have) Water, i.e., ECCS Failed
(Succeeded) in Injection Mode
Description:
Upon vessel failure, the BWST may (if containment pressure is low) provide water to the
reactor cavity via gravity feed. This basic event indicates whether water is retained in
the BWST prior to reactor vessel failure because the safety injection systems failed in
the injection mode.
Quantification:
This basic event is quantified based directly upon information in the core damage
sequence, making use of the logic for those core melt bins involving unavailability of
injection, i.e., NOPDSINJECCS involves only those core melt bin scenarios that involve
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failure of ECCS injection. Figure 5-10 depicts the logic for PDSINJECCS and its
complementary event NOPDSINJECCS:
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Figure 5-10 Logic for Events PDSINJECCS and NOPDSINJECCS
PDSINJSP (NOPDSINJSP)
PDS Indicates that Reactor Building Sprays Are Available (Unavailable) in Injection
Mode
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Description:
The reactor building sprays provide a source of water to the reactor cavity prior to
reactor vessel failure that may have an impact on rapid steam generation when the
vessel fails.
Quantification:
This basic event is represented by system level logic associated with injection of water
from the borated water storage tank (BWST) by the reactor building spray pumps.
Figure 5-11 depicts the logic used for events PDSINJSP and NOPDSINJSP.
Figure 5-11 Logic for Events PDSINJSP and NOPDSINJSP
PDSLOW (NOPDSLOW)
PDS Indicates that Sequence Is (Is Not) a Low Pressure Core Melt
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Description:
The RCS pressure during the core melt and at vessel failure has important implications
for containment challenges and fission product retention.
A RCS pressure of less than or equal to about 300 psia at vessel failure would be
considered a low pressure core melt. The large LOCAs are considered to be low
pressure. The small LOCAs and the cycling relief valve cases are considered high
pressure scenarios.
This basic event does not include the probability of primary depressurization after core
melt and prior to vessel failure (i.e., operator opens the pressurizer PORV, primary
system failure due to heating, etc.); these are included in other CET events and have
the effect of changing some fraction of the high pressure sequences to low pressure
sequences.
Quantification:
This basic event is sequence specific. For large LOCAs, low pressure at core melt is
"certain". For small LOCAs, low pressure at core melt is "impossible," unless
depressurization occurs for another reason independent of the PDS (accounted for
elsewhere in the CET). For medium LOCAs the pressure is "indeterminate" to allow for
gradual transition between large and small core melt bins (CMBs). Thus, the events
PDSLOW and NOPDSLOW are represented by the conditional logic depicted below in
Figures 5-12 and 5-13, which is based on the CMB to which the core damage sequence
belongs.
PDSLOW_5 = 0.5
NOPDSLOW_5 = 0.5
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Figure 5-12 Logic for Event PDSLOW
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Figure 5-13 Logic for Event NOPDSLOW
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PDSNOISL (NOPDSNOISL)
PDS Indicates the Presence of No ISLOCA (Presence of ISLOCA) Initiator
Description:
This basic event is used to indicate that containment has not been bypassed by an
ISLOCA initiator.
Quantification:
This information can be determined directly from the PDS. Figures 5-14 and 5-15
display the logic used to determine the state of this event.
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Figure 5-14 Logic for Event PDSNOISL
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Figure 5-15 Logic for Event NOPDSNOISL
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PDSNOSGTR (NOPDSNOSGTR)
PDS Indicates the Presence of No SGTR (Presence of SGTR) Initiator
Description:
This basic event is used to indicate that containment has not been bypassed by an
SGTR initiator.
Quantification:
Quantification of this event can be determined directly from the core melt bin logic.
Figures 5-16 and 5-17 display the logic used to determine the state of this event.
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Figure 5-16 Logic for Event PDSNOSGTR
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Figure 5-17 Logic for Event NOPDSNOSGTR
PDSPRESSH (NOPDSPRESSH)
PDS Sequence Has (Does Not Have) High Base Pressure in Containment at Reactor
Vessel Failure
Description:
This basic event represents the analyst's confidence that a given core melt sequence
will have a high base pressure in containment at reactor vessel failure. This event is
important in order to analyze the effects of DCH, rapid steam generation, or hydrogen
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burns on the containment. Such high pressures have the potential to increase the
containment failure probability due to DCH, but suppress hydrogen burns by steam
inerting.
Quantification:
The base pressure in containment is based on the core melt sequence and the
availability of containment safeguards. In all cases, if the fan coolers are available, the
base pressure in containment will be low prior to reactor vessel failure. MAAP runs
were used to confirm the effectiveness of the RB fans for maintaining low base
pressure. Also, if the sequence is a large isolation failure, regardless of containment
safeguards, the containment will have a low pressure.
Quantification of this event is based directly upon the CSS/CIS logic [Reference 24] to
determine status of containment pressure. Figure 5-18 displays the logic used to
represent the state of this event.
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Figure 5-18 Logic for Events PDSPRESSH and NOPDSPRESSH
PDSPZRPORV (NOPDSPZRPORV)
PDS Indicates Pressurizer PORV Is (Is Not) Available
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Description:
The status of the pressurizer PORV may be important in assessing the potential for the
operators to reduce RCS pressure prior to vessel failure or tube ruptures.
Quantification:
The quantification of this event is determined directly from the Level 1 portion of the
model that determines PORV status by questioning availability of the PORV. Figure 5-
19 displays the logic used to represent the state of this event.
Figure 5-19 Logic for Events PDSPZRPORV and NOPDSPZRPORV
PDSRCEQSG (NOPDSRCEQSG)
PDS Indicates that RCS Pressure Is (Is Not) Slightly Above or is Below Steam
Generator Pressure
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Description:
This basic event is used in assessing the potential for an induced steam generator tube
rupture. Induced tube ruptures are not expected unless a significant pressure
difference exists across the tube walls.
The pressure difference will be of no concern for large or medium LOCAs, or small
LOCAs with SSHR available. Small LOCAs without SSHR and cycling relief valve
cases have the potential for high tube differential pressure.
Quantification:
Quantification of this event can be determined directly from the core melt bin logic.
Figure 5-20 displays the logic used to determine the state of this event.
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Figure 5-20 Logic for Events PDSRCEQSG and NOPDSRCEQSG
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PDSRCPWR (NOPDSRCPWR)
PDS Indicates that Power Is (Is Not) Available to RCPs
Description:
This basic event represents whether power is available to the reactor coolant pumps.
Quantification:
Quantification of this event can be determined directly from the Level 1 logic for
unavailability of RCP support systems. Figure 5-21 displays the logic used to determine
the state of this event.
Figure 5-21 Logic for Events PDSRCPWR and NOPDSRCPWR
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PDSSGADV (NOPDSSGADV)
PDS Indicates Steam Generator Atmospheric Dump Valves (ADVs) Are (Are Not)
Available
Description:
The steam generator ADVs are important in assessing the potential for the operators to
reduce RCS pressure through secondary depressurization prior to reactor vessel failure.
Quantification:
Quantification of this event can be determined directly from the Level 1 logic for
unavailability of the ADVs. Figure 5-22 displays the logic used to determine the state of
this event.
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Figure 5-22 Logic for Events PDSSGADV and NOPDSSGADV
PDSSPRAY (NOPDSSPRAY)
PDS Indicates that Containment Sprays Are (Are Not) Available
Description:
The containment sprays provide for long-term containment pressure control and fission
product scrubbing. Thus, success for this basic event indicates that sprays are
available in both injection and recirculation modes.
Quantification:
This event was quantified using the system level logic for reactor building (RB) sprays.
The system top event for unavailability of RB sprays was quantified and found to be on
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the order of 1E-3 ("remotely possible"). Therefore, availability of RB sprays is the
complementary value, i.e., 0.999 ("almost certain"). To simplify the Level 2 model, a
single basic event was used that represents availability of the RB spray system early in
the accident scenario. Figure 5-23 displays the logic used to determine the state of this
event.
RBSPRAY = 0.999
Figure 5-23 Logic for Events PDSSPRAY and NOPDSSPRAY
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PDSSSHR (NOPDSSSHR)
PDS Indicates SSHR Is (Is Not) Available
Description:
This basic event establishes the availability of Secondary side heat removal (SSHR)
during the core damage period. SSHR is important in assessing the potential for
secondary depressurization and for preventing tube ruptures.
Quantification:
The quantification of this event can be determined from the Level 1 logic for
unavailability of SSHR. Figures 5-24 and 5-25 display the logic used to determine the
state of this event.
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Figure 5-24 Logic for Event PDSSSHR
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Figure 5-25 Logic for Event NOPDSSSHR
PLATEOUT (NOPLATEOUT)
Confidence that Plateout Will (Will Not) Scrub Fission Products
Description:
This basic event represents the analyst's confidence that plateout will be effective in
reducing fission product releases to the Auxiliary Building for ISLOCAs and isolation
failures.
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Quantification:
NUREG/CR-1989, IDCOR Technical Report 11.6 and 85.2, and MAAP models indicate
that plateout is expected [Reference 9, 20, and 19]. These studies have shown that
aerosol particles will adhere to surface in containment. In addition due to the solubility
of CsI and CsOH in water, their potential for dispersion is essentially eliminated even if
water dries out. However, it was conservatively decided that this phenomenon should
not be credited, such that this basic event is quantified as "impossible" for all PDSs, and
its complementary event as "certain".
PLATEOUT = 0.0
NOPLATEOUT = 1.0
PRISTREN1 (NOPRISTREN1)
Confidence that Containment Can (Cannot) Handle Hydrogen Burn Pressure with High
Base Pressure Prior to Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that containment will remain intact
given a hydrogen burn prior to reactor vessel failure with high base pressure. The
quantification of this basic event reflects only the phenomenon of hydrogen burn.
Quantification:
The quantification for this event is similar to that for AFTSTREN1. This basic event is
quantified as "indeterminate" for all PDSs. The complementary event is also quantified
as "indeterminate".
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PRISTREN1 = 0.5
NOPRISTREN1 = 0.5
PRISTREN2 (PRISTREN2)
Confidence that Containment Can (Cannot) Handle Hydrogen Burn Pressure with Low
Base Pressure Prior to Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that containment will remain intact
given a hydrogen burn prior to reactor vessel failure with low containment base
pressure.
Quantification:
The quantification for this event is similar to that for AFTSTREN2. This basic event is
quantified as "almost certain" for all PDSs, and its complementary event as "remotely
possible".
PRISTREN2 = 0.999
NOPRISTREN2 = 0.001
PRVHPCONF (NOPRVHPCONF)
Confidence that Pressurizer PORV Can (Cannot) Prevent HPME
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Description:
This basic event represents the analyst's confidence that the pressurizer PORV is
capable of sufficiently depressurizing the RCS so that a HPME (primary pressure above
about 300 psia) can be prevented.
Quantification:
At TMI-1, because the pressurizer PORV is small (about 1 inch in diameter), RCS
pressure may not decrease to 300 psia before vessel failure. Although the PORV may
help reduce pressure, its effectiveness is uncertain for severe accident conditions.
Therefore, little credit is being taken for the PORV's effectiveness at this time. Thus,
this basic event is quantified as "remotely possible" for all PDSs, and its complementary
event as "almost certain".
PRVHPCONF = 0.001
NOPRVHPCONF = 0.999
PZPORVCONF (NOPZPORVCONF)
Confidence that Operators Will (Will Not) Manually Open the Pressurizer PORV
Description:
This basic event represents the analyst's confidence that the operators will open the
PORV and depressurize the RCS prior to reactor vessel failure.
The considerations for quantifying this basic event are the same as those already
discussed in EFFDEPRESS. However, the timing is different in that reactor vessel
failure is involved instead of a steam generator tube rupture.
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Quantification:
Quantification depends on emergency procedures. The emergency operating
procedure (for superheated conditions) does instruct operators to depressurize the
primary system (open the PORV). Since the above actions are proceduralized, this
basic event can be quantified as "almost certain" for most PDSs.
The exception, however, is where the PDS indicates the operators have already failed
to go on HPI cooling, in which case the analysis will not give credit for another
opportunity. For those PDSs that involve operator failure to open the PORV, such as to
initiate HPI cooling, it was determined that the conditional probability to later open the
PORV would be 0.0. Therefore, the resulting probability of PZPORVCONF is based on
logic to account for those core damage sequences involving operator failure to initiate
HPI cooling. The success and failure logic for this event is depicted below in Figure 5-
26.
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Figure 5-26 Logic for Events PZPORVCONF and NOPZPORVCONF
PZRPORVDEP (PZRNOPORVDEP)
Confidence that the Pressurizer PORV Can (Cannot) Depressurize the Primary System
to at or Below Steam Generator Pressure
Description:
This event represents the analyst's confidence that the pressurizer PORV can
depressurize the primary system to at or below steam generator pressure, thereby
preventing an induced SGTR.
If the pressurizer PORV can depressurize the primary system to at or below the steam
generator pressure, the pressure drop across the steam generator tubes will be lower.
Therefore, the probability of a SGTR is reduced.
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Quantification:
Opening the pressurizer PORV will help reduce RCS pressure, however, success
depends on the rate of competing steam and gas generation, and the steam generator
pressure. Although the PORV may help reduce pressure, its effectiveness is uncertain
for severe accident conditions. Therefore, little credit is being taken for the PORV's
effectiveness at this time. Thus, this basic event is quantified as "remotely possible" for
all PDSs, and its complementary event as "almost certain".
PZRPORVDEP = 0.001
PZRNOPORVDEP = 0.999
PZRSAFETY (NOPZRSAFETY)
Pressurizer Safety Valve Sticks (Does Not Stick) Open During Core Damage
Description:
This basic event represents the likelihood that a pressurizer safety valve sticks open
while passing the hot gases and aerosols generated during core damage.
Quantification:
Based on a literature search (e.g., NUREG/CR-4551 [Reference 1]) and failure rates
determined by the Level 1 analyst, this basic event is quantified as "unlikely" for all
PDSs, and its complementary event as "likely".
PZRSAFETY = 0.1
NOPZRSAFETY = 0.9
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RECACPRI (NORECACPRI)
Power Is (Is Not) Recovered to RCPs Prior to Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that reactor coolant pump power is
recovered during the time period from the onset of core damage to reactor vessel
failure.
Quantification:
Since this is a fairly short time period, no credit is being taken for this type of recovery.
Therefore, this basic event is quantified as "impossible" for all PDSs, and its
complementary event as "certain".
RECACPRI = 0.0
NORECACPRI = 1.0
RECOVFANSAFT (NORECOVFANSAFT)
Reactor Building Fans Are (Are Not) Recovered Early After Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that the RB fans can be recovered
early after reactor vessel failure.
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Quantification:
Since this is a fairly short time period, no credit is being taken for this type of recovery.
Therefore, this basic event is quantified as "impossible" for all PDSs, and its
complementary event as "certain".
RECOVFANSAFT = 0.0
NORECOVFANSAFT = 1.0
RECOVFANSLT (NORECOVFANSLT)
Reactor Building Fans are Recovered Prior to Late Containment Failure
Description:
This event represents the analyst's confidence that the RB fans can be recovered prior
to late containment failure.
This event is applicable only to those PDSs where fans are unavailable. For those
PDSs where RB fans are available, this event is not applicable.
Quantification:
This event is quantified directly from the system level logic for unavailability of reactor
building fans. If the RB Fan system is unavailable due to power failure, then recovery of
offsite power is based on a recovery (non-recovery) probability based on a previous
analysis [Reference 31]. When power failure was not the cause, then RB fans are
assumed in this analysis to be subject to mechanical and non-power related failures.
Post-LOOP recovery logic was used for determining whether RB fans were actually
unavailable, along with consideration of offsite power restoration. To determine the
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probability in which RB fans do not fail in a post-LOOP scenario, the complement of the
failure probability (quantified to be on the order of 1E-2) was used. Therefore, the event
RECFANSLT was created and assumed to be "almost certain". Figure 5-27 displays
the logic used to determine the state of this event.
RECFANSLT = 0.99
RECOFFSITEPWR = 0.964
NORECOFFSITEPWR = 0.036
Figure 5-27 Logic for Events RECOVFANSLT and NORECOVFANSLT
RECOVFANSPRI (NORECOVFANSPRI)
Reactor Building Fans Are (Are Not) Recovered at or Prior to Reactor Vessel Failure
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Description:
This basic event represents the analyst's confidence that the RB fans can be recovered
at or prior to reactor vessel failure.
Quantification:
Since this is a fairly short time period, no credit is being taken for this type of recovery.
Therefore, this basic event is quantified as "impossible" for all PDSs, and its
complementary event as "certain".
RECOVFANSPRI = 0.0
NORECOVFANSPRI = 1.0
RECOVRV (NORECOVRV)
Recovery of Core Cooling Prevents (Does Not Prevent) Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that core cooling is recovered
following the onset of core damage, but early enough to prevent reactor vessel failure.
Quantification:
At TMI-1, no credit is being taken for this recovery. Therefore, this basic event is
quantified as "impossible" for all PDSs, and its complementary event as "certain".
RECOVRV = 0.0
NORECOVRV = 1.0
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RECOVSPAFT (NORECOVSPAFT)
Containment Sprays Are (Are Not) Recovered Early After Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that the containment sprays can be
recovered early after reactor vessel failure.
Quantification:
Since this is a fairly short time period, no credit is being taken for this type of recovery.
Therefore, this basic event is quantified as "impossible" for all PDSs, and its
complementary event as "certain".
RECOVSPAFT = 0.0
NORECOVSPAFT = 1.0
RECOVSPLT (NORECOVSPLT)
Containment Sprays Are (Are Not) Recovered Prior to Late Containment Failure
Description:
This event represents the analyst's confidence that the containment sprays can be
recovered prior to late containment failure.
This event is applicable only to those PDSs where RB Sprays are unavailable. For
those PDSs where sprays are available, this event is not applicable.
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Quantification:
This event is quantified using the system level logic for unavailability of reactor building
sprays. If the RB spray system is unavailable due to power failure, then recovery of
offsite power is based on a recovery (non-recovery) probability based on a previous
analysis [Reference 31]. The time frame for late recovery of offsite power was assumed
to be 24 hours. When power failure was not the cause for failure of RB sprays, then
mechanical and non-power related failures are assumed. Hence, the post-LOOP
recovery system fault tree event for RB spray (CS01-R) was used for determining
whether RB sprays were actually unavailable, along with consideration of offsite power
restoration. To determine the probability in which RB sprays do not fail in a post-LOOP
scenario, the complement of the failure probability was used (system unavailability was
quantified to be on the order of 1E-3). Therefore, the complementary event
RECSPRAYLT, representing availability of the RB spray system, was created and
assumed to be "almost certain". Figure 5-28 displays the logic used to determine the
state of this event. Also, conditional LOOP logic was added to identify that in order for
power to be restored, it had to have been initially lost, otherwise the RB spray system
would not require restoration of offsite power.
RECSPRAYLT = 0.999
RECOFFSITEPWR = 0.964
NORECOFFSITEPWR = 0.036
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Figure 5-28 Logic for Events RECOVSPLT and NORECOVSPLT
RECOVSPPRI (NORECOVSPPRI)
Containment Sprays Are (Are Not) Recovered Prior to Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that the reactor building sprays can
be recovered at or prior to reactor vessel failure.
Quantification:
This is a fairly short time period. No credit is being taken for this type of recovery.
Therefore, this basic event is quantified as "impossible" for all PDSs, and its
complementary event as "certain".
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RECOVSPPRI = 0.0
NORECOVSPPRI = 1.0
RECOVSSHR (NORECOVSSHR)
SSHR Is (Is Not) Recovered Prior to Reactor Vessel Failure
Description:
The basic event represents the analyst's confidence that SSHR can be recovered
following the onset of core damage, but prior to reactor vessel failure.
Quantification:
At TMI-1, no credit is being taken for this recovery. Therefore, this basic event is
quantified as "impossible" for all PDSs, and its complementary event as "certain".
RECOVSSHR = 0.0
NORECOVSSHR = 1.0
RELLOC (NORELLOC)
Release of Fission Products Is (Is Not) In Lower Sections of Auxiliary Building
Description:
This event is used in conjunction with PLATEOUT to credit fission product scrubbing
within the Auxiliary Building. This event applies to PDSs with isolation failures or
ISLOCAs to the Auxiliary Building. Release at low elevations will allow more time for
plateout on equipment, thus reducing the amount of fission products released to the
environment.
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Quantification:
Quantification of this event is dependent on the location of the isolation failure. The
plant layout at TMI-1 is favorable for plateout for those PDSs that involve isolation
failures and ISLOCAs (except for those that involve steam generator tube rupture).
This event is quantified based directly upon the PDS for each particular core damage
sequence. The logic for determining the state of this event is depicted below in Figures
5-29 and 5-30.
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Figure 5-29 Logic for Event RELLOC
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Figure 5-30 Logic for Event NORELLOC
RSGFANSEFF (NORSGFANSEFF)
Confidence that Reactor Building Fans Can (Cannot) Handle Rapid Steam Production
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Description:
This basic event represents the analyst's confidence that RB fans can control the
containment pressure rise that results when the vessel fails and the corium contacts
water in the cavity.
Quantification:
Analyses with MAAP have shown that the RB fans can be effective in mitigating the
pressure spike for a rapid steam generation (RSG) event. This is mainly due to the time
over which this phenomenon takes place. (Up to 5 hours after reactor vessel failure is
considered early containment failure.) Unlike DCH, an RSG event could take several
hours to develop. Thus, some credit is taken for the potential mitigative effects of the
RB fans.
This basic event is quantified as "indeterminate" for all PDSs. The complementary
event is also quantified as "indeterminate".
RSGFANSEFF = 0.5
NORSGFANSEFF = 0.5
RSGSTREN1 (NORSGSTREN1)
Confidence that Containment Strength Can (Cannot) Handle RSG Pressure Spike with
High Base Pressure
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Description:
This basic event represents the analyst's confidence that containment will remain intact
following rapid steam generation at vessel failure with a high base pressure. The
quantification of this event reflects only the phenomenon of RSG.
Quantification:
The analysis of Oconee's containment response to RSG has shown that given a
sufficient supply of water inside containment, the pressure rise could be 100 to 120 psia
over several hours, without containment safeguards. This result also varied depending
on the heat transfer rate between the dispersed corium and the water pool. This
pressure rise, on top of a high base pressure (40 to 100 psia), could result in pressures
of 140 to 220 psia.
Based on the analyst's judgment, and the containment failure probability curve in the
Oconee PRA (see Appendix C, since the TMI-1 containment ultimate strength is similar
to the Oconee containment ultimate strength), this basic event is quantified as "remotely
possible" for all PDSs, and its complementary event as "almost certain".
RSGSTREN1 = 0.001
NORSGSTREN1 = 0.999
RSGSTREN2 (NORSGSTREN2)
Confidence that Containment Strength Can (Cannot) Handle RSG Pressure Spike with
Low Base Pressure
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Description:
This basic event represents the analyst's confidence that containment will remain intact
following rapid steam generation at vessel failure with a low base pressure inside
containment. The quantification of this basic event reflects only the phenomenon of
RSG.
Quantification:
For low base pressures, the probability of containment failure due to rapid steam
generation is low. Based on the analyst's judgment, this basic event is quantified as
"almost certain" for all PDSs, and its complementary event as "unlikely".
RSGSTREN2 = 0.99
NORSGSTREN2 = 0.01
SEQPRESSH (NOSEQPRESSH)
PDS Indicates (Does Not Indicate) High Base Pressure in Containment Early After
Reactor Vessel Failure
Description:
This event represents the analyst's confidence that a given core melt sequence will
have a high base pressure in containment early after reactor vessel failure.
This event is important in order to analyze the effects of combustible gas burns on the
containment. Such high pressures have the potential to suppress combustible gas
burns by steam inerting. Steam inerting early after reactor vessel failure indicates that
the containment safeguards are not available.
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Quantification:
The base pressure in containment is based on the core melt sequence and the
availability of containment safeguards. If the RB fan coolers are available, the base
pressure in containment will be low because non-condensable gases are not a major
influence on containment pressure at this point. MAAP was used to confirm the
effectiveness of the RB fan coolers for maintaining low base pressure. Also, if the
sequence is a large isolation failure, regardless of containment safeguards availability,
the containment will have a low pressure early after reactor vessel failure.
Quantification of this event is based directly upon the logic that determines the CSS/CIS
portion of the PDS for each core damage sequence. Figure 5-31 depicts the logic used
to model the state of this event.
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Figure 5-31 Logic for Events SEQPRESSH and NOSEQPRESSH
SEQPRESSL (NOSEQPRESSL)
PDS Sequence Indicates (Does Not Indicate) Low Base Pressure in Containment Late
After Reactor Vessel Failure
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Description:
This basic event represents the analyst's confidence that there are low pressures in
containment late after reactor vessel failure.
Quantification:
Late after reactor vessel failure, non-condensable gases as well as steam contribute to
containment pressure. While RB fan coolers are effective for reduction of steam
pressure, they are ineffective for reducing pressure from non-condensable gas
generation. However, RB sprays, if providing a constant flow of water to the cavity, will
preclude pressure buildup from non-condensable gases due to the suspension or
termination of concrete attack. Thus, late low base pressures in containment indicate
that RB Sprays are operating. Also, if the sequence is a large isolation failure,
regardless of containment safeguards availability, the containment will have a low
pressure late after RV failure. Success for this event indicates that RB sprays are
available in both injection and recirculation modes.
Quantification of this event is based directly upon the logic that determines the CSS/CIS
portion of the PDS for each core damage sequence. Figure 5-32 depicts the logic used
to model the state of this event.
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Figure 5-32 Logic for Events SEQPRESSL and NOSEQPRESSL
SGREL (NOSGREL)
Fission Products Are (Are Not) Released to the Steam Generator
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Description:
In the fission product scrubbing tree, this event identifies whether or not the core
damage sequence involves a steam generator tube rupture initiator. That is, this event
determines whether the fission product release is through the steam generators, which
is the available release path following a tube rupture.
Quantification:
Quantification of this event is based directly upon the logic that determines the core melt
bin portion of the PDS for each core damage sequence. Figure 5-33 depicts the logic
used to model the state of this event.
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Figure 5-33 Logic for Events SGREL and NOSGREL
SLOWHTRATE (FASTHTRATE)
Heat Transfer Rate from Corium to Water Pool is Slow (Fast)
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Description:
This basic event represents the analyst's confidence that heat transfer from the corium
to the water in the cavity is not so rapid that containment failure results.
Quantification:
The quantification of this event reflects the analyst's judgment based on a review of the
literature [Reference 7]. The probability of containment failure from quasi-static
overpressure due to a steam explosion is judged to be very small. The term quasi-static
refers to a rapid pressurization of containment that does not result in shock waves that
would dynamically load the containment vessel. The pressurization is a steady but
rapid increase in pressure that statically loads the containment vessel; therefore, in
literature it is referred to as a quasi-static overpressurization.
Therefore, this basic event is quantified as "almost certain" for all PDSs, and its
complementary event as "unlikely".
SLOWHTRATE = 0.99
FASTHTRATE = 0.01
SPRAYEFF (NOSPRAYNOEFF)
Confidence that Sprays Will (Will Not) Scrub Fission Products for a Small Containment
Failure
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Description:
This basic event represents the analyst's confidence that containment sprays are
effective in scrubbing fission products from the containment atmosphere when the
containment is failed.
Quantification:
A review of the literature, NUREG/CR-1989 [Reference 9], has shown that sprays are
very effective at removing fission products from the atmosphere. However, with a small
isolation failure there is some possibility of fission products bypassing the sprays.
Therefore, this basic event is quantified as "likely" for all PDSs, and its complementary
event as "unlikely".
SPRAYEFF = 0.9
SPRAYNOEFF = 0.1
SPRAYEFFLT (NOSPRAYEFFLT)
Confidence that Sprays Will (Will Not) Scrub Fission Products Prior to Release to
Environment
Description:
This basic event represents the analyst's confidence that containment sprays are
effective in scrubbing fission products from the containment atmosphere when there is
no containment failure.
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Quantification:
As discussed above for event SPRAYEFF, sprays are very effective at scrubbing fission
products. Therefore, this basic event is quantified as "almost certain" for all PDSs, and
its complementary event as "unlikely".
SPRAYEFFLT = 0.99
NOSPRAYEFFLT = 0.01
SSHRREVAP (SSHRNOREVAP)
Confidence that SSHR Will (Will Not) Prevent Revaporization
Description:
This basic event represents the analyst's confidence that fission products will plateout
on steam generator tubes given that SSHR is available. This basic event is important
for the revaporization phenomenon.
Quantification:
If SSHR is available, this will create a "cold spot" in the RCS that will attract fission
products released from the core. Analyses with MAAP have shown that this "cold spot"
is very efficient for attracting fission products. Therefore, this basic event is quantified
as "likely" for all PDSs, and its complementary event as "unlikely".
SSHRREVAP = 0.9
SSHRNOREVAP = 0.1
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SSHRRVPREC (SSHRNORVPREC)
SSHR Removal Is (Is Not) Recovered Prior to Revaporization
Description:
This basic event represents the analyst's confidence that SSHR is recovered prior to
revaporization of the fission products.
Quantification:
The quantification of this basic event depends on the probability that the operators will
take the appropriate action to recover SSHR prior to revaporization. At TMI-1, no credit
is being taken for post-core damage recovery of SSHR. Therefore, this basic event is
quantified as "impossible" for all PDSs, and its complementary event as "certain".
SSHRRVPREC = 0.0
SSHRNORVPREC = 1.0
SSHRSGTCOOL (SSHRSGTNOCOOL)
Confidence that SSHR Will (Will Not) Keep Steam Generator Tubes Cool
Description:
This basic event represents the analyst's confidence that SSHR will keep the steam
generators tubes cool in the presence of hot RCS gases. This basic event is important
in assessing the likelihood of an induced steam generator tube rupture event during
severe accident conditions.
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Quantification:
SSHR has been shown to be very effective in cooling the steam generators tubes. In
once-through steam generators (OTSG), the water level does not cover the entire
steam generator tube surface. Hot gases in the RCS coming in contact with the steam
generator tubes may decrease the effectiveness of the tube cooling by SSHR.
Therefore, this basic event is quantified as "indeterminate". The complementary event
is also quantified as "indeterminate".
SSHRSGTCOOL = 0.5
SSHRSGTNOCOOL = 0.5
STREN1H2 (NOSTREN1H2)
Confidence that Containment Can (Cannot) Handle Combustible Gas Burn Pressure
with High Base Pressure Late After Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that containment remains intact
following a late combustible gas burn with a high base pressure in containment. The
quantification of this basic event reflects only the phenomenon of a combustible gas
burn.
Quantification:
The quantification for this event is similar to that for AFTSTREN1. This basic event is
quantified as "indeterminate" for all PDSs. The complementary event is also quantified
as "indeterminate".
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STREN1H2 = 0.5
NOSTREN1H2 = 0.5
STREN2H2 (NOSTREN2H2)
Confidence that Containment Can (Cannot) Handle Combustible Gas Burn Pressure
with Low Base Pressure Late After Reactor Vessel Failure
Description:
This basic event represents the analyst's confidence that containment remains intact
following a late combustible gas burn with a low base pressure in containment. The
quantification of this basic event represents only the phenomenon of a combustible gas
burn.
Quantification:
The quantification for this event is similar to that for AFTSTREN2. This basic event is
quantified as "almost certain" for all PDSs, and its complementary event as "remotely
possible".
STREN2H2 = 0.999
NOSTREN2H2 = 0.001
WALLSURVIV (WALLNOSURVIV)
Containment Wall Survives (Does Not Survive) Contact with Corium
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Description:
This basic event represents the analyst's confidence that the containment wall would
survive direct contact with corium.
Quantification:
In general, direct contact of corium would be expected to cause a failure of the
containment wall. Therefore, this basic event is quantified as "unlikely" for all PDSs,
and its complementary event as "likely".
WALLSURVIV = 0.1
WALLNOSURVIV = 0.9
WATEREFF (NOWATEREFF)
Confidence that Water in Steam Generator Will (Will Not) Scrub Fission Products
Description:
This basic event represents the analyst's confidence that fission products released to
the steam generators will be scrubbed prior to release to the environment.
Quantification:
Review of the literature, NUREG/CR-1989 [Reference 9], has shown that water pools
are an effective way to scrub fission products. Iodine (I2) and cesium iodine (CsI) have
a high affinity for water due to their solubility in water. The degree of retention is
dependent on pH and length of time I2 and CsI is in contact with water. A substantial
amount of these chemical forms of iodine will be retained in the primary system water.
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This basic event is quantified based on both the likelihood that the break is located
underwater, which is based on TMI-1's steam generator design, and the effectiveness of
the water to scrub fission products if the break is underwater. Breaks located near the
upper tube sheets will not be underwater; therefore, no credit can be taken for
scrubbing within the steam generator prior to release to the environment. The likelihood
that the break is located underwater is "indeterminate" for all PDSs. The effectiveness
of scrubbing by water in the steam generator if the break is located underwater is
quantified as "likely" for all PDSs. As a result, the overall quantification of this basic
event is "indeterminate" for all PDSs. The complementary event is also quantified as
"indeterminate".
WATEREFF = 0.5
NOWATEREFF = 0.5
5.4 REMOVAL OF ILLOGICAL CUTSETS VIA RECOVERY RULES
RECZED
Release Category Cutset is Invalid for Reactor Building Spray Logic
Description:
This basic event is assigned a probability of 0.0 and is appended to those cutsets
quantified by FORTE version 2.2f that exhibit illogical or contradictory combinations
involving availability and effectiveness of the RB spray system. This logic resides in the
CAFTA recovery rules fault tree that is used by QRECOVER32 when post-processing
Level 2 cutsets. Four separate conditions were developed to exclude cutsets containing
basic event combinations that appear contradictory, e.g., event NOSPRAYEFFLT in
combination with an event that satisfies the gate FPSCRUBBED. Removal of these
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cutsets was necessary in order to avoid overestimating release category frequencies
associated with fission product scrubbing. These conditions and their recovery rules
logic are depicted below in Figure 5-34.
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Figure 5-34 Recovery Rules Logic to Exclude Illogical Cutsets Regarding the Reactor Building
Spray System
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6.0 SOURCE TERM CALCULATIONS AND RELEASE CATEGORY DEFINITIONS
6.1 INTRODUCTION
This section of the in-plant analysis describes the development of release categories for
TMI-1 using the Generic Level 2 Analysis. This analysis is based on the original 1993
IPE Level 2 PRA. The descriptions that follow relate to that original body of work
utilizing MAAP version 3.0B. Section 6.2 describes the MAAP 3.0B code and the
methods it uses to model accident sequences. Section 6.4 describes selected
sequences that are used to define the radionuclide release fractions, timing and release
characteristics of each release category. Section 6.5 describes how the original release
categories were further binned into the final nine release categories that were used to
calculate the offsite consequences in support of the TMI License Renewal project.
6.2 MAAP COMPUTER MODEL
The phenomological models developed by the IDCOR Program have been incorporated
into an integrated analysis code (MAAP) to analyze the major degraded core accident
scenarios for light water reactors. MAAP is designed to provide realistic assessments
for severe core damage accident sequences, including fission product release,
transport, and deposition. The following sections describe the RCS nodalization,
containment nodalization, and the safety systems modeled in the MAAP-PWR code as
applied to the Oconee large, dry containment design, used as the basis for this generic
study.
6.2.1 MAAP NODALIZATION
The MAAP plant model for a large, dry containment is represented by various nodes.
Nodes exist for the upper compartment (compartment A), lower compartment
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(compartment B), annular compartment (compartment D), reactor cavity (compartment
C), quench tank (pressurizer relief tank), and RCS.
The RCS is divided into fifteen nodes:
1. Core region
2. Upper plenum
3. Reactor dome
4. Downcomer
5. Broken loop cold leg
6. Broken loop hot leg
7. Unbroken loop cold leg
8. Unbroken loop hot leg
9. Pressurizer
10. Broken loop intermediate leg
11. Broken loop cold leg tubes
12. Broken loop hot leg tubes
13. Unbroken loop intermediate leg
14. Unbroken loop cold leg tubes
15. Unbroken loop hot leg tubes
This RCS nodalization permits a detailed accounting of the water which is available for
cooling the core and for reacting with the zircaloy fuel cladding. In addition, this scheme
allows the user to track hydrogen and fission product concentrations through the RCS
and thereby calculate release rates to the containment. The core is further divided into
a user selected number of subnodes; a 4 radial by 16 axial nodalization is used for this
TMI-1 analysis based on the Oconee model.
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6.2.2 SAFETY SYSTEMS MODELED IN MAAP
The safety systems considered in this analysis include the high pressure injection
pumps, low pressure injection pumps, core flood tanks, emergency feedwater,
containment sprays and containment air cooling units. Other components important to
accident progression include the pressurizer and steam generator safety valves and
PORVs. All of these systems can be enabled or disabled by the use of "event codes" in
MAAP at the discretion of the user. The MAAP User's Manual (Reference 27) gives a
complete description of the use of MAAP and also compares the physical models with
pertinent experiments.
6.3 RELEASE CATEGORY PARAMETER ANALYSIS
The in-plant analysis of this TMI-1 Level 2 analysis relies heavily on the Oconee PRA
(Reference 26), NUREG-1150 (Reference 30), and NUREG/CR-4551 (Reference 29) to
analyze the sequences and to develop an understanding of the thermal-hydraulic and
release characteristics of each PDS. Many individual sequences were modeled with the
MAAP code.
Many of the release category definitions discussed in Section 6.4 are derived directly
from the MAAP runs. The MAAP run identifications are shown in the tables, and are on
file at Duke Power Company. The release energies are all taken from similar
sequences in NUREG/CR-4551.
For release categories where no MAAP runs exist, parameters were often derived from
parameters of similar release categories. Scrubbing and plateout were generically
accounted for by a factor of five reduction in all the radionuclide release fractions except
the noble gases. The noble gases are assumed to be unaffected by scrubbing or plateout.
Auxiliary Building is not modeled in MAAP; therefore, scrubbing or plateout in the Auxiliary
Building is not accounted for in the source term calculation by MAAP.
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Ex-vessel release of radionuclides that evolve due to corium concrete attack are accounted
for in MAAP. The operation of containment sprays in the injection and the recirculation
modes terminates corium heat-up, corium concrete attack and ex-vessel release of
radionuclides. The availability of containment sprays is a variable modeled in the MAAP
model. Therefore, this parameter can be manipulated to properly account for the source
term attributed to ex-vessel releases versus no ex-vessel releases. A ratio of similar
sequences is used to account for sequences with and without ex-vessel releases. For
example, ex-vessel releases for ISLOCAs (RC 2.01 - 2.04) were determined from ratios of
SGTR releases (RC 1.01 - RC 1.04).
The following section will describe generic release categories along with representative
sequences which are important because of their unique release characteristics.
6.4 RELEASE CATEGORY DEFINITIONS
This section defines the release categories for the TMI-1 Level 2 analysis.
The parameters that define a release category and are important in the analysis of
offsite consequences are:
1. Time of release
2. Duration of release
3. Energy of release
4. Warning time for evacuation
5. Isotopic fractions released to the environment
With the new modeling methods of the CET described in Section 5.0, each endpoint is
capable of describing a unique sequence with unique release characteristics. As a
result, 39 release categories were identified with all endpoints having a unique release
category designation. These 39 generic release categories are discussed in Section
6.4.1
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A numbering scheme is used to separate major categories:
1 = Containment Bypass with Auxiliary Building Bypass
2 = Interfacing-Systems LOCA
3 = Large Isolation Failures
4 = Small Isolation Failures
5 = Early Containment Failure
6 = Late Containment Failure (Catastrophic)
7 = Late Containment Failure (Benign)
8 = Basemat Melt-Through
9 = No Containment Failure
Different sequences within these major categories were given a designation such as
1.01, 1.02, 2.01, etc.
6.4.1 RELEASE CATEGORY DISCUSSION
The release category discussions are divided into three separate parts. The first part
provides a generic description of the release category as defined by the path it traces
through the generic containment event tree.
Part two, the representative sequence, is based on a review of several completed PRAs
and is verified as generally corresponding with the dominant TMI-1 sequences. The
representative sequence serves as a bench mark from which the TMI-1 release
parameters are actually derived. They are typical of the sequences that dominate the
particular release categories.
Table 7-1 in Section 7.0 presents the frequencies for each of the calculated release
categories as determined by solution of the CET (see Figure 2-1).
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The TMI-1 applicability discussion is the last part of the release category
documentation. This discussion provides some insights into the range of TMI-1
sequences that result in the particular release category. Many other sequences could
follow the same path through the containment event tree and thus arrive at the same
release category. Each release category is comprised of several sequences. The
dominant sequences for TMI-1 are then compared to the representative sequence and
a TMI-1 applicability statement is developed. As necessary, additional MAAP runs for
sequences are performed to benchmark the TMI-1 release parameters for a particular
release category. These sequences can potentially have significant differences in
release parameters. The sensitivity of these release differences, to the TMI-1 results, is
discussed in this part.
Release Category 1.01
Description:
This release category is characterized as a bypass of containment with releases going
outside the Auxiliary Building. Fission product scrubbing is available.
Representative Sequence:
The representative sequence is a steam generator tube rupture. The SGTR can be the
initiating event or it can be induced. During the blowdown of the primary system
through the secondary system, a steam line safety valve sticks open. This allows a
containment bypass pathway throughout the entire accident. Core melt occurs as a
result of an early injection failure. SSHR is available allowing the tubes to be covered
and the fission products to be scrubbed. The characteristics of this release category
can be found in Table 6-4.
TMI-1 Applicability:
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All dominant TMI-1 sequences correspond well with the above representative sequence
and its associated release parameters. A TMI-1 MAAP run (TMI18C) was performed to
develop TMI-1 specific release fractions.
Release Category 1.02
Description:
This release category is characterized as a bypass of containment with releases going
outside the Auxiliary Building. Fission product scrubbing is not available.
Representative Sequence:
The representative sequence is a steam generator tube rupture. The SGTR can be the
initiating event or it can be induced. During the blowdown of the primary system
through the secondary system, a steam line safety valve sticks open. This allows a
containment bypass pathway throughout the entire accident. Core melt occurs as a
result of an early injection failure. SSHR is not available for fission products to be
scrubbed. The characteristics of this release category can be found in Table 6-5.
TMI-1 Applicability:
All dominant TMI-1 sequences correspond well with the above representative sequence
and its associated release parameters. A TMI-1 MAAP run (TMI18C4) was performed
to develop TMI-1 specific release fractions.
Release Category 2.01
Description:
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This release category is characterized as a bypass of containment with releases going
into the Auxiliary Building. Ex-vessel fission products are not released and scrubbing is
available. The representative sequence is an interfacing-systems LOCA to the Auxiliary
Building.
Representative Sequence:
The representative sequence is a medium sized interfacing-systems LOCA to the
Auxiliary Building. An example of this would be a failure of the RCS letdown line
outside containment. This provides a direct pathway from the hot leg to the Auxiliary
Building. A loss of all feedwater, all HPI and LPI injection, and all containment
safeguards is assumed. Since the release is low in the Auxiliary Building credit is taken
for fission product plateout.
A medium sized LOCA is created which will lower the RCS below 400 psia prior to core
melt. With the dispersal of the corium over a large area, the debris bed cools
preventing release of ex-vessel fission products. The characteristics of this release
category can be found in Table 6-6.
TMI-1 Applicability:
The dominant interfacing-systems LOCA for TMI-1 is smaller than the one modeled
above. Actual release fractions from the dominant TMI-1 sequence would be expected
to be somewhat lower than those provided. The TMI-1 MAAP run (TMI19F) performed
assumed a 2.5 inch diameter interfacing-systems LOCA similar to the dominant Oconee
sequence.
Release Category 2.02
Description:
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This release category is characterized as a bypass of containment with releases going
into the Auxiliary Building. Ex-vessel fission products are not released and scrubbing is
not available.
Representative Sequence:
The representative sequence is a medium sized interfacing-systems LOCA to the
Auxiliary Building. An example of this would be a failure of the RCS letdown line outside
containment. This provides a direct pathway from the hot leg to the Auxiliary Building.
A loss of all feedwater, all HPI and LPI injection, and all containment safeguards is
assumed. No credit is taken for fission product plateout.
A medium sized LOCA is created which will lower the RCS below 400 psia prior to core
melt. With the dispersal of the corium over a large area, the debris bed cools
preventing release of ex-vessel fission products. The characteristics of this release
category can be found in Table 6-7.
TMI-1 Applicability:
The dominant interfacing-systems LOCA for TMI-1 is smaller than the one modeled
above. Actual release fractions from the dominant TMI-1 sequence would be expected
to be somewhat lower than those provided. The TMI-1 MAAP run (TMI-19F) performed
assumed a 2.5 inch diameter interfacing-systems LOCA similar to the dominant Oconee
sequence.
Release Category 2.03
Description:
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This release category is characterized as a bypass of containment with releases going
into the Auxiliary Building. Ex-vessel fission products are released, but scrubbing is
available.
Representative Sequence:
The representative sequence is a medium sized interfacing-systems LOCA to the
Auxiliary Building. An example of this would be a failure of the RCS letdown line outside
containment. This provides a direct pathway from the hot leg to the Auxiliary Building.
A loss of all feedwater, all HPI and LPI injection, and all containment safeguards is
assumed. Since the release is low in the Auxiliary Building credit is taken for fission
product plateout.
A medium sized LOCA is created which will lower the RCS below 400 psia prior to core
melt. Low pressure at reactor vessel failure will cause the corium to be retained in the
cavity area. The debris bed will not be coolable and ex-vessel fission products will be
released. The characteristics of this release category can be found in Table 6-8.
TMI-1 Applicability:
The dominant interfacing-systems LOCA for TMI-1 is smaller than the one modeled
above. Actual release fractions from the dominant TMI-1 sequence would be expected
to be somewhat lower than those provided. The TMI-1 MAAP run (TMI-19F) performed
assumed an 2.5 inch diameter interfacing-systems LOCA similar to the dominant
Oconee sequence.
Release Category 2.04
Description:
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This release category is characterized as a bypass of containment with releases going
into the Auxiliary Building. Ex-vessel fission products are released and scrubbing is not
available.
Representative Sequence:
The representative sequence is a medium sized interfacing-systems LOCA to the
Auxiliary Building. An example of this would be a failure of the RCS letdown line outside
containment. This provides a direct pathway from the hot leg to the Auxiliary Building.
A loss of all feedwater, all HPI and LPI injection, and all containment safeguards is
assumed. No credit is taken for fission product plateout.
A medium sized LOCA is created which will lower the RCS below 400 psia prior to core
melt. Low pressure at reactor vessel failure will cause the corium to be retained in the
cavity area. The debris bed will not be coolable and ex-vessel fission products will be
released. The characteristics of this release category can be found in Table 6-9.
TMI-1 Applicability:
The dominant interfacing-systems LOCA for TMI-1 is smaller than the one modeled
above. Actual release fractions from the dominant TMI-1 sequence would be expected
to be somewhat lower than those provided. The TMI-1 MAAP run (TMI19F) performed
assumed an 2.5 inch diameter interfacing-systems LOCA similar to the dominant
Oconee sequence.
Release Category 3.01
Description:
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This release category is characterized as a large isolation failure with releases to the
Auxiliary Building. No ex-vessel release occurs, and fission product scrubbing is
available via significant plateout in the Auxiliary Building.
Representative Sequence:
The representative sequence is taken directly from Oconee for Oconee external events
dominate this category. The Oconee sequence is initiated by an event which fails two
eight inch diameter lines (with direct air-to-air communication paths). This results in a
fission product release to the lower levels of the Auxiliary Building. The event also
causes a loss of all feedwater, all HPI and LPI injection, and all containment
safeguards. RCP seal LOCAs exist on all four pumps fifteen minutes after accident
initiation.
The seal LOCAs will not depressurize the RCS and the core melt will occur at high
pressure. This will lead to an energetic blowdown at reactor vessel failure and dispersal
of the corium into the lower containment area. With the dispersal of the corium over a
large area, the debris bed will cool preventing release of ex-vessel fission products.
The characteristics of this release category are listed in Table 6-10.
TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 3.02
Description:
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This release category is characterized as a large isolation failure with releases to the
Auxiliary Building. No ex-vessel release occurs and fission product scrubbing is not
available( i.e., very little fission product plateout prior to release to the environment).
Representative Sequence:
This sequence is similar to Release Category 3.01, and is initiated by the event which
fails two eight inch diameter lines. This results in a fission product release to the
Auxiliary Building but no credit for fission product plateout is taken. Since MAAP does
not model the Auxiliary Building, no credit for plateout in the Auxiliary Building is taken.
The event also causes a loss of all feedwater, all HPI and LPI injection, and all
containment safeguards. Failure of seal injection will cause RCP seal LOCAs on all 4
pumps 15 minutes after accident initiation.
The seal LOCAs will not depressurize the RCS and the core melt will occur at high
pressure. This will lead to an energetic blowdown at reactor vessel failure and dispersal
of the corium into the lower containment area. With the dispersal of corium over a large
area, the debris bed will cool, preventing release of ex-vessel fission products. The
characteristics of this release category are listed in Table 6-11.
TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 3.03
Description:
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This release category is characterized as a large isolation failure with releases to the
Auxiliary Building. Ex-vessel fission products are released and fission product
scrubbing is available via significant plateout in the Auxiliary Building.
Representative Sequence:
This sequence is similar to Release Category 3.01 and is initiated by the event which
fails two 8 inch diameter lines. This results in a fission product release to the lower
levels of the Auxiliary Building. The event also causes a loss of all feedwater, all HPI
and LPI injection, and all containment safeguards. Failure of seal injection will cause
RCP seal LOCAs on all four pumps 15 minutes after accident initiation.
Even though the seal LOCAs will not depressurize the RCS, failure of the pressurizer
surge line occurs due to overheating. This induced LOCA will cause depressurization of
the RCS and, therefore, should be classified as a low pressure core melt. Low pressure
at reactor vessel failure will cause the corium to be retained in the cavity area. The
RCS and accumulator inventories are not sufficient to achieve long-term coolability of
the debris bed. Therefore, core-concrete interaction will take place and ex-vessel
fission products will be released. The characteristics of this release category are listed
in Table 6-12.
TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 3.04
Description:
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This release category is characterized as a large isolation failure with releases to the
Auxiliary Building. Ex-vessel fission products are released and fission product
scrubbing is not available, e.g., very little fission product plateout prior to release to the
environment.
Representative Sequence:
This sequence is similar to Release Category 3.01 and is initiated by the event which
fails two 8 inch diameter lines. This results in a fission product release to the Auxiliary
Building but no credit for fission product plateout is taken. The event also causes a loss
of all feedwater, all HPI and LPI injection, and all containment safeguards. Failure of
seal injection will cause RCP seal LOCAs on all four pumps 15 minutes after accident
initiation.
Even though the seal LOCAs will not depressurize the RCS, failure of the pressurizer
surge line occurs due to overheating. This induced LOCA will allow depressurization of
the RCS and, therefore, should be classified as a low pressure core melt. Low pressure
at reactor vessel failure will cause the corium to be retained in the cavity area. The
RCS and accumulator inventories are not sufficient to achieve long-term coolability of
the debris bed. Therefore, core-concrete interaction will take place and ex-vessel
fission products will be released. The characteristics of this category are listed in Table
6-13.
TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 3.05
Definition:
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This release category is characterized as a large isolation failure with releases outside
the Auxiliary Building. No ex-vessel release occurs and fission product scrubbing is not
available, i.e., very little fission product plateout prior to release to the environment.
Representative Sequence:
The release parameters for this release category are taken directly from Release
Category 3.02. The representative sequence is similar except the isolation failure is
outside the Auxiliary Building. The characteristics of this category are listed in Table 6-
14.
TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 3.06
Definition:
This release category is characterized as a large isolation failure with releases outside
the Auxiliary Building. Ex-vessel fission products are released and fission product
scrubbing is not available, e.g., very little fission product plateout prior to release to the
environment.
Representative Sequence:
The release parameters for this release category are taken directly from Release
Category 3.04. The representative sequence is similar except the isolation failure is
outside the Auxiliary Building. The characteristics of this category are listed in Table 6-
15.
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TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 4.01
Definition:
This release category is characterized by a small isolation failure with releases to the
Auxiliary Building. No ex-vessel release occurs, and fission product scrubbing is
available via significant plateout in the Auxiliary Building.
Representative Sequence:
The representative sequence is a station blackout. The station blackout causes a loss
of all HPI and LPI injection and a loss of all containment safeguards. SSHR is available
via the turbine-driven Emergency Feedwater (EFW) pump. Cooling to the RCP seals
fails, resulting in seal LOCAs on all four pumps 15 minutes after accident initiation. A
small isolation failure provides a fission product release path to the lower Auxiliary
Building, where significant plateout occurs.
The RCP seal LOCAs will not depressurize the RCS, causing the core melt to occur at
high pressure. This will lead to an energetic blowdown at reactor vessel failure and
dispersal of the corium into the lower containment area. With the dispersal of the
corium over a large area, the debris bed will be coolable preventing the release of ex-
vessel fission products. The characteristics of this release category are listed in Table
6-16.
TMI-1 Applicability:
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Both the medium and small LOCA sequences with small containment isolation failures
dominate this release category for TMI-1. Release parameters and release fractions
are similar for these sequences. A TMI-1 MAAP run (TMI7L) showed similar results to
the representative sequence selected from the Oconee PRA.
Release Category 4.02
Description:
This release category is characterized by a small isolation failure with releases to the
Auxiliary Building. No ex-vessel release occurs and fission product scrubbing is not
available, e.g., very little fission product plateout prior to release to the environment.
Representative Sequence:
The representative sequence is a station blackout. The station blackout causes a loss
of all HPI and LPI injection and a loss of all containment safeguards. SSHR is available
via the turbine-driven EFW pump. Cooling to the RCP seals fails, resulting in seal
LOCAs on all four pumps 15 minutes after accident initiation. A small isolation failure
provides a fission product release path to the Auxiliary Building. No scrubbing is
credited in the Auxiliary Building.
The RCP seal LOCAs will not depressurize the RCS, causing the core melt to occur at
high pressure. This will lead to an energetic blowdown at reactor vessel failure and
dispersal of the corium into the lower containment area. With the dispersal of the
corium over a large area, the debris bed will be coolable preventing the release of ex-
vessel fission products. The characteristics of this release category are listed in Table
6-17.
TMI-1 Applicability:
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Both the medium and small LOCA sequences with small containment isolation failures
dominate this release category for TMI-1. Release parameters and release fractions
are similar for these sequences. A TMI-1 MAAP run (TMI7L) showed similar results to
the representative sequence selected from the Oconee PRA.
Release Category 4.03
Description:
This release category is characterized by a small isolation failure with release to the
Auxiliary Building. Ex-vessel fission products are released and fission product
scrubbing is available via significant plateout in the Auxiliary Building.
Representative Sequence:
This sequence is initiated by a station blackout. The station blackout causes a loss of
all HPI and LPI injection and a loss of all containment safeguards. SSHR is available
via the turbine driven EFW pump. Cooling to the RCP seals fails, resulting in seal
LOCAs on all four pumps 15 minutes after accident initiation. The small isolation failure
provides a fission product release path to the lower Auxiliary Building, where significant
plateout occurs.
Even though the seal LOCAs will not depressurize the RCS, failure of the pressurizer
surge line occurs due to overheating. This induced LOCA will allow depressurization of
the RCS and, therefore, should be classified as a low pressure core melt. Low pressure
at reactor vessel failure will cause the corium to be retained in the cavity area. The
RCS and accumulator inventories are not sufficient to achieve long-term coolability of
the debris bed. Therefore, core-concrete interaction will take place and ex-vessel
fission products will be released. The characteristics of this release category are listed
in Table 6-18.
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TMI-1 Applicability:
Both the medium and small LOCA sequences with small containment isolation failures
dominate this release category for TMI-1. Release parameters and release fractions
are similar for these sequences. A TMI-1 MAAP run (TMI7L) showed similar results to
the representative sequence selected from the Oconee PRA.
Release Category 4.04
Description:
This release category is characterized by a small isolation failure with releases to the
Auxiliary Building. Ex-vessel fission products are released and fission product
scrubbing is not available, i.e., very little fission product plateout prior to release to the
environment.
Representative Sequence:
This sequence is initiated by a station blackout. The station blackout causes a loss of
all HPI and LPI injection and a loss of all containment safeguards. SSHR is available
via the turbine-driven EFW pump. Cooling to the RCP seals fails, resulting in seal
LOCAs on all four pumps 15 minutes after accident initiation. The small isolation failure
provides a fission product release path to the Auxiliary Building. However, no credit is
taken for fission product plateout.
Even though the seal LOCAs will not depressurize the RCS, failure of the pressurizer
surge line occurs due to overheating. This induced LOCA will allow depressurization of
the RCS and, therefore, should be classified as a low pressure core melt. Low pressure
at reactor vessel failure will cause the corium to be retained in the cavity area. The
RCS and accumulator inventories are not sufficient to achieve long-term coolability of
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the debris bed. Therefore, core-concrete interaction will take place and ex-vessel
fission products will be released. The characteristics of this release category are listed
in Table 6-19.
TMI-1 Applicability:
Both the medium and small LOCA sequences with small containment isolation failures
dominate this release category for TMI-1. Release parameters and release fractions
are similar for these sequences. A TMI-1 MAAP run (TMI7L) showed similar results to
the representative sequence selected from the Oconee PRA.
Release Category 4.05
Description:
This release category is characterized by a small isolation failure with release directly to
the environment. No ex-vessel release occurs and fission product scrubbing is
available.
Representative Sequence:
This sequence is initiated by a station blackout. The station blackout causes a loss of
all HPI and LPI injection and a loss of all containment safeguards. SSHR is available
via the turbine-driven EFW pump. Cooling to the RCP seals fails, resulting in seal
LOCAs on all four pumps 15 minutes after accident initiation. A small isolation failure
provides a release path directly to the environment. Plateout within HVAC systems
allows significant fission product scrubbing.
The RCP seal LOCAs will not depressurize the RCS, causing the core melt to occur at
high pressure. This will lead to an energetic blowdown at reactor vessel failure and
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dispersal of the corium into the lower containment area. The availability of the RCS and
accumulator inventory after reactor vessel failure, the debris bed will be coolable,
preventing the release of ex-vessel fission products. The characteristics of this release
category are listed in Table 6-20.
TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 4.06
Description:
This release category is characterized by a small isolation failure with release directly to
the environment. No ex-vessel release occurs and fission product scrubbing is not
available.
Representative Sequence:
This sequence is initiated by a station blackout. The station blackout causes a loss of
all HPI and LPI injection and a loss of all containment safeguards. SSHR is available
via the turbine-driven EFW pump. Cooling to the RCP seals fails, resulting in seal
LOCAs on all four pumps 15 minutes after accident initiation. A small isolation failure
provides a release path directly to the environment. With containment safeguards
failed, fission products will not be scrubbed.
The RCP seal LOCAs will not depressurize the RCS, causing the core melt to occur at
high pressure. This will lead to an energetic blowdown at reactor vessel failure and
dispersal of the corium into the lower containment area. With the dispersal of the
corium over a large area, the debris bed will be coolable, preventing the release of ex-
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vessel fission products. The characteristics of this release category are listed in Table
6-21.
TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 4.07
Description:
This release category is characterized by a small isolation failure with releases directly
to the environment. Ex-vessel release of fission products occurs and fission product
scrubbing is available.
Representative Sequence:
This sequence is initiated by a station blackout. The station blackout causes a loss of
all HPI and LPI injection and a loss of all containment safeguards. SSHR is available
via the turbine-driven EFW pump. Cooling to the RCP seals fails, resulting in seal
LOCAs on all four pumps 15 minutes after accident initiation. The small isolation failure
provides a release path directly to the environment. Plateout within HVAC systems
allows significant fission product scrubbing.
Even though the seal LOCAs will not depressurize the RCS, failure of the pressurizer
surge line occurs due to overheating. The induced LOCA will allow depressurization of
the RCS and, therefore, should be classified as a low pressure core melt. Low pressure
at reactor vessel failure will cause the corium to be retained in the cavity area. The
RCS and accumulator inventories are not sufficient to achieve long-term coolability of
the debris bed. Therefore, core-concrete interaction will take place and ex-vessel
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fission products will be released. The characteristics of this release category are listed
in Table 6-22.
TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 4.08
Description:
This release category is characterized by a small isolation failure with releases directly
to the environment. Ex-vessel release of fission products occurs and fission product
scrubbing is not available.
Representative Sequence:
This sequence is initiated by a station blackout. The station blackout causes a loss of
all HPI and LPI injection and a loss of all containment safeguards. SSHR is available
via the turbine-driven EFW pump. Cooling to the RCP seals fails, resulting in seal
LOCAs on all four pumps 15 minutes after accident initiation. The small isolation failure
provides a release path directly to the environment. With failed containment
safeguards, fission products will not be scrubbed.
Even though the seal LOCAs will not depressurize the RCS, failure of the pressurizer
surge line occurs due to overheating. The induced LOCA will allow depressurization of
the RCS and, therefore, should be classified as a low pressure core melt. Low pressure
at reactor vessel failure will cause the corium to be retained in the cavity area. The
RCS and accumulator inventories are not sufficient to achieve long-term coolability of
the debris bed. Therefore, core-concrete interaction will take place and ex-vessel
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fission products will be released. The characteristics of this release category are listed
in Table 6-23.
TMI-1 Applicability:
There are currently no TMI-1 sequences involving this release category.
Release Category 5.01
Description:
This release category is characterized by an early overpressurization of containment
with no ex-vessel release of fission products. Early containment failures are
catastrophic with no fission product scrubbing.
Representative Sequence:
This release is represented by a small LOCA with an injection failure, with a failure of all
containment safeguard systems and SSHR. This will ultimately lead to a core melt at
high pressures since the RCS cannot be depressurized.
With high pressure in the RCS at reactor vessel failure, the resulting blowdown will
disperse the corium into the lower containment area. With the debris bed spread over a
large area, the debris bed will be coolable, preventing the ex-vessel release of fission
products. The early containment failure does not allow sufficient time to effectively
scrub fission products. The characteristics of this release category are listed in Table 6-
24.
Analysis of the CET quantification has shown that the dominant containment failure
mode for this release category is a hydrogen burn.
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TMI-1 Applicability:
This release category is dominated by plant damage states with containment
safeguards systems available. The dominant TMI-1 sequences contain low probability
containment failures caused by hydrogen burns at low containment base pressure.
Other plant damage states without containment safeguards systems are not as
important because sparks are not available for random hydrogen ignition. The high
pressure containment failures associated with all the dominant TMI-1 sequences should
all result in similar release fractions. A TMI-1 MAAP run (TMI7F3) showed similar
results to the representative sequence selected from the McGuire PRA. All other
dominant sequences should have similar results.
Release Category 5.02
Description:
This release category is characterized by an early overpressurization of containment
with an ex-vessel release of fission products. Early containment failures are
catastrophic with no fission product scrubbing.
Representative Sequence:
This release is represented by a small LOCA with an injection failure, with a failure of all
containment safeguard systems and SSHR. This will ultimately lead to a core melt at
high pressures since the RCS cannot be depressurized.
Even though these failures lead to high pressure at the start of core melt, the circulation
of hot gases will begin heating the RCS piping and subsequently fail the pressurizer
surge line.
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This induced LOCA will depressurize the RCS, producing low pressure at reactor vessel
failure. The low pressure blowdown will cause the corium to be retained in the cavity
area. Even though water inventory is available in the lower containment from the RCS
and accumulator inventories, the geometry of the cavity will prevent this water from
cooling the debris. Therefore, core-concrete interaction will occur and ex-vessel fission
products will be released. The early containment failure does not allow sufficient time to
effectively scrub fission products. The characteristics of this release category are listed
in Table 6-25.
Analysis of the CET quantification has shown that the dominant containment failure
mode for this release category is a hydrogen burn.
TMI-1 Applicability:
This release category is dominated by plant damage states with containment
safeguards systems available. The dominant TMI-1 sequences contain low probability
containment failures caused by hydrogen burns at low containment base pressure.
Other plant damage states without containment safeguards systems are not as
important because sparks are not available for random hydrogen ignition. The high
pressure containment failures associated with all the dominant TMI-1 sequences should
all result in similar release fractions. A TMI-1 MAAP run (TMI7F3) showed similar
results to the representative sequence selected from the McGuire PRA. All other
dominant sequences should have similar results.
Release Category 6.01
Description:
This release category is characterized as an overpressurization of containment which
leads to a catastrophic failure of containment late in the accident sequence. There is no
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ex-vessel release of fission products, no revaporization of fission products, but fission
product scrubbing does take place.
Representative Sequence:
The representative sequence for this release category is a small or medium sized LOCA
with a recirculation failure. Containment sprays continue to operate or are recovered
late, but no containment cooling takes place.
With all the injection water available in the containment, core concrete interaction is
prevented and ex-vessel fission products are not released. The operation of the
containment sprays continues to scrub fission products.
The availability of SSHR prevents any late revaporiza-tion from taking place.
Revaporization is likely to occur if the pressure is low and the temperature is high. The
steam generator provides a heat sink; therefore, the temperature in the steam generator
relative to the rest of the RCS is lower, thus suppressing revaporization. Revaporization
is discussed in Section B.4.8.
The debris bed will boil enough water to overpressurize the containment. Without
containment cooling, this will lead to a catastrophic failure of containment. The
characteristics of this release category are listed in Table 6-26.
TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F) showed similar results to the representative sequence selected
from the Oconee PRA. All other dominant TMI-1 sequences should have similar
results.
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Release Category 6.02
Description:
This release category is characterized as an overpressurization of containment which
leads to a catastrophic failure of containment late in the accident sequence. There is no
ex-vessel release of fission products, no revaporization of fission products and no
fission product scrubbing.
Representative Sequence:
The representative sequence for this release category is a small or medium sized LOCA
with a recirculation failure and a failure of all containment safeguards.
With all the injection water available in the containment, core concrete interaction is
prevented and ex-vessel fission products are not released. Failure of the containment
sprays prevents any scrubbing from taking place.
The availability of SSHR prevents any late revaporization from taking place.
Revaporization is likely to occur if the pressure is low and the temperature is high. The
steam generator provides a heat sink; therefore, the temperature in the steam generator
relative to the rest of the RCS is lower, thus suppressing revaporization. Revaporization
is discussed in Section B.4.8.
The debris bed will boil enough water to overpressurize the containment. Without
containment cooling, this will lead to a catastrophic failure of containment. The
characteristics of this release category are listed in Table 6-27.
TMI-1 Applicability:
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TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F) showed similar results to the representative sequence selected
from the Oconee PRA. All other dominant TMI-1 sequences should have similar
results.
Release Category 6.03
Description:
This release category is characterized as an overpressurization of containment which
leads to a catastrophic failure of containment late in the accident sequence. There is no
ex-vessel release of fission products, and fission product scrubbing is available. There
is a revaporization release of fission products for this release category.
Representative Sequence:
The representative sequence for this release category is a small LOCA with a
recirculation failure and a loss of SSHR. Containment sprays continue to operate or are
recovered late, but no containment cooling takes place.
With all the injection water available in the containment, core concrete interaction is
prevented and ex-vessel fission products are not released. The operation of the
containment sprays continues to scrub fission products.
The small LOCA allows a great deal of fission product retention within the primary
system. The loss of SSHR allows the revaporization to take place.
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The debris bed will boil enough water to overpressurize the containment. Without
containment cooling, this will lead to a catastrophic failure of containment. The
characteristics of this release category are listed in Table 6-28.
TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F) showed similar results to the representative sequence selected
from the Oconee PRA. All other dominant TMI-1 sequences should have similar
results.
Release Category 6.04
Description:
This release category is characterized as an overpressurization of containment which
leads to a catastrophic failure of containment late in the accident sequence. There is no
ex-vessel release of fission products, and fission product scrubbing is not available.
There is a revaporization release of fission products for this release category.
Representative Sequence:
The representative sequence for this release category is a small LOCA with a
recirculation failure and a loss of SSHR. All containment safeguards are also failed.
With all the injection water available in the containment, core concrete interaction is
prevented and ex-vessel fission products are not released. The operation of the
containment sprays continues to scrub fission products.
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The small LOCA allows a great deal of fission product retention within the primary
system. The loss of SSHR allows the revaporization to take place.
The debris bed will boil enough water to overpressurize the containment. Without
containment cooling, this will lead to a catastrophic failure of containment. The
characteristics of this release category are listed in Table 6-29.
TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F) showed similar results to the representative sequence selected
from the Oconee PRA. All other dominant TMI-1 sequences should have similar
results.
Release Category 6.05
Description:
This release category is characterized as an overpressurization of containment which
leads to a catastrophic failure of containment late in the accident sequence. There is
fission product scrubbing, but no revaporization releases. However, ex-vessel fission
products will be released.
Representative Sequence:
The representative sequence for this release category is a small or medium sized LOCA
with an injection failure. Containment sprays continue to operate or are recovered late,
but no containment cooling takes place.
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The cavity geometry prevents sufficient cooling of the debris bed and allows ex-vessel
fission products to be released.
The availability of SSHR prevents any late revaporization from taking place.
Revaporization is likely to occur if the pressure is low and the temperature is high. The
steam generator provides a heat sink; therefore, the temperature in the steam generator
relative to the rest of the RCS is lower, thus suppressing revaporization. Revaporization
is discussed in Section B.4.8.
The debris bed will boil enough water to overpressurize the containment. Without
containment cooling, this will lead to a catastrophic failure of containment. The
characteristics of this release category are listed in Table 6-30.
TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F) showed similar results to the representative sequence selected
from the Oconee PRA. All other dominant TMI-1 sequences should have similar
results.
Release Category 6.06
Description:
This release category is characterized as an overpressurization of containment which
leads to a catastrophic failure of containment late in the accident sequence. There is no
fission product scrubbing and no revaporization releases. However, ex-vessel fission
products will be released.
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Representative Sequence:
The representative sequence for this release category is a small or medium sized LOCA
with an injection failure. All containment safeguards systems are also failed.
The cavity geometry prevents sufficient cooling of the debris bed and allows ex-vessel
fission products to be released.
The availability of SSHR prevents any late revaporization from taking place.
Revaporization is likely to occur if the pressure is low and the temperature is high. The
steam generator provides a heat sink; therefore, the temperature in the steam generator
relative to the rest of the RCS is lower, thus suppressing revaporization. Revaporization
is discussed in Section B.4.8.
The debris bed will boil enough water to overpressurize the containment. Without
containment safeguards, this will lead to a catastrophic failure of containment. The
characteristics of this release category are listed in Table 6-31.
TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F) showed similar results to the representative sequence selected
from the Oconee PRA. All other dominant TMI-1 sequences should have similar
results.
Release Category 6.07
Description:
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This release category is characterized as an overpressurization of containment which
leads to a catastrophic failure of containment late in the accident sequence. There is
fission product scrubbing, and ex-vessel fission products will be released. A
revaporization release is also assumed for this release category.
Representative Sequence:
This release category is represented by a loss of all core cooling, all SSHR, and all
containment cooling. Reactor coolant pump seal cooling is maintained, so no seal
LOCA occurs. Containment sprays are available for scrubbing but no cooling is
provided.
The loss of SSHR will cause the steam generators to boil dry and subsequently cause
the RCS pressure to increase to the pressurizer relief valve setpoint. RCS inventory will
be depleted through the cycling relief valve until the core uncovers.
The cycling relief valve allows a great deal of fission product retention within the primary
system. The loss of SSHR allows the revaporization to take place.
After reactor vessel failure, the accumulator water will discharge onto the debris bed in
the cavity. However, this is not enough inventory to cool the corium. Therefore, the
steaming of the water will start to pressurize containment. This sequence of events will
ultimately lead to core-concrete interaction and subsequent release of ex-vessel fission
products while the containment is being pressurized. The containment will ultimately fail
catastrophically due to overpressurization. The characteristics of this release category
are listed in Table 6-32.
TMI-1 Applicability:
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TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F) showed similar results to the representative sequence selected
from the Oconee PRA. All other dominant TMI-1 sequences should have similar
results.
Release Category 6.08
Description:
This release category is characterized as an overpressurization of containment which
leads to a catastrophic failure of containment late in the accident sequence. There is no
fission product scrubbing but ex-vessel fission products will be released. A
revaporization release is also assumed for this release category.
Representative Sequence:
This release category is represented by a loss of all core cooling, all SSHR, and all
containment safeguards. Reactor coolant pump seal cooling is maintained, so no seal
LOCA occurs.
The loss of SSHR will cause the steam generators to boil dry and subsequently cause
the RCS pressure to increase to the pressurizer relief valve setpoint. RCS inventory will
be depleted through the cycling relief valve until the core uncovers.
After reactor vessel failure, the accumulator water will discharge onto the debris bed in
the cavity. However, this is not enough inventory to cool the corium. Therefore, the
steaming of the water will start to pressurize containment. This sequence of events will
ultimately lead to core-concrete interaction and subsequent release of ex-vessel fission
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products while the containment is being pressurized. The containment will ultimately fail
catastrophically due to overpressurization.
The cycling relief valve allows a great deal of fission product retention within the primary
system. The loss of SSHR allows the revaporization to take place. The characteristics
of this release category are listed in Table 6-33.
TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F) showed similar results to the representative sequence selected
from the Oconee PRA. All other dominant TMI-1 sequences should have similar
results.
Release Category 7.01
Description:
This release category is characterized as an overpressurization of containment which
leads to a benign containment failure late in the accident sequence. There is no release
of ex-vessel fission products, and fission product scrubbing is available. A benign
failure is defined as a failure of the containment structure which does not lead to a rapid
blowdown of the containment atmosphere. Instead, the containment structure relieves
pressure enough such that a continued pressure rise does not occur.
Representative Sequence:
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The representative sequence is identical to that of Release Category 6.01. The benign
containment failure significantly reduces the energy of the release. The characteristics
of this release category are listed in Table 6-34.
TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F), showed similar results to there respective representative
sequences with catastrophic failures of containment, selected from the Oconee PRA.
On the basis of these comparisons all dominant TMI-1 sequences should have similar
results for this release category.
Release Category 7.02
Description:
This release category is characterized as an overpressurization of containment which
leads to a benign containment failure late in the accident sequence. There is no release
of ex-vessel fission products and no fission product scrubbing. A benign failure is
defined as a failure of the containment structure which does not lead to a rapid
blowdown of the containment atmosphere. Instead, the containment structure relieves
pressure enough such that a continued pressure rise does not occur.
Representative Sequence:
The representative sequence is identical to that of Release Category 6.02. The benign
containment failure significantly reduces the energy of the release. The characteristics
of this release category are listed in Table 6-35.
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TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F), showed similar results to there respective representative
sequences with catastrophic failures of containment, selected from the Oconee PRA.
On the basis of these comparisons all dominant TMI-1 sequences should have similar
results for this release category.
Release Category 7.03
Description:
This release category is characterized as an overpressurization of containment which
leads to a benign containment failure late in the accident sequence. Ex-vessel fission
products will be released, but fission product scrubbing is available. A benign failure is
defined as a failure of the containment structure which does not lead to a rapid
blowdown of the containment atmosphere. Instead, the containment structure relieves
pressure enough such that a continued pressure rise does not occur.
Representative Sequence:
The representative sequence is identical to that of Release Category 6.05. The benign
containment failure significantly reduces the energy of the release. The characteristics
of this release category are listed in Table 6-36.
TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
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TMI7F2 and TMI4F), showed similar results to there respective representative
sequences with catastrophic failures of containment, selected from the Oconee PRA.
On the basis of these comparisons all dominant TMI-1 sequences should have similar
results for this release category.
Release Category 7.04
Description:
This release category is characterized as an overpressurization of containment which
leads to a benign containment failure late in the accident sequence. Ex-vessel fission
products will be released, and fission product scrubbing is not available. A benign
failure is defined as a failure of the containment structure which does not lead to a rapid
blowdown of the containment atmosphere. Instead, the containment structure relieves
pressure enough such that a continued pressure rise does not occur.
Representative Sequence:
The representative sequence is identical to that of Release Category 6.06. The benign
containment failure significantly reduces the energy of the release. The characteristics
of this release category are listed in Table 6-37.
TMI-1 Applicability:
TMI-1 late containment failure sequences are dominated by overpressurizations from
the generation of steam and non-condensable gases. TMI-1 MAAP runs (TMI7F,
TMI7F2 and TMI4F), showed similar results to there respective representative
sequences with catastrophic failures of containment, selected from the Oconee PRA.
On the basis of these comparisons all dominant TMI-1 sequences should have similar
results for this release category.
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Release Category 8.01
Description:
This release category is characterized as a late containment failure due to basemat
melt-through. Ex-vessel release of fission products will occur, and fission product
scrubbing is not available.
Representative Sequence:
This release is represented by a small LOCA with an injection failure, with a failure of all
containment safeguard systems and SSHR. This will ultimately lead to a core melt at
high pressures since the RCS cannot be depressurized.
Even though the LOCA will not depressurize the RCS, failure of the pressurizer surge
line occurs due to overheating. This induced LOCA will depressurize the RCS and,
therefore, should be classified as a low pressure core melt. Low pressure at reactor
vessel failure will cause the corium to be retained in the cavity area. With all
containment water sources failed, the RCS and accumulator inventories are not
sufficient to achieve long-term coolability of the debris bed. Therefore, core-concrete
interaction will take place and ex-vessel fission products will be released.
The cavity geometry will prevent the water available in lower containment from reaching
the corium pool. This will have two effects. First, it will prevent long-term cooling of the
debris bed. Second, it will also prevent significant steam production from boiling of the
water. Given this situation, the corium will continue to attack the concrete basemat and
will fail the basemat prior to overpressurization of the containment.
The releases are assumed to be the same as the no containment failure categories
except that the noble gas release fraction will be increased to 1.O, and iodine was
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increased to 0.03 based on NUREG/CR-4551 (Reference 29). Release timing is also
taken from NUREG/CR-4551. The characteristics of this release category are listed in
Table 6-38.
TMI-1 Applicability:
TMI-1 MAAP runs (TMI7F, TMI7F2 and TMI4F), showed similar results to there
respective representative sequences with catastrophic failures of containment, selected
from the Oconee PRA. On the basis of these comparisons all dominant TMI-1
sequences should have similar results for this release category.
Release Category 9.01
Description:
This release category is characterized as no containment failure without ex-vessel
fission products, and fission product scrubbing is available. The release mechanism
from containment is the same as normal leakage.
Representative Sequence:
The representative sequence is taken from the Oconee PRA. This Oconee sequence is
initiated by an ATWS which causes an overpressurization of the RCS, creating a large
LOCA. Based on analysis of allowable stresses, the overpressurization is assumed to
catastrophically fail the RCS. The ATWS event is also assumed to fail the injection
lines into the RCS. Containment safeguards are available.
The large LOCA will cause depressurization of the RCS, leading to a low pressure core
melt. Low pressure at reactor vessel failure will cause the corium to remain within the
cavity area. With containment sprays available, water will be injected into the cavity
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from the refueling pool throughout the accident. This will allow cooling of the debris bed
within the cavity. With the containment fan coolers available, steam produced from the
water-corium interaction will be condensed, preventing containment failure. The
containment sprays will also aid in fission product scrubbing prior to release to the
environment. The characteristics of this release category are listed in Table 6-39.
TMI-1 Applicability:
There may be a wide range of sequences that can lead to core melt with no
containment failure. However, the release category definition is believed to be a
reasonable modeling of all these sequences.
Release Category 9.02
Description:
This release category is characterized as no containment failure without an ex-vessel
fission product release and without fission product scrubbing. The release mechanism
from containment is the same as normal leakage.
Representative Sequence:
This release category is represented by a small LOCA injection failure without
containment safeguards. The small LOCA will not depressurize the RCS, and other
depressurization efforts fail causing the core melt to occur at high pressure. This will
lead to an energetic blowdown at reactor vessel failure and dispersal of the corium into
the lower containment area. With the dispersal of the corium over a large area, the
debris bed will be coolable, preventing the release of ex-vessel fission products.
Recovery of containment fan coolers late in the sequence prevents overpressurization
of the containment. The characteristics of this release category are listed in Table 6-40.
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TMI-1 Applicability:
There may be a wide range of sequences that can lead to core melt with no
containment failure. However, the release category definition is believed to be a
reasonable modeling of all these sequences.
Release Category 9.03
Description:
This release category is characterized as no containment failure with an ex-vessel
fission product release with fission product scrubbing available. The release
mechanism from containment is the same as normal leakage.
Representative Sequence:
This release category is represented by a small LOCA injection failure without
containment safeguards. SSHR is successfully used to depressurize the primary
system. Low pressure at reactor vessel failure will cause the corium to be retained in
the cavity area. The RCS and accumulator inventories are not sufficient to achieve
long-term coolability of the debris bed. Therefore, core-concrete interaction will take
place and ex-vessel fission products will be released. Recovery of containment sprays
late in the accident sequence will prevent overpressurization of the containment and
allows fission product scrubbing to take place. The characteristics of this release
category are listed in Table 6-41.
TMI-1 Applicability:
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There may be a wide range of sequences that can lead to core melt with no
containment failure. However, the release category definition is believed to be a
reasonable modeling of all these sequences.
Release Category 9.04
Description:
This release category is characterized as no containment failure with an ex-vessel
fission product release and no fission product scrubbing. The release mechanism from
containment is the same as normal leakage.
Representative Sequence:
This release category is represented by a small LOCA injection failure without
containment safeguards. SSHR is successfully used to depressurize the primary
system. Low pressure at reactor vessel failure will cause the corium to be retained in
the cavity area. The RCS and accumulator inventories are not sufficient to achieve
long-term coolability of the debris bed. Therefore, core-concrete interaction will take
place and ex-vessel fission products will be released. Recovery of containment fan
coolers late in the accident sequence will prevent overpressurization of the containment.
The characteristics of this release category are listed in Table 6-42.
TMI-1 Applicability:
There may be a wide range of sequences that can lead to core melt with no
containment failure. However, the release category definition is believed to be a
reasonable modeling of all these sequences.
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6.5 FINAL BINNING OF RELEASE CATEGORIES
There are a total of 39 individual Release Categories (RC) described in Section 6.4. In
preparation for calculating the off-site consequences associated with these postulated
severe accidents, the 39 RCs are combined into 9 unique consequence bins. These 9
bins represent the unique fission product release timing and magnitude for all of the
core damage sequences. This section describes the 9 consequence bins.
The endpoint of the CET contains two major pieces of information, which are the
release frequency and the release category designation. The parameters that define a
release category and are important in the analysis of offsite consequences are:
1. Time of release
2. Duration of release
3. Energy of release
4. Warning time for evacuation
5. Isotopic fractions released to the environment
Each CET end point is capable of describing a unique sequence with potentially unique
release characteristics. For TMI-1, 39 release categories were identified in the CET
with most endpoints having a unique release category designation (see Figure 2-1). A
numbering scheme is used to separate major categories:
1 = Containment Bypass with Auxiliary Building Bypass
2 = Interfacing-Systems LOCA
3 = Large Isolation Failures
4 = Small Isolation Failures
5 = Early Containment Failure
6 = Late Containment Failure (Catastrophic)
7 = Late Containment Failure (Benign)
8 = Basemat Melt-Through
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9 = No Containment Failure
Different sequences within these major categories were given a designation such as
1.01, 1.02, 2.01, etc. in order to distinguish between specific details of the containment
response.
As part of the original IPE, the MAAP 3.0B thermal hydraulics code was used to analyze
the plant specific containment responses for each of the CET sequences. The 39 TMI-1
release categories were then reviewed in order to determine how they could be grouped
for the assignment of source terms. It is possible to develop source terms for every
release category in the CET, but in many cases, the results are so similar that
maintaining unique source terms for every release category does not provide any
measurable benefit. As a result, release categories with similar traits were grouped
together and a single source term was used to represent the entire group to streamline
the Level 3 analysis. For TMI-1, nine major source term groups identified above were
found to be an adequate structure for segregating the source terms. The table below
(Table 6-1) correlates the major source term groups to the source term designators and
provides basic descriptions of the representative sequence established for each source
term group:
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TABLE 6-1 REPRESENTATIVE SEQUENCE DESCRIPTIONS FOR SOURCE TERM GROUPS
RELEASE CATEGORY GROUP
SOURCE TERM
DESIGNATOR
GENERAL DESCRIPTION OF CONTRIBUTING SEQUENCES
1: Containment Bypass w/ Aux Bldg Bypass
SGTR This event is initiated with a double ended failure of a steam generator tube with the SG safety valve failed open. All injection is assumed unavailable. Emergency feedwater is available.
2: ISLOCA ISLOCA This event is initiated with a small break outside of containment followed by failure of injection. Emergency feedwater is available.
3: Large Isolation Failure
ISO-LG This scenario is represented by a loss of main feedwater followed by a failure of all injection. A large containment isolation failure is assumed to occur at time zero. Emergency feedwater operates successfully for a period of 6 hours. At 15 minutes into the event, 42 gpm seal leakage is assumed per loop. Core damage occurs at 9.4 hours into the event followed by failure of the hot leg due to creep rupture 36 minutes later. Vessel breach occurs at 16 hrs.
4: Small Isolation Failure
ISO-SM This scenario is represented by a loss of main feedwater followed by a failure of all injection. A small containment isolation failure is assumed to occur at time zero. Emergency feedwater is assumed unavailable. At 15 minutes into the event, 42 gpm seal leakage is assumed per loop. Core damage occurs at 50 minutes into the event followed by failure of the hot leg due to creep rupture 36 minutes later. Vessel breach occurs at 6 hrs.
5: Early Containment Failure
EARLY This scenario is represented by a Station Blackout. Emergency feedwater operates successfully for a period of 6 hours. At 15 minutes into the event, 42 gpm seal leakage is assumed per loop. Core damage occurs at 9 hours into the event. Vessel breach occurs at 11.7 hrs. It is assumed that containment failure occurs at the time of vessel breach.
6: Late Containment Failure (catastrophic)
LATE-LG This scenario is represented by a loss of main feedwater followed by a failure of all injection. Emergency feedwater operates successfully. At 15 minutes into the event, 42 gpm seal leakage is assumed per loop. Core damage occurs at 26 hours into the event followed by failure of the hot leg due to creep rupture 40 minutes later. Vessel breach occurs at 34.8 hrs. The containment fails due to overpressure at 70 hours into the event with an assumed large failure area, resulting a rapid depressurization of containment.
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TABLE 6-1 REPRESENTATIVE SEQUENCE DESCRIPTIONS FOR SOURCE TERM GROUPS
RELEASE CATEGORY GROUP
SOURCE TERM
DESIGNATOR
GENERAL DESCRIPTION OF CONTRIBUTING SEQUENCES
7: Late Containment Failure (benign)
LATE-SM This scenario is represented by a Station Blackout. Emergency feedwater operates successfully for a period of 6 hours. At 15 minutes into the event, 42 gpm seal leakage is assumed per loop. Core damage occurs at 9 hours into the event followed by failure of the hot leg due to creep rupture 50 minutes later. Vessel breach occurs at 16.5 hrs. Containment sprays are assumed to be recovered at 24 hours into the event. The core debris remains covered with water, however, without heat removal, the containment fails due to overpressure at 52 hours into the event. The breach area is assumed to be represented by a leak-before-break and results in a very slow containment depressurization.
8: Basemat Melt-Through
BMMT This scenario is represented by a loss of main feedwater followed by a failure of all injection. Emergency feedwater operates successfully. At 15 minutes into the event, 42 gpm seal leakage is assumed per loop. Core damage occurs at 26 hours into the event followed by failure of the hot leg due to creep rupture 40 minutes later. Vessel breach occurs at 34.7 hrs. All of the core debris is forced to remain in the reactor cavity in order to accelerate the amount of core concrete attack. When concrte erosion has exceeded 6 feet, containment failure is assumed to occur with a representative failure area equal to 1 ft2.
9: No Containment Failure
INTACT This scenario is represented by a loss of main feedwater followed by a failure of all injection. Emergency feedwater operates successfully. At 15 minutes into the event, 42 gpm seal leakage is assumed per loop. Core damage occurs at 26 hours into the event followed by failure of the hot leg due to creep rupture 40 minutes later. Vessel breach occurs at 34.6 hrs. Successful operation of containment sprays and fan coolers prevents containment overpressure failure long term.
Table 6-2 below provides additional accident progression information for the
representative sequences described above, including the time to core damage, time to
containment failure, and notable release fractions.
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In some cases, there were competing contributors to a release category group with
measurable differences in some of the release fractions (e.g., scrubbed vs unscrubbed
releases). The representative source term for the release category is typically chosen
based on the largest frequency, but when the consequences of a source term with a
smaller frequency are more severe, the more severe source term is used if it is believed
that the group would otherwise be underrepresented.
TABLE 6-2 SUMMARY OF REPRESENTATIVE MAAP SEQUENCES FOR TMI-1 SOURCE
TERMS MAAP CASE
NAME DESCRIPTION EFW SEAL LOCA?
SPRAY ON?
FANS ON?
TCUHRS
TCDHRS
HLCRHRS
TVF HRS
TCF HRS
TEND HRS
NG FRAC.
CSI FRAC.
TM0034 INTACT No cont failure, no ex-vessel rel., FP scrubbed
Y Y Y Y 18.8 26.0 26.7 34.6 NA 48 1.2E-01 4.6E-04
TM0035 BMMT Basemat melt w/o debris cooling
Y Y N N 18.7 26.0 26.6 34.7 64.4 48 9.7E-01 8.7E-03
TM0036 LATE - SM
Small late containment failure
6 hrs Y N N 8.2 9.0 9.9 16.5 52.1 72 7.0E-01 6.5E-03
TM0037 LATE-LRG
Large containment failure
Y Y N N 18.8 26.0 26.6 34.8 70.8 72 1.0E+00 6.9E-02
TM0038 EARLY Early containment failure at vessel breach
6 hrs Y N N 8.2 9.3 NA 11.7 11.7 48 1.0E+00 6.0E-02
TM0039 ISO-SM Containment isolation failure - small
N Y N N 0.6 0.8 1.4 6.0 0.0 48 8.3E-01 3.4E-02
TM0040 ISO-LRG Containment isolation failure - large
6 hrs Y N N 8.5 9.4 10.0 16.0 0.0 48 1.0E+00 2.3E-01
TM0041 ISLOCA .003 ft2 break N N N N 15.0 15.8 16.8 24.3 NA 72 9.2E-01 1.8E-01
TM0042 SGTR .0066 ft2 break N N N N 12.7 13.5 16.6 18.3 NA 48 1.0E+00 6.5E-01
The source terms that are used as input to the TMI-1 Level 3 model are a combination
of radionuclide release fractions, the timing of the radionuclide release relative to the
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declaration of a general emergency, and the frequencies at which the releases occur.
This combination of information is used in conjunction with other TMI-1 site
characteristics in the Level 3 model to evaluate the consequences of a core damage
event. Table 6-3 below provides a summary of the TMI-1 source term information,
which includes the following:
• MAAP case identifier (for reference),
• Airborne release for each of the fission product groups provided my MAAP,
• Start time of the airborne release (measured from the time of accident initiation),
• End time of the airborne release (measured from the time of accident initiation).
Three Mile Island PRA - L2 Containment Event Tree Analysis
_______________________________________________________________________________________________
6-52 0467060030-2788-050107
TABLE 6-3 TMI-1 SOURCE TERM SUMMARY
RELEASE CATEGORY INTACT BMMT LATE-SM LATE-LRG EARLY ISO-SM ISO-LRG ISLOCA SGTR
MAAP Case ID TM0034 TM0035 TM0036 TM0037 TM0038 TM0039 TM000040 TM0041 TM0042
Run Duration 48 hr 72 hr 72 hr 120 48 hr 48 hr 48 hr 72 hr 48 hr
Time after Scram when General Emergency is declared (3) 26 hr 26 hr 9 hr 26 hr 9.3 hr 0.8 hr 9.4 hr 15.8 hr 13.5 hr
Fission Product Group:
1) Noble
Total Plume 1 Release Fraction 1.25E-01 3.00E-01 7.00E-01 1.00E+00 1.00E+00 8.30E-01 1.00E+00 9.20E-01 1.00E+00
Start of Plume 1 Release (hr) 26.00 26.00 10.00 70.80 11.70 1.00 10.00 16.00 14.00
End of Plume 1 Release (hr) 48.00 64.00 72.00 70.80 11.70 48.00 20.00 20.00 16.00
Total Plume 2 Release Fraction2 1.00E+00
Start of Plume 2 Release (hr) 64.00
End of Plume 2 Release (hr) 64.00
2) CsI
Total Plume 1 Release Fraction 4.60E-04 8.70E-03 6.50E-03 7.00E-02 6.00E-02 3.40E-02 2.30E-01 1.80E-01 2.00E-02
Start of Plume 1 Release (hr) 26.00 26.00 10.00 70.80 11.70 1.00 10.00 16.00 14.00
End of Plume 1 Release (hr) 30.00 50.00 20.00 100.00 11.70 8.00 16.00 25.00 14.00
Total Plume 2 Release Fraction2 6.50E-01
Start of Plume 2 Release (hr) 34.00
End of Plume 2 Release (hr) 44.00
3) TeO2
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-53 0467060030-2788-050107
TABLE 6-3 TMI-1 SOURCE TERM SUMMARY
RELEASE CATEGORY INTACT BMMT LATE-SM LATE-LRG EARLY ISO-SM ISO-LRG ISLOCA SGTR
Total Plume 1 Release Fraction 4.60E-04 9.00E-03 9.00E-03 2.00E-02 3.80E-02 1.50E-02 2.00E-01 6.00E-02 1.00E-02
Start of Plume 1 Release (hr) 26.00 26.00 10.00 70.80 11.70 1.00 10.00 16.00 14.00
End of Plume 1 Release (hr) 30.00 50.00 20.00 100.00 11.70 8.00 16.00 20.00 14.00
Total Plume 2 Release Fraction2 4.00E-02
Start of Plume 2 Release (hr) 34.00
End of Plume 2 Release (hr) 44.00
4) SrO
Total Plume 1 Release Fraction 7.00E-05 8.50E-04 4.00E-04 5.00E-06 4.50E-03 1.50E-03 1.00E-02 6.00E-03 9.00E-04
Start of Plume 1 Release (hr) 29.00 30.00 10.00 70.80 11.70 1.00 12.00 16.00 14.00
End of Plume 1 Release (hr) 32.00 40.00 20.00 70.80 20.00 8.00 20.00 20.00 24.00
Total Plume 2 Release Fraction2
Start of Plume 2 Release (hr)
End of Plume 2 Release (hr)
5) MoO2
Total Plume 1 Release Fraction 3.50E-04 4.00E-03 2.80E-03 2.00E-05 2.00E-02 2.00E-02 3.50E-02 3.00E-02 6.00E-03
Start of Plume 1 Release (hr) 29.00 30.00 10.00 70.80 11.70 1.00 10.00 16.00 14.00
End of Plume 1 Release (hr) 32.00 40.00 20.00 70.80 11.70 8.00 16.00 20.00 14.00
Total Plume 2 Release Fraction2
Start of Plume 2 Release (hr)
End of Plume 2 Release (hr)
Three Mile Island PRA - L2 Containment Event Tree Analysis
_______________________________________________________________________________________________
6-54 0467060030-2788-050107
TABLE 6-3 TMI-1 SOURCE TERM SUMMARY
RELEASE CATEGORY INTACT BMMT LATE-SM LATE-LRG EARLY ISO-SM ISO-LRG ISLOCA SGTR
6) CsOH
Total Plume 1 Release Fraction 4.50E-04 9.00E-03 5.50E-03 2.00E-02 3.00E-02 1.00E-02 1.50E-01 5.00E-02 2.00E-02
Start of Plume 1 Release (hr) 26.00 26.00 10.00 70.80 11.70 1.00 10.00 16.00 14.00
End of Plume 1 Release (hr) 30.00 50.00 20.00 100.00 11.70 8.00 16.00 20.00 14.00
Total Plume 2 Release Fraction2 9.00E-02
Start of Plume 2 Release (hr) 34.00
End of Plume 2 Release (hr) 44.00
7) BaO
Total Plume 1 Release Fraction 1.80E-04 3.00E-03 1.00E-03 1.20E-05 5.00E-03 9.00E-03 1.50E-02 2.50E-02 2.00E-03
Start of Plume 1 Release (hr) 29.00 30.00 10.00 70.80 11.70 1.00 10.00 16.00 14.00
End of Plume 1 Release (hr) 32.00 40.00 20.00 70.80 20.00 8.00 16.00 20.00 14.00
Total Plume 2 Release Fraction2
Start of Plume 2 Release (hr)
End of Plume 2 Release (hr)
8) La2O3
Total Plume 1 Release Fraction 2.00E-06 5.50E-05 3.00E-05 5.50E-07 5.50E-04 1.00E-04 9.00E-04 2.50E-04 1.00E-04
Start of Plume 1 Release (hr) 29.00 30.00 10.00 70.80 11.70 1.00 14.00 16.00 14.00
End of Plume 1 Release (hr) 32.00 40.00 20.00 70.80 20.00 8.00 20.00 20.00 24.00
Total Plume 2 Release Fraction2
Start of Plume 2 Release (hr)
Three Mile Island PRA - L2 Containment Event Tree Analysis
_______________________________________________________________________________________________
6-55 0467060030-2788-050107
TABLE 6-3 TMI-1 SOURCE TERM SUMMARY
RELEASE CATEGORY INTACT BMMT LATE-SM LATE-LRG EARLY ISO-SM ISO-LRG ISLOCA SGTR
End of Plume 2 Release (hr)
9) CeO2
Total Plume 1 Release Fraction 1.00E-05 5.20E-04 5.00E-04 1.00E-05 1.50E-02 1.50E-03 2.00E-02 1.50E-03 2.00E-03
Start of Plume 1 Release (hr) 29.00 30.00 10.00 70.80 11.70 4.00 14.00 16.00 14.00
End of Plume 1 Release (hr) 32.00 50.00 20.00 70.80 20.00 10.00 20.00 26.00 24.00
Total Plume 2 Release Fraction2
Start of Plume 2 Release (hr)
End of Plume 2 Release (hr)
10) Sb
Total Plume 1 Release Fraction 4.00E-04 1.50E-02 8.00E-03 5.00E-02 1.80E-01 5.00E-02 1.50E-01 1.50E-01 7.00E-01
Start of Plume 1 Release (hr) 29.00 30.00 10.00 70.80 11.70 1.00 10.00 16.00 28.00
End of Plume 1 Release (hr) 32.00 40.00 20.00 120.00 20.00 8.00 20.00 20.00 30.00
Total Plume 2 Release Fraction2
Start of Plume 2 Release (hr)
End of Plume 2 Release (hr)
11) Te2
Total Plume 1 Release Fraction 0.00E+00 1.00E-04 3.00E-05 1.50E-03 2.00E-04 4.00E-03 7.00E-04 9.00E-05 2.00E-04
Start of Plume 1 Release (hr) 30.00 18.00 70.80 11.70 6.00 16.00 30.00 20.00
End of Plume 1 Release (hr) 40.00 20.00 70.80 20.00 16.00 20.00 40.00 24.00
Total Plume 2 Release Fraction2
Three Mile Island PRA - L2 Containment Event Tree Analysis
_______________________________________________________________________________________________
6-56 0467060030-2788-050107
TABLE 6-3 TMI-1 SOURCE TERM SUMMARY
RELEASE CATEGORY INTACT BMMT LATE-SM LATE-LRG EARLY ISO-SM ISO-LRG ISLOCA SGTR
Start of Plume 2 Release (hr)
End of Plume 2 Release (hr)
12) UO2
Total Plume 1 Release Fraction 0.00E+00 5.00E-06 2.80E-06 1.50E-06 1.20E-04 1.00E-05 2.00E-04 5.00E-06 1.00E-05
Start of Plume 1 Release (hr) 30.00 18.00 70.80 11.70 6.00 16.00 30.00 20.00
End of Plume 1 Release (hr) 50.00 20.00 70.80 20.00 16.00 20.00 40.00 24.00
Total Plume 2 Release Fraction2
Start of Plume 2 Release (hr)
End of Plume 2 Release (hr)
Notes:
(1) Puff releases are denoted in the table by those entries with equivalent start and end times.
(2) Plume 2 release fraction is cumulative and includes the initial plume 1 release fraction
(3) General Emergency declaration based on time of core damage per Radiological Emergency Plant for TMI, EP-AA-1009 Revision 7
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-57 0467060030-2788-050107
TABLE 6-4 RELEASE CATEGORY 1.01
CONTAINMENT BYPASS, OUTSIDE THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION
PRODUCTS, WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS
Time of Release 4.0 Hrs.
Duration of Release 1.0 Hrs.
Warning Time 3.0 Hrs.
Energy of Release 1.0E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS
Xe 1.0E+00
I 3.5E-03
Cs-Rb 3.5E-03
Te-Sb 1.8E-03
Ba 4.2E-05
Ru 1.2E-04
La 3.2E-07
Sr 4.3E-06 MAAP Run - TMI18C
TABLE 6-5 RELEASE CATEGORY 1.02
CONTAINMENT BYPASS, OUTSIDE THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS
Time of Release 3.0 Hrs.
Duration of Release 1.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.0E+06 Watts
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-58 0467060030-2788-050107
TABLE 6-5 RELEASE CATEGORY 1.02
CONTAINMENT BYPASS, OUTSIDE THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS
Elevation of Release 10.0 meters RELEASE FRACTIONS
Xe 1.0E+00
I 1.3E-02
Cs-Rb 1.3E-02
Te-Sb 1.6E-03
Ba 8.6E-05
Ru 1.7E-05
La 6.4E-07
Sr 8.5E-06 MAAP Run - TMI18C4
TABLE 6-6 RELEASE CATEGORY 2.01
CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING,WITHOUT EX-VESSEL RELEASE OF FISSION
PRODUCTS, WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 4.0 Hrs.
Duration of Release 1.0 Hrs.
Warning Time 3.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.7E-01
Cs-Rb 1.7E-01
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-59 0467060030-2788-050107
TABLE 6-6 RELEASE CATEGORY 2.01
CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING,WITHOUT EX-VESSEL RELEASE OF FISSION
PRODUCTS, WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Te-Sb 1.8E-02
Ba 1.1E-01
Ru 1.4E-01
La 6.9E-03
Sr 1.7E-02 Used RC 2.03 with FP ratios RC 1.03/RC 1.01
TABLE 6-7 RELEASE CATEGORY 2.02
CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS,
WITHOUT FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 3.0 Hrs.
Duration of Release 1.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 8.5E-01
Cs-Rb 8.5E-01
Te-Sb 9.0E-02
Ba 5.5E-01
Ru 7.0E-01
La 3.5E-02
Sr 8.5E-02 Used RC 2.01 release fractions with factor of 5 (Plateout)
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-60 0467060030-2788-050107
TABLE 6-8 RELEASE CATEGORY 2.03
CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS,
WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 4.0 Hrs.
Duration of Release 1.0 Hrs.
Warning Time 3.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.7E-01
Cs-Rb 1.7E-01
Te-Sb 1.6E-01
Ba 1.1E-01
Ru 1.4E-01
La 9.2E-03
Sr 1.8E-02 Used RC 2.04 release fractions with factor of 5 (Plateout)
TABLE 6-9 RELEASE CATEGORY 2.04
CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS,
WITHOUT FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 3.0 Hrs.
Duration of Release 1.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-61 0467060030-2788-050107
TABLE 6-9 RELEASE CATEGORY 2.04
CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS,
WITHOUT FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 8.5E-01
Cs-Rb 8.5E-01
Te-Sb 8.2E-01
Ba 5.7E-01
Ru 7.2E-01
La 4.6E-02
Sr 9.2E-02 MAAP Run - TMI19F
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-62 0467060030-2788-050107
TABLE 6-10 RELEASE CATEGORY 3.01
LARGE ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION
PRODUCTS, WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 1.5 Hrs.
Duration of Release 2.0 Hrs.
Warning Time 1.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.6E-02
Cs-Rb 2.6E-02
Te-Sb 2.0E-04
Ba 4.6E-03
Ru 1.6E-02
La 6.0E-05
Sr 9.0E-04 MAAP Run - ORAS9R
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-63 0467060030-2788-050107
TABLE 6-11 RELEASE CATEGORY 3.02
LARGE ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 1.5 Hrs.
Duration of Release 2.0 Hrs.
Warning Time 1.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.3E-01
Cs-Rb 1.3E-01
Te-Sb 1.0E-03
Ba 2.3E-02
Ru 8.0E-02
La 3.0E-04
Sr 4.5E-03 MAAP Run - ORAS9R
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-64 0467060030-2788-050107
TABLE 6-12 RELEASE CATEGORY 3.03
LARGE ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITH FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 1.5 Hrs.
Duration of Release 2.0 Hrs.
Warning Time 1.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 4.4E-02
Cs-Rb 4.4E-02
Te-Sb 2.4E-02
Ba-Sr 4.6E-03
Ru 1.8E-02
La 7.0E-05
Sr 1.3E-03 MAAP Run - ORAS9R
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-65 0467060030-2788-050107
TABLE 6-13 RELEASE CATEGORY 3.04
LARGE ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION
PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING RELEASE PARAMETERS
Time of Release 1.5 Hrs.
Duration of Release 2.0 Hrs.
Warning Time 1.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS
Xe 1.0E+00
I 2.2E-01
Cs-Rb 2.2E-01
Te-Sb 1.2E-01
Ba 2.3E-02
Ru 9.0E-02
La 3.5E-04
Sr 6.5E-03 Used RC 3.03 release fractions with factor of five (Plateout)
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-66 0467060030-2788-050107
TABLE 6-14 RELEASE CATEGORY 3.05
LARGE ISOLATION FAILURE, OUTSIDE THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION
PRODUCTS RELEASE PARAMETERS:
Time of Release 1.5 Hrs.
Duration of Release 2.0 Hrs.
Warning Time 1.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.3E-01
Cs-Rb 1.3E-01
Te-Sb 1.0E-03
Ba-Sr 2.3E-03
Ru 8.0E-02
La 3.0E-04
Sr 4.5E-03 Used RC 3.02 release fractions
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-67 0467060030-2788-050107
TABLE 6-15 RELEASE CATEGORY 3.06
LARGE ISOLATION FAILURE, OUTSIDE THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION
PRODUCTS RELEASE PARAMETERS:
Time of Release 1.5 Hrs.
Duration of Release 2.0 Hrs.
Warning Time 1.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.2E-01
Cs-Rb 2.2E-01
Te-Sb 1.2E-01
Ba 2.3E-02
Ru 9.0E-02
La 3.5E-04
Sr 6.5E-03 Used RC 3.04 release fractions
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-68 0467060030-2788-050107
TABLE 6-16 RELEASE CATEGORY 4.01
SMALL ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION
PRODUCTS, WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.0E-03
Cs-Rb 2.0E-03
Te-Sb 2.0E-04
Ba 1.1E-04
Ru 2.0E-04
La 4.0E-07
Sr 2.4E-05 MAAP Run - ORAR7L (Prior to concrete attack) compared to TMI7L.
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-69 0467060030-2788-050107
TABLE 6-17 RELEASE CATEGORY 4.02
SMALL ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.0E-02
Cs-Rb 1.0E-02
Te-Sb 1.0E-03
Ba 5.5E-04
Ru 1.0E-03
La 2.0E-06
Sr 1.2E-04 Used RC 4.01 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-70 0467060030-2788-050107
TABLE 6-18 RELEASE CATEGORY 4.03
SMALL ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITH FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.8E-03
Cs-Rb 3.2E-03
Te-Sb 4.0E-03
Ba 1.3E-04
Ru 1.3E-03
La 1.7E-06
Sr 6.4E-05 Used RC 4.04 with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-71 0467060030-2788-050107
TABLE 6-19 RELEASE CATEGORY 4.04
SMALL ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITH EX-VESSEL
RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.4E-02
Cs-Rb 1.6E-02
Te-Sb 2.0E-02
Ba 6.5E-04
Ru 6.5E-03
La 8.5E-06
Sr 3.2E-04 MAAP Run - ORAR7L (After concrete attack) compared to TMI7L.
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-72 0467060030-2788-050107
TABLE 6-20 RELEASE CATEGORY 4.05
SMALL ISOLATION FAILURE, TO THE ENVIRONMENT, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS,
WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.6E-03
Cs-Rb 2.8E-03
Te-Sb 2.0E-04
Ba 5.0E-04
Ru 2.0E-04
La 1.8E-06
Sr 5.8E-05 Used RC 4.06 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-73 0467060030-2788-050107
TABLE 6-21 RELEASE CATEGORY 4.06
SMALL ISOLATION FAILURE, TO THE ENVIRONMENT, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS,
WITHOUT FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.3E-02
Cs-Rb 1.4E-02
Te-Sb 1.0E-03
Ba 2.5E-03
Ru 1.0E-03
La 9.0E-06
Sr 2.9E-04 MAAP Run - ORAR9L
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-74 0467060030-2788-050107
TABLE 6-22 RELEASE CATEGORY 4.07
SMALL ISOLATION FAILURE, TO THE ENVIRONMENT, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS,
WITHOUT FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 5.0E-03
Cs-Rb 6.2E-03
Te-Sb 7.0E-03
Ba 5.0E-04
Ru 2.4E-03
La 4.6E-06
Sr 1.4E-04 Used RC 4.08 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
6-75 0467060030-2788-050107
TABLE 6-23 RELEASE CATEGORY 4.08
SMALL ISOLATION FAILURE, TO THE ENVIRONMENT, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS,
WITHOUT FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 1.9E+06 Watts
Elevation of Release 0.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.5E-02
Cs-Rb 3.1E-02
Te-Sb 3.5E-02
Ba 2.5E-03
Ru 1.2E-02
La 2.3E-05
Sr 6.9E-04 MAAP Run - ORAR9L
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-76 0467060030-2788-050107
TABLE 6-24 RELEASE CATEGORY 5.01
EARLY CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE
RELEASE PARAMETERS:
Time of Release 3.25 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 2.75 Hrs.
Energy of Release 2.8E+07 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.6E-02
Cs-Rb 1.6E-02
Te-Sb 8.0E-03
Ba 3.5E-04
Ru 9.3E-04
La 1.3E-05
Sr 3.0E-05 MAAP Run - MRA8PI2 compared to TMI7F3.
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-77 0467060030-2788-050107
TABLE 6-25 RELEASE CATEGORY 5.02
EARLY CONTAINMENT FAILURE, WITH EX-VESSEL FISSION PRODUCT RELEASE
RELEASE PARAMETERS:
Time of Release 5.5 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 5.0 Hrs.
Energy of Release 2.8E+07 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.4E-02
Cs-Rb 1.3E-02
Te-Sb 1.2E-02
Ba 8.7E-04
Ru 1.8E-03
La 3.8E-03
Sr 2.2E-04 MAAP Run - MRA7PI6 compared to TMI7F3.
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-78 0467060030-2788-050107
TABLE 6-26 RELEASE CATEGORY 6.01
LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE, WITHOUT REVAPORIZATION, WITH
FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 45.0 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 24.0 Hrs.
Energy of Release 2.8E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
X 1.0E+00
I 8.0E-04
Cs-Rb 1.8E-03
Te-Sb 2.0E-05
Ba 1.4E-05
Ru 4.0E-06
La 2.0E-07
Sr 1.0E-06 Used RC 6.02 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-79 0467060030-2788-050107
TABLE 6-27 RELEASE CATEGORY 6.02
LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION
PRODUCT RELEASE, WITHOUT REVAPORIZATION, WITHOUT FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 45.0 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 24.0 Hrs.
Energy of Release 2.8E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 4.0E-03
Cs-Rb 9.0E-03
Te-Sb 1.0E-04
Ba 7.0E-05
Ru 2.0E-05
La 1.0E-06
Sr 5.0E-06 MAAP Run - ORAS12F, ORAR5E, ORAS13F, ORAR33 compared to TMI7F, TMI7F2, and
TMI4F.
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-80 0467060030-2788-050107
TABLE 6-28 RELEASE CATEGORY 6.03
LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION
PRODUCT RELEASE, WITH REVAPORIZATION, WITH FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 45.0 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 24.0 Hrs.
Energy of Release 2.8E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.0E-02
Cs-Rb 2.0E-02
Te-Sb 2.0E-05
Ba 1.4E-05
Ru 4.0E-06
La 2.0E-07
Sr 1.0E-06 Used RC 6.04 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-81 0467060030-2788-050107
TABLE 6-29 RELEASE CATEGORY 6.04
LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE, WITH REVAPORIZATION, WITHOUT
FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 45.0 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 24.0 Hrs.
Energy of Release 2.8E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.0E-01
Cs-Rb 1.0E-01
Te-Sb 1.0E-04
Ba 7.0E-05
Ru 2.0E-05
La 1.0E-06
Sr 5.0E-06 MAAP Run - ORAS12F, ORAR5E, ORAS13F, ORAR33, NUREG-1150 compared to
TMI7F, TMI7F2, and TMI4F.
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-82 0467060030-2788-050107
TABLE 6-30 RELEASE CATEGORY 6.05
LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT REVAPORIZATION, WITH
FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 45.0 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 24.0 Hrs.
Energy of Release 2.8E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 8.0E-04
Cs-Rb 1.8E-03
Te-Sb 4.0E-03
Ba 1.4E-05
Ru 4.0E-05
La 2.0E-07
Sr 1.0E-06 Used RC 6.06 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-83 0467060030-2788-050107
TABLE 6-31 RELEASE CATEGORY 6.06
LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE,
WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT REVAPORIZATION, WITHOUT FISSION
PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 45.0 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 24.0 Hrs.
Energy of Release 2.8E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 4.0E-03
Cs-Rb 9.0E-03
Te-Sb 2.0E-02
Ba 7.0E-05
Ru 2.0E-04
La 1.0E-06
Sr 5.0E-06 MAAP Run - ORAS12F, ORAR5E, ORAS13F, ORAR34T, NUREG-1150 compared to
TMI7F, TMI7F2, and TMI4F.
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-84 0467060030-2788-050107
TABLE 6-32 RELEASE CATEGORY 6.07
LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE,
WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITH REVAPORIZATION, WITH FISSION PRODUCT
SCRUBBING RELEASE PARAMETERS:
Time of Release 45.0 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 24.0 Hrs.
Energy of Release 2.8E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.0E-02
Cs-Rb 2.0E-02
Te-Sb 4.0E-03
Ba 1.4E-05
Ru 4.0E-05
La 2.0E-07
Sr 1.0E-06 Used RC 6.08 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-85 0467060030-2788-050107
TABLE 6-33 RELEASE CATEGORY 6.08
LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITH REVAPORIZATION, WITHOUT
FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 45.0 Hrs.
Duration of Release 0.5 Hrs.
Warning Time 24.0 Hrs.
Energy of Release 2.8E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.0E-01
Cs-Rb 1.0E-01
Te-Sb 2.0E-02
Ba 7.0E-05
Ru 2.0E-04
La 1.0E-06
Sr 5.0E-06 MAAP Run - ORAS12F, ORAR5E, ORAS13F, ORAR34R, NUREG-1150 compared to
TMI7F, TMI7F2, and TMI4F.
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-86 0467060030-2788-050107
TABLE 6-34 RELEASE CATEGORY 7.01
LATE OVERPRESSURIZATION, WITH BENIGN CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION
PRODUCT RELEASE, WITH FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 14.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 12.0 Hrs.
Energy of Release 1.0E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.0E-04
Cs-Rb 4.0E-04
Te-Sb 2.0E-05
Ba 4.0E-06
Ru 4.0E-06
La 1.0E-07
Sr 4.0E-07 Used RC 7.02 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-87 0467060030-2788-050107
TABLE 6-35 RELEASE CATEGORY 7.02
LATE OVERPRESSURIZATION, WITH BENIGN CONTAINMENT FAILURE, WITHOUT
EX-VESSEL FISSION PRODUCT RELEASE, WITHOUT FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 14.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 12.0 Hrs.
Energy of Release 1.0E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.0E-03
Cs-Rb 2.0E-03
Te-Sb 1.0E-04
Ba 2.0E-05
Ru 2.0E-05
La 5.0E-07
Sr 2.0E-06 MAAP Run - ORAB12F, ORAR33S, ORAB5E compared to TMI7F, TMI7F2, and TMI4F.
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-88 0467060030-2788-050107
TABLE 6-36 RELEASE CATEGORY 7.03
LATE OVERPRESSURIZATION, WITH BENIGN CONTAINMENT FAILURE, WITH EX-VESSEL RELEASE OF
FISSION PRODUCTS, WITH FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 14.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 12.0 Hrs.
Energy of Release 1.0E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 2.0E-04
Cs-Rb 4.0E-04
Te-Sb 2.0E-04
Ba 4.0E-06
Ru 4.0E-06
La 1.0E-07
Sr 4.0E-07 Used RC 7.04 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-89 0467060030-2788-050107
TABLE 6-37 RELEASE CATEGORY 7.04
LATE OVERPRESSURIZATION, WITH BENIGN CONTAINMENT FAILURE, WITH EX-VESSEL RELEASE OF
FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING
RELEASE PARAMETERS:
Time of Release 14.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 12.0 Hrs.
Energy of Release 1.0E+06 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 1.0E-03
Cs-Rb 2.0E-03
Te-Sb 1.0E-03
Ba 2.0E-05
Ru 2.0E-05
La 5.0E-07
Sr 2.0E-06 MAAP Run - ORAB12F, ORAR34S, ORAB5E compared to TMI7F, TMI7F2, and TMI4F.
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-90 0467060030-2788-050107
TABLE 6-38 RELEASE CATEGORY 8.01
CONTAINMENT FAILURE FROM BASEMAT MELT-THROUGH, WITH EX-VESSEL RELEASE OF FISSION
PRODUCTS RELEASE PARAMETERS:
Time of Release 36.0 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 6.0 Hrs.
Energy of Release 0.0 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E+00
I 3.0E-02
Cs-Rb 2.0E-05
Te-Sb 1.4E-05
Ba 8.0E-07
Ru 7.0E-06
La 1.4E-08
Sr 2.5E-07 MAAP Run - ORAR7F/ORAR12F/ORAR13F compared to TMI4F, TMI7F and TMI7F2.
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-91 0467060030-2788-050107
TABLE 6-39 RELEASE CATEGORY 9.01
NO CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE, WITH FISSION PRODUCT
SCRUBBING RELEASE PARAMETERS:
Time of Release 0.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 0.0 Hrs.
Energy of Release 0.0 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E-03
I 7.0E-07
Cs-Rb 7.0E-07
Te-Sb 2.0E-09
Ba 4.0E-08
Ru 2.0E-09
La 2.0E-09
Sr 4.0E-09 MAAP Run - ORAR1A
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-92 0467060030-2788-050107
TABLE 6-40 RELEASE CATEGORY 9.02
NO CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE, WITHOUT FISSION
PRODUCT SCRUBBING RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 0.0 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E-03
I 2.0E-05
Cs-Rb 2.0E-05
Te-Sb 1.0E-06
Ba 8.0E-07
Ru 1.0E-06
La 1.4E-08
Sr 2.5E-07 MAAP Run - ORAR7F/ORAR12F/ORAR13F
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-93 0467060030-2788-050107
TABLE 6-41 RELEASE CATEGORY 9.03
NO CONTAINMENT FAILURE, WITH EX-VESSEL FISSION PRODUCT RELEASE, WITH FISSION PRODUCT
SCRUBBING RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 0.0 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E-03
I 4.0E-06
Cs-Rb 4.0E-06
Te-Sb 2.8E-06
Ba 1.6E-07
Ru 1.4E-06
La 2.8E-09
Sr 5.0E-08 Used RC 9.04 release fractions with factor of 5 (Scrubbing)
Three Mile Island PRA - L2 Containment Event Tree Analysis
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6-94 0467060030-2788-050107
TABLE 6-42 RELEASE CATEGORY 9.04
NO CONTAINMENT FAILURE, WITH EX-VESSEL FISSION PRODUCT RELEASE, WITHOUT FISSION PRODUCT
SCRUBBING RELEASE PARAMETERS:
Time of Release 2.5 Hrs.
Duration of Release 10.0 Hrs.
Warning Time 2.0 Hrs.
Energy of Release 0.0 Watts
Elevation of Release 10.0 meters RELEASE FRACTIONS:
Xe 1.0E-03
I 2.0E-05
Cs-Rb 2.0E-05
Te-Sb 1.4E-05
Ba 8.0E-07
Ru 7.0E-06
La 1.4E-08
Sr 2.5E-07 MAAP Run - ORAR7F/ORAR12F/ORAR13F
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7-1 0467060030-2788-050107
7.0 CONTAINMENT EVENT TREE SOLUTION
The resulting frequencies for each release category described in Section 6.4 are listed
in Table 7-1. PRAQUANT version 4.0a and FORTE version 3.0b were used to calculate
each of the release category cutset files. The CAFTA PRA model, including the
reliability database and support files necessary for quantification of these release
categories, is described in Reference [32]. Also, because the Level 2 analysis required
quantification of both the success and failure branches of the Containment Event Tree,
the quantification time using FORTE was initially interminable. This was attributed to
certain basic events with high probabilities, e.g. 0.9 or higher. Therefore, in an effort to
streamline quantification, certain Level 2 events with probabilities of 0.9 or higher were
set to TRUE using a CAFTA flag file. This could be viewed as being somewhat
conservative in the quantification of the “success” branches of the event tree in that
certain release category frequencies may be somewhat higher, since the generated
cutsets would not contain the events with a probability of 0.9 that were set to TRUE.
However, this conservative assessment was deemed necessary in order to afford
quantification of the Level 2 cutsets within a reasonable time frame.
TABLE 7-1 RELEASE CATEGORY FREQUENCIES
(ITEMS LISTED IN BOLD ARE CONTRIBUTORS TO LERF) RELEASE CATEGORY
DESIGNATOR FREQUENCY (1/YR) PERCENTAGE OF CDF
1-01 4.57E-07 2.00%
1-02 1.59E-06 7.10%
2-01 0.00E-00 0.00%
2-02 1.81E-07 0.80%
2-03 0.00E-00 0.00%
2-04 1.27E-08 0.10%
3-01 9.07E-11 0.00%
Three Mile Island PRA - L2 Containment Event Tree Analysis
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7-2 0467060030-2788-050107
TABLE 7-1 RELEASE CATEGORY FREQUENCIES
(ITEMS LISTED IN BOLD ARE CONTRIBUTORS TO LERF) RELEASE CATEGORY
DESIGNATOR FREQUENCY (1/YR) PERCENTAGE OF CDF
3-02 9.07E-11 0.00%
3-03 1.90E-10 0.00%
3-04 2.88E-10 0.00%
3-05 0.00E-00 0.00%
3-06 0.00E-00 0.00%
4-01 3.90E-08 0.20%
4-02 1.46E-08 0.10%
4-03 8.54E-09 0.00%
4-04 3.16E-07 1.40%
4-05 0.00E-00 0.00%
4-06 0.00E-00 0.00%
4-07 0.00E-00 0.00%
4-08 0.00E-00 0.00%
5-01 7.39E-07 3.30%
5-02 1.66E-07 0.70%
6-01 0.00E-00 0.00%
6-02 0.00E-00 0.00%
6-03 2.20E-08 0.10%
6-04 2.36E-10 0.00%
6-05 2.08E-11 0.00%
6-06 0.00E-00 0.00%
6-07 8.00E-08 0.40%
6-08 1.43E-08 0.10%
7-01 2.25E-07 1.00%
7-02 2.75E-09 0.00%
7-03 7.45E-07 3.30%
7-04 2.89E-07 1.30%
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7-3 0467060030-2788-050107
TABLE 7-1 RELEASE CATEGORY FREQUENCIES
(ITEMS LISTED IN BOLD ARE CONTRIBUTORS TO LERF) RELEASE CATEGORY
DESIGNATOR FREQUENCY (1/YR) PERCENTAGE OF CDF
8-01 3.19E-06 14.20%
9-01 1.20E-05 53.40%
9-02 1.69E-08 0.10%
9-03 2.36E-06 10.50%
9-04 1.91E-08 0.10%
During the quantification of certain release category top events, it was found that
quantification at a truncation limit of 1.0E-11 was impractical due to lengthy computer
run times. Therefore, Table 7-2 identifies those release categories that were evaluated
at a higher truncation (5.0E-11) and compares the results with quantifications performed
at a lower truncation of 1.0E-11. This comparison showed that the additional risk at the
lower truncation as a percentage of the Level 1 CDF (2.37E-05/yr) was minimal.
TABLE 7-2 TRUNCATION LIMIT COMPARISON FOR CERTAIN RELEASE CATEGORIES RELEASE CATEGORY
DESIGNATOR FREQUENCY (1/YR)
AT 1.0E-11 FREQUENCY (1/YR)
AT 5.0E-10 %DIFFERENCE OF
LEVEL 1 CDF
6-08 2.83E-08 1.43E-08 0.1%
7-01 2.94E-07 2.25E-07 0.3%
7-02 5.41E-09 2.75E-09 0.0%
7-03 7.86E-07 7.45E-07 0.2%
7-04 3.86E-07 2.89E-07 0.4%
9-01 1.24E-05 1.20E-05 1.7%
9-03 2.56E-06 2.36E-06 0.9%
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7-4 0467060030-2788-050107
7.1 TREATMENT OF ILLOGICAL CUTSETS
In order to help eliminate cutsets containing illogical combinations of events, e.g.,
recovery of the RB spray system due to loss of offsite power along with a mechanical
failure of the system, recovery rules and QRECOVER32 were employed to exclude
these cutsets. The method employed made use of a recovery event (RECZED) with a
zero probability. Hence, when a particular cutset matches the excluded logic, this event
is appended to the cutset and results in the overall probability being zero. Figure 7-1
below shows the logic used in the recovery rules CAFTA file regarding illogical
combinations of events for the RB spray system.
Figure 7-1 Model Logic Used to Exclude Non-Realistic Cutsets Associated with Reactor
Building Spray
Three Mile Island PRA - L2 Containment Event Tree Analysis
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8.0 REFERENCES
1. Gregory, J. J., et al., Evaluation of Severe Accident Risks: Sequoyah, Unit 1 NUREG/CR-4551, U.S. Nuclear Regulatory Commission, Washington, D.C., December 1990.
2. Kouts, H., Review of Research on Uncertainties in Estimates of Source Terms from Severe Accidents in Nuclear Power Plants, NUREG/CR-4883, U.S. Nuclear Regulatory Commission, Washington, D.C., 1987.
3. Williams, D.C., et al., Containment Loads Due to Direct Containment Heating and Associated Hydrogen Behavior: Analysis and Calculations with the CONTAIN Code, NUREG/CR-4896, U.S. Nuclear Regulatory Commission, Washington, D.C., May 1987.
4. MAAP Modular Accident Analysis Program User's Manual, IDCOR Technical Report 16.2-3, Fauske and Associates, Inc., Burr Ridge, IL, February 1987.
5. Marx, K.D., A Model for the Transport and Chemical Reaction of Molten Debris in Direct Containment Heating Experiments, NUREG/CR-5120, Sandia National Laboratories, Livermore, CA, May 1988.
6. Thompson, R.T., Large-Scale Hydrogen Combustion Experiments, NP-3878, Electric Power Research Institute, Palo Alto, CA, October 1988.
7. Corrandini, M.L., Swenson, D.V., Probability of Containment Failure Due to Steam Explosions Following A Postulated Core Meltdown In An LWR, NUREG/CR-2214, U.S. Nuclear Regulatory Commission, Washington, D.C., 1981.
8. Tarbell, W.W., et al., Sustained Concrete Attack by Low-Temperature, Fragmented Core Debris, NUREG/CR-3024, U.S. Nuclear Regulatory Commission, Washington, D.C., 1987.
9. Haskin, F.E., et al., Analysis of Hypothetical Severe Core Damage Accidents for the Zion Pressurized Water Reactor, NUREG/CR-1989, Sandia National Laboratories, Livermore, CA, October 1982.
10. Technical Report 12.3, Hydrogen Combustion in Reactor Containment Building, IDCOR Program Report, Fauske and Associates, Inc., Burr Ridge, IL, September 1983.
11. Technical Report 15.3, Core Concrete Interactions, IDCOR Program Report, Fauske and Associates, Inc., Burr Ridge, IL, September 1983.
12. Technical Report 86.1, Status of Technical Issue Resolutions, IDCOR Program Report, Fauske and Associates, Inc., Burr Ridge, IL., October 1988.
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13. Chambers, R., Depressurization to Mitigate Direct Containment Heating, Nuclear Technology, Vol. 88, December 1989.
14. Hawley, J.T., et al., Evaluation of the Consequences of Containment Bypass Scenarios, NP-6586-L, Electric Power Research Institute, Palo Alto, CA, November 1989.
15. Letter from J.S. Wetmore (GPUN) to E.H. Domaleski,GPUN File No. 5430-92-0015, Dated April 22, 1992, Including followup Telecon between D.B. Noxon (DE&S) and C.D. Adams (GPUN), DE&S File No. MTS-4042, May 21, 1992.
16. Oconee Nuclear Station Unit 3 Probabilistic Risk Assessment, Duke Power Company, transmitted by letter dated November 30, 1990 from M.S. Tuckman (Duke Power) to NRC Document Control Desk (in response to generic letter 88-20).
17. Blose, R.E., et al., SWISS: Sustained Heated Metallic Melt/Concrete Interactions With Overlying Water Pools, NUREG/CR-4727, Sandia National Laboratories, Livermore, CA, July 1987.
18. Standard Review Plan, Section 6.5.3, NUREG-0800, U.S. Nuclear Regulatory Commission, July 1981.
19. Technical Report 85.2, Technical Support for Issue Resolution, IDCOR Program Report, Fauske and Associates, Inc., Burr Ridge, IL, July 1985.
20. Technical Report 11.6, Resuspension of Deposited Aerosols Following Primary System or Containment Failure, Fauske and Associates, Inc., Burr Ridge, IL, August 1984.
21. Reactor Risk Reference Document, NUREG-1150, U.S. Nuclear Regulatory Commission, February 1987.
22. Technical Report 15.2B, Debris Coolability, Vessel Penetration, and Debris Coolability, Fauske and Associates, Inc., Burr Ridge, IL, August 1983.
23. Allen, M.D., et al., “Experiments to Investigate the Effects of Flight Path on Direct Containment Heating”, Nuclear Technology, Volume 100, 52, October 1992.
24. TMI Level 2 Notebook – Vol. 1, “Level 1 to Level 2 Interface”, TMI-PRA-015.1, Draft Version.
25. Three Mile Island System Notebook, “Reactor Building Isolation System Notebook”, P0467050015-2545, TM-PRA-010.15 (RBIS), Draft version.
26. Oconee PRA, A Probabilistic Risk Assessment of Oconee Unit 3, Revision 1, Duke Power Company, February 1990.
Three Mile Island PRA - L2 Containment Event Tree Analysis
_____________________________________________________________________
8-3 0467060030-2788-050107
27. MAAP Modular Accident Analysis Program User's Manual, IDCOR Technical Report 16.2-3, Fauske and Associates, Inc., Burr Ridge, IL, February 1987.
28. Fission Product Transport in Degraded Core Accidents, IDCOR Technical Report on Task 11.3, December 1983.
29. Evaluation of Severe Accident Risks: Surry Unit 1, NUREG/CR-4551, October 1990.
30. U.S. Nuclear Commission, Severe Accident Risks: An Assessment for Five US Nuclear Power Plants, NUREG/CR-1150, June 1989.
31. PBAPS PRA Initiating Events Notebook, Appendix C, Table C-2, PB-PRA-001, Rev. 1, February 2006.
32. TMI PRA-014 Quantification Notebook, Rev.1, PRA Model TM1042, 2007.
Three Mile Island PRA - L2 Containment Event Tree Analysis Appendix A - Decision Logic for Event Tree Nodes
_____________________________________________________________________ 0467060030-2788-050107
APPENDIX A
Decision Logic for Event Tree Nodes
Three Mile Island PRA - L2 Containment Event Tree Analysis Appendix A - Decision Logic for Event Tree Nodes
_____________________________________________________________________
A-1 0467060030-2788-050107
BYPASS
NOBYPASS
EARLY
NOEARLY
LATE
NOLATE
EXRELEASE
NOEXRELEASE
BASEMENT
NOBASEMENT
LATEREVAP
NOLATEREVAP
FPSCRUBBED
FPNOSCRUBBED
FPSCRUBBED2
FPNOSCRUBBED2
Three Mile Island PRA - L2 Containment Event Tree Analysis Appendix B - Containment Capacity
_____________________________________________________________________ 0467060030-2788-050107
APPENDIX B
TMI-1 Containment Capacity
Three Mile Island PRA - L2 Containment Event Tree Analysis Appendix B - Containment Capacity
_____________________________________________________________________
B-1 0467060030-2788-050107
TMI_ctmt_Capacity
Three Mile Island PRA - L2 Containment Event Tree Analysis Appendix C - ONS Containment Capacity
_____________________________________________________________________ 0467060030-2788-050107
APPENDIX C
ONS Containment Capacity
Three Mile Island PRA - L2 Containment Event Tree Analysis Appendix C - ONS Containment Capacity
_____________________________________________________________________
C-1 0467060030-2788-050107
ONS_ctmt_Capacity
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BASEMELT"(2004 Rev. 2)
Page 1 3/22/2007
Containment Failure fromBasemat Melt-Through
BASEMELT
Corium Pool Does Not SpreadOver Large Area Or Freeze
NOCOREFREEZE
Corium Does Not SpreadAcross Lower Containment
Or Cavity Floor
NOSPREADLOW
Primary System Pressure isLow At RV Failure
LOWPRESS
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOW
PDS INDICATES LOWPRESSURE AT CORE MELT
IS CERTAIN
PDSLOW-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
PDSLOW-2
LOW PRESSURE AT COREMELT IS INDETERMINATE
PDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 2
Operators DepressurizeSteam Generators
OPSSSHR
Steam GeneratorDepressurization and SSHR
Are Available
PORVSSHR
PDS INDICATES OTSGADVS ARE AVAILABLE
PDSSGADV
AV
Secondary Side HeatRemoval is Available
SSHRAVAIL
Page 3
Likelihood That OperatorsDepressurize Steam
Generators
OPSDEPRESS
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
Cavity Geometry Does NotAllow Enough Corium toDisperse For Freezing
NOGEOMFREEZE
Likelihood Corium Does NotFreeze On Lower
Containment or Cavity Floor
NOFREEZELOW
Water Pool Does Not StopConcrete Attack Prior toBasemat Melt-Through
ATTACK
Page 4
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BASEMELT"(2004 Rev. 2)
Page 2 3/22/2007
Operators Depressurize WithPressurizer PORV
PZRPORVPage 1
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
LOGIC FOR OPERATORSOPENING PORV
PZPORVCONF-1
PROB THAT OPERATORSOPEN PORV
PZPORVCONF_99
LOGIC FOR OPERATORSOPENING PORV WHEN HPI
COOLING IS NOTINITIATED
PZPORVCONF-2
Prob. That Operators OpenPORV After Failing to Init
HPI Cooling
PZPORVCONF_0
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
PROBABILITY THAT PORVCAN PREVENT HPME
PRVHPCONF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BASEMELT"(2004 Rev. 2)
Page 3 3/22/2007
Secondary Side HeatRemoval is Available
SSHRAVAILPage 1
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
Prob. that Secondary SideHeat Removal is Recovered
Prior to RV Failure
RECOVSSHR
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BASEMELT"(2004 Rev. 2)
Page 4 3/22/2007
Water Pool Does Not StopConcrete Attack Prior toBasemat Melt-Through
ATTACKPage 1
Water Is Unavailable InCavity Prior To Basemat
Melt-Through
NOBMMWATER
Water Unavailable fromContainment Sprays Via Fuel
Transfer Pool Prior to LCF
NOFTRNSPOOLLT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
CS FAILURE FORINJECTION MODE
CS01
Containment Sprays Are NotRecovered Prior to RV
Failure
NORECOVSPPRI
RB SPRAY POWERSUPPLIES ARE NOT
RECOVERED PRIOR TOLATE CTMT FAILURE
NORECOVSPLT
RB SPRAY UNAVAILABLEDUE TO MECH FAILUREOR NO OFFSITE POWER
NORECOVSPLT-1
CS FAILURE FORINJECTION MODE
(POST-LOOP RECOVERY)
CS01-R
OFFSITE POWER NOTRECOVERED WITHIN 24
HOURS
NORECOFFSITEPWR
IE-LOOP-101
Containment Sprays Are NotRecovered Early After RV
Failure
NORECOVSPAFT
Likelihood That Water Poolin Cavity Will Not Stop
Concrete Attack
MELT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BASEMELT"(2004 Rev. 2)
Page 5 3/22/2007
Name Page Zone Name Page Zone
ATTACK 1 7ATTACK 4 3AV 1 6BASEMELT 1 7BWHBW1-----HP2OA 2 4CAG0005 3 1CAG0005-R 3 2CM-001 1 1CM-002 1 2CM-003 1 2CM-004 1 4CM-005 1 5CM-006 1 4CS01 4 1CS01-R 4 2HEATIML 1 8IE-LOOP-100 3 3IE-LOOP-101 4 4LOWPRESS 1 5MELT 4 3NO-SSHR 3 2NO-SSHR-POSTLOOP 3 2NOBMMWATER 4 2NOCOREFREEZE 1 6NOFREEZELOW 1 7NOFTRNSPOOLLT 4 3NOGEOMFREEZE 1 6NOPDSSPRAY 4 1NORECOFFSITEPWR 4 3NORECOVRV 1 8NORECOVSPAFT 4 4NORECOVSPLT 4 3NORECOVSPLT-1 4 3NORECOVSPPRI 4 2NOSPREADLOW 1 6OPSDEPRESS 1 7OPSSSHR 1 7PDSINDET 1 4PDSLOW 1 2
PDSLOW-1 1 2PDSLOW-2 1 4PDSLOW_5 1 3PDSPZRPORV 2 1PDSSGADV 1 6PDSSSHR 3 1PO-HPICOOL 2 1PORVSSHR 1 6PRVHPCONF 2 4PZPORVCONF-1 2 3PZPORVCONF-2 2 3PZPORVCONF_0 2 3PZPORVCONF_99 2 2PZRPORV 1 3PZRPORV 2 2PZRSAFETY 1 9RECOVSSHR 3 2SSHRAVAIL 1 7SSHRAVAIL 3 2YES-SSHR 3 1
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BYPASS"(2004 Rev. 2)
Page 1 3/22/07
Containment Bypass
BYPASS
Interfacing System LOCA(IS-LOCA)
ISLOCA
PDS INDICATES THATISLOCA EXISTS
NOPDSNOISL
CORE MELT BIN 19
CM-019
Likelihood of InducedIS-LOCA
IISL
SGTR-Containment-BypassSequences
SGTRCB
PDS INDICATES SGTREXISTS
NOPDSNOSGTR
CORE MELT BINREPRESENTS SGTR
CM-15-18
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
InducedSGTR-Containment-Bypass
Sequence
ISGTRCB
Induced SGTR
ISGTR
S/G Tube Temp(s) InduceCreep Rupture Failure
HIGHSGTTEMP
S/G Tube Temperature(s)Are High With SSHR
Available
HIGHSGTSSHRAV
Likelihood That SSHR WillNot Keep Tubes Cool
SSHRSGTNOCOOL
Secondary Side HeatRemoval is Unavailable
SSHRUNAVAIL
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
Page 2
Prob. that Secondary SideHeat Removal is Not
Recovered Prior to RVFailure
NORECOVSSHR
Heat Transfer to S/G Tubesis High
HIGHTUBEHT
Page 3
Primary To Secondary DeltaP Induces Creep Rupture
Failure
HIGHDELP
Page 4
Likelihood That FPs Are NotReleased to ContainmentInstead of the Enviroment
CBREL
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BYPASS"(2004 Rev. 2)
Page 2 3/22/07
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHRPage 1
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BYPASS"(2004 Rev. 2)
Page 3 3/22/07
Heat Transfer to S/G Tubesis High
HIGHTUBEHTPage 1
Primary System NaturalCirculation Heat Transfer to
S/G Tubes is High
HIGHNATHT
Likelihood That NaturalCirculation Heat Transfer is
High
UNEFFNATHT
Reactor Coolant Pumps AreNot Running
RCPUMPOFFPage 6
No Power To The ReactorCoolant Pumps (RCPs)
NORECOVPOWER
PDS INDICATES POWER ISUNAVAILABLE TO RCPS
NOPDSRCPWR
UNAVAILABILITY OFSUPPORT SYSTEMS FOR
RCPS
DPG0003
Power Is Not Recovered tothe RCPs Prior to RV Failure
NORECACPRI
Likelihood That Operators DoNot Start the Reactor Coolant
Pumps
NONCONBYOPS
Primary System ForcedCirculation Heat Transfer to
S/G Tubes is High
HIGHFORCEHT
Likelihood That ForcedCirculation Heat Transfer is
High
NOEFFFORCEHT
Reactor Coolant Pumps AreRunning
RCPUMPON
Page 6
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BYPASS"(2004 Rev. 2)
Page 4 3/22/07
Primary To Secondary DeltaP Induces Creep Rupture
Failure
HIGHDELPPage 1
PDS INDICATES RCSPRESSURE IS
NOTSLIGHTLY ABOVE ORBELOW SG PRESS
NOPDSRCEQSG
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
Operators Do Not Depress.With Pressurizer PORVsPrior to S/G Tube Failure
NOOPSDEPRES2
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
Likelihood That OperatorsFail to Open the PORV Prior
to S/G Tube Failure
NOEFFDEPRESS-1
OPERATOR FAILS TOOPEN PORV
EFFDEPRESS_99-C
CONDITIONAL PROB THATOPERATOR FAILS TO
OPEN PORV
NOEFFDEPRESS-2
Page 5
Likelihood That PressurizerPORV(s) Cannot Depress
Primary System to S/G Press
PZRNOPORVDEP
Primary System Failure DoesNot Reduce RC PressurePrior to S/G Tube Failure
NOPRIMFAILURE
Page 6
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BYPASS"(2004 Rev. 2)
Page 5 3/22/07
CONDITIONAL PROB THATOPERATOR FAILS TO
OPEN PORV
NOEFFDEPRESS-2Page 4
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
OPERATOR FAILS TOOPEN PORV AFTER
FAILING TO INITIATE HPICOOLING
EFFDEPRESS_0-C
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BYPASS"(2004 Rev. 2)
Page 6 3/22/07
Primary System Failure DoesNot Reduce RC PressurePrior to S/G Tube Failure
NOPRIMFAILUREPage 4
Primary System Failure DoesNot Precede S/G Tube
Failure With RCPs Running
NOPRIMFAILPMP
Conf That Primary SysFailure Does Not Precede
S/G Tube Failure W/ RCPsOn
NOEFFPMP
Reactor Coolant Pumps AreRunning
RCPUMPONPage 3
Power To The ReactorCoolant Pumps (RCPs)
RECOVPOWER
PDS INDICATES POWER ISAVAILABLE TO RCPS
PDSRCPWR
UNAVAILABILITY OFSUPPORT SYSTEMS FOR
RCPS
DPG0003
Power Is Recovered to theRCPs Prior to RV Failure
RECACPRI
Likelihood That OperatorsStart the Reactor Coolant
Pumps
NCONBYOPS
Primary System Failure DoesNot Precede S/G TubeFailure With RCPs Off
NOPRIMFAILNPMP
Conf That Primary SysFailure Does Not Precede
S/G Tube Failure W/ RCPsOff
NOEFFNPMP
Reactor Coolant Pumps AreNot Running
RCPUMPOFF
Page 3
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "BYPASS"(2004 Rev. 2)
Page 7 3/22/07
Name Page Zone Name Page Zone
BWHBW1-----HP2OA 5 1BYPASS 1 3CAG0005 2 1CAG0005-R 2 2CBREL 1 6CM-009 4 1CM-010 4 1CM-011 4 1CM-012 4 2CM-013 4 2CM-014 4 2CM-015 1 3CM-016 1 3CM-018 1 4CM-019 1 1CM-15-18 1 3DPG0003 3 1DPG0003 6 1EFFDEPRESS_0-C 5 2EFFDEPRESS_99-C 4 4HIGHDELP 1 6HIGHDELP 4 3HIGHFORCEHT 3 4HIGHNATHT 3 2HIGHSGTSSHRAV 1 4HIGHSGTTEMP 1 5HIGHTUBEHT 1 5HIGHTUBEHT 3 2IE-LOOP-100 2 3IISL 1 2ISGTR 1 5ISGTRCB 1 5ISLOCA 1 2NCONBYOPS 6 3NO-SSHR 2 2NO-SSHR-POSTLOOP 2 2NOEFFDEPRESS-1 4 5NOEFFDEPRESS-2 4 5NOEFFDEPRESS-2 5 2
NOEFFFORCEHT 3 3NOEFFNPMP 6 3NOEFFPMP 6 1NONCONBYOPS 3 3NOOPSDEPRES2 4 4NOPDSNOISL 1 1NOPDSNOSGTR 1 3NOPDSPZRPORV 4 3NOPDSRCEQSG 4 2NOPDSRCPWR 3 1NOPDSSSHR 1 4NOPDSSSHR 2 1NOPRIMFAILNPMP 6 4NOPRIMFAILPMP 6 2NOPRIMFAILURE 4 5NOPRIMFAILURE 6 2NORECACPRI 3 2NORECOVPOWER 3 2NORECOVSSHR 1 5PDSRCPWR 6 1PO-HPICOOL 4 3PZRNOPORVDEP 4 6RCPUMPOFF 3 2RCPUMPOFF 6 4RCPUMPON 3 4RCPUMPON 6 2RECACPRI 6 2RECOVPOWER 6 2SGTRCB 1 4SSHRSGTNOCOOL 1 4SSHRUNAVAIL 1 5UNEFFNATHT 3 1
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 1 3/22/07
Early Containment Failure
EARLY
Containment Failure fromDirect Containment Heating
DCH
Pressure Load of HPME isGreater than Containment
Stregnth
DCHLOAD
Sufficient Fragmentation toCreate Significant Pressure
NODCHFRAG
Reactor Building FansCannot Handle DCH
Pressure Spike
NODCHFANSG
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRI
Page 2
Likelihood That ReactorBuilding Fans Cannot Handle
DCH Pressure Spike
DCHFANSNOEFF
Containment StregnthCannot Handle DCH Event
NODCHSTRENT
Containment StrengthCannot Handle DCH Eventand Base Pressure is High
DCHNOSTRENH
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 3
Likelihood That Cont.Strength Cannot Handle
DCH Press Spike W/ HighBase Press
NODCHSTREN1
Containment StrengthCannot Handle DCH Eventand Base Pressure is Low
DCHNOSTRENL
Page 4
Cavity Geometry AllowsEnough Corium to Disperse
For Freezing
GEOMFREEZE
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 6
Containment Failure fromCombustible Gas Burns
H2BURNS
Page 7
Containment Failure fromRapid Steam Generation
RSG
Page 31
Containment Failure fromDirect Contact of Corium
CONTACT
Page 33
Containment Failure FromMissile
MISSLE
Page 34
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 2 3/22/07
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRIPage 31Page 30Page 3
... see x-ref
PDS INDICATES THAT RBFANS ARE NOT AVAILABLE
AT OR PRIOR TO RVFAILURE
NOPDSFANS
CF
Reactor Building Fans AreNot Recovered At or Prior to
RV Failure
NORECOVFANSPRI
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 3 3/22/07
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSHPage 12Page 31Page 11
... see x-ref
PDS DOES HAVE HIGHBASE PRESSURE IN
CONTAINMENT AT RVFAILURE
PDSPRESSH
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CSS/CIS J
J001
CSS/CIS K
K001
CSS/CIS L
L001
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRI
Page 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 4 3/22/07
Containment StrengthCannot Handle DCH Eventand Base Pressure is Low
DCHNOSTRENLPage 1
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 5
Likelihood That Cont.Strength Cannot Handle
DCH Press Spike W/ LowBase Press
NODCHSTREN2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 5 3/22/07
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSLPage 31Page 11Page 8
... see x-ref
PDS DOES NOT HAVEHIGH BASE PRESSURE IN
CONTAINMENT AT RVFAILURE
NOPDSPRESSH
Page 32
Reactor Building Fans AreAvailable Prior to RV Failure
FANSAT
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRIPage 29
PDS INDICATES THAT RBFANS ARE AVAILABLE AT
OR PRIOR TO RV FAILURE
PDSFANS
CF
Reactor Building Fans AreRecovered At or Prior to RV
Failure
RECOVFANSPRI
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 6 3/22/07
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESSPage 33Page 34Page 25
... see x-ref
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOW
Page 13
Operators Do NotDepressurize Steam
Generators
OPSNOSSHR
Steam GeneratorDepressurization Or SSHR Is
Unavailable
NOPORVSSHR
PDS INDICATES OTSGADVS ARE UNAVAILABLE
NOPDSSGADV
AV
Secondary Side HeatRemoval is Unavailable
SSHRUNAVAIL
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
NO SSHR EXISTS
NO-SSHR
Page 21
Prob. that Secondary SideHeat Removal is Not
Recovered Prior to RVFailure
NORECOVSSHR
Likelihood That Operators DoNot Depressurize Steam
Generators
NOOPSDEPRESS
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORV
Page 15
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 7 3/22/07
Containment Failure fromCombustible Gas Burns
H2BURNSPage 1
Containment Failur From H2Burns Before RV Failure
NOH2PRI
Hydrogen Burns Before RVFailure
PRIBURN
Page 18
Containment StrengthCannot Handle H2 BurnEvent Prior to RV Failure
NOPRISTRENT
Page 8
Containment Failure FromComb. Gas Burns At RV
Failure
NOH2AT
Page 9
Containment Failure FromH2 Burns after RV Failure
NOH2AFTER
Page 12
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 8 3/22/07
Containment StrengthCannot Handle H2 BurnEvent Prior to RV Failure
NOPRISTRENTPage 7
Containment StrengthCannot Handle H2 Burn and
Base Pressure is High
NOPRISTRENH
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 3
Likelihood That Cont. CannotHandle H2 Burn Press W/
High Base Pressure
NOPRISTREN1
Containment StrengthCannot Handle H2 Burn and
Base Pressure is Low
NOPRISTRENL
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 5
Likelihood That Cont. CannotHandle H2 Burn Press W/
Low Base Pressure
NOPRISTREN2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 9 3/22/07
Containment Failure FromComb. Gas Burns At RV
Failure
NOH2ATPage 7
Hydrogen Burns At ReactorVessel Failure
ATBURN
H2 Concentration isSufficient to Cause Burns at
RV Failure
NOATCONC
Sufficient Hydrogen isReleased to Containment at
RV Failure
NOATRELEASE
Hydrogen Does Not BurnBefore RV Failure to Deplete
H2 Concentration
NOPRIBURN
Page 12
In-Vessel H2 Prod. Sufficientto Cause H2 Burns
NOOXIDIZED
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
Hydrogen Has Not BeenReleased to Containment
BOTTLED
Page 13
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
Ignition Source is Availableat RV Failure
NOATIGNITION
Dispersal of Corium FromCavity
DISPERSE
Cavity Geometry Does NotRetain All Corium
NOGEOMH2
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 6
Random Spark is Available atRV Failure
SPARKAT
Containment is Not SteamInerted Prior to RV Failure
NOSTMINERTP
Page 10
Containment StrengthCannot Handle H2 Burns
Event at RV Failure
NOATSTRENT
Page 11
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 10 3/22/07
Containment is Not SteamInerted Prior to RV Failure
NOSTMINERTPPage 9
Page 18
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 11 3/22/07
Containment StrengthCannot Handle H2 Burns
Event at RV Failure
NOATSTRENTPage 9
Containment StrengthCannot Handle H2 Burns and
Base Pressure is High
NOATSTRENH
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 3
Likelihood That Cont CannotHandle H2 Burn Press. W/
High Base Pressure
NOATSTREN1
Containment StrengthCannot Handle H2 Burns and
Base Pressure is Low
NOATSTRENL
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 5
Likelihood That Cont CannotHandle H2 Burn Press. W/
Low Base Pressure
NOATSTREN2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 12 3/22/07
Containment Failure FromH2 Burns after RV Failure
NOH2AFTERPage 7
Combustible Gas BurnsEarly After RV Failure
AFTBURN
Comb. Gas Concentration isSufficient to Cause H2 Burns
Early After Failure
NOAFTERCONC
Sufficient Comb. Gas isAvailable Early After RV
Failure
NOAFTERREL
Previous Burns Do NotDeplete Hydrogen in
Containment
NOPRIGLOBAL
Hydrogen Does Not BurnBefore RV Failure to Deplete
H2 Concentration
NOPRIBURNPage 9
H2 Concentration is NotSufficient to Cause Burns
Before RV Failure
NOPRICONC
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
No Sufficient Hydrogen isReleased to Containment
Before RV Failure
NOPRIRELEASE
Page 13
NO RANDOM SPARK ISAVAILABLE BEFORE RV
FAILURE
NOSPARK
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
NOSPARK-1
Page 16
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
NOSPARK-2
Page 17
Containment is SteamInerted Prior to RV Failure
STMINERTP
Page 12
Hyrdrogen Burns At RVFailure Are Prevented
NOATBURN
H2 Concentration isInsufficient to Cause Burns
at RV Failure
ATCONC
Insufficient Hydrogen isReleased to Containment at
RV Failure
ATRELEASE
Page 18
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
No Ignition Source isAvailable at RV Failure
ATIGNITION
Page 20
Containment is SteamInerted Prior to RV Failure
STMINERTPPage 12
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 3
Ex-Vessel Gas ProductionAfter RV Failure is High
EXVPRODAFTH
Page 22
Cavity Recombination DoesNot Deplete Combustible
Gas Early After RV Failure
NOAFTERRECOM
Page 25
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE
SPARKAFT
Page 28
Containment Is Not SteamInerted After RV Failure
NOSTMINERTAF
Page 29
Containment StrengthCannot Handle Comb. GasBurn Event After RV Failure
NOAFTSTRENT
Page 30
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 13 3/22/07
No Sufficient Hydrogen isReleased to Containment
Before RV Failure
NOPRIRELEASEPage 12
In-Vessel H2 Prod. NotSufficient to Cause H2 Burns
OXIDIZED
Hydrogen Has Not BeenReleased to Containment
BOTTLEDPage 9
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOWPage 6
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2
Page 14
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
NOPDSLOW-1
HIGH PRESSURE AT COREMELT IS INDETERMINATE
NOPDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
Page 20
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORV
Page 15
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 14 3/22/07
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2Page 13
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 15 3/22/07
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORVPage 13Page 6
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
PROBABILITY THAT PORVDOES NOT PREVENT
HPME
NOPRVHPCONF
LOGIC FOR OPERATORSFAILING TO OPEN PORV
NOPZPORVCONF-1
PROBABILITY THATOPERATORS FAIL TO
MANUALLY OPEN PORV
PZPORVCONF_99-C
LOGIC FOR OPERATORSFAILING TO OPEN PORVWHEN HPI COOLING IS
NOT INITIATED
NOPZPORVCONF-2
PROB THAT OPERATORSFAIL TO OPEN PORV
AFTER FAILING TO INITHPI COOLING
PZPORVCONF_0-C
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 16 3/22/07
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
NOSPARK-1Page 12
PROB THAT SPARK IS NOTAVAILABLE BEFORE RVFAILURE WITHOUT RB
SPRAY
NOSPARK_9
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAYPage 28Page 22Page 18
CS FAILURE FORINJECTION MODE
CS01
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 17 3/22/07
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
NOSPARK-2Page 12
PROB THAT SPARK IS NOTAVAILABLE BEFORE RV
FAILURE WITH RB SPRAY
NOSPARK_01
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 18 3/22/07
Insufficient Hydrogen isReleased to Containment at
RV Failure
ATRELEASEPage 12
Hydrogen Burns Before RVFailure
PRIBURNPage 7
H2 Concentration isSufficient to Cause Burns
Before RV Failure
PRICONC
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
Sufficient Hydrogen isReleased to Containment
Before RV Failure
PRIRELEASE
In-Vessel H2 Prod. Sufficientto Cause H2 Burns
NOOXIDIZED
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLED
Page 18
RANDOM SPARK ISAVAILABLE BEFORE RV
FAILURE
SPARK
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
SPARK-1
PROB THAT SPARK ISAVAILABLE BEFORE RVFAILURE WITHOUT RB
SPRAY
SPARK_1
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 16
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
SPARK-2
PROB THAT SPARK ISAVAILABLE BEFORE RV
FAILURE WITH RB SPRAY
SPARK_99
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment is Not SteamInerted Prior to RV Failure
NOSTMINERTP
Page 10
In-Vessel H2 Prod. NotSufficient to Cause H2 Burns
OXIDIZED
Recovery of Core CoolingDoes Prevent Reactor
Vessel Failure
RECOVRV
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLEDPage 18
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOW
Page 20
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 19
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 19 3/22/07
Operators Depressurize WithPressurizer PORV
PZRPORVPage 20Page 18
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
LOGIC FOR OPERATORSOPENING PORV
PZPORVCONF-1
PROB THAT OPERATORSOPEN PORV
PZPORVCONF_99
LOGIC FOR OPERATORSOPENING PORV WHEN HPI
COOLING IS NOTINITIATED
PZPORVCONF-2
Prob. That Operators OpenPORV After Failing to Init
HPI Cooling
PZPORVCONF_0
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
PROBABILITY THAT PORVCAN PREVENT HPME
PRVHPCONF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 20 3/22/07
No Ignition Source isAvailable at RV Failure
ATIGNITIONPage 12
No Dispersal of Corium FromCavity
NODISPERSE
Cavity Geometry DoesRetain All Corium
GEOMH2
Primary System Pressure isLow At RV Failure
LOWPRESSPage 33Page 22Page 22
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOWPage 18
PDS INDICATES LOWPRESSURE AT CORE MELT
IS CERTAIN
PDSLOW-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
PDSLOW-2
LOW PRESSURE AT COREMELT IS INDETERMINATE
PDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDETPage 13
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 19
Operators DepressurizeSteam Generators
OPSSSHR
Page 21
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
Random Spark isUnavailable at RV Failure
NOSPARKAT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 21 3/22/07
Operators DepressurizeSteam Generators
OPSSSHRPage 20
Steam GeneratorDepressurization and SSHR
Are Available
PORVSSHR
PDS INDICATES OTSGADVS ARE AVAILABLE
PDSSGADV
AV
Secondary Side HeatRemoval is Available
SSHRAVAIL
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHRPage 6
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
Prob. that Secondary SideHeat Removal is Recovered
Prior to RV Failure
RECOVSSHR
Likelihood That OperatorsDepressurize Steam
Generators
OPSDEPRESS
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 22 3/22/07
Ex-Vessel Gas ProductionAfter RV Failure is High
EXVPRODAFTHPage 12
Corium Pool Does NotSpread Over Large Area Or
Freeze
NOCOREFREEZE
Corium Does Not SpreadAcross Lower Containment
Or Cavity Floor
NOSPREADLOW
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 20
Cavity Geometry Does NotAllow Enough Corium toDisperse For Freezing
NOGEOMFREEZE
Likelihood Corium Does NotFreeze On Lower
Containment or Cavity Floor
NOFREEZELOW
Concrete Attack ProducesSufficient Combustible Gas
After RV Failure
NOH2SRCAFTER
Water Pool Does Not StopConcrete Attack In Cavity
After RV Failure
ATTKAFT
Water Pool In CavityUnavailable Early After RV
Failure
NOWATERAFTER
Water Does Not Fill CavityFrom Plant Specific Sources
And Paths
NOOTHERWATER
Accumulator Water isUnavailable at RV Failure
ACCUMUNAVAIL
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 20
No BWST Water GravityFeed Into Reactor CavityThrough Failed Reactor
Vessel
NOGRAVFEEDAFT
NO FAILURE OF ECCSINJECTION
NOPDSINJECCS
Page 23
PDS INDICATES HIGHBASE PRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
SEQPRESSH
Page 24
Water Unavailable FromSprays Via Fuel Transfer
Pool Early After RV Failure
NOFTRNSPOOLAFT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 16
Containment Sprays Are NotRecovered Prior to RV
Failure
NORECOVSPPRI
Containment Sprays Are NotRecovered Early After RV
Failure
NORECOVSPAFT
Likelihood That Water Pool inCavity Will Not Stop
Concrete Attack
MELT
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 23 3/22/07
NO FAILURE OF ECCSINJECTION
NOPDSINJECCSPage 22
CORE MELT BIN 2
CM-002
CORE MELT BIN 5
CM-005
CORE MELT BIN 8
CM-008
CORE MELT BIN 11
CM-011
CORE MELT BIN 3
CM-003
CORE MELT BIN 6
CM-006
CORE MELT BIN 10
CM-010
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 24 3/22/07
PDS INDICATES HIGHBASE PRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
SEQPRESSHPage 30Page 22
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CSS/CIS J
J001
CSS/CIS K
K001
CSS/CIS L
L001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 25 3/22/07
Cavity Recombination DoesNot Deplete Combustible
Gas Early After RV Failure
NOAFTERRECOMPage 12
Water Pool In CavityAvailable Early After RV
Failure
WATERAFTER
Water Does Fill Cavity FromPlant Specific Sources And
Paths
OTHERWATER
Water Available From SpraysVia Fuel Transfer Pool Early
After RV Failure
FTRNSPOOLAFT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAY
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment Sprays AreRecovered Prior to RV
Failure
RECOVSPPRI
Containment Sprays AreRecovered Early After RV
Failure
RECOVSPAFT
Accumulator Water isAvailable at RV Failure
ACCUMAVAIL
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 6
BWST Water Gravity FeedInto Reactor Cavity Through
Failed Reactor Vessel
GRAVFEEDAFT
FAILURE OF ECCSINJECTION
PDSINJECCS
Page 26
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSH
Page 27
Likelihood ThatRecombination Cannot
Deplete Comb. Gas Given aDry Cavity
NODRYEFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 26 3/22/07
FAILURE OF ECCSINJECTION
PDSINJECCSPage 25
CORE MELT BIN 1
CM-001
CORE MELT BIN 4
CM-004
CORE MELT BIN 7
CM-007
CORE MELT BIN 9
CM-009
CORE MELT BIN 12
CM-012
CORE MELT BIN 15
CM-015
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 27 3/22/07
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSHPage 25Page 29
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS G
G001
CSS/CIS H
H001
CSS/CIS I
I001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 28 3/22/07
RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE
SPARKAFTPage 12
RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE WITH RBSPRAY
SPARKAFT-1
PROB THAT SPARK ISAVAILABLE EARLY AFTER
RV FAILURE WITH RBSPRAY
SPARKAFT_99
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
RANDOM SPARK ISAVAILABLE EARLY AFTERRV FAILURE WITHOUT RB
SPRAY
SPARKAFT-2
PROB THAT SPARK ISAVAILABLE EARLY AFTERRV FAILURE WITHOUT RB
SPRAY
SPARKAFT_1
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 16
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 29 3/22/07
Containment Is Not SteamInerted After RV Failure
NOSTMINERTAFPage 12
Sequence After RV FailureHas Low Base Pressure In
Containment
AFTPRESSLPage 30
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSH
Page 27
Reactor Building FansAvailable Early After RV
Failure
FANSAFT
Reactor Building FansAvailable Early After RV
Failure
FANSAVAILAFT
Reactor Building Fans DoFunction at RV Failure
NOFANSPRI
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRI
Page 5
Likelihood RB Fans DoSurvive Containment
Enviroment At Or Prior ToRV Failu
EQUALFANSPRI
Reactor Building Fans AreRecovered Early After RV
Failure
RECOVFANSAFT
Likelihood Fans SurviveContainment Environment
Early After RV Failure
EQUALFANSAF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 30 3/22/07
Containment StrengthCannot Handle Comb. GasBurn Event After RV Failure
NOAFTSTRENTPage 12
Containment StrengthCannot Handle Comb. GasBurn and Base Pressure is
High
NOAFTSTRENH
Containment Base Pressureis High
BASEPRESSH
Containment Has High BasePressure Early After RVFailure Without Steam
Inerting
NOINERTAF
Sequence After RV FailureHas High Pressure In
Containment
AFTPRESSH
PDS INDICATES HIGHBASE PRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
SEQPRESSH
Page 24
Reactor Building Fans DoNot Function Early After RV
Failure
NOFANSAFT
Reactor Building FansUnavailable Early After RV
Failure
FANSUNAVAILAFT
Reactor Building Fans DoNot Function at RV Failure
FANSPRI
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRI
Page 2
Likelihood RB Fans Do NotSurvive Containment
Enviroment At Or Prior ToRV Failu
NOEQUALFANSPRI
Reactor Building Fans AreNot Recovered Early After
RV Failure
NORECOVFANSAFT
Likelihood Fans Do NotSurvive Containment
Environment Early After RVFailure
NOEQUALFANSAF
Likelihood That Cont CannotHandle Comb. Gas Burn
Press. W/ High BasePressure
NOAFTSTREN1
Containment StrengthCannot Handle Comb. GasBurn and Base Pressure is
Low
NOAFTSTRENL
Containment Base Pressureis Low
BASEPRESSL
Containment Has Low BasePressure Early After RVFailure Without Steam
Inerting
INERTAF
Sequence After RV FailureHas Low Base Pressure In
Containment
AFTPRESSL
Page 29
Likelihood That Cont CannotHandle Comb. Gas Burn
Press. W/ Low BasePressure
NOAFTSTREN2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 31 3/22/07
Containment Failure fromRapid Steam Generation
RSGPage 1
Rapid Steam GenerationOccurs
RSGOCCUR
Heat Transfer Rate FromCorium To Water Pool is
Fast
FASTHTRATE
Likelihood That WaterReaches Cavity Prior to RV
Failure
WATERCAV
INDICATION WHETHERSPRAYS ARE AVAILABLE
IN INJECTION MODE
PDSINJSP
CS FAILURE FORINJECTION MODE
CS01
Containment Sprays AreRecovered Prior to RV
Failure
RECOVSPPRI
Reactor Building FansCannot Handle Steam
Production
NORSGFANSG
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRI
Page 2
Likelihood That ReactorBuilding Fans CannotHandle Rapid Steam
Production
NORSGFANSEFF
Containment StrengthCannot Handle Rapid Steam
Generation Event
NORSGTRENT
Containment StrengthCannot Handle RSG Eventand Base Pressure Is High
RSGNOSTRENH
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 3
Likelihood That ContStrength Cannot Handle
RSG Press. Spike W/ HighBase Press.
NORSGSTREN1
Containment StrengthCannot Handle RSG Eventand Base Pressure Is Low
RSGNOSTRENL
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 5
Likelihood That ContStrength Cannot Handle
RSG Press. Spike W/ LowBase Press.
NORSGSTREN2
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 32 3/22/07
PDS DOES NOT HAVE HIGHBASE PRESSURE IN
CONTAINMENT AT RVFAILURE
NOPDSPRESSHPage 5
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS G
G001
CSS/CIS H
H001
CSS/CIS I
I001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 33 3/22/07
Containment Failure fromDirect Contact of Corium
CONTACTPage 1
Sufficient Amount of CoriumCan Make Contact With
Containment Wall
COREWALL
Sufficient Amount of CoriumCan Make Contact WithContainment Wall With
LPME
COREWALLLP
Plant Config and LayoutDoes Not Limit Material
Reaching Cont. Wall WithLPM
CWNOLIMITLPME
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 20
Sufficient Amount of CoriumCan Make Contact WithContainment Wall With
HPME
COREWALLHP
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 6
Plant Config and LayoutDoes Not Limit Material
Reaching Cont. Wall WithHPM
CWNOLIMITHPME
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
Containment Wall Does NotSurvive Contact With Corium
WALLNOSURVIV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 34 3/22/07
Containment Failure FromMissile
MISSLEPage 1
Alpha Mode Failure ofContainment Exists
ALPHA
RV Rocket
RVROCKET
Containment Failure FromPressure Generated
Missile(s)
PGENMISSL
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 6
Likelihood That Cont FailureIs Not Prevented Given a
Pressure Generated Missle
MISSLELIKE
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 35 3/22/07
Name Page Zone Name Page Zone
A001 27 1A001 32 1ACCUMAVAIL 25 4ACCUMUNAVAIL 22 4AFTBURN 12 7AFTPRESSH 30 2AFTPRESSL 29 2AFTPRESSL 30 4ALPHA 34 1ATBURN 9 3ATCONC 12 6ATIGNITION 12 7ATIGNITION 20 3ATPRESSH 1 3ATPRESSH 3 2ATPRESSH 8 1ATPRESSH 11 1ATPRESSH 12 8ATPRESSH 31 5ATPRESSL 4 1ATPRESSL 5 2ATPRESSL 8 3ATPRESSL 10 1ATPRESSL 11 3ATPRESSL 31 7ATRELEASE 12 5ATRELEASE 18 6ATTKAFT 22 6AV 6 1AV 21 1B001 27 1B001 32 1BASEPRESSH 30 2BASEPRESSL 30 4BOTTLED 9 2BOTTLED 13 3BWHBW1-----HP2OA 15 4BWHBW1-----HP2OA 19 4C001 27 1
C001 32 1CAG0005 21 2CAG0005-R 21 2CF 2 1CF 5 2CM-001 20 1CM-001 26 1CM-002 20 2CM-002 23 1CM-003 20 2CM-003 23 1CM-004 20 4CM-004 26 1CM-005 20 5CM-005 23 1CM-006 20 4CM-006 23 1CM-007 14 1CM-007 26 1CM-008 14 1CM-008 23 1CM-009 14 1CM-009 26 2CM-010 14 1CM-010 23 2CM-011 14 1CM-011 23 1CM-012 14 1CM-012 26 2CM-013 14 2CM-013 23 2CM-014 14 2CM-014 23 2CM-015 14 2CM-015 26 2CM-016 14 2CM-016 23 2CM-018 14 2CM-018 23 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 36 3/22/07
Name Page Zone Name Page Zone
CM-019 14 2CM-019 23 2CONTACT 1 7CONTACT 33 3COREWALL 33 2COREWALLHP 33 4COREWALLLP 33 2CS01 16 2CS01 31 2CWNOLIMITHPME 33 4CWNOLIMITLPME 33 1D001 3 1D001 24 1DCH 1 4DCHFANSNOEFF 1 3DCHLOAD 1 3DCHNOSTRENH 1 4DCHNOSTRENL 1 5DCHNOSTRENL 4 2DISPERSE 9 4E001 3 1E001 24 1EARLY 1 6EQUALFANSAF 29 3EQUALFANSPRI 29 2EXVPRODAFTH 12 6EXVPRODAFTH 22 4F001 3 1F001 24 1FANSAFT 29 2FANSAT 5 2FANSAVAILAFT 29 2FANSAVAILPRI 5 2FANSAVAILPRI 29 1FANSPRI 30 2FANSUNAVAILAFT 30 2FANSUNAVAILPRI 1 2FANSUNAVAILPRI 2 2FANSUNAVAILPRI 3 3
FANSUNAVAILPRI 30 1FANSUNAVAILPRI 31 3FASTHTRATE 31 1FTRNSPOOLAFT 25 2G001 27 1G001 32 1GEOMFREEZE 1 5GEOMH2 20 1GRAVFEEDAFT 25 6H001 27 1H001 32 1H2BURNS 1 5H2BURNS 7 2HEATIML 18 7HEATIML 20 5HIGHPRESS 1 5HIGHPRESS 6 3HIGHPRESS 9 4HIGHPRESS 25 4HIGHPRESS 33 3HIGHPRESS 34 3I001 27 1I001 32 1IE-LOOP-100 21 3INERTAF 30 3J001 3 2J001 24 2K001 3 2K001 24 2L001 3 2L001 24 2LOWCONCBURN 12 1LOWCONCBURN 12 6LOWPRESS 20 4LOWPRESS 22 1LOWPRESS 22 4LOWPRESS 33 2M001 27 2M001 32 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 37 3/22/07
Name Page Zone Name Page Zone
MELT 22 6MISSLE 1 8MISSLE 34 2MISSLELIKE 34 3N001 27 2N001 32 2NO-SSHR 6 2NO-SSHR 21 2NO-SSHR-POSTLOOP 21 3NOAFTERCONC 12 6NOAFTERRECOM 12 6NOAFTERRECOM 25 4NOAFTERREL 12 5NOAFTSTREN1 30 3NOAFTSTREN2 30 5NOAFTSTRENH 30 2NOAFTSTRENL 30 4NOAFTSTRENT 12 8NOAFTSTRENT 30 3NOATBURN 12 6NOATCONC 9 2NOATIGNITION 9 4NOATRELEASE 9 2NOATSTREN1 11 2NOATSTREN2 11 4NOATSTRENH 11 2NOATSTRENL 11 4NOATSTRENT 9 4NOATSTRENT 11 2NOCOREFREEZE 22 2NODCHFANSG 1 2NODCHFRAG 1 1NODCHSTREN1 1 4NODCHSTREN2 4 2NODCHSTRENT 1 4NODISPERSE 20 3NODRYEFF 25 4NOEQUALFANSAF 30 3NOEQUALFANSPRI 30 2
NOFANSAFT 30 3NOFANSPRI 29 2NOFREEZELOW 22 3NOFTRNSPOOLAFT 22 8NOGEOMFREEZE 22 2NOGEOMH2 9 3NOGRAVFEEDAFT 22 6NOH2AFTER 7 4NOH2AFTER 12 8NOH2AT 7 3NOH2AT 9 4NOH2PRI 7 2NOH2SRCAFTER 22 3NOHEATIML 6 4NOHEATIML 13 3NOINERTAF 30 1NOLOWCONCBURN 9 3NOLOWCONCBURN 12 7NOLOWCONCBURN 18 1NOOPSDEPRESS 6 3NOOTHERWATER 22 3NOOXIDIZED 9 1NOOXIDIZED 18 2NOPDSFANS 2 1NOPDSINJECCS 22 5NOPDSINJECCS 23 2NOPDSLOW 6 1NOPDSLOW 13 2NOPDSLOW-1 13 2NOPDSLOW-2 13 1NOPDSLOW-2 14 2NOPDSLOW_5 13 2NOPDSPRESSH 5 1NOPDSPRESSH 32 2NOPDSPZRPORV 15 1NOPDSSGADV 6 1NOPDSSPRAY 16 2NOPDSSPRAY 18 5NOPDSSPRAY 22 7
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EARLY" (2004Rev. 2)
Page 38 3/22/07
Name Page Zone Name Page Zone
NOPDSSPRAY 28 4NOPDSSSHR 6 2NOPORVSSHR 6 2NOPRIBURN 9 1NOPRIBURN 12 3NOPRICONC 12 2NOPRIGLOBAL 12 5NOPRIRELEASE 12 2NOPRIRELEASE 13 2NOPRISTREN1 8 2NOPRISTREN2 8 4NOPRISTRENH 8 2NOPRISTRENL 8 4NOPRISTRENT 7 2NOPRISTRENT 8 2NOPRVHPCONF 15 2NOPZPORVCONF-1 15 3NOPZPORVCONF-2 15 4NOPZRPORV 6 3NOPZRPORV 13 4NOPZRPORV 15 2NOPZRSAFETY 6 5NOPZRSAFETY 13 5NORECOVFANSAFT 30 3NORECOVFANSPRI 2 2NORECOVRV 1 4NORECOVRV 9 2NORECOVRV 22 7NORECOVRV 31 7NORECOVRV 33 3NORECOVRV 34 4NORECOVSPAFT 22 8NORECOVSPPRI 22 8NORECOVSSHR 6 3NORSGFANSEFF 31 4NORSGFANSG 31 4NORSGSTREN1 31 6NORSGSTREN2 31 8NORSGTRENT 31 6
NOSEQPRESSH 25 6NOSEQPRESSH 27 2NOSEQPRESSH 29 1NOSPARK 12 4NOSPARK-1 12 3NOSPARK-1 16 2NOSPARK-2 12 4NOSPARK-2 17 2NOSPARKAT 20 4NOSPARK_01 17 1NOSPARK_9 16 1NOSPREADLOW 22 2NOSTMINERTAF 12 8NOSTMINERTAF 29 2NOSTMINERTP 9 5NOSTMINERTP 10 1NOSTMINERTP 18 6NOTBOTTLED 18 3NOTBOTTLED 18 7NOWATERAFTER 22 5O001 27 2O001 32 2OPSDEPRESS 21 3OPSNOSSHR 6 2OPSSSHR 20 4OPSSSHR 21 2OTHERWATER 25 1OXIDIZED 13 1OXIDIZED 18 5P001 27 2P001 32 2PDSFANS 5 2PDSINDET 13 3PDSINDET 20 4PDSINJECCS 25 5PDSINJECCS 26 2PDSINJSP 31 2PDSLOW 18 7PDSLOW 20 2
Name Page Zone Name Page Zone
PDSLOW-1 20 2PDSLOW-2 20 4PDSLOW_5 20 3PDSPRESSH 3 2PDSPZRPORV 19 1PDSSGADV 21 1PDSSPRAY 25 1PDSSSHR 21 2PGENMISSL 34 3PO-HPICOOL 15 1PO-HPICOOL 19 1PORVSSHR 21 2PRIBURN 7 1PRIBURN 18 4PRICONC 18 2PRIRELEASE 18 2PRVHPCONF 19 4PZPORVCONF-1 19 3PZPORVCONF-2 19 3PZPORVCONF_0 19 3PZPORVCONF_0-C 15 3PZPORVCONF_99 19 2PZPORVCONF_99-C 15 3PZRPORV 18 8PZRPORV 19 2PZRPORV 20 3PZRSAFETY 18 8PZRSAFETY 20 6Q001 27 2Q001 32 2R001 27 2R001 32 2RBSPRAY 17 2RBSPRAY 18 7RBSPRAY 25 1RBSPRAY 28 2RECOVFANSAFT 29 3RECOVFANSPRI 5 3RECOVRV 18 6
RECOVSPAFT 25 3RECOVSPPRI 25 2RECOVSPPRI 31 3RECOVSSHR 21 3RSG 1 6RSG 31 4RSGNOSTRENH 31 5RSGNOSTRENL 31 7RSGOCCUR 31 2RVROCKET 34 2SEQPRESSH 22 6SEQPRESSH 24 2SEQPRESSH 30 1SPARK 18 5SPARK-1 18 4SPARK-2 18 6SPARKAFT 12 7SPARKAFT 28 2SPARKAFT-1 28 2SPARKAFT-2 28 4SPARKAFT_1 28 3SPARKAFT_99 28 1SPARKAT 9 5SPARK_1 18 4SPARK_99 18 6SSHRAVAIL 21 3SSHRUNAVAIL 6 3STMINERTP 12 5STMINERTP 12 8WALLNOSURVIV 33 4WATERAFTER 25 3WATERCAV 31 2YES-SSHR 21 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EXRELEASE"(2004 Rev. 2)
Page 1 3/22/2007
Ex-Vessel Release of FPs
EXRELEASE
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
Corium Pool Does Not SpreadOver Large Area Or Freeze
NOCOREFREEZE
Corium Does Not SpreadAcross Lower Containment
Or Cavity Floor
NOSPREADLOW
Primary System Pressure isLow At RV Failure
LOWPRESS
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOW
PDS INDICATES LOWPRESSURE AT CORE MELT
IS CERTAIN
PDSLOW-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
PDSLOW-2
LOW PRESSURE AT COREMELT IS INDETERMINATE
PDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 2
Operators DepressurizeSteam Generators
OPSSSHR
Page 3
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
Cavity Geometry Does NotAllow Enough Corium toDisperse For Freezing
NOGEOMFREEZE
Likelihood Corium Does NotFreeze On Lower
Containment or Cavity Floor
NOFREEZELOW
Ex-Vessel Release of FPs toCont. Atmos. or Water Pool is
Unavailable
NORVORPOOL
Ex-Vessel Release of FPs toCont. Atmos. from the Cavity
RVFAILS
Page 4
Water Unavailable In ReactorCavity Prior to Ex-VesselFission Product Release
NOEXFISWATER
Page 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EXRELEASE"(2004 Rev. 2)
Page 2 3/22/2007
Operators Depressurize WithPressurizer PORV
PZRPORVPage 1
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
LOGIC FOR OPERATORSOPENING PORV
PZPORVCONF-1
PROB THAT OPERATORSOPEN PORV
PZPORVCONF_99
LOGIC FOR OPERATORSOPENING PORV WHEN HPI
COOLING IS NOTINITIATED
PZPORVCONF-2
Prob. That Operators OpenPORV After Failing to Init
HPI Cooling
PZPORVCONF_0
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
PROBABILITY THAT PORVCAN PREVENT HPME
PRVHPCONF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EXRELEASE"(2004 Rev. 2)
Page 3 3/22/2007
Operators DepressurizeSteam Generators
OPSSSHRPage 1
Steam GeneratorDepressurization and SSHR
Are Available
PORVSSHR
PDS INDICATES OTSGADVS ARE AVAILABLE
PDSSGADV
AV
Secondary Side HeatRemoval is Available
SSHRAVAIL
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
Prob. that Secondary SideHeat Removal is Recovered
Prior to RV Failure
RECOVSSHR
Likelihood That OperatorsDepressurize Steam
Generators
OPSDEPRESS
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EXRELEASE"(2004 Rev. 2)
Page 4 3/22/2007
Ex-Vessel Release of FPs toCont. Atmos. from the Cavity
RVFAILSPage 1
Water Pool Does Not StopConcrete Attack Prior
ATTKLT
Likelihood That Water Poolin Cavity Will Not Stop
Concrete Attack
MELT
Water Unavailable InReactor Cavity Prior to
Ex-Vessel Fission ProductRelease
NOEXFISWATER
Page 5
Likelihood That OverlyingWater Pool Will Not Scrub
FPs Released From Corium
NOEXSCRUBEFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EXRELEASE"(2004 Rev. 2)
Page 5 3/22/2007
Water Unavailable InReactor Cavity Prior to
Ex-Vessel Fission ProductRelease
NOEXFISWATERPage 4Page 1
Water Unavailable FromSprays Via Fuel Transfer
Pool Early After RV Failure
NOFTRNSPOOLAFT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
CS FAILURE FORINJECTION MODE
CS01
Containment Sprays Are NotRecovered Prior to RV
Failure
NORECOVSPPRI
Containment Sprays Are NotRecovered Early After RV
Failure
NORECOVSPAFT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "EXRELEASE"(2004 Rev. 2)
Page 6 3/22/2007
Name Page Zone Name Page Zone
ATTKLT 4 2AV 3 1BWHBW1-----HP2OA 2 4CAG0005 3 2CAG0005-R 3 2CM-001 1 1CM-002 1 2CM-003 1 2CM-004 1 4CM-005 1 5CM-006 1 4CS01 5 1EXRELEASE 1 4HEATIML 1 5IE-LOOP-100 3 3LOWPRESS 1 4MELT 4 1NO-SSHR 3 2NO-SSHR-POSTLOOP 3 3NOCOREFREEZE 1 5NOEXFISWATER 1 8NOEXFISWATER 4 2NOEXFISWATER 5 2NOEXSCRUBEFF 4 3NOFREEZELOW 1 6NOFTRNSPOOLAFT 5 2NOGEOMFREEZE 1 5NOPDSSPRAY 5 1NORECOVRV 1 1NORECOVSPAFT 5 3NORECOVSPPRI 5 2NORVORPOOL 1 7NOSPREADLOW 1 5OPSDEPRESS 3 3OPSSSHR 1 4OPSSSHR 3 2PDSINDET 1 4PDSLOW 1 2PDSLOW-1 1 2
PDSLOW-2 1 4PDSLOW_5 1 3PDSPZRPORV 2 1PDSSGADV 3 1PDSSSHR 3 2PO-HPICOOL 2 1PORVSSHR 3 2PRVHPCONF 2 4PZPORVCONF-1 2 3PZPORVCONF-2 2 3PZPORVCONF_0 2 3PZPORVCONF_99 2 2PZRPORV 1 3PZRPORV 2 2PZRSAFETY 1 6RECOVSSHR 3 3RVFAILS 1 7RVFAILS 4 2SSHRAVAIL 3 3YES-SSHR 3 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 1 3/22/2007
FPs Are Not Scrubbed PriorTo Release to the Enviroment
FPNOSCRUBBED
Fission Products Are NotScrubbed Prior to Release
From Containment
FPNOSCRUBCON
CONTAINMENT ISOLATIONFAILURE IS NOT SMALL
(FAILURE OF C)
NOTC
LARGE CONTAINMENTISOLATION FAILURE
LARGE-ISO
FP Scrubbing is Not Effectiveor Avail. for Small Contain.
Failures
NOSMALLEFF
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 15
Likelihood That Sprays WillNot Scrub FPs for Small
Containment Failure
SPRAYNOEFF
Fission Products Are NotScrubbed Outside
Containment
NOSCRUBOUT
Aux. Bldg. Scrubbing isUnavail. or FPs are NotReleased to Aux. Bldg.
AUXUNAVAIL
Fission Products Are NotReleased to the Aux. Bldg.
AUXREL
RELEASE IS NOTTHROUGH AUX BUILDING
(FAILURE OF D)
NOTD
PDS INDICATES THATRELEASE DOES NOT GO
THROUGH AUX BLDG
NOTD-1
CSS/CIS A THROUGH F
CSSCIS-A-FPage 4
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CORE MELT BINS 1THROUGH 14
CM-1-14
Page 2
CORE MELT BINREPRESENTS SGTR
CM-15-18
Page 3
Aux. Bldg. Scrubbing isUnavailable
NOAUXSCRUB
Page 4
S/G Scrubbing Is Unavail. orFP's are Not Released to S/G
SGUNAVAIL
Page 5
Likelihood That There is NoFP Scrubbing By Other
Systems Not in Aux. Bldg.
NOOTHERSCRUB
Fission Products Are NotScrubbed Prior to ReleaseFrom Intact Containment
FPNOSCRUBPRI
Page 8
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 2 3/22/2007
CORE MELT BINS 1THROUGH 14
CM-1-14Page 1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 3 3/22/2007
CORE MELT BINREPRESENTS SGTR
CM-15-18Page 1Page 8Page 4
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 4 3/22/2007
Aux. Bldg. Scrubbing isUnavailable
NOAUXSCRUBPage 1
Fission Product ReleasesAre Not Under Water in the
Aux. Bldg.
NOAUXWATER
Fission Product Plateout isNot Effective
NOAUXPLATE
RELEASE OF FISSIONPRODUCTS IS NOT IN
LOWER SECTIONS OF AUXBLDG
NORELLOC
CSS/CIS A THROUGH F
CSSCIS-A-F
Page 1
CORE MELT BINREPRESENTS SGTR
CM-15-18
Page 3
Likelihood That Plateout WillNot Scrub Fission Products
NOPLATEOUT
Likelihood That ScrubbingCapability of Fission
Products Does Not Exist
NOAUXSPRAYS
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 5 3/22/2007
S/G Scrubbing Is Unavail. orFP's are Not Released to
S/G
SGUNAVAILPage 1
FISSION PRODUCTS ARENOT RELEASED TO THE
STEAM GENERATOR
NOSGREL
Page 6
S/G Scrubbing is Unavailable
NOSGSCRUB
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
Page 7
Likelihood That Water in S/GWill Not Scrub Fission
Products
NOWATEREFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 6 3/22/2007
FISSION PRODUCTS ARENOT RELEASED TO THE
STEAM GENERATOR
NOSGRELPage 5
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 7 3/22/2007
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHRPage 5Page 8
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 8 3/22/2007
Fission Products Are NotScrubbed Prior to ReleaseFrom Intact Containment
FPNOSCRUBPRIPage 1
SGTR-Containment-BypassSequences
SGTRCB
PDS INDICATES SGTREXISTS
NOPDSNOSGTR
CORE MELT BINREPRESENTS SGTR
CM-15-18
Page 3
InducedSGTR-Containment-Bypass
Sequence
ISGTRCB
Induced SGTR
ISGTR
S/G Tube Temp(s) InduceCreep Rupture Failure
HIGHSGTTEMP
S/G Tube Temperature(s)Are High With SSHR
Available
HIGHSGTSSHRAV
Likelihood That SSHR WillNot Keep Tubes Cool
SSHRSGTNOCOOL
Secondary Side HeatRemoval is Unavailable
SSHRUNAVAIL
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
Page 7
Prob. that Secondary SideHeat Removal is Not
Recovered Prior to RVFailure
NORECOVSSHR
Heat Transfer to S/G Tubesis High
HIGHTUBEHT
Primary System NaturalCirculation Heat Transfer to
S/G Tubes is High
HIGHNATHT
Likelihood That NaturalCirculation Heat Transfer is
High
UNEFFNATHT
Reactor Coolant Pumps AreNot Running
RCPUMPOFF
Page 9
Primary System ForcedCirculation Heat Transfer to
S/G Tubes is High
HIGHFORCEHT
Likelihood That ForcedCirculation Heat Transfer is
High
NOEFFFORCEHT
Reactor Coolant Pumps AreRunning
RCPUMPON
Page 10
Primary To Secondary DeltaP Induces Creep Rupture
Failure
HIGHDELP
Page 11
Likelihood That FPs Are NotReleased to ContainmentInstead of the Enviroment
CBREL
CONTAINMENT IS NOTISOLATED (FAILURE OF B)
NOTB
Page 14
Scrubbing is Not EffectivePrior to FP Release to
Enviroment
LATEEFF
Page 15
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 9 3/22/2007
Reactor Coolant Pumps AreNot Running
RCPUMPOFFPage 8
Page 13
No Power To The ReactorCoolant Pumps (RCPs)
NORECOVPOWER
PDS INDICATES POWER ISUNAVAILABLE TO RCPS
NOPDSRCPWR
UNAVAILABILITY OFSUPPORT SYSTEMS FOR
RCPS
DPG0003
Power Is Not Recovered tothe RCPs Prior to RV Failure
NORECACPRI
Likelihood That Operators DoNot Start the Reactor Coolant
Pumps
NONCONBYOPS
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 10 3/22/2007
Reactor Coolant Pumps AreRunning
RCPUMPONPage 13Page 8
Power To The ReactorCoolant Pumps (RCPs)
RECOVPOWER
PDS INDICATES POWER ISAVAILABLE TO RCPS
PDSRCPWR
UNAVAILABILITY OFSUPPORT SYSTEMS FOR
RCPS
DPG0003
Power Is Recovered to theRCPs Prior to RV Failure
RECACPRI
Likelihood That OperatorsStart the Reactor Coolant
Pumps
NCONBYOPS
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 11 3/22/2007
Primary To Secondary DeltaP Induces Creep Rupture
Failure
HIGHDELPPage 8
PDS INDICATES RCSPRESSURE IS
NOTSLIGHTLY ABOVE ORBELOW SG PRESS
NOPDSRCEQSG
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
Operators Do Not Depress.With Pressurizer PORVsPrior to S/G Tube Failure
NOOPSDEPRES2
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
Likelihood That OperatorsFail to Open the PORV Prior
to S/G Tube Failure
NOEFFDEPRESS-1
OPERATOR FAILS TOOPEN PORV
EFFDEPRESS_99-C
CONDITIONAL PROB THATOPERATOR FAILS TO
OPEN PORV
NOEFFDEPRESS-2
Page 12
Likelihood That PressurizerPORV(s) Cannot Depress
Primary System to S/G Press
PZRNOPORVDEP
Primary System Failure DoesNot Reduce RC PressurePrior to S/G Tube Failure
NOPRIMFAILURE
Page 13
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 12 3/22/2007
CONDITIONAL PROB THATOPERATOR FAILS TO
OPEN PORV
NOEFFDEPRESS-2Page 11
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
OPERATOR FAILS TOOPEN PORV AFTER
FAILING TO INITIATE HPICOOLING
EFFDEPRESS_0-C
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 13 3/22/2007
Primary System Failure DoesNot Reduce RC PressurePrior to S/G Tube Failure
NOPRIMFAILUREPage 11
Primary System Failure DoesNot Precede S/G Tube
Failure With RCPs Running
NOPRIMFAILPMP
Conf That Primary SysFailure Does Not Precede
S/G Tube Failure W/ RCPsOn
NOEFFPMP
Reactor Coolant Pumps AreRunning
RCPUMPON
Page 10
Primary System Failure DoesNot Precede S/G TubeFailure With RCPs Off
NOPRIMFAILNPMP
Conf That Primary SysFailure Does Not Precede
S/G Tube Failure W/ RCPsOff
NOEFFNPMP
Reactor Coolant Pumps AreNot Running
RCPUMPOFF
Page 9
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 14 3/22/2007
CONTAINMENT IS NOTISOLATED (FAILURE OF B)
NOTBPage 8
SMALL CONTAINMENTISOLATION FAILURE
SMALL-ISO
LARGE CONTAINMENTISOLATION FAILURE
LARGE-ISO
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 15 3/22/2007
Scrubbing is Not EffectivePrior to FP Release to
Enviroment
LATEEFFPage 8
Sprays Are Not Effective ForScrubbing FPs Prior ToRelease to Enviroment
NOLATESPRAY
Sprays Are Unavailable PriorTo Late Containment Failure
NOHTSPRAY
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAYPage 1
CS FAILURE FORINJECTION MODE
CS01
Containment Sprays Are NotRecovered Prior to Late
Containment Failure
NOSPRECOVLT
Containment Sprays Are NotRecovered Prior to RV Failure
NORECOVSPPRI
Containment Sprays Are NotRecovered Early After RV
Failure
NORECOVSPAFT
RB SPRAY POWERSUPPLIES ARE NOT
RECOVERED PRIOR TOLATE CTMT FAILURE
NORECOVSPLT
RB SPRAY UNAVAILABLEDUE TO MECH FAILURE OR
NO OFFSITE POWER
NORECOVSPLT-1
CS FAILURE FORINJECTION MODE
(POST-LOOP RECOVERY)
CS01-R
OFFSITE POWER NOTRECOVERED WITHIN 24
HOURS
NORECOFFSITEPWR
IE-LOOP-101
Likelihood That Spray Will NotScrub FPs Prior to Release to
Environment
NOSPRAYEFFLT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 16 3/22/2007
Name Page Zone Name Page Zone
A001 1 3AUXREL 1 4AUXUNAVAIL 1 5B001 1 3BWHBW1-----HP2OA 12 1C001 1 3CAG0005 7 1CAG0005-R 7 2CBREL 8 5CM-001 2 1CM-001 6 1CM-002 2 1CM-002 6 1CM-003 2 1CM-003 6 1CM-004 2 1CM-004 6 1CM-005 2 1CM-005 6 1CM-006 2 1CM-006 6 1CM-007 2 1CM-007 6 1CM-008 2 2CM-008 6 2CM-009 2 2CM-009 6 2CM-009 11 1CM-010 2 2CM-010 6 2CM-010 11 1CM-011 2 2CM-011 6 2CM-011 11 1CM-012 2 2CM-012 6 2CM-012 11 2CM-013 2 2CM-013 6 2
CM-013 11 2CM-014 2 2CM-014 6 2CM-014 11 2CM-015 3 1CM-016 3 2CM-018 3 2CM-019 6 2CM-1-14 1 4CM-1-14 2 2CM-15-18 1 5CM-15-18 3 2CM-15-18 4 2CM-15-18 8 1CS01 15 1CS01-R 15 3CSSCIS-A-F 1 3CSSCIS-A-F 4 1D001 1 4DPG0003 9 1DPG0003 10 1E001 1 4EFFDEPRESS_0-C 12 2EFFDEPRESS_99-C 11 4F001 1 4FPNOSCRUBBED 1 4FPNOSCRUBCON 1 2FPNOSCRUBPRI 1 6FPNOSCRUBPRI 8 4HIGHDELP 8 4HIGHDELP 11 3HIGHFORCEHT 8 6HIGHNATHT 8 4HIGHSGTSSHRAV 8 2HIGHSGTTEMP 8 3HIGHTUBEHT 8 5IE-LOOP-100 7 3IE-LOOP-101 15 5ISGTR 8 4
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED" (2004 Rev. 2)
Page 17 3/22/2007
Name Page Zone Name Page Zone
ISGTRCB 8 4LARGE-ISO 1 1LARGE-ISO 14 2LATEEFF 8 5LATEEFF 15 2NCONBYOPS 10 3NO-SSHR 7 2NO-SSHR-POSTLOOP 7 2NOAUXPLATE 4 2NOAUXSCRUB 1 5NOAUXSCRUB 4 2NOAUXSPRAYS 4 3NOAUXWATER 4 1NOEFFDEPRESS-1 11 5NOEFFDEPRESS-2 11 5NOEFFDEPRESS-2 12 2NOEFFFORCEHT 8 6NOEFFNPMP 13 3NOEFFPMP 13 1NOHTSPRAY 15 2NOLATESPRAY 15 3NONCONBYOPS 9 3NOOPSDEPRES2 11 4NOOTHERSCRUB 1 7NOPDSNOSGTR 8 1NOPDSPZRPORV 11 3NOPDSRCEQSG 11 2NOPDSRCPWR 9 1NOPDSSPRAY 1 2NOPDSSPRAY 15 1NOPDSSSHR 5 2NOPDSSSHR 7 1NOPDSSSHR 8 2NOPLATEOUT 4 3NOPRIMFAILNPMP 13 4NOPRIMFAILPMP 13 2NOPRIMFAILURE 11 5NOPRIMFAILURE 13 2NORECACPRI 9 2
NORECOFFSITEPWR 15 4NORECOVPOWER 9 2NORECOVSPAFT 15 3NORECOVSPLT 15 4NORECOVSPLT-1 15 4NORECOVSPPRI 15 2NORECOVSSHR 8 3NORELLOC 4 2NOSCRUBOUT 1 5NOSGREL 5 1NOSGREL 6 2NOSGSCRUB 5 2NOSMALLEFF 1 3NOSPRAYEFFLT 15 3NOSPRECOVLT 15 3NOTB 8 4NOTB 14 2NOTC 1 1NOTD 1 4NOTD-1 1 4NOWATEREFF 5 3PDSRCPWR 10 1PO-HPICOOL 11 3PZRNOPORVDEP 11 6RCPUMPOFF 8 5RCPUMPOFF 9 2RCPUMPOFF 13 4RCPUMPON 8 7RCPUMPON 10 2RCPUMPON 13 2RECACPRI 10 2RECOVPOWER 10 2SGTRCB 8 3SGUNAVAIL 1 6SGUNAVAIL 5 2SMALL-ISO 14 1SPRAYNOEFF 1 3SSHRSGTNOCOOL 8 1SSHRUNAVAIL 8 2
Name Page Zone Name Page Zone
UNEFFNATHT 8 4
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED2" (2004 Rev. 2)
Page 1 3/22/2007
FPs Are Not Scrubbed PriorTo Release to the Enviroment
FPNOSCRUBBED2
Fission Products Are NotScrubbed Outside
Containment
NOSCRUBOUT
Aux. Bldg. Scrubbing isUnavail. or FPs are NotReleased to Aux. Bldg.
AUXUNAVAIL
Fission Products Are NotReleased to the Aux. Bldg.
AUXREL
RELEASE IS NOTTHROUGH AUX BUILDING
(FAILURE OF D)
NOTD
PDS INDICATES THATRELEASE DOES NOT GO
THROUGH AUX BLDG
NOTD-1
CSS/CIS A THROUGH F
CSSCIS-A-FPage 3
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CORE MELT BINS 1THROUGH 14
CM-1-14
Page 2
CORE MELT BINREPRESENTS SGTR
CM-15-18Page 3
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
Aux. Bldg. Scrubbing isUnavailable
NOAUXSCRUB
Page 3
S/G Scrubbing Is Unavail. orFP's are Not Released to S/G
SGUNAVAIL
Page 4
Likelihood That There is NoFP Scrubbing By Other
Systems Not in Aux. Bldg.
NOOTHERSCRUB
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED2" (2004 Rev. 2)
Page 2 3/22/2007
CORE MELT BINS 1THROUGH 14
CM-1-14Page 1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED2" (2004 Rev. 2)
Page 3 3/22/2007
Aux. Bldg. Scrubbing isUnavailable
NOAUXSCRUBPage 1
Fission Product ReleasesAre Not Under Water in the
Aux. Bldg.
NOAUXWATER
Fission Product Plateout isNot Effective
NOAUXPLATE
RELEASE OF FISSIONPRODUCTS IS NOT IN
LOWER SECTIONS OF AUXBLDG
NORELLOC
CSS/CIS A THROUGH F
CSSCIS-A-F
Page 1
CORE MELT BINREPRESENTS SGTR
CM-15-18
Page 1
Likelihood That Plateout WillNot Scrub Fission Products
NOPLATEOUT
Likelihood That ScrubbingCapability of Fission
Products Does Not Exist
NOAUXSPRAYS
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED2" (2004 Rev. 2)
Page 4 3/22/2007
S/G Scrubbing Is Unavail. orFP's are Not Released to
S/G
SGUNAVAILPage 1
FISSION PRODUCTS ARENOT RELEASED TO THE
STEAM GENERATOR
NOSGREL
Page 5
S/G Scrubbing is Unavailable
NOSGSCRUB
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
Likelihood That Water in S/GWill Not Scrub Fission
Products
NOWATEREFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED2" (2004 Rev. 2)
Page 5 3/22/2007
FISSION PRODUCTS ARENOT RELEASED TO THE
STEAM GENERATOR
NOSGRELPage 4
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPNOSCRUBBED2" (2004 Rev. 2)
Page 6 3/22/2007
Name Page Zone Name Page Zone
A001 1 1AUXREL 1 3AUXUNAVAIL 1 3B001 1 1C001 1 1CAG0005 4 1CAG0005-R 4 2CM-001 2 1CM-001 5 1CM-002 2 1CM-002 5 1CM-003 2 1CM-003 5 1CM-004 2 1CM-004 5 1CM-005 2 1CM-005 5 1CM-006 2 1CM-006 5 1CM-007 2 1CM-007 5 1CM-008 2 2CM-008 5 2CM-009 2 2CM-009 5 2CM-010 2 2CM-010 5 2CM-011 2 2CM-011 5 2CM-012 2 2CM-012 5 2CM-013 2 2CM-013 5 2CM-014 2 2CM-014 5 2CM-015 1 4CM-016 1 4CM-018 1 5CM-019 5 2
CM-1-14 1 3CM-1-14 2 2CM-15-18 1 4CM-15-18 3 2CSSCIS-A-F 1 2CSSCIS-A-F 3 1D001 1 2E001 1 2F001 1 2FPNOSCRUBBED2 1 4IE-LOOP-100 4 3NO-SSHR 4 2NO-SSHR-POSTLOOP 4 2NOAUXPLATE 3 2NOAUXSCRUB 1 4NOAUXSCRUB 3 2NOAUXSPRAYS 3 3NOAUXWATER 3 1NOOTHERSCRUB 1 5NOPDSSSHR 4 2NOPLATEOUT 3 3NORELLOC 3 2NOSCRUBOUT 1 4NOSGREL 4 1NOSGREL 5 2NOSGSCRUB 4 2NOTD 1 3NOTD-1 1 2NOWATEREFF 4 3SGUNAVAIL 1 4SGUNAVAIL 4 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 1 3/22/2007
FPs Are Scrubbed Prior toRelease to the Enviroment
FPSCRUBBED
Fission Products AreScrubbed Prior to Release
From Containment
FPSCRUBCON
CONTAINMENT ISOLATIONFAILURE IS SMALL(SUCCESS OF C)
C
SMALL CONTAINMENTISOLATION FAILURE
SMALL-ISO
FP Scrubbing is Effective orAvail. For Small Contain.
Failures
SMALLEFF
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAY
Page 13
Likelihood That Sprays WillScrub FPs for SmallContainment Failure
SPRAYEFF
Fission Products AreScrubbed Outside
Containment
SCRUBOUT
Aux Bldg. Scrubbing is Avail.or FPs are Released to Aux.
Bldg.
AUXAVAIL
Fission Products AreReleased to the Aux. Bldg.
NOAUXREL
RELEASE IS THROUGHAUX BUILDING (SUCCESS
OF D)
D
PDS INDICATES THATRELEASE DOES GO
THROUGH AUX BLDG
D-1
CSS/CIS G THROUGH R
CSSCIS-G-R
Page 2
CORE MELT BINS 1THROUGH 14
CM-1-14
Page 3
CORE MELT BIN 19
CM-019
Aux Bldg. Scrubbing isAvailable
AUXSCRUB
Fission Product Releases AreUnder Water in the Aux. Bldg.
AUXWATER
Fission Product Plateout isEffective
AUXPLATE
RELEASE OF FISSIONPRODUCTS IS IN LOWERSECTIONS OF AUX BLDG
RELLOC
CSS/CIS G THROUGH R
CSSCIS-G-R
Page 2
CORE MELT BINS NOTINVOLVING SGTR
SCENARIOS
CMB-NOSGTR
Page 4
Likelihood That Plateout WillScrub Fission Products
PLATEOUT
Likelihood That ScrubbingCapability of Fission Products
Exists
AUXSPRAYS
S/G Scrubbing is Avail. orFP's Are Released to S/G
SGAVAIL
Page 5
Likelihood That There is FPScrubbing By Other Systems
Not in Aux Bldg.
OTHERSCRUB
Fission Products AreScrubbed Prior to ReleaseFrom Intact Containment
FPSCRUBPRI
Page 6
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 2 3/22/2007
CSS/CIS G THROUGH R
CSSCIS-G-RPage 1Page 1
CSS/CIS G
G001
CSS/CIS I
I001
CSS/CIS K
K001
CSS/CIS M
M001
CSS/CIS O
O001
CSS/CIS Q
Q001
CSS/CIS H
H001
CSS/CIS J
J001
CSS/CIS L
L001
CSS/CIS N
N001
CSS/CIS P
P001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 3 3/22/2007
CORE MELT BINS 1THROUGH 14
CM-1-14Page 1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 4 3/22/2007
CORE MELT BINS NOTINVOLVING SGTR
SCENARIOS
CMB-NOSGTRPage 1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 5 3/22/2007
S/G Scrubbing is Avail. orFP's Are Released to S/G
SGAVAILPage 1
FISSION PRODUCTS ARERELEASED TO THE STEAM
GENERATOR
SGREL
CORE MELT BINREPRESENTS SGTR
CM-15-18
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
S/G Scrubbing is Available
SGSCRUB
PDS INDICATES SSHR ISAVAILABLE
PDSSSHRPage 8
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
Likelihood That Water in S/GWill Scrub Fission Products
WATEREFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 6 3/22/2007
Fission Products AreScrubbed Prior to ReleaseFrom Intact Containment
FPSCRUBPRIPage 1
SGTR-Containment-BypassSequences Are Prevented
NOSGTRCB
PDS INDICATES THATSGTR IS NOT PRESENT
PDSNOSGTR
PDSNOSGTR-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
PDSNOSGTR-2
Page 7
InducedSGTR-Containment-Bypass
Sequence is Prevented
NOISGTRCB
No Induced SGTR
NOISGTR
S/G Tube Temp(s) Are TooLow To Induce Creep
Rupture Failure
LOWSGTTEMP
S/G Tube Temperature(s)Are Low With SSHR
Available
LOWSGTSSHRAV
Likelihood That SSHR WillKeep Tubes Cool
SSHRSGTCOOL
Secondary Side HeatRemoval is Available
SSHRAVAIL
Page 8
Heat Transfer to S/G Tubesis Low
LOWTUBEHT
Page 9
Primary to Secondary DeltaP is Too Low to InduceCreep Rupture Failure
LOWDELP
Page 10
Likelihood That FPs AreReleased to ContainmentInstead of the Enviroment
NOCBREL
CONTAINMENT ISISOLATED (SUCCESS OF
B)
B
Page 12
Scrubbing is Effective Priorto FP Release to Enviroment
NOLATEEFF
Page 13
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 7 3/22/2007
PDSNOSGTR-2Page 6
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 8 3/22/2007
Secondary Side HeatRemoval is Available
SSHRAVAILPage 6
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
Page 5
Prob. that Secondary SideHeat Removal is Recovered
Prior to RV Failure
RECOVSSHR
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 9 3/22/2007
Heat Transfer to S/G Tubesis Low
LOWTUBEHTPage 6
Primary System NaturalCirculation Heat Transfer to
S/G Tubes is Low
LOWNATHT
Likelihood That NaturalCirculation Heat Transfer is
Low
EFFNATHT
Reactor Coolant Pumps AreNot Running
RCPUMPOFF
Page 10
Primary System ForcedCirculation Heat Transfer to
S/G Tubes is Low
LOWFORCEHT
Likelihood That ForcedCirculation Heat Transfer is
Low
EFFFORCEHT
Reactor Coolant Pumps AreRunning
RCPUMPON
Page 10
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 10 3/22/2007
Primary to Secondary Delta Pis Too Low to Induce Creep
Rupture Failure
LOWDELPPage 6
PDS INDICATES RCSPRESSURE IS SLIGHTLYABOVE OR BELOW SG
PRESSURE
PDSRCEQSG
Page 11
Operators Depress. withPressurizer PORVs Prior to
S/G Tube Failure
OPSDEPRES2
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
Likelihood That OperatorsOpen the PORV Prior to S/G
Tube Failure
EFFDEPRESS-1
OPERATOR MANUALLYOPENS PORV
EFFDEPRESS_99
CONDITIONAL PROB THATOPERATOR OPENS PORV
EFFDEPRESS-2
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
OPERATOR OPENS PORVAFTER FAILING TO
INITIATE HPI COOLING
EFFDEPRESS_0
Likelihood That PressurizerPORV(s) Can Depress
Primary System to S/G Press
PZRPORVDEP
Primary System FailureReduces RC Pressure Prior
to S/G Tube Failure
PRIMFAILURE
Primary System FailurePrecedes S/G Tube Failure
With RCPs Running
PRIMFAILPMP
Conf That Primary Sys FailurePrecedes S/G Tube Failure
W/ RCPs On
EFFPMP
Reactor Coolant Pumps AreRunning
RCPUMPONPage 9
Power To The ReactorCoolant Pumps (RCPs)
RECOVPOWER
PDS INDICATES POWER ISAVAILABLE TO RCPS
PDSRCPWR
UNAVAILABILITY OFSUPPORT SYSTEMS FOR
RCPS
DPG0003
Power Is Recovered to theRCPs Prior to RV Failure
RECACPRI
Likelihood That OperatorsStart the Reactor Coolant
Pumps
NCONBYOPS
Primary System FailurePrecedes S/G Tube Failure
With RCPs Off
PRIMFAILNPMP
Reactor Coolant Pumps AreNot Running
RCPUMPOFFPage 9
No Power To The ReactorCoolant Pumps (RCPs)
NORECOVPOWER
PDS INDICATES POWER ISUNAVAILABLE TO RCPS
NOPDSRCPWR
UNAVAILABILITY OFSUPPORT SYSTEMS FOR
RCPS
DPG0003
Power Is Not Recovered tothe RCPs Prior to RV Failure
NORECACPRI
Likelihood That Operators DoNot Start the Reactor Coolant
Pumps
NONCONBYOPS
Conf That Primary Sys FailurePrecedes S/G Tube Failure
W/ RCPs Off
EFFNPMP
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 11 3/22/2007
PDS INDICATES RCSPRESSURE IS SLIGHTLYABOVE OR BELOW SG
PRESSURE
PDSRCEQSGPage 10
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 12 3/22/2007
CONTAINMENT ISISOLATED (SUCCESS OF
B)
BPage 6
CONTAINMENT IS NOTISOLATED (FAILURE OF B)
NOTB
SMALL CONTAINMENTISOLATION FAILURE
SMALL-ISO
LARGE CONTAINMENTISOLATION FAILURE
LARGE-ISO
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 13 3/22/2007
Scrubbing is Effective Prior toFP Release to Enviroment
NOLATEEFFPage 6
Sprays Are Effective ForScrubbing FPs Prior toRelease to Enviroment
LATESPRAY
Sprays Are Available Prior ToLate Containment Failure
HTSPRAY
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAYPage 1
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment Sprays AreRecovered Prior to Late
Containment Failure
SPRECOVLT
Containment Sprays AreRecovered Prior to RV Failure
RECOVSPPRI
Containment Sprays AreRecovered Early After RV
Failure
RECOVSPAFT
RB SPRAY POWERSUPPLIES ARE
RECOVERED PRIOR TOLATE CTMT FAILURE
RECOVSPLT
OFFSITE POWERRECOVERED WITHIN 24
HOURS
RECOFFSITEPWR
AVAILABILITY OFCONTAINMENT SPRAYS
WITHOUT POWERDEPENDENCY
RECSPRAYLT
IE-LOOP-101
Likelihood That Sprays WillScrub FPs Prior to Release to
Environment
SPRAYEFFLT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 14 3/22/2007
Name Page Zone Name Page Zone
AUXAVAIL 1 5AUXPLATE 1 6AUXSCRUB 1 6AUXSPRAYS 1 7AUXWATER 1 5B 6 4B 12 1BWHBW1-----HP2OA 10 3C 1 1CAG0005 5 3CAG0005-R 5 3CM-001 3 1CM-001 4 1CM-001 6 1CM-001 11 1CM-002 3 1CM-002 4 1CM-002 6 1CM-002 11 1CM-003 3 1CM-003 4 1CM-003 6 1CM-003 11 1CM-004 3 1CM-004 4 1CM-004 6 1CM-004 11 1CM-005 3 1CM-005 4 1CM-005 6 2CM-005 11 1CM-006 3 1CM-006 4 1CM-006 6 2CM-006 11 1CM-007 3 1CM-007 4 1CM-007 6 2CM-007 11 2
CM-008 3 2CM-008 4 2CM-008 6 2CM-008 11 2CM-009 3 2CM-009 4 2CM-009 7 1CM-010 3 2CM-010 4 2CM-010 7 1CM-011 3 2CM-011 4 2CM-011 7 1CM-012 3 2CM-012 4 2CM-012 7 2CM-013 3 2CM-013 4 2CM-013 7 2CM-014 3 2CM-014 4 2CM-014 7 2CM-015 5 1CM-015 11 2CM-016 5 2CM-016 11 2CM-018 5 2CM-018 11 2CM-019 1 5CM-019 4 2CM-019 7 2CM-019 11 2CM-1-14 1 4CM-1-14 3 2CM-15-18 5 2CMB-NOSGTR 1 6CMB-NOSGTR 4 2CSSCIS-G-R 1 3CSSCIS-G-R 1 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED" (2004 Rev. 2)
Page 15 3/22/2007
Name Page Zone Name Page Zone
CSSCIS-G-R 2 2D 1 4D-1 1 4DPG0003 10 4DPG0003 10 6EFFDEPRESS-1 10 3EFFDEPRESS-2 10 3EFFDEPRESS_0 10 4EFFDEPRESS_99 10 2EFFFORCEHT 9 3EFFNATHT 9 1EFFNPMP 10 8EFFPMP 10 4FPSCRUBBED 1 4FPSCRUBCON 1 2FPSCRUBPRI 1 7FPSCRUBPRI 6 4G001 2 1H001 2 2HTSPRAY 13 2I001 2 1IE-LOOP-100 5 4IE-LOOP-101 13 5J001 2 2K001 2 1L001 2 2LARGE-ISO 12 2LATESPRAY 13 3LOWDELP 6 5LOWDELP 10 3LOWFORCEHT 9 4LOWNATHT 9 2LOWSGTSSHRAV 6 4LOWSGTTEMP 6 4LOWTUBEHT 6 5LOWTUBEHT 9 2M001 2 1N001 2 2NCONBYOPS 10 6
NO-SSHR 5 3NO-SSHR-POSTLOOP 5 4NOAUXREL 1 4NOCBREL 6 5NOISGTR 6 4NOISGTRCB 6 5NOLATEEFF 6 5NOLATEEFF 13 2NONCONBYOPS 10 8NOPDSRCPWR 10 6NORECACPRI 10 7NORECOVPOWER 10 7NOSGTRCB 6 3NOTB 12 2O001 2 1OPSDEPRES2 10 2OTHERSCRUB 1 7P001 2 2PDSNOSGTR 6 2PDSNOSGTR-1 6 2PDSNOSGTR-2 6 3PDSNOSGTR-2 7 2PDSPZRPORV 10 1PDSRCEQSG 10 1PDSRCEQSG 11 2PDSRCPWR 10 4PDSSPRAY 1 2PDSSPRAY 13 1PDSSSHR 5 3PDSSSHR 8 1PLATEOUT 1 7PO-HPICOOL 10 1PRIMFAILNPMP 10 8PRIMFAILPMP 10 5PRIMFAILURE 10 6PZRPORVDEP 10 4Q001 2 1R001 2 2RBSPRAY 13 1
Name Page Zone Name Page Zone
RCPUMPOFF 9 2RCPUMPOFF 10 7RCPUMPON 9 4RCPUMPON 10 5RECACPRI 10 5RECOFFSITEPWR 13 4RECOVPOWER 10 5RECOVSPAFT 13 3RECOVSPLT 13 4RECOVSPPRI 13 2RECOVSSHR 8 2RECSPRAYLT 13 4RELLOC 1 6SCRUBOUT 1 6SGAVAIL 1 6SGAVAIL 5 2SGREL 5 1SGSCRUB 5 4SMALL-ISO 1 1SMALL-ISO 12 1SMALLEFF 1 3SPRAYEFF 1 3SPRAYEFFLT 13 3SPRECOVLT 13 3SSHRAVAIL 6 4SSHRAVAIL 8 2SSHRSGTCOOL 6 3WATEREFF 5 4YES-SSHR 5 3
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED2" (2004 Rev. 2)
Page 1 3/22/2007
FPs Are Scrubbed Prior toRelease to the Enviroment
FPSCRUBBED2
Fission Products AreScrubbed Outside
Containment
SCRUBOUT
Aux Bldg. Scrubbing is Avail.or FPs are Released to Aux.
Bldg.
AUXAVAIL
Fission Products AreReleased to the Aux. Bldg.
NOAUXREL
RELEASE IS THROUGHAUX BUILDING (SUCCESS
OF D)
D
PDS INDICATES THATRELEASE DOES GO
THROUGH AUX BLDG
D-1
CSS/CIS G THROUGH R
CSSCIS-G-R
Page 2
CORE MELT BINS 1THROUGH 14
CM-1-14
Page 3
CORE MELT BIN 19
CM-019
Aux Bldg. Scrubbing isAvailable
AUXSCRUB
Fission Product Releases AreUnder Water in the Aux. Bldg.
AUXWATER
Fission Product Plateout isEffective
AUXPLATE
RELEASE OF FISSIONPRODUCTS IS IN LOWERSECTIONS OF AUX BLDG
RELLOC
CSS/CIS G THROUGH R
CSSCIS-G-R
Page 2
CORE MELT BINS NOTINVOLVING SGTR
SCENARIOS
CMB-NOSGTR
Page 4
Likelihood That Plateout WillScrub Fission Products
PLATEOUT
Likelihood That ScrubbingCapability of Fission Products
Exists
AUXSPRAYS
S/G Scrubbing is Avail. orFP's Are Released to S/G
SGAVAIL
FISSION PRODUCTS ARERELEASED TO THE STEAM
GENERATOR
SGREL
CORE MELT BINREPRESENTS SGTR
CM-15-18
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
S/G Scrubbing is Available
SGSCRUB
Page 5
Likelihood That There is FPScrubbing By Other Systems
Not in Aux Bldg.
OTHERSCRUB
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED2" (2004 Rev. 2)
Page 2 3/22/2007
CSS/CIS G THROUGH R
CSSCIS-G-RPage 1Page 1
CSS/CIS G
G001
CSS/CIS I
I001
CSS/CIS K
K001
CSS/CIS M
M001
CSS/CIS O
O001
CSS/CIS Q
Q001
CSS/CIS H
H001
CSS/CIS J
J001
CSS/CIS L
L001
CSS/CIS N
N001
CSS/CIS P
P001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED2" (2004 Rev. 2)
Page 3 3/22/2007
CORE MELT BINS 1THROUGH 14
CM-1-14Page 1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED2" (2004 Rev. 2)
Page 4 3/22/2007
CORE MELT BINS NOTINVOLVING SGTR
SCENARIOS
CMB-NOSGTRPage 1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED2" (2004 Rev. 2)
Page 5 3/22/2007
S/G Scrubbing is Available
SGSCRUBPage 1
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
Likelihood That Water in S/GWill Scrub Fission Products
WATEREFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"FPSCRUBBED2" (2004 Rev. 2)
Page 6 3/22/2007
Name Page Zone Name Page Zone
AUXAVAIL 1 3AUXPLATE 1 4AUXSCRUB 1 4AUXSPRAYS 1 5AUXWATER 1 3CAG0005 5 1CAG0005-R 5 2CM-001 3 1CM-001 4 1CM-002 3 1CM-002 4 1CM-003 3 1CM-003 4 1CM-004 3 1CM-004 4 1CM-005 3 1CM-005 4 1CM-006 3 1CM-006 4 1CM-007 3 1CM-007 4 1CM-008 3 2CM-008 4 2CM-009 3 2CM-009 4 2CM-010 3 2CM-010 4 2CM-011 3 2CM-011 4 2CM-012 3 2CM-012 4 2CM-013 3 2CM-013 4 2CM-014 3 2CM-014 4 2CM-015 1 6CM-016 1 6CM-018 1 7CM-019 1 3
CM-019 4 2CM-1-14 1 2CM-1-14 3 2CM-15-18 1 6CMB-NOSGTR 1 4CMB-NOSGTR 4 2CSSCIS-G-R 1 1CSSCIS-G-R 1 3CSSCIS-G-R 2 2D 1 2D-1 1 2FPSCRUBBED2 1 5G001 2 1H001 2 2I001 2 1IE-LOOP-100 5 3J001 2 2K001 2 1L001 2 2M001 2 1N001 2 2NO-SSHR 5 2NO-SSHR-POSTLOOP 5 2NOAUXREL 1 2O001 2 1OTHERSCRUB 1 7P001 2 2PDSSSHR 5 1PLATEOUT 1 5Q001 2 1R001 2 2RELLOC 1 4SCRUBOUT 1 5SGAVAIL 1 6SGREL 1 6SGSCRUB 1 7SGSCRUB 5 2WATEREFF 5 2YES-SSHR 5 1
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 1 3/22/2007
Late Containment Failure
LATE
Containment Failure FromLate Combustible Gases
H2LATE
Late Combustible Gas Burns
LTBURN
Combustible GasConcentration is Sufficient toCause Burns Late After RV
Failure
LTCONC
Previous Combustible GasBurns Do Not DepleteCombustible Gas in
Containment
NOLTPRIGLOB
Hydrogen Does Not BurnBefore RV Failure to Deplete
H2 Concentration
NOPRIBURNPage 15
H2 Concentration is NotSufficient to Cause Burns
Before RV Failure
NOPRICONC
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
No Sufficient Hydrogen isReleased to Containment
Before RV Failure
NOPRIRELEASE
Page 2
NO RANDOM SPARK ISAVAILABLE BEFORE RV
FAILURE
NOSPARK
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
NOSPARK-1
Page 5
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
NOSPARK-2
Page 7
Containment is SteamInerted Prior to RV Failure
STMINERTP
Page 1
Hyrdrogen Burns At RVFailure Are Prevented
NOATBURN
H2 Concentration isInsufficient to Cause Burns at
RV Failure
ATCONC
Insufficient Hydrogen isReleased to Containment at
RV Failure
ATRELEASE
Page 8
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
No Ignition Source isAvailable at RV Failure
ATIGNITION
Page 10
Containment is SteamInerted Prior to RV Failure
STMINERTPPage 1
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 13
Combustible Gas Burns EarlyAfter RV Failure are
Prevented
NOAFTBURN
Page 14
Cavity Recombination DoesNot Deplete CombustibleGas Early After RV Failure
NOAFTERRECOM
Page 28
Cavity Recombination DoesNot Deplete Comb. Gas Prior
to Late Containment Failur
NOLTRECOMB
Page 29
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
Containment is Not SteamInerted Late After RV Failure
NOLATEINERT
Page 30
Random Spark is AvailableLate After RV Failure
SPARKLT
Page 31
Containment StrengthCannot Handle Late
Combustible Gas Burn Event
LTNOSTRENT
Page 32
Containment Failure FromSteam Generation
STEAM
Page 35
Containment Failure FromNon Condensable Gases
GASES
Page 37
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 2 3/22/2007
No Sufficient Hydrogen isReleased to Containment
Before RV Failure
NOPRIRELEASEPage 1
In-Vessel H2 Prod. NotSufficient to Cause H2 Burns
OXIDIZED
Hydrogen Has Not BeenReleased to Containment
BOTTLEDPage 15
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOWPage 18
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2
Page 3
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
NOPDSLOW-1
HIGH PRESSURE AT COREMELT IS INDETERMINATE
NOPDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
Page 10
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORV
Page 4
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 3 3/22/2007
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2Page 2
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 4 3/22/2007
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORVPage 2
Page 18
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
PROBABILITY THAT PORVDOES NOT PREVENT
HPME
NOPRVHPCONF
LOGIC FOR OPERATORSFAILING TO OPEN PORV
NOPZPORVCONF-1
PROBABILITY THATOPERATORS FAIL TO
MANUALLY OPEN PORV
PZPORVCONF_99-C
LOGIC FOR OPERATORSFAILING TO OPEN PORVWHEN HPI COOLING IS
NOT INITIATED
NOPZPORVCONF-2
PROB THAT OPERATORSFAIL TO OPEN PORV
AFTER FAILING TO INITHPI COOLING
PZPORVCONF_0-C
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 5 3/22/2007
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
NOSPARK-1Page 1
PROB THAT SPARK IS NOTAVAILABLE BEFORE RVFAILURE WITHOUT RB
SPRAY
NOSPARK_9
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 6
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 6 3/22/2007
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAYPage 5
Page 37Page 25
... see x-ref
CS FAILURE FORINJECTION MODE
CS01
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 7 3/22/2007
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
NOSPARK-2Page 1
PROB THAT SPARK IS NOTAVAILABLE BEFORE RV
FAILURE WITH RB SPRAY
NOSPARK_01
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 8 3/22/2007
Insufficient Hydrogen isReleased to Containment at
RV Failure
ATRELEASEPage 1
Hydrogen Burns Before RVFailure
PRIBURNPage 14
H2 Concentration isSufficient to Cause Burns
Before RV Failure
PRICONC
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
Sufficient Hydrogen isReleased to Containment
Before RV Failure
PRIRELEASE
In-Vessel H2 Prod. Sufficientto Cause H2 Burns
NOOXIDIZED
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLED
Page 9
RANDOM SPARK ISAVAILABLE BEFORE RV
FAILURE
SPARK
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
SPARK-1
PROB THAT SPARK ISAVAILABLE BEFORE RVFAILURE WITHOUT RB
SPRAY
SPARK_1
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 6
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
SPARK-2
PROB THAT SPARK ISAVAILABLE BEFORE RV
FAILURE WITH RB SPRAY
SPARK_99
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment is Not SteamInerted Prior to RV Failure
NOSTMINERTP
Page 15
In-Vessel H2 Prod. NotSufficient to Cause H2 Burns
OXIDIZED
Recovery of Core CoolingDoes Prevent Reactor Vessel
Failure
RECOVRV
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLED
Page 9
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 9 3/22/2007
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLEDPage 8Page 8
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOW
Page 10
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 10
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 10 3/22/2007
No Ignition Source isAvailable at RV Failure
ATIGNITIONPage 1
No Dispersal of Corium FromCavity
NODISPERSE
Cavity Geometry DoesRetain All Corium
GEOMH2
Primary System Pressure isLow At RV Failure
LOWPRESSPage 22
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOWPage 9
PDS INDICATES LOWPRESSURE AT CORE MELT
IS CERTAIN
PDSLOW-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
PDSLOW-2
LOW PRESSURE AT COREMELT IS INDETERMINATE
PDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDETPage 2
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
Operators Depressurize WithPressurizer PORV
PZRPORVPage 9
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
LOGIC FOR OPERATORSOPENING PORV
PZPORVCONF-1
PROB THAT OPERATORSOPEN PORV
PZPORVCONF_99
LOGIC FOR OPERATORSOPENING PORV WHEN HPI
COOLING IS NOTINITIATED
PZPORVCONF-2
Page 11
PROBABILITY THAT PORVCAN PREVENT HPME
PRVHPCONF
Operators DepressurizeSteam Generators
OPSSSHR
Page 12
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
Random Spark isUnavailable at RV Failure
NOSPARKAT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 11 3/22/2007
LOGIC FOR OPERATORSOPENING PORV WHEN HPI
COOLING IS NOTINITIATED
PZPORVCONF-2Page 10
Prob. That Operators OpenPORV After Failing to Init
HPI Cooling
PZPORVCONF_0
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 12 3/22/2007
Operators DepressurizeSteam Generators
OPSSSHRPage 10
Steam GeneratorDepressurization and SSHR
Are Available
PORVSSHR
PDS INDICATES OTSGADVS ARE AVAILABLE
PDSSGADV
AV
Secondary Side HeatRemoval is Available
SSHRAVAIL
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHRPage 19
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
Prob. that Secondary SideHeat Removal is Recovered
Prior to RV Failure
RECOVSSHR
Likelihood That OperatorsDepressurize Steam
Generators
OPSDEPRESS
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 13 3/22/2007
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSHPage 1
PDS DOES HAVE HIGHBASE PRESSURE IN
CONTAINMENT AT RVFAILURE
PDSPRESSH
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CSS/CIS J
J001
CSS/CIS K
K001
CSS/CIS L
L001
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRI
Page 27
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 14 3/22/2007
Combustible Gas Burns EarlyAfter RV Failure are
Prevented
NOAFTBURNPage 1
Comb. Gas Concentration isInsufficient to Cause H2Burns Early After Failure
AFTERCONC
Sufficient Comb. Gas isAvailable Early After RV
Failure
AFTERREL
Previous Burns Do DepleteHydrogen in Containment
PRIGLOBAL
Hydrogen Burns Before RVFailure
PRIBURN
Page 8
Hydrogen Burns At ReactorVessel Failure
ATBURN
Page 15
Ex-Vessel Gas ProductionAfter RV Failure is Low
EXVPRODAFTL
Page 18
Cavity Recombination DoesDeplete Combustible Gas
Early After RV Failure
AFTERRECOM
Page 22
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
NO RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE
NOSPARKAFT
Page 25
Containment Is SteamInerted After RV Failure
STMINERTAF
Page 26
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 15 3/22/2007
Hydrogen Burns At ReactorVessel Failure
ATBURNPage 14
H2 Concentration isSufficient to Cause Burns at
RV Failure
NOATCONC
Sufficient Hydrogen isReleased to Containment at
RV Failure
NOATRELEASE
Hydrogen Does Not BurnBefore RV Failure to Deplete
H2 Concentration
NOPRIBURN
Page 1
In-Vessel H2 Prod. Sufficientto Cause H2 Burns
NOOXIDIZED
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
Hydrogen Has Not BeenReleased to Containment
BOTTLED
Page 2
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
Ignition Source is Availableat RV Failure
NOATIGNITION
Dispersal of Corium FromCavity
DISPERSE
Cavity Geometry Does NotRetain All Corium
NOGEOMH2
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 18
Random Spark is Available atRV Failure
SPARKAT
Containment is Not SteamInerted Prior to RV Failure
NOSTMINERTPPage 8
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 16
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 16 3/22/2007
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSLPage 15
PDS DOES NOT HAVEHIGH BASE PRESSURE IN
CONTAINMENT AT RVFAILURE
NOPDSPRESSH
Page 17
Reactor Building Fans AreAvailable Prior to RV Failure
FANSAT
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRIPage 33
PDS INDICATES THAT RBFANS ARE AVAILABLE AT
OR PRIOR TO RV FAILURE
PDSFANS
CF
Reactor Building Fans AreRecovered At or Prior to RV
Failure
RECOVFANSPRI
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 17 3/22/2007
PDS DOES NOT HAVE HIGHBASE PRESSURE IN
CONTAINMENT AT RVFAILURE
NOPDSPRESSHPage 16
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS G
G001
CSS/CIS H
H001
CSS/CIS I
I001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 18 3/22/2007
Ex-Vessel Gas ProductionAfter RV Failure is Low
EXVPRODAFTLPage 14
Corium Pool Does SpreadOver Large Area Or Freeze
COREFREEZE
Corium Does Spread AcrossLower Containment Or
Cavity Floor
SPREADLOW
Page 35
Likelihood Corium DoesFreeze On Lower
Containment or Cavity Floor
FREEZELOW
Concrete Attack ProducesInsufficient Combustible Gas
After RV Failure
H2SRCAFTER
Water Pool Does StopConcrete Attack In Cavity
After RV Failure
NOATTKAFT
Water Pool In CavityAvailable Early After RV
Failure
WATERAFTERPage 28
Water Does Fill Cavity FromPlant Specific Sources And
Paths
OTHERWATER
Water Available From SpraysVia Fuel Transfer Pool Early
After RV Failure
FTRNSPOOLAFT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAY
Page 35
Containment Sprays AreRecovered Prior to RV
Failure
RECOVSPPRI
Containment Sprays AreRecovered Early After RV
Failure
RECOVSPAFT
Accumulator Water isAvailable at RV Failure
ACCUMAVAIL
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESSPage 15Page 35
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOW
Page 2
Operators Do NotDepressurize Steam
Generators
OPSNOSSHR
Page 19
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORV
Page 4
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
BWST Water Gravity FeedInto Reactor Cavity Through
Failed Reactor Vessel
GRAVFEEDAFT
Page 20
Likelihood That Water Poolin Cavity Will Stop Concrete
Attack
NOMELT
Recovery of Core CoolingDoes Prevent Reactor
Vessel Failure
RECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 19 3/22/2007
Operators Do NotDepressurize Steam
Generators
OPSNOSSHRPage 18
Steam GeneratorDepressurization Or SSHR Is
Unavailable
NOPORVSSHR
PDS INDICATES OTSGADVS ARE UNAVAILABLE
NOPDSSGADV
AV
Secondary Side HeatRemoval is Unavailable
SSHRUNAVAIL
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
NO SSHR EXISTS
NO-SSHR
Page 12
Prob. that Secondary SideHeat Removal is Not
Recovered Prior to RVFailure
NORECOVSSHR
Likelihood That Operators DoNot Depressurize Steam
Generators
NOOPSDEPRESS
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 20 3/22/2007
BWST Water Gravity FeedInto Reactor Cavity Through
Failed Reactor Vessel
GRAVFEEDAFTPage 18
FAILURE OF ECCSINJECTION
PDSINJECCS
CORE MELT BIN 1
CM-001
CORE MELT BIN 4
CM-004
CORE MELT BIN 7
CM-007
CORE MELT BIN 9
CM-009
CORE MELT BIN 12
CM-012
CORE MELT BIN 15
CM-015
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSH
Page 21
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 21 3/22/2007
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSHPage 20
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS G
G001
CSS/CIS H
H001
CSS/CIS I
I001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 22 3/22/2007
Cavity Recombination DoesDeplete Combustible Gas
Early After RV Failure
AFTERRECOMPage 14
Water Pool In CavityUnavailable Early After RV
Failure
NOWATERAFTER
Water Does Not Fill CavityFrom Plant Specific Sources
And Paths
NOOTHERWATER
Accumulator Water isUnavailable at RV Failure
ACCUMUNAVAIL
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 10
No BWST Water GravityFeed Into Reactor CavityThrough Failed Reactor
Vessel
NOGRAVFEEDAFT
NO FAILURE OF ECCSINJECTION
NOPDSINJECCS
Page 23
PDS INDICATES HIGHBASE PRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
SEQPRESSH
Page 26
Water Unavailable FromSprays Via Fuel Transfer
Pool Early After RV Failure
NOFTRNSPOOLAFT
Page 24
Likelihood ThatRecombination Can Deplete
Comb. Gas Given a DryCavity
DRYEFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 23 3/22/2007
NO FAILURE OF ECCSINJECTION
NOPDSINJECCSPage 22
CORE MELT BIN 2
CM-002
CORE MELT BIN 5
CM-005
CORE MELT BIN 8
CM-008
CORE MELT BIN 11
CM-011
CORE MELT BIN 3
CM-003
CORE MELT BIN 6
CM-006
CORE MELT BIN 10
CM-010
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 24 3/22/2007
Water Unavailable FromSprays Via Fuel Transfer
Pool Early After RV Failure
NOFTRNSPOOLAFTPage 22
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 6
Containment Sprays Are NotRecovered Prior to RV
Failure
NORECOVSPPRI
Containment Sprays Are NotRecovered Early After RV
Failure
NORECOVSPAFT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 25 3/22/2007
NO RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE
NOSPARKAFTPage 14
NO RANDOM SPARKEARLY AFTER RV FAILUREFOR RB SPRAY AVAILABLE
NOSPARKAFT-1
PROB THAT SPARK ISUNAVAILABLE EARLY
AFTER RV FAILURE WITHRB SPRAY
NOSPARKAFT_01
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
NO RANDOM SPARKEARLY AFTER RV FAILURE
FOR RB SPRAYUNAVAILABLE
NOSPARKAFT-2
PROB THAT SPARK ISUNAVAILABLE EARLYAFTER RV FAILURE
WITHOUT RB SPRAY
NOSPARKAFT_9
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 6
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 26 3/22/2007
Containment Is SteamInerted After RV Failure
STMINERTAFPage 14
Sequence After RV FailureHas High Pressure In
Containment
AFTPRESSH
PDS INDICATES HIGHBASE PRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
SEQPRESSHPage 22
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CSS/CIS J
J001
CSS/CIS K
K001
CSS/CIS L
L001
Reactor Building Fans DoNot Function Early After RV
Failure
NOFANSAFT
Page 27
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 27 3/22/2007
Reactor Building Fans DoNot Function Early After RV
Failure
NOFANSAFTPage 26Page 35
Reactor Building FansUnavailable Early After RV
Failure
FANSUNAVAILAFT
Reactor Building Fans DoNot Function at RV Failure
FANSPRI
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRIPage 13
PDS INDICATES THAT RBFANS ARE NOT AVAILABLE
AT OR PRIOR TO RVFAILURE
NOPDSFANS
CF
Reactor Building Fans AreNot Recovered At or Prior to
RV Failure
NORECOVFANSPRI
Likelihood RB Fans Do NotSurvive Containment
Enviroment At Or Prior ToRV Failu
NOEQUALFANSPRI
Reactor Building Fans AreNot Recovered Early After
RV Failure
NORECOVFANSAFT
Likelihood Fans Do NotSurvive Containment
Environment Early After RVFailure
NOEQUALFANSAF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 28 3/22/2007
Cavity Recombination DoesNot Deplete Combustible
Gas Early After RV Failure
NOAFTERRECOMPage 1
Water Pool In CavityAvailable Early After RV
Failure
WATERAFTER
Page 18
Likelihood ThatRecombination Cannot
Deplete Comb. Gas Given aDry Cavity
NODRYEFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 29 3/22/2007
Cavity Recombination DoesNot Deplete Comb. Gas Prior
to Late Containment Failur
NOLTRECOMBPage 1
Likelihood ThatRecombination Cannot
Deplete Comb. Gas With aDry Cavity Late
NODRYEFFLT
Water Is Available in CavityArea
STMWATER
Page 35
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 30 3/22/2007
Containment is Not SteamInerted Late After RV Failure
NOLATEINERTPage 1
Sequence Late After RVFailure Has Low BasePressure From Steam
LTPRESSL
Page 32
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 31 3/22/2007
Random Spark is AvailableLate After RV Failure
SPARKLTPage 1
SPARK AVAILABLE;OFFSITE POWER
RECOVERED
SPARKLT-1
RANDOM SPARKAVAILABLE WITH OFFSITE
POWER
SPARKLT-OP
OFFSITE POWERRECOVERED WITHIN 24
HOURS
RECOFFSITEPWR
SPARK AVAILABLE;OFFSITE POWER NOT
RECOVERED
SPARKLT-2
RANDOM SPARKAVAILABLE WITHOUT
OFFSITE POWER
SPARKLT-NOP
OFFSITE POWER NOTRECOVERED WITHIN 24
HOURS
NORECOFFSITEPWR
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 32 3/22/2007
Containment StrengthCannot Handle Late
Combustible Gas Burn Event
LTNOSTRENTPage 1
Containment StrengthCannot Handle Late Comb.
Gas Burn and Base Pressureis High
LTSTRENH
Sequence Late After RVFailure Has High Base
Pressure From GasGeneration
NOINERTLT
Likelihood That Cont CannotHandle Comb. Gas Burn
Press. W/ High BasePressure
NOSTREN1H2
Containment StrengthCannot Handle Late Comb.
Gas Burn and Base Pressureis Low
LTSTRENL
Sequence Late After RVFailure Has Low BasePressure From Steam
LTPRESSLPage 30
Reactor Building FansFunction Prior to LateContainment Failure
FANSLT
Reactor Building Fans AreAvailable Prior to LateContainment Failure
FANSAVAILLT
Reactor Building FansAvailable Early After RV
Failure
FANSAFT
Reactor Building FansAvailable Early After RV
Failure
FANSAVAILAFT
Reactor Building Fans DoFunction at RV Failure
NOFANSPRI
Page 33
Reactor Building Fans AreRecovered Early After RV
Failure
RECOVFANSAFT
Likelihood Fans SurviveContainment Environment
Early After RV Failure
EQUALFANSAF
RB FAN POWER SUPPLIESARE RECOVERED PRIORTO LATE CTMT FAILURE
RECOVFANSLT
OFFSITE POWERRECOVERED WITHIN 24
HOURS
RECOFFSITEPWR
AVAILABILITY OF RB FANSWITHOUT POWER
DEPENDENCY
RECFANSLT
Likelihood That RB FansSurvive Containment
Environment to Prevent LCF
EQUALFANSLT
PDS INDICATES LOWBASE PRESSURE INCONTAINMENT LATEAFTER RV FAILURE
SEQPRESSL
Page 34
Likelihood That Cont CannotHandle Comb. Gas Burn
Press. W/ Low BasePressure
NOSTREN2H2
Sequence Late After RVFailure Has Low Base
Pressure From GasGeneration
INERTLT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 33 3/22/2007
Reactor Building Fans DoFunction at RV Failure
NOFANSPRIPage 32
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRI
Page 16
Likelihood RB Fans DoSurvive Containment
Enviroment At Or Prior ToRV Failu
EQUALFANSPRI
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 34 3/22/2007
PDS INDICATES LOW BASEPRESSURE IN
CONTAINMENT LATEAFTER RV FAILURE
SEQPRESSLPage 32
CSS/CIS A
A001
CSS/CIS D
D001
CSS/CIS G
G001
CSS/CIS J
J001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 35 3/22/2007
Containment Failure FromSteam Generation
STEAMPage 1
There is Sufficient SteamProduced to Pressurize
Containment
STMPROD
Sufficient Steam ProducedFrom Lower Containment
Area
LOWSTM
Corium Does Spread AcrossLower Containment Or Cavity
Floor
SPREADLOWPage 18
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 18
Cavity Geometry AllowsEnough Corium to Disperse
For Freezing
GEOMFREEZE
Water Is Available in CavityArea
STMWATERPage 29
Water Available fromContainment Sprays Via Fuel
Transfer Pool Prior to LCF
FTRNSPOOLLT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAYPage 18
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment Sprays AreRecovered Prior to RV Failure
RECOVSPPRI
Containment Sprays AreRecovered Early After RV
Failure
RECOVSPAFT
RB SPRAY POWERSUPPLIES ARE
RECOVERED PRIOR TOLATE CTMT FAILURE
RECOVSPLT
OFFSITE POWERRECOVERED WITHIN 24
HOURS
RECOFFSITEPWR
AVAILABILITY OFCONTAINMENT SPRAYS
WITHOUT POWERDEPENDENCY
RECSPRAYLT
IE-LOOP-101
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
Reactor Building Fans AreUnavailable Prior to Late
Containment Failure
FANSUNAVAILLT
Reactor Building Fans Do NotFunction Early After RV
Failure
NOFANSAFT
Page 27
RB FAN POWER SUPPLIESARE NOT RECOVEREDPRIOR TO LATE CTMT
FAILURE
NORECOVFANSLT
Page 36
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 36 3/22/2007
RB FAN POWER SUPPLIESARE NOT RECOVEREDPRIOR TO LATE CTMT
FAILURE
NORECOVFANSLTPage 35
RB FANS UNAVAILABLEFOLLOWING POST-LOOP
RECOVERY
CF-R
OFFSITE POWER NOTRECOVERED WITHIN 24
HOURS
NORECOFFSITEPWR
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 37 3/22/2007
Containment Failure FromNon Condensable Gases
GASESPage 1
Water is Unavailable inCavity Prior to LCF
NOSTMWATER
Water Unavailable fromContainment Sprays Via Fuel
Transfer Pool Prior to LCF
NOFTRNSPOOLLT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 6
Containment Sprays Are NotRecovered Prior to RV
Failure
NORECOVSPPRI
RB SPRAY POWERSUPPLIES ARE NOT
RECOVERED PRIOR TOLATE CTMT FAILURE
NORECOVSPLT
RB SPRAY UNAVAILABLEDUE TO MECH FAILUREOR NO OFFSITE POWER
NORECOVSPLT-1
CS FAILURE FORINJECTION MODE
(POST-LOOP RECOVERY)
CS01-R
OFFSITE POWER NOTRECOVERED WITHIN 24
HOURS
NORECOFFSITEPWR
IE-LOOP-101
Containment Sprays Are NotRecovered Early After RV
Failure
NORECOVSPAFT
Likelihood That ContainmentCannot Handle Pressurefrom Non-Condensable
Gases
NCGASES
Likelihood That NonCondensable Gas Production
is High Given a Dry Cavity
NCGASHIGH
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 38 3/22/2007
Name Page Zone Name Page Zone
A001 17 1A001 21 1A001 34 1ACCUMAVAIL 18 4ACCUMUNAVAIL 22 2AFTERCONC 14 3AFTERRECOM 14 3AFTERRECOM 22 3AFTERREL 14 2AFTPRESSH 26 2ATBURN 14 2ATBURN 15 4ATCONC 1 6ATIGNITION 1 7ATIGNITION 10 4ATPRESSH 1 8ATPRESSH 13 2ATPRESSL 15 6ATPRESSL 16 2ATRELEASE 1 5ATRELEASE 8 5AV 12 1AV 19 1B001 17 1B001 21 1BOTTLED 2 3BOTTLED 15 2BWHBW1-----HP2OA 4 4BWHBW1-----HP2OA 11 2C001 17 1C001 21 1CAG0005 12 2CAG0005-R 12 2CF 16 2CF 27 1CF-R 36 1CM-001 10 1CM-001 20 1CM-002 10 2
CM-002 23 1CM-003 10 2CM-003 23 1CM-004 10 4CM-004 20 1CM-005 10 5CM-005 23 1CM-006 10 4CM-006 23 1CM-007 3 1CM-007 20 1CM-008 3 1CM-008 23 1CM-009 3 1CM-009 20 2CM-010 3 1CM-010 23 2CM-011 3 1CM-011 23 1CM-012 3 1CM-012 20 2CM-013 3 2CM-013 23 2CM-014 3 2CM-014 23 2CM-015 3 2CM-015 20 2CM-016 3 2CM-016 23 2CM-018 3 2CM-018 23 2CM-019 3 2CM-019 23 2COREFREEZE 18 2CS01 6 1CS01-R 37 2D001 13 1D001 26 1D001 34 1
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 39 3/22/2007
Name Page Zone Name Page Zone
DISPERSE 15 4DRYEFF 22 4E001 13 1E001 26 1EQUALFANSAF 32 3EQUALFANSLT 32 4EQUALFANSPRI 33 2EXVPRODAFTL 14 3EXVPRODAFTL 18 3F001 13 1F001 26 1FANSAFT 32 2FANSAT 16 2FANSAVAILAFT 32 2FANSAVAILLT 32 3FANSAVAILPRI 16 2FANSAVAILPRI 33 1FANSLT 32 3FANSPRI 27 2FANSUNAVAILAFT 27 2FANSUNAVAILLT 35 6FANSUNAVAILPRI 13 3FANSUNAVAILPRI 27 2FREEZELOW 18 2FTRNSPOOLAFT 18 2FTRNSPOOLLT 35 5G001 17 1G001 21 1G001 34 1GASES 1 10GASES 37 3GEOMFREEZE 35 2GEOMH2 10 1GRAVFEEDAFT 18 5GRAVFEEDAFT 20 2H001 17 1H001 21 1H2LATE 1 8H2SRCAFTER 18 3
HEATIML 9 1HEATIML 10 8HIGHPRESS 15 4HIGHPRESS 18 4HIGHPRESS 35 1I001 17 1I001 21 1IE-LOOP-100 12 3IE-LOOP-101 35 7IE-LOOP-101 37 4INERTLT 32 6J001 13 2J001 26 2J001 34 1K001 13 2K001 26 2L001 13 2L001 26 2LATE 1 9LOWCONCBURN 1 1LOWCONCBURN 1 6LOWCONCBURN 14 4LOWPRESS 10 6LOWPRESS 22 2LOWSTM 35 1LTBURN 1 8LTCONC 1 7LTNOSTRENT 1 9LTNOSTRENT 32 3LTPRESSL 30 1LTPRESSL 32 4LTSTRENH 32 2LTSTRENL 32 5M001 17 2M001 21 2M001 34 1N001 17 2N001 21 2N001 34 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 40 3/22/2007
Name Page Zone Name Page Zone
NCGASES 37 3NCGASHIGH 37 4NO-SSHR 12 2NO-SSHR 19 2NO-SSHR-POSTLOOP 12 3NOAFTBURN 1 7NOAFTBURN 14 4NOAFTERRECOM 1 6NOAFTERRECOM 28 2NOATBURN 1 6NOATCONC 15 2NOATIGNITION 15 4NOATRELEASE 15 2NOATTKAFT 18 4NODISPERSE 10 3NODRYEFF 28 2NODRYEFFLT 29 1NOEQUALFANSAF 27 3NOEQUALFANSPRI 27 3NOFANSAFT 26 3NOFANSAFT 27 3NOFANSAFT 35 5NOFANSPRI 32 1NOFANSPRI 33 2NOFTRNSPOOLAFT 22 5NOFTRNSPOOLAFT 24 2NOFTRNSPOOLLT 37 3NOGEOMH2 15 3NOGRAVFEEDAFT 22 4NOHEATIML 2 3NOHEATIML 18 4NOINERTLT 32 1NOLATEINERT 1 8NOLATEINERT 30 1NOLOWCONCBURN 1 8NOLOWCONCBURN 8 1NOLOWCONCBURN 15 3NOLTPRIGLOB 1 5NOLTRECOMB 1 7
NOLTRECOMB 29 2NOMELT 18 4NOOPSDEPRESS 19 3NOOTHERWATER 22 1NOOXIDIZED 8 2NOOXIDIZED 15 1NOPDSFANS 27 1NOPDSINJECCS 22 3NOPDSINJECCS 23 2NOPDSLOW 2 2NOPDSLOW 18 3NOPDSLOW-1 2 2NOPDSLOW-2 2 1NOPDSLOW-2 3 2NOPDSLOW_5 2 2NOPDSPRESSH 16 1NOPDSPRESSH 17 2NOPDSPZRPORV 4 1NOPDSSGADV 19 1NOPDSSPRAY 5 2NOPDSSPRAY 6 1NOPDSSPRAY 8 5NOPDSSPRAY 24 1NOPDSSPRAY 25 4NOPDSSPRAY 37 1NOPDSSSHR 19 2NOPORVSSHR 19 2NOPRIBURN 1 3NOPRIBURN 15 1NOPRICONC 1 2NOPRIRELEASE 1 2NOPRIRELEASE 2 2NOPRVHPCONF 4 2NOPZPORVCONF-1 4 3NOPZPORVCONF-2 4 4NOPZRPORV 2 4NOPZRPORV 4 2NOPZRPORV 18 4NOPZRSAFETY 2 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATE" (2004Rev. 2)
Page 41 3/22/2007
Name Page Zone Name Page Zone
NOPZRSAFETY 18 4NORECOFFSITEPWR 31 4NORECOFFSITEPWR 36 2NORECOFFSITEPWR 37 3NORECOVFANSAFT 27 3NORECOVFANSLT 35 6NORECOVFANSLT 36 2NORECOVFANSPRI 27 2NORECOVRV 15 2NORECOVRV 35 4NORECOVSPAFT 24 2NORECOVSPAFT 37 4NORECOVSPLT 37 3NORECOVSPLT-1 37 3NORECOVSPPRI 24 2NORECOVSPPRI 37 2NORECOVSSHR 19 3NOSEQPRESSH 20 3NOSEQPRESSH 21 2NOSPARK 1 4NOSPARK-1 1 3NOSPARK-1 5 2NOSPARK-2 1 4NOSPARK-2 7 2NOSPARKAFT 14 4NOSPARKAFT 25 2NOSPARKAFT-1 25 2NOSPARKAFT-2 25 4NOSPARKAFT_01 25 1NOSPARKAFT_9 25 3NOSPARKAT 10 4NOSPARK_01 7 1NOSPARK_9 5 1NOSTMINERTP 8 6NOSTMINERTP 15 5NOSTMWATER 37 2NOSTREN1H2 32 2NOSTREN2H2 32 5NOTBOTTLED 8 3
NOTBOTTLED 8 7NOTBOTTLED 9 2NOWATERAFTER 22 3O001 17 2O001 21 2O001 34 2OPSDEPRESS 12 3OPSNOSSHR 18 3OPSNOSSHR 19 2OPSSSHR 10 7OPSSSHR 12 2OTHERWATER 18 1OXIDIZED 2 1OXIDIZED 8 5P001 17 2P001 21 2P001 34 2PDSFANS 16 2PDSINDET 2 3PDSINDET 10 4PDSINJECCS 20 2PDSLOW 9 1PDSLOW 10 2PDSLOW-1 10 2PDSLOW-2 10 4PDSLOW_5 10 3PDSPRESSH 13 2PDSPZRPORV 10 5PDSSGADV 12 1PDSSPRAY 18 2PDSSPRAY 35 3PDSSSHR 12 2PO-HPICOOL 4 1PO-HPICOOL 10 5PORVSSHR 12 2PRIBURN 8 4PRIBURN 14 1PRICONC 8 2PRIGLOBAL 14 2
Name Page Zone Name Page Zone
PRIRELEASE 8 2PRVHPCONF 10 8PZPORVCONF-1 10 7PZPORVCONF-2 10 7PZPORVCONF-2 11 2PZPORVCONF_0 11 1PZPORVCONF_0-C 4 3PZPORVCONF_99 10 6PZPORVCONF_99-C 4 3PZRPORV 9 2PZRPORV 10 6PZRSAFETY 9 2PZRSAFETY 10 9Q001 17 2Q001 21 2Q001 34 2R001 17 2R001 21 2R001 34 2RBSPRAY 7 2RBSPRAY 8 7RBSPRAY 25 2RBSPRAY 35 3RECFANSLT 32 5RECOFFSITEPWR 31 2RECOFFSITEPWR 32 4RECOFFSITEPWR 35 6RECOVFANSAFT 32 2RECOVFANSLT 32 4RECOVFANSPRI 16 3RECOVRV 8 6RECOVRV 18 5RECOVSPAFT 18 3RECOVSPAFT 35 5RECOVSPLT 35 6RECOVSPPRI 18 2RECOVSPPRI 35 4RECOVSSHR 12 3RECSPRAYLT 35 6
SEQPRESSH 22 4SEQPRESSH 26 2SEQPRESSL 32 4SEQPRESSL 34 2SPARK 8 5SPARK-1 8 4SPARK-2 8 6SPARKAT 15 5SPARKLT 1 9SPARKLT 31 2SPARKLT-1 31 2SPARKLT-2 31 4SPARKLT-NOP 31 3SPARKLT-OP 31 1SPARK_1 8 4SPARK_99 8 6SPREADLOW 18 1SPREADLOW 35 2SSHRAVAIL 12 3SSHRUNAVAIL 19 3STEAM 1 9STEAM 35 4STMINERTAF 14 5STMINERTAF 26 2STMINERTP 1 5STMINERTP 1 8STMPROD 35 3STMWATER 29 2STMWATER 35 4WATERAFTER 18 3WATERAFTER 28 1YES-SSHR 12 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATEREVAP"(2004 Rev. 2)
Page 1 3/22/2007
Late Fission ProductRevaporization Release is
Prevented
LATEREVAP
Insufficient Fission ProductHeating to Cause
Vaporization
NOFPHEATING
Heat Losses From PrimarySystem Are Very Large
HEATLOSS
Amount of Fission ProductsRetained in Primary System
is Small
FPAMOUNT
Primary Retention Is Not Lowfor Low Pressure Core Melt
AMTLOWP
Likelihood That Retention isLow for a Low Pressure Core
Melt
NOLPCMEFF
Primary System Pressure isLow Prior to Core Melt
LOWPRESPCM
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOW
PDS INDICATES LOWPRESSURE AT CORE MELT
IS CERTAIN
PDSLOW-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
PDSLOW-2
LOW PRESSURE AT COREMELT IS INDETERMINATE
PDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
Page 2
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 3
Primary Retention Is Not Lowfor High Pressure Core Melt
AMTHIGHP
Primary System Pressure isHigh Prior to Core Melt
HIGHPRESPCM
Page 4
Likelihood That Retention isNot Low for a High Pressure
Core Melt
NOHPCMEFF
Chemical Form of FissionProducts Has HighVaporization Temp
CHEMICAL
Secondary Side HeatRemoval Prevents
Revaporization
SSHR
Page 6
Recovery of Core CoolingDoes Prevent Reactor Vessel
Failure
RECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATEREVAP"(2004 Rev. 2)
Page 2 3/22/2007
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDETPage 1Page 4
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATEREVAP"(2004 Rev. 2)
Page 3 3/22/2007
Operators Depressurize WithPressurizer PORV
PZRPORVPage 1
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
LOGIC FOR OPERATORSOPENING PORV
PZPORVCONF-1
PROB THAT OPERATORSOPEN PORV
PZPORVCONF_99
LOGIC FOR OPERATORSOPENING PORV WHEN HPI
COOLING IS NOTINITIATED
PZPORVCONF-2
Prob. That Operators OpenPORV After Failing to Init
HPI Cooling
PZPORVCONF_0
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
PROBABILITY THAT PORVCAN PREVENT HPME
PRVHPCONF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATEREVAP"(2004 Rev. 2)
Page 4 3/22/2007
Primary System Pressure isHigh Prior to Core Melt
HIGHPRESPCMPage 1
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOW
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2
Page 5
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
NOPDSLOW-1
HIGH PRESSURE AT COREMELT IS INDETERMINATE
NOPDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
Page 2
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORV
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
PROBABILITY THAT PORVDOES NOT PREVENT
HPME
NOPRVHPCONF
LOGIC FOR OPERATORSFAILING TO OPEN PORV
NOPZPORVCONF-1
PROBABILITY THATOPERATORS FAIL TO
MANUALLY OPEN PORV
PZPORVCONF_99-C
LOGIC FOR OPERATORSFAILING TO OPEN PORVWHEN HPI COOLING IS
NOT INITIATED
NOPZPORVCONF-2
PROB THAT OPERATORSFAIL TO OPEN PORV
AFTER FAILING TO INITHPI COOLING
PZPORVCONF_0-C
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATEREVAP"(2004 Rev. 2)
Page 5 3/22/2007
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2Page 4
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATEREVAP"(2004 Rev. 2)
Page 6 3/22/2007
Secondary Side HeatRemoval Prevents
Revaporization
SSHRPage 1
Secondary Side HeatRemoval is Available Prior to
Revaporization
SSHRRVPAVL
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
Secondary Side HeatRemoval is Recovered Prior
to Revaporization
SSHRRVPREC
Likelihood That SSHR WillPrevent Revaporization
SSHRREVAP
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "LATEREVAP"(2004 Rev. 2)
Page 7 3/22/2007
Name Page Zone Name Page Zone
AMTHIGHP 1 4AMTLOWP 1 2BWHBW1-----HP2OA 3 4BWHBW1-----HP2OA 4 7CAG0005 6 1CAG0005-R 6 2CHEMICAL 1 3CM-001 1 1CM-002 1 2CM-003 1 2CM-004 2 1CM-005 2 2CM-006 2 2CM-007 5 1CM-008 5 1CM-009 5 1CM-010 5 1CM-011 5 1CM-012 5 1CM-013 5 2CM-014 5 2CM-015 5 2CM-016 5 2CM-018 5 2CM-019 5 2FPAMOUNT 1 3HEATLOSS 1 1HIGHPRESPCM 1 3HIGHPRESPCM 4 3IE-LOOP-100 6 3LATEREVAP 1 3LOWPRESPCM 1 2NO-SSHR 6 2NO-SSHR-POSTLOOP 6 2NOFPHEATING 1 2NOHPCMEFF 1 4NOLPCMEFF 1 1NOPDSLOW 4 2NOPDSLOW-1 4 3
NOPDSLOW-2 4 2NOPDSLOW-2 5 2NOPDSLOW_5 4 2NOPDSPZRPORV 4 4NOPRVHPCONF 4 5NOPZPORVCONF-1 4 6NOPZPORVCONF-2 4 7NOPZRPORV 4 5NOPZRSAFETY 4 1PDSINDET 1 4PDSINDET 2 2PDSINDET 4 3PDSLOW 1 2PDSLOW-1 1 2PDSLOW-2 1 4PDSLOW_5 1 3PDSPZRPORV 3 1PDSSSHR 6 1PO-HPICOOL 3 1PO-HPICOOL 4 4PRVHPCONF 3 4PZPORVCONF-1 3 3PZPORVCONF-2 3 3PZPORVCONF_0 3 3PZPORVCONF_0-C 4 6PZPORVCONF_99 3 2PZPORVCONF_99-C 4 6PZRPORV 1 3PZRPORV 3 2PZRSAFETY 1 1RECOVRV 1 5SSHR 1 4SSHR 6 2SSHRREVAP 6 3SSHRRVPAVL 6 2SSHRRVPREC 6 2YES-SSHR 6 1
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOBASEMELT" (2004 Rev. 2)
Page 1 3/22/2007
No Containment Failure FromBasemat Melt-Through
NOBASEMELT
Corium Pool Does SpreadOver Large Area Or Freeze
COREFREEZE
Corium Does Spread AcrossLower Containment Or Cavity
Floor
SPREADLOW
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOW
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2
Page 2
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
NOPDSLOW-1
HIGH PRESSURE AT COREMELT IS INDETERMINATE
NOPDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
Operators Do NotDepressurize Steam
Generators
OPSNOSSHR
Steam GeneratorDepressurization Or SSHR Is
Unavailable
NOPORVSSHR
PDS INDICATES OTSGADVS ARE UNAVAILABLE
NOPDSSGADV
AV
Secondary Side HeatRemoval is Unavailable
SSHRUNAVAIL
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
Page 3
Prob. that Secondary SideHeat Removal is Not
Recovered Prior to RV Failure
NORECOVSSHR
Likelihood That Operators DoNot Depressurize Steam
Generators
NOOPSDEPRESS
Operators Do NotDepressurize with Pressurizer
PORV
NOPZRPORV
Page 4
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
Cavity Geometry AllowsEnough Corium to Disperse
For Freezing
GEOMFREEZE
Likelihood Corium DoesFreeze On Lower
Containment or Cavity Floor
FREEZELOW
Water Pool Stops ConcreteAttack Prior to Basemat
Melt-Through
NOATTACK
Page 5
Recovery of Core CoolingDoes Prevent Reactor Vessel
Failure
RECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOBASEMELT" (2004 Rev. 2)
Page 2 3/22/2007
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2Page 1
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOBASEMELT" (2004 Rev. 2)
Page 3 3/22/2007
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHRPage 1
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOBASEMELT" (2004 Rev. 2)
Page 4 3/22/2007
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORVPage 1
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
PROBABILITY THAT PORVDOES NOT PREVENT
HPME
NOPRVHPCONF
LOGIC FOR OPERATORSFAILING TO OPEN PORV
NOPZPORVCONF-1
PROBABILITY THATOPERATORS FAIL TO
MANUALLY OPEN PORV
PZPORVCONF_99-C
LOGIC FOR OPERATORSFAILING TO OPEN PORVWHEN HPI COOLING IS
NOT INITIATED
NOPZPORVCONF-2
PROB THAT OPERATORSFAIL TO OPEN PORV
AFTER FAILING TO INITHPI COOLING
PZPORVCONF_0-C
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOBASEMELT" (2004 Rev. 2)
Page 5 3/22/2007
Water Pool Stops ConcreteAttack Prior to Basemat
Melt-Through
NOATTACKPage 1
Water is Available in CavityPrior to Basemat
Melt-Through
BMMWATER
Water Available fromContainment Sprays Via Fuel
Transfer Pool Prior to LCF
FTRNSPOOLLT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAY
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment Sprays AreRecovered Prior to RV
Failure
RECOVSPPRI
Containment Sprays AreRecovered Early After RV
Failure
RECOVSPAFT
RB SPRAY POWERSUPPLIES ARE
RECOVERED PRIOR TOLATE CTMT FAILURE
RECOVSPLT
OFFSITE POWERRECOVERED WITHIN 24
HOURS
RECOFFSITEPWR
AVAILABILITY OFCONTAINMENT SPRAYS
WITHOUT POWERDEPENDENCY
RECSPRAYLT
IE-LOOP-101
Likelihood That Water Poolin Cavity Will Stop Concrete
Attack
NOMELT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOBASEMELT" (2004 Rev. 2)
Page 6 3/22/2007
Name Page Zone Name Page Zone
AV 1 4BMMWATER 5 2BWHBW1-----HP2OA 4 4CAG0005 3 1CAG0005-R 3 2CM-004 1 2CM-005 1 3CM-006 1 3CM-007 2 1CM-008 2 1CM-009 2 1CM-010 2 1CM-011 2 1CM-012 2 1CM-013 2 2CM-014 2 2CM-015 2 2CM-016 2 2CM-018 2 2CM-019 2 2COREFREEZE 1 6FREEZELOW 1 6FTRNSPOOLLT 5 3GEOMFREEZE 1 6HIGHPRESS 1 5IE-LOOP-100 3 3IE-LOOP-101 5 5NO-SSHR 3 2NO-SSHR-POSTLOOP 3 2NOATTACK 1 7NOATTACK 5 3NOBASEMELT 1 7NOHEATIML 1 7NOMELT 5 3NOOPSDEPRESS 1 6NOPDSLOW 1 2NOPDSLOW-1 1 2NOPDSLOW-2 1 1NOPDSLOW-2 2 2
NOPDSLOW_5 1 2NOPDSPZRPORV 4 1NOPDSSGADV 1 4NOPDSSSHR 1 5NOPDSSSHR 3 1NOPORVSSHR 1 5NOPRVHPCONF 4 2NOPZPORVCONF-1 4 3NOPZPORVCONF-2 4 4NOPZRPORV 1 6NOPZRPORV 4 2NOPZRSAFETY 1 8NORECOVSSHR 1 6OPSNOSSHR 1 5PDSINDET 1 3PDSSPRAY 5 1PO-HPICOOL 4 1PZPORVCONF_0-C 4 3PZPORVCONF_99-C 4 3RBSPRAY 5 1RECOFFSITEPWR 5 4RECOVRV 1 8RECOVSPAFT 5 3RECOVSPLT 5 4RECOVSPPRI 5 2RECSPRAYLT 5 4SPREADLOW 1 5SSHRUNAVAIL 1 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 1 3/22/2007
No Containment Bypass
NOBYPASS
No Interfacing System LOCA(IS-LOCA)
NOISLOCA
PDS INDICATES THATISLOCA IS NOT PRESENT
PDSNOISL
PDSNOISL-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
PDSNOISL-2
Page 2
CONFIDENCE THAT ANINDUCED ISLOCA IS
PREVENTED
NOIISL
SGTR-Containment-BypassSequences Are Prevented
NOSGTRCB
PDS INDICATES THATSGTR IS NOT PRESENT
PDSNOSGTR
PDSNOSGTR-1
Page 3
PDSNOSGTR-2
Page 4
InducedSGTR-Containment-Bypass
Sequence is Prevented
NOISGTRCB
Page 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 2 3/22/2007
PDSNOISL-2Page 1
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 3 3/22/2007
PDSNOSGTR-1Page 1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 4 3/22/2007
PDSNOSGTR-2Page 1
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 5 3/22/2007
InducedSGTR-Containment-Bypass
Sequence is Prevented
NOISGTRCBPage 1
No Induced SGTR
NOISGTR
S/G Tube Temp(s) Are TooLow To Induce Creep
Rupture Failure
LOWSGTTEMP
S/G Tube Temperature(s)Are Low With SSHR
Available
LOWSGTSSHRAV
Likelihood That SSHR WillKeep Tubes Cool
SSHRSGTCOOL
Secondary Side HeatRemoval is Available
SSHRAVAIL
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHR
Page 6
Prob. that Secondary SideHeat Removal is Recovered
Prior to RV Failure
RECOVSSHR
Heat Transfer to S/G Tubesis Low
LOWTUBEHT
Primary System NaturalCirculation Heat Transfer to
S/G Tubes is Low
LOWNATHT
Likelihood That NaturalCirculation Heat Transfer is
Low
EFFNATHT
Reactor Coolant Pumps AreNot Running
RCPUMPOFF
Page 7
Primary System ForcedCirculation Heat Transfer to
S/G Tubes is Low
LOWFORCEHT
Likelihood That ForcedCirculation Heat Transfer is
Low
EFFFORCEHT
Reactor Coolant Pumps AreRunning
RCPUMPON
Page 7
Primary to Secondary DeltaP is Too Low to InduceCreep Rupture Failure
LOWDELP
Page 7
Likelihood That FPs AreReleased to ContainmentInstead of the Enviroment
NOCBREL
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 6 3/22/2007
NO SSHR EXISTS
NO-SSHRPage 5
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 7 3/22/2007
Primary to Secondary Delta Pis Too Low to Induce Creep
Rupture Failure
LOWDELPPage 5
PDS INDICATES RCSPRESSURE IS SLIGHTLYABOVE OR BELOW SG
PRESSURE
PDSRCEQSG
Page 8
Operators Depress. withPressurizer PORVs Prior to
S/G Tube Failure
OPSDEPRES2
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
Likelihood That OperatorsOpen the PORV Prior to S/G
Tube Failure
EFFDEPRESS-1
OPERATOR MANUALLYOPENS PORV
EFFDEPRESS_99
CONDITIONAL PROB THATOPERATOR OPENS PORV
EFFDEPRESS-2
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
OPERATOR OPENS PORVAFTER FAILING TO
INITIATE HPI COOLING
EFFDEPRESS_0
Likelihood That PressurizerPORV(s) Can Depress
Primary System to S/G Press
PZRPORVDEP
Primary System FailureReduces RC Pressure Prior
to S/G Tube Failure
PRIMFAILURE
Primary System FailurePrecedes S/G Tube Failure
With RCPs Running
PRIMFAILPMP
Conf That Primary SysFailure Precedes S/G Tube
Failure W/ RCPs On
EFFPMP
Reactor Coolant Pumps AreRunning
RCPUMPONPage 5
Power To The ReactorCoolant Pumps (RCPs)
RECOVPOWER
PDS INDICATES POWER ISAVAILABLE TO RCPS
PDSRCPWR
Page 9
Power Is Recovered to theRCPs Prior to RV Failure
RECACPRI
Likelihood That OperatorsStart the Reactor Coolant
Pumps
NCONBYOPS
Primary System FailurePrecedes S/G Tube Failure
With RCPs Off
PRIMFAILNPMP
Reactor Coolant Pumps AreNot Running
RCPUMPOFFPage 5
No Power To The ReactorCoolant Pumps (RCPs)
NORECOVPOWER
PDS INDICATES POWER ISUNAVAILABLE TO RCPS
NOPDSRCPWR
Page 10
Power Is Not Recovered tothe RCPs Prior to RV Failure
NORECACPRI
Likelihood That Operators DoNot Start the Reactor Coolant
Pumps
NONCONBYOPS
Conf That Primary SysFailure Precedes S/G Tube
Failure W/ RCPs Off
EFFNPMP
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 8 3/22/2007
PDS INDICATES RCSPRESSURE IS SLIGHTLYABOVE OR BELOW SG
PRESSURE
PDSRCEQSGPage 7
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
CORE MELT BIN 4
CM-004
CORE MELT BIN 5
CM-005
CORE MELT BIN 6
CM-006
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 9 3/22/2007
PDS INDICATES POWER ISAVAILABLE TO RCPS
PDSRCPWRPage 7
UNAVAILABILITY OFSUPPORT SYSTEMS FOR
RCPS
DPG0003
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 10 3/22/2007
PDS INDICATES POWER ISUNAVAILABLE TO RCPS
NOPDSRCPWRPage 7
UNAVAILABILITY OFSUPPORT SYSTEMS FOR
RCPS
DPG0003
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOBYPASS"(2004 Rev. 2)
Page 11 3/22/2007
Name Page Zone Name Page Zone
BWHBW1-----HP2OA 7 3CAG0005 6 1CAG0005-R 6 2CM-001 1 1CM-001 3 1CM-001 8 1CM-002 1 1CM-002 3 1CM-002 8 1CM-003 1 1CM-003 3 1CM-003 8 1CM-004 1 1CM-004 3 1CM-004 8 1CM-005 1 2CM-005 3 2CM-005 8 1CM-006 1 2CM-006 3 2CM-006 8 1CM-007 1 2CM-007 3 2CM-007 8 2CM-008 1 2CM-008 3 2CM-008 8 2CM-009 2 1CM-009 4 1CM-010 2 1CM-010 4 1CM-011 2 1CM-011 4 1CM-012 2 1CM-012 4 2CM-013 2 2CM-013 4 2CM-014 2 2CM-014 4 2
CM-015 2 2CM-015 8 2CM-016 2 2CM-016 8 2CM-018 2 2CM-018 8 2CM-019 4 2CM-019 8 2DPG0003 9 1DPG0003 10 1EFFDEPRESS-1 7 3EFFDEPRESS-2 7 3EFFDEPRESS_0 7 4EFFDEPRESS_99 7 2EFFFORCEHT 5 6EFFNATHT 5 4EFFNPMP 7 8EFFPMP 7 4IE-LOOP-100 6 3LOWDELP 5 4LOWDELP 7 3LOWFORCEHT 5 6LOWNATHT 5 4LOWSGTSSHRAV 5 2LOWSGTTEMP 5 3LOWTUBEHT 5 5NCONBYOPS 7 6NO-SSHR 5 2NO-SSHR 6 2NO-SSHR-POSTLOOP 6 2NOBYPASS 1 3NOCBREL 5 5NOIISL 1 3NOISGTR 5 4NOISGTRCB 1 5NOISGTRCB 5 4NOISLOCA 1 2NONCONBYOPS 7 8NOPDSRCPWR 7 6
Name Page Zone Name Page Zone
NOPDSRCPWR 10 1NORECACPRI 7 7NORECOVPOWER 7 7NOSGTRCB 1 4OPSDEPRES2 7 2PDSNOISL 1 2PDSNOISL-1 1 2PDSNOISL-2 1 3PDSNOISL-2 2 2PDSNOSGTR 1 4PDSNOSGTR-1 1 4PDSNOSGTR-1 3 2PDSNOSGTR-2 1 5PDSNOSGTR-2 4 2PDSPZRPORV 7 1PDSRCEQSG 7 1PDSRCEQSG 8 2PDSRCPWR 7 4PDSRCPWR 9 1PDSSSHR 5 2PO-HPICOOL 7 1PRIMFAILNPMP 7 8PRIMFAILPMP 7 5PRIMFAILURE 7 6PZRPORVDEP 7 4RCPUMPOFF 5 5RCPUMPOFF 7 7RCPUMPON 5 7RCPUMPON 7 5RECACPRI 7 5RECOVPOWER 7 5RECOVSSHR 5 3SSHRAVAIL 5 2SSHRSGTCOOL 5 1YES-SSHR 5 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 1 3/22/2007
Early Containment Failure isPrevented
NOEARLY
Containment Failure FromDirect Containment Heating
is Prevented
NODCH
Pressure Load of HPME isLess than Containment
Stregnth
NODCHLOAD
Insufficient Fragmentation toCreate Significant Pressure
DCHFRAG
Reactor Building Fans CanHandle DCH Pressure Spike
DCHFANSG
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRI
Page 2
Likelihood That ReactorBuilding Fans Can Handle
DCH Pressure Spike
DCHFANSEFF
Containment Stregnth CanHandle DCH Event
DCHSTRENT
Containment Strength CanHandle DCH Event and Base
Pressure is High
DCHSTRENH
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 32
Likelihood That Cont.Strength Can Handle DCHPress Spike W/ High Base
Press
DCHSTREN1
Containment Strength CanHandle DCH Event and Base
Pressure is Low
DCHSTRENL
Page 3
Cavity Geometry Does NotAllow Enough Corium toDisperse For Freezing
NOGEOMFREEZE
Recovery of Core CoolingDoes Prevent Reactor
Vessel Failure
RECOVRV
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 5
Containment Failure FromCombustible Gas Burns is
Prevented
NOH2BURNS
Page 6
Containment Failure FromRapid Steam Generation is
Prevented
NORSG
Page 32
Containment Failure FromDirect Contact of Corium is
Prevented
NOCONTACT
Page 33
Containment Failure FromMissiles is Prevented
NOMISSLE
Page 36
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 2 3/22/2007
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRIPage 32Page 1Page 4
... see x-ref
PDS INDICATES THAT RBFANS ARE AVAILABLE AT
OR PRIOR TO RV FAILURE
PDSFANS
CF
Reactor Building Fans AreRecovered At or Prior to RV
Failure
RECOVFANSPRI
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 3 3/22/2007
Containment Strength CanHandle DCH Event and Base
Pressure is Low
DCHSTRENLPage 1
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 4
Likelihood That Cont.Strength Can Handle DCHPress Spike W/ Low Base
Press
DCHSTREN2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 4 3/22/2007
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSLPage 16Page 32Page 11
... see x-ref
PDS DOES NOT HAVEHIGH BASE PRESSURE IN
CONTAINMENT AT RVFAILURE
NOPDSPRESSH
Page 17
Reactor Building Fans AreAvailable Prior to RV Failure
FANSAT
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRI
Page 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 5 3/22/2007
Primary System Pressure isLow At RV Failure
LOWPRESSPage 33Page 36Page 1
... see x-ref
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOWPage 13
PDS INDICATES LOWPRESSURE AT CORE MELT
IS CERTAIN
PDSLOW-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
PDSLOW-2
LOW PRESSURE AT COREMELT IS INDETERMINATE
PDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDETPage 6
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 14
Operators DepressurizeSteam Generators
OPSSSHR
Steam GeneratorDepressurization and SSHR
Are Available
PORVSSHR
PDS INDICATES OTSGADVS ARE AVAILABLE
PDSSGADV
AV
Secondary Side HeatRemoval is Available
SSHRAVAIL
Page 34
Likelihood That OperatorsDepressurize Steam
Generators
OPSDEPRESS
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 6 3/22/2007
Containment Failure FromCombustible Gas Burns is
Prevented
NOH2BURNSPage 1
Containment Failure From H2Burns Before RV Failure is
Prevented
H2PRI
Hydrogen Does Not BurnBefore RV Failure to Deplete
H2 Concentration
NOPRIBURNPage 19
H2 Concentration is NotSufficient to Cause Burns
Before RV Failure
NOPRICONC
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
No Sufficient Hydrogen isReleased to Containment
Before RV Failure
NOPRIRELEASE
In-Vessel H2 Prod. NotSufficient to Cause H2 Burns
OXIDIZED
Hydrogen Has Not BeenReleased to Containment
BOTTLEDPage 19
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOWPage 21
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2
Page 7
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
NOPDSLOW-1
HIGH PRESSURE AT COREMELT IS INDETERMINATE
NOPDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
Page 5
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Operators Do NotDepressurize with Pressurizer
PORV
NOPZRPORV
Page 8
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
NO RANDOM SPARK ISAVAILABLE BEFORE RV
FAILURE
NOSPARK
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
NOSPARK-1
PROB THAT SPARK IS NOTAVAILABLE BEFORE RVFAILURE WITHOUT RB
SPRAY
NOSPARK_9
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 9
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
NOSPARK-2
PROB THAT SPARK IS NOTAVAILABLE BEFORE RV
FAILURE WITH RB SPRAY
NOSPARK_01
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment is SteamInerted Prior to RV Failure
STMINERTP
Page 10
Containment Strength CanHandle H2 Burn Event Prior
to RV Failure
PRISTRENT
Page 11
Containment Failure FromComb. Gas Burns At RV
Failure is Prevented
H2AT
Page 12
Containment Failure From H2Burns after RV Failure is
Prevented
H2AFTER
Page 18
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 7 3/22/2007
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2Page 6
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 8 3/22/2007
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORVPage 6
Page 21
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
PROBABILITY THAT PORVDOES NOT PREVENT
HPME
NOPRVHPCONF
LOGIC FOR OPERATORSFAILING TO OPEN PORV
NOPZPORVCONF-1
PROBABILITY THATOPERATORS FAIL TO
MANUALLY OPEN PORV
PZPORVCONF_99-C
LOGIC FOR OPERATORSFAILING TO OPEN PORVWHEN HPI COOLING IS
NOT INITIATED
NOPZPORVCONF-2
PROB THAT OPERATORSFAIL TO OPEN PORV
AFTER FAILING TO INITHPI COOLING
PZPORVCONF_0-C
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 9 3/22/2007
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAYPage 6
Page 28Page 27
... see x-ref
CS FAILURE FORINJECTION MODE
CS01
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 10 3/22/2007
Containment is SteamInerted Prior to RV Failure
STMINERTPPage 12Page 6
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 32
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 11 3/22/2007
Containment Strength CanHandle H2 Burn Event Prior
to RV Failure
PRISTRENTPage 6
Containment Strength CanHandle H2 Burn and Base
Pressure is High
PRISTRENH
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 32
Likelihood That Cont. CanHandle H2 Burn Press W/
High Base Press.
PRISTREN1
Containment Strength CanHandle H2 Burn and Base
Pressure is Low
PRISTRENL
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 4
Likelihood That Cont. CanHandle H2 Burn Press W/
Low Base Press.
PRISTREN2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 12 3/22/2007
Containment Failure FromComb. Gas Burns At RV
Failure is Prevented
H2ATPage 6
Hyrdrogen Burns At RVFailure Are Prevented
NOATBURN
H2 Concentration isInsufficient to Cause Burns
at RV Failure
ATCONC
Insufficient Hydrogen isReleased to Containment at
RV Failure
ATRELEASE
Hydrogen Burns Before RVFailure
PRIBURNPage 18
H2 Concentration isSufficient to Cause Burns
Before RV Failure
PRICONC
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
Sufficient Hydrogen isReleased to Containment
Before RV Failure
PRIRELEASE
In-Vessel H2 Prod. Sufficientto Cause H2 Burns
NOOXIDIZED
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLED
Page 13
RANDOM SPARK ISAVAILABLE BEFORE RV
FAILURE
SPARK
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
SPARK-1
PROB THAT SPARK ISAVAILABLE BEFORE RVFAILURE WITHOUT RB
SPRAY
SPARK_1
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 9
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
SPARK-2
PROB THAT SPARK ISAVAILABLE BEFORE RV
FAILURE WITH RB SPRAY
SPARK_99
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment is Not SteamInerted Prior to RV Failure
NOSTMINERTP
Page 18
In-Vessel H2 Prod. NotSufficient to Cause H2 Burns
OXIDIZED
Recovery of Core CoolingDoes Prevent Reactor
Vessel Failure
RECOVRV
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLED
Page 13
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
No Ignition Source isAvailable at RV Failure
ATIGNITION
Page 15
Containment is SteamInerted Prior to RV Failure
STMINERTP
Page 10
Containment Strength CanHandle H2 Burns Event at
RV Failure
ATSTRENT
Page 16
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 13 3/22/2007
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLEDPage 12Page 12
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOW
Page 5
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 14
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 14 3/22/2007
Operators Depressurize WithPressurizer PORV
PZRPORVPage 5
Page 13
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
LOGIC FOR OPERATORSOPENING PORV
PZPORVCONF-1
PROB THAT OPERATORSOPEN PORV
PZPORVCONF_99
LOGIC FOR OPERATORSOPENING PORV WHEN HPI
COOLING IS NOTINITIATED
PZPORVCONF-2
Prob. That Operators OpenPORV After Failing to Init
HPI Cooling
PZPORVCONF_0
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
PROBABILITY THAT PORVCAN PREVENT HPME
PRVHPCONF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 15 3/22/2007
No Ignition Source isAvailable at RV Failure
ATIGNITIONPage 12
No Dispersal of Corium FromCavity
NODISPERSE
Cavity Geometry DoesRetain All Corium
GEOMH2
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 5
Random Spark isUnavailable at RV Failure
NOSPARKAT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 16 3/22/2007
Containment Strength CanHandle H2 Burns Event at
RV Failure
ATSTRENTPage 12
Containment Strength CanHandle H2 Burns and Base
Pressure is High
ATSTRENH
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 32
Likelihood That Cont CanHandle H2 Burn Press. W/
High Base Press.
ATSTREN1
Containment Strength CanHandle H2 Burns and Base
Pressure is Low
ATSTRENL
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 4
Likelihood That Cont CanHandle H2 Burn Press. W/
Low Base Press.
ATSTREN2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 17 3/22/2007
PDS DOES NOT HAVE HIGHBASE PRESSURE IN
CONTAINMENT AT RVFAILURE
NOPDSPRESSHPage 4
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS G
G001
CSS/CIS H
H001
CSS/CIS I
I001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 18 3/22/2007
Containment Failure From H2Burns after RV Failure is
Prevented
H2AFTERPage 6
Combustible Gas Burns EarlyAfter RV Failure are
Prevented
NOAFTBURN
Comb. Gas Concentration isInsufficient to Cause H2Burns Early After Failure
AFTERCONC
Sufficient Comb. Gas isAvailable Early After RV
Failure
AFTERREL
Previous Burns Do DepleteHydrogen in Containment
PRIGLOBAL
Hydrogen Burns Before RVFailure
PRIBURN
Page 12
Hydrogen Burns At ReactorVessel Failure
ATBURN
H2 Concentration is Sufficientto Cause Burns at RV Failure
NOATCONC
Sufficient Hydrogen isReleased to Containment at
RV Failure
NOATRELEASE
Page 19
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
Ignition Source is Available atRV Failure
NOATIGNITION
Dispersal of Corium FromCavity
DISPERSE
Page 20
Random Spark is Available atRV Failure
SPARKAT
Containment is Not SteamInerted Prior to RV Failure
NOSTMINERTPPage 12
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 4
Ex-Vessel Gas ProductionAfter RV Failure is Low
EXVPRODAFTL
Corium Pool Does SpreadOver Large Area Or Freeze
COREFREEZE
Corium Does Spread AcrossLower Containment Or Cavity
Floor
SPREADLOW
Primary System Pressure isHigh at Reactor Vessel Failure
HIGHPRESS
Page 21
Cavity Geometry AllowsEnough Corium to Disperse
For Freezing
GEOMFREEZE
Likelihood Corium DoesFreeze On Lower
Containment or Cavity Floor
FREEZELOW
Concrete Attack ProducesInsufficient Combustible Gas
After RV Failure
H2SRCAFTER
Water Pool Does StopConcrete Attack In Cavity
After RV Failure
NOATTKAFT
Water Pool In Cavity AvailableEarly After RV Failure
WATERAFTER
Page 22
Likelihood That Water Pool inCavity Will Stop Concrete
Attack
NOMELT
Recovery of Core CoolingDoes Prevent Reactor Vessel
Failure
RECOVRV
Cavity Recombination DoesDeplete Combustible Gas
Early After RV Failure
AFTERRECOM
Page 25
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
NO RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE
NOSPARKAFT
Page 28
Containment Is Steam InertedAfter RV Failure
STMINERTAF
Page 29
Containment Strength CanHandle Comb. Gas Burn
Event After RV Failure
AFTSTRENT
Page 30
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 19 3/22/2007
Sufficient Hydrogen isReleased to Containment at
RV Failure
NOATRELEASEPage 18
Hydrogen Does Not BurnBefore RV Failure to Deplete
H2 Concentration
NOPRIBURN
Page 6
In-Vessel H2 Prod. Sufficientto Cause H2 Burns
NOOXIDIZED
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
Hydrogen Has Not BeenReleased to Containment
BOTTLED
Page 6
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 20 3/22/2007
Dispersal of Corium FromCavity
DISPERSEPage 18
Cavity Geometry Does NotRetain All Corium
NOGEOMH2
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 21
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 21 3/22/2007
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESSPage 33Page 22Page 20
... see x-ref
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOW
Page 6
Operators Do NotDepressurize Steam
Generators
OPSNOSSHR
Steam GeneratorDepressurization Or SSHR Is
Unavailable
NOPORVSSHR
PDS INDICATES OTSGADVS ARE UNAVAILABLE
NOPDSSGADV
AV
Secondary Side HeatRemoval is Unavailable
SSHRUNAVAIL
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
NO SSHR EXISTS
NO-SSHR
Page 35
Prob. that Secondary SideHeat Removal is Not
Recovered Prior to RVFailure
NORECOVSSHR
Likelihood That Operators DoNot Depressurize Steam
Generators
NOOPSDEPRESS
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORV
Page 8
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 22 3/22/2007
Water Pool In CavityAvailable Early After RV
Failure
WATERAFTERPage 18
Water Does Fill Cavity FromPlant Specific Sources And
Paths
OTHERWATER
Water Available From SpraysVia Fuel Transfer Pool Early
After RV Failure
FTRNSPOOLAFT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAY
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment Sprays AreRecovered Prior to RV
Failure
RECOVSPPRI
Containment Sprays AreRecovered Early After RV
Failure
RECOVSPAFT
Accumulator Water isAvailable at RV Failure
ACCUMAVAIL
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 21
BWST Water Gravity FeedInto Reactor Cavity Through
Failed Reactor Vessel
GRAVFEEDAFT
FAILURE OF ECCSINJECTION
PDSINJECCS
Page 23
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSH
Page 24
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 23 3/22/2007
FAILURE OF ECCSINJECTION
PDSINJECCSPage 22
CORE MELT BIN 1
CM-001
CORE MELT BIN 4
CM-004
CORE MELT BIN 7
CM-007
CORE MELT BIN 9
CM-009
CORE MELT BIN 12
CM-012
CORE MELT BIN 15
CM-015
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 24 3/22/2007
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSHPage 22Page 30
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS G
G001
CSS/CIS H
H001
CSS/CIS I
I001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 25 3/22/2007
Cavity Recombination DoesDeplete Combustible Gas
Early After RV Failure
AFTERRECOMPage 18
Water Pool In CavityUnavailable Early After RV
Failure
NOWATERAFTER
Water Does Not Fill CavityFrom Plant Specific Sources
And Paths
NOOTHERWATER
Accumulator Water isUnavailable at RV Failure
ACCUMUNAVAIL
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 5
No BWST Water GravityFeed Into Reactor CavityThrough Failed Reactor
Vessel
NOGRAVFEEDAFT
NO FAILURE OF ECCSINJECTION
NOPDSINJECCS
Page 26
PDS INDICATES HIGHBASE PRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
SEQPRESSH
Page 30
Water Unavailable FromSprays Via Fuel Transfer
Pool Early After RV Failure
NOFTRNSPOOLAFT
Page 27
Likelihood ThatRecombination Can Deplete
Comb. Gas Given a DryCavity
DRYEFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 26 3/22/2007
NO FAILURE OF ECCSINJECTION
NOPDSINJECCSPage 25
CORE MELT BIN 2
CM-002
CORE MELT BIN 5
CM-005
CORE MELT BIN 8
CM-008
CORE MELT BIN 11
CM-011
CORE MELT BIN 3
CM-003
CORE MELT BIN 6
CM-006
CORE MELT BIN 10
CM-010
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 27 3/22/2007
Water Unavailable FromSprays Via Fuel Transfer
Pool Early After RV Failure
NOFTRNSPOOLAFTPage 25
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 9
Containment Sprays Are NotRecovered Prior to RV
Failure
NORECOVSPPRI
Containment Sprays Are NotRecovered Early After RV
Failure
NORECOVSPAFT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 28 3/22/2007
NO RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE
NOSPARKAFTPage 18
NO RANDOM SPARKEARLY AFTER RV FAILUREFOR RB SPRAY AVAILABLE
NOSPARKAFT-1
PROB THAT SPARK ISUNAVAILABLE EARLY
AFTER RV FAILURE WITHRB SPRAY
NOSPARKAFT_01
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
NO RANDOM SPARKEARLY AFTER RV FAILURE
FOR RB SPRAYUNAVAILABLE
NOSPARKAFT-2
PROB THAT SPARK ISUNAVAILABLE EARLYAFTER RV FAILURE
WITHOUT RB SPRAY
NOSPARKAFT_9
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 9
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 29 3/22/2007
Containment Is SteamInerted After RV Failure
STMINERTAFPage 18
Sequence After RV FailureHas High Pressure In
Containment
AFTPRESSH
Page 30
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 30 3/22/2007
Containment Strength CanHandle Comb. Gas BurnEvent After RV Failure
AFTSTRENTPage 18
Containment Strength CanHandle Comb. Gas Burn and
Base Pressure is High
AFTSTRENH
Containment Base Pressureis High
BASEPRESSH
Containment Has High BasePressure Early After RVFailure Without Steam
Inerting
NOINERTAF
Sequence After RV FailureHas High Pressure In
Containment
AFTPRESSHPage 29
PDS INDICATES HIGHBASE PRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
SEQPRESSHPage 25
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CSS/CIS J
J001
CSS/CIS K
K001
CSS/CIS L
L001
Reactor Building Fans DoNot Function Early After RV
Failure
NOFANSAFT
Reactor Building FansUnavailable Early After RV
Failure
FANSUNAVAILAFT
Reactor Building Fans DoNot Function at RV Failure
FANSPRI
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRI
Page 31
Likelihood RB Fans Do NotSurvive Containment
Enviroment At Or Prior ToRV Failu
NOEQUALFANSPRI
Reactor Building Fans AreNot Recovered Early After
RV Failure
NORECOVFANSAFT
Likelihood Fans Do NotSurvive Containment
Environment Early After RVFailure
NOEQUALFANSAF
Likelihood That Cont. CanHandle Comb. Gas Burn
Press. W/ High BasePressure
AFTSTREN1
Containment Strength CanHandle Comb. Gas Burn and
Base Pressure is Low
AFTSTRENL
Containment Base Pressureis Low
BASEPRESSL
Containment Has Low BasePressure Early After RVFailure Without Steam
Inerting
INERTAF
Sequence After RV FailureHas Low Base Pressure In
Containment
AFTPRESSL
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSH
Page 24
Reactor Building FansAvailable Early After RV
Failure
FANSAFT
Reactor Building FansAvailable Early After RV
Failure
FANSAVAILAFT
Reactor Building Fans DoFunction at RV Failure
NOFANSPRI
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRI
Page 2
Likelihood RB Fans DoSurvive Containment
Enviroment At Or Prior ToRV Failu
EQUALFANSPRI
Reactor Building Fans AreRecovered Early After RV
Failure
RECOVFANSAFT
Likelihood Fans SurviveContainment Environment
Early After RV Failure
EQUALFANSAF
Likelihood That Cont. CanHandle Comb. Gas Burn
Press. W/ Low BasePressure
AFTSTREN2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 31 3/22/2007
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRIPage 30Page 32
PDS INDICATES THAT RBFANS ARE NOT AVAILABLE
AT OR PRIOR TO RVFAILURE
NOPDSFANS
CF
Reactor Building Fans AreNot Recovered At or Prior to
RV Failure
NORECOVFANSPRI
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 32 3/22/2007
Containment Failure FromRapid Steam Generation is
Prevented
NORSGPage 1
Rapid Steam GenerationDoes Not Occur
NORSGOCCUR
Heat Transfer Rate FromCorium to Water Pool is
Slow
SLOWHTRATE
Likelihood That No WaterReaches Cavity Prior to RV
Failure
NOWATERCAV
PDS INDICATES THAT RBSPRAYS ARE NOT
AVAILABLE IN INJECTIONMODE
NOPDSINJSP
CS FAILURE FORINJECTION MODE
CS01
Containment Sprays Are NotRecovered Prior to RV
Failure
NORECOVSPPRI
Reactor Building Fans CanHandle Steam Production
RSGFANSG
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRI
Page 2
Likelihood That ReactorBuilding Fans Can HandleRapid Steam Production
RSGFANSEFF
Containment Strength CanHandle Rapid Steam
Generation Event
RGSTRENT
Containment Strength CanHandle RSG Event and Base
Pressure is High
RSGSTRENH
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSHPage 16Page 11Page 1
... see x-ref
PDS DOES HAVE HIGHBASE PRESSURE IN
CONTAINMENT AT RVFAILURE
PDSPRESSH
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CSS/CIS J
J001
CSS/CIS K
K001
CSS/CIS L
L001
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRI
Page 31
Likelihood Strength CanHandle RSG Event and Base
Pressure is High
RSGSTREN1
Containment Strength CanHandle RSG Event and Base
Pressure is Low
RSGSTRENL
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 4
Likelihood Strength CanHandle RSG Event and Base
Pressure is Low
RSGSTREN2
Recovery of Core CoolingDoes Prevent Reactor
Vessel Failure
RECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 33 3/22/2007
Containment Failure FromDirect Contact of Corium is
Prevented
NOCONTACTPage 1
Insufficient Amount of CoriumCan Make Contact WithContainment Wall With
LPME
NOCOREWALLLP
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 5
Plant Configuration andLayout Limits Material
Reaching Cont. Wall withLPME
CWLIMITLPME
Insufficient Amount of CoriumCan Make Contact WithContainment Wall With
HPME
NOCOREWALLHP
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 21
Plant Configuration andLayout Limits Material
Reaching Cont. Wall withHPME
CWLIMITHPME
Recovery of Core CoolingDoes Prevent Reactor Vessel
Failure
RECOVRV
Containment Wall SurvivesContact With Corium
WALLSURVIV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 34 3/22/2007
Secondary Side HeatRemoval is Available
SSHRAVAILPage 5
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHR
Page 35
Prob. that Secondary SideHeat Removal is Recovered
Prior to RV Failure
RECOVSSHR
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 35 3/22/2007
NO SSHR EXISTS
NO-SSHRPage 34Page 21
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 36 3/22/2007
Containment Failure FromMissiles is Prevented
NOMISSLEPage 1
No Alpha Mode Failure ofContainment
NOALPHA
No RV Rocket
NORVROCKET
Containment Failure FromPressure Generated
Missile(s) is Prevented
NOPGENMISSL
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 5
Recovery of Core CoolingDoes Prevent Reactor Vessel
Failure
RECOVRV
Likelihood That Cont Failureis Prevented Given a
Pressure Generated Missle
NOMISSLELIKE
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 37 3/22/2007
Name Page Zone Name Page Zone
A001 17 1A001 24 1ACCUMAVAIL 22 4ACCUMUNAVAIL 25 2AFTERCONC 18 6AFTERRECOM 18 6AFTERRECOM 25 3AFTERREL 18 5AFTPRESSH 29 1AFTPRESSH 30 3AFTPRESSL 30 6AFTSTREN1 30 3AFTSTREN2 30 6AFTSTRENH 30 2AFTSTRENL 30 5AFTSTRENT 18 8AFTSTRENT 30 4ATBURN 18 3ATCONC 12 6ATIGNITION 12 7ATIGNITION 15 2ATPRESSH 1 3ATPRESSH 10 1ATPRESSH 11 1ATPRESSH 16 1ATPRESSH 32 5ATPRESSL 3 1ATPRESSL 4 2ATPRESSL 11 3ATPRESSL 16 3ATPRESSL 18 5ATPRESSL 32 7ATRELEASE 12 5ATSTREN1 16 2ATSTREN2 16 4ATSTRENH 16 2ATSTRENL 16 4ATSTRENT 12 8ATSTRENT 16 2
AV 5 6AV 21 1B001 17 1B001 24 1BASEPRESSH 30 2BASEPRESSL 30 5BOTTLED 6 3BOTTLED 19 2BWHBW1-----HP2OA 8 4BWHBW1-----HP2OA 14 4C001 17 1C001 24 1CAG0005 35 1CAG0005-R 35 2CF 2 1CF 31 1CM-001 5 1CM-001 23 1CM-002 5 2CM-002 26 1CM-003 5 2CM-003 26 1CM-004 5 4CM-004 23 1CM-005 5 5CM-005 26 1CM-006 5 4CM-006 26 1CM-007 7 1CM-007 23 1CM-008 7 1CM-008 26 1CM-009 7 1CM-009 23 2CM-010 7 1CM-010 26 2CM-011 7 1CM-011 26 1CM-012 7 1
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 38 3/22/2007
Name Page Zone Name Page Zone
CM-012 23 2CM-013 7 2CM-013 26 2CM-014 7 2CM-014 26 2CM-015 7 2CM-015 23 2CM-016 7 2CM-016 26 2CM-018 7 2CM-018 26 2CM-019 7 2CM-019 26 2COREFREEZE 18 7CS01 9 1CS01 32 2CWLIMITHPME 33 4CWLIMITLPME 33 2D001 30 1D001 32 4DCHFANSEFF 1 3DCHFANSG 1 2DCHFRAG 1 1DCHSTREN1 1 4DCHSTREN2 3 2DCHSTRENH 1 4DCHSTRENL 1 5DCHSTRENL 3 2DCHSTRENT 1 4DISPERSE 18 3DISPERSE 20 2DRYEFF 25 4E001 30 1E001 32 4EQUALFANSAF 30 7EQUALFANSPRI 30 6EXVPRODAFTL 18 8F001 30 1F001 32 4
FANSAFT 30 6FANSAT 4 2FANSAVAILAFT 30 6FANSAVAILPRI 1 2FANSAVAILPRI 2 2FANSAVAILPRI 4 2FANSAVAILPRI 30 5FANSAVAILPRI 32 3FANSPRI 30 4FANSUNAVAILAFT 30 4FANSUNAVAILPRI 30 3FANSUNAVAILPRI 31 2FANSUNAVAILPRI 32 5FREEZELOW 18 8FTRNSPOOLAFT 22 2G001 17 1G001 24 1GEOMFREEZE 18 7GEOMH2 15 1GRAVFEEDAFT 22 6H001 17 1H001 24 1H2AFTER 6 6H2AFTER 18 8H2AT 6 5H2AT 12 7H2PRI 6 4H2SRCAFTER 18 8HEATIML 5 8HEATIML 13 1HIGHPRESS 18 6HIGHPRESS 20 2HIGHPRESS 21 3HIGHPRESS 22 4HIGHPRESS 33 3I001 17 1I001 24 1IE-LOOP-100 35 3INERTAF 30 4
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 39 3/22/2007
Name Page Zone Name Page Zone
J001 30 2J001 32 5K001 30 2K001 32 5L001 30 2L001 32 5LOWCONCBURN 6 1LOWCONCBURN 12 6LOWCONCBURN 18 7LOWPRESS 1 5LOWPRESS 5 5LOWPRESS 15 2LOWPRESS 25 2LOWPRESS 33 1LOWPRESS 36 3M001 17 2M001 24 2N001 17 2N001 24 2NO-SSHR 21 2NO-SSHR 34 1NO-SSHR 35 2NO-SSHR-POSTLOOP 35 2NOAFTBURN 18 7NOALPHA 36 1NOATBURN 12 7NOATCONC 18 2NOATIGNITION 18 4NOATRELEASE 18 1NOATRELEASE 19 2NOATTKAFT 18 9NOCONTACT 1 7NOCONTACT 33 3NOCOREWALLHP 33 4NOCOREWALLLP 33 2NODCH 1 4NODCHLOAD 1 3NODISPERSE 15 2NOEARLY 1 6
NOEQUALFANSAF 30 5NOEQUALFANSPRI 30 4NOFANSAFT 30 4NOFANSPRI 30 6NOFTRNSPOOLAFT 25 5NOFTRNSPOOLAFT 27 2NOGEOMFREEZE 1 5NOGEOMH2 20 1NOGRAVFEEDAFT 25 4NOH2BURNS 1 5NOH2BURNS 6 5NOHEATIML 6 3NOHEATIML 21 4NOINERTAF 30 1NOLOWCONCBURN 12 1NOLOWCONCBURN 18 2NOMELT 18 10NOMISSLE 1 8NOMISSLE 36 2NOMISSLELIKE 36 4NOOPSDEPRESS 21 3NOOTHERWATER 25 1NOOXIDIZED 12 2NOOXIDIZED 19 1NOPDSFANS 31 1NOPDSINJECCS 25 3NOPDSINJECCS 26 2NOPDSINJSP 32 2NOPDSLOW 6 2NOPDSLOW 21 1NOPDSLOW-1 6 2NOPDSLOW-2 6 1NOPDSLOW-2 7 2NOPDSLOW_5 6 2NOPDSPRESSH 4 1NOPDSPRESSH 17 2NOPDSPZRPORV 8 1NOPDSSGADV 21 1NOPDSSPRAY 6 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOEARLY"(2004 Rev. 2)
Page 40 3/22/2007
Name Page Zone Name Page Zone
NOPDSSPRAY 9 1NOPDSSPRAY 12 5NOPDSSPRAY 27 1NOPDSSPRAY 28 4NOPDSSSHR 21 2NOPGENMISSL 36 3NOPORVSSHR 21 2NOPRIBURN 6 4NOPRIBURN 19 1NOPRICONC 6 2NOPRIRELEASE 6 2NOPRVHPCONF 8 2NOPZPORVCONF-1 8 3NOPZPORVCONF-2 8 4NOPZRPORV 6 4NOPZRPORV 8 2NOPZRPORV 21 3NOPZRSAFETY 6 5NOPZRSAFETY 21 5NORECOVFANSAFT 30 5NORECOVFANSPRI 31 2NORECOVRV 19 2NORECOVSPAFT 27 2NORECOVSPPRI 27 2NORECOVSPPRI 32 3NORECOVSSHR 21 3NORSG 1 6NORSG 32 4NORSGOCCUR 32 2NORVROCKET 36 2NOSEQPRESSH 22 6NOSEQPRESSH 24 2NOSEQPRESSH 30 5NOSPARK 6 5NOSPARK-1 6 5NOSPARK-2 6 7NOSPARKAFT 18 7NOSPARKAFT 28 2NOSPARKAFT-1 28 2
NOSPARKAFT-2 28 4NOSPARKAFT_01 28 1NOSPARKAFT_9 28 3NOSPARKAT 15 3NOSPARK_01 6 6NOSPARK_9 6 4NOSTMINERTP 12 6NOSTMINERTP 18 5NOTBOTTLED 12 3NOTBOTTLED 12 7NOTBOTTLED 13 2NOWATERAFTER 25 3NOWATERCAV 32 2O001 17 2O001 24 2OPSDEPRESS 5 7OPSNOSSHR 21 2OPSSSHR 5 7OTHERWATER 22 1OXIDIZED 6 1OXIDIZED 12 5P001 17 2P001 24 2PDSFANS 2 1PDSINDET 5 4PDSINDET 6 3PDSINJECCS 22 5PDSINJECCS 23 2PDSLOW 5 2PDSLOW 13 1PDSLOW-1 5 2PDSLOW-2 5 4PDSLOW_5 5 3PDSPRESSH 32 4PDSPZRPORV 14 1PDSSGADV 5 6PDSSPRAY 22 1PDSSSHR 34 1PO-HPICOOL 8 1
Name Page Zone Name Page Zone
PO-HPICOOL 14 1PORVSSHR 5 6PRIBURN 12 4PRIBURN 18 1PRICONC 12 2PRIGLOBAL 18 2PRIRELEASE 12 2PRISTREN1 11 2PRISTREN2 11 4PRISTRENH 11 2PRISTRENL 11 4PRISTRENT 6 5PRISTRENT 11 2PRVHPCONF 14 4PZPORVCONF-1 14 3PZPORVCONF-2 14 3PZPORVCONF_0 14 3PZPORVCONF_0-C 8 3PZPORVCONF_99 14 2PZPORVCONF_99-C 8 3PZRPORV 5 3PZRPORV 13 2PZRPORV 14 2PZRSAFETY 5 9PZRSAFETY 13 2Q001 17 2Q001 24 2R001 17 2R001 24 2RBSPRAY 6 7RBSPRAY 12 7RBSPRAY 22 1RBSPRAY 28 2RECOVFANSAFT 30 7RECOVFANSPRI 2 2RECOVRV 1 4RECOVRV 12 6RECOVRV 18 10RECOVRV 32 7
RECOVRV 33 5RECOVRV 36 3RECOVSPAFT 22 3RECOVSPPRI 22 2RECOVSSHR 34 2RGSTRENT 32 6RSGFANSEFF 32 4RSGFANSG 32 4RSGSTREN1 32 6RSGSTREN2 32 8RSGSTRENH 32 5RSGSTRENL 32 7SEQPRESSH 25 4SEQPRESSH 30 2SLOWHTRATE 32 1SPARK 12 5SPARK-1 12 4SPARK-2 12 6SPARKAT 18 4SPARK_1 12 4SPARK_99 12 6SPREADLOW 18 7SSHRAVAIL 5 7SSHRAVAIL 34 2SSHRUNAVAIL 21 3STMINERTAF 18 8STMINERTAF 29 1STMINERTP 6 6STMINERTP 10 1STMINERTP 12 8WALLSURVIV 33 6WATERAFTER 18 9WATERAFTER 22 3YES-SSHR 34 1
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOEXRELEASE" (2004 Rev. 2)
Page 1 3/22/2007
No Ex-Vessel Release of FPs
NOEXRELEASE
Recovery of Core CoolingDoes Prevent Reactor Vessel
Failure
RECOVRV
Corium Pool Does SpreadOver Large Area Or Freeze
COREFREEZE
Corium Does Spread AcrossLower Containment Or Cavity
Floor
SPREADLOW
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOW
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2
Page 2
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
NOPDSLOW-1
HIGH PRESSURE AT COREMELT IS INDETERMINATE
NOPDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
Operators Do NotDepressurize Steam
Generators
OPSNOSSHR
Steam GeneratorDepressurization Or SSHR Is
Unavailable
NOPORVSSHR
PDS INDICATES OTSGADVS ARE UNAVAILABLE
NOPDSSGADV
AV
Secondary Side HeatRemoval is Unavailable
SSHRUNAVAIL
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
Page 3
Prob. that Secondary SideHeat Removal is Not
Recovered Prior to RV Failure
NORECOVSSHR
Likelihood That Operators DoNot Depressurize Steam
Generators
NOOPSDEPRESS
Operators Do NotDepressurize with Pressurizer
PORV
NOPZRPORV
Page 4
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
Cavity Geometry AllowsEnough Corium to Disperse
For Freezing
GEOMFREEZE
Likelihood Corium DoesFreeze On Lower
Containment or Cavity Floor
FREEZELOW
No Ex-Vessel Release of FPsto Cont. Atmos. and Water
Pool Avail.
RVORPOOL
Page 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOEXRELEASE" (2004 Rev. 2)
Page 2 3/22/2007
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2Page 1
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOEXRELEASE" (2004 Rev. 2)
Page 3 3/22/2007
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHRPage 1
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOEXRELEASE" (2004 Rev. 2)
Page 4 3/22/2007
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORVPage 1
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
PROBABILITY THAT PORVDOES NOT PREVENT
HPME
NOPRVHPCONF
LOGIC FOR OPERATORSFAILING TO OPEN PORV
NOPZPORVCONF-1
PROBABILITY THATOPERATORS FAIL TO
MANUALLY OPEN PORV
PZPORVCONF_99-C
LOGIC FOR OPERATORSFAILING TO OPEN PORVWHEN HPI COOLING IS
NOT INITIATED
NOPZPORVCONF-2
PROB THAT OPERATORSFAIL TO OPEN PORV
AFTER FAILING TO INITHPI COOLING
PZPORVCONF_0-C
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOEXRELEASE" (2004 Rev. 2)
Page 5 3/22/2007
No Ex-Vessel Release of FPsto Cont. Atmos. and Water
Pool Avail.
RVORPOOLPage 1
No Ex-Vessel Release of FPsto Cont. Atmos. From the
Cavity
NORVFAILS
Water Pool Stops ConcreteAttack Prior to LateContainment Failure
NOATTKLT
Likelihood That Water Pool inCavity Will Stop Concrete
Attack
NOMELT
Water Available In ReactorCavity Prior to Ex-VesselFission Product Release
EXFISWATERPage 5
Water Available From SpraysVia Fuel Transfer Pool Early
After RV Failure
FTRNSPOOLAFT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAY
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment Sprays AreRecovered Prior to RV Failure
RECOVSPPRI
Containment Sprays AreRecovered Early After RV
Failure
RECOVSPAFT
Likelihood That OverlyingWater Pool Will Scrub FPs
Released From Corium
EXSCRUBEFF
Water Available In ReactorCavity Prior to Ex-VesselFission Product Release
EXFISWATER
Page 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOEXRELEASE" (2004 Rev. 2)
Page 6 3/22/2007
Name Page Zone Name Page Zone
AV 1 4BWHBW1-----HP2OA 4 4CAG0005 3 1CAG0005-R 3 2CM-004 1 2CM-005 1 3CM-006 1 3CM-007 2 1CM-008 2 1CM-009 2 1CM-010 2 1CM-011 2 1CM-012 2 1CM-013 2 2CM-014 2 2CM-015 2 2CM-016 2 2CM-018 2 2CM-019 2 2COREFREEZE 1 6EXFISWATER 5 2EXFISWATER 5 3EXSCRUBEFF 5 3FREEZELOW 1 6FTRNSPOOLAFT 5 2GEOMFREEZE 1 6HIGHPRESS 1 5IE-LOOP-100 3 3NO-SSHR 3 2NO-SSHR-POSTLOOP 3 2NOATTKLT 5 2NOEXRELEASE 1 4NOHEATIML 1 7NOMELT 5 1NOOPSDEPRESS 1 6NOPDSLOW 1 2NOPDSLOW-1 1 2NOPDSLOW-2 1 1NOPDSLOW-2 2 2
NOPDSLOW_5 1 2NOPDSPZRPORV 4 1NOPDSSGADV 1 4NOPDSSSHR 1 5NOPDSSSHR 3 1NOPORVSSHR 1 5NOPRVHPCONF 4 2NOPZPORVCONF-1 4 3NOPZPORVCONF-2 4 4NOPZRPORV 1 6NOPZRPORV 4 2NOPZRSAFETY 1 8NORECOVSSHR 1 6NORVFAILS 5 2OPSNOSSHR 1 5PDSINDET 1 3PDSSPRAY 5 1PO-HPICOOL 4 1PZPORVCONF_0-C 4 3PZPORVCONF_99-C 4 3RBSPRAY 5 1RECOVRV 1 1RECOVSPAFT 5 3RECOVSPPRI 5 2RVORPOOL 1 7RVORPOOL 5 2SPREADLOW 1 5SSHRUNAVAIL 1 5
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 1 3/22/2007
No Late Containment Failure
NOLATE
No Containment FailureFrom Late Combustible
Gases
NOH2LATE
No Late Combustible GasBurns
NOLTBURN
Combustible GasConcentration is Insuff to
Cause Burns Late After RVFailure
NOLTCONC
Previous Combustible GasBurns Deplete Combustible
Gas in Containment
LTPRIGLOB
Hydrogen Burns Before RVFailure
PRIBURNPage 14
H2 Concentration isSufficient to Cause Burns
Before RV Failure
PRICONC
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
Sufficient Hydrogen isReleased to Containment
Before RV Failure
PRIRELEASE
Page 2
RANDOM SPARK ISAVAILABLE BEFORE RV
FAILURE
SPARK
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
SPARK-1
Page 3
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
SPARK-2
Page 5
Containment is Not SteamInerted Prior to RV Failure
NOSTMINERTP
Page 1
Hydrogen Burns At ReactorVessel Failure
ATBURN
H2 Concentration isSufficient to Cause Burns at
RV Failure
NOATCONC
Sufficient Hydrogen isReleased to Containment at
RV Failure
NOATRELEASE
Page 6
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
Ignition Source is Available atRV Failure
NOATIGNITION
Page 8
Containment is Not SteamInerted Prior to RV Failure
NOSTMINERTPPage 1
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSL
Page 9
Combustible Gas Burns EarlyAfter RV Failure
AFTBURN
Page 11
Cavity Recombination DoesDeplete Combustible Gas
Early After RV Failure
AFTERRECOM
Page 26
Cavity RecombinationDepletes Comb. Gas Prior to
Late Containment Failure
LTRECOMB
Page 30
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
Containment is SteamInerted Late After RV Failure
LATEINERT
Page 31
Random Spark is UnavailableLate After RV Failure
NOSPARKLT
Page 33
Containment Strength CanHandle Late Combustible
Gas Burn Event
LTSTRENT
Page 34
No Containment FailureFrom Steam Generation
NOSTEAM
Page 36
No Containment FailureFrom Non Condensable
Gases
NOGASES
Page 38
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 2 3/22/2007
Sufficient Hydrogen isReleased to Containment
Before RV Failure
PRIRELEASEPage 1
In-Vessel H2 Prod. Sufficientto Cause H2 Burns
NOOXIDIZED
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLED
Page 14
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 3 3/22/2007
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
SPARK-1Page 1
PROB THAT SPARK ISAVAILABLE BEFORE RVFAILURE WITHOUT RB
SPRAY
SPARK_1
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 4
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 4 3/22/2007
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAYPage 11Page 36Page 24
... see x-ref
CS FAILURE FORINJECTION MODE
CS01
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 5 3/22/2007
RANDOM SPARK ISAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
SPARK-2Page 1
PROB THAT SPARK ISAVAILABLE BEFORE RV
FAILURE WITH RB SPRAY
SPARK_99
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 6 3/22/2007
Sufficient Hydrogen isReleased to Containment at
RV Failure
NOATRELEASEPage 1
Hydrogen Does Not BurnBefore RV Failure to Deplete
H2 Concentration
NOPRIBURN
Page 11
In-Vessel H2 Prod. Sufficientto Cause H2 Burns
NOOXIDIZED
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
Hydrogen Has Not BeenReleased to Containment
BOTTLED
Page 7
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 7 3/22/2007
Hydrogen Has Not BeenReleased to Containment
BOTTLEDPage 11Page 6
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOWPage 21
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2
Page 12
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
NOPDSLOW-1
HIGH PRESSURE AT COREMELT IS INDETERMINATE
NOPDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
Page 14
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORV
Page 13
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 8 3/22/2007
Ignition Source is Availableat RV Failure
NOATIGNITIONPage 1
Dispersal of Corium FromCavity
DISPERSE
Cavity Geometry Does NotRetain All Corium
NOGEOMH2
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESS
Page 21
Random Spark is Available atRV Failure
SPARKAT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 9 3/22/2007
Sequence Has Low BasePressure in Containment at
RV Failure
ATPRESSLPage 1
PDS DOES NOT HAVEHIGH BASE PRESSURE IN
CONTAINMENT AT RVFAILURE
NOPDSPRESSH
Page 10
Reactor Building Fans AreAvailable Prior to RV Failure
FANSAT
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRIPage 36
PDS INDICATES THAT RBFANS ARE AVAILABLE AT
OR PRIOR TO RV FAILURE
PDSFANS
CF
Reactor Building Fans AreRecovered At or Prior to RV
Failure
RECOVFANSPRI
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 10 3/22/2007
PDS DOES NOT HAVE HIGHBASE PRESSURE IN
CONTAINMENT AT RVFAILURE
NOPDSPRESSHPage 9
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS G
G001
CSS/CIS H
H001
CSS/CIS I
I001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 11 3/22/2007
Combustible Gas BurnsEarly After RV Failure
AFTBURNPage 1
Comb. Gas Concentration isSufficient to Cause H2 Burns
Early After Failure
NOAFTERCONC
Sufficient Comb. Gas isAvailable Early After RV
Failure
NOAFTERREL
Previous Burns Do NotDeplete Hydrogen in
Containment
NOPRIGLOBAL
Hydrogen Does Not BurnBefore RV Failure to Deplete
H2 Concentration
NOPRIBURNPage 6
H2 Concentration is NotSufficient to Cause Burns
Before RV Failure
NOPRICONC
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
No Sufficient Hydrogen isReleased to Containment
Before RV Failure
NOPRIRELEASE
In-Vessel H2 Prod. NotSufficient to Cause H2 Burns
OXIDIZED
Hydrogen Has Not BeenReleased to Containment
BOTTLED
Page 7
NO RANDOM SPARK ISAVAILABLE BEFORE RV
FAILURE
NOSPARK
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
UNAVAILABLE
NOSPARK-1
PROB THAT SPARK IS NOTAVAILABLE BEFORE RVFAILURE WITHOUT RB
SPRAY
NOSPARK_9
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 4
RANDOM SPARK IS NOTAVAILABLE BEFORE RVFAILURE FOR RB SPRAY
AVAILABLE
NOSPARK-2
PROB THAT SPARK IS NOTAVAILABLE BEFORE RV
FAILURE WITH RB SPRAY
NOSPARK_01
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment is SteamInerted Prior to RV Failure
STMINERTP
Page 14
Hyrdrogen Burns At RVFailure Are Prevented
NOATBURN
Page 14
Ex-Vessel Gas ProductionAfter RV Failure is High
EXVPRODAFTH
Page 20
Cavity Recombination DoesNot Deplete Combustible
Gas Early After RV Failure
NOAFTERRECOM
Page 21
No Random LowConcentration Burns PreventSignificant Accumulation of
Comb. Gas
NOLOWCONCBURN
RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE
SPARKAFT
Page 24
Containment Is Not SteamInerted After RV Failure
NOSTMINERTAF
Page 25
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 12 3/22/2007
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2Page 7
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 13 3/22/2007
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORVPage 7
Page 21
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
PROBABILITY THAT PORVDOES NOT PREVENT
HPME
NOPRVHPCONF
LOGIC FOR OPERATORSFAILING TO OPEN PORV
NOPZPORVCONF-1
PROBABILITY THATOPERATORS FAIL TO
MANUALLY OPEN PORV
PZPORVCONF_99-C
LOGIC FOR OPERATORSFAILING TO OPEN PORVWHEN HPI COOLING IS
NOT INITIATED
NOPZPORVCONF-2
PROB THAT OPERATORSFAIL TO OPEN PORV
AFTER FAILING TO INITHPI COOLING
PZPORVCONF_0-C
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 14 3/22/2007
Hyrdrogen Burns At RVFailure Are Prevented
NOATBURNPage 11
H2 Concentration isInsufficient to Cause Burns at
RV Failure
ATCONC
Insufficient Hydrogen isReleased to Containment at
RV Failure
ATRELEASE
Hydrogen Burns Before RVFailure
PRIBURN
Page 1
In-Vessel H2 Prod. NotSufficient to Cause H2 Burns
OXIDIZED
Recovery of Core CoolingDoes Prevent Reactor Vessel
Failure
RECOVRV
Hydrogen Has Already BeenReleased to Containment
NOTBOTTLEDPage 2
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOW
Page 14
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 15
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
Random Low ConcentrationBurns Prevent Significant
Accumulation of Comb. Gas
LOWCONCBURN
No Ignition Source isAvailable at RV Failure
ATIGNITION
No Dispersal of Corium FromCavity
NODISPERSE
Cavity Geometry Does RetainAll Corium
GEOMH2
Primary System Pressure isLow At RV Failure
LOWPRESSPage 26Page 36
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOWPage 14
PDS INDICATES LOWPRESSURE AT CORE MELT
IS CERTAIN
PDSLOW-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
PDSLOW-2
LOW PRESSURE AT COREMELT IS INDETERMINATE
PDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDETPage 7
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 15
Operators DepressurizeSteam Generators
OPSSSHR
Page 16
Prob. that Failure of thePrimary System Occurs Due
to Heating
HEATIML
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
Random Spark is Unavailableat RV Failure
NOSPARKAT
Containment is SteamInerted Prior to RV Failure
STMINERTPPage 11
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSH
Page 18
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 15 3/22/2007
Operators Depressurize WithPressurizer PORV
PZRPORVPage 14Page 14
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
LOGIC FOR OPERATORSOPENING PORV
PZPORVCONF-1
PROB THAT OPERATORSOPEN PORV
PZPORVCONF_99
LOGIC FOR OPERATORSOPENING PORV WHEN HPI
COOLING IS NOTINITIATED
PZPORVCONF-2
Prob. That Operators OpenPORV After Failing to Init
HPI Cooling
PZPORVCONF_0
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
PROBABILITY THAT PORVCAN PREVENT HPME
PRVHPCONF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 16 3/22/2007
Operators DepressurizeSteam Generators
OPSSSHRPage 14
Steam GeneratorDepressurization and SSHR
Are Available
PORVSSHR
PDS INDICATES OTSGADVS ARE AVAILABLE
PDSSGADV
AV
Secondary Side HeatRemoval is Available
SSHRAVAIL
PDS INDICATES SSHR ISAVAILABLE
PDSSSHR
SSHR IS AVAILABLE
YES-SSHR
NO SSHR EXISTS
NO-SSHR
Page 17
Prob. that Secondary SideHeat Removal is Recovered
Prior to RV Failure
RECOVSSHR
Likelihood That OperatorsDepressurize Steam
Generators
OPSDEPRESS
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 17 3/22/2007
NO SSHR EXISTS
NO-SSHRPage 16Page 21
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 18 3/22/2007
Sequence Has High BasePressure in Containment at
RV Failure
ATPRESSHPage 14
PDS DOES HAVE HIGHBASE PRESSURE IN
CONTAINMENT AT RVFAILURE
PDSPRESSH
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CSS/CIS J
J001
CSS/CIS K
K001
CSS/CIS L
L001
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRI
Page 19
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 19 3/22/2007
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRIPage 31Page 18
PDS INDICATES THAT RBFANS ARE NOT AVAILABLE
AT OR PRIOR TO RVFAILURE
NOPDSFANS
CF
Reactor Building Fans AreNot Recovered At or Prior to
RV Failure
NORECOVFANSPRI
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 20 3/22/2007
Ex-Vessel Gas ProductionAfter RV Failure is High
EXVPRODAFTHPage 11
Corium Pool Does NotSpread Over Large Area Or
Freeze
NOCOREFREEZE
Corium Does Not SpreadAcross Lower Containment
Or Cavity Floor
NOSPREADLOW
Page 36
Likelihood Corium Does NotFreeze On Lower
Containment or Cavity Floor
NOFREEZELOW
Concrete Attack ProducesSufficient Combustible Gas
After RV Failure
NOH2SRCAFTER
Water Pool Does Not StopConcrete Attack In Cavity
After RV Failure
ATTKAFT
Water Pool In CavityUnavailable Early After RV
Failure
NOWATERAFTER
Page 26
Likelihood That Water Poolin Cavity Will Not Stop
Concrete Attack
MELT
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 21 3/22/2007
Cavity Recombination DoesNot Deplete Combustible Gas
Early After RV Failure
NOAFTERRECOMPage 11
Water Pool In Cavity AvailableEarly After RV Failure
WATERAFTER
Water Does Fill Cavity FromPlant Specific Sources And
Paths
OTHERWATER
Water Available From SpraysVia Fuel Transfer Pool Early
After RV Failure
FTRNSPOOLAFT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAY
Page 38
Containment Sprays AreRecovered Prior to RV Failure
RECOVSPPRI
Containment Sprays AreRecovered Early After RV
Failure
RECOVSPAFT
Accumulator Water isAvailable at RV Failure
ACCUMAVAIL
Primary System Pressure isHigh at Reactor Vessel
Failure
HIGHPRESSPage 8
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOW
Page 7
Operators Do NotDepressurize Steam
Generators
OPSNOSSHR
Steam GeneratorDepressurization Or SSHR Is
Unavailable
NOPORVSSHR
PDS INDICATES OTSGADVS ARE UNAVAILABLE
NOPDSSGADV
AV
Secondary Side HeatRemoval is Unavailable
SSHRUNAVAIL
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
NO SSHR EXISTS
NO-SSHR
Page 17
Prob. that Secondary SideHeat Removal is Not
Recovered Prior to RV Failure
NORECOVSSHR
Likelihood That Operators DoNot Depressurize Steam
Generators
NOOPSDEPRESS
Operators Do NotDepressurize with Pressurizer
PORV
NOPZRPORV
Page 13
Prob. that Failure of thePrimary System Does Not
Occur Due to Heating
NOHEATIML
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
BWST Water Gravity FeedInto Reactor Cavity Through
Failed Reactor Vessel
GRAVFEEDAFT
FAILURE OF ECCSINJECTION
PDSINJECCS
Page 22
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSH
Page 23
Likelihood ThatRecombination Cannot
Deplete Comb. Gas Given aDry Cavity
NODRYEFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 22 3/22/2007
FAILURE OF ECCSINJECTION
PDSINJECCSPage 21
CORE MELT BIN 1
CM-001
CORE MELT BIN 4
CM-004
CORE MELT BIN 7
CM-007
CORE MELT BIN 9
CM-009
CORE MELT BIN 12
CM-012
CORE MELT BIN 15
CM-015
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 23 3/22/2007
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSHPage 21Page 25
CSS/CIS A
A001
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS G
G001
CSS/CIS H
H001
CSS/CIS I
I001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 24 3/22/2007
RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE
SPARKAFTPage 11
RANDOM SPARK ISAVAILABLE EARLY AFTER
RV FAILURE WITH RBSPRAY
SPARKAFT-1
PROB THAT SPARK ISAVAILABLE EARLY AFTER
RV FAILURE WITH RBSPRAY
SPARKAFT_99
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
RANDOM SPARK ISAVAILABLE EARLY AFTERRV FAILURE WITHOUT RB
SPRAY
SPARKAFT-2
PROB THAT SPARK ISAVAILABLE EARLY AFTERRV FAILURE WITHOUT RB
SPRAY
SPARKAFT_1
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 4
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 25 3/22/2007
Containment Is Not SteamInerted After RV Failure
NOSTMINERTAFPage 11
Sequence After RV FailureHas Low Base Pressure In
Containment
AFTPRESSL
PDS HAS NO HIGH BASEPRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
NOSEQPRESSH
Page 23
Reactor Building FansAvailable Early After RV
Failure
FANSAFT
Page 36
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 26 3/22/2007
Cavity Recombination DoesDeplete Combustible Gas
Early After RV Failure
AFTERRECOMPage 1
Water Pool In CavityUnavailable Early After RV
Failure
NOWATERAFTERPage 20
Water Does Not Fill CavityFrom Plant Specific Sources
And Paths
NOOTHERWATER
Accumulator Water isUnavailable at RV Failure
ACCUMUNAVAIL
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 14
No BWST Water GravityFeed Into Reactor CavityThrough Failed Reactor
Vessel
NOGRAVFEEDAFT
NO FAILURE OF ECCSINJECTION
NOPDSINJECCS
Page 27
PDS INDICATES HIGHBASE PRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
SEQPRESSH
Page 28
Water Unavailable FromSprays Via Fuel Transfer
Pool Early After RV Failure
NOFTRNSPOOLAFT
Page 29
Likelihood ThatRecombination Can Deplete
Comb. Gas Given a DryCavity
DRYEFF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 27 3/22/2007
NO FAILURE OF ECCSINJECTION
NOPDSINJECCSPage 26
CORE MELT BIN 2
CM-002
CORE MELT BIN 5
CM-005
CORE MELT BIN 8
CM-008
CORE MELT BIN 11
CM-011
CORE MELT BIN 3
CM-003
CORE MELT BIN 6
CM-006
CORE MELT BIN 10
CM-010
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 28 3/22/2007
PDS INDICATES HIGHBASE PRESSURE IN
CONTAINMENT EARLYAFTER RV FAILURE
SEQPRESSHPage 26
CSS/CIS D
D001
CSS/CIS E
E001
CSS/CIS F
F001
CSS/CIS J
J001
CSS/CIS K
K001
CSS/CIS L
L001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 29 3/22/2007
Water Unavailable FromSprays Via Fuel Transfer
Pool Early After RV Failure
NOFTRNSPOOLAFTPage 26
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 4
Containment Sprays Are NotRecovered Prior to RV
Failure
NORECOVSPPRI
Containment Sprays Are NotRecovered Early After RV
Failure
NORECOVSPAFT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 30 3/22/2007
Cavity RecombinationDepletes Comb. Gas Prior to
Late Containment Failure
LTRECOMBPage 1
Likelihood ThatRecombination Depletes
Comb. Gas With a Dry CavityLate
DRYEFFLT
Water is Unavailable inCavity Prior to LCF
NOSTMWATER
Page 36
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 31 3/22/2007
Containment is SteamInerted Late After RV Failure
LATEINERTPage 1
Sequence Late After RVFailure Has High Steam
Concentration
LTPRESSH
Reactor Building Fans DoNot Function Prior to Late
Containment Failure
NOFANSLT
Reactor Building Fans AreUnavailable Prior to Late
Containment Failure
FANSUNAVAILLT
Reactor Building Fans DoNot Function Early After RV
Failure
NOFANSAFT
Reactor Building FansUnavailable Early After RV
Failure
FANSUNAVAILAFT
Reactor Building Fans DoNot Function at RV Failure
FANSPRI
Reactor Building Fans AreUnavailable at RV Failure
FANSUNAVAILPRI
Page 19
Likelihood RB Fans Do NotSurvive Containment
Enviroment At Or Prior ToRV Failu
NOEQUALFANSPRI
Reactor Building Fans AreNot Recovered Early After
RV Failure
NORECOVFANSAFT
Likelihood Fans Do NotSurvive Containment
Environment Early After RVFailure
NOEQUALFANSAF
RB FAN POWER SUPPLIESARE NOT RECOVEREDPRIOR TO LATE CTMT
FAILURE
NORECOVFANSLT
RB FANS UNAVAILABLEFOLLOWING POST-LOOP
RECOVERY
CF-R
OFFSITE POWER NOTRECOVERED WITHIN 24
HOURS
NORECOFFSITEPWR
Likelihood That RB Fans DoNot Survive Containment
Enviroment to Prevent LCF
NOEQUALFANSLT
PDS HAS NO LOW BASEPRESSURE IN
CONTAINMENT LATEAFTER RV FAILURE
NOSEQPRESSL
Page 32
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 32 3/22/2007
PDS HAS NO LOW BASEPRESSURE IN
CONTAINMENT LATEAFTER RV FAILURE
NOSEQPRESSLPage 31
CSS/CIS B
B001
CSS/CIS C
C001
CSS/CIS E
E001
CSS/CIS F
F001
CSS/CIS H
H001
CSS/CIS I
I001
CSS/CIS K
K001
CSS/CIS L
L001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 33 3/22/2007
Random Spark isUnavailable Late After RV
Failure
NOSPARKLTPage 1
SPARK UNAVAILABLE;OFFSITE POWER
RECOVERED
NOSPARKLT-1
RANDOM SPARKUNAVAILABLE WITH
OFFSITE POWER
NOSPARKLT-OP
OFFSITE POWERRECOVERED WITHIN 24
HOURS
RECOFFSITEPWR
SPARK UNAVAILABLE;OFFSITE POWER NOT
RECOVERED
NOSPARKLT-2
RANDOM SPARKUNAVAILABLE WITHOUT
OFFSITE POWER
NOSPARKLT-NOP
OFFSITE POWER NOTRECOVERED WITHIN 24
HOURS
NORECOFFSITEPWR
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 34 3/22/2007
Containment Strength CanHandle Late Combustible
Gas Burn Event
LTSTRENTPage 1
Containment Strength CanHandle Late Comb. Gas Burn
and Base Pressure is High
NOLTSTRENH
Sequence Late After RVFailure Has High Base
Pressure From GasGeneration
NOINERTLT
Likelihood That Cont CanHandle Comb. Gas Burn
Press. W/ High BasePressure
STREN1H2
Containment Strength CanHandle Late Comb. Gas Burn
and Base Pressure is Low
NOLTSTRENL
Sequence Late After RVFailure Has Low BasePressure From Steam
LTPRESSL
Reactor Building FansFunction Prior to LateContainment Failure
FANSLT
Reactor Building Fans AreAvailable Prior to LateContainment Failure
FANSAVAILLT
Page 36
Likelihood That RB FansSurvive Containment
Environment to Prevent LCF
EQUALFANSLT
PDS INDICATES LOW BASEPRESSURE IN
CONTAINMENT LATEAFTER RV FAILURE
SEQPRESSL
Page 35
Likelihood That Cont CanHandle Comb. Gas Burn
Press. W/ Low BasePressure
STREN2H2
Sequence Late After RVFailure Has Low Base
Pressure From GasGeneration
INERTLT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 35 3/22/2007
PDS INDICATES LOW BASEPRESSURE IN
CONTAINMENT LATEAFTER RV FAILURE
SEQPRESSLPage 34
CSS/CIS A
A001
CSS/CIS D
D001
CSS/CIS G
G001
CSS/CIS J
J001
CSS/CIS M
M001
CSS/CIS N
N001
CSS/CIS O
O001
CSS/CIS P
P001
CSS/CIS Q
Q001
CSS/CIS R
R001
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 36 3/22/2007
No Containment FailureFrom Steam Generation
NOSTEAMPage 1
There is Insufficient SteamProduced to Pressurize
Containment
NOSTMPROD
Insufficient Steam ProducedFrom Lower Containment
Area
NOLOWSTM
Corium Does Not SpreadAcross Lower Containment
Or Cavity Floor
NOSPREADLOWPage 20
Primary System Pressure isLow At RV Failure
LOWPRESS
Page 14
Cavity Geometry Does NotAllow Enough Corium toDisperse For Freezing
NOGEOMFREEZE
Water is Unavailable inCavity Prior to LCF
NOSTMWATERPage 30
Water Unavailable fromContainment Sprays Via Fuel
Transfer Pool Prior to LCF
NOFTRNSPOOLLT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE UNAVAILABLE
NOPDSSPRAY
Page 4
Containment Sprays Are NotRecovered Prior to RV
Failure
NORECOVSPPRI
RB SPRAY POWERSUPPLIES ARE NOT
RECOVERED PRIOR TOLATE CTMT FAILURE
NORECOVSPLT
RB SPRAY UNAVAILABLEDUE TO MECH FAILUREOR NO OFFSITE POWER
NORECOVSPLT-1
CS FAILURE FORINJECTION MODE
(POST-LOOP RECOVERY)
CS01-R
OFFSITE POWER NOTRECOVERED WITHIN 24
HOURS
NORECOFFSITEPWR
IE-LOOP-101
Containment Sprays Are NotRecovered Early After RV
Failure
NORECOVSPAFT
Recovery of Core CoolingDoes Prevent Reactor Vessel
Failure
RECOVRV
Reactor Building Fans AreAvailable Prior to LateContainment Failure
FANSAVAILLTPage 34
Reactor Building FansAvailable Early After RV
Failure
FANSAFTPage 25
Reactor Building FansAvailable Early After RV
Failure
FANSAVAILAFT
Reactor Building Fans DoFunction at RV Failure
NOFANSPRI
Reactor Building Fans AreAvailable at RV Failure
FANSAVAILPRI
Page 9
Likelihood RB Fans DoSurvive Containment
Enviroment At Or Prior ToRV Failu
EQUALFANSPRI
Reactor Building Fans AreRecovered Early After RV
Failure
RECOVFANSAFT
Likelihood Fans SurviveContainment Environment
Early After RV Failure
EQUALFANSAF
RB FAN POWER SUPPLIESARE RECOVERED PRIORTO LATE CTMT FAILURE
RECOVFANSLT
Page 37
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 37 3/22/2007
RB FAN POWER SUPPLIESARE RECOVERED PRIORTO LATE CTMT FAILURE
RECOVFANSLTPage 36
OFFSITE POWERRECOVERED WITHIN 24
HOURS
RECOFFSITEPWR
AVAILABILITY OF RB FANSWITHOUT POWER
DEPENDENCY
RECFANSLT
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 38 3/22/2007
No Containment FailureFrom Non Condensable
Gases
NOGASESPage 1
Water Is Available in CavityArea
STMWATER
Water Available fromContainment Sprays Via Fuel
Transfer Pool Prior to LCF
FTRNSPOOLLT
PDS INDICATES THATCONTAINMENT SPRAYS
ARE AVAILABLE
PDSSPRAYPage 21
RB SPRAY SYSTEM ISAVAILABLE
RBSPRAY
Containment Sprays AreRecovered Prior to RV
Failure
RECOVSPPRI
Containment Sprays AreRecovered Early After RV
Failure
RECOVSPAFT
RB SPRAY POWERSUPPLIES ARE
RECOVERED PRIOR TOLATE CTMT FAILURE
RECOVSPLT
OFFSITE POWERRECOVERED WITHIN 24
HOURS
RECOFFSITEPWR
AVAILABILITY OFCONTAINMENT SPRAYS
WITHOUT POWERDEPENDENCY
RECSPRAYLT
IE-LOOP-101
Likelihood That ContainmentCan Handle Pressure FromNon-Condensable Gases
NONCGASES
Likelihood That NonCondensable Gas Production
is Not High GIven a DryCavity
NONCGASHIGH
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 39 3/22/2007
Name Page Zone Name Page Zone
A001 10 1A001 23 1A001 35 1ACCUMAVAIL 21 5ACCUMUNAVAIL 26 2AFTBURN 1 7AFTBURN 11 7AFTERRECOM 1 6AFTERRECOM 26 3AFTPRESSL 25 2ATBURN 1 6ATCONC 14 3ATIGNITION 14 7ATPRESSH 14 9ATPRESSH 18 2ATPRESSL 1 8ATPRESSL 9 2ATRELEASE 14 3ATTKAFT 20 4AV 16 1AV 21 3B001 10 1B001 23 1B001 32 1BOTTLED 6 2BOTTLED 7 3BOTTLED 11 3BWHBW1-----HP2OA 13 4BWHBW1-----HP2OA 15 4C001 10 1C001 23 1C001 32 1CAG0005 17 1CAG0005-R 17 2CF 9 2CF 19 1CF-R 31 4CM-001 14 5CM-001 22 1
CM-002 14 6CM-002 27 1CM-003 14 6CM-003 27 1CM-004 14 8CM-004 22 1CM-005 14 9CM-005 27 1CM-006 14 8CM-006 27 1CM-007 12 1CM-007 22 1CM-008 12 1CM-008 27 1CM-009 12 1CM-009 22 2CM-010 12 1CM-010 27 2CM-011 12 1CM-011 27 1CM-012 12 1CM-012 22 2CM-013 12 2CM-013 27 2CM-014 12 2CM-014 27 2CM-015 12 2CM-015 22 2CM-016 12 2CM-016 27 2CM-018 12 2CM-018 27 2CM-019 12 2CM-019 27 2CS01 4 1CS01-R 36 4D001 18 1D001 28 1D001 35 1
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 40 3/22/2007
Name Page Zone Name Page Zone
DISPERSE 8 2DRYEFF 26 4DRYEFFLT 30 1E001 18 1E001 28 1E001 32 1EQUALFANSAF 36 9EQUALFANSLT 34 3EQUALFANSPRI 36 8EXVPRODAFTH 11 5EXVPRODAFTH 20 3F001 18 1F001 28 1F001 32 1FANSAFT 25 2FANSAFT 36 8FANSAT 9 2FANSAVAILAFT 36 8FANSAVAILLT 34 2FANSAVAILLT 36 8FANSAVAILPRI 9 2FANSAVAILPRI 36 7FANSLT 34 3FANSPRI 31 2FANSUNAVAILAFT 31 2FANSUNAVAILLT 31 3FANSUNAVAILPRI 18 3FANSUNAVAILPRI 19 2FANSUNAVAILPRI 31 1FTRNSPOOLAFT 21 2FTRNSPOOLLT 38 3G001 10 1G001 23 1G001 35 1GEOMH2 14 5GRAVFEEDAFT 21 6H001 10 1H001 23 1H001 32 2
HEATIML 14 4HEATIML 14 9HIGHPRESS 8 2HIGHPRESS 21 5I001 10 1I001 23 1I001 32 2IE-LOOP-100 17 3IE-LOOP-101 36 6IE-LOOP-101 38 5INERTLT 34 5J001 18 2J001 28 2J001 35 1K001 18 2K001 28 2K001 32 2L001 18 2L001 28 2L001 32 2LATEINERT 1 8LATEINERT 31 4LOWCONCBURN 1 8LOWCONCBURN 11 1LOWCONCBURN 14 4LOWPRESS 14 8LOWPRESS 26 2LOWPRESS 36 1LTPRESSH 31 4LTPRESSL 34 3LTPRIGLOB 1 5LTRECOMB 1 7LTRECOMB 30 2LTSTRENT 1 9LTSTRENT 34 3M001 10 2M001 23 2M001 35 1MELT 20 4
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 41 3/22/2007
Name Page Zone Name Page Zone
N001 10 2N001 23 2N001 35 2NO-SSHR 16 2NO-SSHR 17 2NO-SSHR 21 4NO-SSHR-POSTLOOP 17 2NOAFTERCONC 11 6NOAFTERRECOM 11 6NOAFTERRECOM 21 4NOAFTERREL 11 5NOATBURN 11 5NOATBURN 14 6NOATCONC 1 6NOATIGNITION 1 7NOATIGNITION 8 2NOATRELEASE 1 5NOATRELEASE 6 2NOCOREFREEZE 20 2NODISPERSE 14 7NODRYEFF 21 5NOEQUALFANSAF 31 3NOEQUALFANSLT 31 4NOEQUALFANSPRI 31 2NOFANSAFT 31 2NOFANSLT 31 4NOFANSPRI 36 7NOFREEZELOW 20 2NOFTRNSPOOLAFT 26 5NOFTRNSPOOLAFT 29 2NOFTRNSPOOLLT 36 5NOGASES 1 10NOGASES 38 3NOGEOMFREEZE 36 2NOGEOMH2 8 1NOGRAVFEEDAFT 26 4NOH2LATE 1 8NOH2SRCAFTER 20 3NOHEATIML 7 3
NOHEATIML 21 6NOINERTLT 34 1NOLATE 1 9NOLOWCONCBURN 1 1NOLOWCONCBURN 1 6NOLOWCONCBURN 11 7NOLOWSTM 36 1NOLTBURN 1 8NOLTCONC 1 7NOLTSTRENH 34 2NOLTSTRENL 34 4NONCGASES 38 3NONCGASHIGH 38 4NOOPSDEPRESS 21 5NOOTHERWATER 26 1NOOXIDIZED 2 1NOOXIDIZED 6 1NOPDSFANS 19 1NOPDSINJECCS 26 3NOPDSINJECCS 27 2NOPDSLOW 7 2NOPDSLOW 21 3NOPDSLOW-1 7 2NOPDSLOW-2 7 1NOPDSLOW-2 12 2NOPDSLOW_5 7 2NOPDSPRESSH 9 1NOPDSPRESSH 10 2NOPDSPZRPORV 13 1NOPDSSGADV 21 3NOPDSSPRAY 3 2NOPDSSPRAY 4 1NOPDSSPRAY 11 5NOPDSSPRAY 24 4NOPDSSPRAY 29 1NOPDSSPRAY 36 3NOPDSSSHR 21 4NOPORVSSHR 21 4NOPRIBURN 6 1
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node "NOLATE"(2004 Rev. 2)
Page 42 3/22/2007
Name Page Zone Name Page Zone
NOPRIBURN 11 4NOPRICONC 11 2NOPRIGLOBAL 11 4NOPRIRELEASE 11 2NOPRVHPCONF 13 2NOPZPORVCONF-1 13 3NOPZPORVCONF-2 13 4NOPZRPORV 7 4NOPZRPORV 13 2NOPZRPORV 21 5NOPZRSAFETY 7 5NOPZRSAFETY 21 7NORECOFFSITEPWR 31 5NORECOFFSITEPWR 33 4NORECOFFSITEPWR 36 5NORECOVFANSAFT 31 3NORECOVFANSLT 31 4NORECOVFANSPRI 19 2NORECOVRV 6 2NORECOVRV 20 5NORECOVSPAFT 29 2NORECOVSPAFT 36 6NORECOVSPLT 36 5NORECOVSPLT-1 36 5NORECOVSPPRI 29 2NORECOVSPPRI 36 4NORECOVSSHR 21 5NOSEQPRESSH 21 7NOSEQPRESSH 23 2NOSEQPRESSH 25 1NOSEQPRESSL 31 5NOSEQPRESSL 32 2NOSPARK 11 5NOSPARK-1 11 4NOSPARK-2 11 6NOSPARKAT 14 8NOSPARKLT 1 9NOSPARKLT 33 2NOSPARKLT-1 33 2
NOSPARKLT-2 33 4NOSPARKLT-NOP 33 3NOSPARKLT-OP 33 1NOSPARK_01 11 6NOSPARK_9 11 4NOSPREADLOW 20 1NOSPREADLOW 36 2NOSTEAM 1 9NOSTEAM 36 6NOSTMINERTAF 11 8NOSTMINERTAF 25 1NOSTMINERTP 1 5NOSTMINERTP 1 8NOSTMPROD 36 3NOSTMWATER 30 2NOSTMWATER 36 4NOTBOTTLED 2 2NOTBOTTLED 14 4NOWATERAFTER 20 3NOWATERAFTER 26 3O001 10 2O001 23 2O001 35 2OPSDEPRESS 16 3OPSNOSSHR 21 4OPSSSHR 14 8OPSSSHR 16 2OTHERWATER 21 1OXIDIZED 11 2OXIDIZED 14 2P001 10 2P001 23 2P001 35 2PDSFANS 9 2PDSINDET 7 3PDSINDET 14 8PDSINJECCS 21 6PDSINJECCS 22 2PDSLOW 14 4
Name Page Zone Name Page Zone
PDSLOW 14 6PDSLOW-1 14 6PDSLOW-2 14 7PDSLOW_5 14 7PDSPRESSH 18 2PDSPZRPORV 15 1PDSSGADV 16 1PDSSPRAY 21 2PDSSPRAY 38 1PDSSSHR 16 2PO-HPICOOL 13 1PO-HPICOOL 15 1PORVSSHR 16 2PRIBURN 1 3PRIBURN 14 1PRICONC 1 2PRIRELEASE 1 2PRIRELEASE 2 2PRVHPCONF 15 4PZPORVCONF-1 15 3PZPORVCONF-2 15 3PZPORVCONF_0 15 3PZPORVCONF_0-C 13 3PZPORVCONF_99 15 2PZPORVCONF_99-C 13 3PZRPORV 14 5PZRPORV 14 7PZRPORV 15 2PZRSAFETY 14 5PZRSAFETY 14 10Q001 10 2Q001 23 2Q001 35 2R001 10 2R001 23 2R001 35 2RBSPRAY 5 2RBSPRAY 11 7RBSPRAY 24 2
RBSPRAY 38 1RECFANSLT 37 2RECOFFSITEPWR 33 2RECOFFSITEPWR 37 1RECOFFSITEPWR 38 4RECOVFANSAFT 36 8RECOVFANSLT 36 9RECOVFANSLT 37 2RECOVFANSPRI 9 3RECOVRV 14 3RECOVRV 36 4RECOVSPAFT 21 3RECOVSPAFT 38 3RECOVSPLT 38 4RECOVSPPRI 21 2RECOVSPPRI 38 2RECOVSSHR 16 3RECSPRAYLT 38 4SEQPRESSH 26 4SEQPRESSH 28 2SEQPRESSL 34 4SEQPRESSL 35 2SPARK 1 4SPARK-1 1 3SPARK-1 3 2SPARK-2 1 4SPARK-2 5 2SPARKAFT 11 7SPARKAFT 24 2SPARKAFT-1 24 2SPARKAFT-2 24 4SPARKAFT_1 24 3SPARKAFT_99 24 1SPARKAT 8 3SPARK_1 3 1SPARK_99 5 1SSHRAVAIL 16 3SSHRUNAVAIL 21 4STMINERTP 11 6
Name Page Zone Name Page Zone
STMINERTP 14 9STMWATER 38 2STREN1H2 34 2STREN2H2 34 4WATERAFTER 21 4YES-SSHR 16 2
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOLATEREVAP" (2004 Rev. 2)
Page 1 3/22/2007
Late Fission ProductRevaporization Release
NOLATEREVAP
Sufficient Fission ProductHeating to Cause
Revaporization
FPHEATING
Heat Losses From PrimarySystem Are Not Large
NOHEATLOSS
Amount of Fission ProductsRetained in Primary System
is Not Small
NOFPAMOUNT
Primary Retention is Low ForLow Pressure Core Melt
NOAMTLOWP
Likelihood That Retention IsNot Low for a Low Pressure
Core Melt
LPCMEFF
Primary System Pressure isLow Prior to Core Melt
LOWPRESPCM
Prob. that Pressurizer SafetyValves Stick Open During
Core Damage
PZRSAFETY
PDS INDICATESSEQUENCE IS A LOW
PRESSURE CORE MELT
PDSLOW
PDS INDICATES LOWPRESSURE AT CORE MELT
IS CERTAIN
PDSLOW-1
CORE MELT BIN 1
CM-001
CORE MELT BIN 2
CM-002
CORE MELT BIN 3
CM-003
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
PDSLOW-2
LOW PRESSURE AT COREMELT IS INDETERMINATE
PDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
Page 2
Operators Depressurize WithPressurizer PORV
PZRPORV
Page 3
Primary Retention is Low ForHigh Pressure Core Melt
NOAMTHIGHP
Primary System Pressure isHigh Prior to Core Melt
HIGHPRESPCM
Page 4
Likelihood That Retention IsLow for a High Pressure Core
Melt
HPCMEFF
Chemical Form of FissionProducts Does Not Have High
Vaporization Temperature
NOCHEMICAL
Secondary Side HeatRemoval Does Not Prevent
Revaporization
NOSSHR
Page 6
Recovery of Core CoolingDoes Not Prevent Reactor
Vessel Failure
NORECOVRV
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOLATEREVAP" (2004 Rev. 2)
Page 2 3/22/2007
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDETPage 1Page 4
CORE MELT BIN 4
CM-004
CORE MELT BIN 6
CM-006
CORE MELT BIN 5
CM-005
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOLATEREVAP" (2004 Rev. 2)
Page 3 3/22/2007
Operators Depressurize WithPressurizer PORV
PZRPORVPage 1
PDS INDICATES PORV ISAVAILABLE
PDSPZRPORV
PO-HPICOOL
LOGIC FOR OPERATORSOPENING PORV
PZPORVCONF-1
PROB THAT OPERATORSOPEN PORV
PZPORVCONF_99
LOGIC FOR OPERATORSOPENING PORV WHEN HPI
COOLING IS NOTINITIATED
PZPORVCONF-2
Prob. That Operators OpenPORV After Failing to Init
HPI Cooling
PZPORVCONF_0
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
PROBABILITY THAT PORVCAN PREVENT HPME
PRVHPCONF
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOLATEREVAP" (2004 Rev. 2)
Page 4 3/22/2007
Primary System Pressure isHigh Prior to Core Melt
HIGHPRESPCMPage 1
Prob. that Pressurizer SafetyValves Do Not Stick Open
During Core Damage
NOPZRSAFETY
PDS INDICATESSEQUENCE IS A HIGH
PRESSURE CORE MELT
NOPDSLOW
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2
Page 5
PDS INDICATESINDETERMINATE
PRESSURE AT CORE MELT
NOPDSLOW-1
HIGH PRESSURE AT COREMELT IS INDETERMINATE
NOPDSLOW_5
PDS INVOLVESINDETERMINATE
PRESSURE AT CORE MELT
PDSINDET
Page 2
Operators Do NotDepressurize withPressurizer PORV
NOPZRPORV
PDS INDICATES PORV ISUNAVAILABLE
NOPDSPZRPORV
PO-HPICOOL
PROBABILITY THAT PORVDOES NOT PREVENT
HPME
NOPRVHPCONF
LOGIC FOR OPERATORSFAILING TO OPEN PORV
NOPZPORVCONF-1
PROBABILITY THATOPERATORS FAIL TO
MANUALLY OPEN PORV
PZPORVCONF_99-C
LOGIC FOR OPERATORSFAILING TO OPEN PORVWHEN HPI COOLING IS
NOT INITIATED
NOPZPORVCONF-2
PROB THAT OPERATORSFAIL TO OPEN PORV
AFTER FAILING TO INITHPI COOLING
PZPORVCONF_0-C
OPERATOR FAILS TOINITIATE HPI COOLING
BWHBW1-----HP2OA
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOLATEREVAP" (2004 Rev. 2)
Page 5 3/22/2007
PDS SHOWS HIGHPRESSURE AT CORE MELT
IS CERTAIN
NOPDSLOW-2Page 4
CORE MELT BIN 7
CM-007
CORE MELT BIN 8
CM-008
CORE MELT BIN 9
CM-009
CORE MELT BIN 10
CM-010
CORE MELT BIN 11
CM-011
CORE MELT BIN 12
CM-012
CORE MELT BIN 13
CM-013
CORE MELT BIN 14
CM-014
CORE MELT BIN 15
CM-015
CORE MELT BIN 16
CM-016
CORE MELT BIN 18
CM-018
CORE MELT BIN 19
CM-019
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOLATEREVAP" (2004 Rev. 2)
Page 6 3/22/2007
Secondary Side HeatRemoval Does Not Prevent
Revaporization
NOSSHRPage 1
Secondary Side HeatRemoval is Unavailable Prior
to Revaporization
SSHRRVPUNAVL
PDS INDICATES SSHR ISUNAVAILABLE
NOPDSSSHR
NO SSHR EXISTS
NO-SSHR
CAG0005
NO SSHR FOR POST-LOOPRECOVERY CONDITIONS
NO-SSHR-POSTLOOP
CAG0005-R IE-LOOP-100
Secondary Side HeatRemoval is Not Recovered
Prior to Revaporization
SSHRNORVPREC
Likelihood That SSHR WillNot Prevent Revaporization
SSHRNOREVAP
TITLE
PAGE NUMBER DATE
TMI Level 2 Logic for Node"NOLATEREVAP" (2004 Rev. 2)
Page 7 3/22/2007
Name Page Zone Name Page Zone
BWHBW1-----HP2OA 3 4BWHBW1-----HP2OA 4 7CAG0005 6 1CAG0005-R 6 2CM-001 1 1CM-002 1 2CM-003 1 2CM-004 2 1CM-005 2 2CM-006 2 2CM-007 5 1CM-008 5 1CM-009 5 1CM-010 5 1CM-011 5 1CM-012 5 1CM-013 5 2CM-014 5 2CM-015 5 2CM-016 5 2CM-018 5 2CM-019 5 2FPHEATING 1 2HIGHPRESPCM 1 3HIGHPRESPCM 4 3HPCMEFF 1 4IE-LOOP-100 6 3LOWPRESPCM 1 2LPCMEFF 1 1NO-SSHR 6 2NO-SSHR-POSTLOOP 6 2NOAMTHIGHP 1 4NOAMTLOWP 1 2NOCHEMICAL 1 3NOFPAMOUNT 1 3NOHEATLOSS 1 1NOLATEREVAP 1 3NOPDSLOW 4 2NOPDSLOW-1 4 3
NOPDSLOW-2 4 2NOPDSLOW-2 5 2NOPDSLOW_5 4 2NOPDSPZRPORV 4 4NOPDSSSHR 6 1NOPRVHPCONF 4 5NOPZPORVCONF-1 4 6NOPZPORVCONF-2 4 7NOPZRPORV 4 5NOPZRSAFETY 4 1NORECOVRV 1 5NOSSHR 1 4NOSSHR 6 2PDSINDET 1 4PDSINDET 2 2PDSINDET 4 3PDSLOW 1 2PDSLOW-1 1 2PDSLOW-2 1 4PDSLOW_5 1 3PDSPZRPORV 3 1PO-HPICOOL 3 1PO-HPICOOL 4 4PRVHPCONF 3 4PZPORVCONF-1 3 3PZPORVCONF-2 3 3PZPORVCONF_0 3 3PZPORVCONF_0-C 4 6PZPORVCONF_99 3 2PZPORVCONF_99-C 4 6PZRPORV 1 3PZRPORV 3 2PZRSAFETY 1 1SSHRNOREVAP 6 3SSHRNORVPREC 6 2SSHRRVPUNAVL 6 2
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