Three Mile Island Unit 1 Level 2 Containment Event Tree ... · Three Mile Island Level 2...

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


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


with Ex-Vessel Release of Fission Products, with Fission Product Scrubbing .... 6-60

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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, with Ex-Vessel Release of Fission Products, with Fission Product

Scrubbing ............................................................................................................ 6-64


Building, with Ex-Vessel Release of Fission Products, without Fission


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






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Building, with Ex-Vessel Release of Fission Products, with Fission Product

Scrubbing ............................................................................................................ 6-70


Building, with Ex-Vessel Release of Fission Products, without Fission


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


without Ex-Vessel Release of Fission Products, without Fission Product

Scrubbing ............................................................................................................ 6-73



Scrubbing ............................................................................................................ 6-74



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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Revaporization, without Fission Product Scrubbing ............................................ 6-79


Containment Failure, without Ex-Vessel Fission Product Release, with






Containment Failure, with Ex-Vessel Release of Fission Products, without






Containment Failure, with Ex-Vessel Release of Fission Products, with





Table 6-34 Release Category 7.01 Late Overpressurization, with Benign


Fission Product Scrubbing .................................................................................. 6-86

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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

Three Mile Island PRA - L2 Containment Event Tree Analysis

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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.


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.


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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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).


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



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).


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.


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


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


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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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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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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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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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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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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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)


with Low Base Pressure Early After Reactor Vessel Failure

Description:


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:


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:


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:


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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5-31 0467060030-2788-050107

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:


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:


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:


fans for TMI-1 and Oconee. The Oconee RB fans are expected to remain functional


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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:


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:


fans for TMI-1 and Oconee. The Oconee RB Fans are expected to remain functional



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


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:


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:


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:


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


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


OXIDIZED = 0.99

NOOXIDIZED = 0.01

PDSFANS (NOPDSFANS)

PDS Indicates that Reactor Building Fans Are Available (Unavailable) at or Prior to


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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5-99 0467060030-2788-050107

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:


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:


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:


given a hydrogen burn prior to reactor vessel failure with low containment base

pressure.

Quantification:


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:




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



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:




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


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:




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




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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.




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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


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:


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:


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


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


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


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)


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:


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)


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:


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.


Description:


outside the Auxiliary Building. Fission product scrubbing is not available.


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.


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.


Description:


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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.


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.


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.


Description:


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into the Auxiliary Building. Ex-vessel fission products are not released and scrubbing is

not available.



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.


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.




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.


Description:


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into the Auxiliary Building. Ex-vessel fission products are released, but scrubbing is

available.






assumed. Since the release is low in the Auxiliary Building credit is taken for fission

product plateout.


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.




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.


Description:


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into the Auxiliary Building. Ex-vessel fission products are released and scrubbing is not

available.






assumed. No credit is taken for fission product plateout.


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.




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.


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.


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.


There are currently no TMI-1 sequences involving this release category.


Description:


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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).


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.




Description:


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Auxiliary Building. Ex-vessel fission products are released and fission product

scrubbing is available via significant plateout in the Auxiliary Building.


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.




Description:


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scrubbing is not available, e.g., very little fission product plateout prior to release to the

environment.


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.


surge line occurs due to overheating. This induced LOCA will allow depressurization of





fission products will be released. The characteristics of this category are listed in Table

6-13.




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.


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.




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.


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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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.


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.



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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.


Description:


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.


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.




corium over a large area, the debris bed will be coolable preventing the release of ex-


6-17.



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Description:

This release category is characterized by a small isolation failure with release to the


scrubbing is available via significant plateout in the Auxiliary Building.


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.








in Table 6-18.


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Description:



scrubbing is not available, i.e., very little fission product plateout prior to release to the

environment.






provides a fission product release path to the Auxiliary Building. However, no credit is

taken for fission product plateout.







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in Table 6-19.







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.






provides a release path directly to the environment. Plateout within HVAC systems

allows significant fission product scrubbing.




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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.




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.






provides a release path directly to the environment. With containment safeguards

failed, fission products will not be scrubbed.




corium over a large area, the debris bed will be coolable, preventing the release of ex-


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6-21.




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.






provides a release path directly to the environment. Plateout within HVAC systems

allows significant fission product scrubbing.


surge line occurs due to overheating. The induced LOCA will allow depressurization of






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in Table 6-22.




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.






provides a release path directly to the environment. With failed containment

safeguards, fission products will not be scrubbed.


surge line occurs due to overheating. The induced LOCA will allow depressurization of






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in Table 6-23.




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.


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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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.


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.





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.


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.


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.


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



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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Description:



ex-vessel release of fission products, no revaporization of fission products and no

fission product scrubbing.



with a recirculation failure and a failure of all containment safeguards.


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.










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results.


Description:



ex-vessel release of fission products, and fission product scrubbing is available. There

is a revaporization release of fission products for this release category.


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.




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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results.


Description:



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.


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.





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The small LOCA allows a great deal of fission product retention within the primary










results.


Description:


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.



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.














results.


Description:



fission product scrubbing and no revaporization releases. However, ex-vessel fission

products will be released.


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6-34 0467060030-2788-050107



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.







containment safeguards, this will lead to a catastrophic failure of containment. The







results.


Description:


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6-35 0467060030-2788-050107


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.


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


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.



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6-36 0467060030-2788-050107





results.


Description:



fission product scrubbing but ex-vessel fission products will be released. A

revaporization release is also assumed for this release category.


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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6-37 0467060030-2788-050107

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.






results.


Description:


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.



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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





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.


Description:


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








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6-39 0467060030-2788-050107









Description:


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











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6-40 0467060030-2788-050107






Description:


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















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6-41 0467060030-2788-050107


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.





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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6-42 0467060030-2788-050107

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 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.


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.


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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6-43 0467060030-2788-050107

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.


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.


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.


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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6-44 0467060030-2788-050107






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.



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.



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6-45 0467060030-2788-050107





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.



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.






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6-46 0467060030-2788-050107

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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6-47 0467060030-2788-050107

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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6-48 0467060030-2788-050107

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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6-49 0467060030-2788-050107

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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6-50 0467060030-2788-050107

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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6-51 0467060030-2788-050107

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).


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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



Total Plume 2 Release Fraction2 6.50E-01



3) TeO2


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6-53 0467060030-2788-050107









4) SrO




Total Plume 2 Release Fraction2

Start of Plume 2 Release (hr)

End of Plume 2 Release (hr)

5) MoO2








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6-54 0467060030-2788-050107



6) CsOH







7) BaO







8) La2O3







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6-55 0467060030-2788-050107




9) CeO2







10) Sb







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



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6-56 0467060030-2788-050107





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




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


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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


CONTAINMENT BYPASS, OUTSIDE THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS






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6-58 0467060030-2788-050107


CONTAINMENT BYPASS, OUTSIDE THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS


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


CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING,WITHOUT EX-VESSEL RELEASE OF FISSION

PRODUCTS, WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:





Elevation of Release 0.0 meters RELEASE FRACTIONS:

Xe 1.0E+00

I 1.7E-01

Cs-Rb 1.7E-01


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6-59 0467060030-2788-050107


CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING,WITHOUT EX-VESSEL RELEASE OF FISSION


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


CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS,

WITHOUT FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:






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)


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6-60 0467060030-2788-050107


CONTAINMENT BYPASS, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS,

WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:






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)









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6-61 0467060030-2788-050107





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


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6-62 0467060030-2788-050107


LARGE ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION







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


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6-63 0467060030-2788-050107


LARGE ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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



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6-64 0467060030-2788-050107


LARGE ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITH FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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



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6-65 0467060030-2788-050107


LARGE ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION

PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING RELEASE PARAMETERS






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)


_____________________________________________________________________

6-66 0467060030-2788-050107


LARGE ISOLATION FAILURE, OUTSIDE THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION

PRODUCTS RELEASE PARAMETERS:






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


_____________________________________________________________________

6-67 0467060030-2788-050107


LARGE ISOLATION FAILURE, OUTSIDE THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION







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


_____________________________________________________________________

6-68 0467060030-2788-050107


SMALL ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION







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.


_____________________________________________________________________

6-69 0467060030-2788-050107


SMALL ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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)


_____________________________________________________________________

6-70 0467060030-2788-050107


SMALL ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITH FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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)


_____________________________________________________________________

6-71 0467060030-2788-050107


SMALL ISOLATION FAILURE, TO THE AUXILIARY BUILDING, WITH EX-VESSEL

RELEASE OF FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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.


_____________________________________________________________________

6-72 0467060030-2788-050107


SMALL ISOLATION FAILURE, TO THE ENVIRONMENT, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS,

WITH FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:






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



_____________________________________________________________________

6-73 0467060030-2788-050107


SMALL ISOLATION FAILURE, TO THE ENVIRONMENT, WITHOUT EX-VESSEL RELEASE OF FISSION PRODUCTS,







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


_____________________________________________________________________

6-74 0467060030-2788-050107


SMALL ISOLATION FAILURE, TO THE ENVIRONMENT, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS,







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



_____________________________________________________________________

6-75 0467060030-2788-050107


SMALL ISOLATION FAILURE, TO THE ENVIRONMENT, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS,







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


_____________________________________________________________________

6-76 0467060030-2788-050107


EARLY CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE

RELEASE PARAMETERS:






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.


_____________________________________________________________________

6-77 0467060030-2788-050107


EARLY CONTAINMENT FAILURE, WITH EX-VESSEL FISSION PRODUCT RELEASE

RELEASE PARAMETERS:






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.


_____________________________________________________________________

6-78 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE, WITHOUT REVAPORIZATION, WITH

FISSION PRODUCT SCRUBBING RELEASE PARAMETERS:






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



_____________________________________________________________________

6-79 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION

PRODUCT RELEASE, WITHOUT REVAPORIZATION, WITHOUT FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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.


_____________________________________________________________________

6-80 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION

PRODUCT RELEASE, WITH REVAPORIZATION, WITH FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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



_____________________________________________________________________

6-81 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE, WITH REVAPORIZATION, WITHOUT







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.


_____________________________________________________________________

6-82 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT REVAPORIZATION, WITH







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



_____________________________________________________________________

6-83 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE,

WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITHOUT REVAPORIZATION, WITHOUT FISSION

PRODUCT SCRUBBING RELEASE PARAMETERS:






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



_____________________________________________________________________

6-84 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE,

WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITH REVAPORIZATION, WITH FISSION PRODUCT

SCRUBBING RELEASE PARAMETERS:






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



_____________________________________________________________________

6-85 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH CATASTROPHIC CONTAINMENT FAILURE, WITH EX-VESSEL RELEASE OF FISSION PRODUCTS, WITH REVAPORIZATION, WITHOUT







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



_____________________________________________________________________

6-86 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH BENIGN CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION

PRODUCT RELEASE, WITH FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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



_____________________________________________________________________

6-87 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH BENIGN CONTAINMENT FAILURE, WITHOUT

EX-VESSEL FISSION PRODUCT RELEASE, WITHOUT FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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.


_____________________________________________________________________

6-88 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH BENIGN CONTAINMENT FAILURE, WITH EX-VESSEL RELEASE OF

FISSION PRODUCTS, WITH FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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



_____________________________________________________________________

6-89 0467060030-2788-050107


LATE OVERPRESSURIZATION, WITH BENIGN CONTAINMENT FAILURE, WITH EX-VESSEL RELEASE OF

FISSION PRODUCTS, WITHOUT FISSION PRODUCT SCRUBBING

RELEASE PARAMETERS:






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.


_____________________________________________________________________

6-90 0467060030-2788-050107


CONTAINMENT FAILURE FROM BASEMAT MELT-THROUGH, WITH EX-VESSEL RELEASE OF FISSION





Energy of Release 0.0 Watts


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.


_____________________________________________________________________

6-91 0467060030-2788-050107


NO CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE, WITH FISSION PRODUCT







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


_____________________________________________________________________

6-92 0467060030-2788-050107


NO CONTAINMENT FAILURE, WITHOUT EX-VESSEL FISSION PRODUCT RELEASE, WITHOUT FISSION

PRODUCT SCRUBBING RELEASE PARAMETERS:






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


_____________________________________________________________________

6-93 0467060030-2788-050107


NO CONTAINMENT FAILURE, WITH EX-VESSEL FISSION PRODUCT RELEASE, WITH FISSION PRODUCT







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



_____________________________________________________________________

6-94 0467060030-2788-050107


NO CONTAINMENT FAILURE, WITH EX-VESSEL FISSION PRODUCT RELEASE, WITHOUT FISSION PRODUCT







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


_____________________________________________________________________

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%


_____________________________________________________________________

7-2 0467060030-2788-050107




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%


_____________________________________________________________________

7-3 0467060030-2788-050107




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%


_____________________________________________________________________

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


_____________________________________________________________________

8-1 0467060030-2788-050107

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.


_____________________________________________________________________

8-2 0467060030-2788-050107

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.


_____________________________________________________________________

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


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


Page 2 3/22/2007


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


Page 3 3/22/2007


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


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


(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


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


Page 2 3/22/07


NOPDSSSHRPage 1

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP


TITLE

PAGE NUMBER DATE


Page 3 3/22/07


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


Page 4 3/22/07


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


Page 5 3/22/07


OPEN PORV

NOEFFDEPRESS-2Page 4


BWHBW1-----HP2OA

OPERATOR FAILS TOOPEN PORV AFTER

FAILING TO INITIATE HPICOOLING

EFFDEPRESS_0-C

TITLE

PAGE NUMBER DATE


Page 6 3/22/07


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


RCPUMPONPage 3

Power To The ReactorCoolant Pumps (RCPs)

RECOVPOWER

PDS INDICATES POWER ISAVAILABLE TO RCPS

PDSRCPWR


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


S/G Tube Failure W/ RCPsOff

NOEFFNPMP


RCPUMPOFF

Page 3

TITLE

PAGE NUMBER DATE


Page 7 3/22/07


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


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


Page 2 3/22/07


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


Page 3 3/22/07


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


FANSUNAVAILPRI

Page 2

TITLE

PAGE NUMBER DATE


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


Page 5 3/22/07


RV Failure

ATPRESSLPage 31Page 11Page 8

... see x-ref

PDS DOES NOT HAVEHIGH BASE PRESSURE IN


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


Page 6 3/22/07


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


SSHRUNAVAIL


NOPDSSSHR

NO SSHR EXISTS

NO-SSHR

Page 21



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


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


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


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


RV Failure

ATPRESSL

Page 5

Likelihood That Cont. CannotHandle H2 Burn Press W/

Low Base Pressure

NOPRISTREN2

TITLE

PAGE NUMBER DATE


Page 9 3/22/07


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


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


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


Page 10 3/22/07


NOSTMINERTPPage 9

Page 18


RV Failure

ATPRESSL

Page 5

TITLE

PAGE NUMBER DATE


Page 11 3/22/07

Containment StrengthCannot Handle H2 Burns

Event at RV Failure

NOATSTRENTPage 9

Containment StrengthCannot Handle H2 Burns and


NOATSTRENH


RV Failure

ATPRESSH

Page 3

Likelihood That Cont CannotHandle H2 Burn Press. W/

High Base Pressure

NOATSTREN1

Containment StrengthCannot Handle H2 Burns and


NOATSTRENL


RV Failure

ATPRESSL

Page 5

Likelihood That Cont CannotHandle H2 Burn Press. W/

Low Base Pressure

NOATSTREN2

TITLE

PAGE NUMBER DATE


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


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


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



LOWCONCBURN

No Ignition Source isAvailable at RV Failure

ATIGNITION

Page 20


STMINERTPPage 12


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


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


Page 13 3/22/07


Before RV Failure

NOPRIRELEASEPage 12

In-Vessel H2 Prod. NotSufficient to Cause H2 Burns

OXIDIZED


BOTTLEDPage 9


PRESSURE CORE MELT

NOPDSLOWPage 6

PDS SHOWS HIGHPRESSURE AT CORE MELT

IS CERTAIN

NOPDSLOW-2

Page 14



NOPDSLOW-1

HIGH PRESSURE AT COREMELT IS INDETERMINATE

NOPDSLOW_5



PDSINDET

Page 20



NOHEATIML


NOPZRPORV

Page 15


During Core Damage

NOPZRSAFETY

TITLE

PAGE NUMBER DATE


Page 14 3/22/07


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


Page 15 3/22/07


NOPZRPORVPage 13Page 6


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


BWHBW1-----HP2OA

TITLE

PAGE NUMBER DATE


Page 16 3/22/07


UNAVAILABLE

NOSPARK-1Page 12

PROB THAT SPARK IS NOTAVAILABLE BEFORE RVFAILURE WITHOUT RB

SPRAY

NOSPARK_9


ARE UNAVAILABLE

NOPDSSPRAYPage 28Page 22Page 18


CS01

TITLE

PAGE NUMBER DATE


Page 17 3/22/07


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


Page 18 3/22/07


RV Failure

ATRELEASEPage 12


PRIBURNPage 7

H2 Concentration isSufficient to Cause Burns

Before RV Failure

PRICONC


Comb. Gas

NOLOWCONCBURN

Sufficient Hydrogen isReleased to Containment

Before RV Failure

PRIRELEASE


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


ARE UNAVAILABLE

NOPDSSPRAY

Page 16


AVAILABLE

SPARK-2

PROB THAT SPARK ISAVAILABLE BEFORE RV


SPARK_99


RBSPRAY


NOSTMINERTP

Page 10


OXIDIZED

Recovery of Core CoolingDoes Prevent Reactor

Vessel Failure

RECOVRV


NOTBOTTLEDPage 18


PRESSURE CORE MELT

PDSLOW

Page 20


to Heating

HEATIML


PZRPORV

Page 19


Core Damage

PZRSAFETY

TITLE

PAGE NUMBER DATE


Page 19 3/22/07


PZRPORVPage 20Page 18


PDSPZRPORV

PO-HPICOOL


PZPORVCONF-1


PZPORVCONF_99



PZPORVCONF-2


HPI Cooling

PZPORVCONF_0


BWHBW1-----HP2OA


PRVHPCONF

TITLE

PAGE NUMBER DATE


Page 20 3/22/07


ATIGNITIONPage 12

No Dispersal of Corium FromCavity

NODISPERSE

Cavity Geometry DoesRetain All Corium

GEOMH2


LOWPRESSPage 33Page 22Page 22


PRESSURE CORE MELT

PDSLOWPage 18


IS CERTAIN

PDSLOW-1

CORE MELT BIN 1

CM-001

CORE MELT BIN 2

CM-002

CORE MELT BIN 3

CM-003



PDSLOW-2


PDSLOW_5



PDSINDETPage 13

CORE MELT BIN 4

CM-004

CORE MELT BIN 6

CM-006

CORE MELT BIN 5

CM-005


PZRPORV

Page 19


OPSSSHR

Page 21


to Heating

HEATIML


Core Damage

PZRSAFETY

Random Spark isUnavailable at RV Failure

NOSPARKAT

TITLE

PAGE NUMBER DATE


Page 21 3/22/07


OPSSSHRPage 20


Are Available

PORVSSHR


PDSSGADV

AV


SSHRAVAIL


PDSSSHR

SSHR IS AVAILABLE

YES-SSHR

NO SSHR EXISTS

NO-SSHRPage 6

CAG0005


NO-SSHR-POSTLOOP



Prior to RV Failure

RECOVSSHR


Generators

OPSDEPRESS

TITLE

PAGE NUMBER DATE


Page 22 3/22/07


EXVPRODAFTHPage 12

Corium Pool Does NotSpread Over Large Area Or

Freeze

NOCOREFREEZE


Or Cavity Floor

NOSPREADLOW


LOWPRESS

Page 20


NOGEOMFREEZE



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


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


ARE UNAVAILABLE

NOPDSSPRAY

Page 16


Failure

NORECOVSPPRI


Failure

NORECOVSPAFT

Likelihood That Water Pool inCavity Will Not Stop

Concrete Attack

MELT


Vessel Failure

NORECOVRV

TITLE

PAGE NUMBER DATE


Page 23 3/22/07


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


Page 24 3/22/07



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


Page 25 3/22/07



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


ARE AVAILABLE

PDSSPRAY


RBSPRAY

Containment Sprays AreRecovered Prior to RV

Failure

RECOVSPPRI

Containment Sprays AreRecovered Early After RV

Failure

RECOVSPAFT

Accumulator Water isAvailable at RV Failure

ACCUMAVAIL


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


NOSEQPRESSH

Page 27

Likelihood ThatRecombination Cannot

Deplete Comb. Gas Given aDry Cavity

NODRYEFF

TITLE

PAGE NUMBER DATE


Page 26 3/22/07


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


Page 27 3/22/07



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


Page 28 3/22/07


RV FAILURE

SPARKAFTPage 12


RV FAILURE WITH RBSPRAY

SPARKAFT-1

PROB THAT SPARK ISAVAILABLE EARLY AFTER


SPARKAFT_99


RBSPRAY

RANDOM SPARK ISAVAILABLE EARLY AFTERRV FAILURE WITHOUT RB

SPRAY

SPARKAFT-2

PROB THAT SPARK ISAVAILABLE EARLY AFTERRV FAILURE WITHOUT RB

SPRAY

SPARKAFT_1


ARE UNAVAILABLE

NOPDSSPRAY

Page 16

TITLE

PAGE NUMBER DATE


Page 29 3/22/07


NOSTMINERTAFPage 12

Sequence After RV FailureHas Low Base Pressure In

Containment

AFTPRESSLPage 30



NOSEQPRESSH

Page 27

Reactor Building FansAvailable Early After RV

Failure

FANSAFT


Failure

FANSAVAILAFT

Reactor Building Fans DoFunction at RV Failure

NOFANSPRI


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


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



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


FANSUNAVAILPRI

Page 2

Likelihood RB Fans Do NotSurvive Containment


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


Containment

AFTPRESSL

Page 29


Press. W/ Low BasePressure

NOAFTSTREN2

TITLE

PAGE NUMBER DATE


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


CS01


Failure

RECOVSPPRI

Reactor Building FansCannot Handle Steam

Production

NORSGFANSG


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


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


RV Failure

ATPRESSL

Page 5

Likelihood That ContStrength Cannot Handle

RSG Press. Spike W/ LowBase Press.

NORSGSTREN2


Vessel Failure

NORECOVRV

TITLE

PAGE NUMBER DATE


Page 32 3/22/07

PDS DOES NOT HAVE HIGHBASE PRESSURE IN


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


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


LOWPRESS

Page 20

Sufficient Amount of CoriumCan Make Contact WithContainment Wall With

HPME

COREWALLHP


Failure

HIGHPRESS

Page 6

Plant Config and LayoutDoes Not Limit Material

Reaching Cont. Wall WithHPM

CWNOLIMITHPME


Vessel Failure

NORECOVRV

Containment Wall Does NotSurvive Contact With Corium

WALLNOSURVIV

TITLE

PAGE NUMBER DATE


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


Failure

HIGHPRESS

Page 6

Likelihood That Cont FailureIs Not Prevented Given a

Pressure Generated Missle

MISSLELIKE


Vessel Failure

NORECOVRV

TITLE

PAGE NUMBER DATE


Page 35 3/22/07


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


Page 36 3/22/07


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


Page 37 3/22/07


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


Page 38 3/22/07


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


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


Vessel Failure

NORECOVRV

Corium Pool Does Not SpreadOver Large Area Or Freeze

NOCOREFREEZE


Or Cavity Floor

NOSPREADLOW


LOWPRESS


PRESSURE CORE MELT

PDSLOW


IS CERTAIN

PDSLOW-1

CORE MELT BIN 1

CM-001

CORE MELT BIN 2

CM-002

CORE MELT BIN 3

CM-003



PDSLOW-2


PDSLOW_5



PDSINDET

CORE MELT BIN 4

CM-004

CORE MELT BIN 6

CM-006

CORE MELT BIN 5

CM-005


PZRPORV

Page 2


OPSSSHR

Page 3


to Heating

HEATIML


Core Damage

PZRSAFETY


NOGEOMFREEZE



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


Page 2 3/22/2007


PZRPORVPage 1


PDSPZRPORV

PO-HPICOOL


PZPORVCONF-1


PZPORVCONF_99



PZPORVCONF-2


HPI Cooling

PZPORVCONF_0


BWHBW1-----HP2OA


PRVHPCONF

TITLE

PAGE NUMBER DATE


Page 3 3/22/2007


OPSSSHRPage 1


Are Available

PORVSSHR


PDSSGADV

AV


SSHRAVAIL


PDSSSHR

SSHR IS AVAILABLE

YES-SSHR

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP



Prior to RV Failure

RECOVSSHR


Generators

OPSDEPRESS

TITLE

PAGE NUMBER DATE


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


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


Page 5 3/22/2007

Water Unavailable InReactor Cavity Prior to

Ex-Vessel Fission ProductRelease

NOEXFISWATERPage 4Page 1



NOFTRNSPOOLAFT


ARE UNAVAILABLE

NOPDSSPRAY


CS01


Failure

NORECOVSPPRI


Failure

NORECOVSPAFT

TITLE

PAGE NUMBER DATE


Page 6 3/22/2007


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


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


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


Page 2 3/22/2007


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


Page 3 3/22/2007


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


Page 4 3/22/2007


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


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


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


NOPDSSSHR

Page 7

Likelihood That Water in S/GWill Not Scrub Fission

Products

NOWATEREFF

TITLE

PAGE NUMBER DATE


Page 6 3/22/2007


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


Page 7 3/22/2007


NOPDSSSHRPage 5Page 8

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP


TITLE

PAGE NUMBER DATE


Page 8 3/22/2007

Fission Products Are NotScrubbed Prior to ReleaseFrom Intact Containment

FPNOSCRUBPRIPage 1

SGTR-Containment-BypassSequences

SGTRCB

PDS INDICATES SGTREXISTS

NOPDSNOSGTR


CM-15-18

Page 3


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


SSHRUNAVAIL


NOPDSSSHR

Page 7



NORECOVSSHR


HIGHTUBEHT


S/G Tubes is High

HIGHNATHT


High

UNEFFNATHT


RCPUMPOFF

Page 9


S/G Tubes is High

HIGHFORCEHT


High

NOEFFFORCEHT


RCPUMPON

Page 10


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


Page 9 3/22/2007


RCPUMPOFFPage 8

Page 13


NORECOVPOWER


NOPDSRCPWR


RCPS

DPG0003


NORECACPRI


Pumps

NONCONBYOPS

TITLE

PAGE NUMBER DATE


Page 10 3/22/2007


RCPUMPONPage 13Page 8


RECOVPOWER


PDSRCPWR


RCPS

DPG0003


RECACPRI


Pumps

NCONBYOPS

TITLE

PAGE NUMBER DATE


Page 11 3/22/2007


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


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


OPEN PORV

NOEFFDEPRESS-2

Page 12

Likelihood That PressurizerPORV(s) Cannot Depress


PZRNOPORVDEP


NOPRIMFAILURE

Page 13

TITLE

PAGE NUMBER DATE


Page 12 3/22/2007


OPEN PORV

NOEFFDEPRESS-2Page 11


BWHBW1-----HP2OA

OPERATOR FAILS TOOPEN PORV AFTER

FAILING TO INITIATE HPICOOLING

EFFDEPRESS_0-C

TITLE

PAGE NUMBER DATE


Page 13 3/22/2007


NOPRIMFAILUREPage 11

Primary System Failure DoesNot Precede S/G Tube

Failure With RCPs Running

NOPRIMFAILPMP


S/G Tube Failure W/ RCPsOn

NOEFFPMP


RCPUMPON

Page 10

Primary System Failure DoesNot Precede S/G TubeFailure With RCPs Off

NOPRIMFAILNPMP


S/G Tube Failure W/ RCPsOff

NOEFFNPMP


RCPUMPOFF

Page 9

TITLE

PAGE NUMBER DATE


Page 14 3/22/2007


NOTBPage 8

SMALL CONTAINMENTISOLATION FAILURE

SMALL-ISO


LARGE-ISO

TITLE

PAGE NUMBER DATE


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


ARE UNAVAILABLE

NOPDSSPRAYPage 1


CS01

Containment Sprays Are NotRecovered Prior to Late

Containment Failure

NOSPRECOVLT

Containment Sprays Are NotRecovered Prior to RV Failure

NORECOVSPPRI


Failure

NORECOVSPAFT



NORECOVSPLT

RB SPRAY UNAVAILABLEDUE TO MECH FAILURE OR

NO OFFSITE POWER

NORECOVSPLT-1



CS01-R


HOURS

NORECOFFSITEPWR

IE-LOOP-101

Likelihood That Spray Will NotScrub FPs Prior to Release to

Environment

NOSPRAYEFFLT

TITLE

PAGE NUMBER DATE


Page 16 3/22/2007


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


Page 17 3/22/2007


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


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


CM-1-14

Page 2


CM-15-18Page 3

CORE MELT BIN 15

CM-015

CORE MELT BIN 16

CM-016

CORE MELT BIN 18

CM-018


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


Page 2 3/22/2007


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


Page 3 3/22/2007


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


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


Page 4 3/22/2007

S/G Scrubbing Is Unavail. orFP's are Not Released to

S/G

SGUNAVAILPage 1


STEAM GENERATOR

NOSGREL

Page 5

S/G Scrubbing is Unavailable

NOSGSCRUB


NOPDSSSHR

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP


Likelihood That Water in S/GWill Not Scrub Fission

Products

NOWATEREFF

TITLE

PAGE NUMBER DATE


Page 5 3/22/2007


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


Page 6 3/22/2007


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-ISO

FP Scrubbing is Effective orAvail. For Small Contain.

Failures

SMALLEFF


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


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


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


Page 3 3/22/2007


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


Page 4 3/22/2007


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


Page 5 3/22/2007


SGAVAILPage 1

FISSION PRODUCTS ARERELEASED TO THE STEAM

GENERATOR

SGREL


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


PDSSSHRPage 8

SSHR IS AVAILABLE

YES-SSHR

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP


Likelihood That Water in S/GWill Scrub Fission Products

WATEREFF

TITLE

PAGE NUMBER DATE


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


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


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


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


Page 8 3/22/2007


SSHRAVAILPage 6


PDSSSHR

Page 5


Prior to RV Failure

RECOVSSHR

TITLE

PAGE NUMBER DATE


Page 9 3/22/2007


LOWTUBEHTPage 6


S/G Tubes is Low

LOWNATHT


Low

EFFNATHT


RCPUMPOFF

Page 10


S/G Tubes is Low

LOWFORCEHT


Low

EFFFORCEHT


RCPUMPON

Page 10

TITLE

PAGE NUMBER DATE


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


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


BWHBW1-----HP2OA

OPERATOR OPENS PORVAFTER FAILING TO

INITIATE HPI COOLING

EFFDEPRESS_0

Likelihood That PressurizerPORV(s) Can Depress


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


RCPUMPONPage 9


RECOVPOWER


PDSRCPWR


RCPS

DPG0003


RECACPRI


Pumps

NCONBYOPS


With RCPs Off

PRIMFAILNPMP


RCPUMPOFFPage 9


NORECOVPOWER


NOPDSRCPWR


RCPS

DPG0003


NORECACPRI


Pumps

NONCONBYOPS

Conf That Primary Sys FailurePrecedes S/G Tube Failure

W/ RCPs Off

EFFNPMP

TITLE

PAGE NUMBER DATE


Page 11 3/22/2007


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


Page 12 3/22/2007

CONTAINMENT ISISOLATED (SUCCESS OF

B)

BPage 6


NOTB


SMALL-ISO


LARGE-ISO

TITLE

PAGE NUMBER DATE


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


ARE AVAILABLE

PDSSPRAYPage 1


RBSPRAY

Containment Sprays AreRecovered Prior to Late

Containment Failure

SPRECOVLT

Containment Sprays AreRecovered Prior to RV Failure

RECOVSPPRI


Failure

RECOVSPAFT

RB SPRAY POWERSUPPLIES ARE


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


Page 14 3/22/2007


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


Page 15 3/22/2007


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


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


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


SCENARIOS

CMB-NOSGTR

Page 4

Likelihood That Plateout WillScrub Fission Products

PLATEOUT

Likelihood That ScrubbingCapability of Fission Products

Exists

AUXSPRAYS


SGAVAIL

FISSION PRODUCTS ARERELEASED TO THE STEAM

GENERATOR

SGREL


CM-15-18

CORE MELT BIN 15

CM-015

CORE MELT BIN 16

CM-016

CORE MELT BIN 18

CM-018


SGSCRUB

Page 5

Likelihood That There is FPScrubbing By Other Systems

Not in Aux Bldg.

OTHERSCRUB

TITLE

PAGE NUMBER DATE


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


Page 3 3/22/2007


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


Page 4 3/22/2007


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


Page 5 3/22/2007


SGSCRUBPage 1


PDSSSHR

SSHR IS AVAILABLE

YES-SSHR

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP


Likelihood That Water in S/GWill Scrub Fission Products

WATEREFF

TITLE

PAGE NUMBER DATE


Page 6 3/22/2007


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


H2 Concentration

NOPRIBURNPage 15


Before RV Failure

NOPRICONC



LOWCONCBURN


Before RV Failure

NOPRIRELEASE

Page 2


FAILURE

NOSPARK


UNAVAILABLE

NOSPARK-1

Page 5


AVAILABLE

NOSPARK-2

Page 7


STMINERTP

Page 1


NOATBURN

H2 Concentration isInsufficient to Cause Burns at

RV Failure

ATCONC


RV Failure

ATRELEASE

Page 8



LOWCONCBURN


ATIGNITION

Page 10


STMINERTPPage 1


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


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


Page 2 3/22/2007


Before RV Failure

NOPRIRELEASEPage 1


OXIDIZED


BOTTLEDPage 15


PRESSURE CORE MELT

NOPDSLOWPage 18


IS CERTAIN

NOPDSLOW-2

Page 3



NOPDSLOW-1


NOPDSLOW_5



PDSINDET

Page 10



NOHEATIML


NOPZRPORV

Page 4


During Core Damage

NOPZRSAFETY

TITLE

PAGE NUMBER DATE


Page 3 3/22/2007


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


Page 4 3/22/2007


NOPZRPORVPage 2

Page 18


NOPDSPZRPORV

PO-HPICOOL


HPME

NOPRVHPCONF


NOPZPORVCONF-1


MANUALLY OPEN PORV

PZPORVCONF_99-C


NOT INITIATED

NOPZPORVCONF-2



PZPORVCONF_0-C


BWHBW1-----HP2OA

TITLE

PAGE NUMBER DATE


Page 5 3/22/2007


UNAVAILABLE

NOSPARK-1Page 1


SPRAY

NOSPARK_9


ARE UNAVAILABLE

NOPDSSPRAY

Page 6

TITLE

PAGE NUMBER DATE


Page 6 3/22/2007


ARE UNAVAILABLE

NOPDSSPRAYPage 5

Page 37Page 25

... see x-ref


CS01

TITLE

PAGE NUMBER DATE


Page 7 3/22/2007


AVAILABLE

NOSPARK-2Page 1



NOSPARK_01


RBSPRAY

TITLE

PAGE NUMBER DATE


Page 8 3/22/2007


RV Failure

ATRELEASEPage 1


PRIBURNPage 14


Before RV Failure

PRICONC


Comb. Gas

NOLOWCONCBURN


Before RV Failure

PRIRELEASE


NOOXIDIZED


NOTBOTTLED

Page 9


FAILURE

SPARK


UNAVAILABLE

SPARK-1


SPRAY

SPARK_1


ARE UNAVAILABLE

NOPDSSPRAY

Page 6


AVAILABLE

SPARK-2



SPARK_99


RBSPRAY


NOSTMINERTP

Page 15


OXIDIZED

Recovery of Core CoolingDoes Prevent Reactor Vessel

Failure

RECOVRV


NOTBOTTLED

Page 9

TITLE

PAGE NUMBER DATE


Page 9 3/22/2007


NOTBOTTLEDPage 8Page 8


PRESSURE CORE MELT

PDSLOW

Page 10


to Heating

HEATIML


PZRPORV

Page 10


Core Damage

PZRSAFETY

TITLE

PAGE NUMBER DATE


Page 10 3/22/2007


ATIGNITIONPage 1


NODISPERSE


GEOMH2


LOWPRESSPage 22


PRESSURE CORE MELT

PDSLOWPage 9


IS CERTAIN

PDSLOW-1

CORE MELT BIN 1

CM-001

CORE MELT BIN 2

CM-002

CORE MELT BIN 3

CM-003



PDSLOW-2


PDSLOW_5



PDSINDETPage 2

CORE MELT BIN 4

CM-004

CORE MELT BIN 6

CM-006

CORE MELT BIN 5

CM-005


PZRPORVPage 9


PDSPZRPORV

PO-HPICOOL


PZPORVCONF-1


PZPORVCONF_99



PZPORVCONF-2

Page 11


PRVHPCONF


OPSSSHR

Page 12


to Heating

HEATIML


Core Damage

PZRSAFETY


NOSPARKAT

TITLE

PAGE NUMBER DATE


Page 11 3/22/2007



PZPORVCONF-2Page 10


HPI Cooling

PZPORVCONF_0


BWHBW1-----HP2OA

TITLE

PAGE NUMBER DATE


Page 12 3/22/2007


OPSSSHRPage 10


Are Available

PORVSSHR


PDSSGADV

AV


SSHRAVAIL


PDSSSHR

SSHR IS AVAILABLE

YES-SSHR

NO SSHR EXISTS

NO-SSHRPage 19

CAG0005


NO-SSHR-POSTLOOP



Prior to RV Failure

RECOVSSHR


Generators

OPSDEPRESS

TITLE

PAGE NUMBER DATE


Page 13 3/22/2007


RV Failure

ATPRESSHPage 1



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


FANSUNAVAILPRI

Page 27

TITLE

PAGE NUMBER DATE


Page 14 3/22/2007


Prevented

NOAFTBURNPage 1

Comb. Gas Concentration isInsufficient to Cause H2Burns Early After Failure

AFTERCONC


Failure

AFTERREL

Previous Burns Do DepleteHydrogen in Containment

PRIGLOBAL


PRIBURN

Page 8


ATBURN

Page 15

Ex-Vessel Gas ProductionAfter RV Failure is Low

EXVPRODAFTL

Page 18

Cavity Recombination DoesDeplete Combustible Gas


AFTERRECOM

Page 22



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


Page 15 3/22/2007


ATBURNPage 14


RV Failure

NOATCONC


RV Failure

NOATRELEASE


H2 Concentration

NOPRIBURN

Page 1


NOOXIDIZED


Vessel Failure

NORECOVRV


BOTTLED

Page 2


Comb. Gas

NOLOWCONCBURN


NOATIGNITION


DISPERSE


NOGEOMH2


Failure

HIGHPRESS

Page 18


SPARKAT


NOSTMINERTPPage 8


RV Failure

ATPRESSL

Page 16

TITLE

PAGE NUMBER DATE


Page 16 3/22/2007


RV Failure

ATPRESSLPage 15



NOPDSPRESSH

Page 17


FANSAT


FANSAVAILPRIPage 33



PDSFANS

CF


Failure

RECOVFANSPRI

TITLE

PAGE NUMBER DATE


Page 17 3/22/2007



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


Page 18 3/22/2007


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


FREEZELOW

Concrete Attack ProducesInsufficient Combustible Gas

After RV Failure

H2SRCAFTER

Water Pool Does StopConcrete Attack In Cavity

After RV Failure

NOATTKAFT


Failure

WATERAFTERPage 28


Paths

OTHERWATER


After RV Failure

FTRNSPOOLAFT


ARE AVAILABLE

PDSSPRAY

Page 35


Failure

RECOVSPPRI


Failure

RECOVSPAFT


ACCUMAVAIL


Failure

HIGHPRESSPage 15Page 35


PRESSURE CORE MELT

NOPDSLOW

Page 2


Generators

OPSNOSSHR

Page 19


NOPZRPORV

Page 4



NOHEATIML


During Core Damage

NOPZRSAFETY



GRAVFEEDAFT

Page 20

Likelihood That Water Poolin Cavity Will Stop Concrete

Attack

NOMELT


Vessel Failure

RECOVRV

TITLE

PAGE NUMBER DATE


Page 19 3/22/2007


Generators

OPSNOSSHRPage 18


Unavailable

NOPORVSSHR


NOPDSSGADV

AV


SSHRUNAVAIL


NOPDSSSHR

NO SSHR EXISTS

NO-SSHR

Page 12



NORECOVSSHR


Generators

NOOPSDEPRESS

TITLE

PAGE NUMBER DATE


Page 20 3/22/2007



GRAVFEEDAFTPage 18


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



NOSEQPRESSH

Page 21

TITLE

PAGE NUMBER DATE


Page 21 3/22/2007



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


Page 22 3/22/2007



AFTERRECOMPage 14


Failure

NOWATERAFTER


And Paths

NOOTHERWATER


ACCUMUNAVAIL


LOWPRESS

Page 10


Vessel

NOGRAVFEEDAFT


NOPDSINJECCS

Page 23



SEQPRESSH

Page 26



NOFTRNSPOOLAFT

Page 24

Likelihood ThatRecombination Can Deplete

Comb. Gas Given a DryCavity

DRYEFF

TITLE

PAGE NUMBER DATE


Page 23 3/22/2007


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


Page 24 3/22/2007



NOFTRNSPOOLAFTPage 22


ARE UNAVAILABLE

NOPDSSPRAY

Page 6


Failure

NORECOVSPPRI


Failure

NORECOVSPAFT

TITLE

PAGE NUMBER DATE


Page 25 3/22/2007


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


RBSPRAY

NO RANDOM SPARKEARLY AFTER RV FAILURE

FOR RB SPRAYUNAVAILABLE

NOSPARKAFT-2

PROB THAT SPARK ISUNAVAILABLE EARLYAFTER RV FAILURE

WITHOUT RB SPRAY

NOSPARKAFT_9


ARE UNAVAILABLE

NOPDSSPRAY

Page 6

TITLE

PAGE NUMBER DATE


Page 26 3/22/2007


STMINERTAFPage 14


Containment

AFTPRESSH



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


Failure

NOFANSAFT

Page 27

TITLE

PAGE NUMBER DATE


Page 27 3/22/2007


Failure

NOFANSAFTPage 26Page 35


Failure

FANSUNAVAILAFT


FANSPRI


FANSUNAVAILPRIPage 13



NOPDSFANS

CF


RV Failure

NORECOVFANSPRI



NOEQUALFANSPRI


RV Failure

NORECOVFANSAFT



NOEQUALFANSAF

TITLE

PAGE NUMBER DATE


Page 28 3/22/2007



NOAFTERRECOMPage 1


Failure

WATERAFTER

Page 18



NODRYEFF

TITLE

PAGE NUMBER DATE


Page 29 3/22/2007

Cavity Recombination DoesNot Deplete Comb. Gas Prior

to Late Containment Failur

NOLTRECOMBPage 1


Deplete Comb. Gas With aDry Cavity Late

NODRYEFFLT

Water Is Available in CavityArea

STMWATER

Page 35

TITLE

PAGE NUMBER DATE


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


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


HOURS

RECOFFSITEPWR

SPARK AVAILABLE;OFFSITE POWER NOT

RECOVERED

SPARKLT-2

RANDOM SPARKAVAILABLE WITHOUT

OFFSITE POWER

SPARKLT-NOP


HOURS

NORECOFFSITEPWR

TITLE

PAGE NUMBER DATE


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



NOSTREN1H2

Containment StrengthCannot Handle Late Comb.

Gas Burn and Base Pressureis Low

LTSTRENL


LTPRESSLPage 30

Reactor Building FansFunction Prior to LateContainment Failure

FANSLT

Reactor Building Fans AreAvailable Prior to LateContainment Failure

FANSAVAILLT


Failure

FANSAFT


Failure

FANSAVAILAFT


NOFANSPRI

Page 33


Failure

RECOVFANSAFT



EQUALFANSAF

RB FAN POWER SUPPLIESARE RECOVERED PRIORTO LATE CTMT FAILURE

RECOVFANSLT


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



NOSTREN2H2

Sequence Late After RVFailure Has Low Base


INERTLT

TITLE

PAGE NUMBER DATE


Page 33 3/22/2007


NOFANSPRIPage 32


FANSAVAILPRI

Page 16



EQUALFANSPRI

TITLE

PAGE NUMBER DATE


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


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


Failure

HIGHPRESS

Page 18


For Freezing

GEOMFREEZE


STMWATERPage 29

Water Available fromContainment Sprays Via Fuel


FTRNSPOOLLT


ARE AVAILABLE

PDSSPRAYPage 18


RBSPRAY


RECOVSPPRI


Failure

RECOVSPAFT



RECOVSPLT


HOURS

RECOFFSITEPWR



RECSPRAYLT

IE-LOOP-101


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


Page 36 3/22/2007


FAILURE

NORECOVFANSLTPage 35

RB FANS UNAVAILABLEFOLLOWING POST-LOOP

RECOVERY

CF-R


HOURS

NORECOFFSITEPWR

TITLE

PAGE NUMBER DATE


Page 37 3/22/2007

Containment Failure FromNon Condensable Gases

GASESPage 1

Water is Unavailable inCavity Prior to LCF

NOSTMWATER



NOFTRNSPOOLLT


ARE UNAVAILABLE

NOPDSSPRAY

Page 6


Failure

NORECOVSPPRI



NORECOVSPLT


NORECOVSPLT-1



CS01-R


HOURS

NORECOFFSITEPWR

IE-LOOP-101


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


Page 38 3/22/2007


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


Page 39 3/22/2007


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


Page 40 3/22/2007


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


Page 41 3/22/2007


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


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


Core Damage

PZRSAFETY


PRESSURE CORE MELT

PDSLOW


IS CERTAIN

PDSLOW-1

CORE MELT BIN 1

CM-001

CORE MELT BIN 2

CM-002

CORE MELT BIN 3

CM-003



PDSLOW-2


PDSLOW_5



PDSINDET

Page 2


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


Failure

RECOVRV

TITLE

PAGE NUMBER DATE


Page 2 3/22/2007



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


Page 3 3/22/2007


PZRPORVPage 1


PDSPZRPORV

PO-HPICOOL


PZPORVCONF-1


PZPORVCONF_99



PZPORVCONF-2


HPI Cooling

PZPORVCONF_0


BWHBW1-----HP2OA


PRVHPCONF

TITLE

PAGE NUMBER DATE


Page 4 3/22/2007


HIGHPRESPCMPage 1


During Core Damage

NOPZRSAFETY


PRESSURE CORE MELT

NOPDSLOW


IS CERTAIN

NOPDSLOW-2

Page 5



NOPDSLOW-1


NOPDSLOW_5



PDSINDET

Page 2


NOPZRPORV


NOPDSPZRPORV

PO-HPICOOL


HPME

NOPRVHPCONF


NOPZPORVCONF-1


MANUALLY OPEN PORV

PZPORVCONF_99-C


NOT INITIATED

NOPZPORVCONF-2



PZPORVCONF_0-C


BWHBW1-----HP2OA

TITLE

PAGE NUMBER DATE


Page 5 3/22/2007


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


Page 6 3/22/2007

Secondary Side HeatRemoval Prevents

Revaporization

SSHRPage 1

Secondary Side HeatRemoval is Available Prior to

Revaporization

SSHRRVPAVL


PDSSSHR

SSHR IS AVAILABLE

YES-SSHR

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP


Secondary Side HeatRemoval is Recovered Prior

to Revaporization

SSHRRVPREC

Likelihood That SSHR WillPrevent Revaporization

SSHRREVAP

TITLE

PAGE NUMBER DATE


Page 7 3/22/2007


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


COREFREEZE


Floor

SPREADLOW


Failure

HIGHPRESS


PRESSURE CORE MELT

NOPDSLOW


IS CERTAIN

NOPDSLOW-2

Page 2



NOPDSLOW-1


NOPDSLOW_5



PDSINDET

CORE MELT BIN 4

CM-004

CORE MELT BIN 6

CM-006

CORE MELT BIN 5

CM-005


Generators

OPSNOSSHR


Unavailable

NOPORVSSHR


NOPDSSGADV

AV


SSHRUNAVAIL


NOPDSSSHR

Page 3


Recovered Prior to RV Failure

NORECOVSSHR


Generators

NOOPSDEPRESS

Operators Do NotDepressurize with Pressurizer

PORV

NOPZRPORV

Page 4



NOHEATIML


During Core Damage

NOPZRSAFETY


For Freezing

GEOMFREEZE



FREEZELOW

Water Pool Stops ConcreteAttack Prior to Basemat

Melt-Through

NOATTACK

Page 5


Failure

RECOVRV

TITLE

PAGE NUMBER DATE


Page 2 3/22/2007


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


Page 3 3/22/2007


NOPDSSSHRPage 1

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP


TITLE

PAGE NUMBER DATE


Page 4 3/22/2007


NOPZRPORVPage 1


NOPDSPZRPORV

PO-HPICOOL


HPME

NOPRVHPCONF


NOPZPORVCONF-1


MANUALLY OPEN PORV

PZPORVCONF_99-C


NOT INITIATED

NOPZPORVCONF-2



PZPORVCONF_0-C


BWHBW1-----HP2OA

TITLE

PAGE NUMBER DATE


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



FTRNSPOOLLT


ARE AVAILABLE

PDSSPRAY


RBSPRAY


Failure

RECOVSPPRI


Failure

RECOVSPAFT



RECOVSPLT


HOURS

RECOFFSITEPWR



RECSPRAYLT

IE-LOOP-101

Likelihood That Water Poolin Cavity Will Stop Concrete

Attack

NOMELT

TITLE

PAGE NUMBER DATE


Page 6 3/22/2007


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



NOISGTRCB

Page 5

TITLE

PAGE NUMBER DATE


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


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


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


Page 5 3/22/2007



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


SSHRAVAIL


PDSSSHR

SSHR IS AVAILABLE

YES-SSHR

NO SSHR EXISTS

NO-SSHR

Page 6


Prior to RV Failure

RECOVSSHR


LOWTUBEHT


S/G Tubes is Low

LOWNATHT


Low

EFFNATHT


RCPUMPOFF

Page 7


S/G Tubes is Low

LOWFORCEHT


Low

EFFFORCEHT


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


Page 6 3/22/2007

NO SSHR EXISTS

NO-SSHRPage 5

CAG0005


NO-SSHR-POSTLOOP


TITLE

PAGE NUMBER DATE


Page 7 3/22/2007

Primary to Secondary Delta Pis Too Low to Induce Creep

Rupture Failure

LOWDELPPage 5


PRESSURE

PDSRCEQSG

Page 8

Operators Depress. withPressurizer PORVs Prior to

S/G Tube Failure

OPSDEPRES2


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


BWHBW1-----HP2OA

OPERATOR OPENS PORVAFTER FAILING TO

INITIATE HPI COOLING

EFFDEPRESS_0

Likelihood That PressurizerPORV(s) Can Depress


PZRPORVDEP

Primary System FailureReduces RC Pressure Prior

to S/G Tube Failure

PRIMFAILURE


With RCPs Running

PRIMFAILPMP

Conf That Primary SysFailure Precedes S/G Tube

Failure W/ RCPs On

EFFPMP


RCPUMPONPage 5


RECOVPOWER


PDSRCPWR

Page 9


RECACPRI


Pumps

NCONBYOPS


With RCPs Off

PRIMFAILNPMP


RCPUMPOFFPage 5


NORECOVPOWER


NOPDSRCPWR

Page 10


NORECACPRI


Pumps

NONCONBYOPS

Conf That Primary SysFailure Precedes S/G Tube

Failure W/ RCPs Off

EFFNPMP

TITLE

PAGE NUMBER DATE


Page 8 3/22/2007


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


Page 9 3/22/2007


PDSRCPWRPage 7


RCPS

DPG0003

TITLE

PAGE NUMBER DATE


Page 10 3/22/2007


NOPDSRCPWRPage 7


RCPS

DPG0003

TITLE

PAGE NUMBER DATE


Page 11 3/22/2007


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


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


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


RV Failure

ATPRESSH

Page 32

Likelihood That Cont.Strength Can Handle DCHPress Spike W/ High Base

Press

DCHSTREN1


Pressure is Low

DCHSTRENL

Page 3


NOGEOMFREEZE


Vessel Failure

RECOVRV


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


Page 2 3/22/2007


FANSAVAILPRIPage 32Page 1Page 4

... see x-ref



PDSFANS

CF


Failure

RECOVFANSPRI

TITLE

PAGE NUMBER DATE


Page 3 3/22/2007


Pressure is Low

DCHSTRENLPage 1


RV Failure

ATPRESSL

Page 4

Likelihood That Cont.Strength Can Handle DCHPress Spike W/ Low Base

Press

DCHSTREN2

TITLE

PAGE NUMBER DATE


Page 4 3/22/2007


RV Failure

ATPRESSLPage 16Page 32Page 11

... see x-ref



NOPDSPRESSH

Page 17


FANSAT


FANSAVAILPRI

Page 2

TITLE

PAGE NUMBER DATE


Page 5 3/22/2007


LOWPRESSPage 33Page 36Page 1

... see x-ref


PRESSURE CORE MELT

PDSLOWPage 13


IS CERTAIN

PDSLOW-1

CORE MELT BIN 1

CM-001

CORE MELT BIN 2

CM-002

CORE MELT BIN 3

CM-003



PDSLOW-2


PDSLOW_5



PDSINDETPage 6

CORE MELT BIN 4

CM-004

CORE MELT BIN 6

CM-006

CORE MELT BIN 5

CM-005


PZRPORV

Page 14


OPSSSHR


Are Available

PORVSSHR


PDSSGADV

AV


SSHRAVAIL

Page 34


Generators

OPSDEPRESS


to Heating

HEATIML


Core Damage

PZRSAFETY

TITLE

PAGE NUMBER DATE


Page 6 3/22/2007

Containment Failure FromCombustible Gas Burns is

Prevented

NOH2BURNSPage 1

Containment Failure From H2Burns Before RV Failure is

Prevented

H2PRI


H2 Concentration

NOPRIBURNPage 19


Before RV Failure

NOPRICONC



LOWCONCBURN


Before RV Failure

NOPRIRELEASE


OXIDIZED


BOTTLEDPage 19


PRESSURE CORE MELT

NOPDSLOWPage 21


IS CERTAIN

NOPDSLOW-2

Page 7



NOPDSLOW-1


NOPDSLOW_5



PDSINDET

Page 5



NOHEATIML


PORV

NOPZRPORV

Page 8


During Core Damage

NOPZRSAFETY


FAILURE

NOSPARK


UNAVAILABLE

NOSPARK-1


SPRAY

NOSPARK_9


ARE UNAVAILABLE

NOPDSSPRAY

Page 9


AVAILABLE

NOSPARK-2



NOSPARK_01


RBSPRAY


STMINERTP

Page 10

Containment Strength CanHandle H2 Burn Event Prior

to RV Failure

PRISTRENT

Page 11


Failure is Prevented

H2AT

Page 12

Containment Failure From H2Burns after RV Failure is

Prevented

H2AFTER

Page 18

TITLE

PAGE NUMBER DATE


Page 7 3/22/2007


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


Page 8 3/22/2007


NOPZRPORVPage 6

Page 21


NOPDSPZRPORV

PO-HPICOOL


HPME

NOPRVHPCONF


NOPZPORVCONF-1


MANUALLY OPEN PORV

PZPORVCONF_99-C


NOT INITIATED

NOPZPORVCONF-2



PZPORVCONF_0-C


BWHBW1-----HP2OA

TITLE

PAGE NUMBER DATE


Page 9 3/22/2007


ARE UNAVAILABLE

NOPDSSPRAYPage 6

Page 28Page 27

... see x-ref


CS01

TITLE

PAGE NUMBER DATE


Page 10 3/22/2007


STMINERTPPage 12Page 6


RV Failure

ATPRESSH

Page 32

TITLE

PAGE NUMBER DATE


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


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


RV Failure

ATPRESSL

Page 4

Likelihood That Cont. CanHandle H2 Burn Press W/

Low Base Press.

PRISTREN2

TITLE

PAGE NUMBER DATE


Page 12 3/22/2007


Failure is Prevented

H2ATPage 6


NOATBURN

H2 Concentration isInsufficient to Cause Burns

at RV Failure

ATCONC


RV Failure

ATRELEASE


PRIBURNPage 18


Before RV Failure

PRICONC


Comb. Gas

NOLOWCONCBURN


Before RV Failure

PRIRELEASE


NOOXIDIZED


NOTBOTTLED

Page 13


FAILURE

SPARK


UNAVAILABLE

SPARK-1


SPRAY

SPARK_1


ARE UNAVAILABLE

NOPDSSPRAY

Page 9


AVAILABLE

SPARK-2



SPARK_99


RBSPRAY


NOSTMINERTP

Page 18


OXIDIZED


Vessel Failure

RECOVRV


NOTBOTTLED

Page 13



LOWCONCBURN


ATIGNITION

Page 15


STMINERTP

Page 10

Containment Strength CanHandle H2 Burns Event at

RV Failure

ATSTRENT

Page 16

TITLE

PAGE NUMBER DATE


Page 13 3/22/2007


NOTBOTTLEDPage 12Page 12


PRESSURE CORE MELT

PDSLOW

Page 5


to Heating

HEATIML


PZRPORV

Page 14


Core Damage

PZRSAFETY

TITLE

PAGE NUMBER DATE


Page 14 3/22/2007


PZRPORVPage 5

Page 13


PDSPZRPORV

PO-HPICOOL


PZPORVCONF-1


PZPORVCONF_99



PZPORVCONF-2


HPI Cooling

PZPORVCONF_0


BWHBW1-----HP2OA


PRVHPCONF

TITLE

PAGE NUMBER DATE


Page 15 3/22/2007


ATIGNITIONPage 12


NODISPERSE


GEOMH2


LOWPRESS

Page 5


NOSPARKAT

TITLE

PAGE NUMBER DATE


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


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


RV Failure

ATPRESSL

Page 4

Likelihood That Cont CanHandle H2 Burn Press. W/

Low Base Press.

ATSTREN2

TITLE

PAGE NUMBER DATE


Page 17 3/22/2007



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


Page 18 3/22/2007

Containment Failure From H2Burns after RV Failure is

Prevented

H2AFTERPage 6


Prevented

NOAFTBURN

Comb. Gas Concentration isInsufficient to Cause H2Burns Early After Failure

AFTERCONC


Failure

AFTERREL

Previous Burns Do DepleteHydrogen in Containment

PRIGLOBAL


PRIBURN

Page 12


ATBURN

H2 Concentration is Sufficientto Cause Burns at RV Failure

NOATCONC


RV Failure

NOATRELEASE

Page 19


Comb. Gas

NOLOWCONCBURN

Ignition Source is Available atRV Failure

NOATIGNITION


DISPERSE

Page 20


SPARKAT


NOSTMINERTPPage 12


RV Failure

ATPRESSL

Page 4


EXVPRODAFTL


COREFREEZE


Floor

SPREADLOW

Primary System Pressure isHigh at Reactor Vessel Failure

HIGHPRESS

Page 21


For Freezing

GEOMFREEZE



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


Failure

RECOVRV



AFTERRECOM

Page 25



LOWCONCBURN


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


Page 19 3/22/2007


RV Failure

NOATRELEASEPage 18


H2 Concentration

NOPRIBURN

Page 6


NOOXIDIZED


Vessel Failure

NORECOVRV


BOTTLED

Page 6

TITLE

PAGE NUMBER DATE


Page 20 3/22/2007


DISPERSEPage 18


NOGEOMH2


Failure

HIGHPRESS

Page 21

TITLE

PAGE NUMBER DATE


Page 21 3/22/2007


Failure

HIGHPRESSPage 33Page 22Page 20

... see x-ref


PRESSURE CORE MELT

NOPDSLOW

Page 6


Generators

OPSNOSSHR


Unavailable

NOPORVSSHR


NOPDSSGADV

AV


SSHRUNAVAIL


NOPDSSSHR

NO SSHR EXISTS

NO-SSHR

Page 35



NORECOVSSHR


Generators

NOOPSDEPRESS


NOPZRPORV

Page 8



NOHEATIML


During Core Damage

NOPZRSAFETY

TITLE

PAGE NUMBER DATE


Page 22 3/22/2007


Failure

WATERAFTERPage 18


Paths

OTHERWATER


After RV Failure

FTRNSPOOLAFT


ARE AVAILABLE

PDSSPRAY


RBSPRAY


Failure

RECOVSPPRI


Failure

RECOVSPAFT


ACCUMAVAIL


Failure

HIGHPRESS

Page 21



GRAVFEEDAFT


PDSINJECCS

Page 23



NOSEQPRESSH

Page 24

TITLE

PAGE NUMBER DATE


Page 23 3/22/2007


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


Page 24 3/22/2007




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


Page 25 3/22/2007



AFTERRECOMPage 18


Failure

NOWATERAFTER


And Paths

NOOTHERWATER


ACCUMUNAVAIL


LOWPRESS

Page 5


Vessel

NOGRAVFEEDAFT


NOPDSINJECCS

Page 26



SEQPRESSH

Page 30



NOFTRNSPOOLAFT

Page 27



DRYEFF

TITLE

PAGE NUMBER DATE


Page 26 3/22/2007


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


Page 27 3/22/2007





ARE UNAVAILABLE

NOPDSSPRAY

Page 9


Failure

NORECOVSPPRI


Failure

NORECOVSPAFT

TITLE

PAGE NUMBER DATE


Page 28 3/22/2007


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


RBSPRAY

NO RANDOM SPARKEARLY AFTER RV FAILURE

FOR RB SPRAYUNAVAILABLE

NOSPARKAFT-2

PROB THAT SPARK ISUNAVAILABLE EARLYAFTER RV FAILURE

WITHOUT RB SPRAY

NOSPARKAFT_9


ARE UNAVAILABLE

NOPDSSPRAY

Page 9

TITLE

PAGE NUMBER DATE


Page 29 3/22/2007


STMINERTAFPage 18


Containment

AFTPRESSH

Page 30

TITLE

PAGE NUMBER DATE


Page 30 3/22/2007

Containment Strength CanHandle Comb. Gas BurnEvent After RV Failure

AFTSTRENTPage 18

Containment Strength CanHandle Comb. Gas Burn and


AFTSTRENH

Containment Base Pressureis High

BASEPRESSH

Containment Has High BasePressure Early After RVFailure Without Steam

Inerting

NOINERTAF


Containment

AFTPRESSHPage 29



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


Failure

NOFANSAFT


Failure

FANSUNAVAILAFT


FANSPRI


FANSUNAVAILPRI

Page 31



NOEQUALFANSPRI


RV Failure

NORECOVFANSAFT



NOEQUALFANSAF

Likelihood That Cont. CanHandle Comb. Gas Burn


AFTSTREN1

Containment Strength CanHandle Comb. Gas Burn and


AFTSTRENL

Containment Base Pressureis Low

BASEPRESSL

Containment Has Low BasePressure Early After RVFailure Without Steam

Inerting

INERTAF


Containment

AFTPRESSL



NOSEQPRESSH

Page 24


Failure

FANSAFT


Failure

FANSAVAILAFT


NOFANSPRI


FANSAVAILPRI

Page 2



EQUALFANSPRI


Failure

RECOVFANSAFT



EQUALFANSAF

Likelihood That Cont. CanHandle Comb. Gas Burn


AFTSTREN2

TITLE

PAGE NUMBER DATE


Page 31 3/22/2007


FANSUNAVAILPRIPage 30Page 32



NOPDSFANS

CF


RV Failure

NORECOVFANSPRI

TITLE

PAGE NUMBER DATE


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


CS01


Failure

NORECOVSPPRI

Reactor Building Fans CanHandle Steam Production

RSGFANSG


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


RV Failure

ATPRESSHPage 16Page 11Page 1

... see x-ref



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


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


RV Failure

ATPRESSL

Page 4

Likelihood Strength CanHandle RSG Event and Base

Pressure is Low

RSGSTREN2


Vessel Failure

RECOVRV

TITLE

PAGE NUMBER DATE


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


LOWPRESS

Page 5

Plant Configuration andLayout Limits Material

Reaching Cont. Wall withLPME

CWLIMITLPME

Insufficient Amount of CoriumCan Make Contact WithContainment Wall With

HPME

NOCOREWALLHP


Failure

HIGHPRESS

Page 21

Plant Configuration andLayout Limits Material

Reaching Cont. Wall withHPME

CWLIMITHPME


Failure

RECOVRV

Containment Wall SurvivesContact With Corium

WALLSURVIV

TITLE

PAGE NUMBER DATE


Page 34 3/22/2007


SSHRAVAILPage 5


PDSSSHR

SSHR IS AVAILABLE

YES-SSHR

NO SSHR EXISTS

NO-SSHR

Page 35


Prior to RV Failure

RECOVSSHR

TITLE

PAGE NUMBER DATE


Page 35 3/22/2007

NO SSHR EXISTS

NO-SSHRPage 34Page 21

CAG0005


NO-SSHR-POSTLOOP


TITLE

PAGE NUMBER DATE


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


LOWPRESS

Page 5


Failure

RECOVRV

Likelihood That Cont Failureis Prevented Given a

Pressure Generated Missle

NOMISSLELIKE

TITLE

PAGE NUMBER DATE


Page 37 3/22/2007


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


Page 38 3/22/2007


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


Page 39 3/22/2007


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


Page 40 3/22/2007


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


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


Failure

RECOVRV


COREFREEZE


Floor

SPREADLOW


Failure

HIGHPRESS


PRESSURE CORE MELT

NOPDSLOW


IS CERTAIN

NOPDSLOW-2

Page 2



NOPDSLOW-1


NOPDSLOW_5



PDSINDET

CORE MELT BIN 4

CM-004

CORE MELT BIN 6

CM-006

CORE MELT BIN 5

CM-005


Generators

OPSNOSSHR


Unavailable

NOPORVSSHR


NOPDSSGADV

AV


SSHRUNAVAIL


NOPDSSSHR

Page 3



NORECOVSSHR


Generators

NOOPSDEPRESS


PORV

NOPZRPORV

Page 4



NOHEATIML


During Core Damage

NOPZRSAFETY


For Freezing

GEOMFREEZE



FREEZELOW

No Ex-Vessel Release of FPsto Cont. Atmos. and Water

Pool Avail.

RVORPOOL

Page 5

TITLE

PAGE NUMBER DATE


Page 2 3/22/2007


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


Page 3 3/22/2007


NOPDSSSHRPage 1

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP


TITLE

PAGE NUMBER DATE


Page 4 3/22/2007


NOPZRPORVPage 1


NOPDSPZRPORV

PO-HPICOOL


HPME

NOPRVHPCONF


NOPZPORVCONF-1


MANUALLY OPEN PORV

PZPORVCONF_99-C


NOT INITIATED

NOPZPORVCONF-2



PZPORVCONF_0-C


BWHBW1-----HP2OA

TITLE

PAGE NUMBER DATE


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


After RV Failure

FTRNSPOOLAFT


ARE AVAILABLE

PDSSPRAY


RBSPRAY


RECOVSPPRI


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


Page 6 3/22/2007


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


PRIBURNPage 14


Before RV Failure

PRICONC


Comb. Gas

NOLOWCONCBURN


Before RV Failure

PRIRELEASE

Page 2


FAILURE

SPARK


UNAVAILABLE

SPARK-1

Page 3


AVAILABLE

SPARK-2

Page 5


NOSTMINERTP

Page 1


ATBURN


RV Failure

NOATCONC


RV Failure

NOATRELEASE

Page 6


Comb. Gas

NOLOWCONCBURN

Ignition Source is Available atRV Failure

NOATIGNITION

Page 8


NOSTMINERTPPage 1


RV Failure

ATPRESSL

Page 9

Combustible Gas Burns EarlyAfter RV Failure

AFTBURN

Page 11



AFTERRECOM

Page 26

Cavity RecombinationDepletes Comb. Gas Prior to


LTRECOMB

Page 30



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


Page 2 3/22/2007


Before RV Failure

PRIRELEASEPage 1


NOOXIDIZED


NOTBOTTLED

Page 14

TITLE

PAGE NUMBER DATE


Page 3 3/22/2007


UNAVAILABLE

SPARK-1Page 1


SPRAY

SPARK_1


ARE UNAVAILABLE

NOPDSSPRAY

Page 4

TITLE

PAGE NUMBER DATE


Page 4 3/22/2007


ARE UNAVAILABLE

NOPDSSPRAYPage 11Page 36Page 24

... see x-ref


CS01

TITLE

PAGE NUMBER DATE


Page 5 3/22/2007


AVAILABLE

SPARK-2Page 1



SPARK_99


RBSPRAY

TITLE

PAGE NUMBER DATE


Page 6 3/22/2007


RV Failure

NOATRELEASEPage 1


H2 Concentration

NOPRIBURN

Page 11


NOOXIDIZED


Vessel Failure

NORECOVRV


BOTTLED

Page 7

TITLE

PAGE NUMBER DATE


Page 7 3/22/2007


BOTTLEDPage 11Page 6


PRESSURE CORE MELT

NOPDSLOWPage 21


IS CERTAIN

NOPDSLOW-2

Page 12



NOPDSLOW-1


NOPDSLOW_5



PDSINDET

Page 14



NOHEATIML


NOPZRPORV

Page 13


During Core Damage

NOPZRSAFETY

TITLE

PAGE NUMBER DATE


Page 8 3/22/2007


NOATIGNITIONPage 1


DISPERSE


NOGEOMH2


Failure

HIGHPRESS

Page 21


SPARKAT

TITLE

PAGE NUMBER DATE


Page 9 3/22/2007


RV Failure

ATPRESSLPage 1



NOPDSPRESSH

Page 10


FANSAT


FANSAVAILPRIPage 36



PDSFANS

CF


Failure

RECOVFANSPRI

TITLE

PAGE NUMBER DATE


Page 10 3/22/2007



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


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


Failure

NOAFTERREL

Previous Burns Do NotDeplete Hydrogen in

Containment

NOPRIGLOBAL


H2 Concentration

NOPRIBURNPage 6


Before RV Failure

NOPRICONC



LOWCONCBURN


Before RV Failure

NOPRIRELEASE


OXIDIZED


BOTTLED

Page 7


FAILURE

NOSPARK


UNAVAILABLE

NOSPARK-1


SPRAY

NOSPARK_9


ARE UNAVAILABLE

NOPDSSPRAY

Page 4


AVAILABLE

NOSPARK-2



NOSPARK_01


RBSPRAY


STMINERTP

Page 14


NOATBURN

Page 14


EXVPRODAFTH

Page 20



NOAFTERRECOM

Page 21


Comb. Gas

NOLOWCONCBURN


RV FAILURE

SPARKAFT

Page 24


NOSTMINERTAF

Page 25

TITLE

PAGE NUMBER DATE


Page 12 3/22/2007


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


Page 13 3/22/2007


NOPZRPORVPage 7

Page 21


NOPDSPZRPORV

PO-HPICOOL


HPME

NOPRVHPCONF


NOPZPORVCONF-1


MANUALLY OPEN PORV

PZPORVCONF_99-C


NOT INITIATED

NOPZPORVCONF-2



PZPORVCONF_0-C


BWHBW1-----HP2OA

TITLE

PAGE NUMBER DATE


Page 14 3/22/2007


NOATBURNPage 11

H2 Concentration isInsufficient to Cause Burns at

RV Failure

ATCONC


RV Failure

ATRELEASE


PRIBURN

Page 1


OXIDIZED


Failure

RECOVRV


NOTBOTTLEDPage 2


PRESSURE CORE MELT

PDSLOW

Page 14


to Heating

HEATIML


PZRPORV

Page 15


Core Damage

PZRSAFETY



LOWCONCBURN


ATIGNITION


NODISPERSE

Cavity Geometry Does RetainAll Corium

GEOMH2


LOWPRESSPage 26Page 36


PRESSURE CORE MELT

PDSLOWPage 14


IS CERTAIN

PDSLOW-1

CORE MELT BIN 1

CM-001

CORE MELT BIN 2

CM-002

CORE MELT BIN 3

CM-003



PDSLOW-2


PDSLOW_5



PDSINDETPage 7

CORE MELT BIN 4

CM-004

CORE MELT BIN 6

CM-006

CORE MELT BIN 5

CM-005


PZRPORV

Page 15


OPSSSHR

Page 16


to Heating

HEATIML


Core Damage

PZRSAFETY

Random Spark is Unavailableat RV Failure

NOSPARKAT


STMINERTPPage 11


RV Failure

ATPRESSH

Page 18

TITLE

PAGE NUMBER DATE


Page 15 3/22/2007


PZRPORVPage 14Page 14


PDSPZRPORV

PO-HPICOOL


PZPORVCONF-1


PZPORVCONF_99



PZPORVCONF-2


HPI Cooling

PZPORVCONF_0


BWHBW1-----HP2OA


PRVHPCONF

TITLE

PAGE NUMBER DATE


Page 16 3/22/2007


OPSSSHRPage 14


Are Available

PORVSSHR


PDSSGADV

AV


SSHRAVAIL


PDSSSHR

SSHR IS AVAILABLE

YES-SSHR

NO SSHR EXISTS

NO-SSHR

Page 17


Prior to RV Failure

RECOVSSHR


Generators

OPSDEPRESS

TITLE

PAGE NUMBER DATE


Page 17 3/22/2007

NO SSHR EXISTS

NO-SSHRPage 16Page 21

CAG0005


NO-SSHR-POSTLOOP


TITLE

PAGE NUMBER DATE


Page 18 3/22/2007


RV Failure

ATPRESSHPage 14



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


FANSUNAVAILPRI

Page 19

TITLE

PAGE NUMBER DATE


Page 19 3/22/2007


FANSUNAVAILPRIPage 31Page 18



NOPDSFANS

CF


RV Failure

NORECOVFANSPRI

TITLE

PAGE NUMBER DATE


Page 20 3/22/2007


EXVPRODAFTHPage 11

Corium Pool Does NotSpread Over Large Area Or

Freeze

NOCOREFREEZE


Or Cavity Floor

NOSPREADLOW

Page 36



NOFREEZELOW

Concrete Attack ProducesSufficient Combustible Gas

After RV Failure

NOH2SRCAFTER

Water Pool Does Not StopConcrete Attack In Cavity

After RV Failure

ATTKAFT


Failure

NOWATERAFTER

Page 26


Concrete Attack

MELT


Vessel Failure

NORECOVRV

TITLE

PAGE NUMBER DATE


Page 21 3/22/2007

Cavity Recombination DoesNot Deplete Combustible Gas


NOAFTERRECOMPage 11

Water Pool In Cavity AvailableEarly After RV Failure

WATERAFTER


Paths

OTHERWATER


After RV Failure

FTRNSPOOLAFT


ARE AVAILABLE

PDSSPRAY

Page 38


RECOVSPPRI


Failure

RECOVSPAFT


ACCUMAVAIL


Failure

HIGHPRESSPage 8


PRESSURE CORE MELT

NOPDSLOW

Page 7


Generators

OPSNOSSHR


Unavailable

NOPORVSSHR


NOPDSSGADV

AV


SSHRUNAVAIL


NOPDSSSHR

NO SSHR EXISTS

NO-SSHR

Page 17



NORECOVSSHR


Generators

NOOPSDEPRESS


PORV

NOPZRPORV

Page 13



NOHEATIML


During Core Damage

NOPZRSAFETY



GRAVFEEDAFT


PDSINJECCS

Page 22



NOSEQPRESSH

Page 23



NODRYEFF

TITLE

PAGE NUMBER DATE


Page 22 3/22/2007


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


Page 23 3/22/2007




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


Page 24 3/22/2007


RV FAILURE

SPARKAFTPage 11



SPARKAFT-1

PROB THAT SPARK ISAVAILABLE EARLY AFTER


SPARKAFT_99


RBSPRAY

RANDOM SPARK ISAVAILABLE EARLY AFTERRV FAILURE WITHOUT RB

SPRAY

SPARKAFT-2

PROB THAT SPARK ISAVAILABLE EARLY AFTERRV FAILURE WITHOUT RB

SPRAY

SPARKAFT_1


ARE UNAVAILABLE

NOPDSSPRAY

Page 4

TITLE

PAGE NUMBER DATE


Page 25 3/22/2007


NOSTMINERTAFPage 11


Containment

AFTPRESSL



NOSEQPRESSH

Page 23


Failure

FANSAFT

Page 36

TITLE

PAGE NUMBER DATE


Page 26 3/22/2007



AFTERRECOMPage 1


Failure

NOWATERAFTERPage 20


And Paths

NOOTHERWATER


ACCUMUNAVAIL


LOWPRESS

Page 14


Vessel

NOGRAVFEEDAFT


NOPDSINJECCS

Page 27



SEQPRESSH

Page 28



NOFTRNSPOOLAFT

Page 29



DRYEFF

TITLE

PAGE NUMBER DATE


Page 27 3/22/2007


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


Page 28 3/22/2007



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


Page 29 3/22/2007





ARE UNAVAILABLE

NOPDSSPRAY

Page 4


Failure

NORECOVSPPRI


Failure

NORECOVSPAFT

TITLE

PAGE NUMBER DATE


Page 30 3/22/2007

Cavity RecombinationDepletes Comb. Gas Prior to


LTRECOMBPage 1

Likelihood ThatRecombination Depletes

Comb. Gas With a Dry CavityLate

DRYEFFLT


NOSTMWATER

Page 36

TITLE

PAGE NUMBER DATE


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


Failure

NOFANSAFT


Failure

FANSUNAVAILAFT


FANSPRI


FANSUNAVAILPRI

Page 19



NOEQUALFANSPRI


RV Failure

NORECOVFANSAFT



NOEQUALFANSAF


FAILURE

NORECOVFANSLT

RB FANS UNAVAILABLEFOLLOWING POST-LOOP

RECOVERY

CF-R


HOURS

NORECOFFSITEPWR

Likelihood That RB Fans DoNot Survive Containment

Enviroment to Prevent LCF

NOEQUALFANSLT

PDS HAS NO LOW BASEPRESSURE IN


NOSEQPRESSL

Page 32

TITLE

PAGE NUMBER DATE


Page 32 3/22/2007

PDS HAS NO LOW BASEPRESSURE IN


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


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


HOURS

RECOFFSITEPWR

SPARK UNAVAILABLE;OFFSITE POWER NOT

RECOVERED

NOSPARKLT-2

RANDOM SPARKUNAVAILABLE WITHOUT

OFFSITE POWER

NOSPARKLT-NOP


HOURS

NORECOFFSITEPWR

TITLE

PAGE NUMBER DATE


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


NOINERTLT

Likelihood That Cont CanHandle Comb. Gas Burn


STREN1H2

Containment Strength CanHandle Late Comb. Gas Burn

and Base Pressure is Low

NOLTSTRENL


LTPRESSL

Reactor Building FansFunction Prior to LateContainment Failure

FANSLT


FANSAVAILLT

Page 36

Likelihood That RB FansSurvive Containment

Environment to Prevent LCF

EQUALFANSLT



SEQPRESSL

Page 35

Likelihood That Cont CanHandle Comb. Gas Burn


STREN2H2

Sequence Late After RVFailure Has Low Base


INERTLT

TITLE

PAGE NUMBER DATE


Page 35 3/22/2007



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


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


Or Cavity Floor

NOSPREADLOWPage 20


LOWPRESS

Page 14


NOGEOMFREEZE


NOSTMWATERPage 30



NOFTRNSPOOLLT


ARE UNAVAILABLE

NOPDSSPRAY

Page 4


Failure

NORECOVSPPRI



NORECOVSPLT


NORECOVSPLT-1



CS01-R


HOURS

NORECOFFSITEPWR

IE-LOOP-101


Failure

NORECOVSPAFT


Failure

RECOVRV


FANSAVAILLTPage 34


Failure

FANSAFTPage 25


Failure

FANSAVAILAFT


NOFANSPRI


FANSAVAILPRI

Page 9



EQUALFANSPRI


Failure

RECOVFANSAFT



EQUALFANSAF


RECOVFANSLT

Page 37

TITLE

PAGE NUMBER DATE


Page 37 3/22/2007


RECOVFANSLTPage 36


HOURS

RECOFFSITEPWR

AVAILABILITY OF RB FANSWITHOUT POWER

DEPENDENCY

RECFANSLT

TITLE

PAGE NUMBER DATE


Page 38 3/22/2007

No Containment FailureFrom Non Condensable

Gases

NOGASESPage 1


STMWATER



FTRNSPOOLLT


ARE AVAILABLE

PDSSPRAYPage 21


RBSPRAY


Failure

RECOVSPPRI


Failure

RECOVSPAFT



RECOVSPLT


HOURS

RECOFFSITEPWR



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


Page 39 3/22/2007


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


Page 40 3/22/2007


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


Page 41 3/22/2007


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


Page 42 3/22/2007


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


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


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


Core Damage

PZRSAFETY


PRESSURE CORE MELT

PDSLOW


IS CERTAIN

PDSLOW-1

CORE MELT BIN 1

CM-001

CORE MELT BIN 2

CM-002

CORE MELT BIN 3

CM-003



PDSLOW-2


PDSLOW_5



PDSINDET

Page 2


PZRPORV

Page 3

Primary Retention is Low ForHigh Pressure Core Melt

NOAMTHIGHP


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


Vessel Failure

NORECOVRV

TITLE

PAGE NUMBER DATE


Page 2 3/22/2007



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


Page 3 3/22/2007


PZRPORVPage 1


PDSPZRPORV

PO-HPICOOL


PZPORVCONF-1


PZPORVCONF_99



PZPORVCONF-2


HPI Cooling

PZPORVCONF_0


BWHBW1-----HP2OA


PRVHPCONF

TITLE

PAGE NUMBER DATE


Page 4 3/22/2007


HIGHPRESPCMPage 1


During Core Damage

NOPZRSAFETY


PRESSURE CORE MELT

NOPDSLOW


IS CERTAIN

NOPDSLOW-2

Page 5



NOPDSLOW-1


NOPDSLOW_5



PDSINDET

Page 2


NOPZRPORV


NOPDSPZRPORV

PO-HPICOOL


HPME

NOPRVHPCONF


NOPZPORVCONF-1


MANUALLY OPEN PORV

PZPORVCONF_99-C


NOT INITIATED

NOPZPORVCONF-2



PZPORVCONF_0-C


BWHBW1-----HP2OA

TITLE

PAGE NUMBER DATE


Page 5 3/22/2007


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


Page 6 3/22/2007

Secondary Side HeatRemoval Does Not Prevent

Revaporization

NOSSHRPage 1

Secondary Side HeatRemoval is Unavailable Prior

to Revaporization

SSHRRVPUNAVL


NOPDSSSHR

NO SSHR EXISTS

NO-SSHR

CAG0005


NO-SSHR-POSTLOOP


Secondary Side HeatRemoval is Not Recovered

Prior to Revaporization

SSHRNORVPREC

Likelihood That SSHR WillNot Prevent Revaporization

SSHRNOREVAP

TITLE

PAGE NUMBER DATE


Page 7 3/22/2007


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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