l!Y!!!! POWER .. -..SYSTEMS
Transcript of l!Y!!!! POWER .. -..SYSTEMS
C·E Power Systems Corntius!1c•:i En9,neer1ng inc 1000 F'rc>s~»:>cr Hri! Hoad ~\' nclsor c (_;~1necr1cu! 0609~1
l!Y!!!! POWER .. -..SYSTEMS
George Knighton, Chief Licensing Branch No. l U.S. Nuclear Regulatory Commission Washington, D.C. 20555
lPI 20J/6881 ]11 T PiPX 99297
September 5. l 9R5 LD-85-046
Subject: Informational Report on the Core Operating Limit Supervisory System (COLSS)
Dear Mr. Knighton:
The Core Protection Calcuhtor (CPC) Oversight Co11111ittee, consisting of Arizona IJuclear Power Project, Arkansas Power & Light Company, Louisiana Power & Light Company and Southern California Edison Company, with Combustion Engineering as its technical consultant, met \<lith the NRC staff on November 8, 1984, March 8, 1985, and April 18, 1985 to discuss a program of CPC Modifications and Methodology Improvements scheduled for implementation in 1986 and 1987. In these meetings the NRC indicated t~eir desire for an informational report describing the scope and methodology of COLSS.
The purpose of this letter is to provide the NRC with the requested report !Attachment (A) proprietary and Attachment (B) non-proprietary].
Combustion Engineering is providing this document on behalf of the CPC Oversight Committee. As the NRC staff (Core Performance Branch) requested, this document is being provided for informational purposes only and no ~IRC review is being requested. The attached document is not dpplicable to any individual licensee until referenced by that licensee for use on their docket.
It is requested that any questions you have about the attached document be addressed to the Chairman of the CPC Ovr.rsight Conwnittee, with copies sent to each committee member and C-E. Enclosed is a list of individuals to whom the copies should be sent.
Attachment (A) contains information considered by C-E to be proprietary in nature. As such, we request that it be withheld from public disclosure in accordance with the provisions of 10 CFR 2.790 and that this material be safeguarded. The reasons for the classification of this material as proprietary are delineated in the affidavit provided in Attachment (C).
fAJcL~ IO. C-02.E .fEpfo(t,..~ 5 C<l5
H. P-ooo / c. '1 .J . LO 1lf:o,.J I C 'I
Jb.B LE! I C..lf
E. l:t. C..J?Tl.4 I c.. 'j
Mr. George Knighton September 5, 1985
LD-85-046 Page 2
Since this document is being provided for informdtional purposes only and no review or approval is being requested, Combustion Engineering believes that no fees are to be incurred as a result of this submittal.
Should you have any questions on the contents of this letter, please feel free to contact me or Mr. H. C. Trwin of my staff at (203) 285-5210.
AES: bk s Enclosure cc: G. Hsi i
Very truly yours,
COMBUSTION ENGINEERING, INC.
~ Director Nuclear Licensing
L. Phillips L. Rubenstein
Attachments: (A) CEN-312-P, "Overview Description of the Core Operating Limit Supervisory Sys tern (COL SS)", September 1985: Copies 000001 thru 000009.
(8) CEN-312-NP: Non-proprietary Version of Attachment (A): 9 Copies Attached.
(C) Affidavit Attesting to the Proprietdry IJature of CDJ-312-P.
Correspondence List
Chairman of the COLSS/CPC Overs i gh._!__~omi t_tee
Mr. C. E. OeOeaux Louisiana Power and Light Company P.O. Box 60340 317 Baronne Street Mail Unit 17 New Orleans, Louisiana 70160
r-!~mbers _Qf_ the CPC Ove_~~gh!__~~n.!11_i_!:_tee
Arkansa~ Power -~-Li ght__f_~mp~ny
A. G. Mansell Arkansas Power & Light Company Post Office Box 551 Little Rock, Arkansas 72203
P . F. C raw l ey Arizond Nuclear Power Project Post Office Box 21666 Mail Station 4090 Phoenix, Arizona 85036
Sou_thern _f_C!] i f<?__r:_ni a l_di s_~_']__(:_~_mpan~
E. J. Donovan Southern California Edison Company Room 316 G. 0. 1 Post Office Box 800 Ro~emead, California 91770
Lou s i an a . P o~!-~~_!_gh ~-i=_<?_mp any
F. J. Drulllllond Louisiana Power & Light Company P.O. Box 60340 317 Baronne Street Mai 1 Unit 17 ~Jew Orleans, Lousiana 70160
Combustio~ineering, Inc.
A. E. Scherer Director, Nuclear Licensing Combustion Engineering, Inc. 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095
Enclosure to LD-85-046
Combustion Engineering, Inc. State of Connecticut County of Hartford
AFFIDAVIT PURSUANT
TO 10 CFR 2.790
SS.:
I, A. E. Scherer, depose and say that I am the Director, Nuclear Licensing,
of Combustion Engineering, Inc., duly authorized to make this affidavit, and
have reviewed or caused to have reviewed the information which is identified as
proprietary and referenced in the paragraph immediately h•?low. am submitting
''1i s J ff i davit in conformance with the provisions of 10 CFR 2. 790 of the
Commission 1 s regulations for withholding this information.
The information for which proprietary treatment is sought is contained in
the following document:
CEN-312-P, Overview Description of the Core OpPrating Limit Supervisory
System (COLSS), ~eptember 1985.
This document has been appropriately designated as proprietary.
I ~ave personal knowledge of the criteria and procedures utilized by
Combustion Engineering in designating information as a trade secret, privileged
or as confidential commercial or financial information.
Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the
Commission 1 s regulations, the following is furnished for consideration by the
Commission in determining whether the information sought to be withheld from
public disclosure, included in the above referenced document, should be
withheld.
-2-
1. The ir:",·::,utiun sought to be withhelct from public disclosurt' are the
methods usej i ! OP- l rH1d CETOP-n. <tnd, specific COL SS methods on density
dependent rar· p~aking. vector average tilt, and sensor cross-checking and
replacement, whir .. h is ownect and has been held in confidence by Combustion
Engineering.
2. The information consists of test data or other similar datd concerning
a process, method or component, the application of which results in a
substantial competitive advantage to Combustion Engineering.
3. The information is of a type customarily held in confidence by
Combustion Engineering and not customarily disclosed to the public. Combustion
[ngineering has a rational basis for determining the types of information
customarily held in confidence by it and, in that connection, utilizes a system
to determine when and whether to hold certain types of information in
confidence. The details of th~ aforementioned system were provided to the
Nuclear Regulatory Commission via letter DP-537 from F.M. Stern to Frank
Schroeder dated December 2, 1974. This system was applied in determining that
the subject document herein are proprietary.
4. The information is being transmitted to the Commission in confidence
under the provisions of 10 CFR 2.790 wfth the understanding that it is to be
received in confidence by tht· rmmtission.
5. The information, to the best of my knowledge and belief, is not
available in public sources, and any disclosure to third parties has been made
pursuant to regulatory provisions or proprietary agreements which pro vi de for
maintenance of the information f n confidence.
6. Public disclosure of the information is 11kely to cause substantial
harm to the competitive position of Combustion Engineering because:
-3-
a. A similar product is manufactured and sold by major pressurized
water reactor competitors of Combustion Engineering.
b. Development of this informdtion by C-E required tens of thousands
of manhours and hundreds of thousands of dollars. To the best of my knowledge
and belief a competitor would have to undergo similar expense in generating
equivalent information.
c. In order to acquire such information, a competitor would also
require considerable time and inconvenience in developing similar methods used
in CETOP-1 and CETOP-D, and, specific COLSS methods on density dependent radial
peaking, vector average tilt, and sensor cross-check111:J rnd replilcem1~nt.
d. The information required significant effort and expense to obtain
the licensing approvals necessary for application of the information.
Avoidance of this expense would decrease a competitor's cost in applying the
information and marketing the product to which the information is applicable.
e. The information consists of methods used in CETOP-1 and CETOP-D,
and, specific COLSS methods on density dependent radial peaking, vector average
tilt, and sensor cross-checking and replacement, the application of which
provides a competitive economic advantage. The availability of such
information to competitors would enable them to modify their product to better
compete with Combustion Engineering, take marketing or other actions to improve
their product's position or impair the position of Combustion Engineering's
product, and avoid developing similar data and analyses in support of their
processes, methods or apparatus.
f. In pricing Combustion Engineering's products and services,
significant research, development, engineering, analytical, manufacturing,
licensing, quality assurance and other costs and expenses must be included.
-4-
The ability of Combustion Engineering's competitors to utilize such information
without similar expenditure of resources may enable them to sell at prices
reflecting significantly lower costs.
g. Use of the information by competitors in the interndtional
marketplace would increase their ability to market nuclear steam supply systems
by reducing the costs associated with their technology development. In
addition, disclosure would have an adverse economic impact on Combustion
Engineering's potential for obtaining or maintaining foreign licensees.
Further ~he deponent sayeth not.
Director Nuclear Licensing
Sworn to before me
this ~s\·f\ day of ·,~-\J\S.W'-'* l 1 \ '-r-:3 c~>
\~-~-~~~(Q ___ b_ ___ ~~(( (}:~J:._ -Hotary Public
OVERVIEW DESCRIPTION
OF THE
CORE OPERATING LIMIT SUPERVISORY SYSTEM
(COLSS)
CEN-312-NP
Revision 00-NP
SEPTEMBER, 1985
COMBUSTION ENGINEERING, INC
Nuclear Power Systems
Power Systems Group
Windsor, Connecticut
8509100308 8~§61 PDR ADOCK 0 PDR p
• B) Due to nozzle outside diameter configuration,
the nozzle side cannot be examined. Due to branch connection, the head side of weld cannot be examined (Weld 3. 744).
C) Due to location of lifting lugs and nozzle configuration, 100~ of welds cannot be examined.
ABSTRACT
A nuclear power plant must be maintained within its limiting conditions for
operation as specified in the plant Technical Specifications to assure safe
operation. The Core Operating Limit Supervisory System (COLSS) aids the
operator in maintaining operating margin to limits on linea~ heat rate and
departure from nucleate boiling. To do so, COLSS uses meJsurements of incore
detector signals, CEA positions and plant thennal/hydraulic properties to
detennine the core power distribution and thermal performance.
This report provides a general description of the scope and methodology of the
COLSS algorithms. It is provided solely for information to be tJSed as a
reference during future reviews of submittals on the dockets of C-E supplied
NSSS's that utilize COLSS.
3
Table of Contents
1.0 Introduction and Su1T1Tiary
2.0 COLSS Description
2.1 Purpose of the COLSS System
2.2 Overview of COLSS Operation
2.2.1 System Inputs
2.2.2 Process Measurement Processing
2.2.3 COLSS Calc~lations
2.2.3.1 Volumetric Flow Calculation
2.2.3.2 Core Power Calculation
2.2.3.3 Power Distribution Calculation
2.2.3.4 Secondary Calorimetric Power Calculation
2.2.3.5 Local Power Density Power Operating Limit Calculation
2.2.3.6 Thermal Margin Power Operating Limit Calculation
2.2.3.7 Core Power Margin Calculation
2.2.~ COLSS Outputs
2.3 Description of COLSS Algorithms
4
2.3.1 Reactor Coolant System Volumetric Flow
2.3.2 Primary Calorimetric Power
2.3.3 Turbine Power
2.3.4 Secondary Calorimetric Power
2.3.4.1 Power in Each Steam Generator
2.3.4.2 Power Adjustments from the NSSS
8
10
10
11
12
12
14
15
15
16
J. 7
17
17
18
18
28
29
30
30
31
32
33
2.3.5
2.3.6
Table of Contents (Cont'd)
Plant Power
Core Power Distribution
2.3.6.1 Conversion of Flux to Power
2.3.6.2 Planar Radial Peaking Factors
2.3.6.3 Axial Power Distribution
2.3.6.4 Hot-Pin Integrated Radial and ASI
2.3.6.5 ~zimuthal Tilt
2.3.6.6 Three-0 Power Distribution
2.3.7
2.3.8
2.3.9
2. 3.10
Linear Heat Rate Power Operating Limit
Thermal Margin Power Operating Limit
Thermal Margin Power Operating Limit Update
Core Power Margin
2.4 Uncertainties
2.4.1 Power Measurement Bias
2.4.2 Power Operating Limit Uncertainties
3.0 Constants and Supporting Data
3.1 Basis for Mechanical and Thermal-Hydrau:ic Constants
3.2 Basis for Core Design Constants
5
3. 2 .1
3.2.2
3.2.3
3.2.4
3.2.5
Conversion of Flux to Power Constants
Planar Radial Peaking Factor look-up Tables
Axial Power Distribution Constants
Azimuthal Tilt Calculation Constants
LHR Limit Constants
34
34
35
36
37
38
38
39
40
41
42
43
44
44
45
46
46
42
48
48
49
50
so
Table of Contents (Cont'd)
3.3 Bdsis for DNB Margin Monitoring Constants
3.3.1 Derivation of the[
Flow Analysis
J from the Loss of
3.3.2 Other Events Analyzed to Confirm Adequate Monitoring
3.3.3 COLSS Penalty Factors Applied for CEA Calculat0rs
50
51
52
Inoperable 54
3.4 Basis for Measurement and Calculational Uncertainty Constants 54
3.5 Basis for Constants Supporting On-Line DNB Calculation 56
4.0 Conclusion
5.0 References
6
58
59
[
AOO
ASI
CEA
CHF
COL SS
CPC
CRT
DNB
ONBR
DNB-OPM
DP
F xy
KW/FT
LCO
LHR
LOCA
LOF
NSSS
PDIL
POL
RCP
RCS
RTD
7
GLOSSARY OF TERMS
Anticipated Operational Occurrence
Axial Shape Index
Control Element Assembly
Critical Heat Flux
Core Operating Limit Supervisory System
Core Protection Calculator
Cathode Ray Tube (display)
Departure From Nucleate Boiling
Departure From Nucleate Boiling Ratio
Departure From Nucleate Boiling Overpower Margin
Differential Pressure
Planar Radial Power Peaking Factor
Kilowatts per Foot
Limiting Condition for Operation
Linear Heat Rate
Loss of Coolant Accident
Loss of Flow (event)
Nuclear Steam Supply System
Power Dependent Insertion Limit
Power Operating Limit
Reactor Coolant Pump
Reactor Coolant System
Resistance Temperature Detector
-J
1.0 Introduction and Surrmary
Maintaining a nuclear power plant within its I imiting Conditions fer Operdtion
(LCO) is a necessary condition for safe operation and acceptable trans'2nt
consequences. These LCOs are delineated in the Technical Specifications.
There are many systems in a nuclear power plant that are used to help the
operators maintain the limiting conditions for operation. One such syster
used in C-E supplied NSSSs is the Core Operating Limit Supervisory System
(COLSS). COLSS is a digital computer based on-line monitoring system that is
used to provide information to aid the operator in complying with the Technical
Specifications operating limits on total core power, peak Linear Heat Rate
(LHR), Departure from Nucleate Boiling Ratio (DNBR), Axial Shape Index (ASI),
and azimuthal power tilt. The C-E Standard Technical Specifications discuss
the importance and purpose of these operating limits in the bases for Section
3.2. The system is used at the following plants:
1) Arkansas Nuclear One Unit 2,
2) San Onofre Nuclear Generating Station Units 2 and 3,
3) Waterford Unit 3, and
4) Palo Verde Nuclear Generating Station Units 1, 2, and 3.
COLSS uses input from selected sensors to determine the plant condition and
displays this condition to the operator in a form which allows easy
interpretation of reactor core status. Audible alarms and visual CRT messages
are provided to alert the operator when an operating limit is exceeded. COLSS
is a monitoring system and does not activate any safety equipment, initiate
8
any automatic actions, or provide any direct input to safety systems. The
major calculations performed by COLSS are:
1) Core Power,
2) Core Power Distribution,
3) Margin to Minimum Departure from Nucleate Boiling Ratio,
4) Margin to Linear Heat Rate Limit, and
5) Core Azimuthal Power Tilt Magnitude.
The purpose of this report is to provide a general description of COLSS for
reference during future review of submittals on the dockets of C-E supplied
NSSSs that utilize COLSS. To meet this purpose, the report describes:
1) COLSS monitoring and alarms which aid the operator in maintaining
the appropriate Technical Specification operating limits,
2) sen~or data and its processing for input to COLSS,
3) COLSS algorithm functions, and
4) determination of constants •or use in COLSS.
The accuracy of the information supplied by COLSS to the operator was orig
inally evaluated in Reference and has been updated in Reference 5. This
subject will not be addressed further in this report.
9
2.0 COLSS Description
2.1 Purpose of the COLSS System
The plant Technical Specifications specify Limiting Conditions for Operation
of plant systems, components, and parameters. Munitoring systems are provided
to assist the operator in meeting these Technical Specification requirements.
COLSS is a monitoring system that assists the p1Jnt operator in maintaining
the Limiting Conditions for Operation (LCD) specified in the following
Technical Specifications:
1) 3.2.1 Linear Heat Rate,
2) 3.2.3 Azimuthal Power Tilt,
3) 3.2.4 DNBR Margin,
4) 3.2.7 Axial Shape Index, and, for some plants,
5) 3.3.3.2 Incore Detector Operability.
An audible alarm and a visual CRT alarm message is initiated whenever any of
the parameters indicated above do not satisfy the LCO conditions required by
the Technical Specifications.
COLSS monitoring is accomplished by performing calculations using incore
detector signals, CEA positions, primary and secondary coolant pressure
measurements, and various temperature measurements and flow measurements to
monitor the following parameters:
10
1) margin to the peak Linear Heat Rate (LHR) limit,
2) margin to the Departure from Nucleate Boiling Ratio (ONBR) 1 in. l t '
3) margin to the licensea total core power,
4) azimuthal tilt, and
5) Axial Shape Index (ASI).
The function of the COLSS in the overall plant monitoring ard protection
system is illustrated in Figure 2-1. The protection function is provided by
the Core Protection Calculators (CPC) which cause a plant trip if necessary to
avoid violation of fuel design limits on LHR or DNBR. The COLSS monitoring
system reviews system behavior and alerts the plant operator to situations
where LHR or DNBR have reached their monitoring limits. In addition, COLSS
alerts the operator when other plant parameters (e.g., azimuthal tilt or axial
shape) are at prespecified limits. The Technical Specifications require
periodic review of specific aspects of the operation of both the monitoring
and protection systems relative to detailed calculations or specific measure
~ents to verify acceptable cperation and recalibrate as required.
2.2 Overview of COLSS Operation
The COLSS algorithms provide an integrated approach to monitoring those system
parameters important to the evalua~ion of LHR and ONBR. Rather than
restricting each parameter individually, COLSS uses its inputs to simulate the
core power distribution which is then used to directly evaluate the current
LHR and DNBR. From this evaluation, the power margin to the ONBR limit, to
the LHR limit, and to the licensed plant power are determined and compared to
11
alarm setpoints which monitor the requirements of the Technical Specifications.
Additional alarm limits are provided on Axial Shape Index (ASI) and azimuthal
tilt. If an alarm setpoint is violated, an alarm sequence is initiated to
alert the operator to the violation. The functional block diagram of Fig•Jre
2-2 illustrates the overall COLSS algorithm.
2. 2 .1 System Inputs
Table 2-1 provides a typical list of COLSS monitored variables. The specific
number of sensors and the sensor ranges can vary from plant to plant depending
on installed instrumentation. Figure 2-3 shows typical COLSS sensor locations.
All COLSS sensors are sampled at one second intervals except for the CEA
position indications, which are sampled at ten second intervals, and the
incore detectors, which are sampled at two second intervals for some plants.
2.2.2 Process Measurement Processing
The plant computer process cont~ol executive program processes system inputs
for use by COLSS. This processing includes taking the measurements, checking
the values against transducer limits, and conversion of measurements to
engineering units. If a measurement exceeds the associated transducer limits,
it is identified as invalid for use in later algorithms.
Additional measurement validity checking is performed internal to COLSS
[ncluding[ J 12
~This checking will alert the operator to the gradual
deterioration of a sensor.
When data being obtained from a sensor is determined to be invalid, the
operator is informed of the sensor failure by alarm and the data is marked
within COLSS as being invalid. To compensate for invalid data from a
particular sensor,~
13
liUIVlt)U~ I IUI~ tl'illll~tt.KINll
PROPRIETARY INFORMATION
2.2.3 COLSS Calculations
Portions cf the COLSS calculations are performed at one, ten, and thirty
second intervals and are synchronized with data acquisition rates. (e,(:.,
incore instruments are polled at 2 second intervals but used in COLSS power
distribution synthesis at 10 second intervals.) Calculations performed at one
second intervals include:
l) measurement process~ng,
2) reactor vessel volumetric flow calculation,
3) plant power calculation based on:
a) turbine first stage pressure
b) reactor coolant temperature rise across the core
4) update of the DNB power operating limit since the latest detai~ed
calculation, and
5) comparison of the plant pcwer to calculated limits.
Calculations performed at ten second intervals include:
1) axial power distribution synthesis,
2) azimuthal tilt calculat~ons,
14
3) local power density power operating limit calculations, anrl
4) comparison of AS! and azimuthal tilt to al1owe1 limits.
Calculations performed at thirty second intervals include:
I) secondary calorimetric calculations of reactor power, and
2) thermal margin power operating limit calculations.
2.2.3.1 Volumetric Flow Calculation
The volumetric flow for a single pump is based on differential pressure across
the pump, pump rotational speed, and water properties from the measured values
of cold leg temperature and primary system pressure. The total reactor
coolant system flow is derived by summing the individudl pump flows.
2.2.3.2 Core Power Calculation
Core power is determined by an auctioneering process between power values
calculated by a primary side calorimetric and a correlation to turbine first
stage pressure, both of which are calibr?.ted periodica11y to the secondary
side calorimetric. The primary calorimetric power is derived from the
calculated volumetric flow and water properties based on measured valu~s of
cold leg temperature, hot leg temperaturP., and reactor coolant system
pressure. The estimate of reactor power from turbine pressure is based on a
third order polynomial fit to turbine first stage pressure. The secondary
15
* ca 1 orimetri c power is derived from the measured va 1 ues of feed.,.,a ter fl ow ,
feedwater temperature. steam f1 ow, and secondary steam pressure. Appropriate
allowances for energy gains and losses are included.
2.2.3.3 Power Distribution Calculation
Signals from the fixed in-core neutron detectors and signalc from the CEA
pulst counter position indicators supply the input to the power distribution
calculations. The calculations performed include:
1) Determination of planar radial pea~ing factors based on CEAs present
in each axial plane,
2) Calculation of a normalized 40 node axial power distribution and a
3-D power peaking factor for use in the calculation of the LHR power
oper·lting limit,
3) Calculation of a core average axial shape index (ASI),
4) Determination of a 20 node hot channel axial power distribu~ion and
associated integrated radia1 peak for use in the calculat;on of the
thermal margin power operating limit to DNB, and
5) Calculation of azimuthal power tilt.
* Both Feedwater flow and steam flow are determined from differential pressure measurement across known flow restrictions.
16
2.2.3.4 Secondary Calorimetric Power Calculation
Secondary calorimetric power is derived from measurements of steam header
pressure. feedwater flow (as differential pressure), feedwater temperature,
and steam flow (as differential pressure). These inputs are used to perform
an energy balance on each steam generator and then the separate results are
added. Corrections are made to the secondary calorimetric power for energy
additions to and losses from the system, including letdown and charging pump
f1ows, reactor coolant pump heat input, pressurizer heat input, and heat
losses from NSSS components.
2.2.3.5 Linear Heat Rate Power Operating Limit Calculation
The power operating limit is based on the core average ful 1 power linear heat
rate, the linear heat rate limits (historically set by the LOCA), the
calculated 3-D power peaking factor, and the calculated azimuthal tilt. The
LHR limit can be provided as a function of both inlet temperat~re and axial
position.
2.2.3.6 Thermal Margin Power Operating Limit Calculation
The thermal margin power operating limit calculation is based on the same
methods used in the C-E developed thermal margin design computer code (CETOP)
and incorporates the CE-1 Critical Heat Flux (CHF) correlation (see References
2 and 3). This calculation uses measured data from cold leg temperature
sensors and reactor coolant system pressure sensors along with the hot channel
axial power distribution and the primary system volumetric flow calculated
17
previously. The detailed calculation is performed at 30 second intervals and
is updated at one second intervals based on changes in reactor coolant system
pressure, cold leg temperature, reactor coolant volumetric flow rate,
azimuthal tilt, and integrated radial peaking factors to provide the operator
with a smoother response to changes in plant conditions.
2.2.3.7 Core Power Margin Calculation
The core power margin calculation compares the actual power to the thermal
margin and LHR power operating limits (POL) and to the licensed power limit.
Two sets of checks are done. The first set consists of two margin calcu
lations using the present value of the core power and two POLs. The second
set consists of three margin calcu1ations using running averages of both the
power and the two POLs and includes calculation of the margin to the licensed
power limit. T:-iese latter three margins are called "smoothed" margins. rn
all, five margins are calculated and compared to appropriate limits. The
smallest of the smoothed margins is selected for disp1ay on tne digital panel
meter and the CRT display.
2,? .4 COLSS Outputs
A typical set of ~edicated COLSS outputs to the plant operator are listed in
Table 2-2. These outputs include displays of core power, power operating
limits, the minimum margin to any power operating limit, the COLSS master
alarm, and the azimuthal tilt alarm. The COL~S master alarm is activated when
licensed power is exceeded, when either power operating limit is exceeded, or
when a valid value of plant power or a power operating limit is unavailable.
18
This alarm is also activated when COLSS has been bypassed for testing. Sample
messages that can be displayed on the COLSS alarm CRT are given in Figur~ 2-4.
Additional displays and reports are incorporated in CJLSS to assist the
operator in monitoring the operation of the NSSS and in evaluating COLSS
alarms. These additional outputs are:
19
1) CRT displays of several hundred internal parameters (Figure 2-5
gives a sample of the types of parameters included),
2) a detailed printed report of all inputs and outputs,
3) an axial power distribution plot as illustrated in Figure 2-6,
4) a COLSS Failed Sensor Report listing all sensor inp~ts that have
failed validity checking, and
5) a Test Mode Report to verify correct operation of the COLSS program.
N 0
Measurement
Core volumetric flow
Core power Primary cal orimetr 1 c
Secondary calorimetric
Core power distribution
Reactor coolant pressure
Turbine power
TABLE 2-1
TYPICAL COLSS MONITORED PLANT VARIABLES
Sensors
Reactor coolant pump rotational speed
Reactor coolant pump differential pressure
Cold leg temperature Narrow range Wide range
Hot leg temperature
F eedwa ter fl ow
Steam fl ow tiP
Feedwater t~nperature
Steam pressure
tn-core monitoring system
CEA position
Pressurizer pressure
Turbine first stage steam pressure
Typical Number
2 per pump
2 per pump
1 per cold leg
1 per hot leg
1 per genera tor
1 per generator
1 per generator
1 per generator
44 to 61 incore assemblies with 5 axially stacked detectors each
l per CEA
2 (on pressurizer)
2 (on turbine)
Typical Range & Units
100 - 1200 RPM
0 - 150 PSID
525 - 62:.>F 0 - 600F
525 - 675F
0 780 in water
0 - 660 in water
100 - 500F
850 - 1050 PSIG
NA (power distribution is provided graphically)
0 - 150 inches
1,500-2,500 PSIA
u-l,000 PSIA
TABLE 2-2 TYPICAL DEDICATED COLSS OUTPUTS
TYPICAL UPDATE OUTPUT OUTPUT QUANTITY DI SPLAY RANGE UNITS FREQUENCY TYPE
Plant Power O to 125 " Power l Sec. Analog .'O
Power Operating Limit 0 to 125 CV Power Sec. Ana 1 og "' based on Linear Heat Rate
Power Operating Limit 0 to 125 <¥ Power 1 Sec. Analog ·"' based on Thermal Margin
Minimum Margin to an -50.0 to 125.9 % Power 1 Sec. Digital Operating Limit 4 Di git
Axial Shape Index -.7 to +.7 10 Sec. Analog
Margin alarm close - open 1 Sec. Contact
CPC Azimuthal Tilt alarm close - open 10 Sec. Contact
Tech. Specification close - open 10 Sec. Contact Azimuthal Tilt alarm
Axial Shape Index out close - open 10 Sec. Contact of limits alarm
21
N N
INPUT:
OUTPUT:
WHERE:
WHEN:
FUNCTION:
FIXED IN-CORE CEA POSITIONS SELECTED HOT & COLO LEG TEMPERATURE DETECTOR SIGNALS PRIMARY SYSTEM PRESSURE SECONDARY
PRIMARY COOLANT FLOW PROPERTIES
I I I I
' . • ,, • t •
IN-CORE REACTOR ANALYSIS -----· ENGINEER CDLSS PROGRAM
,L • I • DETAILED l·D POWER MARGIN TO LCD
DISTRIBUTION VERIFICATION OF LIMITS ON: • THERMAL MARGIN COLSS/CPC OUTPUTS •CORE POWER • TILT MAGNITUDE AND STORED CONSTANTS •PEAK LHR • BURNUP DISTRIBUTION • THERMAL MARGIN
• TILT MAGNITUDE
OFF-LINE (ON-SITE OR UTILITY OR C-E ON-LINE (PLANT OR REMOTE COMPUTER) ENGINEERING STAFF CORE MONITORING
COMPUTER )
ONDEMlllO AS NECESSARY SEVERAL TIMES PER MINUTE (AUTOMATICALLY,
PROVIDE DETAILED VERIFY COLSS/CPC PROVIDE INFORMATION CORE INFORMATION RESULTS ANO TO ASSIST OPERATOR IN FOR ANALYSIS AND ACCEPTABILITY OF MAINTAINING CORE INTERPRETATION BY STORED CONSTANTS CONDITIONS WITHIN ENGINEERING STAFF TECH. SPEC. OPERA TING LIMITS
FIGIJAE 2-1 OVERVIEW Of C-E CORE MONITORING AND PROTECTION SYSTEMS
EX·CORE DETECTOR SIGNALS
(3 SEGMENTS)
' ' , , CPC
J,
TRIP SIGNALS BASED ON: •PEAK LHR • THERMAL MARGIN
ON-LINE (CORE PROTECTION CALCULATORS)
SEVERAL TIMES PER SECONO (AUTOMATICALLY)
PROVIDE AUTOMATIC PROTECTION AGAINST EXCEEDING LHR AND DNBR FUEL DESIGN LIMITS
N UJ FEEOWATER TEMP
~ ST
FE
ST
EAll PRESSURE --EDWATER FLOW -.. EAM FLOW .__.
TU PR
Rll•E ht ST AGE _
ESSURE
TH OT
Tc :OLD
R
R
R
:PSPEEO
:P HEAD
:s PRESS
CE .A POSITIOIS
IN -CORE FLUX
-
-.. -.... -... --
-....
~
~ SECONDARY CALORIMETRIC
f+ POWER
i _.... - AUTOMATIC en CALIBRATION Of
TURBINE POWER ¥ TURBINE POWER u ~ .... ANO AT POWER rti w -:c TO SECONDARY u
CALORIMETRIC > ... POWER Q
:; rt- REACTOR COOLANT -t c
> AT POWER a: ,. i. 0 en z
I ..., -en -COO LAU -~ VOLUMETRIC ... r
~ FLOW RATE -- ... -... -..
- - -... FUEL PIN ANO ... COOLANT CHANNEL -- PLANAR RADIALS - CORE POWER l IMIT
rt BASED ON LOCAL .. POWER DENSITY r AZIMUTHAL TILT r+ MAGNITUDE :: AUDIBLE ALARM
NORMALIZED AXIAL ~ POWER DISTRIBUTION
FIGURE £-2 FUNCTIONAL DIAGRAM OF THE CORE OPERATING
LIMIT SUPERVISORY SYSTEM
ALARM/DISPLAY
t COMPARISON ...
CORE POWER LIMIT BASED
ON DNBA
• SELECTION OF
CORE POWER LIMIT -.. ~
LICENSED P )WER LIMIT
N ~
FIGURE 2·3 COLSS SENSOR LOCATIONS
FIRST STAGE PRESSURE
HIGH PRESS.
P~ESSURIZER PRESSURE
STEAM PRESSURE TURBINE STEAM PRESSURE
FEED· ATER
FLOW AP
STEAM FLOW uP
MAIN STEAM LINE 1
PRIP.4ARV COOLANT LINES
TEMPERATURE
LOOP1A
RPM FEEDWATER LINE 1
MOISTURE SEPARATORS
& REHEATERS
NOTE:
PUMP tlP
LOW PRESS. TURBINES
& CONDENSERS
RCP = REACTOR COOLANT PUMP
CONDENSATE PUMPS
MAIN STEAM LINE 2
TEMPERATURE
LOOP 2A I * \.!_A~RPM
UMP T\ .1P. PUMP AP
PRIMARY COOLANT LINES
TEMPERATURE
LOOP28
HEATERS
RPM
PUMP DP
FEEDWATER PUMPS
' I 8
STM. GEN. No. 2
FEEDWATER FLOW ~p
FEEOWATER LINE 2
----\ HEATERS I \
v FEEDWATER TEMPERATURES
FIGURE 2-4
TYPICAL ALARM CRT MESSAGES
Alann 1 Messages
(TIME)
XX:XX:XX ALARM COL SS DNBR POWER LIMIT EXCEEDED
XX:XX:XX ALARM COL SS KW/FT POWER LIMIT EXCEEDED
XX:XX:XX ALARM COL SS LICENSED POWER LIMIT EXCEEDED
XX:XX:XX ALARM COL SS INSTANTANEOUS DNBR POWER LIMIT EXCEEDED
XX:XX:XX ALARM COL SS INSTANTANEOUS KW/FT POWER LIMIT EXCEEDED
XX:XX:XX ALARM COL SS LPL ALARM DURATION EXCEEDED
XX:XX:XX ALARM COL SS DNBR ALARM DURATION EXCEEDED
XX:XX:XX ALARM COL SS KW/FT ALARM DURATION EXCEEDED
ALARM 2 and ALARM 3 Messages
XX:XX:XX
XX:XX:XX
XX:XX:XX
XX:XX:XX
ALARM
ALARM
ALARM
ALARM
Alarm 4 Messages
XX:XX:XX
XX:XX:XX
ALARM
ALARM
Other Alarm Messages
XX:XX:XX
XX:XX:XX
25
ALARM
AL.A.RM
COL SS
COL SS
COL SS
COL SS
COL SS
COL SS
COL SS
COL SS
CPC TILT LIMIT EXCEEDED
TECH SPEC TILT LIMIT EXCEEDED
CPC TILT ALARM DURATION EXCEEDED
TECH SPEC TILT ALARM DURATION EXCEEDED
ASI OUT OF LIMITS
ASI ALARM DURATION EXCEEDED
REMOVED FROM SERVICE
HOT LEG DEVIATION EXCEEDED
FIGURE 2-5
SAMPLE PARAMETERS FOR CRT DISPLAY
·· Parameter Descri2tion Usage Units
I COLSS HOT LEG TEMP-LOOP 1 INPUT DEG F COLSS HOT LEG TEMP LOOP 2 INPUT DEG F COLSS TURB lST STA~E PRES,PR INPUT PSIA COLSS TURB lST STAGE PRES,AL INPUT PSIA COLSS FW OUTLET TEMP, SGl INPUT DEG F COLSS FW OUTLET TEMP, SG2 INPUT DEG F COLSS FEEDWATER FLOW DP,SGl INPUT IN H20 COLSS FEEDWATER FLOW DP,SG2 INPUT IN H20 COLSS SECONDARY STEAM PR,SGl INPUT PSIG COLSS SECONDARY STEAM PR,SG2 INPUT PSIG COLSS STEAM FLOW OP, SGl INPUT IN H20 COLSS STEAM FLOW DP, SG2 INPUT IN H20 CEA REG GRP 1 MINIMUM POS INPUT IN CEA REG GRP 2 MINIMUM POS INPUT IN CEA REG GRP 3 MINIMUM POS INPUT IN CEA REG GRP 4 MINIMUM POS INPUT IN CEA REG GRP 5 MINIMUM POS INPUT IN CEA REG GRP 6 MINIMUM POS INPUT IN CEA REG GRP 1 MINIMUM PCS INPUT IN CEA REG GRP 2 MINIMUM POS INPUT IN CEA S 0 GRP 1 MINIMUM POS INPUT IN CEA S 0 GRP 2 MINIMUM POS INPUT IN CEA REG GRP 1 DEVIATION INPUT IN CEA REG GRP 2 DEVIATION INPUT IN CEA REG GRP 3 DEVIATION INPUT IN CEA REG GRP 4 DEVIATION INPUT IN CEA REG GRP 5 DEVIATION INPUT IN CEA REG GRP 6 DEVIATION INPUT IN CEA P L GRP 1 DEVIATION INPUT IN CEA P L GRP 2 DEVIATION INPUT IN CEA S D GRP 1 DEVIATION INPUT IN CEA S D GRP 2 DEVIATION INPUT IN DET SENSTVTY CORR FLUX INPUT NV*El4 RCP lA SPEED OUTPUT RPM RCP lB SPEED OUTPUT RPM RCP 2A SPEED OUTPUT RPM RCP 28 SPEED OUTPUT RPM RCP IA DIFF PRESS OUTPUT PSID RCP 18 DIFF PRESS OUTPUT PSID RCP 2A DIFF PRESS OUTPUT PSID RCP 28 DIFF PRESS OUTPUT PSID RCS PRESSRZR PRESS OUTPUT PSIA RCS LOOP IA COLD LEG TEMP OUTPUT DEG F RCS LOOP 18 COLD LEG TEMP OUTPUT DEG F RCS LOOP 2A COLD LEG TEMP OUTPUT DEG F RCS LOOP 28 COLD LEG TEMP OUTPUT DEG F
26
N ...... CORE AVERAGE AXIAL POWER SHAPE FROM COLSS
·-----·-----·-----·-..---·-----·-----·-----·-----·-----·-----·-----·-----·-----·~·------· .......... -----·........ . . .. • • • 1.523 .. 1.484 • 1.445 • 1.406 • 1.367 • 1.327 • 1.288 • 1.249 • 1.210 • 1.171 • 1.132 • 1.093 • 1.054 • • 1.015 •
. 976 •
. 937 •
.898 • •
.859 •
. 820 •
. 781 • •
.742 •
.703 •
. 664 •
. 625 • •
. 586 •
. 547 •
.508 •
. 469 ••
. 429 •
. 390 •
. 351 •
.312 •
.213 •
.234 •
. 196 •
.156 •
.117 •
.078 •
.039. .000 •
• PLR1 • PLR2 • REG1• REG2 • REGJ • REG4 • REGS• REG&•
•
• • • • • • • •
• • • • • • • • •
• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • •
••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••• • • • • • • • • ••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••
10 ·-----·-----·-----·-----·-----·-----·-----·-----·-----·-----·-----·-----·-----·-----·-----·-----·-----· ....... -·-----·-----·· 5
BOTTOM
15 20 25 30 35 40 45 50 55 60 65 10
CORE HEIGHT. PERCENT
FIGURE 2-6 COLSS POWER DISTRIBUTION PLOT
75 80 85 90 95 100
TOP
2.3 Description of COLSS Algorithms
As discussed in the overview (section 2.2), COLSS performs the following major
calculations:
1) calculation of reactor coolant system volumetric flow rate,
2) calculation of core power based on:
a) reactor coolant temperature rise across the core,
b) turbine first stage pressure, and
c) secondary system calorimetric,
3) calculation of core power distribution parameters including:
a) normalized core average axial power distribution.
b) azimuthal tilt magnitude,
c) hot channel integrated planar radial peaking factors and 3-0
peaking factors, and
4) calculation of power limits based on linear heat rate and on the
departure from nucleate boiling ratio (DNBR).
This section provides additional descriptions of these calculations. This
material is intended to provide a general description of the scope and
methodology implemented in the COLSS algorithms. References are provided, as
appropriate, to more detailed reports.
28
2.3.1 Reactor Coolant System Volumetric Flow
The volumetric flow calculation is performed every second and provides the
flow input needed for the calculation of primary calorimetric power and of the
power operating limit based on DNBR. The flow through each pump is ca1culated
based on sensor inputs of
1) pump rotational speed,
2) pump differential pressure,
3) cold leg temperature, and
4) Reactor Coolant System (RCS) pressure.
Following validity checking of sensor inputs, the specific volume of the water
entering the reactor coolant pumps is determined from cold leg temperature and
RCS pressure. The differential pressure is then converted to pump head and
is adjusted for the fraction of rated pump speed at which the pump is opera-
ting. This result is then used to calcuiate volumetric flow in gallons per
minute based on a polynomial fit to pump speed and the ratio of pump head
divided by the square of the fractional pump speed. The coefficients of this
fit are derived from pump testing. Total flow is then calculated as the sum
* of the flows from each of the four pumps . A normalized vessel volumetric
flow is also calculated.
The volumetric flows are also used to determine the mass flow rate for each
cold leg as the ratio of the volumetric flow rate to the specific volume of
the cold leg water. The total vessel mass flow which is the sum of the flows
through the four cold legs, is provided for operator information.
* allowance is made for core bypass flow in the DNBR calculation
29
2.3.2 Primary Calorimetric Power
The primary calorimetric power calculation is performed every second. This
calculation of power uses the volumetric flow already calculated for each pump
plus sensor inputs of:
1) RCS pressure,
2) cold leg temperature, and
3) hot leg temperature.
The primary calorimetric power calculation begins with the compensation of
each of the four cold leg temperature indications for sensor time response and
plenum mixing time. For each cold leg the compensation uses a digital filter
which is implemented using the present and previous values of cold leg
temperature and the previous value of the compensated cold leg temperature.
The coefficients of this fi1ter are explicitly determined from the time
responses and the period of the calculation.
The enthalpy of the water in each hot leg and cold leg is determined from
polynomial fits to the measured hot leg temperatures, the compensated cold leg
temperatures, and the reactor coolant system pressure. Power is then
calculated from the enthalpy change between the cold and hot legs.
2.3.3 Turbine Power
The turbine power calculation is performed every second. The only measured
input to this calculation is turbine first stage pressure. The calculated
30
power is given by a third order polynomial fit to the turbine pressure. All
coefficients in the fit are determined empirically.
2.3.4 Secondary Calorimetric Power
The secondary calorimetric calculation is performed once per 30 seconds using
input values that are averaged over the previous 10 seconds to reduce the
impact of sensor noise*. The measured inputs to this calculation for each
steam generator are:
1) feedwater flow pressure drop,
2) steam flow pressure drop,
3) feedwater temperature, and
4) secondary steam pressure.
The calculated secondary calorimetric power is the sum of the power
transferred to each steam generator and the energy lost from the system, less
energy additions to the system.
* For the Palo Verde COLSS, secondary calorimetric power itself is averaged
over several calculations rather than using averaged input parameters.
31
2.3.4.1 Power in Each Steam Generator
The power transferred to each steam generator is caiculated from feedwater
enthalpy, feedwater pressure, feedwater mass flow rate, feedwater specific
volume, steam mass flow rate, and steam generator pressure. The calculated
feedwater pressure (performed separately for each feed train) is the secondary
steam pressure corrected for pressure losses from the feed injection point
back to the pressure transducer.
The feeawater specif'~ volume (derived from the feedwater temperature and
pressure using sta,dard water properties) is used to convert the measured
feedwater flow prEssure drop to mass flow for each feed train. A small
temperature correc:ion is provided in this conversion to account for changes
in flow resistance .
The measured se 1 ondary pressure for each steam generator is corrected for
pressure losses uetween the steam generator and the sensor to obtain steam
generator pressure. The steam mass flow rate is calculated as the feedwater
flow minus the blowdown mass flow rate (an input constant).
The power transferred to each steam generator is then calculated as the
difference between enthalpy removal via the steam and blowdown mass flows, and
enthalpy entry via the feedwater mass flow. The quality of both steam flow
and blowdown flow are properly accounted for.
32
2.3.4.2 Power Adjustments from the NSSS
The calculated secondary calorimetric power is adjusted for power losses and
power credits to the NSSS. The power losses are determined from input
constants. On-line measured data is not used directly. The power losses that
are included are:
1) letdown mass flow rate and enthalpy,
2) reactor coolant pump seal cooling mass flow rate and enthalpy,
3) cooling water mass flow rate and enthalpy,
4) mass flow rate of other primary coolant water leaving the system and
its enthalpy,
5) power loss from the pressurizer
6) power loss from primary coolant p;ping,
7) combined power loss from steam generators, and
8) other energy losses from the NSSS.
Similarly, the power credits to the system are also based on input constants.
The power credits that are included are:
1) charging pump mass flow rate and enthalpy,
2) total power input from active reactor coolant pumps,
3) power input from pressurizer heaters,
4) other sources of power input from electrical equipment, and
5) all other power input to the NSSS.
33
The final calculation of secondary calorimetric power is then simply the sum
of the two steam generator powers plus the net NSSS power losses (i.e., total
losses minus total gains).
2.3.5 Plant Power
Both primary calorimetric power and turbine power are calibrated using a
correction factor based on the most r~cently performed secondary calorimetric
calculation of power.~
1 The larger of the two calibrated powers is
selected as plant power for display to the operator, for use in margin
calculations, and for use by the Power Dependent Insertion Limit (PDIL) CEA
Application Program. ~
L ] 2.3.6 Core Power Distr;hution
The major steps in deriving the core power distribution include:
1
1) conversion of incore fluy measurements to assembly relative power by
axial region,
2) determination of planar radial peaking factors from CEA position,
3) synthesis of a core average axial power distribution,
34
4) calculation of azimuthal tilts, and
5) synthesis of a pseudo hot pin power distribution.
2.3.6.1 Conversion of Flux to Power
Using methodology that is essentia11y identical to that used in CECOR
(Reference 4 ), this algorithm converts incore detector compensated neutron
flux to assembly relative power at each incore detector location at 10 second
intervals.
The flux to power conversion uses the incore detector compensated fluxes at
each of the five axial levels of each of the incore detector strings along
with the CEA group positions. The CEA group positions are used to provide an
additive correction to the conversion factor to account for shadowing of a
specific detector by a CEA in the same assembly.
For each string a power dependent correction factor is determined as a linear
function of plant power. The final conversion factor for a string is then the
sum of the CEA shadowing correction plus the product. of the burnup dependent
correction factor and the power dependen~ correction.
The burnup dependent component is calculated daily using the integr~ted power
at a detector 1ocation. This "integration" is done stepwise assuming that the
power has been constant over each 10 second interval. The depletion of fuel
in the vicinity of a given detector 'location is taken to be proportional to
35
the integrated power. The burnup dependent flux to power correction factor is
then given by a polynomial in burnup.
2.3.6.2 Planar Radial Peaking Factors
The appropriate planar radial peaking factors are determined for each axial
node by a table lookup pr0cess based on indicated CEA qroup positions. This
calculation is performed once per ten seconds and is done in two parts.
36
1) Planar radial peaking factor tables are stored for each of the
possible CEA configurations. For each configuration, the table
contains the planar radial peaking factG~
2) Penalty factors are applied to the radial peaking factors based on
the determination of out-of-sequence CEA group insertion and
excessive CEA deviations within any group. A pre-calculated CEA
out-of-sequence penalty multiplier is applied if any out-of-sequence
condition exists.
A secund penalty factor accounts for CEA deviations within a group.
The penalty factor for each CEA group is determined as a piece-wise
linear function of the size of the deviation. The final deviation
penalty factor is the product, over all groups, of the individual
penalty factors. The magnitude of the penalty factor applied
repends on the CEA group in which the deviation is occurring.
2.3.6.3 Axial Power Distribution
A forty node core average axial power distribution is calculated based on
in-core detector power signals using a five mode Fourier series expansion.
This calculation is performed once per ten seconds to provide the power
distribution used in the LHR calculation.
For each of the 5 detector levels, the assembly relative powers calculated
previously (see section 2.3.6.1.) are averaged over all incore locations with
valid signals. These average powers at each level are then normalized to have
a sum of 100%. The normalized detector signals are transformed into five
Fourier series weighting coeffici~nts by evaluating the matrix product of a
prestored "coefficient matrix" and the vector of detector signals. This
prestored matrix depends only on the integral of the five Fourier modes over
the axial length of the incore detectors. The 40 node power distribution is
then constructed by forming the sum. at each axial node. of the Fourier
functions (prestored in an array) times their respective coefficients. The
axial power distribution is normalized so that the average value of the axial
distribution is unity.
37
Once the axial power distribution is available, the core average ASI is
determined as the difference between the lower and the upper half core poher
fractions.
1
2.3.6.4 Hot-Pin Integrated Radial and ASI
A[ ]hot pin power distribution is determined as the product of the axial
power distribution and the planar radial peaking factor for each of thef:_ J nodes. The integrated radial peaking factor is then calculated as the average
of the hot pin power distribution over the[ ]axial nodes. The hot pin ASI is
calculated in the same manner as the core average ASI except for the use of
the hot pin power distribution.
2.3.6.5 Azimuthal Tilt
The core average azimuthal tilt is calculated from the assembly average powers
once per 10 seconds using methodology that is essentially identical to that in
CECOR (Reference 4). The incore detectors are divided into "tilt groups" of
38
four detectors with appropriate symmetry properties. Depending on the olant,
axial detector leoel. ~ ~ there are between nine and twelve tilt groups at each
the signals in opposite
~ For each tilt group,
quadrants are calculated.
the sum and difference of
These sums and differences
and a set of detector location dependent constants are used to calculate an
azimuthal tilt for each group.
The average azimuthal tilt at each level is then calculated as an "arithmetic
average 11 of the magnitude of the individual group tilts at that level. In
some plants,~
The core average azimuthal tilt is calculated by averaging the 5 level tilts
using a weighting factor for each level that is based on the number of valid
sets of detectors at that level. If the calculated azimuthal tilt is higher
than either the Technical Specification limit or the allowance used by tne
Core Protection Calculators (CPCs), then an alarm is initiated.
2.3.6.6 Three-D Power Distribution
The 3-0 power peaking factors are calculated for use in the linear heat rate
power operating limit calculation. The 40 node 3-0 power distribution is then
39
determined as the product of the radial peaking factor (Section 2.3.6.2) and
the value of the 40 node core average axial power distribution (Section
2.3.6.3) at each node. The maximum value of these products is the 3-D power
peaking factor which is made available for operator information.
2.3.7 Linear Heat Rate Power Operating Limit
The core power operating limit based on the Linear H~at Rate (LHR) limit is
calculated once per 10 se~onds. This calculation is used to monitor the LHR
limit normally established by Loss of Coolant Accident (LOCA) considerations.
The linear heat rate is calculated for each of the 40 nodes of the 3-0 power
distribution. This linear heat rate is the product of the normalized power
fraction in the node, the core average linear heat rate at rated power, and
the fraction of core power at which the plant is operating. Correction
factors are applied to account for the azimuthal tilt and modeling
uncertainties.
The power operating limit at each node is calculated as the product of plant
power, a correction factor to account for failed incore detectors, and the LHR
limit divided by the calculated linear heat rate for that node. The minimum
value calculated in this manner is the LHR power operating limit. It is this
value which is compared to the current value of plant power.
40
?.3.8 Thermal Margin Power Operating Limit
The thermal margin power operating limit is based on maintaining the calculated
ONBR above a specified minimum value (based on the CE-1 CHF correlation) and
maintaining the fluid quality below a specified maximum value at the point of
minimum ONBR. The thermal hydraulic model used to evaluate this limit is
based on the C-E proprietary code CEiOP and the CE-1 CHF correlation (see
references 2 and 3). This calculation is performed once per 30 seconds with a
dynamic update provided once per second.
The thermal-hydraulic modeling uses[
41
The calculat~on proceeds in an iterative manner in that an estimate of the
power operating limit (POL) is used to detennine the minirrum DNBR and tf,e
quality at that point. If both the DNBR and the quality are within their
respective limits, the algorithm raises the POL estimate and recalculates the
DNBR and quality. Similarly, if either of DNBR or quality are not within
their limits, the POL is lowered. This iteration continues until it finds the
maximum POL that meets both DNBR and quality limits.
The details of this calculation have been amply described in references 2 and
3 and wi11 not be repeated here. The calculation does incorporate an adjust
ment~ .~to account for the margin
required for the Loss of Flow event in the re~ultant POL. This adjustment is
discussed further in Section 3.3.
2.3.9 Thermal Margin Power Operating Limit Update
The detailed thermal margin calculation is only performed once per 30 seconds.
An approximate update to the most recent detailed calculation is performed
once a second tc provide the operator vJith a smoother indication of the cure
performance. The updated DNBR power operating limit (POL) is based on cnanges
in several measured and derived parameters including:
1) primary pressure,
2) maximum compensated cold leg t~mperature,
3) core flow rate,
4) integrated radial peaking factor,
42
5) azimuthal tilt,
6) c 7) quality at the node of minimum DNBR,
J 8) most recently calculated power operating limit.
9) POL derivative with respect to quality, and
10) POL derivative with respect to ONBR.
2.3.10 Core Power Margin
The core power margin calculation compares the actual power to the thennal
margin and LHR power operating limits (POL) and to the licensed power limit.
Two sets of checks are done. The first set consists of two margin
calculations using the present value of the core power and the two POLs. The
second set consists of three margin calculations using running averages of
both the power and the two POLs and includes calculation of the margin to the
licensed power limit. These latter three margins are called "smoothed"
margins. In all, five margins are calculated and compared to appropriate
limits. The smallest of the smoothed values is displayed on the digita1 panel
meter and CRT display and is referred to as MARGIN. If any of the 5
calculated margins is less than its respective limit, an alarm is initiated.
Before being used in these comparisons, the calculated power operating limits
are adjl'sted for power measurement biases. These biases are dependent on the
measured power level and on which of the three calculated powers have been
used to determine plant power (see section 2.4.1).
43
2.4 Uncertainties
The calculation of DNB and LHR power operatinJ limits requires numerous
measured inputs and calculated constants. Each of the measured inputs (i.e.,
temperature, pressure, etc.) and the calculated constants (i.e., fuel and
poison rod bow, system parameters, etc.) can have some uncertainty associated
with it. These uncertainties are applied in a conservative fashion to re]uce
the predicted power operating limits to ensure that adverse combinations of
uncertainties do not prevent alarms when limiting conditions for operations
are violated. References l and 5 describe the methods used to determine these
uncertainties.
2. 4. 1 Power Measurement Bias
The accuracy of the power measurement is a function of the frequency of
calibration and the method for determining the present power output. The
secondary calorimetric is the most accurate m~asure of reactor power and
generally has a net uncertainty of less than or equal to 2~ of rated power
near full power increasing to no more than( ]of rated power at low power.
The primary calorimetric and the turbine first stage pressure determinations
of power are less accurate having a typical uncertainty of 3.5~. Biases are
applied to the POLs to account for these uncertainties. All of the bias terms
are calculated as a function of plant power. Before a bias is applied to the
power operating limits, the validity of each of the three methods for deriving
plant power is determined. If the secondary calorimetric calculation is valid
then its bias is chosen since the other powers are periodically calibrated to
44
match it. If the secondary calorimetric is invalid, but either of the other
two power determinations is valid, then an appropriate, larger bias is
applied. The bias term is subtracted from the calculated power operating
limits to obtain the biased power operating limits.
2.4.2 Power Operating Limit Uncertainties
Other uncertainties associated with the calculation of the power operating
limits are accounted for in a conservative fashion in the power operating
limit algorithms by applying additive and/or multiplicative adjustment factors.
The uncertainty factors considered in the generation of these terms for DNBR
and LHR power operating limit calculations are:
45
I) uncertainty in in-core detector signal measurement,
2) uncertainty in Control Element Assembly (CEA) position measurement,
3) uncertainties in temperature, pressure, and flow measurements,
4) uncertainty in verification of tabulated planar radial peaking
factors (F ) using CECOR, xy
5) impact of the COLSS power distribution synthesis on the LHP
algorithms and DNB overpower margin,
6) uncertainty in COLSS DNB-OPM algorithm with respect to design
calculations,
7) computer processing uncertainties,
8) fuel and poison rod bow uncertainties,
9) global axial fuel densification uncertainty, and
10) engineering factors due to manufacturing tolerances.
The generation of these uncertainty terms is discussed briefly in Section 3.4.
A more detailed description can be found in Reference 5.
3.0 Constants and Supporting Data
To support the COLSS a1gorithms, numerous constants based on plant design
characteristics must be generated for incorporation into COLSS. These
constants can be divided into 5 major categories:
1) cons tan ts related to plant mechanical and thermal hydraulic design,
2) constants related to core design,
3) constants related to monitoring margin to limiting conditions for
operation,
4) constants re 1 a ted to measurement and calculational uncertainties,
and
5) constants required to support on~line ONBR calculations.
Each of these areas will be discussed to provide some background into the
basis for the constants in that area. General descriptions of the types of
analysis used to determine the constants will be provided where appropriate.
3.1 Basis for Mechanical and Thermal-Hydraulic Constants
Calculations of the RCS vol~metric flow rate and calibrated power depend on
constants which are based on the NSSS thermal hydraulic and mechanical design.
The volumetric flow calculation is determined by a polynomial fit to measured
values of RCP differential pressure and pump speeds. The constants for this
46
calculation are based on a curve fit of experimental pump characteristic data
obtained from operation of the RCP's in a test loop. RCS rated flow and RCP
rated speed are also used in the flow calculation.
The primary calorimetric power is based on calculated fluid enthalpies and
measured flows, temperatures, and pressures. No significant constants are
required to support a strictly static primary calorimetric power calculation
beyond standard water property tables. However, this calculation includes a
dynamic compensation of variations in cold leg temperatures. The cold leg
temperature compensation depends on the cold leg temperature sensor time
constant and the calculated plenum time constant based on RCS design (sensor
location, flow path, and .·1uw rate). The core rated power is provided as a
data base constant to permit normalization of the calculated power to percent
of rated power.
The secondary calorimetric power is based on measurements of feedwater flow,
steam flow, feedwater temperature, and steam pressure. Most of the constants
used in the power calculation are derived from or confirmed by field data
obtained during power ascension testing. These constants relate feedwater
pressure to secondary steam pressure and steam flow, relate steam generator
pressure to steam header pressure and steam flow, and quantify energy losses
from and credits to the system (including the gain associated with operation
of the RCP's).
The relationship of feedwater mass flow rate to feedwater temperature,
feedwater flow, and feedwater specific volume is based on venturi
47
characteristic test data. The calculated turbine power is based on a
polynomial which is fit to the data obtained during power ascension testing.
3.2 Basis for Core Design Constants
3.2.1 Conversion of Flux to Power Constants
The conversion of the flux signal for each incore detector to relative power
uses correlation coefficients that reflect detector location, local geometry,
and local burnup. These coefficients are the same as those used in the CECOR
off-line power distribution calculation (Reference 4). The System 80 plants
require additional adjustments in those bundles which have both a CEA and an
in-core detector string. Other C-E plants using COLSS do not have CEA's
entering instrumented assemblies anc do not require these adjustments.
3.2.2 Planar Radial Peaking Factor Look-up Tables
Prior to startup, neutron1cs calculations are performed to determine the
maximum expected planar radial peak for each CEA configuration allowed by the
CEA Power Dependent Insertion Limit (PDIL). Detailed calculations are
generally performed for the unrodded core and for CEA configurations containing
only the part length CEAs and the first two lead regulating banks.
Conservative, boundirg values are determined for other configurations including
those which involve insertir;~ of shutdown CEA banks.
48
The maximum radial peak expected is installed in COLSS for each configuration.
During start-up testing. measurements are performed with CECOR to verif; the
peaking factors for the CEA configurations that are permitted at higher
powers. Adjustments to the stored constants are made if appropriate. CECOR
calculations are performed periodically during the cycle to verify the
continued adequacy of the installed constants as required by the Technical
Specifications.
Pen~lty factor constants for CEA banks out of sequence and CEA misalignment
are determined to assure an alarm if the CEA misoperation degrades the margin
below the allowed LCO. These constants are based on analyses using standard
neutronics methods to determine the change in power distribution due to the
CEA misoperation.
An additional, addressable, multiplicative penalty factor on the radial peak
is available to compensate for special circumstances requiring change after
the COLSS constants have been installed.
3.2.3 Axial Power Distribution Constants
The incore detector signals are converted into a 40 node core average power
distribution using two arrays of constants. The first array converts the
planar averages of the incore detector signals to amplitude coefficients of a
Fourier series approximation of the axial power distribution. These constants
depend only on the axial location and the length of the incore detectors, and
on the Fourier modes used. The second array is a tabulation of Fourier mode
49
values at each of the axial locations which are precalculated to reduce the
COLSS calculation time.
3.2.4 Azimuthal Tilt Calculation Constants
The azimuthal tilt calculation requires detector location dependent constants
for each "tilt group" of four detectors and appropriate averaging factors.
These factors are used primarily to account for geometric effects {detector
location) but also include an average radial tilt sensitivity from 3-0
neutronics calculations. CECOR is run at regular intervals to verify the
accuracy of the COLSS azimuthal tilt calculations.
3.2.5 LHR Limit Constants
The maximum allowed steady state LHR limit specified in the Technical
Specifications and monitored by COLSS is typically based on the Loss Of
Coolant Accident (LOCA}. This limit is specified as a function of core inlet
temperature in the COLSS of some plants.
3.3 Basis for DNB Margin Monitoring Constants
The Limiting Conditions for Operation (LCO) in the Technical Specifications
assure that sufficient margin is available to cover the degradation in DNB
margin that can occur during any Anticipated Operational Occurrence (AOO).
Such a margin loss can be caused by an increase in local power or temperature,
by a decrease in core flow or pressure, or by an adverse change in the core
50
power distribution. The margin assured by the LCO is sufficient to cover
continued adverse changes from the time the event begins until either
corrective action is taken or a power reduction caused by a reactor trip
begins to recover margin.
COLSS monitors the marqin required by the LCOs throuqh the use of anf -
~is sufficient to acconmodate AOOs without violating a fuel
design 1 imi t. The[ Ji s defined as a function of AS I to reflect the
sensitivity of the margin loss during some AOO's to the initial axial power
distribution.
Historically, the Loss of Flow (LOF) analysis has determined the acceptable
J This
section provides a brief overview of the types of AOO that could limit the
thermal margin requirement.
3.3.1 Derivation of the~ ., j f ram the Loss of Fl ow Ana 1 ys is
For C-E plants, the Loss of Flow event has historically been the most limiting
with respect to thermal margin. During the few seconds after the pumps begin
51
slowing down and prior to a significant power reduction due to the CEA
insertion, the reduced flow causes a rapid decrease in ONB margin. SevLral
seconds into the event the heat flux/flow combination results in the minimum
ONBR that will be experienced during the transient. The specific time and
value of this minimum is a function of the axial power distribution and the
i ni tia l thermal and hydraulic cond iti ans in the core. r ] ( ]
Numerous power distributions and initial conditions are used to determine the
[ Jover the AS r range of
interest using the HERMITE (Reference 6) or CESEC (Reference 7) transient
codes. The~ ]calculated in this manner is represented in COLSS by a
piece-wise linear function of ASI which bounds the values determined in the
transient cases.
3.3.2 Other Events AnalyzeJ to Confirm Adequate Monitoring
Other AOO's can also result in degradation of thermal margin. If the safety
analysis were to indicate that one or more AOOs require more margin than the
LOF, then the~ ]would be adjusted so that COLSS monitors the larger thermal
margin requirement. Two events that could require more thermal margin than
LOF are the Asynmetric Steam Generator Transient (ASGT) and the CEA Drop.
An Asymmetric Steam Generator Transient may result from the inadvertent
closure of a Main Steam Isolation Valve. The resulting asyrrmetric core inlet
52
temperature distribution results in increased core power peaking on the cold
side. This event is protected by an asymmetric steam generator tr~nsient trip
on cold leg temperature difference (~T) in the CPC, but also requires that
adequate thermal margin be available to cover temperature asyrr.metries that
occur prior to trip actuation. This event is simulated by design transient
codes for different values of the ~T setpoint. The increase in the radial
peak used in the safety analysis is calculated as a function of the temperature
tilt using standard physics methodology.
available margin monitored by COLss[
for the ASGT event or the( Ji s adjusted
The selected ~T setpoint and the
lare shown to be adequate J
to allow for the extra margin.
Plants which include the Core Protection Calculator System as part of the
Reactor Protection System have the capability to accommodate deviated CEAs via
penalty factors generated by the CEA calculators. If a SAFDL violation is
conservatively predicted by the CPCs following ap~lication of the penalty,
then the reactor will trip. If not, operation can continue in accordance with
the Technical Specifications. As a result, COLSS has not been required to
verify that adequate margin has been set aside to cover margin degradation
during a dropped CEA event when the CEA calculators in the CPCs are operable.
Recent modifications in analysis methods on several plants have demonstrated
that the thermal margin assured by the LCOs is sufficient to accommodate a
dropped CEA event. This has eliminated the need for a reactor trip during CEA
insertion events or CEA drops. Thus, comparisons have been made using standard
neutronic methods to assure that the COLSS DNBR-POL calculation using the[ )
is conservative for these events. For these plants, COLSS has been modified
to incorporate [ J 53
3.3.3 COLSS Penalty Factors Applied for CEA Calculators Inoperable
If the CEA calculators of the CPCs are not in operation, automatic trip
protection for CEA deviation events is not provided. Therefore, adequate
margin must be set aside per the Technical Specifications. In the COLSS
calculation, this margin degradation during CEA related transients is accounted
for by an addressable input constant to the DNB calculation. The va1ue of
this constant is determined by simulating the CEA misoperation distortion
factor using neutronics codes.
3.4 Bas1s for Measurement and Calculational Uncertainty Constants
Two uncertainty penalties are calculated for COLSS; one which is used in
calculating the linear heat rate power operating limit and the other which is
used in calculating the DNBR power operating limit.
The LHR adjustment accounts for the composite modeling uncertainty in the
COLSS determination of the 3-D peak and for the various engineering factors.
This modeling error is determined from a set of several thousand comparison
cases between COLSS and design codes covering suitable ranges of power level,
core burnup, CEA position, and primary system fluid properties. The overall
adjustment factor accounts for the effects of fuel rod bow, poison rod bow,
design code modeling uncertainty, COLSS power algorithm uncertainty, CECOR F xy
54
measurement uncertainty, and computer processing uncertainties. In the COLSS
algorithm, this adjustment is applied as a multiplier to the core average
linear heat rate. This has the effect of reducing the linear heat rate power
operating limit.
Similarly, the ONBR adjustment accounts for the composite modeling uncertainty
in the COLSS determination of the hot pin power distribution and power as well
as the~ ~ [ ~This composite modeling error is based on the same set of comparison
cases between COLSS and design codes used for the LHR uncertainty calculation.
The overall adjustment factor includes the effects of fuel rod bow, poison rod
bow, design code modeling uncertainty, CECOR F measurement uncertainty, xy
COLSS DNB algorithm uncertainty, and computer processing uncertainties.
For most COLSS plants, the system uncertainties are combined statistically and
included in the minimum DNBR limit that is established for use with the CE-1
CHF correlation. The uncertainties accounted for include inlet flow
distribution uncertainties, fuel pellet density uncertainties, fuel pellet
enrichment uncertainties, fuel pellet diameter uncertainties, random and
systematic uncertainties in fuel clad diameter, random and systematic
uncertainties in fuel rod pitch, and CHF correlation uncertainties. In the
cases where the statistical combination method is not used, the various listed
uncertainties are accounted for by a multiplicative adjustment to the power
operating limit.
55
Details of the methodology used to determine the measurement and rJ~culational
uncertainties for COLSS can be found in Volume 3 of Reference 5 for the
statistical combination of uncertainties method or in reference 1 for the
alternate methods.
3.5 Basis for Constants Supporting On-Line DNB Calculations
The DNB calculations performed in COLSS use a simplified, faster running
version of the design CETOP code called CETOP-1. Most of the constants used
in CETOP-1 are identical to those in CETOP or are the product of CETOP
constants which are provided to reduce computer calculation time. Three
significant differences exist between CETOP-1 and CETOP to reduce the computer
run time.
56
4.0 Conclusion
The preceding discussions have provided an overview of the COLSS program as
used in recent C-E NSSS designs. This system uses measurements of incore
detector signals, CEA positions, and plant thermal-hydraulic properties to
provide an on-line determination of the core power distribution and ther~al
margin performance. The results of these calculations are provided to the
plant operator through various displays to aid him in maintaining the plant
within the Limiting Conditions for Operation as specified in the Technical
Specifications.
58
5. 0 References
*
*
1. "Assessment of the Accuracy of PWR Operating Limits as Determined by the
Core Operating Limits as Determined by the Core Operating Limit
Supervisory System (COLSS)", CENPD-169, July 1975.
2. "CETOP-D Code Structure and Modeling Methods for San Onofre Nuclear
Generating Station Units 2 and 311, CEN-160(S)-NP, September 1981.
3. "C-E Critical Heat Flux - Critical Heat Flux for C-E Fuel Assemblies with
Standard Spacer Grids", Part 1 CENPD-162-A September 1976, Part 2
CENPD-207-A December 1984.
4. 11 INCA/CECOR Power Peaking Uncertainty", CENPD-153, Rev. 1-A, May 1980.
5. "Statistical Combination of Uncertainties - Uncertainty Analysis of
Limiting Conditions for Operation of the San Onofre Generating Station
Units 2 and 311, Part 3, CEN-283(S)-NP, October 1984.
6. "HERMITE - A Multi-dimensional Space Time Kinetics Code for PWR
Transients: CENPD-1°8, March 1976.
7. "CESEC - Digital Simulation of a Combustion Engineering Nuclear Steam
Supply System" Enclosure 1-NP to LD-82-001, January 6, 1982.
* Plant specific references which are intended to be typical of similar references appropriate to other plants
59