ML20215M359
| ML20215M359 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 05/06/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215M354 | List: |
| References | |
| NUDOCS 8705130202 | |
| Download: ML20215M359 (4) | |
Text
'
o UNITED STATES g
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g NUCLEAR REGULATORY COMMISSION 7;.* \\
j i: E WASHINGTON,0. C. 20555
'..D SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL REPORT MSS-NAl-P SUPPLEMENT 1
" QUALIFICATION OF REACTOR PHYSICS METHODS FOR APPLICATION TO PRESSURIZED WATER REACTORS OF THE MIDDLE SOUTH UTILITIES SYSTEM"
- 1. 0 INTRODUCTION By letter dated March 20, 1987,~ Louisiana Power and Light Company (LPL) requested NRC review of MSS-NAl-P Supplement 1, " Qualification of Reactor Physics Methods for Application to Pressurized Water Reactors of the Middle South Utilities System".
This supplement provides comparisons between measurements and predictions for Waterford 3 using the physics methods previously approved by the NRC for use by Middle South Services (MSS) for Arkansas Nuclear One (ANO) Units 1 and 2 (Ref. 1).
In the original NRC safety evaluation report of MSS-NAl-P (Ref. 2) the physics methods and reliability factors were not approved for Waterford 3 appli-cation because comparisons of predictions to Waterford 3 measurements were not available at the time.
2.0 SAFETY EVALUATION The model used to analyze the Waterford 3 core is identical to that described in Reference 1.
It has been verified against Waterford 3 measurements made during Cycle 1 to quantify the reliability factors to be used in safety related calculations.
The term " reliability factor" is used to describe the allowances to be used in safety related calculations to assure conservatism.
The reliability factor is always larger than the one sigma uncertainty factor.
A bias factor is also applied to the average difference between the measured and calculated value of a parameter. The reliability factors applicable to the important physics parameters are listed in Table 1.
The control rod worth, soluble boron worth, and delayed neutron parameter reliability factors determined in Reference 1 were found to be applicable to Waterford 3.
For a single control rod or bank, however, a more conser-vative value of 0.10 is used for the reliability factor in Waterford 3.
Comparisons of measured and calculated temperature coefficients included six ANO-1 measurements, five ANO-2 measurements and seven Waterford 3 measurements.
Data from all 18 measurements were used to determine a reliability factor of 3.4 pcm/ F for the temperature coefficient.
Since this is conservatively bounded by the 4.0 pcm/ F reliability factor previously determined (Ref. 1), the staff concurs that a temperature coefficient reliability factor of 4.0 pcm/ F may be applied to Waterford 3.
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8705130202 870506 PDR ADOCK 05000302 P
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A.
y Since it is not possible to directly measure Doppler coefficients at Waterford 3, calculated Doppler coefficients were compared to two Electric Power Research Institute (EPRI) experiments, resulting in uncertainties of 2.90% and 7.23L Based on this, the 0.10 Doppler coefficient reliability factor previously used for AN0-1 and ANO-2 is deemed adequate and conser-vative for Waterford 3.
As in Reference 1, the Waterford 3 model reliability factors for calculat-ing power distributions are based on comparisons of measured and calculated in-core flux detector signals.
The Waterford 3 core is instrumented with 280 self powered rhodium detectors distributed at five axial core levels in 56 different fuel assemblies.
The signals from these detectors are corrected for detector sensitivity, depletion, background and leakage.
Fifteen core state points, representative of beginning, middle, and near end of cycle power shapes were used to compare measured and calculated reaction rates.
The local power distribution uncertainty is associated with the calculation of the peak to average fuel rod peaking within an assembly.
Since Water-ford 3 uses fuel assemblies similiar to ANO-2 with a common 16 x 16 fuel rod design, the ANO-2 uncertainty presented in Reference 1 is applicable to Waterford 3.
The power distribution reliability factors include the local linear heat rate peaking factor Fn and the enthalpy rise hot channel peaking factor F'.
Both are computVd for Waterford 3 using the same procedures des-cNbedinReference1.
Since a bounding value of 0.096 was found for the F reliability factor for Waterford 3, the proposed use of the more conse9vativevalueof0.10,determinedinReference1,isacceptable.
The reliability factor for F was calculated to be 0.046 for Waterford 3.
ThemoreconservativevaNeof0.057foundinReference1,however,will be used in the Waterford 3 safety related analyses.
This also is accept-able.
3.0 CONCLUSION
S The staff has reviewed Supplement 1 to MSS-NAl-P which extends the model validation presented in MSS-NAl-P for ANO-1 and ANO-2 to Waterford 3.
The staff concludes that this supplement adequately provides comparisons between physics parameter measurements and predictions for Waterford 3 and i
establishes appropriate calculational reliability factors for Waterford 3 application.
However, because of the somewhat limited data base used, the staff recommends that Middle South Services perform periodic reevaluations of the model validity as new data becomes available to provide continuing l
assurance of its applicability.
Principal Contributor:
L. Kopp Dated:
1 1
., _ _ _, _ _ ~ _ - - _ _ _ _., _ _.
4.0 REFERENCES
1.
" Qualification of Reactor Physics Methods for Application to Pres-surized Water Reactors of the Middle South Utilities System",
MSS-NAl-P, August 4, 1980.
2.
Letter from Robert A. Clark and John F. Stolz (NRC) to William Cavanaugh (APL), dated August 11, 1982 transmitting " Evaluation of Report MSS-NAl-P".
B
'C TABLE 1 PHYSICS RELIABILITY FACTORS FDR WATERFORD 3 Physics Parameter Reliability Factor Bias F
0.10 0
q F
0.057' 0
g Control Rod Worth (Pattern) 0.05 0
Control Rod Worth (Bank or Single Rod) 0.10 0
Temperature Coefficient 4.0 pcm/ F 0
Doppler Coefficient-0.10 0
Doppler Defect 0.20 0
Boron Worth 0.05 0
Delayed Neutron Parameters 0.03 0
I l
i L
May 6,1987 Docket No. 50-382 Mr. J. G. Dewease Senior Vice President - Nuclear Operations Louisiana Power and Light Company 317 Baronne Street, Mail Unit 17 New Orleans, Lousiana 70160
Dear Mr. Dewease:
SUBJECT:
REACTOR PHYSICS METHODS TOPICAL REPORT MSS-NAl-P (TAC NO. 65141)
By letter dated March 20, 1987 Louisiana Power and Light Company submitted topical report MSS-NAl-P, Supplement 1.
The staff has reviewed this report and concludes that the physics methods and reliability factors previously approved for Arkansas Nuclear One, Units 1 and 2 are also appropriate for application to Waterford 3.
However, because of the somewhat limited data base used, the staff recommends that Middle South Services perform periodic reevaluations of the model validity as new data becomes available to provide continuous assurance of its applicability.
Details of the staff's review are contained in the enclosed safety evalua-tion.
If you have any questions concerning the staff's review, please contact me at (301) 492-9403.
I 5l James H. Wilson, Project Manager ProjectDirectorate-IV Division of Reactor Projects - III, IV, V and Special Projects
Enclosure:
As stated cc w/ enclosure:
See next page cc:
See next page DISTRIBUTION Docket File NRC PDR Local PDR PD4 Reading D. Crutchfield F. Schroeder OGC E. Jordan J. Partlow J. Wilson P. Noonan L. Kopp M. Hodges L. Shao ACRS (10)
PD Plant File J. A. Calvo
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- f Mr. Jerrold G. Dewease Waterford 3 Louisiana Power & Light Company cc
W. Malcolm Stevenson, Esq.
Regional Administrator, Region IV Monroe & Leman U.S. Nuclear Regulatory Commission-1432 Whitney Building Office of Executive Director for New Orleans, Louisiana 70103 Operations 611 Ryan Plaza Drive, Suite 1000 Mr. E. Blake Arlington, Texas 76011 Shaw, Pittman, Potts & Trowbridge
,2300 N Street, NW Carole H. Burnstein, Esq.
Washington, D.C.
20037 445 Walnut Street New Orleans, Louisiana 70118 Mr. Gary L. Groesch Post Office Box 791169 Mr. Charles B. Brinkman, Manager New Orleans, Louisiana 70179-1169 Washington Nuclear Operations Combustion Engineering, Inc.
Mr. F. J. Drummond 7910 Woodmont Avenue, Suite 1310 Project Manager - Nuclear Bethesda, Maryland 10814 Louisiana Power & Light Company 317 Baronne Street Mr. William H. Spell, Administrator New Orleans, Louisiana 70160 Nuclear Energy Division Office of Environmental Affairs Mr. K. W. Cook Post Office Box 14690 -
Nuclear Support and Licensing Manager Baton Rouge, Louisiana 70898 Louisiana Power & Light Company 317 Baronne Street President, Policy Jury New Orleans, Louisiana 70160 St. Charles Parris Mahnville, Louisiana 70057 Resident Inspector /Waterford NPS Post Office Box 822 Killona, Louisiana 70066 Mr. Ralph T. Lally Manager of Quality Assurance Middle South Servies, Inc.
Post Office Box 61000 New Orleans, Louisiana 70161 Chairman Louisiana Public Service Commission One American Place Suite 1630 BatonRouge, Louisiana 70825-1697
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UNITED STATES NUCLEAR REGULATORY COMMISSION n
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1
RELATING TO TOPICAL REPORT MSS-NAl-P SUPPLEMENT'l
" QUALIFICATION OF REACTOR PHYSICS METHODS FOR APPLICATION TO j
PRESSURIZE 0 WATER REACTORS OF THE MIDDLE SOUTH UTILITIES SYSTEM"
1.0 INTRODUCTION
By letter dated March 20, 1987, Louisiana Power and Light Company (LPL) requested-NRC review of MSS-NAl-P Supplement 1, " Qualification of Reactor Physics Methods for Application to Pressurized Water Reactors of the
[
Middle South Utilities System".
This supplement provides comparisons between measurements and predictions for Waterford 3 using the physics methods previously approved by the NRC for use by Middle South Services 4
(MSS) for Arkansas Nuclear One (ANO) Units 1 and 2 (Ref. 1).
In the original NRC safety evaluation report of MSS-NAl-P (Ref. 2) the physics methods and reliability factors were not approved for Waterf6rd 3 appli-cation because comparisons of predictions to Waterford 3 measurements were not available at the time.
].
2.0 SAFETY EVALUATION
[
The model used to analyze the Waterford 3 core is identical to that i
described in Reference 1.
It has been verified against Waterford 3 measurements made during Cycle 1 to quantify the reliability factors to be used in safety related calculations.
The term " reliability factor" is i
used to describe the allowances to be used in safety related calculations i
to assure conservatism.
The reliability factor is always larger than the i
j-one sigma uncertainty factor.
A bias factor is also applied to the
!j average difference between the measured and calculated value of a parameter. The reliability factors applicable to the important physics l
parameters are listed in Table 1.
t The control rod worth, soluble boron worth, and delayed neutron parameter reliability factors determined in Reference 1 were found to be applicable to Waterford 3.
For a single control rod or bank, however, a more conser-j vative value of 0.10 is used for the reliability factor in Waterford 3.
Comparisons of measured and calculated temperature coefficients included i
i six ANO-1 measurements, five ANO-2 measurements and seven Waterford 3 measurements.
Data from all 18 measurements were used to determine a l
reliability factor of 3.4 pcm/*F for the temperature coefficient.
Since this is conservatively bounded by the 4.0 pcm/'F reliability factor previously determined (Ref. 1), the staff concurs that a temperature coefficient reliability factor of 4.0 pcm/*F may be applied to j
Waterford 3.
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, Since it is not possible to directly measure Doppler coefficients at"
-Waterford 3, calculated Doppler coefficients were compared to two Electric Power Research Institute-(EPRI) experiments, resulting in uncertainties of 2.90% and 7.23%.
Based on this, the 0.10 Doppler coefficient reliability factor previously used for'ANO-1 and ANO-2 is deemed adequate and conser-vative for Waterford 3.
I
'As in Reference 1, the Waterford 3 model reliability factors for calculat-ing power distributions are based on comparisons of measured and calculated in-core flux detector signals. The Waterford 3 core is. instrumented with
~
280 self powered rhodium detectors distributed at five axial core levels in 56 different fuel assemblies.
The signals from these detectors are y
corrected for detector sensitivity, depletion, background and leakage.
Fifteen core state points, representative of beginning, middle, and near 1
i end of cycle power shapes were used to compare measured and calculated reaction rates.
The local power distribution uncertainty is associated with the calculation i
of the peak to average fuel rod peaking within an assembly.
Since Water-ford 3 uses fuel assemblies similiar to ANO-2 with a common 16 x 16 fuel 1
rod design, the ANO-2 uncertainty presented in Reference 1 is applicable to Waterford 3.
l The power distribution reliability factors include the local linear heat
... h rate peaking factor F and the enthalpy rise hot channel peaking factor F
BotharecomputSdforWaterford3usingthesameproceduresdes-cNbed in Reference 1.
Since a bounding value of 0.096 was found for the Fn reliability factor for Waterford 3 the proposed use of the more conseYvative value of 0.10, determined in, Reference 1, is acceptable.
The i
reliability factor for F was calculated to be 0.046 for Waterford 3.
~
w The more conservative vaTue of 0.057 found in Reference 1, however, will 1
be'used in the Waterford 3 safety related analyses.
This also is accept-l able.
l
3.0 CONCLUSION
S i
}-
The staff has reviewed Supplement I to MSS-NAl-P which extends the model 1
validation presented in MSS-NAl-P for ANO-1 and ANO-2 to Waterford 3.
The i
staff concludes that this supplement adequately provides comparisons i
between 1hysics parameter measurements and predictions for Waterford 3 and establisles appropriate calculational reliability factors for Waterford 3 i
application.
However, because of the somewhat limited data base used, the i
staff recommends that Middle South Services perform periodic reevaluations j
of the model validity as new data becomes available to provide continuing assurance of its applicability.
Principal Contributor:
L. Kopp l
Dated:
i
.i
4.0 REFERENCES
1.
" Qualification of Reactor Physics Methods for Application to Pres-surized Water Reactors of the Middle South Utilities System",
MSS-NAl-P, August 4, 1980.
2.-
Letter from Robert A. Clark and John F. Stolz (NRC) to William Cavanaugh (APL), dated August 11, 1982 transmitting " Evaluation of Report MSS-NAl-P".
O
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TABLE 1 PHYSICS RELIABILITY FACTORS FOR WATERFORD 3 Physics Parameter Reliability Factor
-Bias F
0.10 0
q F
0.057 0
g Control Rod Worth (Pattern) 0.05 0
Control Rod Worth (Bank or Single Rod) 0.10 0
Temperature Coefficient 4.0 pcm/ F 0
Doppler Coefficient 0.10 0
Doppler Defect 0.20 0
Baron Worth 0.05 0
Delayed Neutron Parameters 0.03 0
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