ML17331B199
| ML17331B199 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/12/1994 |
| From: | Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AEP:NRC:1118G, NUDOCS 9401270213 | |
| Download: ML17331B199 (13) | |
Text
ACCELERATED DI<TRJBUTION DEMONSTRATION SYSTEM W
~ '
W REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9401270213 DOC.DATE: 94/01/12 NOTARIZED: NO DOCKET FACXL:50-315 Donald C.
Cook Nuclear Power Plant, Unit 1, Indiana M
05000315 50-316 Donald C.
Cook Nuclear Power Plant, Unit 2, Indiana M
05000316 AUTH.NAME AUTHOR AFFILIATION FITZPATRICK,E.
(formerly Indiana
& Michigan Ele RECIP.NAME RECIPXENT AFFILIATION R
MURLEY,T.E.
Document Control Branch (Document Control Desk)
I
SUBJECT:
Notifies of LOCA model changes or errors that meet definition of significant as defined in 10CFR50.46.
Determination of effect of LOCA model changes on pint SBLOCA analyses encl.
DISTRIBUTION CODE:
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D RECIPIENT XD CODE/NAME PD3-1 LA HXCKMAN,J INTERNAL: NRR/DE/EELB NRR/DRCH/HICB NRR/DSSA/SPLB NUDOCS-ABSTRACT OGC/HDS2 EXTERNAL: NRC PDR COPIES LTTR ENCL 1
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1 RECIPIENT ID CODE/NAME PD3-1 PD NRR/DORS/OTS B NRR/DRPW NRR/DSSA/SRXB OC LFDCB G
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Indiana Michigan power Company p.O. Box 16631 Columbus, OH 43216 ELM MA NECNEGQM PQWM AEP:NRC:1118G Donald C.
Cook Nuclear Plant Units 1 and 2
Docket Nos.
50-315 and 50-316 License No.
DPR-58 and DPR-74 REPORT OF SIGNIFICANT LOCA EVALUATION MODEL CHANGES PURSUANT TO 10CFR50.46(a)(3)(ii)
U.
S. Nuclear Regulatory Commission Document Control Desk Washington, D.
C.
20555 Attn:
T.
E. Murley January 12, 1994
Dear Dr. Murley:
Pursuant to the requirements of 10CFR50.46(a)(3)(ii), this letter provides notification of LOCA model changes or errors that meet the definition of significant as defined in 10CFR50.46
'pecifically, this letter reports on the effects of an input error in auxiliary feedwater flow rate used to analyze a small break LOCA analysis for Unit 1.
This letter also serves to provide additional information regarding an error in the emergency core cooling system evaluation methodology used by Westinghouse for Units 1 and 2 that was reported previously in our letter AEP:NRC:1118F, dated October 19, 1993.
The original small break LOCA (SBLOCA) analysis for Units 1 and 2 of Donald C.
Cook Nuclear Plant was performed by Westinghouse Electric Corporation (Westinghouse) using their NOTRUMP computer code.
Subsequently, Cook Nuclear Plant Units 1
and 2
were reanalyzed to determine the impact of an increase in the main steam safety valve (MSSV) setpoint tolerance from +
1% to + 3S.
This reanalysis was submitted to the NRC in support of our request to modify the MSSV Technical Specifications via our letter AEP:NRC:1169, dated November 11, 1992.
We have discovered that the Unit 1
SBLOCA reanalysis submitted via letter AEP:NRC:1169 used nominal auxiliary feedwater flow rates (1258 gpm total delivery) instead of minimum auxiliary feedwater flow rates (750 gpm total delivery).
Since minimum auxiliary gggfb P}
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Dr. T. E. Murley AEP:NRC:1118G feedwater flow rates are more limiting, the small break LOCA for Cook Nuclear Plant Unit 1 for + 3S MSSV setpoint tolerance has been reanalyzed using minimum auxiliary feedwater flow rates (750 gpm total delivery).
The reanalysis is being reported under the provisions of 10CFR50.46 because the MSSV setpoint tolerance relaxation SBLOCA analyses have been included in our 10CFR50.46 reports since AEP:NRC:1118D, dated March 12,
- 1993, was submitted.
Therefore, we believe it is appropriate to report errors found in these analyses under 10CFR50.46 even though they have not yet been approved and are not yet part of our licensing basis.
These analyses include in their modeling non-discretionary changes to the SBLOCA model as defined in WCAP-13251.
The MSSV SBLOCA analyses bound currently approved operational limits and are therefore conservative.
The revised analysis shows that the total peak clad temperature (PCT) with minimum auxiliary feedwater flow rate with high head safety injection cross ties closed is 2068 F, while the PCT reported in AEP:NRC:1169 with nominal auxiliary feedwater was 1878.7~F.
Since the absolute value of the change in calculated PCT exceeds 50'F, the change meets the definition of significant provided in 10CFR50.46.
Attachment 1
contains the peak clad temperatures calculated specifically for the small break LOCA analyses for Cook Nuclear Plant Unit 1.
The peak clad temperatures for the large break LOCA (LBLOCA) remain the same as reported to the NRC via our letter AEP:NRC:1118D, dated March 12, 1993 'he licensing basis PCT plus permanent assessment for SBLOCA for Cook Nuclear Plant includes all previous permanent 10CFR50.46 model assessments.
The assessment of this report consists of a PCT change of 189.3 F to the Unit 1 SBLOCA cross ties closed analysis because of the NOTRUMP code input error in auxiliary feedwater flow rate.
The SBLOCA analysis with high head safety injection cross-tie valve open has not yet been performed by Westinghouse.
For this case, the change in PCT for model assessments has been estimated to be 97oF Regarding the second issue of error in Emergency Core Cooling System (ECCS) methodology, our letter AEP:NRC:1118F, dated October 19, 1993, reported errors which Westinghouse discovered in their NOTRUMP computer code used for SBLOCA analysis, and an error in ECCS evaluation methodology used by Westinghouse for
yt g
D
Dr. T.
E. Murley AEP:NRC:1118G both Units 1
and 2
of Cook Nuclear Plant.
The following additional information on the Westinghouse ECCS evaluation methodology has been provided to us by the Westinghouse Owners Group (WOG).
Westinghouse plans to meet with the NRC in January 1994 to present the results of the investigations performed for safety injection (SI) in the broken loop, and discuss the condensation on safety injection test data and resulting correlation.
This meeting will serve as the focal point for determining the level of detail and schedule required by the NRC in order to complete their review of the SI in the broken loop issue.
Should final resolution of the SI in the broken loop issue not change the conclusion of Westinghouse letter ET-NRC-93-3971, dated September 21,
- 1993, and our letter AEP:NRC:1118F, we will not file an additional report.
- However, should final resolution of this issue change the conclusion of our current 30 day report, Indiana Michigan Power Company will provide further information to the NRC under 10CFR50.46 requirements.
Attachment 2 provides replacement pages for pages 1, 2, and 5 of Attachment 3 to our submittal AEP:NRC:1118F, which contained inadvertent errors.
Regarding plans for future analysis, the MSSV analysis will provide the new analysis of record for SBLOCA for Unit 1.
The cross-tie valve open case is scheduled to be completed in June 1994.
S rel E.
E. Fitz atrick Vice President cad Attachments cc:
A. A. Blind - Bridgman G. Charnoff J.
B. Martin - Region III NFEM Section Chief NRC Resident Inspector
- Bridgman J.
R. Padgett
ATTACHMENT 1 TO AEP:NRC:1118G DETERMINATION OF EFFECT OF LOCA MODEL CHANGES ON COOK NUCLEAR PLANT SMALL BREAK LOCA ANALYSES
Attachment 1 to AEP:NRC:1118G Page 1
SMALL BREAK. LOCA PLANT NAME:
DONALD C.
COOK NUCLEAR PLANT UNIT 1 A.
ANALYSIS OF RECORD PCT-2122oF (Comments:
Evaluation Model:
NOTRUMP, FQT-2 32, FdH-1 55, SGTP-15X, Other HHSI Cross Tie Valve Closed 3250 Mwt Reactor Power B.
10 CFR 50.92 SAFETY EVALUATIONS C.
PRIOR PERMANENT LOCA MODEL ASSESSMENTS 8PCT-
-268'Fi 6PCT
-10 F D,
1993 10 CFR 50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin) 1.
Auxiliary Feedwater Flow Rate Input Error in NOTRUMP Computer Code 8PCT-189 3 F E.
LICENSING BASIS PCT + PERMANENT ASSESSMENTS PCT-2033 3'P Note:
As discussed in the body of this submission, prior LOCA model assessments have been absorbed in new analyses performed to support a
request for relaxation of the main steam safety valve (MSSV) setpoint tolerance.
The analyses were submitted for NRC review with our letter AEP:NRC:1169, dated November ll, 1992.
Since these analyses bound currently licensed operating conditions, the resulting changes are being reported in lieu of developing a rack up of evaluations for each issue that has been absorbed in the analyses.
Attachment 1 to AEP:NRC:1118G Page 2
SMALL BREAK LOCA PLANT NAME:
DONALD C.
COOK NUCLEAR PLANT UNIT 1 A.
ANALYSIS OF RECORD PCT 2122 P (Comments:
Evaluation Model:
NCTRUMP, FQT-2 32, FdM-1 55, SGTP 15X, Other'HSI Cross Tie Valve 0 en 3588 Mwt Reactor Power B.
10 CFR 50.92 SAFETY EVALUATIONS C.
PRIOR PERMANENT LOCA MODEL ASSESSMENTS 4PCT
-552 Ps 4PCT
-13 F D.
1993 10 CFR 50e46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin) 1.
Auxiliary Feedwater Flow Rate Input Error in NOTRUMP Computer Code 4PCT ~9 F
E.
LICENSING BASIS PCT + PERMANENT ASSESSMENTS PCT 1654oF Note:
As discussed in the body of this submission, prior LOCA model assessments have been absorbed in new analyses performed to support a
request for relaxation of the main steam safety valve (MSSV) setpoint tolerance, The analyses were submitted for NRC review with our letter AEP:NRC:1169, dated November 11, 1992.
Since these analyses bound currently licensed operating conditions, the resulting changes are being reported in lieu'f developing a rack up of evaluations for each issue that has been absorbed in the analyses.
ATTACHMENT 2 TO AEP'NRC'1118G REPLACEMENT PAGES FOR ATTACHMENT 3 TO AEP'NRC'1118F LETTER
Attachment 3 to AEP:NRC:1118F Page 1
SMALL BREAK LOCA PLANT NAME:
DONALD C.
COOK NUCLEAR PLANT UNIT 1 A.
'122 F
- gOTRUHP, FQT 2 32, FCR 1 55, 3250 Mwt Reactor Power ANALYSIS OF RECORD (Comments:
Evaluation Model:
SGTP 15',
'I Other'HSI Cross Tie Valve Closed B.
10 CFR 50.92 SAFETY EVALUATIONS C.
PRIOR PERMANENT LOCA MODEL ASSESSMENTS hPCT
-268~F~
RPCT 3 F D.
1993 10 CFR 50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin) 1.
Effect of SI in Broken Loop 2.
Effect of Improved Condensation Model 3.
Drift Flux Flow Regime Errors RPCT-150'F 5PCT 150oF RPCT
-13 F E.
LICENSING BASIS PCT + PERMANENT ASSESSMENTS hPCT 1844oF Note:
As discussed in the body of this submission, prior LOCA model assessments have been absorbed in new analyses performed to support a
request for relaxation of the main steam safety valve (MSSV) setpoint tolerance.
The analyses were submitted for NRC review with our letter AEP:NRC:1169, dated November 11, 1992.
Since these analyses bound currently licensed operating conditions, the resulting changes are being reported in lieu of developing a rack up of evaluations for each issue that has been absorbed in the analyses.
Attachment 3 to AEP:NRC:1118F Page 2
SMALL BREAK LOCA PLANT NAME:
DONALD C.
COOK NUCLEAR PLANT UNIT 1 A.
ANALYSIS OF RECORD PCT 2122 P (Comments:
Evaluation Model:
- NCTRCMP, PQT-2 32, FdH-1 55, SGTP 15X, Other'HSI Cross Tie Val~e 0 en 3588 Mwt Reacto Power B.
10 CFR 50.92 SAFETY EVALUATIONS C.
PRIOR PERMANENT LOCA MODEL ASSESSMENTS EPCT
-552 Fs RPCT 0 F D.
1993 10 CFR 50e46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin) l.
Effect of SI in Broken Loop 2.
Effect of Improved Condensation Model 3.
Drift Flux Flow Regime Errors EPCT 150 F RPCT-
-150'F 5PCT
-13 F E.
LICENSING BASIS PCT + PERMANENT ASSESSMENTS EPCT 1557 F Note:
As discussed in the body of this submission, prior LOCA model assessments have been absorbed in new analyses performed to support a
request for relaxation of the main steam safety valve (MSSV) setpoint tolerance.
The analyses were submitted for NRC review with our letter AEP:NRC.'1169, dated November ll, 1992.
Since these analyses bound currently licensed operation conditions, the resulting changes are being reported in lieu of developing a rack up of evaluations for each issue that has been absorbed in the analyses.
Attachment 3 to AEP:NRC:1118F Page 5
SMALL BREAK LOCA PLANT NAME:
DONALD C.
COOK NUCLEAR PLANT UNIT 2 A.
ANALYSIS OF RECORD PCT- ~135 'F (Comments:
Evaluation Model:
NOTRUMP, FQT-2 32, FdH 1 62, SGTP 15X, Other'HSI Cross Tie Valve 0 en 3588 Mwt Reactor Power B.
10 CFR 50.92 SAFETY EVALUATIONS dPCT
~+1 4 F~
C.
1993 10 CFR 50.46 MODEL ASSESSMENTS (Permanent Assessment of PCT Margin) 1.
Effect of SI in Broken Loop 2.
Effect of Improved Condensation Model 3.
Drift Flux Flow Regime Errors hPCT 150OF hPCT
-1500F APCT
-13~F D.
LICENSING BASIS PCT + PERMANENT ASSESSMENTS PCT 15180F Note:
As discussed in the body of this submission, prior LOCA model assessments have been absorbed in new analyses performed to support a
request for relaxation of the main steam safety valve (MSSV) setpoint tolerance.
The analyses were submitted for NRC review with our letter AEP:NRC:1169, dated November ll, 1992.
Since these analyses bound currently licensed operating conditions, the resulting changes-are being reported in lieu of developing a rack up of evaluations for each issue that has been absorbed in the analyses.