ML20235G611
| ML20235G611 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/30/1982 |
| From: | Bandon R, Gridley R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20235G546 | List: |
| References | |
| NEDO-22226, NUDOCS 8709300122 | |
| Download: ML20235G611 (16) | |
Text
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Attachment 5
) NEDO-22226 DRF L12-00553 82NED098
- Class I September 1982 l
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FITZFATRICK NUCLEAR POWER PLANT CYCLE 5 l ANALYSIS FOR OPERATION WITH ONE SAFETY / RELIEF VALVE OUT OF SERVICE F !'
i, Approved: ,
Mg- Y!]{ M Approved:__R. I/. Gridley/, Ma'n'ager
~
. .prandon, Manager Nuclear Services Engineering Fuel and Se dices Licensing 8709300122 870925 gDR ADOCK 05000333 PDR tv NUCLEAR POWER SYSTEMS DIVISION
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL $ ELECTRIC
Attachment 5 NEDO-22226 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric Company (GE) solely for The Power Authority of the State of New York (PASNY) for PASNY's use with the U.S.
Nuclear Regulatory Commission (USNRC) for amending PASNY's operating' license of.the James A. FitzPatrick Nuclear Power Plant. The.information contained in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to GE at the time this report was prepared.
The only undertakings of GE respecting information in this document tre con-tained in the contract between PASNY and GE as per GE Proposal No. 424-TY7 EP1, dated August 27, 1982. The use of this information except as defined y 7 said contract, or for any purpose other than that for which it is intended, is j not authorized; and with respect to any such unauthorized use, neither GE nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
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. Attachment.5 NEDo-22226 CONTENTS
.Page
' l. INTRODUCTION 1
- 2. LOSS-OF-COOLANT ACCIDENT 1 2.1 Gene ~ral Discussion 1 2.2 ' Plant Specific Analysis 2 2.~3 Conclusions- 2
- 3. TRANSIENTS .
2 3.1 General Discussion 2 3.2 Limiting Transient 3 3.3 MSIV Flux Scram 3-
'3.4 Conclusion 4 4 '. OVERALL CONCLUSION 4 5 . .. REFERENCES 9
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Attachment 5 NED0-22226 Tables Table Title Page 1 FitzPatrick - Loss-of-Coolant Accident Analysis One ADS Valve Out-of-Service 5 2 FitzPatrick Cycle 5 S/R Valve Setpoint and Grouping 6 3 FitzPatrick Cycle 5 Core Wide Transient Analysis Results 7 4 FitzPatrick Cycle 5 Overpressurization Analysis
. Summary 8 ILLUSTRATIONS Figure Title Page 1 Vater Level Inside the Shroud and Aeactor Vessel Pressure Following a Small Recirculation Line Break. HPCI Failure, One ADS Valve Out-of-Service, Break Area = 0.0? ft2 9.
2 -Peak Cladding Temperature Following a Small Recirculation Line Break, HPCI Failure, One ADS Valve Out-of-Service, Break Area = 0.07 ft2 to 3 Plant Response to Generator Load Rejection Without Bypass 11 4 Plant Response to MSIV Closure (Flux Scram) 12 l
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Attachment 5 l
- 1. INTRODUCTION The purpose of this report is to provide the technical bases for opera-tion of the FitzPatrick Nuclear Power Plant with one Safety / Relief Valve (S/RV) out-of-service for the current cycle (Cycle 5).
The potential ef fect of one S/RV out-of-service is to change the pressure response of the reactor during transients and postulated accidents which could conceivably impact the margins and/or safety limits for plant operation.
The accident and transient considerations for operation during Cycle 5 with one S/RV out-of-service are presented.
- 2. LOSS-CF-COOLANT ACCIDENT i
2.1 GENERAL DISCUSSION FitzPatrick has 11 combination S/RVs. All of these valves have a pressure-actuated safety function and seven of them have an additional solenoid-air operated relief function automatically actuated by the Automatic Depressurization Systee (ADS).
For Loss of Coolant Accident (LOCA) evaluations, credit is taken for the safety function; however, not all of the S/RVs are actuated during a LOCA.
If the out-of-service valve is one of the S/RVs with the ADS function there can be a potential impact on the calculated Peak Cladding Temperatures (PCT) for small break sizes of less than approximately 0.2 ft2 This may occur because, with a worst case postulated single failure of the High Pres-sure Coolant Injection System (HPCI), the small break transient is dominated by the time required to depressurize the reactor to the operating pressure of the low pressure Emergency Core Cooling System (ECCS).
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Attachment 5 NEDO-22226 For larger postulated break sizes, the LOCA itself depressurizes the
= reactor. vessel rapidly before the ADS actuates, and the number of S/RVs or
~
Jactuation of the ADS is inconsequential to'the calculated PCT. ,
For the FitzPatrick plant, the limiting LOCA is a large break (greater than 1.0 ft2 in size), and the loss'of the S/RV or ADS function hAs no effect on the calculated Maximum Average Planar Linear ~ Heat Generation Rate (MAPLHGR)
. . limit.
2.2 PLANT SPECIFIC ANALYSIS A plant specific analysis was performed.for FitzPatrick with the' approved Appendix K . evaluation models, SAFE and REFLOOD, to determine the increase in the calculated small break PCT and to show that the small break does not bep come limiting'with one S/RV with the ADS function out'of service. ,-
The results of the analyses are shown in Table 1. The most limiting small break in terms'of. PCT is the 0.07 ft' recirculation suction line break.
With one ADS out-of-service, the PCT is less than 1300*F, which is over 900*F below the'2200*F limit. The water level, pressure, and PCT for the worst case small break are shoun in Figures 1 and 2.
2.3 CONCLUSION
S With one S/RV out-of-service, there is no impact on the calculated MAPLHGR limits for FitzPatrick, even if the out-of-service S/RV has .the ADS function.
- 3. TRANSIENTS 3.1 GENERAL DISCUSSION The following limiting abnormal operational transients during Cycle 5 opera-tion with one S/RV out-of-service were performed using the ODYN computer code:
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- a. Load rejection without bypass at 104% power /100% flow.
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y Attachment 5 NED0-22226'
- b. The ASME code overpressure protection event - Main Steam Isolation Valve (MSIV) closure, flux scram.
The S/RV setpoints and groupings for the analysis of these events are presented in Table 2.
Note that the S/RV in the lowest setpoint group is conservatively assumed to be out-of-service in this analysis. Therefore, any particular S/RV which becomes inoperable during Cycle 5 is bounded by this
, analysis.
3.2 LIMITING TRANSIENT The ACPR values for Cycle 5 operation are reported in Reference 1. In terms of 6CPR value, the limiting transient is the Load Rejection Without Byr pass (LRNB) event. -
The LRNB transient was performed with ont S/RV out-of-service. The peak i
neutron flux, the peak heat flux, and the Minimum Critical Powet Ratio (MCPR) remain unchanged for this event. This is because both the peak neutron flux and the peak heat flux occur bet' ore the S/RVs are actuated during this event.
Therefore, the effect of one S/RV out-of-service has no impact on the 6CPR of the limiting transient LRNB event.
The output parameter data for this event is summarized in Table 3 and the time response of this transient is shown in Figure 3. i The non-pressurization events (e.g., rod withdrawal error) are independent of valve setpoint and valve capacity; therefore, the 6CPR values are unchanged as a result of one S/RV out-of-service.
3.3 MSIV FLUX SCRAM The adequacy of the S/RV capacity based on ASME code requirements is demonstrated by the MSIV closure transient with high flux scram. The peak vessel pressure for this event increases by 15 psi as a result of one S/RV out- .
of-service, resulting in a peak pressure of 1290 psig. This shows a margin of i l
85 psi to the ASME code upset limit of 1375 psig. Therefore, the ASME limit l
is still met with one S/RV out-of-service.
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Attachment 5 NEDO-22226 The output parameter data for the MSIV flux scram transient considered in this analysis is summarized in Table 4. The time response of key variables for this transient is shown in Figure 4.
3.4 CONCLUSION
- With one S/RV out-of-service, there is no impact on the MCPR operating limit for FitzPatrick, Cycle 5 operation. Also, the peak vessel pressure for the MSIV flux scram event is well within the ASME code upset limit of 1375'psig.
- 4. OVERALL CONCLUSION The operation of the FitzPatrick Nuclear Plant at full power for the remainder of Cycle 5 with any one S/RV out of service has no impact on op rating limits. '
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- 5. REFERENCES 1.
Supplemental Reload Licensing Submittal for FitzPatrick Nuclear Power Station 4 Y1003J01A25.
}
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Attachment-5 NEDO-22226 Table'l FITZPATRICK - LOSS-OF-COOLANT ACCIDENT ANALYSIS q
.ONE ADS VALVE OUT-OF-SERVICE Recirculation Line Break Upstream of Discharge Valve.
System Failed: HPCI Systems Remaining: '2 LPCS + 2 LPCI + 6 ADS Valves" Reflooding Peak Cladding Break Size 'Uncovery Time Time Temperature
-(ft 2) (Sec) (sec) (*F) 0.05 324.6 412.4' 1103
-0.07 265.6 372.4 1271
-0.10 232.1 330.5 1241 i
"Two of the LPCI systems ' inject into the broken loop and it is conservatively assumed that all of the inje.cted water is lost through the break.
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NED0-22226 Table 2 FITZPATRICK CYCLE 5 S/R Valve Setpoint and Grouping Previous S/RV Grouping New (one S/RV out of service) Grouping Setpoint S/RV Setpoint S/RV No. No.
psig Target Rock psig Target Rock 1090 + 1% 2 1090 + 1% 1 1105 + 1% 2 1105 + 1% 2 1140 + 1% 7 1140 + 1% 7 1
l 6
L-_------_---------
Attachment:
-5 NEDO-22226 Table 3 FITZPATRICK CYCLE 5 CORE-WIDE TRANSIENT ANALYSIS RESULTS Transient ' Load Rejection No Bypass
-Power 104%
Core Flow 100%
Peak Neutron Flux (%NBR) 653 Peak Heat Flux (%) 125 ACPR (nominal) 0.22 (8x8 + 8x8R) 0.24 (P8x8R)
Plant Response Figure 3 i
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. Attachment 5:
. Table 4 FITZPATRICK CYCLE 5-OVERPRESSURIZATION ANALYSIS
SUMMARY
Transient MSIV Closure (Flux Scram)
' Power 104%
Core Flow 100%
P,1- 1255 psig P 1290 psig Plant Response Figure 4 4
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