ML20063A992

Design Verification Test Summary,Model 7567F

ML20063A992
Person / Time
Site:	FitzPatrick
Issue date:	12/02/1976
From:	TARGET ROCK CORP.
To:
Shared Package
ML20063A987	List: ML20063A985 ML20063A989 ML20063A992
References
1834, NUDOCS 8208250145
Download: ML20063A992 (21)

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• COPY NO - . REPORT NO- 183t.

,i PROJECT NO. 75EE I .

DATE 12/2/76 i

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f) t MODEL 7567F DESIGN VERIFICATION TEST StW. MARY

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LIST OF EFFECTIVE DATE BY APR REVISIONS OF PAGES LETTER DE S C RI P TI ON F G.

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1.0 SCOPE 1.1 This report delineates the results of the preproduction steam testing performed on the Target Rock Model 7567F safety S/N 1 production valve. These tests were performed to verify that minor design changes between the production and the prototype valve did not adversely affect performance.

2.0 VALVE CHARACTERISTICS TESTED a) Valve set point repeatability b) Valve response time c) Valve leakage in " clean" sys tem d) Valve operation in " dirty" system e) Affect of leakage on set point and response time.

3.0 VALVE PERFORMANCE CALIBRATION RUN 3.1 This test was perfor=ed to evaluate the valve's perfor=ance in

.a " clean" sys tem. In addition, this test was used as a basis for evaluating the perfor=ance of the valve in a " dirty" system (see Paragraph 4.0) and in a high leakage configuration (see Paragraph 5.0) .

3.2 Tes t Results The valve was actuated in the automatic mode twenty (20) times.

The following valve characteristics were measured:

a) Average valve lif t point - 1105.5 psig

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b) Valve lift point range - 1103 to 1109 psl~g l

c) Valve leakage af ter six (6) cycles - no detectable leakage l

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d) Valve leakage af ter twenty (20) cycles - droplets on mirror

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e) Valve delay time - average 153 ms f) Valve delay time range - 146-162 ms.

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3.3 Conclusion Valve perfor=ance acceptable.

4.0 VALVE PERFORMANCE WITH PIPING C0h7 AMINATION 4.1 This test was performed to detemine the affect of large amounts of contamination (pipe crud) on valve performance.

4.2 Concaminaeion Level Approximately one (1) cubic inch of pipe crud (60-300 micron) was introduced into the pilot section of the valve. (NOTE: Partial valve disasse=bly required to introduce the contaminant.)

4.3 Test Results The valve was automatically actuated ten (10) times. The follow-ing valve characteristics were measured:

'.. a) Average valve lif t point - 1109.6 psig ,

b) Valve lift point range - 1101 to 1116 psig c) Valve delay time average - 162 ms d) Valve delay time range - 148 to 170 ms.

4.4 Conclusion Valve performance not affected.

5.0 VALVE PERFORMANCE WITH VARIOUS AMOUNTS OF INDUCED LEAKAGF 5.1 This test was performed to determine the affect of large amounts of leakage on valve performance.

5.2 Leakage Levels Tested

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The following amounts of Icakage were introduced by scorihg and machining leakage paths into the pilot disc. (Valve disassembly was required to remachine the pilot disc to achieve the various  ;

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Configuration Induced Leakage at 1000 psig A 2.6 #/hr l

B 17.0 #/hr C 27.3 #/hr D 198.2 #/hr 5.3 Test Results The valve was actuated auta=atically five (5) times for each can-figuration.

The following valve characteristics were measured:

a) Average valve lift point:

Configuration Average Lif t Pressure A 1108 psig B 1109 psig C 1100 psig 1

' D 1104 psig b) Valve lif t point range:

Configuration Lif t Pressure Range A 1104 to 1111 psig B 1107 to 1115 psig .

C 1091 to 1106 psig

. D 1101 to 1106 psig

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c) Valve delay time average:

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Configuration Average Delay A 168 ms

' - B 273 'ms i

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d) Valve delay time range:

Configuration Delay Range

• A 109 to 145 ms B 210 to 440 ms C 220 to 680 ms D 300 to 520 ms

• Higher values are result of condensation. Improved drainaga should be investigated.

5.4 conclusion Leakage rates up to 200 #/hr do not adversely affect set point repeatability or response time.

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• .,'b4' a y h.;! b~ INSTALLATION AND f;v. t. a ;5hCGPL g SERVICE ENGINEERING I GENERAL ELECTRIC COMPANY , . . . .3532 JAMES STREET, P. O. BOX 4841 DIVISION SYRACUSE. NEW YORK 13221 G-EP1-2-088 August 11, 1982 f

Mr. R. A. Burns Vice President - BWR Support Power Authority of the State of New York 123 Main Street White Plains, NY, 10601

SUBJECT:

CONTINUED PLANT OPERATION WITH ONE SAFETY RELIEF VAI.VE OUT OF SERVICE

- TRANSMITTAL OF ENGINEERING REPORT

Dear Mr. Burns:

At the request of Power Authority of the State of New York, General Electric-Company has perfomed an engineering analysis to justify continued operation of James A. FitzPat'ick r Nuclear Power Plant with one safety relief valve out of service for the current cycle (cycle 5). The engineering report on the anal-ysis is attached. It is concluded from the analysis that the operation of the FitzPatrick Nuclear Power Plant at-full power for the remainder of cycle 5 with any one safety relief valve out of service has no impact on safety limits. A licensing report will follow in approximately four weeks.

We expect that your purchase order for this work will be forwarded at your ear-liest convenience.

Very truly yours,

@) WL- d.-

Morris-Service Supervisor - Nuclear JMM:NMV: dmo attachment-cc: J. Gray C..A. McNeill, Jr.

T. Landers L. Guagull

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• ' FITZPATRICK SRV OUT OF SERVICE ANALYSIS

• I INTRODUCTION ,

The purpose of this report is to justify continued operation of the FitzPatrick Nuclear Power plant with one Safety Relief Valve (SRV) out of service for the current cycle:(cycle 5).

The potential effect of one relief valve out of service is to change the pressure i- . response of the reactor during trans ients and postulated accidents which could

conceivably impact the margins and/or safety-limits for plant operation.

LOSS OF COOLANT ACCIDENT (LOCA)

General Discussion Having one satety or relief valve out of service has no effect on LOCA analysis since no credit for the power actuated relief function is taken in the basis thermal hydraulic analysis.

For LOCA evaluations credit is taken for the safety relief function, however not all

• of the sarety valves are actuated during LOCA. Furthermore, the appendix K LOCA analyses are insensitive to changes in safety valve setpoints, d

If, however, the valve out of service is one of the SRVs with the Automatic i

Dapressurization (ADS) function there can be a potential impact on calculated Peak Cladding Temperatures (PCT) for small break sizes of less ,than approximately 0.2 f t2 This may occur because, with a worst case postulated single f ailure of .the High Pressure Coolant Inj ection System (HPCI), the small break transient is dominated by the time required to depressurize the reactor to the operating pressure of the low

, pressure ECCS.

For larger postulated breaks sizes, the LOCA itself depressurizes the reactor vessel rapidly before the ADS actuates and the number of SRVs or actuation 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 i size) and the loss of SRV or ADS function has no effect on the calculated MAPLHGR limit.

Gsneral Electric has performed sensitivity studies that show that the calculated small break PCT can increase approximately 100*F for a ten percent reduction in overall ADS j capacity (e.g., the loss of one out of ten valves) .for a typical BWR 4 plant similar to FitzPatrick. However, these breaks are much less limiting than the DBA have then is no resulting impaact on MAPLHGR.

i Plant Specific Analysis l A plant specific analysis was performed for FitzPatrick with the approved Appendix K ~

j evaluation modes SAFE and REFLOOD to determine the increase in calculated small break PCr and to show that the small break does not become limiting with one SRV with ADS

'fanction out of service.

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The results of ths enalyses are shown in Table 1. The sna11 break (loss then 0.2 f t2) witn the highest PCT is the 0.07 f t2 break. With ona ADS out of service ths PCT is Icss than 1300*F which is over 900*F below the 2200*F limit. Figure 1 and 2 show the 4

ccter level, pressure, and PCT for the wrost case small break.

Conclusion With one SRV out of service, there is no impact on the calculated MAPLHGR limits for FitzPatrick, even if the SRV out of service has the ADS function.

TRANSIENTS General Discussion The following limiting abnormal operational transients during cycle 5 operation with cne S/RV out of service were examined:

1. Load rejection without bypass at 104% power /100% flow.

2. Feedwater controller failure, maximum demand at 104% power /100% flow.

The ASME code overpressure protection event - MSIV closure, flux scram - was also reevaluated.

The S/RV setpoints and groupings for analysis of these events are presented in Table

2. Note the S/RV in the lowest setpoint group is conservatively assumed to be out of ssrvice in this analysis. Therefore, any particular S/RV which becomoes inoperable during cycle 5 is bounded by this anlaysis.

Limitina Transients The MCPR values for cycle 5 operation are reported in Reference 1. Depending on the fuel type and exposure range, the limiting transients in terms of MCPR valve are eitner Load Rej ection Without Bypass (LRNB) or Rod Withdrawal Error (RWE) . The next limiting event is Feedwater Controller, Maximum Demand (FMCF) event.

As shown in Figure 3, and 4 (Taken from Reference 1), both the peak neutron flux and paak heat flux occur before the S/RV are actuated to open during LRNB and FWCF events. Therefore, the effect of one S/RV out of service has no significant impact on the transient change in critici power ratio (ACPR) and the operating limit minimum critical power rat!.o for these transients since the most limiting conditions occur prior to any S/RV opening.

The RWE transient is independent of valve setpoint and valve capacity, therefore the MCPR operating limit is unchanged as a result of one S/RV valve out of service.

MSIV Flux Scram The adequacy of the safety valve capacity based on ASME code requirements is dtmonstrated by the MSIV closure transient with high flux scram, which shows a margin l

of 85 psi to ASME code upset limit of 1375 psig. The peak vessel pressure for this event increase by 15.psig as a result of one S/RV valve out of service.

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. is The output parameter data for MSIV flux scram transientThe time response of key varia summarized in Table 3.The analysis was performed using '0DYN' code.

shown in Figure 5.

Conslusion thermal margin such that there is no With one S/RV out of service there is suf ficient impact on MCPR for Fitzpatrick, cycle 5 operation.

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, . OVERAII CONCLUSION ,

' The operation of the FitzPatrick Nuclear Plant at full power for the remainder of Cycle 5 'with any one SRV out of service has no impact on operating limits.

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REFERENCES

1. Supplemental Reload Licensing Submittal for FitzPatrick Nuclear Power Station 4, Y1003J01A25.

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. TABLE 1 FITZPATRICK - LOSS OF COOLANT ACCIDENT ANALYSIS 1 - ADS VALVE OUT OF SERVICE Recirculation Line Break Upstream of Discharge Valve.

System Failed: HPCI Systems Remaining: 2 LPCS + 2 LPCI + 6 ADS Valves Reflood S.B.M.

Break Uncovery Reflooding Peak Cladding Size Time Time Temperature (ft*) (sec) (sec) (*F) 0.05 324.6 412.4 1103 0.07 265.6 372.4 1271 0.10 232.1 330.5 1241 2 of the LPCI systems inj ect into the broken loop and it is conservatively assumed that all the inj ected water is lost through the break.

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TABLE 2 ,

FITZPATRICK CYCLE 5 S/R Valve Setpkint and Grouping Previoni SRV Grouning New (one S/RV out of service) Groupina Setnoint S/RV Setnoint S/RV No/ No/

d Tarnet Rock 2sig Tarnet Rock 1090+1% 2 1090+1% 1 .

1105+1% 2 1105+1% 2 1140+1% 7 1140+1% 7 I

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. . TABLE 3 FIT?PATRICE CYCLE 5 OVERPRESSURIZATION ANALYSIS

SUMMARY

4 Transient MSIV C1soure (Flux Scram)

Pcwer 104%

Core Flow 100%

j P 1255 psig d ,

P 1290 psis y

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1 VESSEL PRES RISE (PSI) f F I wuina nur 2 S/RV .-?DW, CROUP 3*

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• • Croup 1 S/RV Set Point = 1090 psig +1% 7 Croup 2 S/RV Set Point = 1105 psig +1% .

Figure 3a. Plant Response to Cenerator Load Rejection Without Bypass, EOC 5 (Ref. 1)

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9 Figure 5 Plant Response of MSIV (Flux Scram) .

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