ML20138G479
| ML20138G479 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/07/1985 |
| From: | Mussel W Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20138G467 | List: |
| References | |
| PSE-SE-Z-018, PSE-SE-Z-18, NUDOCS 8510250484 | |
| Download: ML20138G479 (3) | |
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-2-018 TITLE: MSIV FULL ISOLATION, TEST 23B - EXPANSION OF TEST WINDOW Date: OCT 7 1985 1.0 PUR POS E The purpose of this Safety Evaluation is to document the results of the evaluation of the proposal to expand the allowable MSIV full isolation test window to between 70%
and 100% power should an inadvertent MSIV full isolation in this power range occur during the power ascension test program.
2.0 SCOPE The scope of this test is the transient behavior of the plant and associated systems during and following main steam isolation.
3.0 REFERENCES
- 1. Regulatory Guide 1.68, Revision 2, August 1978
- 2. Hope Creek Final Safety Analysis Report (FSAR)
Chapter 14
- 3. General Electric Startup Test Specification, 23A4137, Revision 0 4.
Power Ascension Test Procedure TE-SU.AB-252(O) MSIV ,
Full Isolation Test 4.0 DISCUSSION Paragraph S.m.m of Reference 1 specifies requirements for demonstration of the dynamic response of the plant for a MSIV full closure transient at 100% power. Power Ascension Test 23B, MSIV Full Isolation, is scheduled to be performed at Test Condition 6 to satisfy the regulatory requirements.
However, if an inadvertent full isolation occurs between 70 and 100% power, the inadvertent transient results will be substituted for the Test Condition 6 Test.
8510250484 851017 PDR ADOCK 05000354 A PDR PSE-SE-2-018 1 of 3
4, s Response of the reactor plant and its associated systems during and following an automatic closure of all main steam isolation valves (MSIVs) is determined by comparing the results of Test 23B to the acceptance criteria in Reference 3. The Level I criteria require that reactor dome pressure and simulated heat flux not exceed the predicted values by more than a given amount. As these predicted values are referenced to actual test conditions of initial power and dome pressure, expanding the test window to between 70% and 100% power will not af fect the adequacy of the test.
Other Level 1 criteria require that MSIV closure times meet timing specifications, that the reactor must scram, and that the feedwater control system settings must prevent flooding of the main steam lines. The reactor scram and the MSIV closure are not affected by modifying the window and the feedwater control settings will be analyzed to predict their response at 100% power based on their performance at lower power.
The Level 2 criteria are predicted based on actual test conditions and are not adversely affected by expansion of the test window. The automatic initiation of HPCI and RCIC and the recirculation pump trip at L2 will be verified if L2 is reached.
This approach was- used successfully by Grand Gulf when an inadvertent scram occurred at approximately 75% power.
The NRC accepted the analyzed results providet .n lieu of actual performance of the full isolation test ut. 100%
power.
5.0 CONCLUSION
The full MSIV isolation test is simplified if an inadvertent scram occurs at >70% power but <100% power.
If all the necessary information is obtained (i.e., MSIV closure times, peak dome pressure, etc.) analysis can predict what plant response would be at 100% power and no additional test would need to be perforned. No Technical
' Specification changes are required and the safe operation of the plant and associated systems is not affected.
Based on the above no unreviewed safety question exists.
6.0 DOCUMENTS GENERATED None I
PSE-SE-2-018 2 of 3
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7.0 RECOMMENDATIONS Revisions to Hope Creek's FSAR and startup test procedures shall be made as necessary to reflect the expansion of the test window for MSIV full isolation testing to 70% to 100% power.
8.0 ATTACHMENTS None 9.0 SIGNATURES Originator '
_ s .NL l' O' Verifier '
7i u /i/y 9(
Group Head SSE) bt dw bd f'. E , IPhf9 '
Systems Analysis Group Head O ,ld . L 7. 6. 10/4 5 Site Engineering Manager ,[]) huvf w k ( (5' l Ddt'e PSE-SE-Z-018 3 of 3
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l W)g3 Womahmed MISSISSIPPI P8WER & LIGHT COMPANY Helping Mulld Mic els elppi P. C. 5 0X 194 0, J A C K 6 ON. MIS SIS SIP PI 30210-1540 April 23, 1985 . 3.ms.. ucamo a saerty us ant so U. $. Nuclear Regulatory Commission . Office of Nuclear Reactor Regulatten Washingten. D. C. 20$$5 Attentions Mr. Harold K. Denton. Director Daar Mr. Denton: SU5 JECT: Crand Culf Nuclear Station Unit 1 Docket No. 50-416 Licenas No. NPP-29 File: 0290/L-360.0 Proposed Medification to FAAR Startup Test Program: Operating License Condition 2.C.(31) AECH-85/0127 . In accordance with OGNS Operating License condition 2.C.(31), Mississippi Power & Light (MPSL) hereby requests a modification to the PSAR Chapter 14 Startup Test Progrs= for your review and approval. The proposed modification involves certain test conditions and acceptance criteria for Startup Test Number 253-Tu11 Reactor teolation (PSAR Section 14.2.12.3.22.2). These preposed changes will allow MP&L to take credit for recoct test data to east the startup testing objectives for this test. MP&L classifies the requested modification as a major change to the test ptsgrat: due to the different power level than that currently prescribed in 75AR Chapter 14 and the associated acceptance criteria Changes. MP&L expeets to enter Test Ceadition 6 in mid-May 1985. In order to provide sufficient time to incorporate the proposed changes into the OGNS startup test procedures and to acces:nedate revised testing schedules, your review and approval is respectfully requested by May 13, 1983. The bases and justification for this request is contained in the attachments to this letter. This request has been reviewed and approved by OGNS Plant Safety Review Cons:ittes. In accordance with the provietens of 10CPR170, an application fee of $150 is attached. Flesse advise, if you require additional inforsation. Yours truly, W Wrwl L. F. Dale Director - e8 , 8AB/JGCivos Attechsents cc (See Next tage) p.s e o m g g r)[( h ) U' l J*"I uv s++ k'
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J0P13A!CM85041702 - 1
c .o u i t. n tois :+;- .c. ..: m . : , ;,1 ,, . Attachzent 1 h) Freposed Hodification to GGNS Btartup Test Fregran e Test: Test No. 2SB - Full Reactor teolation . 7 SAP section: 14.2.12.3.22.2 Requiremcot GONS Operating License Condition 2.C.(31) requires prior NRC permission for major modifications to the startup test program. Introduction Certain changes to the Grand Culf Nuclear station startup test prograr, are proposed here to take credit for a full closure of the M51Vs at approxirastely 751 reactor power. These proposed changes will eliminate what MP61 considers to be redundant testing (i.e. two MS!V closure events at 75% and 100% reactor power respective 1v) while still meeting the startup testing objectives of verifying proper plant design and response. Additional benefits and considerations include (1) challenges to plant safety systems will be reduced by accepting the 751 MSIV closure data; (2) a demanding and thorough startup test program will not be further extended while still fully verifying the plant's ability to respond to operational transiental and (3) pionned sustained full power operation can ecuenes in accordance with schedules which satisfy HP&l, corporate ob,i e ctive s . These changes are proposed with careful consideration of plant safety and fully estisfying the test progra't's objectives. The proposed changes are considered in the public interest by avoiding additional delays in p!scing the Benerating station into public service while maintaining safe plant operatione and meeting the test program objectives. l Basis For FSAR Paragraph 14.2.12.3.22.2(c) definas the test procedures as Proposed ". . . simultaneous full closure of all main steam isolation Medifiestion: valves (Ms:vs) will be performed at 95 to 100 percent of rated i thermal power." l
/a present. the CCliS startup test program includes the !
prformance of two startup tests which result in a plant trip from 100% pever. These tests are startup Test #255, Full l Reactor f aciation (MSIV closure test) and 8tersup Test #27, Turbine Trip and Cenerator toad Rejection Test. NFat, is
~
proposing that the 1002 power full Resster feelation (#258) be modified to allow performance of the test at 70 to 752 of rated thernal power. This modification is considered to be a es,ior , modification to the startup test progres as defined in the above i referenced license condition. This proposal involves the inedification of acceptance criteria and the performance of the Full Reactor Isolation Test at a power level different frate ## that currently prescribed in FSAR Chapter 14. J0P13 MISC 85041701 - 1 l l - _ -
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On April 7. 1985 Grand Gulf Unit 1 scrasaned from approximately 75% power due to an MSIV closure trip caused by a spurious isolation signal in conjunction with surveillance testing. The j transient data was recorded and analysed. The results indicate . ! that the transient le within the predicded analytical results l based on beginning of cycle plant conditions for MSIV closure
- l from 75% power. A key observation from the data is that the i staan shutoff characteristics of the MSIVs are less severe them
- originally predicted. These demonstrated results have been i analytically extrapolated to the closure event fro = 1002 power and are also bounded by the origins 11y predicted results for thfe event (See Figures 1 through 6 contained in Attachment 3).
Based en the analysis of available infor=ation, performance of the plant and all protection systems is expected to be essentially the sane at 100% power as they hava already been observad during the 7SI event. As further assuracco of acceptable plant response to this pressurization transient, the full power Turbine Trip and Generator Lead Rejection Test (d27) causes a r. ore rapid shutoff of sein steam then the MSIV closure event. CGN3 73AR Figures j 15.2-2 " Generator Lead Rejection, Trip Scrat, typasa-On" and 15.2-6 "Three-second closure of All Hain Stest: tine Isolation Valve with Position Switch Scram Trip" depicts this conclusion. Thess two pressurization events result in similar system response and behavior with the notable exception of the more rapid response for the generator lead rejection event. All FSAR Level 1 acceptance criteria fer the MSIV closure test are included in the Level 1 acceptance criteria for the Turbin Trip and Generator Lead Rejection Test (f27). However, each of these criteria were met with sufficient margin during the actual 73% event; no significantly greater challenge is predicted at full power. The Level 2 acceptance criteria for the Fuli 1 Reactor Isolation (except for the group 2 $RV sctuation and the low water level automatic actions) were demonstrated by the 731 i event. The group 2 3RV actuation will be demonstrated during the generator load rejection. The criteria related to RPV low water level (level 2) NPCS actuation and ATVS Recirculation Pump Trip have been adequately deeenstrated and documented during previous plant transient events. Although OGNS has not experienced an on-line RCIC level 2 wuto start, the RCIC system has been the subject of extensive startup and surveillance testing. Therefore, MPM. taintains an extremely high confidence in the systez's ability to properly actuate and to achieve design flow
. rate.
The slower coastdown than predicted for the feedvetor systes . provided good water level control during the ?St event. Long-ter= maintenance of water level was easily demonstrated during the 75% event by utilicing manual initiatien-of RCIC. No significant difference is expected for a full power event. ,s I J0?13 MISC 65041701 2 l _ _
-- n u.u m .s n 1c, :,: m :s s a c.1, Attachmant 1
. 1093 Regarding M5!V operabilityr observance of MS!V operation has been previously demonstrated by tests already performed and will be further demonstrated upon completion of Test No. 25A.
H$IV Funccional Tests. Additionally. MSIV closure times are not expected to be significantly different at 75% power than at 1002 power. , i Vibration measuresents on the sain stos: piping were not cenducted during the MS V closure event. However. Performing main steam piping vibration testing concurrently with Test' No. 27 has been shevn to be an acceptable alternative, since the generator load rejection from 100% power will represent a verse case plant transient for measuring vibration of the main steam line piping. In su= ary: a) The MS!v closure event which occurred on April 7. 1985 was well within the predicted analytical results. The steam ehutoff characteristics were less severe and feedwater supply was acre favorable than originally predicted. b) F.xtrapolatten of the April 7 data to 100% power is well within the originally predicted results. e) The Turbine Trip and Cenerator Load Rejection Test. (' scheduled for TC6, will provide further verification of plant systems operation to rett full design capability and to show that analytical models conservatively predict actual plant conditions during and following transient events, d) Running the MSIV closure test at 1002 power would provide minimal additional significant information since all of the objastives of the test, with the exception of the group 2 3RV actuation and those objectives associated with i l reaching RPV low water level were achieved during the 751 HSIV closure evente as previously discussed. (e) The acceptance of the 75% test data avoids an additional plant transient and the subsequent challenge to various plant saft ty " systems. Proposed FSAR Attachment 2 provides the proposed FIAT. changes modifying Test Changest No. 25B to redefine the test procedura as " . . . simultaneous full closure of all M5Iva v111 be performed at 70 to 75 percent 1 of rated thermal power" in FSAR Paragraph 14.2.12.3.22.2(c). , Paragraph 14.2.12.3.22.2(d). Acceptance Criteria, tevel 2. item
- 1. will be sodified so that the Initial conditions for I power will be 70% to 753 and Dome Pressure will be 1000 to 1010 pela with the resulting sceeptance criteria changed to en increase,in J0P13 MISC 85041701 - 3
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~ 0 3 Heat Flux of 01 and an increase in Dese Fressure of 125 psi.
The proposed Level 2 acceptance criteria is based on the original prediction for executing the MSIV closure event at approximately 75% power (Sea Attachment 3 for additional discussion of the various enelysts conducted). The existing Acceptance Criteria, tevel 2, iter. 2 will be eliminated and . Level 2. iten 3 will be modified to read " Recirculation runback shall occur " t
- FSAR table 14.2-3 will also be revised to perform Test No. 25 (Full Reactor Isolation) at TC3 instead of TC6. The testing for Test No. 33 (which is presently required by note 13 te be cenducted in conjunction with Test No. 25) will be performed concurrently with Test No. 27. Turbine Trip and Generator Lead Rejection, Fending NRC approval of these proposed modifications to the test program. FSAR changes will be distributed to all MP&L 78AR senual holders as an interim change. The subject FSAR ehanges will be included in the FSAR Update later this year.
FSAR Q&Rs, 423.15, 423.33 and 423.43 discuss the performance of the full reactor (MSIV) isolation test at 100% power during Tc6. If the NRC concurs with this position, the responses conceined in these Q&Rs are essentially supereeded for GCNS Unit 1. en , JOP13 MISC 85041701 - 4 l
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Attachment 2 l , o h3 t I I l I l 1 RIVISTD GGN5 FSAR PACES PROPOSED FOR STARTL*F TEST NtHBER 253 - RM REACTOR ISOLATION I
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. F8AR Level 2 _. .
During full closure of individual valves, peak vessel' pressure must be 10 psi (0.7 kg/cm2) below scram, peak ' neutron flux must be 7.5 percent below scram, and steam 34 flew in individual lines must be 10 percent below the isolation trip setting. The peak heat flux must be S l percent less than its trip point. 14.2.12.3.22.2 Test Nu.mber 25B - Full Reactor Isolation
- a. Test objective ,
The purpose of this test is to determine the reactor transient behavior that results from the simultaneous full closure of all M51Vs.
- b. Prereeuisites The preoperational tests have been completed; the FSRC has reviewed and approved the test procedures and initiation of testing. Instrur.entation has been checked or calibrated as appropriate.
- c. Test Procedure
- 1 78 75 l A test of the simultan@us fvil closure of all MsIVs W/#g will be performed at M to M+ percent of rated thermal lU '
power. Correct performance of the RcIc and relief w valves will be shown. Esactor process variables will be monitored to determine the transient behavior of the system during and following the main steam i
. line isolation.
l
- d. M f._ance Criteria Level 1
- 1. The positive change in vessel dome pressure occurring within 30 seconds after closure of all MSIV valves must not exceed the Level 2 criteria by more than 25 psi. The positive change in simulated heat flux shall not exceed the Laval 2 criteria by more than 2 i
percent of rated value.
- 2. Feedwater control system tettings must prevent .
flooding of the steam lines. - Level 2
- 1. The RCIC syt. tem shall adequately take over water level ' ES protection. The relief valves must reclose properly (without leakage) following the pressure transient.
Amend. 55 4/02 14.2-151
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- I oo FSAR pi);g3 erfiew o9 M 'h5t
- For the full MsIV closure -
'"' ; r- , predicted ) -
F analytical results based on beginning of cycle design-basis analysis, assuming no equipment failures and applying appropriate parametric corrections, will be used as the basis te which the actual transient is . compared. The following table specifies the upper limits of these criteria during the first 30 seeends , following initiation of the indicated conditions. l Initial Conditions Criteria
~Done Increase In Ihcrease In Power Pressure Heat riux Dome Pressure (psial_ (%) (psi) 34
(%)_ _ C I} G rpcM /eco-sete M>& LMe' / 7 ! 1 n;;..J i o. ~mi e u-+y W y 4 - n--i,.. Lp. d
- 2. Initial action of the RCIC and MPCS shall be automatic when water Level 2 is rasched, and eyster performance hall be within specifications. - Murrt
, 1/. Recircultp n runback shall_ occur M irculation anurares wnen Level 2 _is reache .
ep mp snm ue a stTc 14.2.12.3.22.3 Test Number 250 - Main Steam Line Flow Venturi Calibration
- a. Test _0_biective The purpose of this test is to calibrata the main steam flow venturis at selected power levels over the 55 entire core flow range. The final calibration takes place with the data accumulated along the 100 percent rod line.
- b. Prerequisites The preoperational tests have been completed; the PSRC has reviewed and, approved the test procedures and initiation of testing. Instrumentation has been checked or calibrated as appropriate.
- c. Test Procedure Beginning at approximately 40 percent core thermal power, pertinent plant data will be taken along the .
75 percent rod line at selected power levels. The same process will be repeated along the 100 percent rod line. The accumulated data will then be compared Amend. 55 4/82 14.2-152
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Results for H5IV C1cours~ The folleving figures provide a cesperison of the actus1 75% power test results to that eritinally predicted for the 1003 power test. Comparisons , have been made for dome pressure increase, percent of rated povar and steam flow rated versus time af ter iselation at both 75% and 100% power as listed j below and attached. Definiticas (as used in attached figures): TEST DATA - Data in figures represents actual plant performance during April 7, 1983 event. CRIGINAL PRIDICTION - Data in figures represents results fres analysis conducted by Genersi Electric to predict plant performance given the HSIV closure event. This set of analyses (75: and 1002 runn) were based on standard Ceneral 11ectric conservative assumptions for expected performance of CCNS, e.g., seram time, MSIV clesure time and steam shutoff rate, feedvetor coastdown, etc. The original Prediction analyses for 1002 power were perforsed prior to 4/7/85 in preparation for the exsected execution of the test at full power. The Original Frediction analyses for 75% power were parforr.ed following the 4/7/65 event in support of this MP&L submittal. POS: TI5! - Following the evaluation of the 4/7/45 event data, several FREDICTION observations were made regarding actusi plant performance. These observations were incorporated into the original ! (75% & 100%) ! r.edel by revising certain input information, e.R., stans ' shutoff characteristics, feedwater systre coastdown, etc. The data labeled POST TEST PREDICTION represente results of analyses performed emploving these refinements which ensbie the predictive model to more closely describe actual plant performance. Analyses were run at 75r and 100% power and compared with original prediction and/or actual event data, depending en the situation. ;
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-- od lAa 10:E a6' 10. a.1)i! 21 ;t r.13 Figurs 1 - MSIV Closure at 752 Fever, 'k h ) k ATTACMED - 1 FIC'.'RIS Deee Pressure Ineresse (PSI) i Figure 2 - MSTV Closure at 75% Power, Power (! of Rated)
Figure 3 - MEIV Closure at 75% Power. 8 team Flow Rate (I of Initial) f Figure 4 - MSIV Closure at 100! Power. Dor.e Pressure Increase (PSI) Figure 5 - MS!v closure at 1001 Power. - Power (1 of Rated) Figure 6 - MEIV Closure at 100% Pcwor, Steae Flow Rate (X of Initial) 4 e e
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i GG!ERAL ELECTRIC COMPANY TECilNICAL ANALYSIS
. . . IIOPE CREEK GENERATING STATION TEST NUMBER 23B - MSIV FULL ISOLATION TEST SIMPLIFICATION OBJECTIVE:
Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraph 5.m.m requires demonstration that the dynamic response of the plant is in accordance with design for the case of automatic closure of all main steam isolation valves. Test Number 23B, MSIV Full Isolation, will perform a simultaneous full closure of all MSIVs at Test Condition 6. Transient behavior of the plant and associated systems will be determined during and following the main steam isolation. Correct relief valve performance will also be demonstrated for the test. It is currently planned to conduct this test at Test Condition 6; however, should an inadvertent MSIV full isolation occur at some lower power level (> 70%) it is proposed to substitute the results from the lower power eve.it, with appropriate analysis and extrapolation, for the higher power level results. This proposal, in effect, expands the " test window" to 70-100% power. The proposed testing and analysis will demonstrate that Regulatory Guide 1.68 objectives are met. DISCUSSION: Response of the plant and its associated systems during and following an automatic closure of all main steam isolation valves is determined by analyzing test data and comparing the results to acceptance criteria, which define the required system performance. These criteria require that vessel dome pressure, MSIV closure times, and simulated heat flux are within specified limits. In addition, the feedwater control systen response must prevent flooding of the steam lines, the reactor must scram, and proper operation of relief valves, HPCI, RCIC, and recirculation pump trip must occur if initiated. The MSIV full closure at Test Condition 6 will provide data that will verify compliance with the acceptance criteria. Howeve r, should an inadvertent full closure of the MSIVs occur a t a lower power level (> 70%) supporting analysis will confirm that the criteria will still be met. Analysis will be performed using actual plant conditions during the inadvertent trip to verify that the results can be adequately predicted by the model. Using the above inputs, the analysis will then be performed at rated power conditions to demonstrate that the criteria would be met at Test Condition 6 had the test been performed at f ull powe r. The full power predictions will also be compared to other BWR plant tests at full power to verify 1
the adequacy of the extrapolated results. Support for this expanded test window (70-100 % powe r) has already been provided for another startup test program. CONCLUSION: By using analysis of an inadvertent MSIV full closure at a lower power level (> 70%) and extrapolation of these results to rated power, the objectives of Regulatory Guide 1.68 (Revision 2; August 1978), Appendix A, paragraph 5.m.m can still be satisfied withcut requiring additional testing at Test Condition 6. Performance of the reactor and its associated systems for this test is well documented, and based on data obtained from previous plant startups it can be shown that the results from the test at a lower power level may be extrapolated to the high power criteria. Expanding the test window for MSIV full closure will not adversely affect any safety related system or the safe operation of the plant and therefore does not involve an unreviewed safety questi in. Therefore, Test Number 23, MSIV Full Isolation, can be simplified by expanding the test window to 70-100% power.
REFERENCES:
- 1. Letter, L. F. Dale (MP&L) to H. R. De n to n ( N T< C ) , " Grand Gulf Nuclear Station Unit 1 Proposed Modification to FSAR Startup Test Program; Operating License Condition R.C.(31)", Docket No. 50-416, April 23, 1985.
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r I* GENERAL ELECTRIC COMPANY TECHNICAL ANALYSIS i G HOPE CREEK GENERATING STATION TEST NUMBER 24 - RELIEF VALVES TEST SIMPLIFICATION - REDUCED NUMBER OF TESTS OBJECTIVE: Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraphs 4.p and 5.t require that the operability, response times, relieving capacities, setpoints, and reset pressures for main steam line relief valves be verified. Test Number 24, Relief Valves, demonstrates the proper operation of the main steam relief valves. Operability testing is currently planned to be performed at low pressure (250-500 psig) and between Test Conditions 2 and 3. It is proposed to replace the startup testing during heatup and between Test Conditions 2 and 3 with testing at Test Condition 1. DISCUSSION: Actuation of the relief valves at low pressures has been identified as a contributor to valve seat damage caused by reseating against abnormally low pressure (INPO newsletter of September 28, 1984). Therefore, relief valve testing at low reactor pressure should be minimized to reduce unnecessary damage to the valve seats. Acceptable response of the relief valves is determined by analyzing test data and comparing to acceptance criteria which define the required system performance. The criteria require that there is positive indication of steam discharge during a manual actuation of each valve, that the pressure control system response is stable, and that the discharge temperature remains within acceptable limits. These criteria can be demonstrated during single valve testing at rated reactor pressure with steam flow greater than relief valve capacity. Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraph 4.p requires demonstration of relief valve operability at rated temperature during low pcwer testing (defined as "normally at less than 5% power"). To provide adequate control of system pressure, the testing must be performed at a steam flow that is greater than the individual relief valve capacity. Since the relief valve capacity is typically 5-7% steam flow, testing is proposed to be done at approximately 10-20% power (Test Condition 1), approximately the test power level of paragraph 5.t of Regulatory Guide 1.68. Previously, testing at 250 psig was performed to ensure that re. lief valves would function properly at rated pressure if depressurization was required. This testing responded to 1 1 1
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- F occurrences where high air supply pressure or incorrectly wired logic circuits resulted in failure of energized solenoids to de-energize and resultant valve opening. In response to these instances, qualification tests performed on these valve designs have demonstrated that the design satisfies specified requirements. In addition, bench tests are performed on each relief valve to provide assurance that each assembly will perform satisfactorily and preoperational tests check out the adequacy of electrical power supply, logic, and air supply.
Testing of the ADS valves is required by Plant Technical Specification Surveillance Requirements to ensure that depressurization capability exists. Overpressure protection during transients at power levels below 25% power can be adequately handled by the ADS valves. Therefore, single valve testing at Tes t Condition 1 meets the objective of demonstrating relief valve operability required by Regulatory Guide 1.68, Appendix A, paragraphs 4.p and 5.t. CONCLUSION: Testing of the relief valves at Test Condition 1 demonstrates the operability of the relief valves. This proposed testing change does not adversely affect any safety systems or safe operation of the plant and therefore does not involve an unreviewed safety question. Test Number 24, Relief Valves, can therefore be simplified by replacing the testing during heatup at low pressure (250-500 psig) and between Test Conditions 2 and 3 with testing at Test Condition 1 with steam flow greater than relief valve capacity. 2}}