ML20086U243
| ML20086U243 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 12/30/1991 |
| From: | Labruna S Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20086U248 | List: |
| References | |
| NLR-N91219, NUDOCS 9201070282 | |
| Download: ML20086U243 (23) | |
Text
{{#Wiki_filter:._ -_ - _____________ If , > ew a % ge l 1941f 2t' fMJGM C Orn) . p n y Stanley LaBrum Ntx Sevce Umoc ou Ga3 D:m;wy P O tw ?W Hw ons 1%e.10 WO3ti wb3Blao u , n, <,..1 ,.n. DEC 3 0 1991 NLR-N91219 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlement RESPONSE TO SAFETY EVALUATION ON STATION DLACKOUT HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Public Service Electric and Gas (PSE&G) submits the attached information in response to the recommendations and questions identified in the NRC " Safety Evaluation of Hope Creek Generating Ststion Response to the Station Blackout Rule" (dat11d October 25, 1991). Atts hment 1 contains a response to each individual roconnencation stated in the Safety Evaluation.
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In response to the NRC Safety Evaluation, a reevalW tion of containment isolation capability during a Station ilackout (SBO) event was performed. PSE&G has determined from tha ni rect Aluation that modifications associated with the main steam dbain line, drywell floor drain line and drywell equipment drain lines would be prudent to ensure containment isolation capabilit y during a SBO event. A comprehensive schedule for completion of all proc 93nre changes necessary for Hope Creek to cope with a station Bla:kout a1d the proposed modifications contained in Attachment 1 wiil be completed by March of 1992. Upon completion of thii schedule, a copy will be submitted to the NRC. Preliminary evthuation indicates that these modifications and procedure changes will be complete by October of 1993 in accordanco with 10CFR50.63 (c) (4) . If you have any questions regarding this submittal please contact us. Sincerely,
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l l . f . e Document Control Desk 2 N1R-N91219 . 3 E C Mr. Stephen Dembok Project Manager Mr. T. P. Johnson Senior Resident Inspector Mr. T. Martin, Administrator Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 h 9
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i I f- , ') , i i ATTAClE4ENT 1 ; NLR-N91219 l 2.1 STATION BIACKOUT DURATION 1 Recommendationst 1 i The licensee should provide site-specific data and ! analyses to demonstrate that the plant should be in ESW l Group "2", otherwise the plant will be placed in ESW ! Group "4" in accordance with Table 3-2 of NUMARC 87-00, which in turn will place the plant in Group "P2". If ' the licensee can not provide site-specific data and analyses to demonstrate that the plant should be in ESW Group "2", then the licensee should change the target EDG reliability from 0.95 to 0.975 in order to be a 4-hour coping plant instead of an 8-hour coping plant.
-Response ;
Site specific weather data and analysis of Hope Creek plant han been performed by Report No. NUS-5175 ;
" Estimated Frequency of Loss of Offsite Power Due to i Extremely Severe Weather (ESW) and severe Weather (SW) ;
of Salem and Hope Creek Generating Station"(retar to -
-Attachment 3 for report). This report demonstrates that the estimated frequency of loss of offsite power [
due to extremely severe weather (ESW) for Hope Creek i plant-is 6.5xE-4 which in accordance with Table 3-1 of ' NUMARC.87-00 places Hope Creek in ESW Group 2. Report i No. NUS-5175 is controlled under PSE&G's document I control program. 2.2.2 CIASS 1E IETTERY CAPACITY r Recommendations ! The licensee should re-evaluate the battery capacity considering more than one start attempt of the EDGs. The licensee should address the operation of the SB0 , equipment at the final terminal voltage of 105 V and i 210 V for the 125 V and 250 V de batteries,
-respectively. In addition, the licensee should '
evaluate.the battery capacity when there is no heat available in the battery room. .
-Responset l-h a. For Hope Creek's SDo battery load profile, the l -diesel generator field flashing load has been L considered for the complete first minute-(0-1 L minute) and also for the complete last minute 1 (239-240 minute). Hope Creek's Standby Diesel ,
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, 's Generators are designed to start and attain ratad voltage within 10 seconds of the receipt of the starting signal (Refer to Hope Creek UFSAR section 0.3.1.1.3.10). Following the occurrence of a SB0 ovent, the Hope Creek 125V D.C. Class 1E battorios have sufficient capacity to allow the diesel to start more than once in the first minuto and also start during the last minuto of the SBo four hour durataon.
- b. Calculation Nos. El.4(Q) Rev. I and E4.2(Q) Rev. 1 ensure that the minimum voltage reached at the terminals of 250V and 125V D.C. bhtteries during SBO duty cycle, as por Calculation Hos. E45.001(Q) and E45.002(Q), will provido adequate operating voltage for the SB0 equipment during SDO coping duration.
Reference ,
- 1. Calculation No. E1.4(Q) - Hope Creek 125V and 250V Class 1E D.C. systems: Short Circuit and Voltage Drop Studies.
- 2. Calculation No. E4.2(Q) - Hope Creek Class 1E D.C. Equipment and Coraponents Voltage Study.
- 3. Calculation No. E45.002(Q)-HCGS 125V D.C.
Battery Capacity verification Calculatic.n (SBO).
- 4. Calculation No. E45.001(Q) - HCG 3 250V ll.C.
Battery Capacity Verification Calculatio) (SBO). g vf
- c. Hope Creek UFSAR Section 9.4.1.1.4 states th6, the Control Equipment Room Supply System is desighbd to maintain the battery room temperature at 77t3'T during normal plant operating condition. The #25V and 250V battery room temperatures are maintained by safety related thermostatically controlled temperature elements. The setpoints of the temperature elements are 77'F for 125V Class 1E battery rooms and 7713*F for 250V Class 1E battery rooms. The manufacturer's accuracy is i l'F. A weekly surveillance program also exists at the Hope Creek station to record the battery room temperatures. The 250V and 125V battery SB0 battery calculations, E45.001(Q) and E45.002(Q)
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% , 'n have considered 72*P as the battery electrolyte temperature which is lower than the lowout electrolyte temperature anticipated under normal operating conditions. (Refer to Section 7.2.2 of NUMARC 87-00). It is alsL snticipated that during the station blackout event, tne electrolyte temperature will increase since the battery will not be floating but will be discharging at a higher rate to supply power to the station blackout loads. NUMARC 87-00, Section 2.7.2 (2) (B) statast "Also, the mass of battery electrolyte is sufficient to resist significant temperature drops over a four hour period due to lower battery room temperatures since battery cell materials are not efficient thermal conductors. Therefore, a decrease in battery capacity due to temperature decreases in electrolyte under station blackout conditions does not warrant further consideration." On the pretext of the above considerations, it can be concluded that the Hope Creek 250 Volt and 125 Volt batteries have adequate capacity to supply the station blackout loads for a four hour SBO coping period. 2.2.4 EFFECTS OF IDSS OF VENTILATION Recommendation: Question #1: Provide and justify that the initial temperatures used in the heat-up calculations are the maximum allowable and that the heat load accurately reflects those during a SBO event. Responset The initial room and wall temperatures and heat loads during the Station Blackout for the Dominant Areas of Concern (DAC) for Hope Creek Station are as tabulated belows 3 l
I I I 1 4 l ttKN l RCKM l IW111AL Wall l TOT AL ht AT LDAD l l DEStalPl!ON l WO. l ftMP OF (*C) l (MICH AND EttC1) l l l l l DL*lhG $80 IN WAf f l l 1 I I i l 1. RCic Pep som l 4110 l 13.9 (93*F) l 22,476* " l l 2. HPCI Pw p som l 4111 l 33.7 (92.5'F) l 75,392 "
- l l 3. Battery Rocsa l 5104 l 33.3 (92'F) l 23P *
- l l 4. Battery Rocra l 5128 l 33.3 (92'F) j 112*** l l 5. Control l 5302 l 37.8 (100*F) l 76,778 " l l twigrnent Room l l l l l 6. Irwerter tocss j 5447 l 34.7 (94.5'F) l 1,009 " l l 7. Irwerter tocan j 5448 ) 34.7 (94.5'F ) l 6,600 " l l 8. Electrical Access l 5501 l 34.7 (94.5'F) l 15,954 " l l Rooa l l I I l 9. Control toern j 5510 l 32.8 (91'F) l 26,130 " l l 10. Irwerter econ [ 5615 l 34.7 (94.5'F) l 3,016 " l 1 I i i I
- Taken f rom Catt. #H 1 CK TOC 0734
" inken frorn Cate. #H 1 CK POC-0735
"* 1aken f rcnn Csic. 8H 1 GK POC-07M ord H 1 GR 60C 0733 The initial room temperatures for the above noted areas have been taken from the UFSAR.
For DACs that have an initial air temperature higher than the adjacent rooms, the initial wall temperature for the loss of ventilation calculativn is based on the maximum normal operating air temperature for the DAC. For DACs that have an initial air temperature less than the adjacent rooms, the initial wall temperature is based on an average of the maximum normal uperating air temperature for the DAC and adjacent rooms. The most limiting adjacent room temperature has been considered. Where necessary, the temperature in each adjacent room han been calculated for each wall on an individual basis. Impact of Solar Radiation is incorporated for Room 5501. These effects are negligible for all other rooms due to location, thickness of slabs, or the initial wall temperatures envelope the solar effects. The initial wall temperatures are defined based on the surface temperatures expected during normal operation. When the wall temperature is expected to be less than the initial DAC temperature, the wall is specified as equal to the initial room temperature. When the DAC is surrounded by hotter 4
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, 'e j rooms and_ areas, the wall will be expected to be i at a higher temperature. Th? wall temperature has t been defined as the avetage of the inside and ;
outside surface temperatures. To estimate heat generation rates in a room, heat , generation rates were calculated by a detailed - evaluation of the equipment expected to be powered and operable following a SBO event (Refer Cales. fH-1-GK-MDC-0735 and H-1-GR-MDC-0733). These heat generation rates are conservative since all identified components will not be operating continuously during a Station Blackout. Recommendation: Question #2: Discuss'the modification / procedure for removing the acoustic cei' ling tiles in order to provide i sufficient cooling'in the control room. : Responses-Acoustic. ceiling tiles will be removed in the ' Control Room in order to provide sufficient cooling. Station Procedures will be developed to address the number of tiles to be removed and timeframe for removal. Evaluations are currently in progress to determine the timeframe for removal of the acoustic ceiling tiles in accordance with'the SBO coping analysis study. Recommendation: Question #3: Describe-what is being proposed-for reducing the heat load and temperature heat-up in the control equipment room and state which temperature is being considered when performing the assessment of equipment-operability-in the room. '
Response
i In order to reduce the heat load in the Control EquipmentLRoom (5302) several circuits powered by p non-Class lE batteries must be de-energized. The l E circuits to be de-energized and their locations ! are: . 5 !
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IDCATION PANEb NO. CIRCUIT 3449 1AD318 12, 27, 28 1BD318 12, 19, 24, 27 D123'34 -1CD318 11, 12, 14, 27 D124'34 1DD318 11, 12, 14, 27 5102 1AJ483 7- 15, 21 1BJ483 6, 22, 23 1CJ483 6 F163'68 1AJ484 1, 2, 3, 4, 5, 6, 8, 9, 14 1BJ484 1, 2, 3, 4, 5, 6, 14 1CJ484 1, 2, 3, 4, 5, 6, 14 1DJ484 1, 2, 3, 4, 5, 6 5623 1CJ492 5, 14, 17 1DJ492 3, 4, 5, 7, 9, 16 In. order to reduce temperature, Door 5310C which opens from the control Equipment Room to the unoccupied area (5318) located on the North wall of Room 5302 will be opened. The maximum steady state temperature calculated for this room is 118.3' F with reduced heat load and door open as noted above. Station Procedures will be developed to address the de-energization of the circuits powered by non-class 1E batteries and to open door 5318C. ' Recommendation: Question #4: Demonstrate that all the assumptions made for drywell calculation are conservative and that the assumed initial conditions accurately reflect those expected during a SBO event.
Response
References:
- 1. Bechtel Calculation No. 1402
- 2. Engineering Evaluation H-1-GS-MEE-0514 l-6
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e Initially, Bechtel had been contracted by PSE&G to perform an analysis to determine containment response during a Station Blackout (SBO), Reference 1. Howevet, independent review by a PSE&G contractor indicated some deficiencies and inaccuracles in Bechtel's calculation; hence, the Probabilistic Hisk Assessment (PRA) group of PSE&G was requested to determine if Reference 1 needed to be revised or replaced. Upon review of Reference 1, the PRA group determined that the input file that was used in tho Bechtel proprietary computer code was not the Hope Creek Generating Station (HCCS) plant specific data and that the cases studied were not adequate for this investigation. The PRA group recommended that Reference 1 be kept in the file as a reference only, and initiated a new Enginaering Evaluation, Reference 2, not only to evaluate the containment response based on the present Emergency Operating Procedure ( EOP) , but also to determine whether the ROPs needed to be revised. The PRA group used the Modular Accident Analysis Program (MAAP) , Version 3.0B - Revision 6.05, to determine the HCGS containment responso during the SBO. As is documented in Reference 2, the PRA group studied 13 different cases. Also, this study included SD0 scenarios during which certain required systems, such as the accumulators for the Safety Relief Valves (SRV's) would become inoperable. (These additional runs were not necessary from the NUHARC 07-00 point of vicw; however, in our opinion, this extra effort was essential from the Individual Plant Examination (IPE) point of flew.) The HCGS input parameter file and input data sets used in MAAP were developed and reviewed by PSE&G personnel and independently verified by the contractor. As is customary in our analysis, the most conservative plant data were used during the SB0 study. For example:
- 1. A pressure dependent leakage of 100 GPM from the vessel was assumed at time zero and at reactor pressure of 1020 PSIA. Using a reactor de-pressurization rate of 100*F/HR, this leakage averaged about 54 GPM for the entire four hours. However, it should be noted that the manufacturer data, along with data provided in response to item II.K.3.25 7
of NUREG-0737, .,uggests that no appreciable leakage will occur during the first two hours, and that HUMARC 87-00 requires us to assume only 18 GPM leakage por recirculation pump, and the Technical Specifications suggest that a total of 30 GPM of identified and unidentified leakage from the reactor is porsible.
- 2. It was assumed that the SDo occurs during the summer days. Thorofore, the water in the Condensate Storage Tank who assumed to be 90'F. However, in one case we assumed that the water temperature in the CST would be about 100*F. This did not change any of the previous results, appreciably.
- 3. The volumetric analysis based on the temperature ptofile in the primary containment during normal operation indicated that the initial drywell temperature was above 115'F; however, we chose the initial pedestal and drywell temperature to be 135'F
- the Technical Specification limit. Also, the initial torus water temperature was assumed to be 95'F, the maximum allowable by the Technical Specifications.
- 4. Initial containment pressure is normally 0.5 PSIG, and it was assumed to be the condition in our analysis. However, in one case we raised the pressure to 1.5 PSIG - the Technical Specification limit - and did not detect any noticeable change in the results.
- 5. In all the cases it was assumed that the initial reactor power was 3293 MWT; however, in one case the sensitivity ana is was performed to show that if the ps er is raised to 3323 MWT; there would not be any noticeable change in the results.
- 6. De-pressurization takes place using two SRV's and, according to the EOPs, both the Reactor Core Isolation Cooling (RCIC) and the High Pressure Cooling Injection (HPCI) pumps inject to the vessel from the Condensate Storage Tanx (CST); however, HPCI was assumed to be tripped 30 minutes into the run. The results 8
P l I *. - indicate that, depending on the sensitivity of the SRV deadbands, the amount of makeup needed for maintaining the normal reactor water is anywhere between 120,000 gallons to 131,000 gallons; well within the 135,000 gallon minimum allowed capacity of the CST. In conclusion, using the most conservativo input data, wa determined that the present procedures at HCGS are adequata. Using these procedures, the maximum drywell and torus temperature in foor hours woald reach 220'F and 185'F, respectively - this is well within the design limit of the plant. The containment pressure during the SBO was determined to roach about 10 PS7G in four hours. Also, it was determined that the maximum makeup capability that would be needed during the SBO would be about 131,000 gallons. Recommendation: Question #5: Provide procedure for opening the instrumentation and concrol cabinet doors within 30 minutes of the oncet of a SBO in accm/ dance with NUMARC 87-00, Appendix F.5.
Response
Cabinet doors for panels containing SBO equipment that are required to be opened during a SBO event vill be opened in accordance with the SBO coping analysis study. Procedures to address this issue will be written by October of 1993 in accordance with 10CFR50.63 (c) (4) . Recommendation: Question #6:
~ Provide procedures for opening the doors to rooms where the heat-up calculations were performed with the credit taken for opening the area doors.
Response
The doors, for which credit has been taken, have been identified and will be opened following an SBO. Procedures to address this issue will be written by October of 1993 in accordance with 10CFR50. 63 (c) (4 ) . 9 ; 1
n, <. Recommendation: Question #7: Verify that there are no valvec in the main steam tunnel which need to be operable should containment isolation become necessary.
Response
There is one valve in the Main Steam Tunnel which requires manual operation (Refer to Response 2.5). Since the motor operated valve will bs closed by the local handwheel, Environmental Qua:ification for the motor operator and associated cabling is not a concern. 2.2.5 CONTAINMENT ISOIATION Recommendation: The licensee should list the Containment Isolation Valves (CIVsj which are either normally closed or normally open, and fail as-is upon loss of ac power, and cannot be excluded by the criteria given in. Regulatory Guide (RG) 1.155. The
.icensee should provide in procedures the actions that must be taken to ensure that the above cited 14 valves can be verified to be closed during a SBO. This information and verification should be included in the documentation suppcrting the SB0 submittals that is to be maintained by the licensee.
Response
In the March 28, 1991 Revised Station Blackout coping Analysis submittal by PSE&G, a total of 26 valves were excluded from the list of CIVs requiring closure during a SBO event based on exclusion criterion ". aased upon the SER issued, it is apparent that tooru details a're required to be provided f; review and justification for RG 1.155 exclusion. The 26 valves excluded, based on water seals, were reevaluated using only the exclusion criteria identified in RG 1.155. These 26 valves belong to ESF category systems required to provide continued I 10
l safa shutdown upon restoration of AC power (exception --Torus Water Cleanup System is not required for continued safe shutdown). These . systems meet exclusion criteria #4 (closed ! loop systems) in that they are normally non-radioactive systems which do not communicate s directly with the containment atmosphere. Our assessment of these systems e cludes credit for line submergence to provide iwolation from the containment atmosphere. All lines have been verified to terminate below the Technical Specification minimum torus water level. The following piping systems were evaluated as closed loop systems meeting the RG 1.155 containment isolation exclusion criterlat RHR Core Spray HPCI RCIC Torus Water Cleanup
- 1. These systems with the exception of portions of the Torus Water Cleanup System are classified / qualified Seismic Category I, therefore system piping failures were not considered.
- 2. These systems are' continuous closed systems outside the containment with submerged-suction and discharge lines preventing direct communication with the Torus atmosphere.
- 3. These systems are considered non-radioactive during normal operation (13. they are isolated from.the reactor pressure vessel).
- 4. With the exception of the Torus Water cleanup, those systems are classified as Engineered Safety Features (Quality Group Classification A & B, as per RG 1.26) which might reasonably be called upon to function in order to mitigate Station Blackout. Torus Water. cleanup piping is classified as Quality Groups D & D. As such these systema should not be isolated.
Hope Creek will provide priority isolation to 12 of the 14 previously identified valves that require closure during a SB0 in the Abnormal Procedure dedicated to Station Blackout, HC.OP-AB.ZZ-0136(Q). The other two (of the 14) 11
i identified valves will not be closed unless l conditions exist that require their associated l penetrations to be isolated. Specifically, HPCI and RCIC Turbine Steam Supply isolation valves F3HV-F003 and FCHV-F008 respectively, will not be closed as tl.ey are required to be open during SB0 conditions for reactor level and pressure control. However, appropriate guidance will be provided in HC.cP-AB.ZZ-0135(Q) for isolation of these two penetrations as necessary. Three additional valves of the original 14 listed also have unique isolation considerations. Refer to discussion under Response 2.5 for detailed information. 2.5 PROPOSED MODIFICATION Recommendation: The' licensee should explain why modi"ications to the
-inboard MSIV drain line, drywell sump drain line, and drywell equipment drain lines are no longer required.
Response
All containment isolation valves required to meet containment isolation capability have been re-evaluated. Hope Creek has determined that all valycs are accessible and can be closed / verified closed locally. However, the location of the three valves in question pose challenges to the operator's ability to perform containment isolation. To provide additional defense in depth for containment isolation, Hope Creek proposes a_ modification to address the following valves. VALVE VALVE NAME MODIFICATION HV-F019 Main Steam Drain Install access hatch in existing grating above the valve to facilitate manual operation. In lieu of isolating the following two valves, a more practical approach is described below. 12
1 i *.. VALVE VALVE NAME MODIFICATION HV-F004 Drywtell Floor Drain
- Change dovnstream
, Sump Discharge air operated valve to fail closed on loss of air or AC power HV-F020 Drywell Equipment Drain
- Change downstream Sump Discharge air operated valve to fail closed on loss of air or AC power
- The drywell floor and equipment drain sump discharge lines have inboard and outboard MOVs providing normal containment isolation function. PSE&G proposes to changa the air operated valve located approximately twenty-five feet downstream of the 3" mo* r-operated outboard containment isolation valve to i failed closed valve (see generic configuration below). A one inch capped drain line is the only branch connection between the two valves. The air operated valve is not classified as a safety-related valve and is not contained in Hope Creek's ASME Section XI IST program for valves.
From wJto guno M Joracmaste f u contatrvnent t
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Change to FC on isolation loss of power valve logg of er The proposed modifications on the drywell floor and equipment drain lines will place these lines into containment isolation exclusion category #2, i.e. valves that fail closed on a loss of power. NOTE: this eliminates these two valves from the list of 14 valves. 13
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2.6 QUALITY ASSURANCE AND TECHNICAL SPECIFICAT1oNS Recommendation: The licensee should verify that the SBo equipment is covered by an appropriate QA program consistent with the guidance of RG 1.155.
Response
2.6.1 DESIGN OF EQUIPMENT LIST An equipment list was compiled and consolidated from Reference 4. This Equipment List provides a consolidated listing of electrical, mechanical, control and instrumentation equipment, and components located in the various areas of the plant that are required for coping with a SBo event. The Equipment List also identifies the QA category of each equipment. Where non-nuclear safety related equipment has been used for SBo it has been assigned a QA Requirement of RG 1.155 Appendix A/B. The station blackout responso equipment list is compiled to show for each piece of equipment, the:
- Name;
- Location; Power source / PATH (Protection Channel);
- Function; QA category; and operating and monitoring location other data is provided where readily available.
1 The field descriptions for equipment database are provided below. Appropriate abbreviations and definitions are provided herein for proper identification of source information. 1 l 14
I t .. STA'0 ION BLEKOUT EQUIPMENT DATABASR_INEQBMAT10H FIELD /NAME DESCRIPTION FUNC FUNCTION SYST SYSTEM PATH PATH (PROTECTION CHANNEL) COMP COMPONENT DESCR DESCRIPTION
*NP NORMAL POSITION
*FLP FAILED LOSS OF POWER
*FLA FAILID LOSS OF AIR
*DBO DESIRED BLACKOUT PWRShC POWER SOURCE CONMON CONTROL MONITOR DEVICE LOCDEV LOCATION OF CONTROL /
MONITOR DEVICE LOC LOCATION OF THE COMPONENT PNL PANEL
%KR BREAKER 2F REFERENCE COMM COMMENTS
- NOT DISPLAYED ON EQUIPMENT LIST REPORT IN SECTION 5.0 FIELD /NAME DESCRIPTION INFORMATION SOURCE QAR SAFETY RELATED MMIS Q - NUCLEAR SAFETY RELATED N - NON NUCLEAR SAFETY RELATED A - SBO EQUIPMENT RG. 1.155 APP. A APPLIES QGC NUCLEAR CLASS OF COMPONENT MMIS QA QA TESTING LZQUIREMENT -- MMIS YES/NO EQ ENVIRONMENTAL CLASSIFICATION MMIS SEIS SEISMIC CATEGORY MMIS
'RC RECOVERY OR COPING COMPONENTS LISTED IN TENER7 REPORT (REF. 4.4)
R - DENOTES RECORDS TAKEN FROM TENERA'S RECOVERY EQUIPMENT LIST (APPENDIX II) C - DENOTES RECORDS TAKEN FROM TENERA'S HOT SHUTDOWN EQUIPMENT LIST (APPENDIX I) 15 b ---_________________ _ __ _
- x. ..
The Managed Maintenance Information System (MMIS) database was utilized to determine the equipment field information including the QA classification designation for SBO coping and recovery components. References 2 (RG 1.155) and 5 were used in determining the QA classifications of components when the appropriate QA classification code was not specified by MMIS. Any non-nuclear safety related equipment used for coping with a SB0 event are identified with N, A.
"N" indicating non-nuclear safety related and "A:
indicating that Regulatory Guide 1.155 Appendix A quality assurance program applies. 2.6.2 Equipment List See Attachment 2 for sample page from equipment list. 2.6.3 Conclusion All equipment necessary to cope with a SBO event is covered by a QA program consistent with the guidance provided by RG 1.155, 2.6.4 References 1 10CFR.50.63: Loss of All Alternating Current Power 2 USNRC Regulatory Guide 1.155 (Task SI 501-4), August 1988 3 NUMARC 87-00, November 1987; Guidelines and Technical Bases for NUMARC Initiatives Addressina Station Blackout At Light Water Reactors 4 PSBP No. 312451-01: Station Blackout Technical Report for Hope Creek Generating Station, dated April 14, 1989. 5 PSE&G Procedure No. DE-AP.ZZ-0019 (Q), R3v. 2: Dasign Classification of Structures, Systems and Components for Hope Creek Generating Station. l 1 16
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2.7 EDG RELIABILITY PROGRAM Recommendation The licensee should implement an EDG reliability program which meets the guidance of RG 1.155, Section 1.2. If an EDG reliability program currently exists, the program should be evaluated and adjusted in accordance with RG 1.155. Confirmation that such a program is in place or will be implemented should be included in the documentation supporting the SBO submittals that is to be maintained by the licensee. Responset
- 1. Surveillance that identifies. EDG support systems and subcomponents, frequency and scope of testing, and incorporates manufacturer's recommendations.
At Hope Creek, procedures are in place to insure that proper line up/ operation of EDG sub components and support system exist as described in NUMARC 87-00 tables E.1-1 and E.1-2. The procedures are in the form of shift and weekly logs. Status lights (2) and an annunciator alarm provide " ready for auto-start" indication in the control room and the local panel in the engine room. This is in addition to the dynamic monthly, 6 month and 18 month surveillances. HC.OP-DL.ZZ-0006(Q) DAILY HC.OP-DL.ZZ-0016(Q) WEEKLY HC.OP-DL.ZZ-0001(Q) WEEKLY HC.OP-ST.KJ-0001-4(Q) MONTHLY (6 MONTH) - HC.OP-ST.KJ-0005-8(Q) 18 MONTH The equipment operator assigned to the diesel building visits all EDG cells an a daily to weekly basis. System engineering is made aware of any EDG system or subsystem degradation that may take place outside of normal working hours. Fuel oil and lube oil surveillances are performed on a weekly, monthly or as required basis with the results going to the system engineer for his review and action if required.
- 2. Performance monitoring of important parameters on an ongoing basis to obtain information on the condition of the EDG and key components so that precarious conditions can be identified prior to failure.
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mt 1 Surveillance test data is reviewed and trended by the system engineer. Graphic presentation of the trended data provides a means to assess EDG performance and implement / plan corrective maintenance prior to a potential failure, oil analysis-is used as a predictive technique to provide insight into the internal condition of bearings and other moving parts. Jacket water analysis is used to monitor the rust inhibitor in solution and condition of the water side of the EDG.
- 3. Maintenance designed for both preventive and corrective actions based upon operating history and past mainten3nce activitics, vendor recommendations, and the results of surveillance testing.
MMIS is the active repository for past, present and future Preventive Maintenance (PM) and ' Corrective Maintenance-(CM). Vendor recommendation identified in the vendor manual and service informative letters are reflected in the PM program. Components or sub components identified as requiring CM during surveillance testing are entered in MMIS for next available maintenance period or.are corrected immediately depending on their impact to system operability.
- 4. Failure analysis and root cause investigation to assist in developing effective corrective actions to prevent reoccurrence of failures.
All EDG failures in which a LCO/LER results, has a
. failure-analysis and root cause investigation performed. An action tracking item is opened to track this action to completion. An additicnal Component Failure Analysis Report (CFAR) provides component failure comparison for this utility as compared to the nuclear industry.
5.. EDG problem closcout process to ensure the resolution of a failure or a problem is properly implemented and successful. DR's (Non Conformance Reports) and CFARS are , formalized. PM/CM problems that are identified by planners, maintenance and operations are routed to the responsible system engineer for resolution. Tracking is done through the action tracking i 18
lW* L. b .
*d.
,.i.
system or: daily planning package. -Progress is tracked at the morning ~and afternoon.-planning-meeting. Problem closure'by planners, system
. engineers;and operationsz are. learned-
- organizational ^ behavior and-are not-formalized.
6.- EDG rollability data-system to ensure the o availability and reliability of important data and information relating lto EDG reliability. Although~m' centralized method'of capturing-and retrieving data important to thn EDG does not
-presently exist at HCGS, all of the information zlisted'in NUMARC 87-00 Rev. 1, Section E,3, is collected,-maintained available, and utilized.
The above information indicates Hope Creek's' compliance with RG 1.155 Sectis'n 1.2. .Thisl program will be adjusted in accordance with RG 1.155 and will be
-documented 11n the Hope Creek. Coping Analysis Study.
1
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ATTACHMENT 3 NLR-N91219 REPORT NO. NUS-5175 __ _ _ _ _ _ _ _ _ _ _ __}}