ML20087H873

Forwards Responses to Listed 831207 Requests for Info Re Environ Qualification of Electrical Equipment,Including Resolution of Deficiencies & TMI Action Plan Items (NUREG-0737) to Be Installed at End of Cycle 10 Outage

ML20087H873
Person / Time
Site:	Oyster Creek
Issue date:	03/16/1984
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-03-05.B, TASK-2.B.3, TASK-2.D.3, TASK-2.E.4.1, TASK-2.F.1, TASK-2.K.3.19, TASK-3-5.B, TASK-RR, TASK-TM NUDOCS 8403220012
Download: ML20087H873 (18)

Text

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GPU Nuclear Corporation Nuclear

=:r388 Forked R.ver, New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:

March 16, 1984 Mr. Dennis M. Crutenfield, Chief Operating Reactors Branch #5 U. S. Nuclear Regulatory Commission Washington, D. C. 20055

Subject:

Oyster Creek Nuclear Generating Station (OCNGS)

Docket No. 50-219 Environmental Qualification of Electrical BIuipment During our meeting concerning the subject matter on December 7,1983, the NRC staff requested UPUN representatives to provide:

1. Resolution of environmental qualification deficiencies at OCNGS identified in a Technical Evalution Report (TEit), prepared by Franklin Research Center under contract to the NRC.

2. Justification for continued op3 ration (JCO) for those items to be replaced, modiiied or evaluated after the current Cycle 10 refueling outage.

3. A list of TMI action plan items (NUREG 0737) to be installed by the end of the current (Cycle 10) refueling outage and identification of their qualification documentation.

4. Confirmation that all design-basis events at OCNGS which could result in a potentially harsh environment, including flooding outside containment, were addressed in identifying safety related electrical equipment.

5. Description of the method used to identify electrical equipment within the scope of paragraph (b)(2) of 10CFP.50.49 (i.e., "Nonsafety-related electrical equipment whose failure under postulated environm7ntal conditions could prevent satisfactory accomplishment of safety functions

.....").

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GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

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6. Identification of electrical equipment and their q.talification documentation for post-accident monitoring system as required by R~julatory Guide 1.97.

7. Reasons for extension of ceplacement or modification to the second refueling outage after March,1982.

Our responses to these items are provided below.

Resolution of Deficiencies Identified in TER Attachment I to this letter contains a table which lists proposed solutions for deficiencies identified by the TER for various safety-related components.

The table was reviewed in taa December 7, 1983 meeting. It has been revised to incorporate caninents generated in the meeting and additional information requested by th0 NRC staff.

JustificationforContinuedOperation(JC.i .

JCO's for those safety related equig ent in a harsh environment without complete qualification documentation were provided in Chapter 7 of our report to the NRC dated November 1, 1980. Subsequently, the JCO's were rcviewed by Franklin Research Canter and were found to be adequate. By our letter dated March 16, 1983 GPUN transmitted the revised Chapter 7 with additional JCO's.

The additional JCO's were reviewed in the December 7, 1983 meeting and the NRC staff requested further clarification. W e aforementioned TER concludes that some of the eqaipment which we considered to be qualified, still lacks complete documentation. The staff, therefore, requested JCO's for these equipments. Attachment II to this letter prcvides the required JCO's for your review.

TMI Action Plan Items (NUREG 0737)

GPUN letter dated October 23, 1981 listed the electrical equipment originally scheduled to be installed during the current. Cycle JO refueling outage in compliance with the NUREG 0737. W e letter also listed the clectrical equipment installed during the previous outage. Attachment III provides updated status of the NUREG 0737 items that were previously reported in the October 23, 1981 letter. JCO's for operational equipment without complete documentation are given in Attachment II.

DesignBasisEvents(DBEl Oyster Creek FDSAR analysis of the DBE IDCA shows that peak drywell temperatures do not exceed 285 'F. Wis is for a double-er;3ed rupture of a recirculation line with containment spray. The peak containment pressure is about 38 psig which is the saturation pressure at the peak containment temperature. n ese values are the highest containment temperature and pressure for breaks which occur below the core mixture level. Higher temperature and pressure in a containment atmosphere could be reached when the blowdown is pure steam, since the increased heat capacity of a droplet laden atmosphere reduces superheat conditions. On this basis GPUN performed plant specific analyses for steam line breaks above the core mixture !evel in determining the containtant temperature and pressure profile for environmental qualification of equignent. Methodology and results of the analy.as are provided in Chapter 2 of Environmental Qualification Report dated November 1, 1980 which was sutxaltted to you.

Plant specific high energy line break accidents outside containment were also evaluated to identify safety-related equipment which may te exposed to a harsh environment. Detailed discussion and results of the evaluation are given in Chapter 3 of the November 1, 1980 report. An ana]ysis was performed to evaluate the reactor building flood lewls for the oyster Creek Nuclear Generating Station following a high energy line break outside cotainment.

Description of the analysis and results are included in Cnapter 4 of the November 1, 1980 report. Werefore, the design-basis events and high ener.gy line break accidents at OCNGS were considered in the identification of the saftey related equignents which are essential to mitigate the postulated accidents and to achieve cold shutdown.

Affect of Nonsafety Equignent on Safety BIuipment JCP&I/GPU letter dated October 5,1979 in response to IE Information Notice 79-22 states that our evaluation of interactions between nonsafety systems and safety systems did not identify any adverse impact which would increase the consequences of any accidents analyzed in the FDSAR.

In addition, GPUN plans to conduct verification of proper selective coordination of protective devices or circuit breakers and fuses on vital i

buses to ensure that an electrical fault developed in nonsafety systems due to.

harsh envirotraent will not be transmitted to the safe shutdown systems. mis work will be conducted in the first half of the 1984 calendar year as part of the evaluation for the Fire Protection Program (10CFR50 Appendix R).

Isolation of the reactor protection system from nonsafety systems has been reviewed by the MC staff also in the Systematic Evaluation Program (SEP) for OCNC6 (SEP 'Ibpic No. VII-1A).

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Fosh Accident Monitoring Equignent (Reg. Guide 1.97, Rev. 2) l As indicated in Attachment III (NUREG 0737 Item II.B.3), four new isolation valves will be installed to mee[ the Post Accident Sampling capability requirement. Other modifications required by Reg. Guide 1.97 are still under investigation. As soon as the type of modification and electrical equipent involved are determined, GPUN will transmit SCEW sheet for those equipments.

Replacement Schedule Section (g) of 10CFR50.49 requires licensees to qualify the electric equipment in harsh enviror. ment by the end of the second refueling outage after March 31, 1982 or by March 31, 1985, whichever is earlier. Tne section (g) also states, "Tne Director of the Office of Nuclear Reactor Regulation may grant requests for extensions of this deadline to a date no later than November 30, 1985, As we have indicated in our previous submittals and in the December 7, 1983 meeting with your staff, we plan to replace partc and components without complete documentation during the second refueling outage after March 31, 1982. The second refueling outage will most likely take place after the March 31, 1985 deadline. However, every effort will be made to replace by Maren 31, 1985 those parts and components whose replacement activity does not place plant operation in an unsafe condition .

Very truly yours,

$ A j' Y H 'f)A&

er a. Fledler Vice President and Director Oyster Creek 1r/0122e cc: Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pa. 19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, N. J. 08731

Pags.1 of 8 ATTACHMENT 1 SPECIFIC EQUIPMENT EQ DEFICIENCIES O.C.

NRC PP.O POSED TEE CAT. DEFICIENCY SGLUTION NO.- (X)MPONENT 1,3,8 Motorized Valve Actuator II.A Documentation, Similarity, Qustified Units will be replaced this 9,10 V-17-54, V-16-1, V-17-19 Life, Spray, Steam Exposure, Radiation outage w/ Qualified Limitorque

.V-14-30, 31, 32, 33, 34, Model.

35, 36, 37 2,6,7_ Matorized Valve Actuator II.A Documentation, Qualified Life, Limitorque's Report fB0058, V-20-15, 21,<40, 41 Radia tion which will be referenced in the V-21-1, 3, 5, 7, 9, 11 SCEW sheet, describes the generic qualification and qualified life of subject actuator. Limitorque's Report

' d'J 0003 includes Radiation exposures.

II.C Qualified Life V-16-7 & 14 will be replaced 4 Motorized Valve Actuator V-16-2, 14 . this outage 'w/ Qualified Limit-orque Model.

V-16-61 V-16-61 is equipped w/ Reliance Class B Motor, instead of Peer-less Class B as mentioned in the TER. Limitorque's Report fB0058, which will be referenced ir t' e SCEW sheet, addresses th qualified life of subject actuator.

'S- ' Motorized Valve Actuator II.A Qualified Life, Radiation V-17-1,2,3 will be replaced this

-V-17-1, 2, 3 outage w/ Qualified Limitorque V-17-55, 56, 57 Meiel.

V-17-55,56,57 requires additional Documentation for Peerless Class B DC Motor.

' See Attachment 11 for JCO.

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ATTACHfElT 1 SPECIFIC EQUIPMENP EQ DEFICIENCIES O.C.

TER NRC PROPOSED NO. COMPCNENT CAT. DSFICIENCY SOLUTION 11 Motorized Valve Actuator II.A Documents. ion, Qualified Life, Spray Actuators are equipped with V-5-166, 148 Reliance Class RH A.C. Motors.

Limitorque's Report #B0058, which will be referenced in the SCEW sheet describes the generic qualification and qualification tests. Chemical Solution (Sodium Chromate) will be replaced w/ demineralized water.

12 Motorized Valve Actuator II.A Documentation, Similarity This motor is a Reliance Class V-1-106, 107 RH instead of Class B as men-tioned in the TER. Limitorques Report #'s 600456 and 80058 will ba referenced in the SCEW sheet.

13,15 Motorized Valve Actuator II.A ' Qualified Life, Radiation Additional Documentation for V-21,13, 17 Peerless Class B D.C. Motor is required. See Attachment II for JCO.

'14 Motorized Valve Actuator II.C Qualified Life Limltorque's Report 180058, V-5-167, 147 which will be referenced in the SCEW Sheet, addresses the qualified life of subject actuator.

16,17,18 Solenoid Valve II.C Qualified Life ASCO Report #AQR-67368/Rev. O, 19,20 .V-24-29 which ' vill be referenced in the 4 NS03A, B SCEW Si.eet, addresses the NSO4AL1, L2, L3 qualified life of subject valve..

NSO4BL1, L2, L3 o -

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ATTACHMENT 1 SPECIFIC EQUIPMENT EQ DEFICIENCIES 0.C.

TER NRC PRO POSED

- NO. COMPONENT CAT. DEFICIENCY SOLUTION 21 Solenoid Valve - 1.B Evaluation of Aging Degradation, Replaced Age Sensitive Non-V-38-9, 10, 16, 17 Qualified Life, Program to Identify Metallic Parts this outage.

Aging Degrada t ion

- 27 Solenoid Valve II.C Qualified Life lhe qualified life was covered

- V-26-16,-'18 in reference document, Wyle Report No. 17451-13.TER evalu-ation (on page 5f) does not' apply.

-22,23,24- Solenoid Valve 1.B Documentation Replace w/ Qualified Valve next 25,26,28 -V-23-13, 15, 16, 17, 18 outage.

31,35,36 19, 20,- 21, 22 V-27-1, 2, 3, 4 V-28-17, .18, 4 7 29,30,32 Solenoid Valve 1.B Documentation Replaced this outage with -

33,34 V-38-22, 23 Qua1ified ASCO Va1ve.

V-31-2.

V-24 V-22-1, 2,,28,- 29 V-11-34,--36 37 Solenoid Valve II.A Eva lua tion of Aging Degrada tion, Additional Documentation NR-108 A,B,C,D,E Qualified Life, Radiation and eva lua tion.

38,39,40 Pressure, Level, Flow 1.B Documenta t i on Replace with Qualified

- 41,42,43 Tra nsmi t te r Transmit ter next outage.

44,45 - ID46 A,B; ID45

- IP05'A,B,C,D; IAl2 ID13 A,B; IP07 IG06 A,B; RV26 A,B

IP03 A,B 1

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SPECIFIC EQUIPMENT EQ DEFICIENCIES 0.C.

TER NRC PR0l'OSED NO. COMPONENT CAT. DEFICIENCY SOLUTION 46,48,49 Pressure, Level Switch I.B Documentation Replace with Qualified 50,53,54 'RE23 A, B,C,D ; RE17 A, B ,C,0 Analog Trip System next -

55,56,57 REIS A,B,C,D ; RE03 A,B,C,D outage 58,60,61 REOS A,B; RE22 A,B,C,D,E

( 62,63 RE18 A,B,C,D; RE05-19 A,B l RE02 A, B, C,D ; RV46 A, B,C,D

( IA83 A,B,C,D,E; IP15 A,B,C,D IBil-Al,A2,B1,B2;

( IB05-Al,A2,B1,B2 I

47 Pressure Suitch II.A Evaluation of Aging Degradation, Environmental Qualification not PS-153 Qualified Life, Rad ia t ion required. The automatic bypass i design feature in the control I I logic for venting the drywell'

! on high N2 Pressure was elimi- l l

nated by replacing the existing l switch with a spring return-switch. This was done in order to ensure that deliberate oper-ator action is required to open the drywell vent valves after l

L the isolation signal has been I reset.

51,52 Pressure Switch .I.B Documentation Replace w/ Qualified Pressure RV29 A,B,C,D Switch next outage.

RV40 A,B,C,D 1 59 Level Switch 1.B -Documentation Replaced this outage w/Qualifici, RD08 A,B,C,D,E,F Level Switch.

64 - Limit Switch I.B Documentation Replaced this outage w/Qualifie'.

NSO4A-1,-2 Limit Switch.

NSO4B-1,-2 l

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Page 5 of 8 ATTACHMENT 1 SPECIFIC EQUIPMENT EQ DEFICIENCIES O.C.

TSR NRC PROPOSED NO. COMPONENT CAT. DEFICIENCY SOLUTION 65 Temperature Swit.ch II.A Similarity, Adequate Pressure, We referenced document, Wyle IB10 A thru P Adequate Steam Exposure Reports No. 17451-18 and No.

43854-1, provides all the neces-sary information which estab-lished the specific relationship between the tested switch model '

and installed unit.

In addition, recent Wyle letter (which will be referenced in the SCEW sheet)-outlined the applicable portions and photo-graphs from the referenced re-ports about the similarity of the switch and verified that subject switch's body and wirings were exposed to steam environment during the I.OCA testing. A re-evaluation of the pressure requirements in the Pain Steam 'Iunnel area, where IB-10E thru P switches are located, reveals that the pressure will be about 18.2 psia following accident con-dition. his information is found in EDS Report No. 02-0990-1085, Rev. O dated July 1981, wil] be referenced in the SCEW Sheet.

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Page 6 of 8 ATTACriMENT 1 ,

SPECIFIC EQUIPMENT EQ DEFICIENCIES O.C.

TER NRC PROPOSED NO. COMPONENT CAT. DEFICIE!CY SOLUTION 66 Temperature Switch I.B Documentation In 1979, Jersey Central Power IB06A,B,C,D and Light 03mpany (JCPEL) per-formed a high energy line break (HELB) analysis of the Oyster Creek Emergency (bndenser System (ECS) piping outside containment and concluded that a pipe break could cause damage to the ECS isolation valves and controls. JCP&L provided thia conclusion to the NRC. The NRC (SEP Branch) performed an on-site inspection as a part of the evaluation of the SEP Tapic No. III-5B, confirmed JCP&L's findings, and requested JCP&L to provide adequate protection against the effects of a postulated HELB. GPUN (now licensed operator for the OCNGS) performed and submitted an analysis to demonstrate that the ECS piping would leak before a significaat break could occur. During the integrated assessment of the SEP topics, the NRC staff accepted the result of the analysis and requested GPUN to provide leak detection capability (i.e., visual inspections or automated devices). GPUN is currently evaluating the type of the capability to be utilized.

Since GPUN will provide improved capability of an early ECS leak detection, the subject temperature switch is no longer needed.

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A'RTACHMEft' 1 SPECIFIC EQUIPMElff EQ DEFICIENCIES 0.C.

TER NRC PROPOSED NO. COMPONENT CAT. DEFICIENCY SOLUTION 67 Electric Motor II.C Qualified Life Inbrication changeout is NZ01-A,B,C,D covered by Plant PM program.

70 Electric Motor II.A Steam Exposure Motors are located inside the 1-1 corner rocus. Tnere are no 1-2 steam lines in corner rooms.

1-3 These motors will not be ex-1-4 posed to steam environment.

TER evaluation (page SF) does not apply.

71 Terminal Block I.B Documentation, Evaluation of Aging Replaced with Qualified Block TB#'s63-242 Degradation, Qualified Life, steam this outage.63-246 Exposure, Spray, Radiation 22-389 22-390 71-412 22-640 22-641 72 Electrical Penetration II.A Aging Degradation, Qualified Life Additional Analysis and X10, X13, X18 Spray Radiation documentation.

73,74 Motor Control Centers II.A Documentation pdditional Analysis.

DC-1, lA21B, LAB 2, 1821A, 1821B ,

75 fetor (bntrol Center II.A Documentation This equipment is located DC-2. in floor El. 75' in about the same area where the temperature switches under TER No, 66 are located. Therefore, the pro-posed solution described under TER No. 66 also applies.

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ATTACHMENT 1 SPECIFIC EQUIPMENT EQ DEFICIENCIES 0.C.

NRC PROPOSED '

TER

' CAT. DEFICIENCY SOLUTION NO. COMPONENT II.A Documentation The Rockbestos Report No.

80 Electrical Cable Rockbestos Firewall EP QA 1804 will be referenced in the SCEW sheet. ,

i II.A Locumentation, Similarity, Aging The E.I. Dupont Report dated j 81 Electrical Cable Tensolite Deg ra da t i on, Qualified Life, Radiation August 1, 1974, " Tests of Elec-tric Cables Insulated and '

Jacketed With TEFZEL 280 Flouropolyuer Under IEEE 383-1974," will be referenced in the SCEW Sheet.

Documentation Franklins Report d's F-C2770

, 82 ,

Elect rical Cable II.A dated October 1970 and F-C2737 t Kerite dated April 15, 1970 will be l referenced in the SCEW Sheet.

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N/A Temperature Switch N/A See Attachment II for JCO.

IP18A IP18B 4

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(014 7M)

ATTACINEN T II ~l I

JUSTIFICATION OF WNTINUED OPERATION TER NO .

V-17-3 5 Shutdown Cooling Valves V-17-5 6 V-17-57

'lhese valves are installed a t the discharge side of the shutdown heat exchangers respectively. If these valves should fail, the alternate shutdown cooling me thad is to direc t the flew to the main condenser by way of the cleante system lytdown and maintaining reactor water level using the ccadensate system via the feedwater string (s) to makeup to the reactor vessel, Currently, anotiwr shutdown cooling method is being developed whi ch will include the Elec tromatic Relie f Valve s. If this me thod becanes official and supersedes the preceding, we chall advise accordingly.

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TER No. 13, 15 l V-21-13 Containment Spray Test Valves ,

V-21-17 j These are the Containment Spray System dynamic test valves under each system loop respectively. They are normally closed and are used for full-flow '

testing of the Containment Spray System without wetting the drywell. The flow from these valves is directed to the suppression chamber through connections into the vacuum breaker line.

The operating circuits of these valves are interlocked so that both loops can not be in the test position at the same time. .Each loop has the heat removal capacity to hold primary containment pressure below the design

, pressure of 62 psig, and to reduce the pressure to essentially atmospheric within about 8 days following the accident. Also, the control circuitry for the containment spray system provides automatic reset from dynamic test to automatic start readiness when process conditions indicate impending need for the containment spray.

If any of the test valves fail to close while the containment spray system is on, the net flow of that loop will be reduced by about 20%. However, the flow of one containment spray loop in either loop is more than ample to provide the necessary heat removal capacity.

(0147M)

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ADDITIONAL ITIM (not covered by TER)

I P A Contalment Spray Temperature Switches IP-18-B he containment spray system consists of two independent full-capacity cooling loops. Ei ther loop is capable of removing the fission product decay heat fro:n the contalment.

he containment spray temperature switches, located outside the Drywell a t dif ferent locations, are used to sense the water temperature from the hea t exdiangers in loops I and II respectively, as a failure monitor. Failure of the temperature switch of start loop will autanatically disable the contaiment spray pump of that loop. We contairinent spray pumps of the standby loop can be manually started using their respective controls'in the Control Roan.

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ADDITIONAL ITEM (Not covered by TER)

IUREG 0737 Item II.D.3

Direct Indication of Relief and Safety Valve Position 2

I j' These acoustic monitoring systems were operational at the OCNGS prior to this IUREG requirement which includes environmental qualification. The environmental gitalification will be established by the B&W Owner's Group Test Program, in which the OCNGS is a participant.

The testing has been concluded and the final report and recommendations are scheduled for release in the first quarter of 1984. The results of the

, testing show that the basic components are satisfactory for in containment IDCA and MSLB. However, some additional protection of the driver / amplifier and connections to it probably will be required at the OCNGS.

4 For the relief valves there are also position indicating switches with Control Room displays. For the safety valves there are tenperature readouts t

In the Reactor Building which indicate if a safety valve has lifted.

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ADDITIONAL ITEM (not covered by TER)

This is in addition to previously submitted JCO No. 51 - Torus Vacuum Relief System.

a) IP-12 Pressure Transmitters PT-52 Wese transmitters are physically located in different locations outside the Drywell. The cables and conduits run separately, and power supply to these transmitters comes from two different sources.

b) DPS-66A Pressure Switch DPS-66B 2ese switches control vacuum breaker valves V-26-16 and V-26-18 respectively to prevent excessiva vacuum in the suppression chartber.

If both switches should fail under a single failure situation, the Control Room Operator can monitor the pressure in the suppression chamber with the Torus Pressure indicatiaa device (IP13) and operate the vacuum breaker valves manually with the Control Switch at the llF panel.

ATTACHMENT III TMI ACT70N PLAN, NURF 0737 ITEMS NUREG 0737 Item TI.B.3 Post Accident Sampling Capability Of the four solenoid valves :equired, one, V-40-6, has been installed, but is not operable. The remaining valves will be installed consirtent with our post accident system operational comitment.. These four valves are environmentally qualified and SCEW Gnesta will be submitted.

NUREG 0737 Item II.D.3 Direct Indication of Relief and Safety Valve Position The environmental qualification will be established by the B&W Owner's Group Test Program, in which OCNGS is a participant.

The testing has been concluded and the final report and recomendation are scheduled for release in the first quarter of 1984. The re.3ults of the testing show that the basic components are satisfactory for in containment IOCA and MSLB. However, some additional protection of the driver / amplifier and connections to it probably will be required at the OCNGS.

NUREG 0737 Item II.E.4.1

< Dedicated Hydrogen Pet.etrations GPUN required exemption of this item. Refer to the letter dated December 15, 1902.

NUREG 0737 Item II.F.1 Subparts (1) through (6)

Additional Accident Monitoring Instrumentation Rese items will be operable by the end of the Cycle 10 refuel 4ng outage.

They are fully qualified and the applicable Sch:W sheets will be submitted.

NUREG 0737 Item II.K.3.19 Interlock on Recirculation Pump toops mis item will be installed during Cycle 11 refueling outage.

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