ML20205T533

Application for Amend to License NPF-57,revising Tech Spec 4.4.3 to Extend Specified Valve Local Leak Rate Tests Until First Refueling Outage Scheduled to Begin on 880201.Fee Paid

ML20205T533
Person / Time
Site:	Hope Creek
Issue date:	04/03/1987
From:	Corbin McNeil Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20205T535	List: ML20205T533 ML20205T541
References
LCR-87-06, LCR-87-6, NLR-N87047, NUDOCS 8704070316
Download: ML20205T533 (12)

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e Public Service Electric and Gas Cornpany Cctbin A. McNeill, Jr. Public Service Electric and Gas company P.O. Box 236. Hancocks Bridge, NJ 08038 609 339-4800 Vice President -

N uclear April 3, 1987 NLR-N87047 LCR 87-06 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REQUEST FOP AMENDMENT FACILITY OPERATING LICENSE NPF-57 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 ,

In accordance with the requirements of 10 CFR 50.90, Public Service Electric and Gas Company (PSE&G) hereby transmits a request for amendment of Facility Operating License NPF-57 for Hope Creek Generating Station (HCGS). In accordance with the requirements of 10 CPR 170.21, a check in the amount of $150.00 is enclosed. In accordance with the requirements of 10 CFR 50.91(b) (1), a copy of this request for amendment has been sent to the State of New Jersey as indicated below.

This amendment requests revision of Technical Specification Section 4.4.3 to extend specified valve local leak rate tests (LLRTs) until the first refueling outage, currently scheduled to begin February 1, 1988. Currently these tests will become overdue beginning June 11, 1987. A request for exemption from the Appendix J requirements regarding LLRT requirements is being submitted concurrently under separate cover. Depending on future operating conditions, additional extensions may be requested covering other types of surveillances and inspections.

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This submittal i'ncludes.three (3) signed originals and forty.(40) copies. .. ,

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Should you have'any questions on the subject transmittal, please don't hesitate to contact us.

Sincerely,

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8704070316 870403 -'~~ 4 PDR ADOCK 05000354 P PDR 00)

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Document Control Desk 2 4/3/87 C Mr. D. H. Wagner USNRC Licensing Project Manager Mr. R. W. Borchardt USNRC Senior Resident Inspector Director, Bureau of Radiation Protection Department of Environmental Protection 380 Scotch Road Trenton, NJ 08628 i

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Ref: LCR 87-06 STATE OF NEW JERSEY )

) SS.

COUNTY OF SALEM )

Corbin A. McNeill, Jr., being duly sworn according to law deposes and says:

I am Vice President of Public Service Electric and Gas Company, and as such, I find the matters set forth in our letter dated April 3, 1987 concerning Facility Operating License NPF-57, Hope Creek Generating Station, is true to the best of my knowledge, information and belief, k ~

Subscribed and Sworn o before me this J d day of U , 1987

/

bARJ A DELORtSILIMegg Notary Public of New Jersey A Notary Pubbe of NewW

" U"W My Commission expires on 94

-a_-

LCR 87-06 Pcga 1 of 4 Description of Change This proposed amendment to the Hope Creek Generating Station Technical Specification requests a one time extension to the local leak rate testing intervals for certain primary containment isolation valves specified in Technical Specifications 4.4.3.2.2.a and 4.6.1.2.d (Attachment 3).

Additionally, this proposal requests, relief from Section XI of the ASME code per 10 CFR 50, 50.55a(g) and its requirements to leak rate test Category A valves not less than once every two years. Although the ASME code test interval is larger than Technical Specification 4.4.3.2.2 interval noted below, the Section XI interval will be exceeded if this request is granted.

A request for exemption from the requirements of 10CFR50, Appendix J, Section III.D.3 is being submitted under separate cover to support this LCR.

Justification for Chance Technical Specification 4.4.3.2.2.A requires local leak rate tests for Reactor Coolant System Pressure Isolation valves listed in Table 3.4.3.2-1 to be leak tested at least once per 18 months. Past NRC reviews for Hope Creek have allowed Type "C" gas tests at 18-month intervals to be performed in lieu of water tests to satisfy these requirements (NUREG 1048 Supplement 5 Section 3.9.6). In addition to satisfying 4.4.3.3.2.a requirements these tests would also satisfy the l

Type "C" test requirements for these valves per Section 4.6.1.2.d. An additional 25 percent may be added to the 18-month interval per Specification 4.0.2.a. The valves for which this relief is requested are listed in Attachment 1 and are identified by having an overdue date entered in the 4.4.3.2.2.a (18-month) column. It should be noted that these dates are 18 months plus 25 percent dates.

l Technical Specification 4.6.1.2.d requires local leak rate tests (Type C tests) on the primary containment isolation valves listed in Table 3.6.3-1 to be performed at intervals no greater than 24 months except for containment isolation valves in hydrostatically tested lines penetrating the primary containment, which shall be leak tested at least once per 18 months. The Commission's Regulations (10 CFR 50, Appendix J, Section III.D.3) require local leak test (Type C tests) to be performed during each reactor shutdown for refueling, but in no case at intervals greater than two years.

The end of the initial 18-month and 24-month testing intervals for some of the Hope Creek Generating Station containment isolation valves is approaching. Type C tests have been performed within the test interval for those valves that did not require plant shutdown for testing. However, in order to

LCR 87-06 Pcgn 2 of 4 satisfy the test interval requirements for 12 tests covering twenty-seven valves (see Attachment 1), it would be necessary to shut down the plant for approximately two weeksprior to the first refueling outage solely for this purpose.

A containment entry is required to test the valves that cannot be tested at power. Testing of these valves at power poses a personnel hazard due to ALARA considerations, high ambient temperatures existing within containment and entry into the inerted containment. Additional restraints to testing some of the valves at power include the need to drain the "A" and "B" RHR loops, as well as the "A" core spray loop. For specific descriptions by valve, refer to Attachment 2.

The requested license change is' temporary and became necessary as a result of delays in attaining full power operation common to initial startup activities.

The current schedule is for the first refueling outage to begin approximately February 1, 1988. Testing for these valves will be performed prior to startup from this outage.

A two week outage required to perform this testing prior to the refueling outage would result in a net increase in overall outage time and would subject the plant equipment and systems to the detrimental effects inherent in an additional shutdown and startup operation.

Therefore, PSE&G requests an extension of the Type C test interval and certain 18-month leak tests for the specified primary containment isolation valves listed in Table 3.6.3-1, Part A, B, and C that require the plant to be shutdown for testing. PSE&G also requests relief from ASME Section XI per Section 50.55a(g) for leak rate testing interval requirements for the applicable Category A valves (see Attachment 1). The proposed change as shown on enclosed Technical Specification pages 3/4 4-12 and 3/4 6-4 (Attachment 3) would extend the test interval for these valves into the first refueling outage.

Additionally, it is requested the requirements of Technical Specification Section 4.0.2 be waived for this one time license change. For additional information, see the associated Appendix J Exemption Request.

Significant Hazards Consideration Determination '

The Commission'had provided guidance concerning the application of standards in 10 CFR 50.92 for determining whether license amendments involve significant hazards consideration by providing certain examples which were published in Federal Register on April 6, l 1983 (48 FR 14870). One of the examples j (vi) of an action involving no significant hazards J

consideration is a change which may in some way reduce a safety margin, but where the results of the change are clearly within )

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'LCR'87-06 Paga 3~of.4:

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'all acceptable criteria. The foregoing requested _ change _and exemption fits this example. Postponing:the aforementioned local leak rate tests until.the end of.the first. refueling outage'would allow.for continued operation of the plant and would have little or no~effect_on containment integrity as l g discussed below. 1

-(1) The containment 1 isolation valves listed on Attachment 1 were all_ tested.successfully in late

, 1985. The_ total of the measured Type C. leakage

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rates for'these valves is not a significant j portion (4.9%) of the. allowable-leakage limit .( 0 .' 6 Ls). Deterioration in the overall. integrity of  ;

the containment penetration'is normally;a gradual process.

f NUREG/CR 4330, Review of Light Water Reactor.

2 Regulatory-requirements, has shown-that i

containment leakage is'a relatively. minor.

contributor to overall plant risk. In addition inherent BWR design features will maintain the offsite doses below 10CFR100 limits even with

leakage above current limits.

1 (2) The intent of the Technical-Specifications leak rate testing intervals for pressure isolation

valves and containment isolation valves is to 4

require testing of the isolation valves once every.

fuel cycle. A. normal reactor fuel. load is designed to provide an 18-month cycle with approximately 16 months of full power operations. .

Consequently, the primary containment-isolation. _

valves are normally exposed to 18 months of rated temperature conditions.between each Type C_ test.

Since the initial Type C tests-at the Hope Creek

, Generating Station, these valves.will have been I subjected to rated temperature _ conditions l approximately equal to one 18-month operating cycle by the first refueling outage. An extension ~

of the Type C test interval ~to the first refueling

outage is not unrealistic when the intent of the Type "C" test schedule specified by the Technical.

, Specifications and Appendix.J is considered.

! - a For these reasons, the proposed temporary amendment to'the Hope a Creek Operating License does not constitute'a significant

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hazards consideration in thatiit would.not e 1. Involve a significant increase in the probability or

consequences of an accident previously' evaluated since.

i the extension of the surveillance intervals.is i

consistent with the intent of_the regulations, when

. considering the operating conditions to which the subject valves have been exposed; or

3 LCR 87-06 Pcg3 4 of 4.

2. Create the possibility of a new type of accident or a different kind of accident:from any accident previously analyzed in that current analyses assume certain values of containment leakager. therefore, new accident scenarios are not credible based upon scheduling of this testing alone; or

3. Involve a significant reduction in the margin of safety because, based on initial LLRT results, these valves have exhibited a high degree of leak tight reliability. Additionally, the valves'will be exposed to operating conditions consistent with those normally experienced between testing intervals.

The requested amendment concerns schedular relief for surveillance testing of a limited number of containment isolation valves and will not result in a significant change in the amounts or types of effluents that may be released offsite.

There will be no significant increase in individual or cumulative occupational radiation exposure as a result of the requested amendment which merely requests to delay testing.

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Attachment 1 18-Month 24-Month 4.4.3.2.2.a 4.6.1.2.d TECH. SPEC.REF.

Overdue Overdue Valve No. Date Date Table 3.6.3-1 i

1) BC-V014 HV-F050B 6/23/87 Part C, Group 39-F

2) BC-V013 HV-F015B 6/23/87 A, 3-E

3) BC-Vll8 HV-F122B 6/23/87 B, 26-F

4) BC-V015 HV-F027B_ 8/12/87 A, 3-B

5) BC-Vll3 HV-F017A 7/16/87 B, 26-A

6) BC-Vll4 HV-F041A 7/16/87 C, 39-E

7) BC-V119 HV-F146A 7/16/87

8) BC-V110 HV-F015A 7/25/87 A, 3-E

9) BC-Vlll HV-F050A 7/25/87 C, 39-F

10) BC-Vil7 HV-F122A 7/25/87 B, 26-F

11) BC-V116 HV-F021A 6/11/87 B, 26-B

, 12) BE-V003 HV-F005B 8/6/87 B, 25-A

13) BE-V002 HV-F006B 8/6/87 A, 10-B

14) BE-V072 HV-F039B 8/6/87 - B, 25-D

15) BE-V007 HV-F005A 8/5/87 B, 25-A

16) BE-V006 HV-F006A 8/5/87 A, 10-B

17) BE-V071 HV-F039A 8/5/87 B, 25-D

18) BE-V001 HV-F006 8/5/87 B, 22-A

19) KL-V026 HV-5152B 8/7/87- A, 15-A i; 20) BC-V021 HV-F023 9/3/87 A, 3-D j 21) BC-V020 HV-F022 9/3/87 A, 3-D l 22) FD-V001 HV-F002 9/6/87 A, 5-A

23) FD-V051 HV-F100 9/6/87 A, 5-A

24) FD-V002 HV-F003 9/6/87 A, S'-A

25) FC-V001 HV-F007 9/18/87 A, 6-A

26) FC-V048 HV-F076 .

9/18/8'7 A, 6-A

27) FC-V002 HV-F008 9/18/87 A, 6-A 11

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ATTACHMENT 2 SPECIFIC DESCRIPTIONS Items 1-3 "B" Shutdown Cooling Return to "B" Recirc Loop, PEN 4A

1) Valve BC-V014, Stop Check 12" (HV-F0508)

2) Valve BC-V013, Globe 12" (HV-F015B)

3) Valve BC-Vil8, Globe 2" (HV-F122B)

Original tests.were performed on 8/8/85. Maximum leakage was 18 SCCM(1). These valves would require an extension of 32 weeks

• beyond the 18-month (+25 percent) - P.I.V. test interval.

Technical Specification Air Tests would require containment entry, RHR B Loop isolation and drainage.

Item 4 Torus Spray Supply, PEN 214A

4) Valve BC-V015, Gate 6" (HV-F027B)

Original test date was 8/7/85 with leakage of 296 SCCM(1).

This valve would require an extension of 25 weeks

• beyond the 24-month Type "C" test interval. This test would be performed with the test for Items 1 through 3, and would require isolation and drainage of the "B" RHR Loop.

l Items 5-7 "A" LPCI Injection, PEN P6C s

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5) BC-V113, Gate 12" (HV-F017A)

6) BC-Vil4, Stop Check 12", (HV-F041A)

7) BC-Vil7, Globe 2", (HV-F146A)

Based on assumed February 1, 1988 outage start date i

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Original tests were performed on 7/9/85 and 7/29/85. Maximum leakage was 305 SCCM(1). These valves would require a 29-week

• extension beyond the 18 month (+25 percent) P.I.V. test interval. Test Specification Air Tests would require containment entry, isolation and drainage of the RHR "A" Loop.

Items 8-10 "A" Shutdown Cooling Return to "A" Recirc Loop, PEN P4B

8) BC-V110, Globe 12", (HV-F015A)

9) BC-Vlll, Stop Check 12", (HV-F050A)

10) BC-V117, Globe 2", (HV-F122A)

Original tests were performed on 8/14/85 and 9/3/85. Maximum leakage was 165 SCCM(1). These valves would require a 28 week

• extension beyond the 18-month (+25 percent) P.I.V. test interval.

Technical Specification Air Tests would require containment entry, and isolation and drainage of the "A" RHR Loop.

Item 11 RHR "A" Loop Containment Spray Supply, PEN P24B

11) BC-Vll6, 16" Gate, (HV-F021A)

Original test date was 6/7/85 with maximum leakage of 880 SCCM(1). This valve would require a 34-week

• extension to the 24-month Type "C" test interval. Technical Specification Air Tests would require isolation and drainage of the "A" RHR Loop.

It is intended that this test be performed along with Items 8-10.

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Items 12-14 "B" Core Spray Injection, PEN'P5A 12)- BE-V003, Gate 12", (HV-F005B).

- 13) BE-V002, Stop Check 12", (HV-F006B)

14) BE-V072, 2" Globe, (HV-F0398)

Original test date was 9/10/85 with maximum leakage 920 SCCM(1). These valves would require a 26-week *~ extension to the 18-month.(+25 percent) P.I.V. test ~ interval. Technical-Specification Air Tests would require containment entry, drainage and isolation or Core Spray Loop "B."

Items 15-18 "A" Core Spray Injection, PEN P5B

15) BE-V007, Gate 12", (HV-F005A)

16) BE-V006, Stop Check 12", (HV-F006A)

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17) BE-V071, Globe 2", (HV-F039A)

18) BE-V001, Gate 14", (HV-F006) i l Original test date was 9/17/85 with maximum leakage of 902.4 SCCM(1). These valves would require a 26-week

• extension.to the ,

18-month (+25 percent) P.I.V. test interval. Technical Specification Air Tests would require containment entry, drainage and isolation of Core Spray Loop "A."

- Item 19 "B" Instrument Gas Header, PEN P28A f

19) , KL-V026, 2" Globe, (HV-SIS 2B), ,

Original test date was 8/2/85 with maximum leakage of 24 j SCCM(1). This valve would require a 26-week

• extension to the 24-month Type "C" test interval. Technical Specification Air

, Tests would require containment entry and loop isolation.

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Item 20-21 Hnad Sprcy

20) BC-V0211, 6" Gate, (HV-F023)

21) BC-V020, 6" Globe (HV-F022)

Orignial test date was 10/12/87 with maximum leakage of 27 SCCM(1). These valves would require a 22-week

• extension to the 18-month (+25 percent) P.I.V. test interval. Technical Specification Air Tests would require plant shutdown to remove the head spray piping spool above the seal plate so that a blind flange could be installed to permit testing. The reactor well plugs and the drywell head would also have to be removed. This test can only be performed during a refueling outage.

Items 22-24 HPCI Steam Supply

22) FD-V001, 12" Gate, (HV-F002)

23) FD-V051, 2" Globe, (HV-F100)

24) FD-V002, 12" Gate, (HV-F003)

Original tests were performed on 9/1/85 with maximum leakage of 192.1 SCCM(1). These valves would require a 21-week

• extension to the Type "C" test interval. Testing would require containment entry.

Item 25-27 RCIC Steam Supply

25) FC-V001, 4" Gate, (HV-F007) )

26) FC-V048, 2" Globe, (HV-F076)

27) FC-V002, 4" Gate (HV-F008)d

'. Original tests were performed on 9/13/85 with maximum leakage of 8.6 SCCM(1). These valves would require a 20-week

• extension to the Type "C" test interval. Testing would require containment entry.

Reference (1) Public Service Electric & Gas Company, Primary Reactor Containment Integrated Leakage Rate Test for the Hope Creek Generating Station (Final Report March 1986).

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