ML20216A857

Forwards Request for Exemptions from 10CFR50,App J,Option,As Documented in NRC Reg Guide 1.163, Performance Based Containment Leak Test Program, Dtd Sept 1995

ML20216A857
Person / Time
Site:	Oyster Creek
Issue date:	03/31/1998
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-REGGD-01.163, RTR-REGGD-1.163 1940-98-20179, NUDOCS 9804130307
Download: ML20216A857 (8)

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{ GPU Nuciear. Inc.

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U.S. Route #9 South NUCLEAR

• Post Office Box 388 Forked River, NJ 087310388 Tel 609-9714300 March 31,1998 1940-98-20179 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington DC 20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Request for Exceptions to Appendix J Testing By letter dated September 3,1997, the USNRC issued Amendment 186 to the Operating License for the Oyster Creek Nuclear Generating Station. Amendment 186 approved the use of Option B to Appendix J of 10 CFR 50 at the Oyster Creek Nuclear Generating Station.

As part of an ongoing program to upgrade the previous testing requirements and implement the new program, two areas needing exception from Reg Guide 1.163 " Performance Based Containment Leak Test Program", dated September 1995, have been identified. Attachment I to this cover contains the request for exceptions.

If any additional information or assistance is required, please contact Mr. John Rogers of my i staff at 609.971.4893.

' I l l Very truly yours, j

} ($ \ l Michael B. Roche

, Vice President and Director

_: L c rb Oyster Creek MBR/JJR cc: Administrator, Region I NRC Project Manager Senior Resident Inspector 9804130307 980331 PDR ADOCK 05000219 P PDR..

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l Attachment I l

Request for Exceptions from Regulatory Guide 1.163 Request No.1 Requirement:

The requirements for implementation of 10 CFR 50, Appendix J, Option B, were documented in Regulatory Guide 1.163, " Performance Based Containment Leak Test Program", Revision 0, dated September 1995. This Regulatory Guide endorsed Nuclear Energy Institute document NEI 9.4-01, " Industry Guideline for Implementing

. Ferformance Based Option of 10 CFR 50, Appendix J" Revision 0, dated July 26,1996, I

and ANSI /ANS 56.8-1994, " Containment System l_eakage Testing Requirements".

l 10 CFR 50, Appendix J, Sections II.F and II.H define in part:

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"' Type A Tests' means tests intended to measure the primary reactor containment overall leakage rate..." l

"' Type C Tests' means tests intended to measure containment isolation i valve leakage rates. The containment isolation valves included are those l

that.. 2. Are required to close automatically upon receipt of a containment isolation signal in response to controls intended to effect containment isolation:..."

Rcquested Exception The following valves are in the Shutdown Cooling (SDC) System:

V-17-19 Inboard Supply Isolation Valve; Automatic closure Containment Isolation Valve V-17-54 Inboard Return Isolation Valve; Automatic closure Containment Isolation Valve V-17-1,2,3 Outboard Supply Isolation Valves; Manual electric control, No Automatic features V-17-55,56,57 Outboard Return Isolation iaivec- Manual electric control No Automatic features An exception is requested to not perform Type C testing on these eight valves.

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1940-98-20179 Attachment I Page 2 of 7 System Configuration 4

The SDC is a closed loop system. The entire system is rated for full primary pressure  !

(1250 psig). The portion of the system inboard of the outboard isolation valves is rated j to full primary temperature (575 F), and is Seismic Class I as defined in the design l

bases documented in the current UFSAR. The portion of the system between the ]

outboard isolation valves is rated for 350 F, and is rated Seismic Class II. Both the I Supply and Return lines for the SDC system connect to the E Recirculation Loop, inboard of the loop isolation valves. The inboard isolation valves are non-isolable from the reactor vessel.

The supply line is a 14" pipe connected to the E Recirculation Loop Discharge Line at approximate elevation 42'6". It proceeds through the drywell, through inboard isolation valve V-17-19 at the same elevation, and exits at a penetration at approximate elevation i 41'10". Outside of the drywell, the 14 inch header rises to approximate elevation l 45'10" before descending to approximate elevation 39'. The header then splits into three 10" feeders which all loop upwards to approximate elevation 46'2" where outboard isolation valves V-17-1,2, and 3 are located.

The return lines are three 8" pipes outside of the drywell at approximate elevation  ;

52'10" and contain outboard return isolation valves V-17-55,56, and 57. The three

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lines ascend t;pward to rejoin into a 14" header at approximate elevation 72'. The header then descends to a penetration at approximate 58'1". The pipe continues inside contaito.nt, through inboard isolation valve V-17-19 at approximate elevation 41',

to connat to the E Recirculation Loop Suction Line at approximate elevation 28' 6". -

The inboard isolation valves automatically close on a containment isolation signal.

However, they are also interlocked to close (or remain closed) above 350 F. The outboard isolation valves do not automatically close on a containment isolation signal.

The SDC design does not include a simultaneous design basis Loss of Coolant Accident (LOCA) and design basis seismic event.

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. Historical Review and Basis for Relief l

None of these eight valves has ever received an Appendix J Type C test. The SDC system was designed prior to the issuance of the 10 CFR 50, Appendix A, General Design Criteria. At the time of the original design and licensing, the two mboard valves were considered to be " containment isolation valves", while the six outboard valves were considered to be " valves with isolation provisions".

By letter dated November 22,1978, Jersey Central Power and Light (a predecessor to GPU Nuclear, Inc.) stated that the six outboard valves were not being tested and that no exemption from testing was necessary. Additionally, a request for an exemption from testing the inboard two valves was submitted.

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1940-98-20179 Attachment i Page 3 of 7 Exemptions for these valves were addressed in both the discussion on Type A and Type C testing. The Type A requirement stated:

"Those portions of fluid systems that are part of the reactor coolant pressure boundary and are open directly to the containment atmosphere under post accident conditions and become an extension of the boundary of the containment shall be opened or vented to the containment atmosphere prior to and during the test.. .All l

vented systems shall be drained of water or other fluids to the extent necessary to assure exposure of the system containment isolation valves to containment air test pressure and to assure they will be subjected to the post accident differential pressure."

However, the same requirement contained the pre-approved exemption:

" Systems that are required to maintain the plant in a safe condition during the test shall be operable in their normal mode, and need not be vented."

The SDC system is regetred to cool the reactor during shutdown conditions, therefore, the purpose of the ensuing discussiori was to request the relief from venting a system required to be operable during the Type A test. The letter dated November 22,1978, requested an exemption correctly stating that:

"This system (SDC) is in service during the Type A test to provide decay heat removal. No type C testing is performed since this system does not contribute to post accident leakage. The system isolation valves are closed during operation. In

! cddition, the system is closed outside containment...and is protected from the

) effects of pipe ruptures. We request an exemption from the requirement for l venting and draining this system."

Unfortunately, the same exemption contained a statement which was not correct. The exemption request stated that the SDC system was Seismic Class I. The SDC system is Seismic Class I from the reactor vessel to the outboard isolation valves, as defined in the design bases documented in the current UFSAR. The piping between the outboard isolation valves is Seismic Class II. The system piping, supports and components of the pressure boundary were analyzed to FSAR and NUREG-0800 criteria for an SSE event at a conservative operating temperature of 225 F, and were found to meet the criteria for maintaining of the integrity of the pressure boundary with a few exceptions. Those exceptions were reviewed against the guidance provided in Generic Letter 91-18 and found to meet operability criteria. To bring the system into full compliance with FSAR and NUREG-0800 criteria, some support modifications will be installed. As the radiation levels during system operation (outages) are unacceptably high for system modifications, it is planned to complete the modifications during the present or next operating cycle. This commitment will be specified in the Integrated Schedule.

.. .i 1940-98-20179 Attachment 1 Page 4 of 7 The calculations and analysis supporting this evaluation are in the process of design verification. And will complete by April 30,1998. By this letter, GPU Nuclear dockets the correction of the 1978 error.

By letter dated March 4,1982, the NRC docketed the approval of the Appendix J program for the Oyster Creek Nuclear Generating Station. The NRC safety evaluation was based on contractor report dated March 21,1981. That report, in Section 3.1.1.3, exempts Oyster Creek Nuclear Generating Station from draining the SDC during a Type A test, based on the evaluation in Section 3.1.4.1 for Type C testing.

Section 3.1.4.1 addressed the request to exempt V-17-19, and 54 (the inboard isolation valves). The report identifies the basis for the exemption request as:

...the system is not expected to be a source of post-accident leakage. The valves i are closed during power operation. The system is closed outside containment, is seismically designed, and is protected from the effect of pipe rupture "

GPU Nuclear has evaluated this statement, including the consideration that the portion of . I the system between the outboard isolation valves is not Seismic Class I. The inboard valves are automatic containment isolation valves and are interlocked closed above 350 F. The system is closed outside containment and is protected from the effects of pipe rupture. If a seismic event were to occur and cause damage to the SDC outside of the outboard valves, the results would be non significant, as the sune seismic event could not cause a LOCA. Alternatively, if a LOCA were to occur, there would be no release path through the SDC system. Therefore, although the original exemption included an incorrect assumption, the correction of that error does not affect the previously granted exemption.

The next paragraph of the contractor report states that:

"Since the shutdown cooling system connects to the reactor coolant pressure

' boundary through the reactor recirculation system, these valves will remain water covered throughout the post-accident period by the water level in the reactor vessel..."

i This statement did not include one possible scenario. If the LOCA were to occur from  !

the E Recirculation loop, then the inboard side of the inboard isolation valves could become exposed to containment atmosphere. However, as the entire SDC system is closed outside of containment, there is no containment-to-environment leak path.

Additionally, the elevations of the inboard and outboard valves ensure that the outboard valves will remain water covered to provide additional isolation. Therefore, the new j scenario does not introduce a new release pathway. I

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.. - . . 1 1940-98-20179 Attachment I Page 5 of 7 Justification for Exception 1 GPU Nuclear has determined that the request for exception from the Type A draining requirements and Type C testing requirements for the eight identified valves would be acceptable based on the following:onsiderations:

1. The SDC system is a closed loop system and does not provide a containment- i to-environment release path.

2. The SDC system is required to be in operation for core cooling during a Type A containment leak rate test.

3. The system is protected from the effects of pipe rupture.

4. The portion of the system from the recirculation lines to the outboard isolation valves is Seismic Class I as defmed in the design bases documented in the current UFSAR.

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5. The inboard isolation valves are interlocked such that they will not open when the reactor vessel is above 350 F. This relates to a system pressure of approximately 150 psig. At these temperatures and pressures, any leakage volume outside of containment would be bounded by the design basis LOCA described in the FSAR. Additionally, as the SDC is not placed into operation for a significant time after plant shutdown (and cooldown), the source term L would be greatly reduced. . Therefore, the offsite release would be bounded by l the existing FSAR calculations.

6. Any leakage which could occur between the outocard SDC valves with the system in operation would be annunciated in the Control Room by either a i high temperature alarm or a high radiation alarm. Additionally, the decrease )

in inventory would cause a decrease in reactor water level. Emergency Operating Procedures use these parameters as indicators of a primary leak and direct the closing of all eight valves. However, even in this scenario, there is no containment-to-environment leak path. Therefore, no 10 CFR 50, Appendix J requirements should apply.

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1940-98-20179 Attachment 1 Page 6 of 7 R'equest No. 2 Requirement 10 CFR 50 Appendix J, Section II.G requires in part:

"' Type B Tests' means tests intended to detect local leaks and to measure leakage across each pressure-containing or leakage-limiting boundary for the following primary reactor containment penetrations...l. Containment penetrations whose l design incorporates... piping fitted with expansion bellows..." 1 Reauested Excention The feedwater piping penetrates the primary containment using expansion bellows for limiting leakage. The penetrations are a single bellows design and are, therefore, untestable. An exception is requested to not test tne feedwater bellows.

System Configuration Two separate 18" feedwater lines penetrate the containment at approximate elevation 33'. To allow for thermal expansion and contraction of the feedwater lines, an expansion bellows was installed at the penetration during original construction. The drywell wall at the penetration is curved, and the piping is not concentric to the drywell opening.

Historical Review and Basis for Relief  !

The bellows in the feedwater system in this exception request are identical to the bellows l in the steam system. By letter dated August 12,1976, Jersey Central Power and Light  !

j requested an exemption from the Type B testing requirements for the steamline bellows.

The basis for the exemption was that: 1) the bellows were non-testable; 2) experience had shown that there was no significant leakage through this type of penetration; and I

3) the Type A test would provide acceptable leak rate testing. The NRC approved this i exemption by letter dated December 3,1976.  ;

The two feedwater lines and two steamlines all penetrate the containment in close proximity to each other. It is currently believed that the 1976 exemption request for the steamline bellows was intended to be a generic request for all four bellows. This letter clarifies the applicability of the previous exemption.

, .s 1940-98-20179 Attachment i Page 7 of 7 Justification for Exception 2 GPU Nuclear has determined that the request for exception from the Type B testing requirements for the two feedwater bellows would be acceptable based on the following considerations:

1. The feedwater bellows can not be individually leakrate tested. As the piping is not centered in either penetration, no conventional inflatable seal could be l used. Additionally, as the drywell wall is curved at the penetration, no coupling device exists which could hold the seal in place.

l l 2. The feedwater penetration configuration (openings, reinforcement, and expansion joints) meet both the original 1%2 ASME Vill and subsequent 1980 ASME Ill requirements for the drywell.

I 3, Over 25 years of experience has shown that the bellows type of penetration does not leak. There has never been an instance of containment leakage through either a steamline or feedwater line bellows.

4. The feedwater bellows are tested during the performance of the Type A test.

Thceefore, any penetration leakage would be d:tected at that time.

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