ML20128F479

Safety Evaluation Accepting Rev to Inservice Testing Program Re Leakage Testing of Containment Isolation Valves

ML20128F479
Person / Time
Site:	Oyster Creek
Issue date:	10/01/1996
From:	NRC (Affiliation Not Assigned)
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ML20128F460	List: ML20128F456 ML20128F479
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NUDOCS 9610080092
Download: ML20128F479 (7)

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ye t UNITED STATES 1 .

-t j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 200eH001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE INSERVICE TESTING PROGRAM i

GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NUMBER 50-219 i

1.0 INTRODUCTION

The Code of Federal Regulations, 10 CFR 50.55a, requires that inservice testing (IST) of certain American Society of Mechanical Engineers (ASME) Code

' Class 1, 2, and 3 pumps and valves be performed. in accordance with Section XI of the ASME Boffer and Pressure Vessel Code (the Code) and applicable addenda,  !

except where alternatives have been authorized or relief has been requested by the licensee and granted by the Commission pursuant to Sections (a)(3)(1),

(a)(3)(ii), or (f)(6)(i) of 10 CFR 50.55a. In proposing alternatives or requesting relief, the licensee must demonstrate that: (1) the proposed i

alternatives provide an acceptable level of quality and safety; (2) compliance would result in hardship or unusual difficulty without a compensating increase in-the level of quality and safety; or (3) conformance is impractical for its facility. Section 50.55a authorizes the Commission to approve alternatives  !

and to grant relief from ASME Code requirements upon making the necessary ,

findings. Additionally, paragraph (f)(4)(iv) of Section 50.55a provides that IST of pumps and valves may meet the requirements set forth in subsequent editions and addenda of the Code that are incorporated by reference in paragraph (b) of Section 50.55a, subject to the limitations and modifications  !

listed therein, and subject to Commission approval. Portions of editions or j

addenda may be used provided that all related requirements of the respective editions or addenda are met. Guidance related to the development and i implementation of IST programs is given in Generic Letter (GL) 89-04,

" Guidance on Developing Acceptable Inservice Testing Programs," issued April 3,1989, and its Supplement 1 issued April 4,1995. Also see NUREG-1482, " Guidelines for Inservice Testing at Nuclear Power Plants."

4 The 1989 Edition of the ASME Code is the latest edition incorporated by reference in paragraph (b) of Section 50.55a. Subsection IWV, which gives the requirements for IST of valves, references Part 10 of the American National ,

Standards Institute /ASME Operattons and #aintenance .itandards (OM-10) as the rules for IST. OM-10 replaces specific requirements in previous editions of Section XI, Subsection IWV, of the ASME Code.

2.0 BACKGROUND

The NRC published a final rule change to 10 CFR 50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," in the 9610080092 961001 PDR ADOCK 05000219 G PDR

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Federal Register on September 26, 1995 (60 FR 49495). The final rule became  !

effective October 26, 1995. The revised regulations provide a performance-

based option for leakage-rate testing of containments (" Option 8"). Licensees 1 1 may voluntarily adopt the option in lieu of compliance with the prescriptive  !

] requirements in the regulation (" Option A"). The NRC issued the change as i part of an effort to improve the focus of regulations by eliminating prescriptive requirements that are marginal to safety. The final rule allows  :

leakage test intervals to be based on system and component performance. Thus, '

licensees have greater flexibility for cost-effective implementation methods i

in satisfying regulatory safety objectives. j

GPU Nuclear Corporation (GPU) requested Technical Specification Change Request i i (TSCR) 242 for the Oyster Creek Nuclear Generating Station in its letter of i i February 23, 1996. On June 19, 1996, GPU submitted Revision 1 to TSCR 242.

l On July 17, 1996, as supplemented August 28, 1996 GPU submitted a revised TSCR l

! 242 (Revision 2). The NRC issued Amendment 186 to the Oyster Creek Technical  !

j Specifications (TS) on September 3, 1996. Amendment 186 allows GPU to implement Option B of Appendix J and to establish a performance-based leakage-4 1

rate test interval for the containment isolation valves subject to leakage l

testing, j l 3.0 REVISION RELATED TO IST LEAKAGE TESTING OF VALVES l By letter dated March 28, 1996, GPU submitted a revision to the_ Oyster Creek Nuclear Generating Station IST Program. GPU requested NRC approval of an

alternative to the valve leakage testing requirements of the 1986 Edition of the ASME Code. The alternative applies to the IST program for the Oyster Creek Nuclear Generating Station and would allow the licensee, GPU, to use a 1ater edition of the ASME Code for leakage testing containment isolation valves. Effectively, approval of the IST program revisions would allow GPU to implement Option B consistent with NRC approval for implementing the revised TS. In the IST program revision, GPU proposes to use the requirements of OH-1 10 related to leakage testing of valves, including containment isolation valves. GPU makes the request pursuant to the provisions of paragraph

(f)(4)(iv) of Section 50.55a for the use of portions of later editions of the ASME Code. GPU's proposal includes all related requirements for valve seat leakage-rate testing.

4 3.1 Discussion The 1986 Edition Section XI of the ASME Code includes requirements for valve leak rate testing in paragraph IWV-3420. These rules are applicable to all Category A valves (i.e., valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their function). Position 10 of GL 89-04 indicated that all containment isolation i valves included in the Appendix J program should be included in the IST ,

program as Category A valves (or Category A/C for check valycs that have a i leaktight function for containment isolation). In Position 10, the NRC said  !

j that the valve leakage-rate testing requirements of Appendix J were considered

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. equivalent to the requirements of IWV-3421 through IWV-3425, but that licensees must comply with the analysis of leakage rates and corrective action requirements of IWV-3426 and IWV-3427(a).

The requirements of IWV-3421 through IWV-3425 apply to te scope, frequency,

! differential test pressure, seat leakage measurement, and -est medium.

Paragraph IWV-3422 requires a test frequency of at least once every 2 years.

The leakage-rate testing of valves in Appendix J at the time GL 89-04 was issued (and the current Option A of Appendix J) requires tests shall be performed during each reactor shutdown for refueling but in no case at intervals greater than 2 years. The performance-based interval in the new Option B of Appendix J cannot be considered equivalent to the frequency required by IWV-3422.

Paragraph 4.2.2.1 of OH-10 specifies the scope of valve seat leakage-rate tests as follows:

Category A valves shall be leakage tested, except that valves which function in the course of plant operation in a manner that demonstrates functionally adequate seat leak-tightness need not be additionally leakage tested.

In such cases, the valve record shall provide the bases for the conclusion that operational observations constitute satisfactory demonstration.

Paragraph 4.2.2.2 of OH-10 specifies the requirements for containment isolation valves as follows:

Category A valves, which are containment isolation valves, shall be tested in accordance with Federal Regulation 10 CFR 50, Appendix J. Containment isolation valves which also provide a reactor coolant system pressure isolation function shall additionally be tested in accordance with para. 4.2.2.3.

Paragraph 4.2.2.3 of OH-10 gives the requirements for leakage-rate testing for valves other than containment isolation valves, including frequency, differential test pressure, test medium, analysis of leakage rates, and corrective action. The frequency requirements for containment isolation valves would be specified by Appendix J. Paragraph (b)(2)(vii) of Section 50.55a modified the requirements of OH-10 for IST of containment isolation valves. Specifically, paragraph (b)(2)(vii) requires that, when using OM-10 for IST, leakage rates for Category A containment isolation valves that do not provide a reactor coolant system pressure isolation function must be analyzed in accordance with paragraph 4.2.2.3(e) of OM-10 and corrective actions for these valves must be made in accordance with paragraph 4.2.2.3(f) of OH-10. These provisions identify related requirements for using OM-10 leakage requirements. The regulations take no other exceptions to Qe provisions of OH-10. Therefore, conducting IST in accordance with G1-10 does not precluce the use of Option B of Appendix J for establishing a performance-

! based leakage monitoring schedule for leak testing containment isolation valves.

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} Containment isolation valves that have another leaktight safety function

! (e.g., pressure isolation valves, or train separation where flow diversion l could be a critical parameter) may also be subject to paragraph 4.2.2.3 provisions for testing the second function. This additional function is not

, deferred to Appendix J requirements by paragraph 4.2.2.2 of OM-10. Tests for verification of a check valve's capability to close are required as part of an i IST program. Tests to exercise check valves are often deferred from quarterly i to cold shutdowns or refueling outages because of impractical conditions.

l Many licensees use the local leak-rate test performed to reet the requirements l of Appendix J to verify that a check valve is capable cf closing, as discussed

in Section 4.1.4 of NUREG-1482. Closure verification does not require the

rigor of a local leak-rate test. Where the test is currently used to verify '

closure during each refueling outage, a licensee may determine that a less-rigorous leak-rate test is capable of verifyinf that a check valve is capable i of closing, or may determine another means of exercising the valve.

i For Oyster Creek Nuclear Generating Station, where closure of check valves is

verified by the Appendix J 1eak test, a number of relief requests have been i changed to delete references to a local leakage-rate test per Appendix J. GPU will continue to rely on a leak test (or another positive means, including

] nonintrusive techniques), but not necessarily an Appendix J local leakage-rate a test. The purpose and intent of these relief requests are to defer testing

from quarterly or during cold shutdown to during refueling outages. The j intent of the relief requests is not changed, and the revisions are therefore i acceptable. These relief requests were granted in NRC's Safety Evaluation

! dated September 24, 1992, based on the impracticality of exercising the l applicable valves quarterly. Relief was granted to defer testing to refueling i outages so that the valves could be leak tested to verify closure. The I revised relief requests are granted for the same alternative frequency

pursuant to 10 CFR 50.55a(f)(6)(1) based on the impracticality of testing the

! valves quarterly. It would be a burden on the licensee to impose the

] requirements to test the valves quarterly, necessitating design changes or i valve replacement. The applicable valve relief requests (RR) are RR-2, RR-3, RR-9, RR-15, RR-17, RR-31, and RR-47.

l i Also related to the change involving Appendix J 1eakage rate testing are j revisions to RR-50 and RR-51. RR-50 and RR-51 were revised to allow an option i of local- observation or parameter response to verify accurate position j indication for the applicable containment isolation valves; whereas i previously, these referred to the use of an Appendix J 1eakage-rate test to

! verify the position indication. It is impractical to verify valve position by I

! local observation as required by the Code. The revision continues to meet the l

! requirements to verify remote position indication every 2 years, specifically i i relying on system parameters (e.g., flow, pressure, system operation). These l relief requests are authorized pursuant to 10 CFR 50.55a(a)(3)(1) based on the I

} acceptable level of quality and safety provided by the alternative methods of verifying valve position.

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3.2 Reauest for Use of Portions of OH-10 l GPU currently uses the testing requirements of Appendix J in accordance with the guidance of Position 10 of GL 89-04 for containment isolation valve I leakage-rate monitoring. The requirements of IWV-3426 and IWV-3427(a) for the analysis of leakage rates and corrective action are also imposed for the IST j of these valves. A test frequency of at least once every 2 years (typically ,

during refueling outages) is currently required for the valves that are " Type C" tested per Appendix J (i.e., tests intended to measure containment  :

isolation valve leakige rates). i i

l In Section 3.0 of the Oyster Creek Nuclear Generating Station, GPU proposes to l

! use OM-10, Section 4.2.2, " Valve Seat Leakage Rate Test," for IST leakage l

testing of valves, including containment isolation valves. Specifically, for  !

l containment isolation valves, the related requirements include paragraphs 4.2.2.2, 4.2.2.3(e), and 4.2.2.3(f) of Section 4.2.2. GPU will conduct leakage-rate testing of containment isolation valves according to the provisions of Appendix J, including Option B, according to revised technical specifications and where applicable. GPU will implement the analysis of leakage rates and corrective action requirements imposed by the regulations l for conteineent isolation valves. Issuance of Amendment 186 to the Oyster

, Creek TS, dated September 3,1996, in conjunction with NRC approval of GPU's IST program proposal will enable implementation of the Option B initial testing during the refueling outage scheduled for September 1996.

3.3 Evaluation The 1989 Edition of the ASME Code was incorporated by reference in rulemaking effective September 8, 1992 (57 FR 34C66). The NRC recommended that licensees update their IST program to the OM Standards referenced in the 1989 Edition of the Code (see NUREG-1482) as alternative requirements to those in earlier editions of the Code. Several plants are conducting valve IST programs according to the provisions of OM-10, including plants that revised their program to meet the updating provisions of Section 50.55a (i.e., at each 120-month interval) and plants that voluntarily implemented the requirements pursuant to paragraph (f)(4)(iv) of Section 50.55a as recommended in NUREG-1482. For valves subject to seat leakage-rate testing, other than containment isolation valves, the requirements of OM-10 are considered essentially equivalent to the requirements of IWV-3421 through IWV-3427 of earlier editions of the Code.

For plants using OM-10 for IST of valves, no conflict exists between Appendix J and OM-10 for leakage testing of containment isolation valves. For plants that have not yet updated to the requirements of OM-10, there is a conflict in the test frequency that would preclude the use of Option 8 of Appendix J if no alternative is available. In issuing the Appendix'J rule change, the NRC did not intend to create a conflict for the plants continuing to use earlier editions of the Code.

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1 l Option 8 specifies that the periodic schedule for Type B and Type C testing be a

based on the safety significance and historical performance of each boundary i and isolation valve to ensure the integrity of the overall containment system  :

as a barrier to fission product release to reduce the risk from reactor j accidents. Performance criteria are given in the regulations. A similar i

scheduling scheme based on risk-assessment and performance-based criteria is  !

under development for IST, but rules have not yet been promulgated in the IST i regulations, Section 50.55a.

! Because the requirements of Appendix J are acceptable for leakage-rate testing l of containment isolation valves, and because GPU has already received TS

. amendment for implementing Option B, it would be appropriate to allow the i licensee to apply the performance-based criteria to the valves as opposed to

! the requirements in an earlier edition of the Code. The Appendix J rule

! change assessed safety concerns related to the extended test intervals and i determined that the extended intervals are acceptable based on the assessment

of historical data on leakage test results, leakage trends, valve performance, 4 and failure rates. Data support extension of the leakage rate test without adversely impacting valve performance and leak-tightness capability.

l Therefore, pursuant to 10 CFR 50.55a(f)(4)(iv), the licensee may use the portions of OM-10 that relate to leakage testing of containment isolation

] valves to remove the inconsistency in the requirements. These related

requirements of OM-10 are

Paraaraoh Number Title

4.2.2.2 Containment Isolation Valves j 4.2.2.3(e) Analysis of Leakage Rates

4.2.2.3(f) Corrective Action

! These paragraphs apply whether the licensee uses Option A (current rules) or j Option B of Appendix J. The testing method, frequency, acceptance criteria, l test medium, and leakage measurements must meet the requirements in i Appendix J. The analysis of leakage rates and corrective action must meet the i requirements of both Appendix J and OM-10, as applicable. There are no other

related requirements in OH-10 applicable to the portions listed above.

3 3.4 Conclusion GPU's proposal to use portions of the latest edition of the ASME Code i incorporated by reference in paragraph (b) of Section 50.55a is acceptable for

IST leakage-rate testing of valves, including containment isolation valves.

1 All related requirements applicable to the testing are included in GPU's

, implementation of the portions of OM-10 as required by the regulations.

l Approval is pursuant to the provisions in 10 CFR 50.55a(f)(4)(iv) that allow l that IST of pumps and valves may meet later editions and addenda incorporated by reference in paragraph (b) of Section 50.55a, subject to Commission approval, and provided that all related requirements are met. The authorization or RR-50 and RR-51 is in accordance with the provisions of l

10 CFR 50.55a(a)(3)(1). The relief granted for RR-2, RR-3, RR-9, RR-15,

! RR-17, RR-31, and RR-47 is in accordance with the provisions of 10 CFR l 50.55a(f)(6)(1) and is in consideration of the burden on the licensee if the  ;

l requirements were imposed. These conclusions are authorized by law and in the interest of public health and safety.

Principal Contributor: P. Campbell l

Date: October 1, 1996 l

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