ML17263A977

Application for Amend to License DPR-18,to Revise TS 4.4.2.4.a to Replace Specific Leakage Testing Frequencies for Containment Isolation Valves

ML17263A977
Person / Time
Site:	Ginna
Issue date:	03/13/1995
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17263A976	List: ML17263A975 ML17263A977 ML17263A978
References
NUDOCS 9503160208
Download: ML17263A977 (10)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

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Rochester Gas and Electric Corporation ) Docket No. 50-244 (R.E. Ginna Nuclear Power Plant)

APPLICATION FOR AMENDMENT TO OPERATING LICENSE AND UALITY ASSURANCE PROGRAM Pursuant to Section 50.90 of the regulations of the U.S. Nuclear Regulatory Commission (NRC), Rochester Gas and Electric Corporation (RGE E), holder of Facility Operating License No. DPR-18, hereby requests that the Technical Specifications set forth in Appendix A to that license be amended. This request for change is to replace specific leakage testing frequencies for containment isolation valves with a reference to 10 CFR 50, Appendix J, as modified by approved exemptions.

A description of the amendment request, necessary background information, justification of the requested change, safety evaluation and no significant hazards and environmental considerations are provided in Attachment A. This evaluation demonstrates that the proposed change does not involve a significant change in the types or a significant increase in the amounts of effluents or any change in the authorized power level of the facility. The proposed change also T

does not involve a significant hazards consideration.

A marked up copy of the current Ginna Station Technical Specifications which shows the requested change is set forth in Attachment B. The proposed revised Technical Specifications are provided in Attachment C.

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WHEREFORE, Applicant respectfully requests that Appendix A to Facility Operating License No. DPR-18 be amended in the form attached hereto as Attachment C.

Rochester Gas and Electric Corporation By Robert C. Hecredy Vice President Nuclear Production Subscribed and sworn to before me on this 13 day of Narch 1995.

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Attachment A R.E. Ginna Nuclear Power Plant License Amendment Request Containment Isolation Valve Testing Frequency This attachment provides a description of the amendment request and necessary justification for the proposed change. The attachment is divided into seven sections as follows. Section A identifies all changes to the current Ginna Station Technical Specifications while Section B provides the background and history associated with the change being requested. Section C provides detailed justification for the proposed change including a comparison to Improved Technical Specifications as applicable. A safety evaluation, significant hazards consideration evaluation, and environmental consideration of the requested change are provided in Sections D, E, and F, respectively. Section G lists all references used in this attachment.

A. DESCRIPTION OF AMENDMENT REQUEST This License Amendment Request (LAR) proposes to revise Ginna Station Technical Specifications (TS) as summarized below. Attachment B contains a marked up copy of the current Ginna Station Technical Specifications showing the requested change. The Technical Specifications following the proposed change is provided in Attachment C.

Technical Specification 4.4.2.4.a

i. Replace specific leakage testing frequencies for containment isolation valves with a reference to 10 CFR 50, Appendix J, as modified by approved exemptions.

B. BACKGROUND

1. Hi story On February 21, 1995, the NRC published a proposed revision to 10 CFR 50, Appendix J for public comment (Ref. 1). The proposed revised rule would provide an option for containment leakage testing which is predicated on using performance based acceptance criteria. As such, containment leakage testing frequencies would be developed based on the results of previous tests instead of the current prescriptive frequencies which are now provided in 10 CFR 50, Appendix J. This change is being proposed as a result of extensive industry and NRC evaluations of the actual containment leakage testing results in the nuclear power industry which demonstrates that the current prescriptive testing frequencies are marginal to safety and that a performance based leakage test program is acceptable, and in many ways, preferable. This rule is anticipated to be issued in the fall of 1995.

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The current (and proposed) 10 CFR 50, Appendix J is organized based on three types of tests: Type A, Type B, and Type C. Type A tests refer to the overall containment integrated leakage rate. Type B tests refer to those containment penetrations which use isolation barriers that incorporate resilient seals, gaskets, sealant compounds, bellows, or flexible metal seal assemblies. Type C tests refer to containment isolation valves. This LAR proposes to revise the technical specifications related to Type C penetrations (i.e., containment isolation valves) to remove the prescriptive testing frequency in TS 4.4.2.4.a of "once every reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than two years." This will be replaced with a testing frequency "in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions."

Ginna Station is currently on 12 month refueling cycles such that all containment isolation valves are tested every 12 months, whereas plants with longer operating cycles have longer testing intervals as allowed by 10 CFR 50, Appendix J. Replacing the current prescriptive wording of TS 4.4.2.4.a enables RG&E to revise the testing frequencies of containment isolation valves as allowed by 10 CFR 50. 12 without also having to request changes to technical specifications, since 10 CFR 50, Appendix J must still be met. This eliminates unnecessary duplication between the technical specifications and the Code of Federal Regulations and provides more efficient future use of RG&E and NRC resources.

RG&E is pursuing the requested exemptions to 10 CFR 50, Appendix J, Section III, D.3 concurrent with this LAR such that containment isolation valve leakage testing would be exempted during the 1995 refueling outage.

RG&E anticipates that Ginna Station will leave Cold Shutdown on April 27, 1995 at which time containment integrity is required by LCO 3.6. 1. Since Type C testing is required to ensure containment integrity is available, this LAR and the necessary exemptions must be approved prior to going above Cold Shutdown conditions. As such, RG&E requests that this LAR be approved prior to, or on, April 27, 1995.

2. Hardware Modifications There are no plant modifications which are required to implement the changes requested in this LAR. This,LAR only replaces specific leakage testing frequencies for containment isolation valves with a reference to 10 CFR 50, Appendix J, which provides equivalent testing requirements.

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~ C. JUSTIFICATION This LAR proposes to replace the current specific testing frequencies for containment isolation valves in TS 4.4.2.4.a with a reference to 10 CFR 50, Appendix J, as modified by approved exemptions. Section III, D.3 of Appendix J provides equivalent testing frequencies of containment isolation valves with the exception that "or other convenient intervals" is not specified. Since all operating licenses for water-cooled power reactors are required to meet the requirements of 10 CFR 50, Appendix J, referencing the actual requirement instead of reiterating the requirements within technical specifications provides equivalent testing frequencies.

In addition, removing the actual requirements from technical specifications prevents the need to revise technical specifications as the rule changes, or as exemptions to the rule are granted by the NRC.

Removing the reference to "or other convenient intervals" also eliminates potential conflict with 10 CFR 50, Appendix J since this is not an allowed option for Type C tests. The proposed change is also consistent with LCO 3.6. 1, "Containment," of NUREG-1431 (Ref. 2) in that Surveillance SR 3.6. l. 1 requires testing of containment isolation valves "in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions."

D. SAFETY EVALUATION The change requested in this LAR involves replacing specific leakage testing frequencies for containment isolation valves with a reference to 10 CFR 50, Appendix J, as modified by approved exemptions. This change is consistent with the NRC approved Improved Standard Technical Specifications for Westinghouse plants (Ref. 2). Therefore, the safety impact of making this change has been previously considered and found generically acceptable by the NRC. The plant specific evaluation of this change is provided below.

There is no increase in the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the UFSAR related to this change. 10 CFR 50, Appendix J provides equivalent testing frequencies as the current Ginna Station technical specifications such that containment integrity requirements will still be maintained.

Any change to these testing frequencies still requires prior NRC approval such that there is no increase in the probability or consequences of an accident or malfunction of equipment.

The possibility of a new or different kind of accident or a malfunction of a different type than previously evaluated in the UFSAR is not created by the proposed change. The containment isolation valves are still required to be tested to ensure containment integrity can be met above Cold Shutdown; only the testing frequency is being replaced with a reference to the Code of Federal Regulations. As such, a new or different kind of accident or malfunction is not created by referencing instead of reiterating the testing frequency within technical specifications.

There is no reduction in the margin of safety as defined in the basis for any Technical Specification since the containment isolation valve leakage testing requirements are only being replaced with a reference to 10 CFR 50, Appendix J which provides equivalent controls.

Based on the above, the proposed amendment does not involve an unreviewed safety question and will not adversely affect or endanger the health and safety of the general public.

SIGNIFICANT HAZARDS CONSIDERATION EVALUATION The proposed changes to the Ginna Station Technical Specifications do not involve a significant hazards consideration as discussed below:

Operation of Ginna Station in accordance with the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The change is consistent with NUREG-1431 and has therefore, been previously evaluated and accepted by the NRC. The change involves no technical change to the existing Technical Specification since 10 CFR Appendix J provides equivalent testing frequencies as those currently specified in TS 4.4.2.4.a. There is no impact to initiators of analyzed events or assumed mitigation of accident on transient events. Implementation of this change is expected to result in more efficient use of RG&E and the NRC resources without any reduction in safety.

2. Operation of Ginna Station in accordance with the proposed change does not create the. possibility of a new or different kind of accident from any accident previously evaluated. The change is consistent with NUREG-1431 and has therefore, been previously evaluated and accepted by the NRC. The change does not involve physical alterations of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The change does not impose or eliminate any new or different requirements since 10 CFR 50, Appendix J provides equivalent testing frequencies as those currently specified in TS 4.4.2.4.a.

3. Operation of Ginna Station in accordance with the proposed changes does not involve a significant reduction in a margin of safety. All requirements in the technical specifications related to containment isolation valves remain the same with the exception that a reference to 10 CFR 50, Appendix J is being provided in place of specific leakage testing'equirements. The change has no impact on any safety analysis assumptions.

P Based upon the above information, it has been determined that the proposed change to the Ginna Station Technical Specifications does not involve a significant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of a new or different kind of accident previously evaluated, and does not involve a significant reduction in a margin of safety. Therefore, it is concluded that the proposed changes meet the requirements of 10 CFR 50.92(c) and do not involve a significant hazards consideration.

ENVIRONMENTAL CONSIDERATION RGLE has evaluated the proposed changes and determined that:

1. The change does not involve a significant hazards consideration as documented in Section E above;

2. The change does not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite since the proposed change does not modify any containment integrity requirements; and

3. The change does not involve a significant increase in individual or cumulative occupational radiation exposure since containment isolation valve testing frequencies as contained in 10 CFR 50, Appendix 0 are equivalent to those in the current technical specifications.

Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed changes is not required.

REFERENCES

1. Federal Register, Volume 60, page 9634, February 21, 1995.

2. NUREG-1431, Standard Technical Specifications, westinghouse Plants, September 1992.