ML18153A676

Proposed Tech Specs Re App J Option B,performance-based Containment Leakage Rate Testing

ML18153A676
Person / Time
Site:	Surry
Issue date:	11/20/1995
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18153A675	List: ML18153A674 ML18153A676
References
NUDOCS 9511270429
Download: ML18153A676 (7)

Text

• Attachment 2 Technical Specifications Change Surry Power Station 0 ~11270429 951120, PDR ADO~K 05000280 * .. ,' * *

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. 4A . CONTAINMENT TESTS

• TS 4.4-1 Applicability Applies to containment leakage testing.

Objective To assure that leakage of the primary reactor containment and associated systems is held within allowable leakage rate limits; and to assure that periodic surveillance is performed to assure proper maintenance and leak repair during the service life of the containment.

Specification A; Periodic and post-operational integrated leakage rate tests of the containment shall be performed in accordance with the requirements of 10 CFR 50, Appendix J, "Reactor Containment Leakage Testing for Water Cooled Power Reactors."

B. Containment Leakage Rate Testing Requirements

1. The containment and containment penetrations leakage rate shall be demonstrated by performing leakage rate testing in accordance with 10 CFR 50 Appendix J, Option B, as modified by approved exemptions, and Regulatory Guide 1.163.

Leakage rate acceptance criteria are as follows:

a. An overall integrated leakage rate of less than or equal to La, 0.1 percent by weight of containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at calculated peak pressure (Pa).

b. A combined leakage rate of less than or equal to 0.60 La for all penetrations and valves subject to Type B and C testing when pressurized to Pa.

Prior to entering an operating conditions where containment integrity is required the as-left Type A leakage rate shall not exceed 0.75 La and the combined leakage rate of all penetrations subject to Type Band C testing shall not exceed 0.6 La."

2. The provisions of Specification 4.0.2 are not applicable.

The leak tightness testing of all liner welds was performed during construction by welding a structural steel test channel over each weld seam and performing soap bubble and halogen leak tests.

Amendment Nos.

The containment is designed for a maximum pressure of 45 psig. The containment is TS 4.4-2 maintained at a subatmospheric air partial pressure consistent with TS Figure 3.8-1 depending upon the cooldown capability of the Engineered Safeguards and will not rise above 45 psig for any postulated loss-of-coolant accident.

The initial test pressure for the Type A test is 4 7 .0 psig to allow for containment expansion and equalization. A review was performed to determine the effects o~

pressurizing containment above its design pressure of 45.0 psig. This review was based on the original containment test at 52 psig. During that test, the calculated stresses were found to be well within the allowable yield strength of the structural reinforcing bars, therefore performance of the Type A test at 47 psig will have no detrimental effect on the containment structure.

All loss-of-coolant accident evaluations have been based on an integrated containment leakage rate not to exceed 0.1 % of containment volume per 24 hr.

The above specification satisfies the conditions of 10 CFR 50.54(0) which stated that primary reactor containments shall meet the containment leakage test requirements set forth in Appendix J.

The limitations on closure and leak rate for the containment airlocks are required to meet the restrictions on containment integrity and containment leak rate. Surveillance testing of the airlock seals provides assurance that the overall airlock leakage will not become excessive due to seal damage during the intervals between airlock leakage tests.

References UFSAR Section 5.4 Design Evaluation of Containment Tests and Inspections of Containment UFSAR~Section 7:5:-1 *---*-*Design-Bases ofEngineered*Safeguards*Instrumentation UFSAR Section 14.5 Loss of Coolant Accident 10 CFR 50 Appendix J "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors" Amendment Nos.

• TS 6.6-11 C. Special Reports In the event that the Reactor Vessel Overpressure Mitigating System is used to mitigate a RCS pressure transient, submit a Special Report to the Commission within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or the administrative controls on the transient and any corrective action necessary to prevent recurrence.

FOOTNOTES

1. A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

2. This tabulation Sl!pplements the i;~quiJernents _of Section 20.2206 of 10 CFR Part 20.

3. A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

Amendment Nos.

Attachment 3 Significant Hazards Consideration Determination Surry Power Station

tO CFR 50.92 EVALUATION - BASIS FOR NO SIGNIFICANT* HAZARDS DETERMINATION The Nuclear Regulatory Commission has amended its regulations to provide a performance-based option for leakage-rate testing of containments. This testing option is available in lieu of compliance with the prescriptive requirements contained in Appendix J regulations. In order to implement the performance-based leakage-rate testing option the Technical Specifications must be changed to eliminate reference to the prescriptive Appendix J requirements. Therefore, Virginia Electric and Power Company (Virginia Power) is proposing a change to the Surry Technical Specifications to eliminate the current prescriptive requirements for leakage rate testing of the containment and reference Option B to 10 CFR 50 Appendix J and NRC Regulatory Guide 1.163, "Performance-Based Containment Leakage-Test Program."

This change will permit use of the performance-based surveillance testing, Option B, of 10 CFR 50 Appendix J. Specifically, operation of Surry Power Station with the proposed change will not:

1. Involve a significant increase in either the probability of occurrence or consequences of any accident or equipment malfunction scenario which is important to safety and which has been previously evaluated in the Updated Final Safety Analysis Report (UFSAR).

Plant systems and components will not be operated in a different manner as a result of the proposed Technical Specifications change. The proposed change permits a performance-based approach to determining the leakage-rate test frequency for the containment and containment penetrations (Type A, B, and C tests). There are no plant modifications, or changes in methods of operation. Therefore, the changes in testing intervals for the containment and containment penetrations have no affect on the probability of occurrence of a LOCA. Since the proposed change only affects the test frequency for containment and the containment penetrations, and the as-found test acceptance criteria at Surry the probability of occurrence and the consequences of an accident are not affected by the proposed changes in the leak-rate test interval.

The proposed change increases the probability of a malfunction of equrpment important to safety due to the longer intervals between leakage tests. It has been estimated that the longer test intervals will increase the overall accident risk to the public by approximately 0.7% and 2.2% (for changes in the frequency of Type A tests and Type B and C tests, respectively). However, this increase in accident risk has been judged to be insjgnificant. This increa?e ha? been reviewed and.juqged to be acceptable by the NRC as documented in NUREG-1493 and the recent rulemaking to 10 CFR 50 Appendix J.

The containment and other safety system remain operable as assumed in the accident analysis. Changing the as-found acceptance criterion to 1.0 La at Surry does not increase the probability or consequences of an accident, since the accident analysis assume a leakage rate of La for Design Basis Page 1 of 2

Accidents. The as-left Type A test acceptance criterion remains at less than

0. 75 La. Since the proposed changes do not affect the Limiting Conditions for Operation for the containment, the containment penetrations, or the other safety systems, the consequences of an accident are not affected by the changes in test frequency.

Therefore, the probability of an accident or consequences of an accident are not adversely affected as a result of this change.

2. Create the possibility of a new or different type of accident than those previously evaluated in the UFSAR.

Implementing the proposed Technical Specifications change to remove the prescriptive testing requirements and permit use of Appendix J, Option B, performance-based testing of containment and its penetrations does not create the possibility of an accident of a different type than was previously evaluated in the UFSAR. Plant systems and components will not be operated in a different manner as a result of the proposed Technical Specifications changes. Thus, the proposed Technical Specifications changes in leakage-rate test frequency do not introduce any new accident precursors or modes of operations. The containment and containment penetrations will not be operated any differently as a result of the proposed changes. Therefore, the possibility for an accident of a different type than was previously evaluated in the Safety Analysis Report is not created by the proposed Technical Specifications change.

3. Involve a significant reduction in a margin of safety.

The proposed Technical Specifications change, which replace the present prescriptive testing requirements with Appendix J, Option B, performance-based testing of containment and its penetrations, will continue to ensure that the existing accident analysis assumptions are maintained. The containment and containment penetrations will not be operated or tested any differently. The leakage rate test frequency is being changed as a result of the proposed change. Changing the as-found acceptance criterion to 1.0 La at Surry does not increase the consequences of an accident, since the accident analysis assume a leakage rate of La for Design Basis Accidents.

The as-left Type A test acceptance criterion remains at less than 0.75 La, which maintains the operating margin. The operational leakage-rate test acceptance criteria and the operability requirements are not being changed.

Therefore, the margin of safety as defined in the Technical Specifications bases is unaffected.

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