ML20248K278

Application for Amend to License DPR-16,requesting That TS 4.5.A.1 Be Revised Such That First Type a Test Required by Primary Containment Leakage Rate Testing Program Be Performed During Refueling Outage 18R

ML20248K278
Person / Time
Site:	Oyster Creek
Issue date:	05/28/1998
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20248K276	List: ML20248K272 ML20248K278 ML20248K285
References
NUDOCS 9806100061
Download: ML20248K278 (5)

Text

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GPU Nuclear, Inc.

Oyster Creek Nuclear Generating Station Facility License No. DPR-16 Technical Specification Change Request No. 254 Docket No. 50-219 Applicant hereby submits a change to Appendix A Technical Specification page e. 5.1.

By:

6 Michael B. Roche Vice President and Director Oyster Creek

~

Sworn and Subscribed to before me this 41 Od ay ofMay,1998.

1 Jb Y Notary Public ofNJ GEf%LDINE E. LEVIN NOTAR19tnUCOF09iRNEY g m m _L.7.20o0 9906100061 990528 "

PDR ADOCK 05000219 P PDR,

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Attachment 1 Oyster Creek Nuclear Generating Station Technical Specification Change Request No. 254 L Chanste Requested GPU Nuclear requests that Appendix A Technical Specification page 4.5.1 be changed to J require the first Type A test of Primary Containment Leakage be done in 18R.  !

I H. Discussion of Proposed Channe The proposed change involves revising Technical Specification (TS) 4.5.A.1 such that the first Type A test required by the Primary Containment Leakage Rate Testing Program (established to meet Option B of 10 CFR 50 Appendix J) be performed during refueling outage 18R. Currently, TS 4.5.A.1 requires that the test be performed during refueling outage 17R.

' IIL Safety Assessment Technical Specification Change Request (TSCR) No. 254 requests that TS 4.5.A.1, be revised such that the first Type A test required by the Primary Containment Leakage Rate Testing Program be performed during refueling outage 18f. rather than 17R.

2 The function of the Primary Containment System is to accommodate, with a minimum of leakage, the pressures and temperatures resulting from the break of any enclosed process pipe; and, thereby, to limit the release of radioactive fission products to values which will insure offsite dose rates well below 10CFR100 guideline limits. The Type A tests are performed to verify the integrity of the Primary Containment System in its LOCA configuration such that the release of fission products to the environment under these postulated accident conditions do not exceed the limits established in 10CFR100.

GPU Nuclear assessed the risk associated with deferring the Type A test until 18R and has provided that assessment, entitled; " Risk Evaluation of the Deferral of the Integrated Leak L

Rate Test (ILRT)" as Attachment No. 3. The assessment is based, in part, on NUREG-1493, entitled;" Performance-Based Containment Leak-Test Program".

NUREG-1493 found that the effect of containment leakage on overall accident risk is small since risk is dominated by accident sequences that result in failure or bypass of the containment. The major contributor to the total identified leakage from Primary Containment comes from Type B and C tested components. Only a small portion of the total leakage is detectable solely through Type A testing. This is typical of the information found throughout the nuclear industry. As noted in GPUN's July 17,1996 submittal concerning the implementation of 10 CFR 50, Appendix J, Option B, the last three Type A tests performed at Oyster Creek have all been within specification.

Attachment 1 Page 2 of 4 i NUREG-1493 presents summaries of risk analyses performed in support of determining the cost benefit associated with extending Appendix J testing intervals, including Type A testing. The analyses included evaluations of the Peach Bottom Nuclear Power Station.

The Peach Bottom Nuclear Power Station is a General Electric Boiling Water Reactor I

(BWR-4) with a Mark I type containment. Oyster Creek is a _ General Electric Boiling Water Reactor (BWR-2) with a Mark I containment. A full comparison of plant features is provided in Attachment No. 3 entitled; " Risk Evaluation of the Deferral of the

, Integrated Leak Rate Test (ILRT)". Given the similarity in containment type, the analysis contained in NUREG-1493 is judged to be appropriate in addressing the deferral of the ILRT for Oyster Creek. GPUN's risk assessment, Attachment No. 3, states "The deferral of the Oyster Creek ILRT test from the 17R to the 18R refueling outage results in an imperceptible increase in risk. This is based on the conclusions contained in NUREG-1493 which states that increases in the testing frequency of Type A (n,RT) tests from three in ten to one in 20 years results in an imperceptible increase in risk. Other factors, l such as the monitoring of nitrogen usage and the drywell to torus vacuum breaker testing further ameliorate this small increase by allowing for the detection of gross leakage."

The Containment Inerting System maintains an atmosphere of nitrogen gas within the drywell and torus air space during power operations in order to maintain the oxygen level below 4% by volume. As noted above, Oyster Creek has two means of detecting gross l: containment leakage. The first is by monitoring the use of nitrogen. If excessive nitrogen I consumption is discovered the cause of the use is investigated. A gross leakage path will l be identified due to the use of nitrogen and will trigger an investigation into the cause.

l The second way of determining gross leakage is by performance of the torus to drywell

!' vacuum breaker leak test. This periodic test monitors drywell and torus pressure for a one-hour period to calculate bypass leakage across the vacuum breakers.- The test results L .would also provide an indication if an external leakage path from either the torus or l drywell exists. . This too will trigger an investigation into the cause of the excessive ,

i leakage. 'While not testing the structure at accident pressurc, these methods prove that l l there has been no gross structural degradation since the last ILRT. .

1 Based on the above, it is concluded that the proposed change does not adversely affect nuclear safety or safe plant operations.

IV. Information Suonertine a Finding of No Significant Hazards Consideration 1

GPU Nuclear has concluded that the proposed change to defer the first Type A test required by the Primary Containment Leakage Rate Testing Program until 18R does not involve a Significant Hazards Consideration. In support of this determination, an i

' evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below. l l~

F Attachment 1 Page 3 of4

l. The oroposed TS channe does not involve a significant increase in the probability or i- consequences of an accident oreviously evaluated.

' l l The proposed change does not alter the design, function or manner of operation of any structures, systems or components. As a result, the proposed change does not affect any of the parameters or conditions that could contribute to initiation of any

i. accidents.

NUREG-1493 found that the effect of containment leakage on overall accident risk is small since risk is dominated by accident sequences that result in failure or bypass of the containment. The major contributor to the total identified leakage from Primary Containment comes from Type B and C tested components. Only a small portion of the total leakage is detectable solely through Type A testing. The leaks that have been found by Type A tests have been only marginally above existing requirements. In addition, Oyster Creek has two means (monitoring t

nitrogen use and performing torus to drywell vacuum breaker leak tests) of

- detecting gross containment leakage. The proposed change does not alter the requirements to perform Type B and C testing in accordance with the Primary Containment Leakage Rate Testing Program and does not affect the ability of the facility to mitigate the consequences of an accident.

Therefore, the proposed TS change does not involve an significant increase in the

probability or consequences of an accident previously evaluated.

2. ' The proposed TS channe does not create the possibility of a new or different kind of accident from any accident previousiv evaluated.

' Deferring the Type A test for an operating cycle does not alter the design, function

- or manner of operation of any structures, systems or components. The proposed change does not affect any of the parameters'or conditions that could contribute to initiation of any accidents nor does it introduce any new mechanisms which could contribute to the creation of a new or different kind of accident than previously evaluated.

3. The oroposed TS channes do not involve a significant reduction in a margin of safety.

The proposed change does not alter the design, function or manner of operation of any structures, systems or components. The proposed change does not impact the l primary containment system's ability to provide a barrier against the uncontrolled release of fission products in the event of a break in the reactor coolant system nor does the proposed change impact the primary containment accident leak rate. In addition, NUREG-1493's Summary of Technical Findings states " Reducing the l

Attachment 1 Page 4 of 4 frequency of Type A tests (ILRTs) from the current three per 10 years to one per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements." Therefore, the proposed TS change does not involve a significant reduction in a margin of safety.

V. Information Sunnortine an Environmental Assessment i

An environmental assessment is not required for the proposed change since the proposed change conforms to the criteria for " actions eligible for categorical exclusion" as specified in 10 lj CFR 51.22(c)(9). The proposed change will have no impact on the environment. The proposed change does not involve a significant hazards consideration as discussed in the preceding section. The proposed change does not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed change does not involve a significant increase in individual or cumulative occupational radiation exposure.

VL Conclusion The proposed change to the TS have been reviewed in accordance with Section 6.5 of the Oyster Creek Technical Specifications and it has been concluded there are no unreviewed safety questions. As discussed above, using the standards in 10 CFR 50.92, GPU Nuclear has determined that there are no Significant Hazards Considerations involved with the proposed changes.

Vil IMPLEMENTATION It is requested that the amendment authorizing this change become effective upon issuance.

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