ML18089A259
| ML18089A259 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 07/22/1983 |
| From: | Liden E Public Service Enterprise Group |
| To: | Varga S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18089A260 | List: |
| References | |
| LCR-83-04, LCR-83-4, NUDOCS 8307290180 | |
| Download: ML18089A259 (6) | |
Text
OPS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department July 22, 1983 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Mr. Steven Varga, Chief Operations Reactors Branch 1 Division of Licensing Gentlemen:
REQUEST FOR AMENDMENT FACILITY OPERATING LICENSES DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NOS. 50-272 Ref:
LCR 83-04 In accordance with the Atomic Energy Act of 1954, as amended and the regulations thereunder, we hereby transmit copies of our request for amendment and our analyses of the changes to Facility Operating Licenses DPR-70 for Salem Generating Station, Unit No. 1.
This request consists of an extension of the interval of a surveillance requirement for integrated leak rate testing of the containment building.
This change involves a single safety issue and is, therefore, determined to be a Class III Amendment as defined by 10CFR 170.22.
A check in the amount of $4,000 is enclosed.
Pursuant to the requirements of 10 CFR50.9l(b)(l), a copy of this request for amendment has been sent to the State of New Jersey as indicated below.
8307290180 830722 PDR AOOCK 05000272 P
POR The Energy People 95-2168 (SOM) 11-82
Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7/22/83 This submittal includes three (3) signed originals and forty (40) copies.
Enclosure CC:
Mr. Donald C. Fischer Licensing Project Manager Mr. Leif Norrholm Senior Resident Inspector Very truly yours,
- A. Liden Manager -
Nuclear Licensing and Regulation Mr. Frank Cosolito, Acting Chief Bureau of Radiation Protection Department of Environmental Protection 380 Scotch Road Trenton, New Jersey 08628
Ref:
LCR 83-04 STATE OF NEW JERSEY SS.
COUNTY OF SALEM RICHARD A. UDERITZ, being duly sworn according to law deposes and says:
I am a Vice President of Public Service Electric and Gas Company, and as such, I find the matters set forth in our letter dated July 22, 1983, concerning, Request for Amendment to Facility Operating Licenses DPR-70, are true to the best of my knowledge, information and belief.
NC>tary_Public of New Jersey My Commission expires on~
PROPOSED CHANGES TECHNICAL SPECIFICATION SALEM NO. 1 UNIT DESCRIPTION OF CHANGE Ref:
LCR 8 3-04 On a one-time basis, extend the 40 + 10 month interval of Tech-nical Specification 4.6.l.2a during-the first-10 year s-e*r.,/ice -
period to permit the second inservice integrated leak rate test to be performed during the fifth refueling outage by adding a footnote to the specification that reads:
"The second inservice Integrated Leak Rate Test shall be performed at the fifth refueling outage."
REASON FOR CHANGE Paragraph 4.6.1.2 of the Salem Unit 1 Technical Specification defines the surveillance requirements for the overall integrated containment leakage rate, including the schedule for conducting the necessary surveillance tests, to be* in conformance with Appendix J of 10CFR 50.
Specifically the Technical Specifica-tion paragraph states that three Type "A" tests be conducted at 40 + 10 month intervals during each ten year service period, and that the third test of each set be conducted when the plant is shutdown for the ten year plant inservice inspections.
Neither the Technical Specification nor Appendix J define the term "service period", but th0re is such a definition in Section XI of the ASME Boiler and Pressu~e vessel Code which governs inservice inspection at Salem.
Paragraph lWA-2400 of Section XI states that the (10 year) inspection "intervals" represent calendar years after the reactor facility has been placed into Commercial Service.
Due to an earlier interpretation of the Technical Specif ica-tions, the first inservice Type "A" test for Salem Unit 1 was completed on August 13, 1979, approximately 25 1/2 months after the date of commercial operation (approximately 32 months after initial criticality).
Scheduling of the second Type "A" test at the maximum limit of the 40 + 10 month duration from the August, 1979 date would require the test be conducted in October, 1983.
Because of the unforeseen delays encountered during the restart of Salem Unit 1 from its fourth refueling outage, scheduling of the test in October, 1983 would require a four to five week shutdown of the plant for the single purp~se ~f performing the test, after only approximately 5 months of operation.
I
2 -
Ref:
LCR 8 3-0 4 SAFETY EVALUATION
- a.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report will not be increased.
e Paragraph 4.6.1.2 of the Unit No. 1 Technical Specification states that the tests shall be in conformance with the criteria specified in* Apperidlx J of 10CFR 50.
The logic in the intervals specified in Appendix J by the NRC is to have a set of three tests performed at approximately equal intervals.
Waiting until the next scheduled refueling outage to perform the second periodic Type "A" test does not appear to be inconsistent with the intent of the periodic retest schedule specified in Appendix J.
o Acceptable integrated leakage tests have been performed for both the preoperational Type "A" test and for the first Type "A" retest.
The preoperational Type "A" test resulted in a total leakage at the 95% confidence level of 0.718 La; where La is defined as 0.1 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at design pressure (47.0 psig).
The first Type "A" retest resulted in a total leakage at the 95% confidence level of 0.62 La.
- b.
The possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report will not be created.
e This LCR is not related to any plant modification.
- c.
The margin of safety as defined in the basis for any Technical Specification is not reduced.
e Unit No. 1 has not experienced any unusual temperature or pressure excursions within the Reactor Containment Building since the last Type "A" test and we therefore have no reason to suspect that the containment liner integrity has in any way been reduced.
6 A complete local leak rate test program was completed on all penetrdtions and valves requiring Type "B 11 and "C" testing during the most recent refueling outage.
At the end of that outage, the combined leakage from all Type "B" and "C" penetrations and valves was well within the allowable limit of 0.6 La.
Type "B" and "C" tests will be repeated within the 24 month interval specified in the Technical Specification.
3 -
Ref:
LCR __ 8 3-0 4 e
The limiting dose values of 10CFR 100 for purpose of licensing are 25 rem whole body and 300 rem thyroid.
The present design bases LOCA calculations yield a whole body dose of 3 rem and a thyroid dose of 96 rem at the minimum exclusion boundary.
The assumed design bases leak rate of.001 containment free volume/day was utilized in this calculation.
Considerable margin exists between the calculated radiological dose resulting from the design bas is containment leakage and the radfolog-ical dose rates specified in 10CFR 100.
Our evaluation of the conditions described herein enable us to determine that this change introduces no Unreviewed Safety Questions and involves no Significant Hazards Consideration.