ML20004A063
| ML20004A063 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/28/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20004A053 | List: |
| References | |
| NUDOCS 8105110635 | |
| Download: ML20004A063 (6) | |
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O SAFETY EVALUATION SY-fRE OFFICE OF NUCLEAR REACTCR REGULATION METROPCLITAN EDISON CCMPANY JERSEY CENTRAL PCWER AND LIGHT CCMPMY PENNSYLVANIA ELECTRIC CCMPANY C0CKET NO. 50-320 TBREE MILE ISLAND dUCLEAR STATION, UNIT N0. 2 Intrcduction Metrcpolitan Edisen Coccany, Jersey Central Power and Lignt Cccpany and Pennsylvania Electric Cccoany (collectively, the Licensee) are the holders of Facility Operating License No. OPR-73, which had authcrized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power levels up to 2772 megawatts thermal.
By Order for Modification of License, dated July 20, 1979, the Licensee's authority to operate the facility was suspended and the Licensee's authcrity was limited to maintenance of the facility in the present shutdown ccoling code (24 Fed. Reg. 45271).
By further Order of the Directer, Office of Nuclear Reacter Regulation, dated Fecruary 11, 1980, a new set of formal license recuirements were imocsed to reflect the pcst-accident condition if the f acility l
and to assure the continued maintenance of the current safe, s.aole, icng-term cooling conditien of the f acility (45 Fed. Reg.11232).
These requirements were memorialized in the form cf propcsed Tecnnical Specifications set forth in an attachment to tne Order.
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.. Several request", for a hearing have been filed in connection with the Order and granted by the presiding Atomic Safety and Licensir.g Board estaolished to rule on such requests and to preside over any eventual hearings.
These parties have sought to introduce a numoer of issues involving the proposed Technical Specirications.
These include concerns regarding the reactor coolant systam pressure safety limit 'preposed Technical Specification 2.1.3),
remote shutdcwn monitoring instrumentation (proposed Technical Specification 3.3.3.5), reactor coolant system pressure / temperature limits (preposed Technical Specificatien 3.4.9.1), and record retention (preposed Technical Specifications 5.10.1 and 6.10.2). Consistent with the Commission's regulations which encourage settlemer.t of potential issues in a proceeding (see 10 CFR 12.759), the Staff has modified the preposed Technical Specifications in a manner agreed upoa by the principals and described hereafter.
Evaluation The February 11, 1980 Order establisned, in the form of preposed Technical Specification 2.1.3, a reactor coolant system pressure safety limit of 2750 psig.
Tne basis for this safety limit was the cesign criteria and asscciated ASME Boiler and P ressure Vessel Code requirements acplicable to the reactor coolant system prior to the March 28, 1979 acci dent. This Order also set a reactor coolant system limiting condition for cperatien of 500 psig contained in proposed Technical Specification 3.4.9.1.
The basis fer this lianting condition for operation was to preclude the pcssibility of a ncncuctile failure of the reactor coolant system.
The accident l
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l s suojected portions of the reactor coolant system to unknown environmental conditions and, therefore, the pressure retaining ability of the reactor coolant system is somewhat uncertain.
However, the ability of the reactor coolant system to witnstand a pressure of 600 psig was demonstrated by its operation for extended time intervals at 800-1050 psig curing April 1979 (Reference 1). Furthermore, 10 CFR 550.36(c)(1)(1)(A) of the Commission's regulation requires, in part, that, in the event a safety limit is exceeded, the reactor shall be shut down and that operation shall not be resumed until authorized by the Commission.
Since the TMI-2 reactor is alreacy shut down, and since the licensee's authority to operate TMI-2 in other than its present shutoown condition was suspended by the Order for Modification of License dated July 20, 1979, a reactor coolant system safety limit is not required and can be eliminated from the proposed Technical Specifications. Along with eliminating this safety limit, and to clarify the actions to be taken by the licensee in the event the 600 psig limit is exceeded, we have also modified the Action statement for proposed Technical Specifi-cation 3.4.9.1 to explicitly identify the respcnsive action wnich must be taken if the pressure limit estaolished for tne reactor coolant system, 600 psig, is exceeced.
One of the parties in this matter contended that the allowable out-of-service time in the Action statement of proposed Technical Specification 3.3.3.5 was exces-sively long at 30 days and should De snortened to 7 days. We have not enanged this allcwaole out-of-service time since it is consistent witn the requirements of the Stancarc Tecnnical Specifications for Baecock and Vilcox Pressurized Water Reactors (NUREG-0103). Mcwever, we have supplemented the Action statement for proposed Tech-nical Specification 3.3.3.5 to require the licensee to report the incperability of
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o one of these channels to the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This additional provision will ensure that the NRC is promptly notified if one or more of the Remote Shutdown Instrumentation channels becomes inoperable. The NRC could then initiate any additional actions which aay be appropriate.
Two of the parties seeking a hearing contended that the record retention require-ments of proposed Technical Specifications 6.10.1 and 6.10.2 were inadequate and that the subject records should be retained for longer than the requirements of these pro-posed Technical Specifications.
Since some of these records may have historical value, proposed Technical Specification 6.10.2 has been augmented to include most of the records previously included in proposed Technical Specification 6.10.1.
The records designated in proposed Technical Specification 6.10.2 must be retained as long as the Licensee has a NRC license to operate or possess the TMI facility.
Environmental Consideration We have determined that the modification does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the modificiation involves an action which is insignificant from the j
stancpoint of environmental impact and, pursuant to 10 CFR Section 51.5(d) (4), that an environmental impact statement or negative declaration ard environmental impact l
appraisal need not be prepared in connection with the issuance of the modification.
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-3 Conclusion As discussed toove, :ne socification te preusec Tecnnical Specifications 3.3.3.5, 3.4.9.1, 5.10.1 ana 5.10.2 and :ne celetien of progesec Tecnnical Speci-fication 2.1.3 do not lessen (anc in some cases augment) ne affectec requirements of the Director's Fecruary 11,19e0, Order. Therefere, we have conclucec :nat:
(1) :ne socificatiens cc not involve a significant increase in :ne prenanility er censecuences of accioents previcusly censicerec anc co not involve a sienificant hazarcs consiceraticn, (2) Onere is reascnacle assurance :na: :ne health anc safety of :ne puolic will no: se encangerec sy coeration in :ne accifiec sanner, anc (3) sucn activities will :e ccncucted in ccac11ance with :ne Ccausissicn's regulaticns and :ne issuance of :nis socification will no: te inimical to :ne ccamen cefense anc security cr to tne health and safety of the puslic.
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REFERENCES 1.
Grapn attached to letter free Steven C. Golcters, UShRC, to 'dilliam A. Locnstet, cated August 13, 1980.