ML20141G125

Exemption from 10CFR50.61 Requirements for Protection Against Pressurized Thermal Shock Events

ML20141G125
Person / Time
Site:	Three Mile Island
Issue date:	12/30/1985
From:	Bernero R Office of Nuclear Reactor Regulation
To:	GENERAL PUBLIC UTILITIES CORP.
Shared Package
ML20141G115	List: ML20141G108 ML20141G125 ML20141G157 ML20141G169 ML20141G184
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR NUDOCS 8601090785
Download: ML20141G125 (6)

Text

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Enclosure 1 UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of GENERAL PUBLIC UTILITIES NUCLEAR ) Docket No. 50-320 CORPORATION )

)

(Three Mile Island Nuclear Station )

Unit 2) )

EXEMPTION 1.

GPU Nuclear Corporation, Metropolitan Edison Company, Jersey Central Power and Light Company and Pennsylvania Electric Company (collectively, the licensee) are the holders of Facility Operating License No. DPR-73, which has authorized operation of the Three Mile Island Nuclear Station, Unit 2 (TMI-2) at power levels up to 2772 megawatts thermal. The facility, which is located in Londonderry Township, Dauphin County, Pennsylvania, is a pressurized water reactor previously used for the commercial generation of electricity.

By Order for Modification of License, dated July 20, 1979, the licensee's authority to operate the facility was suspended and the licensee's suthority was limited to maintenance of the facility in the present '

shutdowncoolingmode(44 Fed. Reg.45271). By further Order of the Director, Office of Nuclear Reactor Regulation, dated February 11, 1980, a new set of formal license requirements was imposed to reflect the post-accident condition of the facility and to assure the continued maintenance of the current safe, stable, long-term cooling condition of the facility (45 Fed. Reg. 11292). The license provides, among other things, that it is subject to all rules, regulations and Orders of the Comission now or hereafter in effect. ,

8601090785 B51230 PDR ADOCK 05000320 P PDR

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11.

By letter dated August 27, 1985, the licensee requested exemptions from 10 CFR 50.61 requiring the submission to the U.S. Nuclear Regulatory Commission of projections, analyses, schedules and other steps necessary to protect against pressurized thermal shock events. Specifically, Paragraph (b)(1) of 10 CFR 50.61 requires licensees to submit projected values for

, Reference Temperature for each weld and plate or forging in the reactor vessel beltline and Paragraph (b)(3) requires an analysis and schedule for implementation of a flux reduction program if the projected values of Reference Temperature are expected to exceed the pressurized thermal shock screening criteria set forth in Paragraph (b)(2) of 10 CFR 50.61.

Additionally, the rule requires certain steps be taken if the flux reduction program does not result in reducing the value of the Reference Temperature below that of the pressurized thermal shock screening criteria.

III.

Nuclear plant pressure vessels are fabricated from ferritic steels. A pressure vessel must be designed to maintain fracture toughness of the vessel material for the life of the plant. The pressure vessel of a nuclear plant can be subjected to a pressurized thermal shock (PTS) when an extended cooling transient to the vessel wall is accompanied by primary system pressurization. Under these conditions repeated thermal and pressurization stresses on the internal surfaces of the vessel can cause the formation of cracks. An adequate level of fracture toughness provides assurance that small cracks will not propagate in a " brittle" manner as a result of stresses during an abnormal transient such as a PTS event.

Failure in a brittle manner could fracture the vessel wall and lead to severe failure of the primary pressure boundary in the core area. Due to irradiation damage, older pressure vessels generally have a greater probability of shifting the fracture toughness curve to higher temperatures, thereby increasing the probability of nonductile or brittle vessel failure.

For a pressurized thermal shock to result in a significant nonductile failure the following conditions must be present:

The nuclear plant pressure vessel must exhibit significant loss of fracture toughness through neutron irradiation.

An overcooling transient must occur that is of sufficient duration to cause a steep thermal gradient across the vessel wall and cooling to the low-toughness temperature range.

A flow must be present of sufficient size and be located at a critical vessel beltline location where reduced fracture toughness and high thermal stress exist.

A simultaneous high reactor coolant pressure must be present.

l IV.

The staff has reviewed the past and present condition of the damaged TMI-2 reactor and has determined that:

The plant went critical on March 28, 1978 and went into commercia's operation on December 30, 1978. The accident at THI-2 occurred on March 28, 1979. Neutron irradiation damage to the vessel is minimal.

1 Since the middle of July 1982, the Reactor Coolant System (RCS) has been essentially vented to the reactor building. Since July of 1984, the reactor pressure vessel head has been removed. With the reactor vessel head removed the RCS cannot be pressurized. The licensee has no plans at this time to repressurize the RCS.

As of the middle of September 1985, the incore thermocouple readings range from 70*F to 91*F with an average of 79 F. The average cold leg temperature is 54 F. The incore temperature continues to drop over time. RCS cooling is by natural heat loss to the reactor building ambient atmosphere. No future increase in temperature is expected but rather continued slow cooldown.

1 With the licensee readying for the commencement of fuel removal, the lack of pressure in the RCS and essentially ambient core and RCS temperatures, a pressurized thermal shock is not a credible event. Therefore, the determination of projected values for Reference Temperature for each weld and plate or forging in the reactor vessel beltline and the development of mitigative actions should the Reference Temperature exceed the screening criteria are not warranted. Undertaking the analyses and other actions required by 10 CFR 50.61 would impose an unnecessary burden and expense on the licensee with no concomitant benefit.

V.

Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12, an exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

The Commission hereby grants an exemption from the requirements of 10 CFR 50.61.

It is further determined that the exemption 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. In light of this determination and as reflected in the Environmental Assessment and Notice of Finding of No Significant Environmental Impact prepared pursuant to t

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10 CFR 51.21 and 51.30 through 51.32, issued on December 19,1985, it was concluded that the instant action is insignificant from the standpoint of environmental impact and an environmental impact statement need not be prepared.

FOR THE NUCLEAR REGULATORY COMMISSION

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Robert M. Bernero, Acting Director Office of Nuclear Reactor Regulation Effective Date: December 30, 1985 Dated at Bethesda, Maryland Issuance Date: December 30, 1985 k

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