ML20064D492

Presents Details of Stretch Power Effort in Advance of Execution to Enable Facility to Operate at 2,700 Mwt Following Second Refueling Outage

ML20064D492
Person / Time
Site:	Millstone
Issue date:	11/01/1978
From:	Counsil W NORTHEAST UTILITIES
To:	Reid R Office of Nuclear Reactor Regulation
References
TAC-46174, NUDOCS 7811070196
Download: ML20064D492 (2)

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November 1, 1978 Docket No. 50-336 Director of Nuclear Reactor Regulation Attn:

Mr. R. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Conraission Washington, D. C.

20555

References:

(1)

W. G. Counsil letter to R. Reid dated July 28, 1978.

(2)

D. C. Switzer letter to G. Lear dated January 12, 1978.

Gentlemen:

I Millstone Nuclear Power Station, Unit No. 2 Stretch Power In Reference (1), Northeast Nuclear Energy Company (NNECO) discussed its intent to increase licensed core thermal power from 2560 MWt to 2700 MWt, the FSAR de-sign maximum power level, starting with the beginning of Cycle 3 operation.

Since the date of Reference (1), eff orts have continued regarding detailed develop-ment of specific tasks, evaluations, and analyses which will be performed, as well as the schedule for these efforts. The results of informal discussions with the Staf f have been incorporated into the program. The purpose of this letter is to present the details of the stretch power effort to the Staff in advance of their execution, such that subsequent to the second refueling outage, Millstone Unit No. 2 will be licensed to operate at 2700 MWt.

R is currently anticipated that Cycle 2 could terminate as early as March 10, 1979; the duration of the refueling outage will be approximately eight (8) to ten (10) weeks. Thus, NRC issuance of the refueling license amendment would tentatively be required as early as May 1,1979. Analyses and evaluations associated with a normal refueling are being combined with the atretch power effort and will be submitted in several stages according to the following proposed schedule:

(1) Environmental impact Review - A sectionalized review of the Final Environ-mental Statement, and evaluation of radiological consequences - December 15, 1978..

(2) Non-LOCA Safety Analyses - February 1, 1979.

This submittal would address the scope of a normal refueling ef fort, but i

in greater detail because of the stretch power ef fort. All required Techni-cal Specification changes would be included. Probable results of LOCA analyses would also be discussed. Any relevant information regarding a

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review of the balance-of-plant to support 2700 MWt would be supplied.

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. (3) Formal Large Break LOCA Results - March 15, 1979.

(4) Formal Small Break LOCA Results - April 25, 1979.

Several other topics merit further discussion regarding the stretch power effort at this time. It is NNECO's intention to have operational at the start of Cycle 3 a Reactor Coolant Pump Speed Sensing System (RCPSSS). This system, fully qualified as an addition to the Reactor Protective System (RPS), would replace the current steam generator op system for protection against the four pump loss of flow inci-dent only. This matter was discussed briefly with the NRC Staff in Bethesda, and will be the subject of additional correspondence in the very near future.

With regard to the resolution of the waterhole peaking issue, it is NNECO's under-standing that subsequent to the events summarized in Reference (1), representatives of Combustion Engineering and the NRC have been involved in further technical discussions. It is also NNECO's understanding that assumed uncertainties of 6.0%

and 7.0% for F[ and Ff, respectively, will be acceptable to the NRC Staf f at this time for safety /setpoint analyses performed with the TORC /CE-1 code. For the apptc;riate analyses, this code will be utilized throughout the stretch power effort. The use of assumed uncertainties of 6.0% and 7.0% is designed to preclude further negotiations on this subject on the Millstone Unit No. 2 docket specifically concerning Cycle 3 operation.

Lastly, it is recognized that the proposed schedule for submittal of formal small break LOCA results is not optimized from the perspective of NRC Staff review time.

However, considerable progress has been made in quantifying the peak clad tempera-ture (PCT) results for the limiting small break for Cycle 3.

As reported in Reference (2),2the limiting senll break for Cycle 2 operation was determined to be the 0.05 ft break, yielding a PCT of 1931*F. The use of the Combustion Engineering CEFLASH-4AS Code is expected to identify a limiting break of 0.1 f t2, with a PCT of less than 2000*F. One change from assumptions made in the Reference (2) analysis is that in these preliminary evaluations, credit has been taken for operation of the charging pumps. Conservatively assuming the failure of one diesel generator and postulating the most adverse break location, fif ty percent of the flow from one charging pump is available for core cooling. Note that the three installed charging pumps are safety-related, a minimum of two are auto-matically sequenced onto the diesel generators in the event of a loss of offsite power, and two of the three are required to be operable by Technical Specifications.

It is emphasized that the above information is supplied to advise the Staff of stretch power efforts currently in progress, such that relevant Staff concerns can be addressed in a timely manner. Should the Staff require further amplifica-tion of any of the above items, either in the form of written correspondence or at a meeting, NUSCO and NNECO are prepared to support such efforts.

As there are considerable economic merits associated with this program, your expedi-tious review and comment on the above would be greatly appreciated.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPAh"I l

6 95 W. G. Counsil Vice President