ML18087A528

Informs That Review & Plans Re Cycle 5 Reload Core Is Expected to Achieve Burnup of 11500 Mwd/Mtu.Startup Physics Testing Will Include Control Rod Drive Tests & Droptime & Critical Boron Concentration Measurements

ML18087A528
Person / Time
Site:	Salem
Issue date:	10/14/1982
From:	Liden E Public Service Enterprise Group
To:	Varga S Office of Nuclear Reactor Regulation
References
NS-EPR-3545, NUDOCS 8210220275
Download: ML18087A528 (4)

Text

PS~G Public Service Electric and Gas Company P.O. Box E Hancocks Bridge, New Jersey 08038 Nuclear Department Ref 1) NS-EPR-3545, 1/20/82 October 14, 1982 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Steven Varga, Chief Operating Reactors Branch #1 Division of Licensing Gentlemen:

CYCLE 5 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Salem Unit No. 1 is expected to conclude its fourth cycle of operation and commence a refueling outage on October 15, 1982.

It is anticipated that Cycle 4 will achieve a cycle burnup of 6000 MWD/MTU. Startup of Cycle 5 is scheduled to occur in mid-December. This letter is to inform you of PSE&G' s review - - -

and plans regarding Salem No. 1 Cycle 5 reload core which is expected to achieve a burnup of 11500 MWD/MTU.

The Cycle 5 reload core will consist of 52 new Westinghouse 17xl7 fuel assemblies. The assemblies will have an enrichment of 3.4 w/o (see attached figure).

A review was performed on Salem No. 1 Cycle 5 reload core which addressed those incidents analyzed and reported in the Salem FSAR that could potentially be affected by the fuel reload.

This review was performed in accordance with the Westinghouse reload methodology as outlined in the March 1978 Westinghouse topical report "Westinghouse Reload Safety Evaluation Method-ology" (WCAP-9272). All Cycle 5 assemblies are of the same mechanical, nuclear and thermal hydraulic design as the *cycle 4 assemblies. The total peaking factor envelope is the same for both Cycle 4 and Cycle 5. Kinetic parameter values for Cycle 5 remain within the bounds of the current limits *

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Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 10/14/82 Since all parameters for Cycle 5 remain within the bounds of the current limits, no accident reanalysis was required. How-ever, the dropped RCCA event was analyzed according to the new dropped rod methodology described in Reference 1. Results show that the DNB design basis is met for all dropped rod events in-itiated from full power so that the interim operating restric-t ions are no longer necessary. However until formal NRC noti-fication is received to remove them, the plant shall continue to operate under the interim restrictions.

The locations of the Optimized Demonstration Fuel Assemblies are shown in the attachment. Both are instrumented with a movable incore flux detector and thermocouple.

The criteria with respect to F6H and F Q used to determine the location of the Optimized Assemblies are as follows:

1. Demonstration assemblies are placed in the core such that the lead power fuel rods operate at least 6%

lower in F6H than the maximum allowed design value for the standard assemblies. .

2. Demonstration assemblies are located such that they operate with Fo values at least 0.10 lower than the design value for standard assemblies.

PSE&G has reviewed the bases of the reload analysis and the Westinghouse Reload Safety Evaluation (RSE) Report with West-inghouse. The review of all incidents demonstrated that the results of all the postulated events are within allowable limits. The reload safety evaluation demonstrated that Tech-nical Specification changes are not required for operation of Salem Unit 1 at rated thermal power during Cycle 4. Salem's Station Operations Review Committee has concluded that no un-reviewed safety questions as defined by 10CFR50.59 are in-volved with this reload. Therefore, based on this review application for amendment to the Salem Unit 1 operating license is not required.

The reload core design will be verified during the startup physics testing program. This program will include, but not be limited to, the following tests:

1. Control rod drive tests and drop time

2. Critical boron concentration measurements

3. Control rod bank worth measurements

Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 10/14/82

4. Moderator temperature coefficient measurement

5. Power coefficient measurement, and

6. Startup power distribution measurements using the incore flux mapping system.

Very truly yours,

~

A A. Liden nager - Nuclear ensing & Regulation Attachment CC: Mr. Leif Norrholm Senior Resident Inspector Mr. William Ross Licensing Project Manager

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