ML18086B254

Discusses Review & Plans Re Cycle 4 Reload Core.Results of Postulated Events Are within Allowable Limits,Tech Spec Changes Not Required & No Unreviewed Safety Questions Involved

ML18086B254
Person / Time
Site:	Salem
Issue date:	01/07/1982
From:	Liden E Public Service Enterprise Group
To:	Varga S Office of Nuclear Reactor Regulation
References
NUDOCS 8201190303
Download: ML18086B254 (3)

Text

PS~G Public Service Electric and Gas Company 80 Park Plaza Newark, N .J. 07101 Phone 201 /430-7000 January 7, 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 4 RELOAD ANALYSIS FACILITY OPERATING LICENSE DPR-70 UNIT NO. 1 SALEM GENERATING STATION DOCKET NO. 50-272 Salem Unit No. 1 has concluded its third cycle of operation, and commenced a refueling outage on January 1, 1982. Cycle 3 operation was terminated with a cycle burnup of 9558 MWD/MTU.

Startup of Cycle 4 is expected to occur in Mid-March 1982.

This letter is to advise you of PSE&G's review and plans re-garding Salem No. 1 Cycle 4 reload core. Cycle 4 is expected to achieve a burnup of 6800 MWD/MTU.

The Cycle 4 reload core will consist of 56 new Westinghouse 17 x 17 fuel assemblies. The assemblies will have an enrichment of 3.40 w/o (see attached figureJ.

A review was performed on Salem No. l Cycle 4 reloa,d core which addressed those incidents analyzed and reported in the S'alerri FSA,R that could potentially be affected by the fuel reloa,d. This review was performed in accorda,nce with the Westinghouse reload methodology as outlined in the March *1978 Westinghouse topica,l report "Westinghouse Reload Safety Evaluation Methodology" (WCAP-9272). All Cycle 4 assemblies are of the same mechanical, nuclea:r and thermal hy_d:talilic design as the cycle 3 assemblies. The total peaking factor envelope is the same for both *cycle 3 a,nd Cycle 4.

There are no changes in the kinetic parameteis from Cycle j to Cycle 4.

The locations of the Optimized Demonstration 'Fuel .A,ssemblie$ are shown in the attachment. The assembly in the H-5 loca,tion is instrumented with a thermocouple. The assembly in the H-1i location is instrumented with a movable incore flux detector.

The criteria with respect to F and F used to determine the L1H Q location of the Optimized Assemblies are as follows:

Thr:- .-82011 qo303 s20101 PDR ADOCK 05000272 p _ PDR ~- -

95-2001 (350M) 4-80

Cycle 4 Reload Analysis January 7, 1982

1. Demonstration assemblies are placed in the core such that the lead power fuel rods operate at least 6% lower in F than the maxium allowed design value for the LlH standard assemblies.

2. Demonstration assemblies are located such th.at they operate with F values at least 0.10 lower that the Q

design value for standard assemblies.

PSE&G has reviewd the bases of the reload analysis and the Westinghouse Reload Safety Evaluation (RSE) Report with Westing-house. The review of all incidents demonstrated that the results of all the postulated events are within allowable limits. The reload safety evaluation demonstrated that Tedhnical Specifica-tion changes are not reguired for operation of Salem Unit 1 at rated thermal power during Cycle 4. Salem's Station Operations Review Committee and the Nuclear Review* Board have concluded that no unreviewed safety questions as defined by 10 CFR 50.59 are involved 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 th~ startup physics testing program. This program will include, but not be limited to, the following tests:

1. Control rod drive tests a,nd drop time

2. Critical boron concentration measurements

3. Control rod bank worth measurements

4. Moderator temperature coefficient measu;r;errierit

5. Power coefficient measurement, and

6. Startup power distribution measurenjerits usi.ng the incore flux mapping system.

Ver~ truly yours,

,,,~/21 / ~/I

./(%~

__,/E

• A. Lid en

/ Manager - Nuclear Licensing & R.egula,tion cc Mr. Leif Norrholm Senior Resident rnspector Mr. Gary c. Meyer Licensing Projedt Manager

I "

• . -ALEM CY~LE Region* 3 - Sta,ndard * 'll 7Xl 7 l (3. 29\'7/o}

4 . . e Region SA - Standard ( l 7Xl 7) (2. 80w/o)

• Region 4 - Standard (17Xl7) (2.80w/6) Region SB - f)ta,nda,;r;-d (J.7X17LL3.41'WfOL

~egion 4 A- Optimized (17Xl7) (2.80w/o) Region 6 7- Sta;ndard (J.7:X17L(3.40.w/oJ_

R p.

• N M K L J H G F E D c B A l 2 3 4 5 6' 7 12 14 i l 13 b lf I I I lr l

6 6 6 4 6 6 6 6 6 6 6 2 6 6 24 3 20 3 20 3 24 6 6 SB 3 6 SB 3 4 SA 3 SS 3 SA 4 3 SB 6 4

6 3 SA SA 4 SA SA SA 4 SA SA 3 6 6 ,__

6 24 4 SA 4 SA 4A SA SA 6 s 4 4 4 4 24 6

._,___ 6 6 3 SA 4 SA 3 SA SA SA 3 SA 4 SA 3 6 6 SB. 6 ......_ 7 6 20 3 SA 4 SA 4 24 4 SA 4 SA 3 20 6 SB SB ,_

8 4 3 SB SA 4 SA 24 3 24 SA 4 SA SB 3 4 6 20 6

3 SA 4 SA 4 SB 24 4 r:;A 4 s:;A ~

6

?n ~

- 9

. SA 1-- 10 6 3 SA J 4 3 SA SA SA 3 SA 4 SA 3 6 i 6 24 6

4 SA 4 SA 4 4A 4- SA 4. SA 4 6

24 6 i---

11 6 3 SA SA 4 SA SA SA 4 SA SA 3 6 12 SB 13 6 SB 3 4 SA 3 SS 3 SA 4 3 SB 6 6 6 6 6 14 6 6 24 3 20 3 20 3 24 6 6 6 6 6 4 6 6 6 15 X Region

• Y .Number of Burnable Poison Rods SS Secondary Source Rods 4A Optimized Demonstration Assemblies