ML18087A668
| ML18087A668 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/31/1983 |
| From: | Liden E Public Service Enterprise Group |
| To: | Varga S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18087A669 | List: |
| References | |
| LCR-83-02, LCR-83-2, NUDOCS 8302180435 | |
| Download: ML18087A668 (7) | |
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e PS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Nuclear Department Ref. LCR 83-02 January 31, 1983 Director of Nuclear Reactor Regulation
- u. s. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20014 Attention:
Mr. Steven A. Varga, Chief Operating Reactors Branch #1 Division of Licensing
Dear Mr. Varga:
REQUEST FOR ADMENDMENT AND CYCLE 2 RELOAD ANALYSIS SALEM GENERATING STATION UNIT NO. 2 DOCKET NO. 50-311 Salem Unit No. 2 has concluded its first cycle of operation and commenced a refueling outage on January 21, 1983.
Cycle 1 achieved a burnup of 16697 MWD/MTU.
The Cycle 2 startup is presently scheduled for late May and is expected to achieve a burnup of 9500 MWD/MTU.
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The purpose of this letter is to inform you of the Cycle 2 fll)O design, the results of the Reload Safety Evaluation, and to J I A' request amendm~nt of the Unit 2 Technical Specifications.
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The Cycle 2 reload design*'* will incorporate 72 new Region 4 feed/,oo0*00 fuel assemblies with an enrichment of 3.4 w/o.
The Cycle 2 reload pattern is attached.
The mechanical design of the Region 4 assemblies is the same as the Region 3 assemblies except that the Region 4 backfill pressure is 3700 psig and a grid corner modification was implemented to minimize fuel handling problems.
The Cycle 2 Reload Safety Evaluation (RSE) was performed by Westinghouse utilizing the methodology described in WCAP-9273, "Westinghouse Reload Safety Evaluation Methodology".
This evaluation identified the need for two amendments to the Unit 2 8302180435 830181 -
PDR ADOCK 05000311 P
PDR The Energy People 95-2168 (80 M) 11-82
Mr. Steven A. Varga, Chief U.S. Nuclear Regulatory Commission 1/31/83 Operating License.
The proposed amendments include changes to the Fxy and Rod Insertion Limits.
The requested amendments are transmitted as an attachment to this letter.
The Cycle 2 RSE also assumes the approval of a previously submitted request for amendment (LCR 82-05, October 5, 1982) to the F 6H limit which would change the power dependent multiplier from 0"2 to 0.3.
Assuming the approval of the above requests for amendment, the RSE addressed those incidents analyzed and reported in the FSAR which could potentially be affected by this reload.
PSE&G has reviewed the results of the Westinghouse RSE with Westinghouse.
Based on this review and independent PSE&G calculations, PSE&G concludes that the effects of the reload on the design basis and postulated incidents analyzed in the FSAR were accommodated within the conservatism of the assumptions used in the previous applicable safety analyses.
The dropped RCCA event was analyzed according to the new Westinghouse Dropped Rod Methodology.
The results show that the DNB design basis is met for all dropped rod events initiated from full power so that the interim operating restrictions are no longer necessary.
However, until formal NRC notification is received to remove them, the plant shall continue to operate under the interim restrictions.
As described above, the Cycle 2 Reload Safety Evaluation is predicated on the NRC approval of these amendments to the Facility Operating License.
In accordance with the Atomic Energy Act of 1954, as amended, and the regulations thereunder, we transmit as attachments to this letter copies of our request for amendment (LCR 83-02) and our analysis of these changes to Facility Operating License DPR-75 for Salem Unit 2.
Also transmitted is the Unit 2, Cycle 2 Peaking Factor Limit Report associated with the proposed Fxy amendment.
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
- 4.
Moderator temperature coefficient measurement
- 5.
Power coefficient measurement, and
- 6.
Startup power distribution measurements using the incore flux mapping system
Mr. Steven A. Varga, Chief U.S. Nuclear Regulatory Commission 1/31/83 This change is deemed to not involve a significant hazards consideration and is, therefore, a Class III Amendment as defined in 10CFR170.22.
A check in the amount of $4,000 is enclosed.
This submittal includes three (3) signed originals and forty (40) copies.
Attachments CC:
Mr. D. Fischer Very truly yours, Liden Manager -
Nuclear Licensing and Regulation NRC Licensing Project Manager Mr. L. Norrholm NRC Senior Resident Inspector
Ref. LCR 8 3-0 2 STATE OF NEW JERSEY SS.
COUNTY OF SALEM COUNTY OF SALEM RICHARD A. UDERITZ, being duly sworn according to law deposes and says:
I am.a vice President of Public Service Electric and Gas Company, and as such, I find the matters set forth in our Request for Amendment LCR 83-02 dated January 31, 1983, are true to the best of my knowledge, information and belief.
~HARD A. UDERITZ7 "Otarypublic of New Jersey My Commission expires on
PROPOSED CHANGES HEAT FLUX HOT CHANNEL FACTOR -
FQ(Z), AND CONTROL ROD INSERTION LIMITS TECHNICAL SPECIFICATIONS SALEM UNIT NO. 2 Descriptlon of Change Ref. LCR 83-02 Revise Section 4.2.2.2e of the Fxy surveillance specifica-tion to ref er to A Radial Peaking Factor Limit Report for cycle specific F~JP limits.
A copy _of the revisions to page 3/4 2-7 of the Technical Specifications is attached.
Revise Section B 3/4.2.3 ot the Basis for Power Distribu-tion Limits to refer to A Radial Peaking Factor Limit Report for cycle specific F~JP limits.
A copy of the revisions to page B,3/4 2-5 of the Technical Specifications is attached.
Add Section 6.9.1.10 of Administrative Controls to describe the contents and distribution of the Radial Peaking Factor Limit Report.
A copy of the revisions to page 6.-17 of the Technical Specifications is attached.
Change Figure 3.1-1 of Section 3.1.3.5, Control Rod In-sertion Limits, to reflect a reduction in the allowable insertion for cycle 2.
A copy of the revisions to page 3/4 1-21 of the Technical Specifications is attached.
Ref. LCR 83-02 Reason for Changes The Fxy(Z) limits represent surveillance limits and not a limiEing condition for operation.
These surveillance limits are associated with the radial peaking factors used in the cycle-by-cycle reload safety evaluation9.
Comparisons of measurements to these limits therefore serve as a form of design verification.
The specific changes for Cycle 2 reflect the normal variation usually observed from initial cycle to the first reload~
The implementation of the Radial Peaking Factor Limit Report provides a mechanism to accomplish cycle-by-cycle design variations without the need to request additional changes to the Technical-Specifications.
The change in the Power Distribution Insertion Limits for Cycle 2 is needed to maintain the Nuclear Epthalpy Hot Channel Factor, FlH within Technical Specification Limits.
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Ref. LCR 83-02 Safety Evaluation The proposed Fxy Technical Specification change does not change the probability of occurrence or the consequence of an accident since the change involves the surveillance of Fxy as a verification of the design models.
This change does not require any equipment or hardware changes.
Second, the possibility for any accident or malfunction of a different type than previously evaluated is not generated.
The actual margin of safety as defined in the basis for the FQ technical specification remains unchanged, since the Fg limit is unchanged.
The radial peaking factor Fxy(ZJ is measured periodically to provide assurance that the hot.channel factor ~Q(Z) remains within its limit.
The proposed change to the Power Dependent Insertion (PDIL) Technical Specification does not change the pro-b~bility of occurrence or the consequence of ~n accident since the change does not introduce new reactor conf igu-rations.
This change does not require any equipment or hardware changes.
Secondly, the possibility for any accident or malfunction of a different type than previously evaluated is not generated.
The change to the power dependent insertion limits is proposed in order to assure that the margin of safety as defined in the Technical Specification basis is main-tained for power peaking.
Additionally, margins for shutdown and initial conditions for a postulated ejected rod accident as described in the Final Safety Analysis Report are maintained.
It is therefore concluded that the proposed Te~hnical Specification changes do not create any undue risk to the health and safety of station personnel or the general public.
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