ML20041D180
| ML20041D180 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/05/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20041D179 | List: |
| References | |
| NUDOCS 8203040443 | |
| Download: ML20041D180 (3) | |
Text
{{#Wiki_filter:e 1 f[ o,, UNITED STATES y g NUCLEAR REGULATORY COMMISSION 7; j WASHINGTON, D. C. 20555 o,
- p SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NOS. 3 7 AND 17 TO FACILITY OPERATING LICENSE NOS. NPF-4' AND NPF-7
_ VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA PCWE'R ST'ATION, UNITS NO. 1 AND NO. 2 00CK'ETtNOS. 50-338 AND 50-339 e
== Introduction:== By lett'er dated December 15,1981 (Serial No. 627A), the Virginia Electric. and Power Company (the licenseel. requested changes to the Technical Specifications (.TS1 for the Nortit Anna Power Station, Units 1 & 2 (NA-1&2). The proposed changes would remove specific values of the Fx and-the axial power distributian surveillance linili:s, Pm, from the NA y&2'P - ~ 1 TS. The specific '/alues for thes,e limits would be provided to the NRC-in a special Core Surveillance Report 60 days prior to a reload cycle { startup for either NA-1 or NA-2. Discussion: The TS contain limits on the total heat flux peaking factor FgxK(Z), which are. established by the Loss of Coolant Accident.(LOCA) anal.ysis. A' specific evaluation is made during each reload analysis as to whether the analytically predicted total heat flux peaking factor (F ) values as a Q function of core height are below the FQxK(Z) limit curve. The NRC approved Westinghouse methodolog'y used in the generation of the analytically predicted FQ values is well established, and this methodology has been used and approved in almost all of the safety reload evaluations for Westinghouse reactofs in the past-several years. For predicting the FQ values, load following calculations are performed for the axia'l power distribution. These calculations are combined with the predicted axially dependent maximum values of the horizontal plane peaking factor, F Q fall below tNe, to obtain,, the Fg values.FgxK(Z) limit, then the operation If the predicted values x of F will not exceed the peaking factor limits assumed in thq LOCA analysis. Power distribution TS have been written to ensure that facility' operation is in conformance with' the peaking factor analysis. Most of these TS are not cycle dependent and, therefore, will not be discussed in this safety evaluation. 8203040443 820205 M PDR ADOCK 05000338 P PDR t
- M ~ 2-Two of the parameters spec,ified in the NA-l&2 TS are a function of FQ values predicted for each ' fuel cycle, and are,therefore, the subject of 4 the licensee's proposed change. The first is the set of axially dependent Fxy values. The second is an axial power distribution limit, Pm, expressed as a percentage of full power level. When the Pm limit is exceeded, axial power distribution surveillance. is required. This surveillance occurs in those cases where the predicted Fg exceeds the F xK[Z1 limit curve. In such cases, adherence to the 0 assumptions in the predicted peaking factor analysis will not ensure that the operating Fg value wi,ll' not' exceed the F xK(Z) limit. To ensure that Q this limit will not Se exceeded, the axial power limit, Pm, is designated ,as a power level fraction' which is the limiting ratio of the predicted 'Fg values to the FgxK(Z1 limit curve as a function of core elevation. Above the power level, Pm, conformance with the F xK(Z) limit is ensured Q by axial power distribution surveillance which measures the actual Fg values in the reactor as a function of core elevation. Axial power distribution surveillance requirements have been stipulated in the NA-l&2 TS since initial issuance as Appendix A to the facility operating licenses for NA-l&2 The licensee's proposed changes would..r.emove the specific values of Fxy-and Pm 'from the NA-1&2 TS. - Both quantitidsiwill remain in the TS as generalized parameters.. However, a Section 6.9.1;10 would be added to the NA-l&2 TS which would require a Core Surveillance Report be submitted to the NRC 60 days prior to a specific need for NA-1&2. This report would include the specific values for Fxy, P, and the results m of the F0 analysis supporting the values for F and Pm. The report would be specified to be sent to the Regional Adminis5rator, Region 2 with a copy to be sen.t to the Chief of the Core Performance Branch in the Office of Nuclear Reactor Regulation. Existence of the Core Surveillance Rep 3rt would allow cognizant NRC personnel to maintain records of trends ~ in the affected parameters, and to request further information should any concerns develop. 1 Evaluation: The licensee',s proposed changes As discussed above are administrative in - nature because the affected parameters, F and P, will continue to be xy m specified and will be determined by the same NRC approved Westinghouse methodology used in prior approve,d changes. Therefore, these changes do not result in any unreviewed safety question and do not involve a signifi-cant decrease in a safety margin. Based on the above, we find these changes to be acceptable. ~ only (since Pm was not cycle dependent) Also, a similar change, but for F [mented in the TS for the Joseph M. Farley has already been approved and impf Nuclear Plant, Unit No. 2. e w e e nes 4 ~1~ see -w =
3- ~ The proposed changes require that several NA-l&2 TS be modified where the parameters F y, Pm, and a Figure 3.2-1 (a curve which is a function of x Pm) are speciTied. All of these changes have Been reviewed and are acceptable. Finally, the licensee's ' December 15, 1981 submittal has Surveillance Reports for NA-1&2 with the current F xK(Z)provided Core 2.10, and for a proposed limit of FnxKLZ) where En=Q.14. limits where Eq= 2 The report for. the current limit (f =2.101. is acceEtable for impYementation on the date 0 that the above proposed changes. Secome effective by issuance of the respective license amendments for NA-l&2. The reports for the p.oposed FgxKCZ1 limit where Fg=2.14 (not yet approved) are acceptable for use in the present NA-1&2 operating cycles 60 days after the date they were sub-mitted, which was December 15, 1,981. Environmental Consideration: We have determined that the amendments do 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. Having made this determination,. we have further concluded that the. amendments.jnvolve an action whichris insignificant from the standpoint of environmental impact and, pursuant-to 10 CFR 551.5(dl(4), that an environmental' impact statement or negative declaration and environmental impact appraisal need not be prepared in. ~ connection with the issuance of'these amendments. ,Concl usion: We have concluded, based on the considerations discussed above, that: (1).-.because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public. Dated: FEB .: ES2 Principal Contributors: M. 5. Dunenfeld L. B. Engle 8 9 g m r e =}}