ML20137W918
| ML20137W918 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 09/25/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20137W916 | List: |
| References | |
| NUDOCS 8510040419 | |
| Download: ML20137W918 (4) | |
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' NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555
} SAFETY EVALUATION BY THE OFFICE OF RELATED TO AMENDMENT NO.45 TO FACILITY OPERATING LICENSE NPF-12 SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 I.
INTRODUCTION By letter dated March 6,1985 (Ref.1), South Carolina Electric and Gas Company submitted a request for an amendment to the Virgil C. Summer Technical Specifications to reflect a thermal design flow reduction of 1.9%.
Revised calculations supporting but not changing this request were submitted by letter dated April 30, 1985, (Ref. 2). By letter dated August 9, 1985, (Ref. 3), a note was added to the Technical Specifications bases section to clarify that the available generic design margins are what offset the penalty associated with the thermal design flow reduction of 1.9%.
This note did not substantially change the amendment consisting of a thermal design flow reduction of 1.9% as noticed (50 FR 16014) on April 23, 1985, but simply makes clear that the flow reduction is covered by the available margins in the design calculations. Therefore, this amendment request was not renoticed.
II.
EVALUATION LOSS-0F-COOLANT-ACCIDENT ANALYSIS The licensee provided an evaluation on the effect of reduced design reactor coolant flow on the postulated loss-of-coolant-accident (LOCA).
The licensee only analyzed and evaluated double ended cold leg guillotine (DECLG) breaks since these breaks were identified previously as limiting cases that result in the highest peak cladding temperature. The DEClG break analyses were perfomed with 102% of design thermal power of 2775 MWt and total. peaking factor of 2.32.
A discharge coefficient of 0.4 was used for the limiting case analysis since the sensitivity study shows that the DECLG break with a discharge coefficient of 0.4 results in the highest peak cladding temperature.
The analyses were perfomed by using a, modified version of the 1981 Westinghouse ECCS evaluation model (Ref. 4). This evaluation model uses the Standard PAD Fuel Thermal Safety Model (Ref. 5) for the calculation of the initial fuel rod conditions, the SATAN-VI code for the thermal-hydraulic transient analysis for the RCS during blowdown, the WREFLOOD code for the entlysis of the refill and reflood transient period, the COCO code for the containn.ent pressure transient, and the LOCTA-IV code for the calculation of the peak cladding temperature.
The modified version of the ECCS evaluation model uses the approved BART code (Ref. 3) to calculate the reflood heat transfer coefficient normally performed by the WREFLOOD code. This code takes no credit for the effects of the grids in increasing reflood heat transfer.
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. The staff has reviewed the large break LOCA analysis. The calculated peak cladding temperature is 2189.2*F, the maximum local metal water reaction is 5.7% and the total core metal-water reaction is less than 0.3 percent. We, therefore, conclude that the results presented are acceptable since the approved methods and computer codes were used and the analytical results show that the peak cladding temperature, metal-water reaction and clad oxidation are within the acceptance criteria of 10 CFR 50.46.
THERMAL-HYDRAULIC DESIGN Since the proposed changes to Technical Specifications involve the decrease of thermal design flow of 1.9%, the impact of operating at this lower flow on thennal margin is evaluated.
The licensee has detennined that 1.9% flow reduction will result in a DN8R penalty of 3.0%.
This is derived from using a previously approved sensitivity factor for the rate of change of DNBR with respect to the flow reduction and is acceptable.
The licensee has also recalculated the rod bow penalty on DNBR by using the approved method (Ref. 6). The maximum calculated rod bow penalty is 2.3% for fuels in the Sunener reactor core.
Since the W-3 correlation was used to establish the operating DNBR limit for the Sunsner reactor core, the generally approved DNBR margin of 9.1%
is applicable to the core. This margin is sufficient to compensate for the 2.3% rod bow penalty and penalty of 3.0% DNBR associated with the reduced design flow.
The.1[censeehasalsoevaluatedtheimpactofthereduceddesignflow on'ONB and non-DNBR related transient responses. As a result of the evaluation the licensee concludes that, even with the design flow reduced by 1.9%, the FSAR conclusion that no safety criteria will be violated during transients remains valid.
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Based on our review of the lfcensee's evaluation process and results, we conclude that the reduction of design flow by 1.9% is acceptable for the transient responses.
TECHNICAL SPECIFICATIONS The specific Technical Specification changes and the reasons for their acceptability are:
Table 2.2-1 This table has been modified to include the reduction by 1.9% for the loop design flow. This change is supported by the analysis for safe operation of the core and is acceptable.
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. Figure 3.2-3 This figure has been modified to remove the operating area which has the dependence of operating power and reactor coolant flow on rod bow penalty. The changes cause the operating band to be more restrictive compared with the previously approved operating band and the changes are acceptable.
Changes in pages 83/4.2-4, B3/4.2-5, 3/4.2-8, 3/4.2-9, 2-2 and 3/4.2-11 are editorial. The changes are consistent with the changes related to the removal of dependence of operating band on rod bow penalty (as shown in Figure 3.2-3) and are acceptable.
In conclusion, the staff has reviewed the proposed changes to the Summer Plant Technical Specifications involving a reduction of thermal
' design flow by 1.9% and finds that they are acceptable.
REFERENCES 1.
I.etter from O. W. Dixon, Jr. (SCE8G) to H. R. Denton (NRC) dated March 6, 1985.
2.
Letter from O. W. Dixon, Jr. (SCE&G) to H. R. Denton (NRC) dated April 30, 1985.
3.
I.etter from O. W. Dixon, Jr. (SCE&G) to H. R. Denton (NRC) dated August 9, 1985.
4 I.etter from C. O. Thomas (NRC) to E. P. Rahe (W), " Acceptance for
_ Jeferencing of licensing Topical Report WCAP-Y561 - BART A-1: A Computer Code for Best Estimate Analysis of Reflood Transients,"
dated December 21, 1983.
5.
E. P. Rahe, WCAP-9220: Westinghouse ECCS Evaluation Model, 1981 Version, Revision 1, 1981.
6.
Letter from J. F. Stolz (NRC) to T. M. Anderson (W), " Review of WCAP-8720, Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations."
7.
J. Skaritka, WCAP-8691 (Revision 1), " Fuel Rod Bow Evaluation, dated July 1979.
III.
ENVIRONMENTAL CONSIDERATION This amendment involves a change in the use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no signift-cant increase in individual or cumulative occupational radiation i
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~ exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Sec. 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
l IV. CONCLUSION l
The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Register (50 FR 16014) on April 23, 1985, and consulted with the state of South Carolina. No public comments were received, and the state of South Carolina did not have any comments.
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regula-tions and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Jon B. Hopkins, licensing Branch No. 4 DL Summer B. K. Sun, Core Performance Branch, OSI Dated: September 25, 1985
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