ML19323G433
| ML19323G433 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/16/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323G432 | List: |
| References | |
| NUDOCS 8006020443 | |
| Download: ML19323G433 (5) | |
Text
.
O uu,
,,f, u
UNITED STATES
.g NUCLEAH HEGULA TORY COMMISSION y
g C
WASHING T ON, D. C. 20555
'$9 *....,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 58 TO LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER C0MPANY SURRY POWER STATION UNIT 2 DOCKET NO. 50-281 Introduction By letter dated May 31,1979 (Reference 1), as supplemented October 16, 1979 (Reference 2), October 25,1979 (Reference 3), January 11, 1980 (Reference 4) and February 20,1980 (Reference 5), Virginia Electric and Power Company (the licensee) requested an amendment to Operating License No. DPR-37 for Surry Power Station, Unit 2 References 1 and 2 contain a LOCA analysis and proposed Technical Specification changes in connection with the operation of Unit 2 with 3 percent of steam generator tubes plugged, with mcdifications made during the steam generator replace-ment operation and with a peaking factor Fg of 2.19.
The changes to the Technical Specifications for Unit 2 requested by the licensee are the following:
(a) Change of the heat flux hot channel factor, FQ to 2.19, the limiting enthalpy rise hot channel factor to 1.55 and the corresponding limiting assembly enthalpy rise factor to 1.476 for plant operation -
with 3 percent of steam generator tubes plugged.
-(b) Change of the Hot Channel Factor Normalized Operating Envelope for Unit 2 (TS Fig. 3.12-8h).
(c) Removal of the specifications for the augmented surveillance of core power distribution and change of the axial flux difference limits (TS Fig. 3.12-10).
Since the limiting value of FQ is below the level at which the excore detectors could provide reliable readings, the licensee has analytically predicted the maximum values for the total peaking factor, Fg(Z) using the "3 case analysis"- methodology (Reference 6). These predicted values lie below the LOCA predicted limits (Reference 5) and augmented power distribution surveillance is therefore not required during Cycle 5 operation.
4 8006020 gp
-P.
Evaluation The licensee has provided an evaluation of the performance of Emergency Core Cooling System (ECCS) for Unit 2 corresponding to the hot channel peaking factor value of Fo=2.19 and assuming a steam generator plugging level of 3 percent.
In addition, the following changes in plant operational parameters were introduced into the LOCA analysis:
(a) Change of the Low Head Safety Injection flow due to NPSH consideration.
(b) Modification of the Containment Spray System.
(c)
Change of the containment initial temperature from 90*F to 80'F.
(d) Changes of the steam generator model parameters reflecting the modi-fication caused by the replacement of the steam generators.
g The change in the low safety injection flow was needed in order to meet the NPSH requirement of the LPSI pumps. The flow was limited by means of a venturi flow restrictor. This change resulted in a slightly lower safety injection flow.
However, the licensee has demonstrated (Reference 3) that this flow is still higher than the value assumed in the LOCA analysis.
The modification of the contaiment spray system consisted of adding addi-tional spray header capacity. This additional capacity increased the fill time and resulted in a greater time to actuation of the spray system. The time assumed in the analysis was therefore conservatively changed from 52 to 59 seconds. The additional modification consisted of removing the flow reducing device, which had been installed as a part of the interim NPSH sol ution. This required increasing the assumed pump runout flow rates from 2250 to 3500 gpm. Although this modification caused the actuation time to change from 410 to 365 seconds, the results of the analysis were not affected because peak cladding temperature occurred at a much earlier time.
The reduction of the containment initial temperature from 90 to 80 F is in itself a conservative change in LOCA analysis.
It is consistent with the minimum value of the allowable contaiment temperature range restrictions required by the NPSH considerations.
The changes introduced to the steam generator model parameters were mostly in thermal-hydraulic area and were caused by the modified reactor coolant side pressure drop and by flow area and tube lengthchanges (References 7 and 8).
The magni tude of these changes was small and it did not significantly in-fluence the results of the LOCA analysis, s
The LOCA analysis was performed using the February 1978 version of the Westinghouse Evaluation Model (Reference 9 ) which was reviewed and approved by us.
It was performed for a spectrum of three double ended cold leg guillotine breaks (DECLG) with discharge coefficients of C =0.4, 0.6 and 1.0.
D The input parameters assmed in the analysis are listed below:
i
Core Power:
102 percent of 2441 MWt (rated power).
Peak Linear Power: 102 percent of 13.59 Kw/ft.
Peaking Factor: 2.19 Accumulator Water Volume:
975 cu ft/each The results of the analysis indicate a peak cladding temperature of 2190 F, a maximum local Zr-water reaction of 7.99 percent and a total Zr-water reaction of less than 0.3 percent, all these values occurring at the critical break size of C =0.4 D
The licensee did not include a small break analysis since neither the change in the steam generator tube plugging level, nor the other modifi-cations introduced to the plant affected significantly the results of the original analysis.
The licensee has addressed (Reference 4) the potential impact of the recent concerns related to fuel performance model changes included in draft report NUREG-0630 (Reference 10). The licensee has shown that
' though these model changes by themselves could cause quite significant peak cladding temperature increases,'there are at least two compensating effects which could provide credits required to offset the penalties causing these increases. These effects are due to the changes involving the slip and break flow models which have been approved by us for UHI plants after an extensive review. As a result the effects produced by the fuel performance model change could be excluded from the present ~LOCA analysis without. reducing its degree of conservatism.
The licensee has predicted the maximum values of the total peaking factor reached by the Surry Unit 2 plant during its Cycle 5 operation. The prediction was made using the NRC approved, "3 case FAC" methodology (Reference 11) and the axial flux difference limits specified in the licensee's submittal (Reference 2). The predicted peaking factor, Fg(Z) is below the LOCA determined limit and no augmented core power distri-bution surveillance is required during the Cycle 5 operation. The licensee is therefore justified in removing the augmented surveillance requirement from the plant's Technical Specifications.
Based on our review of the submittal documents, we conclude that the results of the LOCA analysis performed with Fg=2.19 are conservative relative to the 10 CFR 50.46 criteria. We consider the resultant changes to the Technical Specifications acceptable for operating Unit 2 with up to a maximum 3 percent of steam generator tubes plugged.
The submitted LOCA analysis was-reviewed for Unit 2, however, the evaluation could be extended to Unit 1 after the unit is suitably modified to comply with the assumptions made in the analysis, i
._________m
. Environmental Consideration We have determined ~that the amendment does 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 amendment is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR Sl.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with this amendment.
_ Conclusion We have concluded, based on the consideration discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: May 16, 1980
. References 1.
Letter from C. M. Stallings (VEPCO) to H. R. Denton (NRC), Serial No.
388, dated May 31, 1979.
2.
Letter from C. M. Stallings (VEPCO) to H. R. Denton (NRC) Serial No.
820, dated October 16, 1979.
3.
Letter from C. M. Stallings (VEPCO) to H. R. Denton (NRC)
Serial No.
388 A, dated October 25, 1979.
4.
Letter from C. M. Stallings (VEPCO) to D. G. Eisenhut (NRC), Serial No.
039, dated January 11, 1980.
5.
Letter from C. M. Stallings (VEPCO) to H. R. Denton (NRC), Serial No.
388 B, dated February 20, 1980.
6.
Letter from C. Eicheldinger (Westinghouse) to J. F. Stolz (NRC), Serial No. NS-CE-1749, dated April 6,1978.
7.
Letter from C. M. Stallings (VEPCO) to E. G. Case (NRC), Serial No. 351, dated ~ August 17, 1977.
8 Letter from A. Schwencer (NRC) to W. L. Proffitt (VEPCO), dated December 15, 1978.
9.
WCAP-9220, Westinghouse ECCS Evaluation Model, February 1978 Version.
February 1978.
10 NUREG-0630 (Draft), Cladding Swelling and Rupture Models for LOCA Analysis, Novenber 1979.
11, Letter from R. A. Wiesemann (Westinghouse) to J. F. Stolz (NRC), Serial No. AW-78-34, dated April 6,1978.
.