ML19263E823
| ML19263E823 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/09/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19263E822 | List: |
| References | |
| NUDOCS 7906250333 | |
| Download: ML19263E823 (6) | |
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- . 9 :3 3 TC '4 C.,0.;;!!5 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NOS. 49 AND48 TO FACILITY OPERATING LICENSE NOS. DPR-32 AND OPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY NUCLEAR POWER STATION, UNIT N05. 1 AND 2 DOCKET NOS. 50-280 AND 50-281 Introduction By letter dated December 26,1978 (Reference 1) as supplemented January 9,1979 (Reference 2) Virginia Electric and Power Company (the Licensee) requested changes to the Technical Specifications to Operating License Nos. OPR-32 and DPR-37 for the Surry Nuclear Power Station, Unit Nos. I and 2.
The proposed changes are in response to the Exemption and Order for Modification of License issued on June 30,1978 (Reference 3) and April 28,1978 (Reference 4) for Unit Nos.1 and 2, respectively, which were issued as a result of the discovery of an error in tne Zr-water reaction model in the eval-uation model computer codes used in the LOCA analysis (Reference 5).
The proposed Technical Specification changes are supported by the LOCA reanalysis performed with the approved February 1978 version of the ECCS evaluation model (Reference 6) which included the correction of the error and other approved model changes. The reanalysis also included input assumptions of 28 percent of steam generator tubes plugged and reduced accumulator water volume.
It was performed with the total peaking factor, FQ of 2.05.
In addition, the licensee has presented several other related changes to the Technical Specifications dealing with the new limits for core power distribution. Most of these changes were in the conservative direction and, for the one change which decreased the degree of conservatism, the licensee has provided an appropriate justification.
Evaluation On March 21, 1978, an error was discovered in the Westinghouse October 1975 ECCS evaluation model. The error involved the calculated heat generation resulting from the Ir-water reaction and affected the cal-culated cladding temperatures after a LOCA. Following discovery of this error, the licensee administrative 1y reduced the total peaking factor limits for Unit Nos.1 and 2 from Fg=1.85 to Fg=1.79.
This new value ef-FQ was intended to conservatively accommodate the error and 2214 152 7906250 3 9
. was applicable for up to 25 percent of the steam generator tubes plugged (Reference 7). The licensee also committed to provide a new LOCA-ECCS analysis, which was to be performed with an acceptable evaluation model.
As noted in the Order for Modification of License, issued for the Surry Power Station. Unit 2 (Reference 4) the NRC conditionally approved the total peaking factor limit of Fg=1.79, but reouested the licansee to provide, as soon as possible, a valid ECCS reanalysis to confirm the conservatism of this limit.
In response to this request the licensee submitted an interim LOCA analysis applicable to Surry Unit Nos. I and 2 (Reference 8) performed with the October 1975 version of the Westinghouse evaluation model corrected for the Zr-water reaction e rror.
In that analysis, in addition to the error correction, the licensee assumed an FQ of 1.94 and reduced the accumulator water volume to 975 cu ft.
The analysis also included consideration of several other plant specific input assumptions which partially offset the penalty resulting from correction of the Zr-water reaction error.
The submittal was reviewed by us and approved for Unit No. 2 operation subject to the licensee submitting LOCA reanalysis performed with a fully approved version of the ECCS evaluation model. Also, based on the results of this interim analysis, an Exemption was granted for the Surry Power Station, Unit No. I to operate with Fg=1.94 and with the new (reduced) value of accumulator water volume (Reference 3).
On September 13, 1978 in a letter to the licensee (Reference 10) we reiterated our request that the ECCS analysis, performed with a fully corrected and approved model, be provided. On October 11, 1978 (Reference 11) the licensee committed to provide such an analysis by January 1979. On December 26, 1978 the requested analysis was submitted (Reference 1). The analysis was performed with the NRC approved Fecruary 1978 version of the Westinghouse evaluation model (Reference 6) which in addition to including the correction of the Zr-water reaction error and several code maintenance and analytical improvements, con-tained the following changes: modification of the input to the con-tainment code, modified accumulator model, steam dynamic cooling and
.an improved 15 x 15 FLECHT heat transfer correlation. The subniitted analysis was based on the assumptions listed below:
(1) Limiting value of hot channel peaking factor of Fa=2.05 (2) Core inlet temperature of 534.5'F (3)
Initial fuel temperature based on generic values of fuel char-acteristics (4) Modified containment parameters (5) 28 percent of steam generator tubes plugged 2214 153 (6) Accumulator water volume of 975 cu ft.
. The proposed value of FO was justified by the LOCA analysis whi h c
indicated that with FQ=2.05 and with the assumptions listed above the peak cladding temperature and the local and total Zr-water reactions were within the limits set ' orth in 10 CFR 50.46.
The values of these parameters are listed below Peak Cladding Temperature: 2172* F Local Zr-water Reaction:
7.815 Total Zr-Water Reaction:
<0.3%
The core inlet temperature assumed in the analysis represented the best estimate value and was lower by 4.5'F fns.: the value used in the interim analysis (Reference 8). The use of the best estirate inlet temperatures is consistent with our position which accepts the use of these temperatures in LOCA analyses.
In addit on, the current i
analysis was based on generic values of fuel characteristics which were more conservative than the as-built values previously used and therefore would permit the analysis to be referenced for future cores loaoed with fuel having similar fuel characteristics.
Although the containment heat sinks used in the analysis were somewhat different from those assumed in the previous interim analysis, the licensee has shown that they represent in a more realistic manner the plant specific characteristics of the containment. Also, in the analysis the licensee took credit for paint on some of the contain-ment components, the existence of which reduces the flow of heat to the containment structure.
This assumption decreases the steam condensation rates and results in higher calculated containment back-pressures after a LOCA.
We have reviewed the results of the LOCA analysis submitted by the licensee and have concluded that the safe operation of the plants with FQ1 05 and steam generator tube plugging levels of 128 percent 2
has been adequately demonstrated. We concur therefore with the following proposed changes to the Technical Specifications:
(1)
Peaking factor change from F -1.94 for Unit No. I and F =1.79 Q
Q for Unit No. 2 to F =2.05 for both units, and Q
(2) Accumulator water volume change from 1075 cu ft minimum and 1089 cu ft maximum to 975 cu ft minimum and 989 cu ft maximum for both units.
Although the ECCS analysis demonstrates compliance with the regulations, the amount of steam generator plugging for Unit No. 2 steam generators, which are being refurbished, will be limited to 5% at this time.
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. Since the new value of total nuclear peaking factor would remain below 2.32, the licensee has provided for Unit No. I an applicable "18 case FAC analysis" (Reference 12) which indicates that the total nuclear peaking factor would not exceed the value of 2.05 during normal plant operation, including load follow maneuvers and therefore has justified that the use of APDMS is not required. For Unit No. 2 the licensee is committed to use APDMS surveillance.
In addition to the changes resulting directly from the LOCA reanalysis, tha licensee has proposed other Technical Specification changes which are related to the power distribution limits it. the core. These changes are listed below:
(1) Removal of the steam generator tube plugging limits below which values specified for the maximum assembly and rod enthalpy rise factors are not applicable.
(FAH/assm. andaH/rud )
LOCA F
LOCA (2) Change in the procedures for evaluating F (Z) during plant operation.
n For Unit No. 2, where the APDMS surveillance is required, the and the values for F (Z) hp(ec)ified in Table 3.12-1B would BEY (Z) procedure for obtaining F Z from the measured values of F retained. For Unit i o. I which we condude does not require APDMS surveillance, this procedure would be deleted.
(3) New limits for the axial flux difference. This proposed change would specify the new limits for the allowable axial flux differ-ence by:
(a) assigning lower value (88%) for the reactor power level above which the axial flux difference must be maintained within a +5% target band; (b) defining the operational conditions for which axial flux difference could remain outside the pre-scribed limits without reduction of reactor power, and (c) specifying new allowable axial flux difference limits for reactor operation below 88% of its rated power.
With the exception of the new allowable axial flux difference limits, all the changes listed above would make the Technical Specifications more restrictive.
The last change extends the allowable axial flux difference limits and results from a plant specific analysis (Reference
- 2) performed by the licensee to justify this change.
The current allowable axial flux difference limits were derived by a Westinghouse generic analysis which resultec in limits more restrictive than the
- orocosed limits resulting from the analysis cerforec specifically for Surry Units 1 and 2.
The croposed limits are based on plant specific parameters and are conservative. The orocosed limits, derived by previously approved methods, assure that the power distribution in the core will be maintained within specific bounds (F, FaH) and all Q
safety limits ano criterla will be met. We conclude that because this change does not involve a significant increase in the probability or consequences of accidents creviously considered and does not involve a significant decrease in a safety margin, it does not involve a significant hazards censideration.
2214 155 Based on the review of the submitted documents, we conclude from the results of the ECCS reanalysis performed with the creviously approved February 1978 version of the Westinghouse evaluation model, that oceration of Surry Unit Nos.1 and 2 at a peaking factor limit of 2.05 with reduced accumulator water volume will be in conformance with the 10 CFR 50.46 criteria. We consider the ECCS analysis and all the changes to the plant Technical Specifications resulting from this analysis and from the proposed modifications of the core power distribution limits acceptable for operating the plant with up to a maximum of 28 percent of steam generator tubes plugged. Unit No. 2, which will return to operation with refurbished steam generators that have no plugged tubes, will be limited for the present to 55 plugging.
Environmental Imoacts of procosed Action The plugging of 3 to 5 percent of the steam generator tubes as would be allowed by the proposed action would impact the total plant occupational exposure. We have estimated the total exposure resulting from plugging each additional 1 percent of the tubes to be 92 man-rems Our estimate is based on the licensee's historical data per reactor.
regarding total number of tubes plugged and occupational exposure.
For 3 to 5 percent, the estimated exposure would be 275 to 459 man-rems /per reactor. We have examined other pWRs that have plugged a significant number of tubes and find that experience, as a function of percent of tubes plugged, is typical. Therefore, we conclude that there will be no significant environmental impact associated with the proposed action.
Environmental Consideration We have determined that these 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 involve an action which is insignificant from the stand-point of environmental impact and, pursuant to 10 CFR 51.5(d)(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.
CONCLUSION 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 Comission's regulations and the issuance of these ame-cents will not be inimical to the common defense and security or the health and safety of the public.
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~ References 1.
Letter from Vepco (C. M. Stallings) to NRC (H. R. Denton), dated December 26, 1978, Serial No. 736, transmitting: ECCS Analysis (attachment 1); Proposed Technical Specifications (attachment 2).
2.
Letter from Vepco (C. M. Stallings) to NRC (H. R. Denton), dated January 9,1979, Serial No. 019.
3.
Letter from NRC (A. Schwencer) to Vepco (W. L. Proffitt), dated June 30,1978 transmitting Exemption to 10 CFR 50.46 (a) (1) for Surry Power Station, Unit 1.
4.
Letter from NRC'(A. Schwencer) to Vepco (W. L. Proffitt), dated April 28,1978, transmitting Order for Modification of License for Surry Power Station, Unit 2.
5.
Letter from Westinghouse Electric Corporation NS-CE-1751 (C.
Eicheldinger) to NRC (J. F. Stolz) dated April 7,1978, transmitting LOCA-ECCS Analysis with Zirc/ Water Reaction Correction.
6.
WCAP-9220-P-h, Westinghouse ECCS Evaluation Model, February 1978 Version, February 1978.
7.
Letter from Vepco (C. M. Stallings) to NRC (A. Schwencer) dated April 7,1978, Serial No.197.
8.
Letter from Vepco (C. M. Stallings) to NRC (E. G. Case), dated May 26, 1978 Serial No. 303, transmitting large Break LOCA-ECCS Reanalysis for Surry Power Station, Units 1 and 2.
9.
WCAP-8622 (Proprietary), WCAP-8623 (Nonproprietary), " Westinghouse ECCS Evaluation Model-October 1975 Version," November 1975.
- 10. Letter fmm NRC (A. Schwencer) to Vepco (W. L. Profitt), dated September 13, 1978.
- 11. Letter fmm Yepco (C. M. Stallings) to NRC (H. R. Denton), dated October 11, 1978.
- 12. Letter from Vepco (C. M. Stalling) to NRC (E. G. Case), dated April 4,1978, Serial No. 1 71-032-78.
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