ML20090D608
| ML20090D608 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/03/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20090D606 | List: |
| References | |
| NUDOCS 9203060337 | |
| Download: ML20090D608 (4) | |
Text
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c v.as. % c m N o c e m s hfl SAFETY EVAlVATION BY THE OFFICE OF NUCLEAR REAC10R REGULATION RELATED TO AMENDMENT NO. 153 ffA GLITY OPERAllNG llCENSE NO. NPT-4 VIRGINTA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATIDN. UNIT _J0,.1 DOCKET N0_,_ 50-33B
1.0 INTRODUCTION
By letter dated January 28, 1992, as supplemented February 27, 1992, the Virginia Electric and Power Company (the licensee) proposed changes to the Technical Specifications (TS) for the North Anna Power Station, Unit No I (NA-1).
Specifically, the proposed changes would increase the -team generator tube plugging (SGTP) limit value up to 35% for the most restrict 1.a SG.
The proposed changes to the operating license would limit maximum reactor power to 95% of rated thermal power for the interim period of c,:cration until SG replacement in 1993, by adding a footnote to license condition 2.D.(1),
Maximum Power level, which states that maximum reactor power level shall be limited to 95% of rated thermal power for the period of operation until SG replacement in 1993.
The proposed cb rges to the TS would also impose more restrictive equipment operability requirements for the Emergency Core Cooling System (ECCS) by adding a footnote to Action Statement "a" of TS 3.5.2, "ECCS Subsystems - Tavg greater than 350*F," which requires that the charging pump in each ECCS subsystem be operable to comply with the requirements of the action statement if either low head safety injection pump is inoperable.
These proposals are necessary to accommodate the interim effects of increased SGTP on the large break loss of coolant accident (LOCA) analysis.
NA-1 is currently involved in a mid-cycle SG inspection outage. An extensive eddy current inspection of the NA-1 SG tubes is being performed using conservative analysis guidelines and plugging criteria.
A substantially increased number of tubes are expected to be plugged.
By letter dated Febrc.ry 27, 1992, the licensee requested that the amendment be issued on March 3, 1992, but noted that the 30-day notice period does not end until March 6, 1992. However, the steam generator tube inspection and plugging processes have been performed more rapidly than expected, and NA-1 is now scheduled to restart on March 3, 1992.
In addition, NA-2 was shut down on february 26, 1992, and Surry Unit I was shut down on February 28, 1992.
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1 If the amendment is not issued to support a timely startup of NA-1, the licensee could be faced with a potentially adverse power supply situation with three of the four nuclear units out of service.
Due to these changed circumstances, the staff has determined that the amendment can be issued prior to the end of the 30-day notice period.
2.0 fjALUATION There are a number of areas of plant design which are potentially impacted by the operation with extended SGlP. Westinghouse performed reviews of components and systems within their design responsibility to confirm that operation with the proposed conditions remain in compliance with the applicable codes and standards. Westinghouse concitled that all Nuclear Steam Supply System (NSSS) and components will remain within the bounds of existing design analysis results for operation with up to 40% of the tubes plugged in any or all SGs. Stone & Webster Engineering Corporation evaluated balance of plant (B0P) systems and components to determine the effect of extended SGTP operation.
They concluded that the effect on operation with extendeo SGTP will remain within the bounds of existing design analyses for operation with up to 37% average SGTP.
The licensee assessed the impact of exnded SGlP operation upon the NSSS accident analyses. With the exception of the large break LOCA, the existing analyses are valid for operation of NA-1 at rated ti.crmal power of 2893 MWt with up to 35% SGTP in any or all SGs.
The licensee performed a reanalysis of the ECCS performance for the postulated large break LOCA in compliance with the Appendix K of 10 CFP 50.46.
This analysis was performed v:ith the NRC-approved version of the Westinghouse ECCS-LOCA evaluation model, BASH, WCAP-10266-P-A, Rev. 2 "The 1981 Version of the Westinghouse ECCS L<aluation Model Using the BASH Code," Mai h 1987.
The analytical techniques are in full compliance with 10 CFR 50.46, Appendix K.
Based on sensitivity studies in WCAP-8356, " Westinghouse ECCS Plant Sensitivity Studies," July 1974, the licensee postulated a double-ended cold leg guillotine pipe break as the most limiting case. The analysis assumed that 35% of the tubes in each SG are plugged which resulted in a reduced RCS total flowrate of 264,400 gpm.
This value bounds the expected RCS flow associated with 35% SGTP.
In addition, Westinghouse sensitivity studies set forth in WCAP-8471-P-A, "The Westinghouse ECCS Evaluation Model:
Supplementary Information," April 1975, have demonstrated that the limiting single failure is the assumntion that one low head safety injection pump fails.
This assumption, combined with Appendix K requirements, leaves flow available from two high head and one low head safety injection pumps end flow from both containment spray systems.
Using these assumptions in the BASH ECCS evaluation model, it was determined that operation at maximum power of 2748 MWt (i.e., 95% of rated thermal power) with SGTP of up to 35% in any or all SGs will comply with the 10 CFR 50.46, Appendix _K criteria. The LOCA reanalysis results show that a peak cladding temperature of 2140.8'F, a maximum local cladding oxidation level of 7.22% and a total core metal-water reaction of less than 1% will satisfy Appendix K criteria.
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.,UMMARY Based on the licensee evaluation of NSSS/ components, 80P/ components and a reanalysis of LOCA, the NRC staff concludes that the proposed TS changes are acceptable.
4.0 flNAL NO SIGNIflCANT HAZARDS CONSIDERATION DETERMINAT104 The Commission's regulations in 10 CFR 50.0? state that the Commission may make a final determination that a license amendment involves no significant hazards considerations if operation of the facility in accordance with the amendment would not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
The Commission has determined that the amendment involves no significant hazards consideration per 10 CFR 50.92, based on the licensee's analysis provided in their January 28, 1992 letter and presented below:
1.
(The proposed change] d(9s not involve a significant increase in.the probability or consequences of an accident previously evaluated.
The impact of the increased level of [SG) tube pluggiag (up to 35%
peak) with a maximum reactor power of 95% on the large break LOCA was analyzed.
The analysis demonstrated that operation with increased (SG] tube plugging will not result in more severe consequences than those of the currently applicable analyses.
The arobability of occurrence of these accidents is not increased, aecause an increased 'evel of (SG] tube plugging as an initial condition for the accident has no bearing on the probability of occurrence of these accidents.
2.
[The proposed change] does not create the possibility of a new or different-kind of accident from any accident previously evaluated.
The implementation of the increased (SG) tube plugging large break LOCA analysis into the (NA-1] design basis will not create the possibility of an accident of a different type than was previously evaluated in the [ Updated final Safety Analysis Report (UFSAR)].. No changes to plant configuration or modes of operation are implemented by the revised accident analysis.
Therefore, no new mechanisms for the initiation of accidents are created by the implementation of the analysis.
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3.
(The proposed change] does not involve a significant reduction in a margin of safety.
The (NA-1] operating characteristics, and accident analyses which support (NA-1] operation, have been fully assessed.
The results of the revised large break LOCA analysis
[ demonstrate) that the consequences of this accident are not
increased as a result of the increased [SG) tube plugging up to 35%
with a maximum reactor power of 95%.
The results of the accident analysis remain below the limits established by the currently applicable [UFSAR) analyses.
Therefore, there is no significant reduction in the margin af safety.
The NRC staff has reviewed the licensee's analysis and, based on this review, concludes that the analysis demonstrates that the applicable criteria are met.
Accordingly, the Commission has made a final determination that the amendment involves no significant hazards consideration.
5.0 STATE CONSyLTATION in accordance with the Commission's regulations, the Virginia State official was notified of the proposed issuance of the amendment.
The State official had no comment, 6.0 ENVIRONMENTAt CONSJDERATION This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The NRC 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 significant increase in individual or cumulative occupational radiation 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 (57 FR 4503).
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 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.
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7.0 CONCLUSION
The Commission has concluded, based on the considesations 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
K. Desai Date: March 3, 1992