ML20210A450
| ML20210A450 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/28/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20210A421 | List: |
| References | |
| NUDOCS 8705050088 | |
| Download: ML20210A450 (3) | |
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.',o UNITED STATES
~g NUCLE AR REGULATORY COMMISSION
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W ASHINGTON, D. C. 20555 5
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
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SUPPORTING AMENDMENT N0f6 TO FACILITY OPERATING LICENSE NO. NPF-12 SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 DOCKET NO. 50-395
1.0 INTRODUCTION
By letter dated December 10, 1986, as supplemented March 17, 1987, and April 3,1987, South Carolina Electric and Gas Cortpany requested a change to the Technical Specifications for the Virgil C. Sumer Nuclear Station to allow an increase in the steam generator tube plugging limit from 6 p(ercent to 16 percent by changing their Heat Flux Hot Channel Fact F ) from 2.32 to 2.25. The letters contained proposed Technical i
Specification changes, revised FSAR pages, and a safety evaluation. The n
letters of March 17 and April 3,1987, provided additional information and did not change the amendment request as noticed on March 12, 1987 (52FR7696). Therefore, the amendment was not renoticed.
2.0 EVALUATION ~
1 The licensee's submittal dated April 3, 1987, provided a revised emergency i
core cooling, system (ECCS) analysis for V. C. Summer. Changes in the analysis assumptions included 16 percent steam generator tube plugging, a decrease in F from 2.32 to 2.25, and an assumption of up to 10 percent g
asymmetric tube plugging.
In addition, the licensee has assumed that at 16 percent steam generator tube plugging the thermal design flowrate will not be reduced below the level of 96,200 gpm/ loop. This is a reasonable assumption and in any event flow rate is measured when resuming full power operation after a refueling outage.
The decrease in F is intended to limit the peak clad temperature, in theeventofalahgebreaklossofcoolantaccident(LOCA),toavalue less than the 2200*F upper limit imposed by 10 CFR 50.46. The ECCS analysis utilized the BART-WREFLOOD computer model for calculating the effects of a large break LOCA and WFLASH for the analysis of the small n
break LOCA. It is noted that the ccrrected version of BART-WREFLOOD per mEE WCAP-9561, Addendum 3, Revision 1 was used.
reon.
o8 The licensee's choice of 10 percent for the maximum allowable value for tube S8 pluggine asymetry is based upon the licensee's assumption that values of 10 percent or lower will not invalidate the computer evaluation models. For example temperature channeling in the core is not a concern nor are asymmetric mg mo T
values.
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E The licensee has investigated the accident scenarios of FSAR Chapter 15. The Og majority of non-LOCA accidents are not dependent on thermal design flow or reactor coolant system volume and ere, therefore, not dependent on steam m a.a.
generator tube plugging level. The affected accidents are large and small
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break LOCA and the baron dilution event. The total primary system vclume lost if all three steam generators are plugged to 16 percent is 301.5 cubic feet. The resultant impact on the boron dilution event is a reduction in operator response time before a return to criticality. With maximum tube plugging, the operator's response time is 40.7 minutes. This is well ~in excess of the hRC's 15 minute criterion.
of 2.25 resulted in a The evaluation of the large break LOCA using an Fn peak clad tempertaure of 2023.6*F. The amount of cladding that reacts chemically with steam or water did not exceed 1 percent of the total amourt 4
of Zircaloy in the reactor, and localized claddino oxidation did not exceed 17 percent after quenching.
The licensee's small break LOCA analysis used the KFLASH computer model.
This model has been demonstrated to be insensitive to steam generator tube plugging levels up to 20 percent. The NOTRUMP evaluation model yielos.
v substantially the same results as WFLASH for tube plugging below 20 percent.
The licensee maintains that a lower peak clad temperature would result if a reanalysis using NOTRUMP were conducted, but the impact of the reducticn r
j in F used in the large break LOCA analysis significantly outweighs the n
differences between the two models. Therefore, the analysis using WFLASH which provides the licensing basis for the plant will remain valid. The small break LOCA analysis yielded a peak clad temperature of 1822*F and a J
limiting break size of 3 inches.
TherevisedFSARfagesnotedthatthereductionincoolantsystemvolume results in a decrease ir. time to change from cold leg recirculation to hot leg recirculation. The time decreases from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />.after the l
LOCA. Emergency operating procedures affected by this time reduction 1
will be revised.
The staff has reviewed the licensee's submittals dated December 10, 1956, March 17, 1987 and. April 3, 1987, and concludes that the proposed changes to from 2.32 to 2'25 and the V. C. Sumer Technical Specifications to reduce F
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toincreasethesteamgeneratortubeplugginglimitfhom6percentto16 percent is satisfactory provided the thermal design flow, independent of flow measurement uncertainty, is not reduced below 96,200 gallons per minute per loop. The licensee is already in the process of revising energency operating p,rocedures related to these changes (e.g., hot leg
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recirculationcrossover).
3.0 ENVIRONMENTAL COESIDERATION This amendment involves a change in the installation and use of a facility ~
corrponent located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant in-l crease in the amounts, and no significant change in the types, of any ef-l fluents that may be released offsite, and that there is ao significart i
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. Accordingly, this amendment meets the eligibility
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. criteria for categorical exclusion set forth in 10 CFR Section 51,22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental icpact statement or,envircn-mental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
The Commission made a proposed determination that the amendment involves no significant hazards consideration, which was published in the Federcl Register (52 FR 7696) en March 12, 1987, and consulted with the State of South Carolina. No public comnents were received, and the State of South Carolina did not have any comments.
We have concluded, based on the considerations discussed abcVe, that:
(1) there is reasonabic assurance that the health and safety of the public will not be endangered by operatien in the proposed manner, and
- 2) such activities will be conducted in compliance with the Commission's tv3;ulations and the issuance of this arendment will not be inimical to the cokr.on defense and security or to the health and safety cf the public.
Dated: April 28, 1987 Principal Contributors:
J. E. Hopkins, Project Directorate 11 ~ ~
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