ML20086C917
| ML20086C917 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/06/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20086C908 | List: |
| References | |
| NUDOCS 9111250160 | |
| Download: ML20086C917 (5) | |
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![,c urg'g UNITED STATES NUCLEAR REGULATORY COMMISSION 3
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_E WASHINoToN, D. C. 20565 SAFETY EVALUATION BY THE OFFICE Or NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 51 TO FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION
_DCCKET NO. 50-482
- 1. 0 INTRODUCTION By letter dated March 5, 1991 (Ref. 1), Wolf Creek Nuclear Operating Corporation (WCNOC) (the licensee), requested an amendment to facility Operating Licente NPF-42 for the Wolf Creek Generating Station (WCGS) Unit 1.
The proposed amendment presented changes to the Technical Specifications due to proposed modifications to reduce the reactor coolant system (RCS) thermal design flow, tu replace the W-3 critical heat flux (CHF) correlation with the WRB-1 CHF correlation, and to increase the low pressurizer pressure reactor trip setpoint
- limit, These measures have been taken in enticipation of the need to provide compensatory thermal margin to accommodatc any future actual RCS flow degradation due to steam generator tube plugging.
2.0 DISCUSSION The proposed change in RCS thermal design flow (TDF) is from the current Technical Specification value of 95,700 gpm/ loop to a new value of 93,750 gpm/ loop, a reduction of approximately 2 percent.
The proposed reduced RCS flow requirement was chosen to reasonably bound potential future need to account for steam generator tube plugging or sleeving of up to 4 percent of the tubes in each steam generator without requiring extensive reanalysis.
The
-critical heat flux correlation was changed f rom h-3 to WRB-1 to obtain mere margin to offt.et the proposed decrease in RCS TOF.
Also, the low pressurizer
.nressure reactor trip setpoint safety analysis limit (SAL) was increased from 1860 psia to 1900 psia to ensure protoction against vessel exit boiling with reduced RCS flow.
2.1 Core Thermal Limits In light of the potential decrease in RCS flow, the licensee re-calculated core thermal limits with the WRB-1 critical heat flux correlation.
The WRB-1 critical heat flux (CHF) correlation was used in place of the W-3 correlation which was used for analysis documented in the current Updated Safety Analysis Report (USAR).
The WRB-1 correlation is less limiting and offsets the proposed decrease in required RCS flow.
This CHF correlation has been previously reviewed and approved by the NRC (Ref. 2) and is therefore acceptable.
The h
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9 licensee found that the WRB-1 CHF correlation provides wufficient margin to offset the effects of the proposed RCS thermal design flow.
The margin was partially utilized to maintain the existing DNB limits, and the balance is identified as generic DNB margin available for use in future applications.
Changing from the W-3 based correlation to the WRD-1 CHF correlation while maintaining the existing DNB core thermal and axial offset limits results in a redefinition of the safety analysis limit DNBR f rora 1.30 to 1.32.
This results in an N rease in the generic DNBR margin (i.e., the margin between the WRB-1 CHF relation limit DNBR (1.17) and the safety analysis limit DNBR (1.32)).
<e have found this application of the WRB-1 CHF correlation for thermal hydraulic analysis to be acceptable.
The low pressurizer pressure reactor trip setpoint SAL was increased from 1860 psia to 1900 psia in this evaluation, This change was made to ensure that vessel exit boiling limits (VLBL) would not be exceeded during depressurization transients with the reduced RCS flow rate and is therefore acceptable.
This reduction in low pressurizer pressure reactor trip setpoint SAL is reflected in this troposed Technical Specification change.
- 2. 2 Evaluation of Non-LOCA Accidents Previously Analyzed The licensee stated that all non-LOCA tranaients and accidents included in the USAR were reevaluated for sensitivity to and the potential effects of reduced RCS thermal design flow.
Critical statepoints in the original transient analyses, i.e., the transient thermal-hydraulic conditions at the time of minimum DNBR, were identified and reevaluated with a 2 percent flow reduction and reduced inlet temperatures using WRB-1 CHF correlation.
In all cases, the evaluations gave acceptabla results when compared with the revised SAL DNBR of 3.32.
For the transients where the critical statepoint conditions fell outside the range of applicability of the WRB-1 CHF correlation, the statepoints were reevaluated'using the W 3 CHF correlation assuming a 2 percent reduction in the flow rate.
Sufficient accident specific margin was found to be available for these transients to accommodate botn the penalty from reduced flow and
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the increased generic DNB margin included in this evaluation.
2.3 LOCA and LOCA Related Analysis Loss-of-Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks with the Reactor Coolant Boundary - (USAR 15.6.5)
(a) Large Break LOCA The licensee stated that the large break LOCA and long-term core cooling calculations were previously analyzed in the USAR for a reduced RCS thermal design flow (TDF) of 93,200 gpm/ loop and steam generator plugging levels of up to 10 percent.
Therefore the request for a reduction in TDF from 95,700 gpm/ loop to a reduced value of 93,750 gpm/ loop is bounded by
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the 93,200 spm/l]op analysis.
We therefore find this acct.ptable.
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(b) Small Break LOCA The licensee stated that the originally licensed small break LOCA analysis for WCGS was performed using the Westinghouse WFLASH Evaluation Model and assumed a thermal design flow of 95,700 gpm/ loop with no steam generator tube plugging.
Subsequent analysis were performed per Three Mile Island Action Plan Item II.K.3.31 with tt NOTRUMP small break LOCA evaluation mVel to demonstrate that the WFI evaluation model was bounding.
This a.alysis was reviewed and app-
, % NRC for application to WCGS (Ref. 5).
Subsequent generie v
> 1 been performed using the NOTRUMP code to assess the ei
% mal design flow of 93,200 gpm/ loop and 10 percent steam t.
lugging on small break LOCAs (Refs. 6 and 7).
These analyses v
t-(1) steam generator plugging levels _ up to 10 percent
'de effective heat sink to the primary side w'th reduced I.
' lugging leveis of up to 15-20 percent would have no effe w cry end therefore no effect on peak cladding temperatures.
Based on the above, it is concluded that the 2 percent reduction of RCS thermal design flow to 93,750 gpm/ loop will have no significant effect on the PCT for small break LOCA which is already well below that for the large break LOCA, i.e., 1,917.6*F versus 2,163.5*F, 3.0 EVALUATION As a result of the modifications associated with the reduction in RCS thermal design flow, increase in low pressurizer pressure reactor trip setpoint limit and change in the critical heat flux correlation (CHF) from the W-3 to the WRB-1 CHF, changes to the plant's Technical Specifications were proposed.
The following Technical Specifications were examined.
Figure 2.1-2, page 2-2
" Reactor Core Safety Limit - Four Loops in Operation" The parameter of pressure in the RCS Tavg versus Fraction of Rated Power curve was redrawn to increase the 1860 psia value to 1900 psia. This was to reflect the modifications in TDF and the CHF correlation and a shift in the steam generator safety valve actuation line.
This is acceptable as discussed above in Section 2.0.
Table 2.2-1, page 2-4, " Reactor Trip System Instrumentation Trip Setpoints"
-Functional Unit 9
" Pressurizer Pressure Low" was modified.
The trip setpoint was changed from equal or greater than 1875 psig to equal or greater than 1915 psig and t..e allowable value was changed from equal or greater than 1866 psig to equal or greater than 1906 psig. The footnote for loop design flow was changed from 95,700 gpm to 93,750 pgm.
These changes were found to be a::ceptable as discussed above in Section 2.0.
.o Figure 3.2-3, page 3/4 2-9, "RCS Total Flow Rate Versus R - Four Loops in Operation" The total flow rate value of 392,400 gpm was changed to 384,440 gpm.
This is acceptable as it is four times the new TDF value of 93,750 gpm/ loop and includes the flow measurement uncertainty value of 2.5 percent.
Bases 2.1.1 Reactor Core, page B 2-1 l
This page deletes reference to the W-3 CHF correlation.
An insert explains the use of the WRB-1 CHF correlation in place of the W-3 CHF correlation.
This editorial change is acceptable as discussed above in Section 3.0.
Page B 3/4 2-4, Heat Flux Hot Channel Factor, and RCS Flow Rate and Nuclear Hot Channel Factors A portion of a sentence was changed from "below the design DNBR value" to "below the safety analysis DNBR value." Also reference to the generic margin of 9.1 percent DNBR was eliminated together with a listing of the margins and replaced by a generic margin of 11.4 percent ONBR.
A sentence was added which stated "This is the margin between the correlation DNBR limit (1.17) and the safety analysis limit DNBR (1.32)."
These editorial changes are acceptable as discussed in Section 2.0.
Bases 3/4.2.0, DNB Parameters, page 3/4 2-6 A sentence with reference to a " minimum DNBR of 1.30" was changed to "0NBR above the safety analysis limit DNBR (1.32)." This editorial change is acceptable as discussed in Section 2.0.
Bases 3/4.4.1, Reactor Coolant Loops and Coolant Circulation, page B 3/4 4-1 A sentence with "DNBR above 1.30" was changed to "DNBR above the safety analysis limit DNBR (1.32)."
This editorial change is acceptable as discussed in
_Section 2.0.
Thd impact of changing:
(1) the RCS thermal design flow, (2) the low pressurizer pressure reactor trip setpoint limit, and (3) the critical heat flux correlaticn from W-3 to WRB-1 for the Wolf Creek plant on the UFSAR Chapter 15 accidents has been evaluated by the licensee.
The staff has found that the former
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conclusions in the UFSAR remain valid and the Technical Specification changes have been determined to be acceptable as described in Sections 2.0 and 3.0,.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Kansas State official was notified of the proposed issuance of the amendment.
The State official had no comments.
6 5.0= ENVIRONMENTAL CONSIDERATION The 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,
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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 the amendment involves no significant. hazards _consideratica, and there has been no public comment.
on such finding (56 FR-13673).
Accordingly, the 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_the amendment.
- 6. 0 CONCLUSION The Commission has 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, (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.
REFERENCES 1.
- Letter,-Forrest T. Rhodes, Wolf Creek Nuclear Operating Corporation, to NRC, March 5,-10 1.
2.
Letter, John Stolz, NRC, to C. Eicheldinger, Westinghouse Electric Corporation, April-19,.1978.
3.
SLNRC-85-01, " Steam Generator Single-Tube Rupture Analysis for SNUPPS Plants Callaway'and Wolf Creek," dated January 8, 1986.
4.
WCAP-10961aP, "Steamline Break Mass / Energy Releases for Equipment Environmental Qualification Outside-Containment," October 1985.
5.
Letter from Paul W. O'Connor, NRC, to Glenn L. Koester, KG&E, November 17, 1986.
6.
Ciani, S., et a1., " Simulation of Small Break Type Behavior of PUN and-SPES'using the NOTRUMP Code," Proceedings of_the Specialist Meeting on Small Break LOCA Analyses in LWR's, Pisa, Italy, June 1985.
7.
Lee, N.
" Limiting Countercurrent Flow Phenomena in Small Break LOCA Transients," Proceedings of the Specialist Meeting on Small Break LOCA Analyses in LWR's, Pisa, Italy, June 1985.
Principal Contributor:
Harry Balukjian, SRXB Date:
November 6,-1991 l
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