ML20247N732
| ML20247N732 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 09/15/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247N729 | List: |
| References | |
| NUDOCS 8909260286 | |
| Download: ML20247N732 (5) | |
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g UNITED STATES
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j WASHINGTON, D. C. 20655
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I SAFETY EVALUATION BY THE OFFICE OF, NUCLEAR REACTOR REGULATION 1
RELATED TO AMENDMENT NOS.12 AND 3 TO l
FACILITY OPERATING LICENSE N05. NPF-76 A_MD NPF-80 u
HOUSTON LIGHTING & POWER COMPANY I
CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY l
l CITY OF AUSTIN. TEXAS DOCKET N05. 50-498 AND 50-499 SOUTH TEXAS PROJECT, UNITS 1 AND 2 l
1.0 INTRODUCTION
By application dated April 18, 1989 (Ref. 1), Houston Lighting & Power Company (HL&P), et. al., (the licensee) requested changes to the Final Safety Analysis Report (FSAR) for the South Texas Project (STP), Units 1 ani 2.
The proposed changes document the results of safety evaluations that accoent for the effects of the reactor coolant system (RCS) flow anomaly.
These amendments are being issued pursuant to the requirements of 10 CFR 50.59(c) because the licensee identified the changes as an unreviewed safety question. No change to the technical specifications is required by these amendments.
2.0 BACKGROUND
The RCS flow anomaly is a thermal-hydraulic instability in the recctor I
vessel which results in a slight decrease in the coolant flow in certain areas of the reactor core. InaletterofOctober3,1988(Ref.2),HL&P submitted a licensee event report regarding the existence and effects of the RCS flow anomaly in STP Unit 1.
One of the effects of the flow anomaly is a reduction in the margin to departure frons nucleate boiling 1
(DNB). To regain sufficient DNB margin to offset the effect of the anomaly, HL&P, as an interim measure until further analysis were completed, revised the STP operting procedure to maintain the RCS flow above 400,000 gpm when operating at 100% power.
I Westinghouse has since completed a detailed evaluation for STP Units 1 and 2 which concluded that the existing Technical Specification (TS) RCS I
minimum flow of 395,000 gpm is still valid, and that, for the Condition IV accidents such as a locked reactor coolant pump rotor and the RCCA ejection transients, the peak cladding temperature will be increased by less than
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10*F as a result of the RCS flow anomaly. The FSAR was revised and, after applying the three tests under 1J CFR 50.59, the licensee concluded that 50.59(a)(2)(iii))ges represented an unreviewed safety question (10 CFR the proposed chan The changes were submitted to the staff for review.
3.0 EVALUATION The RCS flow anomaly resulting from the formation of vortices in the reactor lower' plenum is a generic problem for the Westinghouse four-loop plants with flat core support plates. The RCS ficw anomaly was first observed at Union Electric's Callaway unit in late November 1986. Subsequent investigation of the root causes and impacts on reactor safety was documented in Westinghouse Topical Report WCAP-11528, "RCS Flow Anomaly Investigation Report," dated April 1988.
In addition to identifying the root cause to be the formation of multiple rotational flow cells in the lower plenum which cause local disturbances in the core inlet flow and flow anomaly in the core, the investigation report provided the following major conclusions:
(1) The core inlet flow maldistribution results in a DNB penalty, with its maximum magnitude depending on whether the reactivity feedback is considered.
(2) The flow anomaly has no impact on large and small break loss-of-coolant accident analyses.
(3) The structural integrity of the reactor internals is not expected to be affected by the flow anomaly since sufficient fatigue margin exists and the cross flow effects on the fuel rods are within the acceptable limits.
While the operating procedure was revised to maintain higher RCS flow as an interim measure to offset the DHB margin reduction due to the flow
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anomaly. HL&P proceeded to change the critical heat flux (CHF) correlation l-used in the safety analysis for STP Units 1 and 2.
In a letter cif March 30, 1989 (Ref. 3), HL&P submitted to the Nuclear Regulatory Commission (NRC) revisions to Chapters 4 and 15 of the FSAR and the affected Technical Specifications Bases by adopting the WRB-1 CHF correlation with a DNB ratio (DNBR) of 1.17 (Ref. 4) in place of the W-3 R-grid correlation with an DNBR limit of 1.3 used in the original thermal hydraulic and safety analyses. The WRB-1 correlation and its DNBR limit of 1.17 had previously been approved by the NRC for application to the Westinghouse optimized fuel assemblies and standard R-grid fuel assemblies. Therefore, it is acceptable to use WRB-1 for the STP core having fuel assemblies with comparable grid design. The change from the DNBR limit of 1.3 for the W-3 R-grid correlation to 1.17 for WRB-1 reflects an improvement in the accuracy of the CHF prediction by WRB-1 rather than a reduction in safety margin. This is because the DNBR limits for these correlations were established based on the variances of the correlations such that there is a'95% probability at a 95% confidence level that DWB will not occur when the calculated DNBR is at the DNBR limits.
Westinghouse, on behalf of HL&P, has reevaluated the FSAR Chapter 15 transients with the existing Technical Specification value of RCS flow and the WRB-1 correlation. For these analyses, a DNBR limit of 1.27 wes used to provide enough margin to offset the DNBR penalties of the RCS flow anomaly end fuel rod bow. The results showed the same conclusion as the existing analyses with the W-3 R-grid correlation. That is, for all the transients analyzed, except for the events of single RCCA withdrawal at power, locked RC pump rotor and RCCA ejection, the calculated minimum DNBRs are above the DNBR lirit. For the single rod withdrawal at power transient, which is an ANS Condition III event, the existing upper bound of 5% fuel rods experiencing DNB and fuel failure remains unchanged. For the locked rotor and rod ejection accidents, which are Condition IV events, Westinghouse analyses showed that the RCS flow anomaly will result in an increase in peak cladding temperature (PCT) of 10*F or lesr above the existing FSAR values.
In the April 18, 1989 submittal (Ref. 2), the STP FSAR was revised accordingly to reflect the effect of the 2CS flow anomaly on the PCT. For the locked rotor transient, the revised PCT is 1685*F. For the RCCA ejection transient, the revised maximum PCT is 2432*F at the end of cycle. However, the maximum average energy density at the fuel hot spot is still below the short-term and long-term core coolability criterion of 280 cal /g. For the radiological consequence calculation, the previous calculation had shown less than 10% of fuel were in DNB and, therefore, clad failure. The dose release was calculated to
.be within the limits of 10 CFR Part 100. These results are.not affected by the flow anomaly and change in CHF correlation. Therefore, the licensee concluded that there is no significant increase in radiological consequence due to the RCS flow anomaly.
In the evaluation of Significant Hazards Consideration for the RCS flow anomaly in the April 18, 1989 letter, HL&P indicated that the PCT of less than 2700*F ensures that fuel failure will not occur. The staff does not agree with the use of 2700*F PCT as the fuel failure criterion. However, the STP FSAR assumed that fission products are released from the gaps of all rods entering DNB, consistent with the fuel failure criteria is the 95/95 DNBR limit stated in both Regulatory Guide 1.77 and Standard Review Plan Section 15.4.8.
Therefore, the amount of fuel failure is correctly calculated in the STP FSAR.
4.0
SUMMARY
The staff has reviewed the effects of the RCS flow anomaly observed in STP Unit 1 and the corresponding revisions on the STP FSAR.
It has found that use of WRB-1 CHF correlation in place of W-3 R-grid correlation is acceptable, that there is sufficient margin to offset the DNBR penalty due to the RCS flow anomaly, that the radiological consequence analyzed ir the existing FSAR for the limiting accidents remains valid, and that I
the proposed changes to the affected sections of the STP FSAR are acceptable.
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5.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on September 14, 1989 (54 FR 38013).
Accordingly, based upon the environmental assessment, the Comission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.
6.0 CONCLUSION
Based upon its evaluation of the proposed changes to the South Texas Project, Units 1 and 2, FSAR, the staff has concluded that:
(1)thereis endangered by operation in the proposed manner, and (2)public will not be reasonable assurance that the health and safety of the such activities will be conducted in compliance with the Comission's regulations and the issuance of the amendments will not be inimical to the comon defense and security or to the health and safety of the public. The staff, therefore, concludes that the proposed changes are acceptable, and are hereby incor-porated into the South lexas Project, Units 1 and 2 FSAR.
Date: September 15, 1989 Principal Contributor:
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, REFERENCES 1.
Letter from S. L. Rosen (HL&P) to llSNRC Document Control Desk, " South Texas Project Electric Generating Station, Units 1 and 2. Docket Nos.
STN 50-498, STN 50-499, Proposed License Amendment Concerning the Effects of the Westinghouse Generic Reactor Coolant System Flow Anomaly,"
j ST-BL-AE-3040, April 18, 1989.
2.
Letter from G. E. Vaughn (HL&P) to USNRC Document Control Desk, " South Texas Project Electric Generating Station, Unit 1 Docket Mo. STN 50-493, Licensee Event Report 88-052 "Regarding the Effects of the Westinghouse Generic Reactor Coolant System Flow Anomaly," STP-HL-AE-2800, October 3, 1988.
3.
Letter from H. A. McBurnett (HL&P) to USNRC Document Control Desk, " South Texas Project Electric Generating Station, Units 1 and 2, Docket Nos.
STP 50-498, STN 50-499, " Change to Replace W-3 CHF Correlation with the l
WRB-1 Correlation," STP-ill-AE.-3021, March 30,1989.
4.
WCAP-8762-P-A, "New Westinghouse Correlation WPB-1 for Predicting Critical Heat Flux (CHF) in Rod Bundles with Mixing Yane Grids."
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