ML20235S938
| ML20235S938 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 09/25/1987 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC |
| References | |
| FRN-52FR7950, RTR-NUREG-1150, RTR-NUREG-4550 52FR7950-00025, 52FR7950-25, 87-230, NUDOCS 8710090193 | |
| Download: ML20235S938 (6) | |
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87 SEP 30 A9:31 September 25, 1987
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'? 9'so Chief, Rules and Procedures Branch Serial No.87-230 Division of Rules and Records N0/JD11:vlh Office of Administration U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Gentlemen:
VIRGINIA 17LECTRIC AND POWER COMPANY COMMENTS ON DRAFT NUREG-1150 REACTOR RISK REFERENCE DOCUMENT We appreciate the opportunity to comment on draft NUREG-1150, " Reactor Risk Reference Document" which was published for public comment on March 2,
1987.
The document is of particular interest to us because Surry is one of the five facilities discussed in the document in detai3. Our comments, both general and specific, are enclosed.
Very truly yours, b b b 'N W. L. Stewart l
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ENCLOSURE Comments on Draft NUREG-1150,
" Reactor Risk Reference Document" Virginia Elcetric and Power Company W...
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The draft report " Reactor Risk Reference Document, NUREG-1150, provides the results of updated risk analyses for five different U.S. light-water reactors (Surry, Zion, Sequoyah, peach Bottom, and Grand Gulf). Virginia Electric and power Company has performed a limited review of this draft report.
A more comprehensive review was performed of NUREG-4550/ Volume 3 entitled " Analysis of Core Damage Frequency From Internal Events:
Surry Unit 1".
This report contains the results of the updated accident sequence analyses for Surry Unit I including the overall core damage frequency estimate and accompanying plant J
damage state frequencies forming the basis for the risk analyses presented in l
NUREC-1150. General comments are provided on the contents of draft NUREG-1150 and specific comments are provided on NUREG-4550.
General Comments on NUREG-1150 1)
The report demonstrates that there remains a large uncertainty in containment performance and sou..e terms for the plant damage states identified. As structured, the document has limited use for the purposes stated in the Executive Summary.
A, central estimate of risk is needed, not merely a statement of the range of possible uncertainties.
2)
The method of presentation of results has led to public misconception of the probability of containment failure and estimates of risk following a severe accident. For example, the early containment failure probability presented for Surry ranges from.01 to
.9.
The fact that the most likely probability of early containment failure is on the order of.05, is s
obscured by the method of presentation.
3)
The effect of expert judgements and the process by which they were incorporated is not clear in the document.
4)
Appropriate credit was not always given for equipment not governed by the Technical Specifications.
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.i 5)
PWR reactor coolant pump seals continue to be evaluated as having a significant potential for causing a loss of coolant accident if cooling is lost. This issue should have been resolved by the French / Westinghouse I
tests.
6)
The document is not complete without some treatment of external events.
We understand that the planned revision is to incorporate an analysis of these events.
l 7)
The draft NUREG-1150 was prepared with only limited consultation with the designers and operators of the plants analyzed.
The results are more
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1 pessimistic than the results obtained under the Industry Degraded Core Rulemaking (IDCOR) program.
In general, it appears that the models are overconservative at d the risk estimates overstated.
Comments on NUREG-4550 Virginia Electric and Power Company had performed a review of the NUREG-4550 for its contents and the methodology employed for the assessment of core I
damage earlier, and our comments were forwarded to the authors in October, 1986. The authors of NUREG 4550/Vol. 3 agreed with some of our comments as was noted in the appendix of that report. However, due to time constraints of the publication of the report, the analysis was not revised.
Therefore, we believe that based on our comments the analysis should be revised, especially the areas which have a substantial effect on the results. The following items summarize the more significant of these comments.
1)
The updated PRA does not reflect the fact that new service water valves have been installed on Surry Unit I to avoid previous problems nor give enough credit for the full range of possible measures that could be taken in the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> available to obtain flow through at least one of the four l
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valves. While we understand that the frequencies of effected sequences have been reduced by recognition of the fact that core damage resulting from containment failure is unlikely, the potential risk significance of
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these sequences makes it important that the above factors be adequately l
reflected in the base case "best estimate" analysis.
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2)
The seal LOCA model doe, not adequately reflect current knowledge about seal behavior.
The model should include a probabilistic treatment of both time to failure and flow rates similar to that presented to the NRC by the Westinghouse Owners Group. The high probability of the maximum possible flow, given a seal LOCA, is unrealistic and should not be the
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base case.
3)
The PRA utilizes a generic representation of loss of offsite power rather than a plant / site specific analysis. Application of the methodology of NUREG-1032 to plant specific conditions yields a loss of offsite power versus duration curve considerably lower than that used in the PRA.
Use of the generic value is inappropriate for the Surry base case.
4)
The FRA assumes that one diesel generator is necessary for t.he safe shut down of each unit at Surry. No credit is given for the ability to bring and maintain both Unit I and 2 to safe shutdown conditions utilizing a l
single diesel generator. Procedures and equipment for cross connecting the Unit I and 2 electrical buses and heat removal systems are maintained in order to achieve this capability. This has a significant impact on the results and should be properly considered in the PRA.
In addition, the PRA should include the nost recent Surry diesel generator test results in arriving at the values used in the analysis.
The failure to start probability based on data for 1983 - 1985 is about one half of that used in the PRA.
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5)
The analysis of the interfacing LOCA includes use of check valve failure rates which are conservative (including leaks which are within pressure relief capacity) or are inappropriate (using a failure to close mode when the valve is tested following refueling or cold shutdown).
While interfacing LOCA is not a dominant core damage contributor, the potential risk significance requires as realistic an analysis as possible in the "best estimate" base case.
6)
While a number of these issues are covered by sensitivity studies, the base case should present the worst case realistic assessment
("best estimate").
Incorporation of the changes suggested in these comments will substantially change the "best estimate" results of the study for Surry.
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