ML20004D386

Forwards Comments in Response to NRC Re Conclusions of Reactor Safety Study Methodology Program Evaluation of Facility.Conclusions Reached in Study Are Not Appropriate for Identifying Corrective Actions

ML20004D386
Person / Time
Site:	McGuire, Mcguire
Issue date:	06/05/1981
From:	Parker W DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8106090333
Download: ML20004D386 (4)

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Mr. Harold R.-Denton, Director 3[ {L' l.b q

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M, Attention:

Ms. E. G. Adensam, Cnief g/f/

Licensing Branch No. 4

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Re: McGuire Nuclear Station Docket Nos. 50-369, 50-370

Dear Mr. Denton:

The following comments are provided in response to Mr. Robert L. Tedesco's letter of May 26, 1981 concerning the conclusions of the Reactor Safety Study Methodology Applications Program evaluation of Sequoyah Unit 1.

Duke Power Company supports the use of probabilistic analysis as an aid in evalcating the level of safety associated with nuclear power plants, and parti-cularly in the application of available resources for improving s,afety where they can be of the most benefit. However, reasonable application of these techniques requires a meaningful standard for comparison. Considering the level of details in the Sequcyah RSSMAP Study, we do not believe that studies such as the Sequoyah RSSMAP provide a good de facto standard. Therefore, we do not feel that the comparison of McGuire to the conclusions reached in RSSMAP is particularly appropriate in identifying necessary corrective actions.

Nevertheless, we have reviewed each of the conclusions contained in pages 4-8 and 4-9 of the Sequoyah RSSMAP report with respect to the McGuire plant, and our assessment is provided in the attachment to this letter.

Very truly yours, William O. Parker, Jr.

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Ms. M. J. Graham Mr. James P. O'Reilly, Director Resident Inspector U.S. Nuclear Regulatory Commission McGuire Nuclear Station Region II o

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McGUIRE NUCLEAR STATION ASSESSMENT OF SEQUOYAH RSSMAP CONCLUSIONS WITH RESPECT TO McGUIRE NUCLEAR STATION RSSMAP CONCLUSION An important accident sequence occurring for the Sequoyah plant results from the potential for blockage or closure of the drains between the upper and lower compartments. This causes a common-mode failure of the ECRS and CSRS when the sump runs dry (sequences S HF and S,HF).

The probability of these 7

sequences could be reduced by improved checking procedures and improved fault detection capabilities.

COMMENTS The McGuire containment differs from that of Sequoyah in this respect. At

'McGuire there are six drain lines, rather than two.

Any two of the drain lines provide sufficient flow path between the compartments for successful ECR and CSR. Therefore, there is a significantly lower likelihood that failure of recirculation for McGuire will result from blockage of the drain lines.

In addition, sufficient administrative controls exist to provide assurance against the common-mode failure due to leaving the drain line valves closed following refueling. These controls include the following:

Each drain line contains one valve which is normally locked open.

Double verification that each of the valves is open is performed following refueling operations.

Confirmation of the above double verification is required by the precritical valve alignment verification procedure.

These lines are checked for indications of blockage during per-formance of the periodic surveillance procedure.

RSSMAP CONCLUSION Failure of the ECRS alone caused by component failures other than the drains also results in some important accident sequences.

C0beENTS This conclusion is not surprising for most plants in light of the importance of ECRS function during LOCA accidents. A detailed analysis would be required to quantify the ECRS failure probability, to identify important contributors to its failure probability, and to investigate the system's importance in dominant accident sequences.

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~RSSMAP CONCLUSION Sequence V, in which check valve failues cause the high-pressure primary coolant to fail _the low-pressure piping outside containment, remains an important sequence for Sequoyah. This sequence could be improved by a more-strategic testing procedure of the check valves over the limited testing i

capability which now exists.

l COMMENTS For McGuire periodic leak testing to verify the integrity of both the inboard-and outboard check valves in the LPIS is performed, as required by the Techni-cal Specifications. Thus, the likelihood of occurrence of the Event V se-quence at McGuire is significantly lower than that reported in the Sequoyah RSSMAP report, which is apparently based on monitoring the leakage of the outboard check valve only.

RSSMAP CONCLUSION Unlike larger containmente, core melting caused by failure of'ECIS or ECRS j

fail the lower pressure, smaller ice condenser containment by' overpressure even though the containment cooling system continues to operate properly.

The analysis of accident processes by Battelle Columbus Laboratories revealed that the smaller containment pressure and volume design would not withstand

,t the pressure exerted by the noncondensible gases generated in the core melt-down accidents.

(This result was similar to the RSS findings for the RSS BWR design.)

COMMENTS As documented in previous communications with NRC (refer to my letters of November 17, 1980'and February 17, 1981 to Harold R. Denton), Duke Power Com-pany has expended a significant effort in examining the causes and effects of generation of noncondensible gases during reactor accidents. Those letters also provide ' details of measures which have been taken for prevention and

- mitigation of such events.

In addition, the McGuire containment has been evaluated to have a substantially higher functional capability (67.5 psig for McGuire versus 30 psig reported for Sequoyah).

RSSMAP CONCLUSION 4

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- RSS,.~does not appear to be' as significant to risk for Sequoyah due to the lower unavailability of on-site 'ac power.

COMMENTS A detailed review of the McGuire system would be required to comment on the applicability of this conclusion to McGuire;. however, since this conclusion

. pertains to'a positive feature of the Sequoyah plant, a detailed evaluation of McGuiro is not deemed necessary.

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t RSSMAP CONCLUSION Failure of the containment cooling system causing core meltdown following a small LOCA-(the S C sequence in the RSS) does not appear to lead to core 2

- meltdown at Sequoyah due to the difference in sump water temperature at the time of containment failure.

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COMENTS.

The containment heat removal capability and presence of the ice condenser i

are expected to result in a similar conclusion for McGuire.

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