ML20207F465

Forwards Amended Pages for NUREG/CR-4540,substituting Pages 43,44 & 60

ML20207F465
Person / Time
Site:	Seabrook
Issue date:	11/17/1986
From:	Khatibrahbar BROOKHAVEN NATIONAL LABORATORY
To:	Lyon W NRC
Shared Package
ML19306D588	List: ML20197D222 ML20205L245 ML20207A140 ML20207F465 ML20207Q051 ML20207Q052 ML20207S785 ML20207T164 ML20207T271 ML20207T335 ML20207T387 ML20207T396 ML20207T407 ML20207T416 NUREG/CR-4540, Forwards BNL 861117 Request to Amend Page 60 of NUREG/CR-4540, Review of Seabrook Psa:Containment Modes & Radiological Source Terms, to Clarify Wording Re Conditional Failure Probability
References
CON-FIN-A-3778, RTR-NUREG-CR-4540 NUDOCS 8701060026
Download: ML20207F465 (4)

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,9r h BROOKHAVEN NATIONAL LABORATORY

{ ASSOCIATED UNIVERSITIES. INC. Upton Long Island. New York 11973 DEC 311986 (516) 282s I Department of Nuclear Energy FTS 666' 2626 Building 130 i November 17, 1986 Mr. Warren Lyon U.S. Nuclear Regulatory Commission Mail Stop 544 (Phillips Building) Washington, D.C. 20555 RE: FIN A-3778

Dear Warren:

Enclosed please find the amended pages for the Seabrook report (NUREG/CR-

     . 4540), which should be substituted for pages 43, 44 and 60.

Please feel free to call me,1f you have any questions or comments. Sincerely,

                                                                          '1         k' Mohsen Khatib-Rahoar, Group Leader Accident Analysis Group MKR/df Enclosures

^ cc: W.T. Pratt R.A. Bari W.Y. Kato I 7'7 0.!G 6 0 R74; 4f f'

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concrete interaction for a dry cavity situation. For sequences in which early and intermediate failure is not expected to occur, and' for which containment sprays are inoperable, failure is expected to be a certainty. The conditional probability for a late overpressurization failure with a vaporization release (dry cavity) is shown to be 0.60 for large break 1.0CAs and-0.89 for small breaks and transients. This results from the relative com-petition between the late overpressure failure and the basemat penetration (TW) for accident sequences without the containment sprays for both low pres-sure (AE) and high pressure (SE, TE) scenarios. The failure time for the late overpressurization failure mode is mgch longer than previously calcul ated for other large dry containment.l. ." This is as a result of the very high failure pressure for the Seabrook con-tainment. As a consequence of this high containment failure pressure (median pressure of 211 for wet and 187 psia for dry

• sequences) it is difficult to challenge the containment integrity by any conceivable event.

Hydrogen deflagration early in the accident sequence or later after vessel failure when steam condensation occurring as a result of reactivation o sprays (due to regaining of ac power), or other natural heat sink mechanisms,{ which can produce a deinerted atmosphere is not expected to challenge the con-tainment integrity. The impact of changes in the containment failure distribution discussed in 3.2.5.4 is not significant for late failures. Basemat Penetration Failure (S4, TW) can result in the absence of containment

heat removal system (sprays) for a dry cavity. A 26-inch high curb surrounds the reactor cavity that prevents the entry of water into the cavity unless all of the water from the RWST has been injected. The conditional probability of the basemat melt through is usually less than the late overpressurization failure, this is particularly true for Seabrook where there is a natural bed rock formation directly under the basemat foundation. Therefore, the basemat penetration failure probabilities are conservatively assigned.

No Failure (S5, T!i) would result for all sequences with full spray operation. The radiological release are thus limited to the design basis leakage with essentially negligible off-site consequences. Containment Isolation Failure (S6, 3Ti) is represented by an 8-inch diameter l purge line. The accident sequences where the containment is either not isola-ted or bypassed (Event V) are assigned to these release categories. , In the SSPSA, the conditional probability for failure to isolate contain-ment (p-failure mode) is assumed to be negligibly small . This is believed to be an optimistic assumption on the part of the applicant, because even for subatmospheric containment the p-failure mode is expected to have a condi-

                 *For dry sequences, only primary system water inventory is available in the
!                containment.                   In this case, the containment atmosphere becomes superheated and, at failure, the temperature can exceed 700*F.

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                ,                                                                                                                                                                  l tional failure probability ranging from 4x10 " to about 2x10 3; therefore, one expects the conditional p-failure mode probability r a large dry containment to be somewhat higher, and perhaps approaching =10-An interfacing systems LOCA (V sequence) results from valve oisc rupture or disc failing open for series check valves that normally separate the high pressure system. This event results in a LOCA in which the reactor coolant bypasses the containment and results in a loss-of-coolant outside the contain-ment.                    Furthermore, the concurrent assuned loss of RHR and coolant make-up capability leads to severe core damage. In the SSPSA, only three possible interfacing systems LOCA sequences have been found and discussed. These are:

1. Disc rupture of the check valve in the cold-leg injection lines of the RHR.

2. Disc rupture of the two series motor-operated valves in the normal RHR hot-leg suction.

3. Disc rupture of the motor-operated valve equipped with a stem mounted

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limit switch and " disc failing open while indicated closed" in the other motor-operated valve in the normal RHR hot-leg suction. For the V-sequence, the core melts early with a low RCS pressure and a dry reactor cavity at vessel melt-through. The containment sump remains dry and recirculation is not possible. The core and containment phenomenology used to arrive at the split frac-tions for the containment event tree and thug "the C-matrix are in general i agreement with the other previous studies , ,- for PWRs with large dry containments. Furthermore, the' claimed unusually high strength of the Sea-brook containment reduces the impact of sensitivity caused by uncertainties in the severe accident progression. However, should the claimed strength of the containment be reduced to levels comparable to some of the other large dry containments, the impact of uncertainties may becpe significantly more pro-nounced, as discussed on our review of the MPSS-3. 3.6 Release Category Frequencies Based on the containment class frequencies in Table 3.6 and the contain-ment failure matrix of Table 3.9, the release frequencies were computed and are summarized in Table 3.10. Table 3.10 indicates that only eight of the release categories dominate the total release frequency. Tables 3.11 and 3.12 set forth the contribution to core melt frequency from the various containment response classes and release categories, respec-tively. It is seen that containment classes 2, 4, and 5 dominate the co're melt frequency while the release categories S5 (containment intact), 3T and S3V dominate the source term frequency. c- ,,w=-.-,.,aw,-,--- -*- e -w--  %--,y . + - - - - - - - - - -+--+--e-. - - - - - - ---,-r---- ---------w-y,w - m - - r-- -=g-w we

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5.

SUMMARY

AND CONCLUSIONS , i The purpose of this report is tc describ'e the technical review of the Sea-brook Station Probabilistic Safety Assessment and to present an assessment of containment performance, and radiological source term estimates for severe core melt accidents. The containment response to severe accidents is judged to be an important factor in mitigating the severe accident risk.- There is negligible probability of prompt containment failure. Failure during the first few hours of overpressure failure is

                      'after corecompared very long   melt is also        unlikely to the       and Reactor   the timing Safety  Study (WASH-1400).Most core melt accidents would be effectively mitigated by containment spray operation. A comparison of SSPSA and RSS containment failure frequencies is given in Table 5.1.
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Our assessment of the containment failure characteristics indicate that, there is indeed a tendency to fail containment through a realistic benign mode compared with the traditional gross failures. The . point-estimate release fractions used in the SSPSA are comparable in magnitude to those used in the RSS. In those cases where comparisons can be made to the more mechanistic source term study carried out by the Accident Source Term Program Office ( ASTP0) of the NRC and reported in BMI-2104 it was found that.the SSPSA releases were either higher than or for the most part similar to the recent release fractions. It was also found that the energy of release was somewhat higher in the SSPSA than for other existing studies. l r l i

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