ML16341G769

Forwards RAI Re IPE Results Submitted 920414

ML16341G769
Person / Time
Site:	Diablo Canyon
Issue date:	11/19/1992
From:	Peterson S Office of Nuclear Reactor Regulation
To:	Rueger G PACIFIC GAS & ELECTRIC CO.
References
TAC-M74403, TAC-M74404, NUDOCS 9211240240
Download: ML16341G769 (16)

Text

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November 19, 1992 Docket Nos.

50-275 and 50-323 Hr. Gregory H. Rueger Senior Vice President and General Manager Nuclear Power Generation, B14A Pacific Gas and Electric Company 77 Beale Street, Room 1451 P.O.

Box 770000 San Francisco, California 94177

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON DIABLO CANYON NUCLEAR GENERATING STATION INDIVIDUAL PLANT EXAMINATION (IPE)

SUBMITTAL (TAC NOS.

H74403 AND H74404)

By letter dated April 14, 1992, you submitted the Diablo Canyon IPE results for NRC review.

Based on our review of your submittal, we have determined that we need additional information to continue with our review.

The enclosed list of questions identifies the information we need.

Please review these, questions so that we can schedule a conference call in about 30 days to discuss your responses.

Following the call, a determination will be made as to whether we need a meeting to discuss any items further.

We will require a

written response within 60 days.

Please contact us should you have any questions regarding this request.

Sincerely, Ortginat sign'ed by Sheri

Peterson, Project Manager Project Directorate V

Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Enclosure:

Request for Additional Information cc w/enclosure:

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 November 19, 1992 Docket Nos.

50-275 and 50-323 Hr. Gregory M. Rueger Senior Vice President and General Manager Nuclear Power Generation, B14A Pacific Gas and Electric Company 77 Beale Street, Room 1451 P.O.

Box 770000 San Franci sco, Cal iforni a 94177

SUBJECT:

RE(VEST FOR ADDITIONAL INFORMATION ON DIABLO CANYON NUCLEAR GENERATING STATION INDIVIDUAL PLANT EXAMINATION (IPE)

SUBMITTAL (TAC NOS.

M74403 AND H74404)

By letter dated April 14,

1992, you submitted the Diablo Canyon IPE results for NRC review.

Based on our review of your submittal, we have determined that we need additional information to continue with our review.

The enclosed list of questions identifies the information we need.

Please review these questions so that we can schedule a conference call in about 30 days to discuss your responses.

Following the call, a determination will be made as to whether we need a meeting to discuss any items further.

We will require a

written response within 60 days.

Please contact us should you have any questions regarding this request.

Sincerely,

Enclosure:

Request for Additional Information cc w/enclosure:

See next page Sheri

Peterson, Project Manager Project Directorate V

Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Hr. Gregory H. Rueger Pacific Gas and Electric Company Diablo Canyon CC:

NRC Resident Inspector Diablo Canyon Nuclear Power Plant c/o U.S. Nuclear Regulatory Commission P. 0.

Box 369 Avila Beach, California 93424 Dr. Richard Ferguson, Energy Chair Sierra Club California 6715 Rocky Canyon Creston, California 93432 Hs.

Sandra A. Silver Mothers for Peace 660 Granite Creek Road Santa Cruz, California 95065 Hs. Jacquelyn C. Wheeler 3303 Barranca Court San Luis Obispo, California 93401 Managing Editor The County Telegram Tribune 1321 Johnson Avenue P. 0.

Box 112 San Luis Obispo, California 93406 Chairman San Luis Obispo County Board of Supervisors Room 370 County Government Center San Luis Obispo, California 93408 Christopher J.

Warner, Esq.

Pacific Gas 5 Electric Company Post Office Box 7442 San Francisco, California 94120 Mr. Hank Kocol Radiologic Health Branch State Department of Health Services Post Office Box 942732 Sacramento, California 94234 Regional Administrator, Region V

U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 Hr. Peter H. Kaufman Deputy Attorney General State of California 110 West A Street, Suite 700 San Diego, California 92101 Ms. Nancy Culver 192 Luneta Street San Luis Obispo, California 93401 Michael H. Strumwasser, Esq.

Special Assistant Attorney General State of California Department of Justice 3580 Wilshire Boulevard, Room 800 Los Angeles, California 90010 Diablo Canyon Independent Safety Committee ATTN:

Robert T. Wellington, Esq.

Legal Counsel 857 Cass Street, Suite D

Monterey, California 93940

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ENCLOSURE RE VEST FOR ADDITIONAL INFORMATION UESTIONS ON DIABLO CANYON INDIVIDUALPLANT EXAMINATION IPE SUBMITTAL In Section 4.1.1.5, page 4.1-3, of the IPE submittal, the possibility is discussed of high-efficiency particulate air (HEPA) filter plugging due to suspended aerosols in the containment atmosphere.

Reference is made to a planned design change to remove HEPA filters.

Without HEPA filters to remove the aerosols, have you considered the effects of aerosols on the availability of safety equipment?

The failure mode descriptions, as given in Table 4.4-1, agree with the text on page 4.4-2.

However, in Table 4.4-2, 14 failure modes (A

through N) are listed for which there is no supporting discussion.

What are the relationships between these two sets of failure modes?

Seven areas of uncertainty associated with direct containment heating are listed in Subsection 4.5.2, page 4.5-6, of the submittal.

Please explain how these uncertainties were considered (if at all) in containment event tree (CET) split fractions.

Subsection 4.6. 1.2. 1, page 4.6-13, of the IPE submittal stated "When the blockage model is used, it has the effect of reducing gas circulation through the core as the core degrades."

Please describe qualitatively the effect of reducing gas circulation on core degradation.

If induced RCS hot leg or surge line failure (Top Event IP of CET)

occurs, what is the assumed likelihood of successful in-vessel cooling of debris by coolant injection?

Subsection 4.6.2, paragraph 3,

page 4.6-49, of the IPE submittal responds to CPI Program recommendations.

Describe the analyses that were performed in this regard.

The IPE takes credit for many long-term pressurization sequences.

The probability of this recovery action is noted as 90 percent.

What is the basis for this probability value?

Does this probability apply to both CFCUs and sprays?

A discrepancy appears in the IPE submittal with respect to the impact of CFCU recovery in preventing late containment failure.

In one place in the submittal, the contribution of CFCUs in arresting late containment failure was assessed as marginal, as shown in Table 4.8-2.

Earlier, in Subsection 4.8.2.5, it could be inferred that, by taking credit for this recovery, a 20-percent reduction in late containment' failures could occur.

Please explain this apparent discrepancy.

In the discussion of the containment performance vulnerability, no reference is made to con'tainment isolation failure.

Please provide a discussion of the containment performance vulnerability as it relates to containment isolation failures.

What is your best estimate for containment isolation failure?

Have any actions been taken to reduce the likelihood of containment isolation failure?

As noted in Subsection 4.8.4.3, page 4.8-10, of the IPE submittal, a

relatively simple design modification would allow cavity flooding.

What design modifications have you considered in light of the sensitivity analysis results, which show a 50-percent reduction in large, early containment failures?

Are there any threats to containment integrity caused by recovering the CFCUs, thereby de-inerting the containment?

For induced large LOCAs, described on page 4.6-20, you postulate a

large vessel failure due to pressurized thermal shock at the time of RWST water injection.

For this scenario, you claim that the availability of recirculation water and CFCUs ".

. would prevent this fuel damage accident from evolving into a severe accident Please discuss this sequence and the analysis that supports it.

Please concisely describe the quantification process of the CETs.

In view of the fact that the IPE cites as an objective the identification of differences between the Level 2 results at Zion and

DCPP, please discuss any insight gained by comparing these results.

What back-end improvements were examined in IPE?

Please discuss the disposition of the back-end improyements examined.

Please identify any important operator actions considered in the back-end analysis, and how these operator actions were modeled in the IPE analyses.

Section 4.6. 1.3.1 (SXYAI) indicates that the analysis assumes that the operators will initiate RHR cooling after vessel failure.

Discuss the rationale for the assumed successful (failure rate

0) initiation of this action.

Please describe any other operator actions taken credit for in this manner in the containment analysis portion of the IPE that have not been explicitly modeled.

Discuss the rationale for this treatment.

Discuss the sensitivity of the results to these assumptions.

Are these actions covered under current procedures?

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The submittal identifies a number of walk-throughs for the Level 2

analysis and flooding. It also identifies that a review of design change packages were done in addition to others, to confirm that the analysis represents the "Current Plant Information".

For the plant changes incorporated since the DCPRA-1988, please discuss the method used to verify that the changes identified in the documentation were physically incorporated as specified and as modeled in the analysis.

The submittal indicates that due to the "RISKHAN" software enhancements, the 178 support system states identified in DCPRA-1988 were no longer necessary.

Since this change in method of addressing the impact of support systems on frontline systems is significant, please describe the impact (if any) of the new method on the results.

Did the core damage frequency increase or decrease7 Have you performed any analysis to assess the impact on the results due to changes in the codes'lease discuss.

4 The submittal indicates that a frequency of a flood scenario is compared to other initiating events that have a similar impact, and if that flood frequency is insignificant compared to the initiating event frequency, then the flood scenario is screened out.

The criteria for this comparison is not discussed.

Please discuss the criteria and its application in making this judgement.

Since the assessment of core damage due to flooding is subject to much uncertainty in the parameters used in the assessment, please discuss your consideration of the sensitivity of the credit taken for operator

action, address the flood scenarios that dropped below the core damage frequency screening criteria because the frequency has been reduced by more than an order of magnitude due to credit taken for operator

action, and discuss how the application of the criteria you'e chosen assures the capture of potential vulnerabilities due to flooding.

Is the core damage frequency (CDF) from all the screened out flood scenarios equal to or greater than the CDF from the non-screened out scenarios7 23.

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25.

The loss of instrument air system (LOIA) is not addressed as a

separate initiating event.

Is the frequency of LOIA as an initiating event subsumed in another event frequency (e.g.

TLHFW)7 Please discuss your treatment of this event.

The list of initiating events does not include loss of power to a single AC Bus.

Describe your investigation of these types of initiating events and discuss the rationale (including impact and frequency) for not including this as an initiating event for your IPE.

The discussion under top event SE indicates that if top event PR fails (i.e.,

PORV opened and failed to reseat),

the top event SE (RCP seal cooling) is not asked because a

LOCA is then already known to have

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occurred.

If seal cooling is lost subsequent to PR, is the resultant size of the LOCA from both leaks significant to success criteria,

timing, CDF, and release category?

There is no description of the Seal LOCA Model provided in Sections 3.1.3 or Section 3.1.4.

Provide a discussion of the Seal LOCA Model used for the IPE submittal including the various leak rates, the probability of their occurrence and the timing of seal failure.

In the dependency

matrices, notes for the impact of LOIA indicate that many valves and dampers fail in the "safe" or a beneficial mode on loss of instrument air.

Since the analysis assumes that instrument air is always failed (Page 3.1-88),

please address whether all these valves and dampers are considered as always successful in moving to their "fail safe" position or are they modeled in their respective systems as having some probability of failure.

Discuss your assessment of the impact of this assumption on the results of the analysis.

While you have indicated that common cause failure and maintenance were included in the analysis, there is no discussion of their impact on the results.

Please describe the impact of these considerations on the results as they contribute to those systems, split fractions and major accident types whose contributions are significant.

Discuss the insights you have derived from the results.

The plant damage state matrix does not explicitly address the timing of core melt.

Please discuss how core melt timing is included in your analysis, its correlation with other parameters (such as pressure) and its impact on the results (potential releases) of your analysis.

The discussion on page 3. 1-98 on PDS parameters indicates that only a limited amount of water will be in the reactor cavity prior to "melt-through" even if the RWST is injected.

'However, Table 3. 1.6-1 (PDS Definitions) indicates that it is assumed that water is present if the RWST is injected.

Please address. the rationale for this assumption and describe its impact on the results of the analysis especially for vessel failure at low RCS pressure when it is unlikely that the cavity is filled with water (Section 4. 1.3. 1).

It is indicated that all other data used in the quantification has been updated to December 1989, except the alpha factor distributions for common cause failures.

Discuss your investigation of plant-specific common cause failures and the rationale for not incorporating the latest data available for common cause failures.

32.

Please discuss any prevention or mitigation measures which address the significant contributors from those systems and/or actions that have been identified in Tables 3.4.2-1 through 3.4.2-7.

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Please provide a discussion addressing the sensitivity of the DHR systems to failure '(partial and total) of the support systems and assumptions for relevant operator actions.

Section 3.3.3.4.3, indicates that "Restoration of offsite power and operation of the turbine driven AFW Pump are assumed to continue after loss of D.C. power if a portable generator is successfully installed to provide 120V AC power for instrumentation".

Table 3.3.3-2, human action failure rate distributions, does not appear to contain the above action.

Was credit given for this action?

'What was the failure rate assigned and what impact did it have on the analysis results'5.

36.

Section 3.3.3.4.3 (scenario specific calculations) indicate that based on plant operating data the batteries are estimated to be available for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> upon complete loss of AC power.

However, Section 4.6.1. 1.2 (HANNI: RCP seal failure; stuck open PORVs and PSVs; AFW for four hours) states that the four hours that the AFW is available corresponds to the time that it takes to deplete the batteries after station blackout.

Please explain the difference in the battery depletion time identified in each case and identify the value that is used throughout the IPE for battery depletion time.

Were multiple values used'f so, provide the basis.

The discussion in the support tree section for top event GF indicates that the mission time for the diesel generators is 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, but does not provide any basis for using 6 instead of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Please discuss your basis as it relates to offsite power recovery.

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