ML19322C877
| ML19322C877 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Sequoyah, Diablo Canyon, McGuire, Zimmer, Crane  |
| Issue date: | 10/12/1979 |
| From: | Major R, Savio R Advisory Committee on Reactor Safeguards |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19322C876 | List: |
| References | |
| ACRS-MS-0585, ACRS-MS-585, NUDOCS 8002010162 | |
| Download: ML19322C877 (5) | |
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UNITED STATES
'g NUCLEAR REGULATORY COMMISSION
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y ADVISORY COWNITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C,20566
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October 12, 1979 D. B. Vassallo, Acting Director, Division of Project Management ACRS QUESTICNS REIATING TO CERMIN 1 EAR-TEIM OL PIANTS (DIABLO
SUBJECT:
CANYON, ZDGER, SEQUORH, AND MCGUIRE)
Attached for your use are the questions which have been prepared for con-sideration during the anticipated ACRS 9_hittee meetings for near-tenn operating licenses on the Diablo Canyon, Zimmer, McGuire, and Sequoyah nuclear plants.
Please let me know when your staff can be prepared to discuss these issues so that we can proceed with ACRS consideration of these projects.
A written response is preferred but the 91Wittee would be prepared to discuss these matters orally to facilitate progress on these matters.
P,s-Richard Savio Staff Engineer ACRS Staff Richard K.
jor Reactor Engineer ACRS Staff
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,O ACRS QUESTIONS RELATING TO CERTAIN NEAR-TERM OL PLAN'IS (DIABLO CANYON, ZD94ER, SEQUOYAH, AND MCQUIRE) i l
I I.
DIABLO CANYON AND SIMIIAR PIANTS:
i W at if anything does the NRC Staff think may warrant special l
1.
consideration because of seismic considerations?
Has the Staff considered whether special considerations are re-2.
quired with regard to operator response for a severe earthquake?
Wat anomalies in system behavior during an earthquake should 3.
operators be trained to handle? Is anything special required because of the failure of non-seismically qualified equipment?
m at are the assumptions concerning the failure of non-seismic 4.
Class I piping?
To whe.t extent can the failure of such piping be tolerated?
Is anything special required with regard to reliability of con-5.
nections to the Refueling Water Storage Tank for the earthquake situation? What criteria do the connections meet?
m at significance is attached (if any) to recent cases of pipe 6.
cracking in stagnant borated water lines as it applies to earthquakes?
7.
mat consideration does the Staff believe appropriate for systen degradation, such a; the recent feedwater nozzle cracking ex-perience, as it applies to inservice inspection programs for plants in areas of high seismic activity?
8.
m at are the seismic classes of:
a.
PORV b.
Block Valve Equipment related to the operability of these devices c.
d.
Pressurizer heaters and related equipment mat are the specific recomendations for the Auxiliary Feedwater 9.
Systan at Diablo Canyon? What is the dependence on AC power?
Are there seismic effects in the control room which require attention?
10.
Has special consideration been given to structures, equipment, and instrumentation in the control room for an earthquake situation?
Pbr example: Has the ceiling been analyzed? Will lighting be ade-quate? Will lighting fixtures and lights remain in place?
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.. 11. mat is the status of control room instnanentation displays,
.how rapid is the plant process computer (delay between printing and real time), and what effect would an earthquake have?
How comprehensive are tests for electrical transients during 12.
What is the an earthquake and what is the effect on equipnent?
reliability of both off-site and on-site power?
- 13. W at are the Staff conclusions regarding technical support capa-bilities for Diablo Canyon.
ZD9ER AND SIMIIAR PIRf1S:
1.
mat are the Staff's specific recomendations on BMts as they related to 1MI Implications - Lessons Imarned? How will they be implemented?
2.
mat has been the review for the reliability of decay heat removal systems for ar.amalous transients?
3.
Has the Staff reviewed procedures for transients and accidents?
Wat has been the conclusion?
7 4.
W at has been the NRC Staff's consideration of the advantages and disadvantages of a filtered and vented containment (see attached,
" Additional Carments" from Interim Report No. 3 on Three Mile Island Nuclear Unit 2, hay 16, 1979)?
5.
Are thera procedures for cases s ere one train of a system is down for maintenance and the other train fails?
6.
During transients, what actions could an operator take to fur-ther aggrevate a situation? Are there procedures noting actions operators should not take?
7.
W at in the way of a systems interactions study has been performed at Zimer? How is the potential for adverse systems interactions explored?
W at is the type of display and use of instrumentation describing 8.
the status of the core?
How fast can the plant process computer respond to severe transients?
9.
How close to real time is the printed output in the control room?
To what extent has the Staff considered anomalous feedwater transients, 10.
such as overfill, in BWRs?
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-3 SEQUOYAH AND POGUIRE AND SIMIIAR ICE CONDENSER /UF.1.PI.MRS III.
Discussion as to special features of the ice condenser /URI plants 1.
which would require considerations in the light of the 'IMI-2
'!he discussion should include hydrogen control accident.
and inert gas blockages in the primary systen.
Discussion as to the information which would be available to the 2.
operator in the control room in the event of a severe transient.
Discussion of the NRC Staff's review of the emergency control 3.
room procedures for Sequoyah and McGuire.
Status of the NRC Staff's review of possible plant transients, 4.
including a discussion as to what extent the Staff has looked at ways in which operator action may increase the consequences of the transients.
Discussion as to the reliability of the decay heat removal systems 5.
following anomalous transients.
itat has been the NRC Staff's consideration of the advantages and 6.
disadvantages of a filtered and vented containment?
What are the specific recommendations for the auxiliary feedwater 7.
What is the dependence on AC systems for Sequoyah and McGuire?
power?
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Honorable Joseph M. Hendrie May 16, 1979 Additional coments by Messrs. B. Imwis, D. Neller, D. Okrant, and J. Ray are presented below.
S neerely, Max W. Carbon Chairman 4
Additional Coments by Messrs. H. Lewis, D. Moeller, D. Okrent, and J. Ray The potential for a reduction in risk to the public in the case of a ser-fous reactor accident by the implementation of a means for controlled, filtered venting of a containment which could retain particulates and the bulk of the iodine has been recognized for more than a decade. Se concept was recomended for study more recently in the American Physical Society Report on Ifght-water reactor safety and in the Ford Foundation-Mitre Report, ' Nuclear Power - Issues and Choices." It is a high pri-ority item in the NRC plan cubnitted to Congrttss for Research to Improve the Safety of Light-Water Nuclear Power Plants (NUREG-0438). Se study performed for the State of California on underground siting concluded that filtered, vented contairunent was a favored option to explore in con-nection with possib)e means to mitigate the consequences of serious re-actor accidents. However, little progress has been made on the develop-ment of sufficiently detailed design information on which to evaluate the efficacy and other factors relevant to a decision on possible implementa-tion of such consequence ameliorating systems.
n e 1MI-2 accident suggests that the probability of a serious accident in which a filtered vented containment could be useful is larger than many had anticipated.
We recomend that the Comission request each power reactor licensee and construction permit holder to perform design studies of a system d ich j
adds the option of filtered venting or purging of containment in the j
event of a serious accident. Se system should be capable of withstand-1 ing a steam and hydrogen environment and of removing and retaining for j
as long a time as necessary radioactive particulates and the great bulk i
of the iodine for accidents involving degraded situations up to and in-1 cluding core melt. Such studies could be done generically for several reactor-containment types, and should evaluate the practicality, pros and cons, the costs, and the potential for risk reduction. A period of about twelve months for a report to the ac by licensees and construction permit holders appears to represent a possible schedule.
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