ML20044B741
| ML20044B741 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/23/1993 |
| From: | Andersen J Office of Nuclear Reactor Regulation |
| To: | Opeka J CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| GL-88-20, TAC-M74432, NUDOCS 9303030327 | |
| Download: ML20044B741 (6) | |
Text
February 23, 1993 Docket No. 50-245 Di st ribution:
Docket File RHernan NRC & Local PDRs CAder Mr. John F. Opeka PD I-4 Plant JFlack Executive Vice President, Nuclear SVarga Connecticut Yankee Atomic Pcwer Company JCalvo Northeast Nuclear Energy Company SNorris Post Office Box 270 JAndersen Hartford, Connecticut 06141-0270 OGC ACRS (10)
Dear Mr. Opeka:
LIDoerflein, RI
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON MILLSTONE NUCLEAR POWER STATION, UNIT 1 - INDIVIDUAL PLANT EXAMINATION (IPE) SUBMITTAL -
GENERIC LETTER 88-20 (TAC NO. M74432)
By letter dated March 31, 1992, Northeast Nuclear Energy Company (NNECO) submitted the Millstone Nuclear Power Station, Unit 1 IPE results for NRC review.
Based on the NRC staff's review of your submittal, we have determined that additional information is needed to continue the review.
The enclosed list of questions identifies the information needed. We have found with other licensees that a conference call is very useful prior to submitting your formal response to these questions.
Please review these questions so that a conference call can be scheduled in about 2C-30 days to discuss your responses. The purpose of the call is to give your staff and ours an opportunity to understand the issues and what information the NRC needs to finish this review.
It is likely that the call can result in more direct and possibly shorter written answers to the questions.
We will require NNECO's written response within 60 days of the date of this letter.
Please contact me should you have any questions regarding this request.
This requirement affects one respondent and, therefore, is not subject to Office of Management and Budget review under P.L.96-511.
Sincerely, Original signed by James W. Andersen, Acting Project Manager Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
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.o Mr. John F. Opeka Millstone Nuclear Power Station Northeast Nuclear Energy Company Unit I cc:
Gerald Garfield, Esquire R. M. Kacich, Director Day, Berry and Howard Nuclear Licensing Counselors at Law Northeast Utilities Service Company City Place Post Office Box 270 Hartford, Connecticut 06103-3499 Hartford, Connecticut 06141-0270 W. D. Romberg, Vice President D. O. Nordquist Nuclear, Operations Services Director of Quality Services Northeast Utilities Service Company Northeast Utilities Service Company Post Office Box 270 Post Office Box 270 Hartford, Connecticut 06141-0270 Hartford, Connecticut 06141-0270 Kevin McCarthy, Director Regional Administrator Radiation Control Unit Region I Department of Environmental Protection U.S. Nuclear Regulatory Commission State Office Building 475 Allendale Road Hartford, Connecticut 06106 King of Prussia, Pennsylvania 19406 Allan Johanson, Assistant Director First Selectmen Office of Policy and Management Town of Waterford Policy Development and Planning Division Hall of Records 80 Washington Street 200 Boston Post Road Hartford, Connecticut 06106 Waterford, Connecticut 06385 S. E. Scace, Nuclear Station Director P. D. Swetland, Resident Inspector Millstone Nuclear Power Station Millstone Nuclear Power Station Northeast Nuclear Energy Company c/o U.S. Nuclear Regulatory Commission Post Office Box 128 Post Office Box 513 Waterford, Connecticut 06385 Niantic, Connecticut 06357 H. F. Haynes, Nuclear Unit Director Millstone Unit No. I Northeast Nuclear Energy Company Post Office Box 128 Waterford, Connecticut 06385 Nicholas S. Reynolds Winston & 5trawn 1400 L Street, NW Washington, DC 20005-3502 a
RE0 VEST FOR INFORMATION
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RELATED TO INDIVIDUAL PLANT EXAMINATION FOR MILLSTONE NUCLEAR POWER STATION. UNIT I DOCKET ND. 50-245 This request for information includes the following:
I.
Describe the process you used to collect and examine plant-specific operating data (e.g., licensee event reports and equipment operating histories) and the findings of your examination. Describe how you used the plant-specific data in your IPE.
2.
Provide your estimate related to the medium loss of coolant accident frequency. Also, provide the event tree (along with event sequence diagrams, if any) for the main steam isolation valve closure event (T4 event) or its equivalent.
I 3.
A discussion related to the frequency estimate of the inadvertent opening of relief valve event is not provided.
Provide such a discussion.
4.
Provide the rationale for not developing a separate event tree for the loss of a single 125V DC train, including a summary of failure causes for the 125V DC train.
5.
Discuss the process used (failure mode effects analysis) to identify i
applicable system / component failures and their root causes.
Include in your discussion a few systems as examples.
6.
Discuss the modeling of the failure combinations of the safety actuation systems and the reactor protection systems (for example, reactor water level sensor failure).
Have you examined overfilling of the reactor vessel and related consequential (main steam line) failures? Discuss your findings and conclusions.
7.
Provide the IPE results on the heating, ventilation, and air conditioning (HVAC) dependencies.
Include in your discussion the dependencies of air
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handling units, chillers, temperature control circuits, and dampers.
Discuss how the IPE treated a loss of HVAC in a room which contains essential equipment.
8.
Is the addition of alternate makeup water to the depressurized vessel from-the diesel-driven fire water pump practiced periodically in the facility j
simulator or its equivalent? Discuss the problems, if any that the operators might experience.
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Also, discuss the operator actions for overriding interlocks in a scenario which eventually leads to the use of diesel-driven fire water pump. Your discussion should include considerations such as availability of equipment used to override interlocks (for example, jumper cables of various sizes),
i availability of associated procedures, and adequacy of hands-on training.
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_2 9.
Is AC power shared between Unit I and Unit 2 under station-blackout conditions?
If so, please discuss how the inter-unit electrical connection is treated in the IPE.
- 10. What is the freeze date for data analysis for the IPE?
- 11. Discuss how your IPE process evaluated test and maintenance (T&M)-
activities and data.
List T&M unavailabilities for applicable systems.
- 12. Discuss your rationale for using random failure data for failure-to-close probability of valves IC-1 and IC-2 following a downstream line break in the isolation condenser system. Discuss the effects (if any) on valve operability because of environmental conditions.
- 13. NUREG-1335 states that:
"the submittal should contain as a minimum, a description of the internal review performed, the results of the review team's evaluation, and a list of review team members." Consistent with NUREG-1335 reporting guidelines, discuss your staff's involvement in the in-house peer review process. Describe, for example, the level of effort, personnel involved, types of activities performed (spot checks, audit calculations, etc.), utilization of any procedures or checklists, overall findings and conclusions.
- 14. With respect to the flood analysis, have you examined the isometric drawings related to the drainage piping system (floor drains and equipment drains), their check valve failures (including errors introduced during test and maintenance actions), and connecting pipes between two rooms containing independent safety components? If so, discuss your results and conclusions with respect to the risk from internal flood.
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- 15. Per Section 2.1.6 Item 5 in NUREG-1335, provide a definition of criteria used to characterize a vulnerability.
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- 16. With respect to the decay heat removal evaluation (Section 3.4.3),
i feedwater coolant injection (FWCI) system was mentioned, where a pre-l designated train of the feedwater system would provide the emergency injection water. How does unavailability of the FWCI system impact the l
quantification of core damage frequency due to various accident sequences (for example, the loss of feedwater transients or overcooling events)?
Please discuss your findings and conclusions.
- 17. Numerous assumptions are made throughout the back-end analysis. The bases for the assumptions are frequently not pravided.
Page 2-10 of NUREG-1335 l
states that " General assumptions used in the modeling of phenomenology are just as important as the models themselves and therefore should be fully described." Please describe the important assumptions used in the modeling of phenomenology in the back-end analysis.
- 18. Discuss how survivability of key systems and components were considered in the back-end analysis.
Your discussion should include various environmental effects.
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- 19. Discuss how you treated dual usage in your IPE (e.g., use of low pressure coolant injection and shutdown cooling when only one train is available).
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- 20. Provide a complete list of the types of dependency considered in your IPE
- 21. Phenomenological effects resulting from the accident progression can have a large impact on the resultant source term.
It is not clear how l
phenomenological effects were addressed in the containment event tree i
(CET) quantification, particularly since small CETs were developed with supporting fault trees. Discuss how phenomenological effects impacted the l
applicable containment failure modes and mechanisms, and describe how they were integrated into the accident progression of the CETs.
In Appendix F, probabilities can be extracted from the fault trees for various phenomena, e.g., missiles generated pierce drywell, pedestal failur,e at reactor pressure vessel failure causes drywell failure, high pressure blowdown overwhelms vapor suppression, direct impingement of molten material on drywell shell causes failure, energetic core coolant interaction in-vessel or ex-vessel, direct heating due to high pressure-blowdown dispersal causes failure, with values of IE-9, IE-3, IE-9, IE-3, IE-4 and IE-3, respectively. The bases for these values are not, however, clear.
List all applicable phenomena considered in the CETs with their j
values and provide bases for the probabilities.
- 22. Provide a discussion of " representative" pathways of radioactive material for each CET describing the accident progression and how containment failure occurred.
- 23. Containment failure pressure probabilities are provided in the submittal for failure location and size at various temperatures. A distribution for median and lower failure pressures are also provided. Discuss how these various containment failure locations and sizes were incorporated into the CET analysis.
It is not clear how the " uncertainty" of containment failure was considered in the study. When determining if a containment failure would occur, were the lower pressures of the distribution considered, or was only the " median" evaluated?
- 24. It is not clear if personnel locks and equipment hatches were addressed in the analysis. Discuss how these items were accounted for in the containment failure characterization.
- 25. It is not clear in the documentation if the various phenomena, though concluded not to present a challenge to containment integrity, could nonetheless challenge system responses. Discuss how phenomenological effects potentially impacted system responses.
- 26. Table 2.2 of NUREG-1335 provides a list of potential containment failure modes and mechanisms.
It is not clear that thermal attack of containment penetration was considered. Discuss the treatment of this potential l
containment failure mechanism in the IPE.
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- 27. Provide a copy of Reference 4.4.7-1: Khalil, Y.F., " Evaluation of Millstone Unit-1 Containment Isolation Failure," Northeast Utilities, r
April 1991.
- 28. "Small" CETs were developed with supporting fault trees. Computer codes l
CAFTA and ETA were used to quantify the CETs. Two problems can be encountered when using this approach: (1) the rare event approximation in CAFTA can lead to incorrect estimations, and (2) with failures and successes and therefore the use of "NOT" gates, dependencies among events can be difficult to model resulting in erroneous cut sets. Discuss how these two items were addressed in the quantification process.
- 29. In Section 4.4.9 and 4.4.10, conditional probabilities of IE-3 and IE-4 were assigned for early containment failu,e from direct containment heating and for core concrete interaction, respectively.
Provide the bases for these events.
- 30. It is not clear what were the specific human actions considered in the back-end analysis.
Provide a list of human actions, the associated probabilities, and the bases for the values.
It is not clear in the documentation if the various phenomena, though concluded not to present a challenge to containment integrity, could nonetheless challenge human responses. Discuss the effects of phenomena on human responses in the back-end analysis.
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