ML20029D043
| ML20029D043 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 04/15/1994 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9405030324 | |
| Download: ML20029D043 (8) | |
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April 15, 1994 Docket No.52-003 Mr. Nicholas J. Liparulo Nuclear Safety and Regulatory Activities Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230
Dear Mr. Liparulo:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON THE AP600 As a result of its review of the June 1992, application for design certifica-tion of the AP600, the staff has determined that it needs additional informa-tion in order to complete its review. The additional information is needed in the areas of auxiliary systems (Q410.106)* and shutdown events (Q440.53-Q440.72).
Enclosed are the staff's questions.
Please respond to this request by June 30, 1994, to support the staff's review of the AP600 design.
The questions on shutdown events are general in scope and are not all-inclusive. As discussed with your staff during the March 24, 1994, meeting, the Combustion Engineering submittal on shutdown risk is an example of the documentation that the staff is requesting.
Because of the importance of safety implications, General Electric Company also included dedicated SSAR chapters documenting a comprehensive assessment on shutdown risk in their advanced reactors design certification application.
You have requested that portions of the information submitted in the June 1992, application for design certification be exempt from mandatory public disclosure.
While the staff has not completed its review of yow request in accordance with the requirements of 10 CFR 2.790, that portion of the submit-ted information is being withheld from public disclosure pending the staff's final determination.
The staff concludes that this request for additional information does not contain those portions of the information for which exemption is sought.
However, the staff will withhold this letter from public disclosure for 30 calendar days from the date of this letter to allow Westing-house the opportunity to verify the staff's conclusions.
If, after that time, you do not request that all or portions of the information in the enclosures be withheld from public disclosure in accordance with 10 CFR 2.790, this letter will be placed in the NRC's Public Document Room.
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Mr. Nicholas J. Liparulo April 15, 1994 This request for additional information affects nine or fewer respondents, and therefore, is not subject to review by the Office of Management and Budget under P.L.96-511.
If you have any questions regarding this matter, you can contact me at (301) 504-1120.
Sincerely, 1
Origittaf%~
Thomas ' Xenyon, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
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- HOLD CENTRAL FILE COPY FOR 30 DAYS OFFICIAL RECORD COPY: SHUTDOWN.RAI i
Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation.
AP600 cc: Mr. B. A. McIntyre Mr. Raymond N. Ng, Manager Advanced Plant Safety & cicensing Technical Division Westinghouse Electric Corporation Nuclear Management and Energy Systems Business Unit Resources Council P.O.-Box 355 1776 Eye Street, N.W.
Pittsburgh, Pennsylvania 15230 Suite 300 Washington, D.C.
20006-3706 Mr. John C. Butler Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Box 355 Pittsburgh, Pennsylvania 15230 Mr. M. D. Beaumont Nuclear and Advanced Technology Division Westinghouse Electric Corporation One Montrose Metro 11921 Rockville Pike Suite 350 Rockville, Maryland 20852 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.
20585 Mr. S. M. Modro EG&G Idaho Inc.
Post Office Box 1625 Idaho Falls, Idaho 83415 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Room 8002 Washington, D.C.
20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 Mr. Victor G. Snell, Director Safety and Licensing AECL Technologies 9210 Corporate Boulevard Suite 410 Rockville, Maryland 20850
REQUEST FOR ADDITIONAL INFORMATION ON THE WESTINGHOUSE AP600 DESIGN AUXILIARY SYSTEMS 410.106 Provide oversize P& ids for the circulation water, the condensate, and feedwater systems, SHUTDOWN EVENTS 440.53 The Diablo Canyon event of April 10, 1987, and the loss of ac power at the Vogtle plant on March 20, 1990, led the staff to issue NUREG-1449, " Shutdown and Low Power Operation at Commercial Nuclear Power Plants in the United States," providing an evaluation of the shutdown risk issue.
The scope of NUREG-1449 includes subjects such as operating experiences as documented in generic letters, operator training, technical specifications, residual heat removal capacity, temporary reactor coolant boundaries, rapid boron dilution, contain-ment capacity, fire protection, outage planning and control, and instrumentation.
The staff recognizes that some of the issues discussed in NUREG-1449 are the plant owners' responsibility because they relate to opera-tion, maintenance and refueling plans, procedures, and risk manage-ment. However, the staff believes that the level of defense-in-depth against shutdown events will be improved if clear guidance _is provided to the areas discussed above by the plant designer.
Therefore, perform a systematic assessment of the shutdown risk issue to address areas identified in the NUREG-1449 (particularly, Sections 5, 6 and 7 of the report) as they are applicable to the AP600 design and document the results in a dedicated section in the SSAR.
For the items that are not applicable to the AP600 design, provide technical bases to justify their non-applicability.
This analysis should include (a) an assessment of shutdown and low power risk, identifying design specific vulnerabilities and weakness, and (b) a demonstration showing consideration and incorporation of design features which minimize shutdown and low power risk probabil-ity.
For any deviations from the scope and guidance of NUREG-1449, provide bases to justify the technical adequacy of these deviations (see also Q440.55, Q440.56, Q440.58, Q440.71, and Q440.72).
440.54 The staff addressed concerns relating to rapid boron dilution during a PWR startup raised by the French regulatory-authority in NUREG/
CR-5819. The French authority postulated a scenario that starts with the highly borated reactor being deborated as part of a startup procedure.
The reactor is at hot condition with the reactor coolant pumps (RCPs) running and the shutdown banks removed.
Unborated or diluted water is being pumped by charging pumps from the volume control tank into the cold leg.
Enclosure
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The initiating event is a loss of offsite power that results 'in tripping the RCPs and charging pump, and scramming the shutdown i
rods.
The charging pump comes back on line quickly when the diesel generators start up.
Charging continues until the volume control tank is empty.
This diluted water is assumed to accumulate in the
-lower plenum.
It is then assumed that the offsite power is recov-ered and the RCPs are restarted. The RCP restart causes the slug of diluted water to rapidly pass through the core and results-in.a potential to cause a power excursion sufficiently large.to. damage the core.
Another variation to this scenario includes'an event i
i having the slug of deborated water through the core by inadvertent blowdown of an accumulator.
In light of these potential rapid boron dilution scenarios, show the adequacy of the AP600 design.by demon-strating that the rapid boron dilution events are incredible, the results are not serious if they occur, or proposing protective measures.
440.55 As stated in Section 6.7 of NUREG-1449, the staff has noted instances in which the failure of temporary RCS boundaries-(such as freeze seals used to temporarily isolate fluid systems, temporary plugs for nuclear instrument housings, and nozzle dams installed in the hot-leg and cold-leg penetrations to steam generators) can lead to a rapid non-isolable loss of reactor coolant.
This concern is of special importance in PWRs because the emergency core cooling system (ECCS) is not designed to automatically mitigate accidents initiated at pressures below a few hundred psig and is not normally. fully available for manual use during these conditions.. Address this concern with respect to failure of temporary boundaries in the AP600 design (see also Q440.53, Q440.56, Q440.58, Q440.71, and Q440.72).
440.56 a.
Provide a description of plant instrumentation designed to operate properly during shutdown and mid-loop operations.
The instrument accuracy, availability, appropriateness of key parameters (RCS level, RCS temperature, and RHR system perfor-mance), and the intended monitoring ranges should be addressed-for shutdown operations.
b.
Identify any devia+ ions, and provide the technical bases and justification for these deviations, from the guidance of NUREG-1449 (Page xiii, Sections 6.6.1.1 and 7.3.3 of the report) that requests each plant to' provide an_ independent and-diverse means of accurately monitoring RCS water level, the capability to continuously mo'nitor decay.- heat removal system (DHR) when a DHR system is being used for cooling the RCS, and visible and audible' indications of abnormal conditions in temperature, level, and DHR system performance-(see also Q440.53, 0440.55, Q440.58, Q440.71, and Q440.72).
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Identify safety-and non-safety-related instruments used during shutdown operations.
For the safety-related instruments, confirm that the instruments will be within the scope of envi-ronmental qualifications and quality assurance criteria.
For non-safety-related instruments, provide a description of the quality assurance program that will be used to provide instru-ments with accurate information in the expected ranges of shutdown measurement that will enhance operator confidence in the instruments, and the training program for operators to understand and interpret data provided by the instruments.
440.57 Provide a description of the Westinghouse emergency procedures guidelines (EPG) for the AP600 for the development of emergency operating procedures (E0P) for conditions including shutdown and mid-loop operations.
Specifically, address the adequacy of the EPGs (existing or to be proposed) for shutdown conditions when many systems will be out for maintenance and the plant is in a configura-l tion different from the normal plant operation.
440.58 Describe what changes have been incorporated into Chapter 16,
" Technical Specifications," of the SSAR for AP600 to deal with shutdown operations.
Identify any deviations from the guidance specified in NUREG-1449 (Sections 6.5 and 7.3.2) for shutdown Technical Specifications and justify the deviations with appropriate technical bases (see also Q440.53, Q440.55, Q440.56, Q440.71,-and Q440.72).
440.59 Provide a description of the design features that are incorporated into the AP600 design to increase the allowable water level operat-ing band during mid-loop operation to prevent a loss of decay heat removal capability and water flooding the steam generator and containment.
440.60 Provide a description of the design features and procedure guidance that are incorporated into the AP600 design to minimize the likeli-hood and consequences of loss of ac power during outage activities.
440.61 Provide a description of'the. design features that minimize shutdown-risk.
The description should include non-safety-related systems for normal shutdown operations, passive safety-related systems,:and the functional capability and availability of the active cystems' to-ensure defense-in-depth, accident mitigation, and core damage prevention capability. The design bases, functions and supporting-analyses for each system used to minimize shutdown risk should be discussed.
440.62 Provide a discussion of design features of non-safety-related systems for normal shutdown operations, procedures, and technical specifications that will reduce the risk to challenge passive safety systems and active non-safety-related systems for accident mitiga-tion during shutdown operations.
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440.63 Provide a discussion of analyses that demonstrate that the design of passive safety-related systems and active non-safety-related systems is adequate for mitigating the consequences of loss of ac power, inadvertent RCS drainage, and loss of normal residual heat removal system events during shutdown and mid-loop operaticns.
The informa-tion should include the assumptions and methods used in the analyses with a discussion of testing data to support the adequacy of the assumptions and methods, and analytical results.
l 440.64 Will there be procedures to prevent conducting maintenance activi-ties during reduced inventory operation that could disturb the RCS inventory or lead to a loss of non-safety-and/or passive safety-related RHR systems given a single malfunction?
l' 440.65 Outage and maintenance activities require only minimal isolation of important systems and components.
How does the AP600 design take this into consideration?
440.66 Transient and accident analyses presented in safety analysis reports typically concentrate on power operation.
The recent experience from the events in operating reactors indicate that further evalua-tion for the plant lower modes is needed.
Confirm whether each of the transient and accident analyses included in the SSAR for the-AP600 is applicable to modes 1 through 6, or provide a discussion of any plans, if any, with respect to the transient and accident analysis at lower operation modes.
440.67 Section 9.2 of Chapter 3 of the EPRI ALWR Requirements Document identifies requirements regarding the RHR system during mid-loop I
operation.
For example, paragraph 9.2.1.3 states that a single failure in the RHR system with reactor vessel head removal should not cause the water in the RCS or the reactor cavity to boil; paragraph 9.2.2.1.1 states that analysis should be performed for all potential RHR conditions that properly account for sources of error during mid-loop operation; paragraph 9.3.1.2 lists various features to prevent or mitigate the effects of losing suction to the RHR pumps when the RCS level is lower.
Table B.1-2 of Appendix B to Chapter 1 of the EPRI Requirements Document indicates that designers meeting the guidance of this document should comply with Generic letter 88-17 regarding loss of decay heat removal during mid-loop operation.
Address compliance of the AP600 RHR system during mid-loop operation with the EPRI Requirements Document, identify any deviations, and provide justification for each of the deviations identified (see also Q440.72).
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440.68 Section 2.3.3.7 of Chapter 1 of the EPRI ALWR Requirements Document states that all operating conditions (including shutdown operations) should be taken into account in the probabilistic risk analysis (PRA). Appendix A to Chapter 1 of the EPRI Requirements Document specifies guidance on the scope and contents of a PRA for shutdown conditions. Address compliance of the AP600 design to this guidance regarding an evaluation to identify system vulnerability for shut-down or mid-loop operation.
440.69 Generic Letter (GL) 87-12 requested information regarding lower reactor coolant system (RCS) inventory operation.
How does the AP600 design comply with GL 87-12?
440.70 a.
NUREG-1269 addressed the containment closure issue resulting from the Diablo Canyon occurrence on April 10, 1987, and iden-tified the need for procedures to reasonably ensure the capa-bility for containment closure in the event of progression of an accident to core damage conditions.
Address this contain-ment closure issue. The discussion should include design considerations such as the need for removal of the equipment hatch and improvements in the AP600 design that facilitate rapid replacement of the hatch should the need arise.
Simi-larly, address other containment penetrations and potential bypass paths, b.
NUREG-1269 states that "the design of the nuclear steam supply system did not appear to provide detailed provisions for mid-loop operation."
Identify and discuss each of the design i
changes in the AP600 design that establishes the adequacy of the AP600 design for reduced RCS inventory operation.
440.71 As discussed in Chapter 2 of NUREG-1449, the staff reviewed operat-ing experience to ensure that its evaluation encompassed the range of events encountered during shutdown and low-power operation. The data base considered by the staff includes licensee event reports, I
NRC internal reports, and various inspection reports to determine the types of events that take place during refueling, cold and hot i
i shutdown, and low-power operation.
Discuss the range of operating experience during shutdown operations considered in the AP600 design and demonstrate that the AP600 design adequately encompasses the shutdown events occurred ir, the industry (see also Q440.53, Q440.55, Q440.56, Q440.58, and Q440.72).
440.72 Generi: Letter 88-17 identified. action items to reduce shutdown risk.
Tne 14 action items as summarized in Table 5.2 of NUREG-1449 address issues ranging from operations, events, hardware design procedures, analyses and instrumentation. Address the proposed resolution to each of these action items for the AP600 design during shutdown and mid-loop operations (see also Q440.53, Q440.55, Q440.56, Q440.58, Q440.67, and Q440.71).
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