ML20202E595
| ML20202E595 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 10/22/1997 |
| From: | Joseph Sebrosky NRC (Affiliation Not Assigned) |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9712080086 | |
| Download: ML20202E595 (6) | |
Text
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October 22, 1997
= Mr. Nichclas J. Liparulo, Manager -
Nuclear Safety-and Regulatory Analysis l
Nuclear.and Advanced Technology Division
.Westinohouse Electric Corporation P.O. Box 355 Pittsburgh, PA 15230
SUBJECT:
OPEN ITEMS ASSOCIATED WITH CHAPTER-19 0F THE AP600 SAFETY EVALUATION REPORT (SER)
Dear Mr. Liparulo The Containment Systems and Severe Ancident~ Branch has provided an SER for a portion of Chapter 19. However, the input to these sections contained some
-open items. These open items have been extracted from the SER and can been found in the enclosure to this letter.
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_ your request in accordance with the requirements of 10 CFR 2.790, that portion of
.the submitted information is being withheld from public disclosure pending the staff's final determination. The staff concludes-that these follow on ques-tions do 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 Westinghouse the opportunity to verify the-staff's-conclusions.
If, after that time, you do not request that all or portiens 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 Nuclear Regulatory Commission Public Document Room.
If you have any questions regarding this' matter, you may contact me at (301) 415-1132.
Sincerely, original signed by:
Joseph M. Sebrosky, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Rea: tor Regulation Docket No.52-003 k g pqr m -m omny MMJNS U U4 MM#
Enclosure:
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=QISTRIBUTION: See'next page
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Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Ms. Cindy L. 4aag Advanced Plant Safety & Licensing Advanced Plant Safety & Licensing Westinghouse Electric Corporation Westinghou,e Electric Corporation Energy Systems Business Unit Energy Systems Business Unit P.O. Box 355 Box 355 Pittsburgh, PA 15230 Pittsburgh, PA 15230 Enclosure to be distributed to the following addressees after_the result of the proprietary evaluation is received from Westinghouse:
Mr. Russ Bell Ms. Lynn Connor Senior Project Manager, Programs DOC-Search Associates Nuclear Energy Institute Post Office Box 34 1776 i Street, NW Cabin John, MD 20818 Suite 300 Washington, DC 20006-3706 Mr. Robert H. Buchholz GE Nuclear Energy Dr. Craig D. Sawyer, Manager 175 Curtner Avenue, MC-781 Advanced Reactor Programs San Jose, CA 95125 GE Nuclear Energy 175 Curtner Avenue, MC-754 Mr. Sterling Franks San Jose, CA 95125 U.S. Department of Energy NE-50 Barton Z. Cowan, Esq.
19901 Germantown Road Eckert Seamans Cherin & Mellott Germantown, MD 20874 600 Grant Street 42nd Floor Pittsburgh, PA 15219 Mr. Charles Thompson, Nuclear Engineer AP600 Certification Mr. Frank A. Ross NE-50 U.S. Department of Energy, NE-42 19901 Germantown Road Office of LWR Safety and Technology Germantown, MD 20874 19901 Germantown Road Germantown, MD 20874 Mr. Ed Rodwell, Manager PWR Design Certification Electric Power Research Institute 3412 Hillview Avenue Palo Alto, CA 94303
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OPEN ITEM ASSOCIATED WITH CHAPTER 19 720.418F Core Debris Coolability (Section 19.2.3.3.3)
Westinghouse has modified the design of the reactor cavity floor and the I
concrete curb surrounding the reactor cavity sump, and has not submitted any information regarding the updated design. Westinghouse needs to submit the design details and drawings for cavity floor and sump, and confirm'that:
(1) the elevations of the reactor cavity subcompartment, RCDT subcompartment and interconnecting ventilation duct, as well as the flow areas between these regions, are consistent with the Argonne National Laboratory analysis, (2) core debris will not pass into the sump via interconnecting pipelines embedded in the concrete. floor and/or sump curb, and (3) the sump curb (height and width) is adequately sized to prevent molten core debris from overflowing or ablating through the curb. The latter point should be demonstrated for:
the ccre debris masses in Table B-5 of the PRA, both RPV failure modes, and both concrete types considered in Appendix B of the PRA. Westinghouse should also explain why neglecting the impact of the RCDT supports / sump on debris spreading / height is appropriate. This is Open Item 19.2.3.3.3-1.
720.419F Core Debiis Coolability (Section 19.2.3 3.3)
Westinghouse's results for time of liner and basemat penetration are based on information provided in response to RAI 720.411, and need to be incorporated into Appendix B of the PRA.
In addition, Westinghouse should confirm the location of liner and basemat penetration for each RPV failure scenario and concrete type.
This is Open Item 19.2.3.3.3-2.
720.420F High-Pressure Core Melt Ejec' ion (19.2.3.3.4)
The staff notes several inconsistencies in Westinghouse's characterization of the pathways for debris transport to the upper compartment, specifically:
(1) the area around the reactor vessel flange is identified as a debris flow path in Section B.2 and Table B-2 of Appendix B, but RPV insulation drawings indicate that there is a permanent cavity seal ring here, and no flow path, (2) the flow area specified in Table B-2 for he annular openings between the coolant loops and the biological shield (16 m}) is substantially greater-than the steam relief area through the same pathway reported in,)the RPY vessel insulation discussion in Appendix K of DOE /ID-10460 (0.7 m, and (3) the impact of the RPV insulation and boro-silicone neutron shie'id blocks on the flow paths and areas does not appear to have been considered.
Westinghousa needs to resolve these apparent inconsistencies. The total flow area for debris transport into the steam generator compartment, and the flow area between the steam generator compartment and upper compartment assumed in the calculation should also be clarified.
This is Open Item 19.2.5.3.4-1.
Enclosure
. 720.421F Accident Management (19.2.5)
Westinghouse has adequately addressed design features to facilitate (or eliminate the need for) accident management in the AP600 design with the exception of containment venting. Although venting is not expected to be necessary in most sequences in AP600, it may be needed in the event of reactor vessel failure (since deterministic calculations indicate that early containment failure from steam explosion is not likely). Westinghouse has indicated that the AP600 has no containment vent. However, in Appendix D of the PRA (equipment survivability, Table D.6-1) high level actions to vent and to depressurize containment are called out, but related equipment is not identified or discussed. Also, in WCAP-13914, Revision 2 (Section 5.9), it is indicated that methods that may be used to vent the AP600 containment will be investigated during a later phase of the development of the severe accident management guidance, but nothing more has been provided. Although the development of detailed guidance and procedures is the responsibility of the COL applicant, the capability to vent and any associated equipment specifications should be established by the designer prior to Design Certification.
This is Open Item 19.2.5-1.
720.422f Accident Management (19.2.5)
WCAP-13914, Revision 2 does not address the need for the COL applicant to develop guidance and procedures for: (1) powering the hydrogen igniters from batteries, (2) containment venting in the event of core concrete interactions, and (3) post-72 hour actions. The report should be revised to include guidance in this regard.
This is Open Item 19.2.5-2.
720.423F PRA input to the Regulatory Treatment of Nonsafety-Related Systems (19.1.7)
In meeting the RTNSS criteria, credit was taken for external reactor vessel cooling (ERVC) as a strategy for retaining molten core debris in-vessel. This results in the majority of core melt accidents (~90 percent) being arrested in-vessel, thereby avoiding RPV failure and associated containment challenges from ex-vessel phenomena.
Successful RCS depressurization and reactor cavity flooding are prerequisites for ERVC, and credit for these aspects of ERVC in the focussed PRA is appropriate since both functions are fulfilled by safety-related systems. However, the nonsafety-related RPV thermal insulation system is also requited for successful ERVC.
The thermal insulation system limits thermal losses during normal operations, but provides an engineered pathway for supplying water cooling to the vessel and venting steam from the reactor cavity during severe accidents.
Attributes of the system include specific RPV/ insulation clearances and water / steam flow areas based on scaled tests, integral ball-and-cage check valves and buoyant steam vent dampers which change position during flood-up of the reactor cavity, and insulation panel and support members designed to withstand the hydrcdynamic loads associated with ERVC, l
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. If credit for ERVC is reduced, large release frequency and CCFP would increase proportionally since all RPV breaches are assumed to lead to early containment failure in the PRA. Under the most limiting assumption of no credit for ERVC, the large release frequency would approach the core melt frequency and CCFP would approach 1.0.
In view of the reliance on-ERVC to meet the Commission's large release frequency goals, the staff will require an appropriate level of regulatory oversight of the RPV thermal insulation system. This oversight should provide reasonable assurance that the as-built insulation system-conforms with design specifications contained in Chapter 39 of the PRA, and that the operability of the system is confirmed through periodic surveillance.
The RPV insulation design description and functional requirements are not currently included in the SSAR, ITAAC, or reliability assurance program. The design description and functional requirements for the RPV insulation should be added to the SSAR, and important criteria associated with the insulation design should be incorporated into the ITAAC, including information related to the necessary clearances / flow areas, and the check valves and steam vent dampers.
The system should be included as a risk-significant SSC in the reliability assurance program, and reliability / availability controls and goals should be provided, consistent with maintenance rule guidelines, to assure that operability of the system and moving parts is maintained.
This is an Open Item.
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Letter to Mr. Nichalas J:'lioarulo.' Dated: October 22. 1997:
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