ML20203B202
| ML20203B202 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 12/04/1997 |
| From: | Huffman W NRC (Affiliation Not Assigned) |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9712120371 | |
| Download: ML20203B202 (5) | |
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[a December 4, 1997 4
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, i Mr. NichotAs J. Liparulo, Manager
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i Nuclear Safety and Regulatory Analysis Nuclear and Advanced Technology Division i J Westinghouse Electric Corporation -
P.O. Box 355
,Pittsburgh, PA 15230
SUBJECT:
OPEN lYEMS ASSOCIATED WITH THE AP600 SAFETY EVALUATION REPORT (SER) ON THE AP600 TRANSIENT AND ACCIDENT ANALYSES
Dear Mr. Liparulo:
The Reactor Systems Branch of the U.S. Nuclear Regulatory Commission has provided e partial SER input to the Standardization Project Directorate on the AP600 transient and accident analyses, SER Chapter 15. The input has open nems which have been extracted and desig-nated as an final safety evaluation report open items in the enclosure to this letter, it you have any questions regarding this matter, you may contact me at (301) 415-1141.
Sincerely, original signed by:
Wiliam C. Huffman, Project Manager l
Standardization Project Directora* s Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003
Enclosure:
As stated cc w/ encl: See next page DISTRIBUTION:
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l Mr. Nicholas J. Liparuto Docket No. 52 003 Westinghouse Electric Corporation AP600 cc:
Mr. B. A. McIntyre Mr. Russ Bell Advanced Plant Safety & Licensing Senior Project Manager, Programs Westinghousa Electric Corporation Nuclear Energy Institute Energy Systems Business Unit 1776 i Street, NW P.O. Box 355 Suite 300 Pittsburgh, PA 15230 Washington, DC 20006-3706 Ms. Cindy L. Haag Ms. Lynn Connor Advanced Plant Safety & Ucensing Doc-Search Associates Westinghouse Elec.tric Corporation Post Office Box 34 Energy Systems Business Unit Cabin John, MD 20818 Box 355 Pittsburgh, PA 15230 Dr. Craig D. Sawyer, Manager Advanced Reactor Programs Mr. Sterling Franks GE Nuclear Energy U.S. Department of Energy 175 Curtner Avenue, MC-754 NE 50 San Jose, CA 95125 19901 Germantown Road Germantown, MD 20874 Mr. Robert H. Buchholz GE Nuclear Energy Mr. Frank A. Ross 175 Curtner Avenue, MC-781 U.S. Department of Energy. NE-42 San Jose, CA 95125 Office of LWR Safety and Technology 19901 Germantown Road Barton Z. Cowan, Esq.
Germantown, MD 20874 Eckert Seamans Cherin & Mellott 600 Grant Street 42nd Floor Mr. Charles Thompson, Nuclear Engineer Pittsburgh, PA 15219 AP600 Certification NE-50 Mr. Ed Rodwell, Manager 19901 Germantown Road PWR Design Certification Germantown, MD 20874 Electric Power Research Institute 3412 Hillview Avenue Palo Alto, CA 94303 I
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SER CHAPTER 15 OPEN ITEMS ASSOCIATED WITH THE AP600 TRANSIENT AND ACCI.
DENT ANALYSES 440.740F in RAI 440.26, the staff asked Westinghouse to submit an ATWS analysis demonstrating that the AP600 ATWS response is within the bounds considered by the staff during its deliberations leading to the ATWS rule (10 CFR 50.62). In its response, Westinghouse provided the results of an ATWS analysis for the staff to review. After the staff reviewed Westinghouse's analysis, the staff asked Westinghouse to provide more information. As noted in RAI 440.655, it is the staffs understanding that Westinghouse had used a complete loss of normal feedwater (LONF) event for the ATWS analysis because the LONF event was previously identified as the limiting case that produced the maximum RCS pressure for conventional Westinghouse PWRs. Since the AP600 passive design is differer,t from conventional PWRs (for example, AP600 relies mainly on the PRHR system instead of the auxiliary feedwater system t.s required by the existing PRWs tc remove the decay heat during an ATWS event), Westinghouse was requested to show that the results of and the methodology used for the existing ATWS analyses were applicable to AP600 for analysis of the worst ATWS case. The acceptanco criterion used by Westughouse and approved by the NRC staff for the ATWS analysis is:
The ATWS must show that the unfavorable exposure time (UET), given the cycle design (including the MTC), will be less than 5 percent, or equivalently, that the ATWS pressure limit will be met for at least 95 percent of the cycle. The UET is the time during the cycle when reactivity feedback was sufficient to maintain pressure under 22.06 MPa (3203 psi) for a given reactor state.
This acceptance criterion was stated in a letter dated July 27,1995, from the NRC to D. L. Farer of the Commonwealth Edison Company. In RAI 440.659, Westinghouse was requested to show that its ATWS analysis for AP600 complied with the acceptance criterion stated above.
Subsequently, the staff discussed this position with Westinghouse. Westinghouse proposed to identify the worst risk significant ATWS scenarios, and to perform plant analyses of these scenarios to demonstrate that the LONF scenario is bounding. The staff will review this response when it has been submitted. This is an open item.
440.741F The basic requirements for the pressurized-water reactor manufactured by Westinghouse are specified in paragraphs (c)(1) and (c)(2) of 10 CFR 50.62 (ATWS rule), which state, in part:
Each pressurized water reactor must have equipment from sensor output to final actuation device, that is diverse from the reactor trip system, to automatically initiate the auxiliary (or emergency) feedwater and initiate a turbine trip under conditions indicative of an ATWS... [and] must have a diverse scram system from the sensor output to interruption of power to the control rods.
The AP600 design has included a control-grade ATWS mitigation i,ystem actuation circuitry (AMSAC) to provide an attemate reactor trip signal and an altemate decay heat removal actuation signal separate and diverse from the safety-grade reactor trip system. The staffs detailed review of Westinghouse's compliance with the ATWS rule is discussed in Sec-tions 7.7.1.12 and 7 7.2 of this report. The staff concludes that Westinghouse has met the intent Enclosure
i 2-of the ATWS rule by relying on the passive residual heat removal system to remove the decay heat. However, Westinghouse should submit a request for exemption from the 10 CFR 50.62 requirement for an auxiliary or emergency feedwater as an attemate system for decay heat removal during an ATWS event.
440.742F in RAI 440.723, the staff has questioned the operability of the pressurizer safety valves to function for extended periods of time (up to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) during pressurization transients (loss of AC power, loss of normal feed, inadvertent operation of the ECCS, and CVS malfunctions). In Westinghouse's response to this RAI (NSD-NRC-97 5384 dated October 17,1997), Westing-house stated that "there is insufficient stored energy in the system to cause the valves to fully open, and valve operation can be characterized as weeping. An evaluation of a typical spring-loaded safety valve under these conditions shows that the valves will be fluttering near the seat with a small amplitude. The effects of this motion near the seat may be that the valves willleak after tha transient. However, there will be no structural damage that will cause the valves to fail to reseat. Safety valve operation under these conditions is acceptable because their operation is with steam at the inlet to the valve. Operation under these conditions are within their capabilities."
As discussed with Westinghouse during a telephone conversation on October 23,1997, the staff does not agree with Westinghouse's conclusions. The staff believes that the safety valves will fully open when the safoty valve setpoint is reached and significant blowdown will take place.
Westinghouse has not demonstrated that the resultant level swellin the pressurizer will not result in the passing of liquid through the safeties which can result in valve failure. This remains and open item.
440.743F The staff has requested quantitative analyses of the AP600 for a steam line break (SLB) at power (RAI 440.722) to demonstrate that the SLB under no-laod conditions is bounding.
Westinghosue has not yet responded to this RAI and it remains an open issue.
440.744F Recent experimental data show failures of high bumup fuels at lower enthalpy than the Regula-tory Guide 1.77 fuel failure enthalpy limits. However, analyses performed by Westinghouse that assume low enthalpy fuel failures indicate that the radiological consequences of rod aection accidents might not exceed design basis limits for cores opereting witnin current NRC approved bumup limits. The analyses are docurnented in a Westinghouse submittal, NTD-NRC 95-4438,
" Westinghouse Assessment of Topical Report Validity for Reactivity insertion Accidents with High Bumup Fuel". The applicant has not documented the applicability of these analyses to the AP600. This is an open item.
440.74SF The staff has reviewed the consequences of the spectrum of postulated fuelloading errors and found that the analyses provided by the applicant shows, for each case considered, that either the error will be detectable by the available instrumentation (and hence remediable) or the error will be undetectable, but the offsite consequences of any fuel rod failures is a small fraction of
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3-10 CFR Part 100 guidelines, thus satisfying the acceptance criteria of SRP 15.4.7. Therefore the staff concludes that the analysis is acceptable. However, tne SSAR should document a COL action item for the plant specific applicant to confirm that the available in-core instrumentation will be used before the start of a fuel cycle to search for fuelloading errors. This is an open item.
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