ML20196G045

Forwards Request for Addl Info Re Review of Questions Q440.242 & 440.262 to RESAR SP/90 Pda Application.Response Requested within 60 Days of Ltr Date

ML20196G045
Person / Time
Site:	05000601
Issue date:	03/02/1988
From:	Kenyon T Office of Nuclear Reactor Regulation
To:	Johnson W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 8803040329
Download: ML20196G045 (13)

Text

i March 2, 1988 Docket No. 50-601 Mr. W. J. Johnson Nuclear Safety Department Westinghouse Electric Corporation Water Reactor Division Box 355 '

Pittsburgh, PA 15230

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON RESAR SP/90 1

As a result of our ongoing review of the RESAR SP/90 PDA application, we-require additional information in order to complete our review of the reactor systems aspects of the design. Enclosed are review questions Q440.242-440.262.

Please respond to this request within 60 days of the date of this letter. If I you have any questions regarding this matter, call me at (301) 492-1120.

Sincerely, Driginal Signed By: i Thomas J. Kenyon, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

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Docket No. 50-601 Mr'. W. J. Johnson Nuclear Safety Department Westinghouse Electric Corporation Water Reactor Division Box 355 Pittsburgh, PA 15230 .

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON RESAR SP/90 As a result of our ongoing review of the RESAR SP/90 PDA application, we require additional information in order to complete our review of the reactor systems aspects of the design. Enclosed are review questions Q440.242-440.262.

Please respond to this request within 60 days of the date of this letter. If ,

you have any questions regarding this matter, call me at (301) 492-1120.

Sincerely,  ;

} yhi~. -

Tholnas J. Keny n, Project Manager i Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated cc: See next page 1

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RESAR-SP/90 Docket No. 50-601 cc: Brookhaven National Laboratory Building 130 .

Upton, New York 11973 Attention: Dr. Robert Bari b

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. ENCLOSURE 1 PE0 VEST FOR ADDITIONAL INF0PMATION ON THE PDA APPLICATION FOR WESTINGHOUSE ADVANCED PPESSURIZED WATER REACTOR (WAPWR)

DOCKET NIIPRER 50-601

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RESAR-SP/90 440.?'2 (Volume 1, Page 3-32) Discuss the safety design classification of the pressurizer heaters. If they are not designed to safety grade standards, confirm that the reactor can be brought to cold shutdown conditions without the operation of pressurizer heaters. BTP RSP 5-1 reouires that the plant he able to reach cold shutdown conditions using only safety grade ecuipment.

440.243 (Volume 1, page 3-11) You have indicated that Westinghouse will cutline a program for emergency response cuide*;ine development prinr to receiving a PDA for the WAPWR desion. Prrvide the above s?.ated outline for the staff review.

440.244 (Volume 1, page 3-42) You have stated that the feed and bleed mode for reactor coolant system operation can be used to remove decay heat from the reactor. Has any thermal-hydraulic analysis been performed to confirm the viability of the feed and bleed process based on the plant configuration of the WAPWR' If so, olease submit the results of the analysis, a40.245 THIS QUESTION INTENTIONALLY LEFT BLANK.

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440.?d6 (Volume 1, page 4-35) You have stated that the current desion basis is to be able' te accommodate a loss of all ac power for a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with an ultimate goal of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. . Please explain how the ultimate goal of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> station blackout could he achieved if the currgn,t design has only 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> capability for a complete loss of ac power.

440.?A7 (Volume 1, page 4-381 You have stated that the primary side and secondary side safeguards system options being considered for inclusion in the WAPWP design provide the capability of removing decay heat from the reactor core while maintaining sufficient water inventory to ensure adeouate core cooling. Confirm that all systems and components used to perform the above stated function will be designed to safety grade sisndards.

440.?48 (Volume 1, page 5-?ti Expand your discussion on the design criteria for the main steam and main feedwater isolation valves. Confirm that these valves will be designed to safety orade standards or that the transient and accident analyses will not give credit to these valves for performino their safety related function, ad0.249 THIS QUESTION INTENTIONALLY LEFT BLANK.

440.?50 (Volume 1, page 5-80) You have stated that passive failures which are considered to have a low probability may not be considered.

Please discuss the criteria for identifying the passive failures with low probability.

440.251 (Valume 1, item 21) Per the requirements of RTP PSR 5-1, please confirm that a boron mixino and natural circulation test will be performed in the first plant with WAPWP desion.

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440.252 (Module 1, Section 6.3.2) Yoy stated that the ECCS pumps are protected against low flow or no flow-operation by the miniflow l path. It is ont clear that how pump protection could be achieved by the miniflow lines under no flow or low suction pressure conditions.

It is the staff's position that the ECCS pump protection should be provided by a safety grade low flow alarm system and to assure that j the pump could withstand those operatirg conditions during'the tine l delay for operator actions to manually trip the pump in response to the alarms.

l 440.253 (Module 1. Section 15.6.4) The Westinghouse LOCA evaluation model dpproved by the staff may not be applicable to WAPWR desian with respect to plant specific confiourations in node arrangement and control systems. Confirm that a new LOCA evaluation model will be prepared for the WAPWR design. 1 440.25d (Module 3, Section 1.1.1.2) You stated that the WAPWR design includes a NSSS with a thermal rating of 3816 macawatts, which includes a core thermal power of 3800 magawatts plus 16 magawatts from the reactor coolant pump heat. Are the primary coolant heat losses included in calculating the NSSS thennal rating? If not, why not?

(Module 4, Section 5.7.?) Section 5.2 ? on page 5.?-3 states that 440.255 the liouid relief valves of the residual heat removal system (RPPS) are used to protect PCS at low temperaturee when the RHRS is in l

operation. Section 5.2.7.10 states that the pressurizer PORVs will

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be used for the low temperature overpressure protection (LTOPi l function. Please clarify the LTOP desian for the WAPWP.

440.?56 (Module 4, Section 5.P.?) Expand this section to address the )

assumptions used for a mass addition event relative to the LTOP l

system design. I l

4 440.257 (Module 4, page 5.2-11) Item'A states that to preclude inadvertent ECCS actuation during heatup and cooldown, blockage of the safety in,iection signal actuation looic below 1975 psia is required.

Discuss the impact of this design relative to a t.00A during modes 3 and 4,.

440.258 (Chapter 15) For transient and accident analyses of WAPWR, provide the following:

a) A list of all transient and accident analyses cross referencing the modules of the PDA application where each event is addressed.

b) Acceptance criteria for each transient and accident analyzed should be clearly stated, c) Initial conditions for each event including consideration of all modes of plant operation.

di For each event, include the best estinate analysis usino realistic plant data and emercency operating procedures and the licensing anal.vsis using most conservative assunptions and only safety grade eouipment for event mitigation, i

e) Identify the most limiting single failure used for the analysis o' each event with respect to different acceptance criteria of the event (e.g. Peak RCS pressure, ONR, radiolooical consequencesi.

440.259 (Module 9, page 15.0-3) item 2a indicates that plants may be operated at power with a rector coolant pump out of service. You l should provide analyses for N-1 loop operation to support the WAPWR design. l l

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5 440.260 (Podule 9, page 15.0-91 Discuss why the nominal plant operatina parameters are assumed for the analyses of transients and accidents which are DNR limited. For the events leading to increase of RCS pressure, why isn't a hiaher RCS pressure assumed as the initial ,

condi, tion. For each event, with respegt to peak RCS pressure or fuel performance, it is nessary to assure different initial conditions in the analyses in order to predict the worst consequences of the event.

440.261 (a) Generic letter (GL) 871.? requested information reaarding lowered PCS inventory operation. Please provide a res;nnse to the generic letter with respect to the PESAP-SP/90.

(b) Ple&se describe instrumentation provided to the operator during  ;

shutdown operations which characterire the state of the reactor  !

coolant system (RCS). include RCS level, RCS temperature, and residual heat removal IPPDi system performance and provide a >

description of the appropriateness and accuracy of each instrument with respect to its intended function. Also, include identification of audible and visual alarms used to delineate out-of-range conditions, includino the values which f constitute those conditions.

Ic) The staff has identified that Diablo Canyon,linit ?, was in a condition not previously analyzed by the NPC staff during the loss of RHR event of April 10, 1987 (NUREG-1?69). Please describe the steps that have been taken and the future plans which will be taken to alleviate this situation for the SP/90.

(d) NUREA-1269 contains the staterent "Design o# the nuclear steam supply system (NSSSI did not appear to provide detail provisions for mid-loop operation." Please address this identified deficiency in PWR design with respect to the SP/90 l

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design. Include identif,icetion of and discussion of each of l the design. changes in the SP/90 which represents and improvement over existing designs and establish the adecuacy of the SP/90 l design for lowered RCS inventory operation.

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l (e) NUREF-1?69 identified that containment was open throuahout the  !

April 10, 1987 event, and there were no procedures to reasonably  ;

assure containment closure in the event of prooression of the  !

accident to a core damage condition. Address this situation with respect to the SP/90 design and the anticipated methods that will be used to operate the plant. Include such desinn considerations as the need for removal of the equipment hatch and improvements in the SP/90 desion which facilitate rapid )

replacement of the hatch should the need arise. Similarly address other containment penetrations and potential bypass paths.

(f) The Diablo Canyon event and subsequently obtained information has shown operating procedures to he inadequate for lowered RCS inventory operation, k' hat plans exist for recommending inproved procedures and administrative controls to SP/90 l owners / operators so that this situation is eliminated in the SP/90 l (q) What equipment exists in the SP/90 that can he used to assure adequate enre cooling in the event o# a complete loss of PHP' (h) Evidence exists that certain Technical Specifications (TSsi nay not be optimum when consideration is given to operation during non-power conditions. For exanple, requirerents for PPP

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. 7 i suction valve interlocksi impact upon RHR reliability, PHP flow rate recuirements may overly restrict flow rate rance and increase the likelihood of loss of RHR due to vortexing, and TSs written on the basis of time (such as one may remove PHR

,fr om operation for an hour) perhaps are more reasonable when written on the basis of the state of the NSSS and/or of containment. Please address this topic with respect to the SP/90 design and provide recomendations for improvement, particularly with respect to the unique design aspects of the SP/90.

(ii Safety analysis reports (SARs) typically concentrate on power operation when consideration is given to many of the potential operational transients. The recent experience from the na iblo Canyon event indicated th-t further evaluation for plant operation at lower modes may be required. Hence, it may be prudent to sddress non-power operation in more depth than has been traditional. What plans exist, if any, with respect to this topic and the SP/90 program?

440.26? Our review has identified several areas in which unioue aspects of the SP/90 design do not appear to have been exploited to achieve the maximum reasonable safety. These include:

(a) The diesel start and loading time requirements of a few seconds do not appear recessary with the SP/00 FCCS design. The staff believes that lonaer start times will j enhance sefety by reduction of stress and wear to the diesels. Please discuss why such short loading time are l necessary.

8 (b) The fnur train primary side safeguards system was oricioally conceived, with one option, as having one diesel with each system. What are the quantitative difference in plant cost and safety when this is changed to the present two diesel design. Please also address the possibility that a four diesel approach may offer. a diverse diesel design possibility that has not been included in the two diesel concept.

(c) Please address the use of four diesels of diverse design and with relaxed start and load time reouirements with respect to the fraction of severe accidents associated with loss of all ac power.

(d) Early conceptual desion of the RCS included laroe dianeter cornections which could be used for rapid depressuri7ation. Vhy was this ccrability removed and what is the impt.ct of the change on accident mitioation and upon risk?

(ei The containment design ray allow coolino via a few nozzles which direct wa :er onto the outside contairment surface.

Was consideration given to such a system of pre-installed piping and noz71es with a connection which could be used, for example, by a fire truck as a source of pumped water' If not, what would be the cost and impact upon safety if such a system were installed?

l (f) Early versions of the SP/90 design included a non-safety related "pump-house" for each of the primary side safeguards systems. This appeared to offer many advantages over the present desian under severe accident i

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l conditions and 'or, control of release outside containment under.a wide range of conditions. What is the cost differential (details please) and impact upon both safety.

and~ releases between the early concept and the present design?

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