ML20217J491

Forwards RAI Re W AP600 Multiple SG Tube Rupture Analysis Rept

ML20217J491
Person / Time
Site:	05200003
Issue date:	08/07/1997
From:	Quay T NRC (Affiliation Not Assigned)
To:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 9708140298
Download: ML20217J491 (6)

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NUCLEAR REGULATORY COMMISSION WAaHINGTON, D.C. feceHoot s;

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Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Analysis Nuclear and Advanced Technology Division l Westinghouse Electric Corporation i P.O. Box 355 Pittsburgh, PA 15230

SUBJECT:

AP600 MULTIPLE STEAM GENERATOR TUBE RUPTURE ANALYSIS REQUEST FOR ADDITIONAL INFORMATION

Dear Mr. Liparulo:

NRC letter dated December 6,1996, provided a list of potential critical path issues for the AP600 design certification process. One of the issues (key issue #24) involves the need for Westinghouse to assess the AP600 design features which mitigate containment bypass due to steam generator tube rupture (SGTR) events. The basis of this issue is described in detail in SECY-93-M7,

" Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor Designs." In order for the staff to complete its review of the AP600 design for mitigating containment bypass during SGTR events, the staff has requested Westinghouse to provide an evaluation of steam generator tube ruptures involving up to 5 tubts, in addition, the staff needs to have a best estimate understanding of how the AP600 will respond during a multiple SGTR event assuming the affected steam gener: tor PORV fails to open, the condenser is unavailable, and the active high pressure charging system (CVS) continues to inject.

Westinghouse letter NSD-NRC-97-5035 dated March 24, 1997, provides a quantita-tive analysis of multiple SGTR events using its PRA thermal-hydraulic analysis code MAAP4. The Westinghouse analysis examines the effects of various system failure assumptions including those of interest to the staff. However, in order for the staff to complete its evaluation, additional information is needed on the thermal-hydraulic evaluation. Enclosed with this letter are requests for additional information (RAls) on the Westinghouse AP600 multiple steam generator tube rupture analysis report.

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Mr. Nicholas J. Liparulo August 7, 1997 l If you have any questions regarding this matter, you can contact Bill Huffman  !

at (301) 415-1141. i Sincerely, l

original signed by: i Theodore R. Quay Director Standardization Project Directorate '

Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003

Enclosure:

As stated  :

cc w/enclo See next page

)lSTRIBUTION:

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Mr. Nicholas J. Liparulo .

Docket No.52-003

. . Westinghouse Electric Corporation AP600 cc Mr. B. A, McIntyre .

Mr. Ronald Simard, Director  ;

Advanced Plant Safety & Licensing Advanced Reactor Programs Westinghouse Electric Corporation Nuclear Energy Institute '

Energy Systems Business L' nit 1776 Eye Street, N.W.

P.O. Box 355 Suite 300 Pittsburgh, PA 15230 Washington, DC 20006-3706 Ms. Cindy L. Haag Ms. Lynn Connor Advanced Plant Safety & Licensing Doc-Search Associates  !

Westinghouse Electric Corporation Post Office Box 34 Energy Systems Business Unit Cabin John, MD 20818 Box 355 Pittsburgh, PA 15230 Mr. James E. Quinn, Projects Manager LMR and SBWR Programs Mr. M. D. Beaumont GE Nuclear Energy Nuclear and Advanced Technology Division 175 Curtner Avenue, M/C 165 Westinghouse Electric Corporation San Jose, CA 95125 One Montrose Metro 11921 Rockville Pike Mr. Robert H. Buchholz Suite 350 GE Nuclear Energy Rockville, MD 20852 175 Curtner Avenue, MC-781 4 San Jose, CA 95125 i

Mr. Sterling Franks U.S. Department of Energy Barton Z. Cowan, Esq.

NE-50 Eckert Seamans Cherin & Mellott 19901 Germantown Road 600 Grant Street 42nd Floor Germantown, MD 20874 Pittsburgh, PA 15219 Mr. S. M. Modro Mr. Ed Rodwell, Manager Nuclear Systems Analysis Technologies PWR Design Certification Lockheed Idaho Technologies Company Electric Power Research Institute Post Office Box 1625 3412 Hillview Avenue Idaho falls, 10 83415 Palo Alto, CA 94303 Mr. Frank A. Ross Mr. Charles Thompson, Nuclear Engineer U.S. Department of Energy, NE-42 AP600 Certification Office of LWR Safety and Technology NE-50 19901 Germantown Road 19901 Germantown Road Germantown, MD 20874 Germantown, MD 20874 9

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RAls ON THE AP600 MSGTR ANALYSIS REPORT DATED 3/24/97 1

440.675 Page 3 of the SGTR analysis report states that since the RCS coolant never saturates or volds, two-phase modeling is not important. What is the basis to conclude that the RCS coolant never saturates or voids during a steam generator tube rupture event?

440.676 Page 3 of the SGTR analysis report states that the MAAP4 core makeu) tank behavior, as well as RCS thermodynamics and break flow modeling, are benchmar(ed in the MAAP4/NOTRUMP benchmarking exercise performed in support of the resolution of the passive system reliability issue. However, the MAAP4 calculations of other important systems and phenomena, such as the passive residual heat removal system, and the SG secondary side response (e.g., SG liquid level and pressure), during MSGTR are not mentioned. Provide discussions on all MAAP4 benchmarking exercises relevant to MSGTR, including references, the results, and justifications on why these benchmark exercises are sufficient to address appropriateness of MAAP4 calculation for the MSGTR/ containment bypass issue.

440.677 In sensitivity study cases SG5 max, SG5 min, and SG5p, the conclusion the "the overall results are not sensitive to break elevation," appears to be a repetition of the conclusion from Case SG5b, and therefore, inconsistent with the respective cases studied. Are the same conclusions drawn from the respective studies?

440.678 Page 4 of the report states that MAAP4 cannot model the heat removal through the steam condenser, whereas Table 1 (p.15) Column 8 indicates that majority of cases i

were run with "Sec PORV/ Condenser" available. This appears to be inconsistent because MAAP4 cannot model the condenser. Are these cases really analyzed with the steam condenser available, or what is the intent of the heading of "Sec PORV/ condenser" availability?

4 440.679 The MAAP4 analysis of the MSGTR events (page 4 of the report) assumes the HSlv closure at the time of turbine trip. In the AP600 design, the MSIV is automatically actuated on low steamline pressure, low-2 RCS temperature, Hi-1 containment pressure, or lor-2 SG narrow range level in the downcomer,

a. For a SGTR event, none of these signals will be reached to actuate the MSIV. Are there other signals to automatically close the MSIV?

b. If not, the faulted SG isolation must be done manually. What are the symptoms, signals, or parameters (and at what time?) that will alert the operator to isolate the MSIV?

c. Is the assumption of MSIV closure at turbine trip a conservative assumption from the point of view of containment bypass study? and why?

_d. If the MSly closure is not made at turbine trip, would the continued steam release through the turbine bypass valves (TBVs) to the condenser, or would the TBVs be closed by what signal at what , time?

Enclosure E

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440.680 The analyses of Cases SGl. SG2 thru SGS, SGSP, etc, did not assume the actuation of the chemical and volume control system (CVS), and the results appear to indicate that if the CVS injection is not actuated, the passive safety system with PRHR can prevent the SG pressure from reaching the safety valve set pressure whether or not the PORY opens (SGSP). However, this assumption is inconsistent with the AP600 design where the CVS injection automatically actuates on pressurizer low pressure and low level following a SGTR event. The analyses of these cases resulted in accelerated RCS depressurization, earlier primary-secondary pressure balance and termination of break flow than actually happened in the plant. Because of this non-conservative and non-realistic assumption, these cases appear to provide no useful i results to address the containment bypass issue. Are there any conclusions that can

be drawn from these cases to help address the MSGTR/ containment bypass issue?

440.681 The results of Case SG5cvs demonstrates that, when the CVS injection is modeled at the reach maximum the safety flowvalve rate of 170 gpm setpoint (at PORV if the time of is rupture)ional.,

operat the SGthe

.On pressure other hand, will not Case SG5stk indicates that, if the CVS injection is actuated at the time of tube rupture, and the PCRV fails to open, the safety valve will be lifted within 13 minutes of tube rupture. What would be the results if the CVS injection is actuated automati-cally n the low pressurizer pressure or level, and the PORV fails to open at

demand? Would the steam pressure reach the safety valve setpoint at & later time (how long?) that may provide sufficient time for operator action to cope with the problem?

i 440.682 Case SG5stk results shows that if safety valve open and fail to reseat, eventually ADS will actuate, and the break flow will stop or even reverse. However, the steam release continue to vent through the stuck open safety valve at a rate of 5 lbm/s.

a. Why does the steam release rate increase from about I to 5 lbm/s at about 25000 seconds when the gravity recirculation lines are open due to low IRWS7 level?

b. What is the basis to state (page 13 of the report) that the steam release through the stuck-open safety valve is " clean water?"

440.683 The analysis report did not appear to provide sufficient argument with respect to the measures to prevent the lifting of safety valves in the event of MSGTR, or the mitigating features in the event of a stuck-open safety valve, to address the MSGTR/ containment bypass issue,

s. For a more realistic calculation with the assum)tions that the CVS actuated at low pressurizer pressure or level, and the turbine )ypass steam dump system available until the MSIV closure, a loss of off-site power, or a loss of condenser vacuum, would the steam pressure reach the safety valve setpoints, and at what time? Would there be sufficient time for opero,or actions (such as diagnosis of a SGTR event, isolation of CVS injection, realignment of the CVS for pressurizer spray) to prevent safety valve lifting if PORY fails to open?

b. Since the AP600 PORV is a non-safety related system, what design features and other measures are there to ensure the PORY will automatically open on demand to prevent safety valve from opening in a MSGTR event? What is the probability of PORV failure on demand?

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c. It is stated (page 8 of the report) that the assumption of safety valve sticking- .

open is highly conservative since the safety valve will not relieve water with the -

4 AP600 steam generator overfilling protection design of automatic isolation of CVS and SFW at 79% SG narrow. range. W1at available data are there to support that safety valve will not stick open for steam release?

d. What measures are there in the AP600 design to mitigate the consequence of a stuck-open safety valve?

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