ML20062J425

Requests for Addl Info to Support Review Activities Re Sbwr Design Certification.Record Copy

ML20062J425
Person / Time
Site:	05200004
Issue date:	09/28/1993
From:	Malloy M Office of Nuclear Reactor Regulation
To:	Marriott P GENERAL ELECTRIC CO.
References
NUDOCS 9311050148
Download: ML20062J425 (6)

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

%."/ .+ September 28, 1993 Docket No.52-004 Mr. Patrick W. Harriott, Manager Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING THE SIMPLIFIED BOILING WATER REACTOR (SBWR) DESIGN The staff has determined that it needs additional information to support its independent assessment program activities related to the SBWR design certifi-cation review. Some additional information is needed-on the isolation condenser system performance (Q950.27-Q950.34).* Please respond to the enclosed questions within 60 days of the date of this letter.

You have previously requested that portions of the information submitted in the August 1992 application for design certification of the SBWR plant, as supplemented in February 1993, be exempt from mandatory public disclosure.

The staff has not completed its review of your request in accordance with the requirements of 10 CFR 2.790; therefore, that portion of the submitted information is being withheld from public' disclosure pending the staff's final' determination. The staff concludes that this RAI does not contain those portions of-the information for which you are seeking exemption. However, the staff will withhold this letter from public disclosure. for 30 ~ calendar days -

from the date of this letter to ' allow GE Nuclear Energy. 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 enclosure be withheld .from public-disclosure in accordance with 10 CFR 2.790, this letter will be placed in the NRC's Public Document Room.

This request for additional information affects nine or . fewer respondents, land L therefore, is not subject to review by the Office of Management and Budget.

under P.L.96-511.

NBC RLE CENTER COPY

• The numbers in parentheses designate the tracking numbers. assigned to the

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questions.010024

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. i Mr. Patrick W. Marriott September 28, 1993 If you have any questions regarding this matter, please contact me at (301) 504-1178 or Mr. Son Ninh at (301) 925-1125.

Sincerely, (Original signed by)

Melinda Malloy, Project Manager Standardization Project Directorate '

Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation ,

Enclosure:

RAI on the SBWR Design cc w/ enclosure:

See next page Distribution (w/ enclosure):

• Central. File PDST R/F -

• PDR MMalloy SNinh i DMcPherson, 8E2 RElliott, 8H2 Alevin, 8E21 Y-Schen, NLN344 TLee, NLN353 MFinkel stein,15B18 ATGody, Jr. ,17G21 JMonninger, 8H2 . MSnodderly, 8H2 MCaruso, 8El Shou, 7H15 SAli Syed, 7H15 Slee, 7H15 .

JHan, NLN353 CTinkler, NLN344 ANotafrancesco, NLN344 Distribution (w/o enclosure):

DCrutchfield/WTravers RBorchardt JNWil son PShea TMurley/FMiraglia,12G18 WRussell,12G18 AThadani/MVirgilio, 8E2 RJones, 8E21 RBarrett, 8H2

• BSheron/TKing, NLS007 FEltawila, NLN344 LShotkin, NLN353 JMoore, 15B18 ACRS (11) GSuh (2), 12E4 GBagchi, 7H15  ;

• To be held for'30 days I ., .

OFC LA)fDh PM:PDST _

PM:,PDSJ SRX,B,f SGS M ) YPDhT NAME PShek- 'MN1oktz SNriik R$ed Rk)t( JNWiYson DATE 09/b93 09/r)/93 09/.b/S3 09/O93 09/ky/93 09/ 28/93 DOCUMENT N  : BWR9317.MM

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d-Mr. Patrick W. Marriott Docket No.52-004 General Electric Company cc: Mr. Laurence S. Gifford GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852  :

Director, Criteria & Standards Division Office of Radiation Programs U.S. Environmental Protection Agency 401 M Street, S.W.

Washington, D.C. 20460 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C. 20585 Mr. John E. Leatherman SBWR Licensing Manager GE Nuclear Energy 175 Curtner Avenue, M/C 781 San Jose, California 95125 Mr. Frank A. Ross Program Manager, ALWR Office of LWR Safety & Technology U.S. Department of Energy NE-42 19901 Germantown Road Germantown, Maryland 20874 I

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REQUEST FOR ADDITIONAL INFORMATION (RAI) ON THE SIMPLIFIED BOILING WATER REACTOR (SBWR) DESIGN Isolation Condenser System Performance In the following questions, the term " capacity" is intended to denote either  :

the heat removal rate (Watts) or the steam condensation rate (kg/s), as  !

appropriate. The questions are structured to support a modeling approach by  :

which a basic heat exchanger capacity is determined each calculational time step depending upon the current isolation condenser system (ICS) or passive containment cooling system (PCCS) pool temperature. Inis basic capacity would ,

then be operated upon by correction factors to reflect the effect of flow  :

through the ICS vent line, the presence of noncondensable gas in the inlet stream, and the effect of varying pressure in the reactor vessel.

It is recognized that other approaches are feasible, and that the detailed information available at GE may be more appropriate to an alternate approach.

If this is the case, then the provision of the needed information in a different form, such as graphs or tables, would be acceptable.

950.27 From Sections 5.4.7 and 15.2** of the standard safety analysis -

report (SSAR), it is clear that no operator action is required to initiate ICS vent line operation when the reactor vessel is at pressure, cycling between 7.584 HPa [1100 psig (vent valve closure pressure)] and 8.618 HPa [1250 psig (set point for safety reli,ef i valve opening)] if significant noncondensable gas is present. '

However, after vessel depressurization, it is unclear to what extent the ICS condensers would effect the subsequent course of the accident. ,

It is recognized that an actuation signal for IC operation is Level 2 in the reactor pressure vessel (RPV). Automatic depressuri-zation system (ADS) operation begins on a Level I signal. Thus, the ICS should be in operation when the reactor coolant system  !

(continued on next page)

• This is the effect of the flow upon the heat transfer coefficient. The model will determine whether or not the heat exchanger tubes have become bound by, noncondensable gases and will calculate the clearing of noncondensable gases from the tubes when the vent valves are opened. i j

• • See items 15.2.3.2(5),15.2.4.2(2),15.2.5.2(4),15.2.6.2(4),and 15.2.7.2(3).  ;

• • • It is recognized that the ICS vent valves open at 7.653 MPa (1100 psig) and  ;

that at full capacity, ICS operation should prevent reactor vessel pretsure from reaching the safety relief valve set point. Furthermore, in the absence '

of noncondensable gases, ICS operation does not depend upon vent valve operation and the vessel should reach pressures well below 7.584 HPa. .

Enclosure  ;

)

y RAI ON THE SBWR DESIGN depressurizes, should conditions proceed that far. Provide information is requested concerning the effectiveness of the ICS after ADS actuation (with depressuriration valve opening) and vessel depressurization.

950.28 Provide information concerning the effect of reactor vessel pressure on ICS performance over the entire pressure range that the ICS is expected to operate, including severe accident conditions. If the isolation condensers are assumed to remain functional after vessel depressurization (either because the vent line valves fail open or because of operator intervention to open the vents), significant degradation in ICS performance would be expected. Therefore, extend the range of the performance versus pressure curve to include anticipated drywell pressure under severe accident conditions.

950.29 How does the ICS heat exchanger innerus' rface heat transfer. coeffi-cient increase when (choked) vent line flow is initiated? If it is expected that the ICS would remain functional after DPV operation because of manual opening of the vent valves, then also provide information concerning the increase in ICS performance with increas-ing vent line flow. Alternatively, how does the ICS performance increase with the reactor vessel-to-pressure suppression pool pressure differential when the vent valves are open at low pressures?

950.30 The basic ICS capacity is assumed to be affected by noncondensable gases in the inlet stream and changes in the pool temperature in a manner similar to that-indicated by the corresponding information.

previously provided for the PCCS. If the performance __ degradation factors for changes in th + noncondensable gas mole fraction and in the IC/PCCS pool temperatures are not the same as for the PCCS, provide values appropriate to the ICS.

950.31 Provide information concerning any special sensitivities of the ICS performance to hydrogen. (The paper entitled " Local Heat Transfer Coefficients _ for Fo'rced-Convection Condensation of Steam-in a Vertical Tube in the Presence of a Noncondensable Gas" by Mansoor Siddique, Michael W. Golay, and Mujid S. Kazimi, Nuclear Technology, _

Volume 102, June 1993, has been reviewed.)

950.32 With respect to the ICS bottom vent valves, SSAR Section 5.4.6.2.3 states "...when the RPV gauge pressure decreases below 7.584 MPa (1100 psi) .(reset value) and after a time delay to avoid too many cycles,_ these valves close." What is the length of this time delay?

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.i RAI ON THE SBWR DESIGN 950.33 SSAR Section 5.4.6.2.2 indicates "A catalytic converter is provided to recombine noncondensable gases (hydrogen and orf]en) under normal plant operation (ICS standby condition)....The catalytic converter' is located on the steam distributor cover at the top end of the steam supply line to the isolation condenser." What is the capacity of this converter and would it be expected to have any significant effect under severe accident conditions?

950.34 SSAR Section 5.4.6.2.2 states "...a vent [ purge line as described in SSAR Section 17.3.11] is provided that takes a.small stream of gas from the top of the isolation condenser and dumps it downstream of ,

the RPV on the main steamline upstream of the MSIVs (main steamline isolation valves)." Will the purge line be opened during operation' .

of the ICS? If so, could the purge line provide a mechanism for noncondensable gas removal or buildup during a severe accident sequence?

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