ML20059D200

Forwards Request for Addl Info Re Sbwr Design Certification & Requests That Written Response to Encl Questions Be Provided within 60 Days of Dtd Ltr

ML20059D200
Person / Time
Site:	05200004
Issue date:	11/29/1993
From:	Malloy M Office of Nuclear Reactor Regulation
To:	Marriott P GENERAL ELECTRIC CO.
References
NUDOCS 9401070060
Download: ML20059D200 (10)

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j4Aj-[ NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 @ 01 g- g

, , November 29, 1993

, Docket No.52-004

-Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy 17S Curtner Avenue San Jose, California 95125

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING THE SIMPLIFIED B0ILING WATER REACTOR (SBWR) DESIGN The staff has determined that it raeds additional information to support its review activities related to the SEVR design certification. Some additional-information on the discussions of r.idiological impacts contained in Chap-ters 2, 6, and 15,of SBWR standard safety analysis report (SSAR) is needed (Q470.1-Q470.37). Please provide a written response to the enclosed ques-tions 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 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 RAI affects nine or fewer respondents, and therefore, is not subject to review by the Office of Management and Budget under P.L.96-511.

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940107 060 931129 $~ t t PDR. ADOCK0520g4'-

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Mr. Patrick W. Marriott November 29, 1993 If you have any questions regarding this matter, please contact me at (301) 504-1178 or Mr. Son Ninh at (301) 504-1125.

Sincerely, (Original signed by)

Melinda Halloy, 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:

• Central; File- PDST R/F MMalloy _

• PDR FHasselberg SNinh LCunningham, 10D4 TEssig, 10D4 JLee, 10D4

-JWigginton, 10D4 CHinson, 1004 RBarrett, 8H7 RJones, 8E23 GBagchi, 7H15 MFinkelstein, 15B18 GSuh-(2),12E4 WDean, 17G21 JMoore, 15B18 DCrutchfield/WTravers RBorchardt JNWilson-PShea TMurley/FMiraglia,12G18 WRussell,12G18 BSher>n/TKing, NLS007 ACRS (11) {

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GE Nuclear Energy: .

12300 Twinbrook Parkway Suite:315 Rockville, Maryland 20852 1 Director,-Criteria &' Standards Division

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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 Mr. Victor G. Snell, Director Safety and Licensing AECL Technologies 9210 Corporate-Boulevard Suite 410 Rockville, Maryland 20850 f

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REQUEST FOR ADDITIONAL INFORMATION (RAI) ON THE SIMPLIFIED BOILING WATER REACTOR (SBWR) DESIGN ,

Meteorology (Chapter 2) 470.1 Section 2.0, " Site Characteristics," of the SBWR standard safety analysis report (SSAR) defines the envelope of site-related para-meters which the SBWR standard plant is designed to accomodate.

Provide a table showing the envelope of SBWR standard plant site design parameters, including but not limited to the (1) tornado design basis and (2) bounding atmospheric relative concentrations (X/Q) for the exclusion area boundary and for the low population zone. The bounding X/Q values should provide assurance that (1) the radiological effluent release limits associated with normal reactor operation (specified in 10 CFR Part 50, Appendix I) will be met, and (2) the radiological consequences of a range of postulated acci-dents, up to and including the limiting design-basis accident (DBA) ,

considered, will be acceptable for an individual located at the -

nearest boundary of the exclusion area for a specified time.

470.2 Local meteorology in SSAR Section 2.3.2 and the on-site meteorologi-cal measurements program in Section 2.3.3 should 'e designated as '

combined operating license (COL) applicant action items.

Control Room Habitability (Chapter 6) 470.3 The SBWR main control room is located on the ground level and within the reactor building, adjacent to the service and turbine buildings.

During and following a loss-of-coolant accident (LOCA), which .is the controlling DBA for the radiological consequence to the control room operators, the radiation exposures to the operators will consist of contributions from airborne fission-products entered into the control room and direct gama radiation from the surrounding build-ings and process equipment. For determination of the gama radia-tion dose to the control room operators, state the major gama radiation sources, including the main steam lines, and the shielding provided (floors and control room wall thicknesses). ,

470.4 SSAR Section 6.4.2 states that the emergency breathing air system (EBAS) is automatically initiated upon automatic isolation of the sealed emergency operation area (SE0A) by a high radiation signal from the normal control room ventilation system. Show this feature in SSAR Figures 21.6.4-1 and 21.9.4-1. ,

470.5 In the analysis of the control room operator radiation doses follow-ing a LOCA, full credit is taken for fission-product removal by the control room envelope heating, ventilation, and air conditioning (CREHVAC) system after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> into a LOCA. This is unacceptable since the CREHVAC system is neither classified nor qualified as an engineered safety feature (ESF) system and needs to be addressed.

(The SBWR CREHVAC system is a single train non-safety-grade system.)

Enclosure

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410.6 'SSAR Table 15.6-13 shows the control room atmospheric relative i concentrations (X/Q) used and the resulting control room operator ,

doses. List the major parameters, assumptions, and methodologies used in determining the X/Q values and control room operator doses.

470.7 The staff's evaluation of Electric Power Research Institute's (EPRI's) Light-Water Reactor Utility Requirements Document for passive plant designs (EPRI's Passive Requirements Document) accepts the use of a passive, safety-grade control room pressurization -

system that would use bottled air to keep the operators' doses within the limit of General Design Criterion (GDC).19_ and Standard Review Plan (SRP) Section 6.4 (Revision 2) for the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of.

the DBA. The evaluation addresses use of safety-grade connections for the pressurization system to allow the use of offsite portable air supplies after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to minimize the operators' doses for the entire duration of a DBA (30 days). In either case (use'of bottled air for the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> only or the use of offsite portable air after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the entire duration of a DBA), GE should specify in the SBWR Tier 1 Design Certification Document; i.e., in the inspections, tests, analyses, and acceptance criteria (ITAAC); that the feasibility and capability J the safety-grade bottled air supply system to maintain positivc pressure in the control room envelope should be demonstrated.

470.8 SSAR Sections 6.2.3 and 6.5.3.3 describe the SBWR safety envelope design and SSAR Table 15.6.9 lists the safety envelope leakage rate and its air mixing efficiency as 25 percent per day and 50 percent, respectively. The leakage rate and air mixing efficiency are used in mitigation of offsite and control room operator radiological consequence assessments. Section 6.2.3 further states that the safety envelope is designed to be capable of periodic testing to assure that performance requirements are met.

Provide (1) the safety envelope free air volume, (2) detailed technical justifications for the 50-percent air mixing efficiency assumed, and (3) the leakage testing criteria, frequency, and proce- '

dures. The staff will require the SBWR Tier 1 Design Certification Document (ITAAC) to specify the safety envelope leakage rate and air mixing efficiency and the COL applicant technical specifications to specify the periodic integrated leakage rate testing.

470.9 SSAR Section 6.5.3.3 describes fission-product holdup, as well as the plate out mechanism, in the safety envelope. However, the staff noticed that holdup and mixing credits (for decay) are claimed, but that no credit is taken for fission-product plate out. Clarify this apparent discrepancy.

470.10 SSAR Section 6.2.2 describes the passive containment cooling system .

(PCCS). The PCCS removes the core decay heat rejected-to the containment after a LOCA. Should the PCCS heat exchanger tubes fail, the PCCS will provide a potential bypass pathway for the SBWR containment, releasing radioactive fission-products from the con-tainment atmosphere to the reactor building through the passive a

containment cooling pool water. Provide the radiological conse- ,

quence assessments, complete with the major assumptions and parame-ters used, for the PCCS heat exchanger tube failure. .

Radiological Consequences of DBAs (Chapter 15) 470.11 SSAR Section 15.6.2 describes failure of a small line carrying primary coolant outside containment, and SSAR Table 15.6-1 lists the assumptions and parameters used in the radiological consequence assessment for this failure. Provide the technical bases, complete with applicable references, for the following assumptions:

(1) the period of 10 minutes for the operator to detect the event (2) the period of 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for the operator to scram the reactor to reduce reactor pressure (3) 13,000 kg of reactor primary coolant released into the reactor building (4) 5,000 kg (out of 13,000 kg lost) reactor primary coolant being flashed to steam (5) magnitudes of iodine spiking (6) iodine plate out fraction of 50 percent (7) reactor building leak rate of 200 percent per hour 470.12 In the response to Part (5) of Question 470.16 above regarding .

iodine spiking, state the chemical forms of iodine assumed to spike  !

and state the reasons for not considering spiking of other nuclides such as cesium.

470.13 In the titles of SSAR Tables 15.6-1 through 15.6-3, change "Instru-ment" to "Small," so that the titles will read "Small Line Break Accident Parameters."

470.14 SSAR Section 15.6.4 describes a main steam pipe break accident outside containment and SSAR Table 15.6-5 lists the major assump-tions and parameters used in the radiological consequence assessment of this accident. Provide the technical bases, complete with applicable references, for the following parameters used:

(1) air exchange rate of 6000 per day in the steam tunnel (provide the free air volume of the steam tunnel and break location) ,

(2) 12,000 kg of steam mass released (3) 2,400 kg of water mass released (4) iodine concentration in the reactor coolant based on offgas release rates of 0.2 Ci/sec and 0.05 Ci/sec.

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470.16 For each release phase (gap release, early in-vessel, ex-vessel, and late in-vessel) and for each nuclide listed in draft NUREG-1465, '

provide a comparison table listing the SBWR, EPRI Passive Require-ments Document, and draft NUREG-1465 values for fission-product release fractions into the containment following a DBA.

470.17 The primary containment leakage rate (0.475 percent) and bypass  :

leakage rate (0.025 percent) provided in SSAR Section 15.6.5.5 should be expressed as those values per weight percents per day rather than in volume percents per day.

470.18 SSAR Section 15.6.5.5 states fission-product release timing and durations for each release phase. List all SBWR accident' sequences which are considered to significantly impact the source term and identify the controlling accident scenario and sequence for fuel rod failure (gap release) and fuel melt (early in-vessel release). (The accident scenarios should consider but not be limited to break and non-break of the reactor coolant system, small-break LOCA, large-break LOCA, and the leak before break.)

470.19 SSAR Section 15.6.5.5 describes and SSAR Table 15.6-9 lists the amount of organic iodide source term as 0.15 percent of the core iodine inventory. This amount is inconsistent with a value devel-oped by the staff and currently undergoing management review. In draft NUREG-1465, the Nuclear Regulatory Commission (NRC) did not evaluate the formation of organic iodide in the containment follow-ing a DBA. The staff realizes, however, that organic iodide can be produced by the reaction of fission product iodine with organic materials present in the-containment. The NRC estimates that no more than 5 percent of the airborne elemental iodine will be con-verted into organic species. This amount of organic iodide would correspond to about 0.25 percent of the core iodine inventory (i.e.,

5 percent of 5 percent is 0.25 percent). Final NUREG-1465 will address this issue. In the interim, GE should reassess the radio-logical consequences and resubmit for staff review the resulting offsite and control room operator doses using 0.25 percent organic iodide. If it desires, GE may retain the consequence analysis based on the 0.15 percent organic iodide for reference purposes, in addition to the analysis requested herein.

470.20 SSAR Section 15.6.5.5 states that the SBWR design is capable of injecting buffering agents into the pool water through various '

systems to maintain the pH of the pool water above 7.0. Describe in detail the (1) type and amounts of chemicals to be used, (2) provi-sions designed for chemical injection, (3) volume of pool water in the containment, (4) provisions designed for pool water mixing, (5) pool water sampling and analysis provisions, (6) pH monitoring, and (7) chemical storage facilities and their locations in build-  ;

ings.

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p 470.21 Provide calculated pH values of pool water _ in the primary contain-ment following the controlling accident sequence selected in Ques-tion 470.18 above as a function of time over the entire duration of the accident (30 days). .

470.22 The most important acids in the primary containment following a DBA are nitric acid, produced by irradiation of water and air, and hydrochloric acid, produced by irradiation or heating of electrical cable insulation. State the chemical properties and estimated amounts of electrical cable insulator used in the SBWR primary containment. State the calculated amounts of hydrochloric acid produced by radiolysis and pyrolysis of electric cable insulators and resulting pool water pH.

470.23 SSAR Section 15.6.5.5 describes and SSAR Table 15.6-9 lists the primary containment aerosol deposition rate of 0.6 per hour. _ It states that this rate is based on prior analyses of boiling water reactors (BWRs) under similar circumstances. Provide the prior analyses.

470.24 The containment aerosol removal rates are plant design specific and will vary depending on, but not be limited to, the containment geometry, containment size, surface area, steam quality, and containment cooling mechanisms. Provide assumptions and parameters used in the determination of the SBWR ( ntainment aerosol removal rates and the computer codes used to calculate the aerosol removal rates.

470.25 Provide the primary containment aerosol removal rates as a function of time following the controlling DBA accident sequence considered in Question 470.18 above, over the entire duration of the accident (30 days).

470.26 SSAR Section 15.6.5.5 refers to a BWR Owners Group report (NEDC-31858P, February 1991) for fission-product holdup, deposition, and resuspension rates used for fission-product mitigation in the ,

SBWR main steam lines and condenser. The report is based on the Hope Creek Nuclear Station design using the TID-14844 source term.

For each iodine species for the entire duration of a DBA (30 days),

provide specific values used for fission-product (iodine) mitigation in the SBWR design for the following:

(1) iodine deposition rates (2) iodine fixation rates (3) iod'ene reuspension rates (4) main steam line temperatures (5) total integrated iodine release to the main steam lines (6) total integrated iodine release to the condenser -

(7) total integrated iodine release from the condenser 470.27 SSAR Se : tion 15.6.5.5 (page 15.6-15, last paragraph) states that specific values used and the results of the main steam line leakage t

• analysis are given in SSAR Table 15.6-8. They are not included in this table. Provide these values and the results.

'_470.28 Among other things, SSAR Table 15.6-9 lists the SBWR condenser parameters used for fission-product mitigation. Provide the techni-cal bases for the following parameters: ,

(1) condenser leakage rate of 11 percent per day (2) iodine removal factor of 99.3 percent 470.29 State the operability status of the main steam.drair valves from the main control room following a DBA. To facilitate the main steam leakage pathway, the drain valves should be able to be manually opened from the main control room following a DBA by means of a safety-related power source.

470.30 Among other things, SSAR Table 15.6-9 lists reactor building leakage parameters. Provide the technical bases for the following '

parameters:

(i) reactor building mixing efficiency of 50 percent (2) reactor building leakage rate of 3600 percent per day  ;

470.31 SSAR Section 10.6.6 describes the feedwater line break accident (outside containment) and SSAR Table 15.6-17 lists the major para-meters used in its radiological consequence assessment. Provide the technical basis for the following parameters used:

(1) 320,000 kg of condensate released from the break (2) 10,000 kg of condensate flashed to steam from the break (3) two percent carryover factor of iodines in the condensate to  ;

flashed steam 470.32 Provide the postulated primary coolant leakages from ESF components (valve stems and pump seals) that are located outside of the primary t containment to the. secondary containment (safety envelope) and to the reactor building following a DBA.

470.33 Provide a radiological consequence assessment for the SBWR reactor ,

water cleanup pipe break accident (outside the primary containment).  ;

For this assessment, GE may assume that the break is instantaneous, circumferential, and on the downstream side of the outmost contain-ment isolation valve, but on the upstream side of the reactor water cleanup demineralizers.

470.34 Provide a radiological consequence assessment for the offgas system failure accident. GE stated in SSAR Section 15.7.1 that the NRC has deleted this SRP section and, therefore, did not provide this analysis. The NRC has transferred (but not deleted) SRP Sec-tion 15.7.1, " Waste Gas System Failure'" to SRP Section 11.3, 4

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• " Gaseous Waste Management System Failure," as Branch Technical Position 11-5, " Postulated Radioactive Releases due to a Waste Gas System Leak or Failure," so the assessment must still be provided.

In its assessment, GE should assume an inadvartent bypass of all charcoal beds due to an operator error or system computer error, in addition to the failure of the automated air-operated downstream isolation valve. GE should also assume that'during this accident, the plant is operating at :nd continues to operate at the maximum permissible offgas release rate (measured at offgas recombiner effluent) as specified in the SBWR technical specifications.

470.35 SSAR Section 15.7.4 describes fuel handling accidents, and SSAR Table 15.7-4 lists the major parameters assumed in the radiological consequence assessment for the fuel handling accidents. Provide.the.

technical bases for the following parameters used:

(1) 62 hours7.175926e-4 days <br />0.0172 hours <br />1.025132e-4 weeks <br />2.3591e-5 months <br /> of minimum decay time prior to fuel movement (2) reactor building release rate of 500 percent per hour Also state the fuel pool decontamination factors assumed for the fission-products other than noble gases and iodines.

470.36 SSAR Section 15.7.5 describes the spent fuel cask drop accident, and SSAR Table 15.7-8 lists the major parameters assumed in the radio-logical consequence assessment for the accident. Provide the technical bases for the following parameters used:

(1) cask drop distance of 24 meters (2) 120 days of minimum decay time prior to spent fuel cask movement (3) 2.5 air exchange per hour in the reactor building (4) the maximum capacity (number of spent fuel rods) of spent fuel cask (5) fuel pool decontamination factors assumed for the airborne fission-products 470.37 Provide the radiological consequence assessment for a heavy load drop accident dropping a heavy object onto the fuel in the reactor vessel during fueling and refueling operations. In the assessment, GE should evaluate (1) whether the reactor building crane system design meets single-failure-proof criteria, and (2) the provisions  ;

provided for prevention of load unbalancing which could potentially defeat the single-failure-proof criteria. State whether the SBWR reactor building crane follows the guidance provided-in NUREG-0554,

" Single-Failure-Proof Cranes for Nuclear Power Plants," for design, fabrication, installation, and testing.

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