ML20058A274
| ML20058A274 | |
| Person / Time | |
|---|---|
| Site: | 05200004 |
| Issue date: | 11/18/1993 |
| From: | Liza Cunningham Office of Nuclear Reactor Regulation |
| To: | Borchardt R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9312010024 | |
| Download: ML20058A274 (8) | |
Text
NOV 181993 MEMORANDUM FOR:
Richard W. Borchardt, Director Standardization Project Directorate Associate Director for Advanced Reactors and License Renewal, NRR FROM:
LeMoine J. Cunningham, Chief Radiation Protection Branch Division of Radiation Safety and Safeguards, NRR
SUBJECT:
REVIEW OF SBWR STANDARD SAFETY ANALYSIS REPORT (SSAR) REQUEST OF ADDITIONAL INFORMATION The Radiation Measurement and Health Effects Section of the Radiation Protection Branch (PRPB) has completed its review of the following sections of the SBWR/SSAR:
Chapter 2 Section 2.3,
" Meteorology" Chapter 6 Section 6.4,
" Control Room Habitability" Chapter 15 Section 15.4, " Radiological Consequences of DBAs" On the basis of this review, we are enclosing the attached requests for additional information (RAls). The RAls from the Facilities Radiation Protection Section of the PRPB, based or 8ts partial review, have been transmitted to you earlier with my transmittal memorandum dated October 29, 1993.
This review was performed by Jay Lee (504-1080).
Original signed by LeMc:ne J. Cunningham LeMoine J. Cunningham, Chief Radiation Protection Branch, DRSS, NRR
Enclosure:
As stated Distribution:
I Central - File,. P1 37' PRPB R/F PRPB S/F, SBWR PDR, LL6 F. Congel R. Erickson L. Cunningham P. McKee T. Essig J. Wigginton C. Hin',un J. Lee R. Anderson, TTC M. Malloy, llH3 0FC PRPy g S RPB:DRSgSC RPB:g:BC gE$L TESSIGhLgNNENbAM NAME
[DATE$1/17 /93 11/d/93 II/n%/93 0FFICIAL RECORD COPY Document Name:RAI E t U M.Pa B m 'ir 9312010024 93111s PDR ADOCK 05200004 A
POR I
ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION RADIATION MEASUREMENT AND HEALTH EFFECT SECTION RADIATION PROTECTION BRANCH GENERAL ELECTRIC SIMPLIFIED B0ILING WATER REACTOR Chapter 2 470.1 Section 2.0, " Site Cl aracteristics" defines the envelope of site related parameters wtich the SBWR Standard Plant is designed to i
accommodate.
Provide a table showing envelope of SBWR standard plant site design parameters including, but not limited to, (1) tornado design basis, and (2) bounding atmospheric relative concentrations (X/Q) for the exclusion area boundary and for the low population zone. Tne 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 accidents, up to and including the limiting design basis accident 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 Section 2.3.2 and on-site meteorological measurements program in Section 2.3.3 should be designated as COL items.
Chapter 6 i
470.3 The SBWR main control room is located on the ground level and within the reactor building adjtcent to service and turbine buildings.
During and following a 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 gamma radiation from the surrounding buildings and process equipment.
For determination of. gamma radiation dose to the control room operators, state the major gamma radiation sources, including the main steam lines, and shielding provided (floors and control room wall thickness).
470.4 Sections 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 high radiation signal from the normal control room ventilation system. Show this feature in P&ID-Figures 21.6.4-1, and 21.9.4-1.
470.5 In your analysis of the control room operator adiation doses following a LOCA, you have taken' full credit for fission-product.
removal by the control room envelope HVAC ' system (CREHVAC) after _72 hours into a LOCA. Th's is unacceptable to the staff since the CREHVAC system is neither classified nor qualified as an engineered safety features system. The SBWR/CREHVAC system is a single train-non-safety-grade system.
2 470.6 Table 15.6-13 shows the control room atmospheric relative concentrations (X/Q) used and resulting control room operator doses.
List the major parameters and assumptions, and methodologies used in determining the X/Q values and control room operator doses.
470.7 In review of the EPRI passive design requirements document, the NRC has accepted the use of a passive, safety-grade control room pressurization system that would use bottled air to keep operator's dose within the limit of GDC 19 and 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, and 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 operator 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 (ITAAC) that the feasibility and capability of the safety-grade bottled air supply system to maintain positive pressure in the control room envelope should be demonstrated.
470.8 Sections 6.2.3 and 6.5.3.3 describes the SBWR safety envelope design and 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 50 percent air mixing efficiency assumed, and (3) leakage testing criteria, frequency, and procedures.
The staff will require the safety envelope leakage rate and air mixing efficiency to be specified in the SBWR Tier 1 Design Certification Document (ITAAC) and periodic integrated leakage rate testing to be specified in the COL technical specifications.
470.9 Section 6.5.3.3 describes fission-product holdup, as well as plate out mechanism in the safety envelope. However, the staff noticed that GE claimed only holdup and mixing credits (for decay), and not for fission-product plate out. Clarify this discrepancy.
470.10 Section 6.2.2 describes 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 containment-atmosphere to the reactor building through the passive containment cooling pool water.
3
.4 Provide the radiological consequence assessments, complete with the major assumptions and parameters used, for the PCCS heat exchanger tube failure.
Chapter 15 470.11 Section 15.6.2 describes failure of small line carrying primary coolant outside containment and Table 15.6-1 lists the major assumptions and parameters used in 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 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 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 responding to iodine spiking in Question 470.16 (5) above, 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 Make editorial changes to the titles of Tables 15.6-1 through 15.6-3 to read Small Line Break Accident Parameters (rather than Instrument Line break).
470.14 Section 15.6.4 describes main steam pipe break accident outside containment and Table 15.6-5 lists the major assumptions and parameters used in radiological consequence assessment of this accident.
Provide technical bases (complete with applicable references) for the following parameters used:
(1) air exchange rate of 6000 per day. in steam tunnel (provide free air volume of 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. 470.15 Section 15.6.5.5 states that reactor accident source terms used are consistent with those specified in EPRI Passive Requirements Document except where noted.
Itemize and list in a table, those exceptions taken.
4 470.16 Provide a comparison table listing GE(SBWR), EPRI Passive Requirement Document, and Draft NUREG-1465 values.for fission-product release fractions into the containment following a DBA for each nuclide listed in draft NUREG-1465, and for each release phases (gap release, early in-vessel, ex-vessel, and late in-vessel).
470.17 The primary containment leakage rate (0.475 percent) and bypass leakage rate (0.025 percent) provided in Section 15.6.5.5 should be expressed as those values per weiaht percents per day rather than volume percents per day.
470.18 Section 15.6.5.5 states fission-product release timing and durations for each release phases.
List all SBWR accident sequences which are considered to be 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 limited to, break-and non-break of the reactor coolant system, small-break LOCA, large-break LOCA, and the leak before break (LBB).
470.19 Section 15.6.5.5 describes and Table 15.6-9 lists' amount of organic iodide source term as 0.15 percent of _ the core iodine inventory.
l This value is unacceptable to the NRC. In draft NUREG-1465, the'NRC l
did not evaluate the formation of organic iodide in the containment following 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 converted 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. Meanwhile, GE should reassess the radiological consequences and resubmit the resulting offsite and control room operator doses using 0.25 organic iodide to NRC for its_ review.
470.20 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 pH of the pool. water above 7.0.
Describe in detail (1) the j
type and amounts of chemicals to be used, (2) the provisions 1
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, (5) pH; monitoring,- and-(6)-
chemical-storage facilities and_ their locations inL buildings.
470.21 Provide calculated pH values of pool, water in the primary containment following the controlling' accident sequence selected in Question 470.23 above as a function of time over.the entire duration.
of the accident (30 days).
1 V
3
5 1
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 chemical properties and estimated amounts of electrical cable insulator used in the SBWR primary containment.
State calculated amounts of hydrochloric acid produced by radiolysis and pyrolysis of electric cable insulators and resulting pool water pH.
470.23 Section 15.6.5.5 describes and Table 15.6-9 lists the primary containment aerosol deposition rate of 0.6 per hour, and it states that this rate is based on prior analyses of 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 limited to, the containment geometry, containment size, surface area, steam quality, and containment cooling mechanisms.
Provide assumptions and parameters used in determination of the SBWR containment aerosol removal rates, and 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 470.23 above, over the entire duration of the accident (30 days).
470.26 Section 15.6.5.5 refers fission-product holdup, deposition, and resuspension rates used for fission-product mitigation in the SBWR main steam lines and condenser to the BWROG report (NEDC-3185P) dated February 1991. The report is based on Hope Creek Nuclear Station design using TID-14844 source term.
Provide the following specific values used for fission-product (iodine) mitigation in the SBWR design for each iodine species for the entire duration of a DBA-(30 days):
(1) iodine deposition rates (2) iodine fixation rates (3) iodine resuspension rates (4) Main steamline temperatures (5) total integrated iodine release to main steamlines (6) total integrated iodine release to condenser (7) total integrated iodine release from condenser 470.27 Section 15.6.5.5 (SBWR/SSAR page 15.6-15, last paragraph) states that specific values used and the results of the main steamline leakage analysis are given in Table 15.6-8.
They are not found in the table. Provide these values and results.
i
a 6
1 470.28 Table 15.6-9 lists, among other things, the parameters for the SBWR condenser used for fission-product mitigation.
Provide the technical 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 operability of the main steam drain valves from the main control room following a DBA. The drain valves should be able to open from the main control room manually following a DBA via a safety-related power source to facilitate main steam leakage pathway.
470.30 Table 15.6-9 list, among other things, reactor building leakage parameters.
Provide technical bases for the following parameters:
(1) reactor building mixing efficiency of 50 percent (2) reactor building leakage rate of 3600 percent per day 470.31 Section 15.6.6 describes feedwater line break accident (outside containment) and Table 15.6-17 lists the major parameters used in its radiological consequence assessment.
Provide 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 containment to the secondary containment (safety envelope) and to the reactor building following a DBA.
470.33 Provide radiological consequence assessment for the SBWR reactor water cleanup pipe break accident (outside the primary containment).
GE may assume for this assessment, the break to be instantaneous, circumferential, and at downstream side of the outmost containment isolation valve but the upstream side of the reactor water cleanup demineralizes.
470.34 Provide radiological consequence assessment for the offgas system failure accident. The NRC has transferred (but not deleted) Section 15.7.1, " Waste Gas System. Failure," of the Standard Review Plan (SRP) to Section 11.3, " Gaseous Waste Management System Failure," of the SRP as Branch Technical Position 11-5, " Postulated Radioactive Releases due to a Waste Gas System Leak or Failure." GE stated in the SBWR/SSAR Section 15.7.1 that the NRC has deleted this'section and therefore, GE did not provide this analysis. This is incorrect.
o.
7 i
In its assessment, GE should assume an inadvertent 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 and continue to operate at the maximum permissible offgas release rate (measured at offgas recombiner effluent) as specified in the SBWR technical specifications.
470.35 Section 15.7.4 describes fuel handling accidents and Table 15.7-4 lists the major parameters assumed in the radiological consequence assessment for the fuel handling accidents.
Provide 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 State fuel pool decontamination factors assumed for the fission-products other than noble gases and iodines.
470.36 Section 15.7.5 describes spent fuel cask drop accident and Table 15.7-8 lists the major parameters assumed in the radiological consequence assessment for the accident.
Provide 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 fuelitig and refueling operations.
In its assessment, evaluate (1) the reactor building crane system design if it 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 if the SBWR reactor building crane follows the guidances provided in NUREG-0554, " Single-Failure-Proof Cranes for Nuclear Power Plants" for design, fabrication, installation, and testing.