ML20058L223
| ML20058L223 | |
| Person / Time | |
|---|---|
| Site: | 05200004 |
| Issue date: | 11/09/1993 |
| From: | Malloy M Office of Nuclear Reactor Regulation |
| To: | Marriott P GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 9312160190 | |
| Download: ML20058L223 (9) | |
Text
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i UNITED STATES k
NUCLEAR REGULATORY COMMISSION O
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wasencrow o.c.rosss
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e November 9, 1993 i,
Docket No.52-004 l
Mr. Patrick W. Marriott, Manager i
Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125
Dear. Mr. Marriott:
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SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING THE SIMPLIFIED l
BOILING WATER REACTOR (SBWR) DESIGN' j
The staff has determined that it needs additional information to support its i
review activities related to the SBWR design certification.. Some additional information on the radiation protection information contained.in Chapte SBWR standard safety analysis report (SSAR) is needed (Q471.1-Q471.31).r 12 of 1
Please provide a written response to the enclosed question within 90 days of the date of this letter.
The enclosed questions represent a fairly comprehensive. review of SSAR Chap-t' ter 12, with the exception of the zone drawings referenced in SSAR Sec-tion 12.3.
The review of the radiation-protection-related information in i
SSAR Chapters 9 and 11 is still under way, and the staff expects to forward I
questions on this material in the near future.
1 You have previously requested that portions of the information submitted in the August 1992, application for design certification of.the SBWR plant, as i
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 i
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:
i staff will withhold this letter from public disclosure for 30 calendar days.
l 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-j the information in the enclosure be withheld from public disclosure in accordance with 10 CFR 2.790, this-lc.tter will be placed in the NRC's Public l
Document Room.
9312160190 931109 [P I
DR ADOCK 0520000 PDRf
- The numbers in parentheses designate the tracking numbers assigned to the questions.
1 NBC HLE CENTE N,
130024
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Mr. Patrick W. Marriott November 9, 1993 i
'I 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.
i If you have any questions regarding this matter, please contact me at (301) 504-1178 or Mr. Son Ninh at (301) 504-1125.
i Sincerely,
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l (Original signed by)
[
Melinda Malloy, Project Manager.
Standardization Project Directorate Associate Directorate for Advanced Reactors'-
l l
and License Renewal.
Office of Nuclear Reactor Regulation' t
tnclosure:
i RAI on the SBWR Design i
cc w/ enclosure:
i See next page j
l Distribution (w/ enclosure)-
- Central File PDST R/F
- PDR MMalloy SNinh i
LCunningham, 10D4 JWigginton, 1004 CHinson,'10D4
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TEssig, 10D4 MFinkelstein, 15B18 GSuh (2), 12E4
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- To be held for 30 days l
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JNWWson RBodhaht NAME PShea ()r3 tz DATE 11/h/93 11/3/93 11/4/93 11/$/93 ~
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.I DOCUMENT NAME: SBWR9401.MM
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i Mr. Patrick W. Marriott Docket No.52-004 l
General Electric Company l
cc:
Mr. Laurence S. Gifford GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 Rockville, Maryland -20852 f
f Director, Criteria & Standards Division Office of Radiation Programs U.S. Environmental Protection Agency 401 M Street, S.W.
Washington, D.C.
20460 1
Mr. Sterling Franks U.S. Department of Energy i
NE-42 Washington, D.C.
20585 4
Mr. John E. Leatherman SBWR Licensing Manager i
GE Nuclear Energy 175 Curtner Avenue, M/C 781 San Jose, California 95125 i
Mr. Frank A. Ross Program Manager, ALWR l
Office of LWR Safety & Technology U.S. Department of Energy NE-42 19901 Germantown Road Germantowm Maryland 20874 4
Mr. Victor G. Snell, Director Safety and Licensing AECL Technologies 9210 Corporate Boulevard Suite 410 Rockville, Maryland 20850 l
i l
l
I REQUEST FOR ADDITIONAL INFORMATION (RAI) ON THE SIMPLIFIED BOILING WATER REACTOR (SBWR) DESIGN Radiat,a Protection i
i 471.1 Table 1.9-2 in the SBWR standard safety analysis report (SSAR) provides a listing of regulatory guides (RGs) which are applicable i
i to the SBWR. Modify this table to include the following regulatory l
1 guides which are listed in Chapter 12 of NUREG-0800, " Standard Review Plan."
" Guide for Administrative Practices in' Radiation l
Monitoring" RG 8.3
" Film Badge Performance Criteria" RG 8.15
" Acceptable Frograms for Respiratory Protection" RG 8.20
" Applications of Bioassay for 17125 and I-131" l
" Applications of Bioassay for Fission and j
Activation Products" RG 8.27
" Radiation Protection Training for Personnel at i
Light-Water-Cooled Nuclear Power Plants" j
1 RG 8.28
" Audible-Alarm Dosimeters" RG 8.29
" Instructions Concerning Risks From Occupational l
Radiation Exposure" 1
The staff is in the process of revising some existing regulatory I
guides and developing additional RGs to address some of the new issues contained in the revised 10 CFR Part 20.
Some of the new or revised RGs which pertain to Chapter 12 of the SBWR SSAR and which should also be addressed in Table 1.9-2 are listed below.
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" Instructions for Recording and Reporting 3
3 Occupational Radiation Exposure Data" (Revised) l RG 8.9
" Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program" (Revised)
)
" Air Sampling in the Workplace" RG 8.34
" Monitoring Criteria and Methods to Calculate Occupational Radiation Doses" RG 8.35
" Planned Special Exposures" RG 8.36
" Radiation Dose to the Embryo / Fetus" RG 8.38
" Control of Access to High and Very High Radia-tion Areas in Nuclear Power Plants" Enclosure
i 471.2 As of January 1, 1994, all licensees must comply with the revised 10 CFR Part 20. Accordingly, all references to sections of 10 CFR Part 20 in the SSAR, should reference the revised section numbers.
h 471.3 SSAR Section 12.1.2.1 lists several gene al design considerations for as-low-as-is-reasonably-achievable (ALARA) exposures.
In order to minimize job set-up time in radiation areas, needed services should be readily available at the job site. Discuss plans for.
l providing service outlets for all service systems (compressed air, breathing air, communications, service water, demineralized water, electricity, vacuum cleaning system, video and data collection cables) on all plant elevations in the vicinity of where maintenance or operating activities are expected to be performed, i
t 471.4 For complex jobs to be performed in plant. areas having high radia-1 tion levels, the use of mockups serves to familiarize workers with the exact operations that they will perform at the job site. SSAR Section 12.1 should % modified to describe how mockups will be used i
as an ALARA preplan._g aid to train workers.
471.5 SSAR Section 12.1.2.2 states that past experience has been factored into current equipirent designs to reduce personnel exposures.
However, the only example of this that is provided in the SSAR is the redesign of the steam relief valves. Amend this section of the SSAR to include other examples of-how experience from past designs t
and operating plants has been incorporated into the design of equipment and facility layout for the SBWR design.
The response i
should include varied examples of equipment / design improvements throughout the plant.
471.6 SSAR Section 12.1.2.2 lists some equipment design considerations to limit time spent in radiation areas.
This section should be amended to describe what considerations have been given to the use of radiation resistant materials in high radiation areas.
Use of such
" radiation hardened" materials in these areas would reduce the need for frequent replacement and thereby reduce personnel radiation i
exposure.
471.7 SSAR Section 12.1.2 describes several design considerations that will be used in the SBWR to maintain in-plant radiation exposures as low as is reasonably achievable. Amend this section to discuss how l
robotics will be utilized in the SBWR design to minimize personnel i
doses.
(Examples of some areas where robotics can be used to reduce personnel doses include remote weld inspections, in-service inspec-tions, and surveys of high radiation areas.)
In addition,-discuss how the SBWR design / layout will facilitate the use of robotics (e.g., use of ramps versus stairs in certain areas, accessibility of j
reactor pressure vessel belt line to perform weld inspections, use of wide doorways, etc.).
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471.8 Regulatory Guide 1.70, " Standard Format and Content of Safety Analy-sis Reports for Nuclear Power Plants," states that the SSAR should i
contain sufficient source-term characterization (i.e., radionuclide concentration, component location and geometry, component and cubicle dimensions, composition of adjacent shielding, etc.) of contained sources for the staff to be able to perform confirmatory calculations to determine dose rates in potentially occupied areas adjacent to these components. Although the SSAR provides the radionuclide concentrations and volumes for several of the contained sources, the SSAR should be amended to include the additional information needed (e.g., component location, geometry, and material composition) for the staff to be able to perform the necessary confirmatory shielding calculations.
471.9 Regulatory Guide 1.70 states that the SSAR should include a tabula-tion of the calculated concentrations of airborne radioactive materials, by nuclide, expected during normal operation and.antici-i pated operational occurrences, for equipment cubicles, corridors, and operating areas normally occupied by operating personnel.
Since this information is not contained in SSAR Section 12.2, state how you plan to comply with this guidance.
471.10 SSAR Section 12.2.1.2 states that the reactor water concentration at the reactor core exit (decay corrected since the SBWR does not have an external recirculation loop) will have an estimated value of 5.1 MBq/gm. Clarify whether this concentration is for N-16 alone or for all the radionuclides contained in the coolant.
If this concentra-t tion is for N-16 alone, justify the apparent discrepancy between this value of 5.1 MBq/gm and the value of 7.0 MBq/gm for N-16 in reactor water contained in SSAR Table 12.2-4.
471.11 The footnote to SSAR Table 12.2-4 states that the steam concentra-tion for N-16 should be increased by a factor or four during opera-tion with hydrogen water chemistry.
1.
Explain why the N-16 concentration in reactor water. should not i
also be increased by a factor of four during operation with hydrogen water chemistry.
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1 2.
Discuss whether the SBWR intends to operate using hydrogen water chemistry at all times.
3.
State whether the effect of hydrogen water chemistry on the N-16 concentration (increasing the level by a factor of four) was incorporated into the plant shielding design.
4.
Section 12.4.5 of the SSAR states that N-16 levels are increased about five times by the use of hydrogen water chemistry.
Clarify the apparent discrepancy between this value and the I
value stated in the SSAR Table 12.2-4 footnote. )
471.12 Regulatory Guide 1.70 states that the SSAR should include a listing of source activities contained in the spent fuel pool water.
Provide a table in Chapter 12 listing the source activities, by radionuclide, in the spent fuel pool water for both normal and design-basis conditions.
l 471.13 In SSAR Section 12.2, provide graphs showing the calculated dose rates in water axially above and from the side of the maximum spent fuel assembly. On these graphs, indicate how the dose rates vary as a function of time out of the core.
471.14 SSAR Section 12.3.1 states that the cobalt content will be limited to no more than 0.05 percent in the XM-19 alloy used in the SBWR control rod blades and in the Inconel X750 used in the SBWR fuel assemblies.
Several operating facilities have replaced their control rod blades with replacement blades having a cobalt content in the range of 0.015 to 0.020 percent. Justify your reasoning for not committing to a cobalt content of between 0.015 and 0.02C per-cent for components which will be in contact with reactor coolant in the reactor coolant system.
l 471.15 SSAR Section 12.3.3.2 describes how heating, ventilation, and air conditioning (HVAC) filters for the control room ventilation system will be maintained.
Verify that adjacent filter trains (for the control room and for other areas of the plant) are adequately shielded from each other so that personnel changing out filters in i
one train are not exposed to radiation from the adjacent filter train.
l 471.16 Area radiation monitors located in high noise areas should have visual as well as audible alarms.
State bow you plan to comply.
471.17 State the reasons why several of the area radiation monitors listed in SSAR Tables 12.3-2, 12.3-3, and 12.3-4 are not provided with local area alarms.
471.18 SSAR Section 9.4 should be amended to include air flow diagrams for each of the SBWR systems.
For each ventilation system which services an area having a potential for airborne radioactivity, i
i these air flow diagrams should indicate the locations of the air-borne radioactivity monitors. These monitors should be located j
upstream of the filter trains so that they monitor representative radioactivity concentrations from the areas being sampled.
471.19 Section 12.3 of NUREG-0800 states that the plant airborne radioac-tivity monitors should be able to detect ten MPC-hours of particu-i late and iodine radioactivity from any compartment which has a l
possibility of containing airborne radioactivity and which normally may be occupied by personnel, taking into account dilution in the ventilation system. Verify that the airborne radioactivity monitors at the SBWR will have this sensitivity.
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471.20 The. plant radiation zone maps (Figures 21.12.3-1 through 21.12.3-4) should be revised to indicate the following:
1.
Normal personnel traffic patterns used in entering the radiation controlled area and accessing various areas of the plant during routine plant walkdowns.
2.
Personnel traffic patterns used in accessing plant vital areas
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durir.g post-accident conditions.
3.
Boundaries for the radiation / contamination control areas.
4.
Location of the health physics facilities (including the onsite t
counting labs and the post-accident sampling station) and their design-basis radiation levels.
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471.21 The SSAR section entitled " Post-Accident Radiation Zone Maps" l
(page 12.3-20) makes reference to Figure 12.3-2, which appears to i
be missing from the SSAR.
Provide Figure 12.3-2.
471.22 SSAR Section 12.3 should include (1) a listing in the text and i
(2) the location on the appropriate radiation zone layout drawing, t
of all accessible plant areas where personnel could receive a radiation dose of I gray (100 rads) or more in one hour during i
normal and anticipated operational occurrences.
In addition, the SSAR should describe design considerations that will ensure that potentially lethal overexposures of personnel will not occur.
Particular attention should be focused on transient very high radiation areas, as well as areas with intense sources of radiation J
continuously present. An example of a transient very high radiation area is the upper drywell area which could experience elevated dose 1
rates in the event a spent fuel bundle were to be dropped onto the reactor flange during fuel movement.
471.23 10 CFR 50.34(f)(2)(vii) (Item II.B.2 of NUREG-0660 and NUREG-0737) states that the post-accident design dose rates should be such that the dose to plant personnel should not exceed 5 x 10E-2 sieverts (5 rem) whole body, or its equivalent to any part of the body, for the duration of the accident, per General Design Criteria (GDC) 19.
The dose rate in areas requiring continuous occupancy (vital areas) should be less than 15 x 10E-5 sieverts/hr (15 mrem /hr) averaged over 30 days. Verify that personnel exposures will meet GDC 19 and NUREG-0737 guidelines following a design-basis accident.
471.24 10 CFR 50.34(f)(2)(vii) states that systems containing high levels j
of radioactivity in post-accident situations be identified. State how you plan to comply with this requirement.
471.25 Verify that the Containment Atmospheric Monitoring system described in SSAR Section 12.3.4 complies with all of the requirements of 10 CFR 50.34(f)(2)(xvii) (Item II.F.1-3 of NUREG-0660 and NUREG-l 0737), including detector range, response, redundancy, separation, location, in-situ calibration, and environmental design qualifica-tion. '
f 471.26 10 CFR 50.34(f)(2)(xxvii) (Item III.D.3.3 of NUREG-0660 and NUREG-0737) states that all plants shall have equipment and associ-ated training and procedures for accurately determining the airborne iodine concentration in potentially occupied areas during an acci-dent.
State how you plan to comply with this requirement.
471.27 SSAR Section 12.4 provides man-hour estimates and average dose rates I
for various tasks performed at typical BWRs. These values are then modified, based on SBWR design improvements, to arrive at projected man-hour estimates and average dose rates for the SBWR design (listed in SSAR Table 12.4-1).
Provide your bases (number of plants in the data base and currentness of data used) for arriving at the typical BWR person-hour and dose rate values.
471.28 Regulatory Guide 1.70 states that the SSAR should address in-plant personnel exposures due to airborne radioactivity.
SSAR Sec-l tion 12.4 should list the peak airborne concentrations, estimated person-hours of occupancy, and estimated inhalation exposures for all areas of the plant accessed by plant personnel. This section should also list the assumptions used to determine airborne radioac-tivity in each building.
471.29 SSAR Section 12.4.2 states that instrumentation work in the reactor building requires 1000 person-hours per year at a typical BWR. This section then says that reactor building instrumentation work at the SBWR should take about the same effort. However, SSAR Table 12.4-1 indicates that reactor building incirumentation work for the SBWR will only require an estimated 60( person-hours.
Clarify this apparent discrepancy.
471.30 SSAP, Section 12.4.5 states that the average dose rate for work at power at a typical BWR is 6.6 rrem/hr. This section does not indicate that the dose rate for work at power at the SBWR will be any different. However, SSAR Table 12.4-1 states that the dose rate for work at power for the SBWR will be only 3.5 mrem /hr. Amend Section 12.4.5 to sta.,why the expected SBWR dose rate for work at power will be less thi.' the typical BWR dose rate.
471.31 Make the following editorial changes in SSAR Section 12 Page 12.2-12: There appears to be'some text missing in the last
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sentence of the section entitled " Radioactive Sources in the Main Steam and Feedwater Lines."
Page 12.2-16: The word " Guide" should be inserted after
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" Regulatory" in the third line of Section 12.2.6.4.
Table 12.2-16:
Provide the units to be used for the data in this table.
Page 12A-4:
Equation 12A-5 is missing a " lambda" in the second parenthetical term "(
+ R,). "
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