ML081650030

Request for Additional Information Boiling Water Reactor (BWR) Owners Group (BWROG) Licensing Topical Report (LTR) NEDO-33349, Revision 1, BWR Application to Regulatory Guide 1.97 Revision 4

ML081650030
Person / Time
Site:	Boiling Water Reactor Owners Group
Issue date:	08/04/2008
From:	Michelle Honcharik NRC/NRR/ADRO/DPR/PSPB
To:	Bunt R BWR Owners Group
References
NEDO-33349, Rev 1, RG-1.097, Rev 4, TAC MD6697
Download: ML081650030 (17)

Text

August 4, 2008 Mr. Randy C. Bunt, Chair BWR Owners= Group Southern Nuclear Operating Company 40 Inverness Center Parkway/Bin B057 Birmingham, AL 35242

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RE: BOILING WATER REACTOR (BWR) OWNERS= GROUP (BWROG) LICENSING TOPICAL REPORT (LTR) NEDO-33349, REVISION 1, BWR APPLICATION TO REGULATORY GUIDE 1.97 REVISION 4 (TAC NO. MD6697)

Dear Mr. Bunt:

By letter dated August 31, 2007 (Agencywide Documents Access and Management System Accession No. ML072470741), the BWROG submitted for U.S. Nuclear Regulatory Commission (NRC) staff review LTR NEDO-33349, Revision 1, BWR Application to Regulatory Guide 1.97 Revision 4. Upon review of the information provided, the NRC staff has determined that additional information is needed to complete the review. On June 25, 2008, Mr. Mike Iannantuono, Project Manager, and I agreed that the NRC staff will receive your response to the enclosed Request for Additional Information (RAI) questions by September 26, 2008. If you have any questions regarding the enclosed RAI questions, please contact me at 301-415-1774.

Sincerely,

/RA/

Michelle C. Honcharik, Senior Project Manager Special Projects Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project No. 691

Enclosure:

RAI questions cc w/encl: See next page

August 4, 2008 Mr. Randy C. Bunt, Chair BWR Owners= Group Southern Nuclear Operating Company 40 Inverness Center Parkway/Bin B057 Birmingham, AL 35242

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RE: BOILING WATER REACTOR (BWR) OWNERS= GROUP (BWROG) LICENSING TOPICAL REPORT (LTR) NEDO-33349, REVISION 1, BWR APPLICATION TO REGULATORY GUIDE 1.97 REVISION 4 (TAC NO. MD6697)

Dear Mr. Bunt:

By letter dated August 31, 2007 (Agencywide Documents Access and Management System Accession No. ML072470741), the BWROG submitted for U.S. Nuclear Regulatory Commission (NRC) staff review LTR NEDO-33349, Revision 1, BWR Application to Regulatory Guide 1.97 Revision 4. Upon review of the information provided, the NRC staff has determined that additional information is needed to complete the review. On June 25, 2008, Mr. Mike Iannantuono, Project Manager, and I agreed that the NRC staff will receive your response to the enclosed Request for Additional Information (RAI) questions by September 26, 2008. If you have any questions regarding the enclosed RAI questions, please contact me at 301-415-1774.

Sincerely,

/RA/

Michelle C. Honcharik, Senior Project Manager Special Projects Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project No. 691

Enclosure:

RAI questions cc w/encl: See next page DISTRIBUTION:

PUBLIC RidsNrrDpr PSPB Reading File RidsNrrDprPspb RidsNrrLADBaxley RidsAcrsAcnwMailCenter RidsNrrPMMHoncharik RidsOgcMailCenter RidsNrrDeEicb RidsNrrDssSrxb Kevin Williams, NSIR Kathryn Brock, NSIR Barry Marcus Steve LaVie, NSIR Tai Huang Aron Lewin RidsNrrDirsItsb ADAMS ACCESSION NO.: ML081650030

• No major changes from input.

NRR-106 OFFICE PSPB/PM PSPB/LA EICB/BC NAME MHoncharik DBaxley WKemper DATE 08/04/08 08/04/08 04/22/08 OFFICE SRXB/BC NSIR/BC PSPB/BC NAME GCranston*

KWilliams*

SRosenberg DATE 6/11/08 4/23/08 08/04/08 OFFICIAL RECORD COPY

BWR Owners= Group Project No. 691 Mr. Doug Coleman Vice Chair, BWR Owners= Group Energy Northwest Columbia Generating Station Mail Drop PE20 P.O. Box 968 Richland, WA 99352-0968 DWCOLEMAN@energy-northwest.com Mr. Richard Libra Executive Vice Chair, BWR Owners= Group Exelon Generation Co, LLC 200 Exelon Way Mail Code KSA 2-N Kennett Square, PA 19348 Rick.libra@exeloncorp.com Mr. Richard Anderson BWROG, Executive Chair FPL Energy (DAEC)

Duane Arnold Energy Center 3277 DAEC Road Palo, IA 52324 richard_l_anderson@fpl.com Mr. James F. Klapproth GE-Hitachi Nuclear Energy M/C A50 3901 Castle Hayne Road Wilmington, NC 28401 james.klapproth@ge.com Mr. Joseph E. Conen Regulatory Response Group Chair BWR Owners= Group DTE Energy-Fermi 2 200 TAC 6400 N. Dixie Highway Newport, MI 48166 conenj@dteenergy.com Mr. J. A. Gray, Jr.

Regulatory Response Group Vice-Chair BWR Owners= Group Entergy Nuclear Northeast 440 Hamilton Avenue Mail Stop 12C White Plains, NY 10601-5029 JGray4@entergy.com Frederick Emerson BWROG, Project Manager Frederick.emerson@ge.com Mr. Ken A. McCall, Program Manager GE Energy M/C F12 3901 Castle Hayne Road Wilmington, NC 28401 kenneth.mccall@ge.com Mr. Tim E. Abney GE Energy M/C A-16 3901 Castle Hayne Road Wilmington, NC 28401 tim.abney@ge.com Oscar Limpias Entergy Nuclear Northeast 1340 Echelon Parkway.

Jackson, MS 39213-8202 olimpia@entergy.com Joe Donahue Progress Energy Inc.

410 S. Wilmington St.

PEB 6A Raleigh, NC 27601-1849 joe.w.donahue@pgnmail.com Paul J. Davison Hope Creek Generating Station P.O. Box 236 Hancocks Bridge, NJ 08038 paul.davison@pseg.com Dennis Madison, Hatch Vice President Southern Nuclear Operating Co.

11028 Hatch Parkway North Baxley, GA 31515-2010 drmadiso@southernco.com 3/18/08

ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION LICENSING TOPICAL REPORT (LTR) NEDO-3349, REVISION 1, BWR [BOILING WATER REACTOR] APPLICATION TO REGULATORY GUIDE (RG) 1.97 REVISION 4 Instrumentation and Controls Branch

1.

On Page 2-1, Section 2.2 includes Level Control as a Type B function instead of the Type B Core Cooling Function in Revision 3 of RG 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants. Discuss the differences between the RG 1.97 Revision 3 Type B, Core Cooling Function and LTR NEDO-33349 Type B, Level Control function.

2.

On Page 2-4, Section 2.6 should discuss the qualification and design of Type E instrumentation that monitors radiological releases, in terms of being qualified to operate in the environment present when the instrument would be called upon to operate.

3.

On Page 3-1, Section 3.1, second paragraph, second sentence, the phrase but total conversion may be considered as part of plant control room upgrades including the use of digital systems, may lead to a conclusion that total conversion would only be part of a control room upgrade. This conclusion would be incorrect. A plant could convert to RG 1.97 Revision 4 without a control room upgrade. Section 3.1 should be revised to allow conversions to RG 1.97 Revision 4 without a control room upgrade.

4.

On Page 3-2, Section 3.6 should discuss the applicability of RGs that are referenced by RG 1.97 Revision 4 that also reference industry codes and standards.

5.

On Page 4-4, Section 4.1.3 should discuss the impact of common cause failures of digital systems.

6.

On Pages 4-6 thru 4-17, Sections 4.15 thru 4.5.2 call for variables that may not be the most direct variable for monitoring each function. Discuss the selection of non-direct variables in relation the IEEE-497, Clause 6.9 statement, To the extent practical, a direct variable shall be selected to monitor the related function. A less direct variable may be substituted for the most direct variable if justified by analysis. The analysis shall account for misinterpretation of the less direct variable as well as availability of reliable instrumentation, by direct variables.

7.

On Page 4-9, Section 4.2.3 should provide greater detail for the justification for how Suppression Pool Temperature fulfills the NEDO-33349 Type B key variable for the Primary Containment Control function.

8.

On Page 4-9, Section 4.2.3 should provide greater detail for the justification for how Suppression Pool Water Level fulfills the NEDO-33349 Type B key variable for the Primary Containment Control function.

9.

On Page 4-9, Section 4.2.3 should provide greater detail for the justification for Primary Containment Isolation Valve Position no longer being a Type B key variable for the RG 1.97 Revision 3 Maintaining Containment Integrity function.

10.

On Page 4-9, Section 4.2.3 should provide greater detail for the justification for the use of Drywell Pressure instead of Primary Containment Pressure as a Type B key variable for the Containment Control function.

11.

On Page 4-9, Section 4.2.3 should describe the difference between the RG 1.97 Revision 3 Type B, Maintaining RCS Integrity function and the NEDO-33349 Type B, Pressure Control function.

12.

On Page 4-9, Section 4.2.3 should provide greater detail for the justification of Drywell Pressure no longer being a Type B key variable for the RG 1.97 Revision 3 Maintaining RCS Integrity function or the NEDO-33349, Type B, Pressure Control function.

13.

On Page 4-11, Section 4.3.4 should provide greater detail for the justification for how Reactor Water Level fulfills the Type C key variable for the Fuel Cladding function.

14.

On Page 4-11, Section 4.3.4 should provide greater detail for the justification for how Reactor Water Level fulfills Type C key variable for the Reactor Coolant Pressure Boundary function.

15.

On Page 4-11, Section 4.3.4 should provide greater detail for the justification for how Suppression Pool Temperature fulfills the Type C key variable for the Reactor Coolant Pressure Boundary function.

16.

On Page 4-11, Section 4.3.4 should provide greater detail for the justification for Drywell Drain Sump Level no longer being a Type C key variable for the Reactor Coolant Pressure Boundary function.

17.

On Page 4-11, Section 4.3.4 should provide greater detail for the justification for RCS Pressure no longer being a Type C key variable for the Primary Containment function.

18.

On Page 4-11, Section 4.3.4 should provide greater detail for the justification for the use of Drywell Pressure instead of Primary Containment Pressure as a Type C key variable for the Primary Containment function.

19.

On Page 4-11, Section 4.3.4 should provide greater detail for the justification for Suppression Pool Water Level as a Type C key variable for the Primary Containment function.

20.

On Page 4-11, Section 4.3.4 should provide greater detail for the justification for Suppression Pool Temperature as a Type C key variable for the Primary Containment function.

21.

On Page 4-15, Section 4.4.5 should include statements concerning the EQ of position switches for containment isolation valves located inside and outside of containment in response to a pipe break outside of containment.

22.

On Page 5-1, Section 5, add to the last sentence of the discussion on SQ the words, following a seismic event.

23.

On Page 5-1, Section 5, the last sentence alludes to the concept that the NEDO-33349 methodology could be implemented consistent with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.59. This section should address the fact that a plant-specific conversion to RG 1.97 Revision 4 would be a change in a licensees commitment and would need to be reviewed by the NRC staff. Additionally, any plant-specific deviations from either NEDO-33349 or RG 1.97 Revision 4 would need to be submitted along with detailed justifications for those deviations for NRC staff review.

24.

Section 5 should address the impact of NUREG-0737, Clarification of TMI [Three Mile Island] Action Plan Requirements, on items that are included in both NUREG-0737 and RG 1.97.

25.

On Pages 5-1 and 5-7, Tables 5-1 and 5-2 should explain the concept of a generic list of Type A variables in consideration of the definition of Type A variables that indicates that Type A variables are plant specific.

26.

On Pages 5-1 and 5-7, Tables 5-1 and 5-2 should provide greater detail for the justification for Suppression Pool Water Level no longer being a Type D key variable for the status of Containment Related Systems.

27.

On Pages 5-1 and 5-7, Tables 5-1 and 5-2 should provide greater detail for the justification for Suppression Pool Temperature no longer being a Type D key variable for the status of Containment Related Systems.

28.

On Pages 5.1 and 5.7, Tables 5-1 and 5-2 should provide greater detail for the justification of Neutron Flux becoming a Type D variable to monitor Safety System Performance for the Reactor Protection System and the Control Rod Drive System.

29.

On Pages 5-1 and 5-7, Tables 5-1 and 5-2 should provide greater detail for the justification for Drywell Pressure no longer being a Type D key variable for the status of Containment Related Systems.

30.

On Pages 5-1 and 5-8, Tables 5-1 and 5-2 should provide greater detail for the justification for Drywell Spray Flow no longer being a Type D key variable for the status of Containment Related Systems.

31.

On Pages 5-2 and 5-8, Tables 5-1 and 5-2 should provide greater detail for the justification of Control Rod Position becoming a Type D variable to monitor Safety System Performance for the Reactor Protection System and the Control Rod Drive System.

32.

On Pages 5-3, 5-4, 5-9, and 5-10, Tables 5-1 and 5-2 should provide greater detail for the justification for the addition of Safety System Performance position switches as Type D variables.

33.

On Pages 5-4 and 5-9, Tables 5-1 and 5-2 should explain the BWR/4 and BWR/6 differences in the Classification Basis for Other RPV Normally Closed Isolation Valve Position Switches on valves that do not require opening for either a LOCA or pipe break outside of containment.

34.

On Pages 5-4 and 5-9, Tables 5-1 and 5-2 should explain the BWR 4 and BWR 6 differences in the Classification Basis for Normally Closed Containment Isolation Valve Position Switches on valves inside or outside containment that do not require opening for a loss-of-coolant accident.

35.

On Pages 5-5 and 5-11, Tables 5-1 and 5-2 should detail information concerning a generic alternate means for providing Cooling Water Temperature to Engineering Safety Feature (ESF) System Components as a Type D key variable to monitor operation of the Cooling Water System. Otherwise the review of alternate means should be plant specific.

On Page 7-2, Section 7.1 address plant specific deviations for Cooling Water Temperature to ESF System Components and Cooling Water Flow to ESF System Components from Category 2 to Category 3 when ESF Room Temperature and essential service water (ESW) Pump Running are used as Category 2 variables. On Pages 5-5 and 5-11, Tables 5-1 and 5-2 list Equipment Area Cooling System Cooling Water Temperature and Essential Service Water System Flow, but Cooling Water Temperature to ESF System Components and Cooling Water Flow to ESF System Components are not listed. Is this supposed to be a generic change? If so, provide detailed information concerning this generic change. Otherwise Cooing Water Temperature to ESF System Components and Cooling Water Flow to ESF System Components should be included in Tables 5-1 and 5-2.

36.

On Pages 5-5 and 5-11, Tables 5-1 and 5-2 should explain the apparent discrepancy between the EQ and the SQ for DC Power Status between BWR/4 and BWR/6. For BWR/4 EQ and SQ are listed as Yes and for BWR 6 EQ and SQ are listed as No.

37.

On Pages 5-6 and 5-12, Tables 5-1 and 5-2 should provide greater detail for the justification for the addition of Off Gas System Release Point Radiation Level, Ambient Air Temperature, and Control Room Area Radiation Monitors as Type E variables.

38.

On Page 6-1, Section 6 discusses guidelines for application to specific plants. Although the accident monitoring instrumentation in currently licensed plants meet the plants licensing basis, licensees converting to RG 1.97 Revision 4 will need to clearly identify design differences and equivalent variables in their plant specific applications for use of RG 1.97 Revision 4 and document any deviations. This topic should be addressed in Section 6.

39.

On Page 6-4, Section 6.5 discusses compliance with IEEE-497 referenced standards.

When IEEE standards are issued they reference the latest versions of issued IEEE standards. The NRC staff recognizes that existing plants current licensing bases predate many of these referenced standards. While it is anticipated that existing plants will maintain their current licensing basis, it is expected that each licensee will document deviations from RG 1.97 Revision 4 in their plant-specific applications for the use of RG 1.97 Revision 4. This topic should be addressed in Section 6.5.

40.

On Page 7-1, Section 7.1 references Standard Review Plan Branch Technical Position (BTP) HICB 10-5. The SRP was updated in March 2007. As part of this update, BTP HICB 10-5 was renumbered as BTP 7-10. Section 7.1 should be updated to change HICB-10-5 to 7-10 (Reference 5), and to reference the information in the March 2007 update of the SRP.

41.

On Page 7-1, Section 7.1 discusses the acceptance of RG 1.97 Revision 3 Category 3 alternate instrumentation in lieu of Category 1 Drywell Sump and Drywell Drain Sump Level instrumentation. This acceptance was based on meeting certain conditions as listed in Table 1 of SRP BTP 7-10. Section 7.1 should either include these conditions or include a determination that all existing BWRs meet these conditions, if that is the case.

42.

On Page 7-1, Section 7.1 should include details concerning the basis for each of the deviations and clarifications listed in this section.

43.

On Page 7-2, Section 7.1 includes a list of plant-specific deviations. If these deviations are applicable to the majority of BWRs, the justification for applicability should be included along with references to plant-specific safety evaluations where these deviations were approved. If these deviations are not applicable to the majority of BWRs, why is a list of plant-specific deviations included?

44.

On Page 7-3, Section 7.2 should clarify the conclusion that the five variables listed should be in the TSs along with any additional plant-specific Type A, Type B, or Type C variables.

45.

On Page 7-4, Section 7.3.2 lists Containment Area High-Range Radiation as a RG 1.97 Revision 3 Type C Category 1 variable. This is not correct. Containment Area Radiation is a RG 1.97 Revision 3 Type C Category 3 variable and Containment Area High-Range Radiation is a RG 1.97 Revision 3 Type E Category 1 variable. Licensees should be reminded that even if RG 1.97 Revision 4 allows different design and qualification criteria for Containment Area High-Range Radiation instrumentation, the criteria of NUREG-0737 Item II.F.1 are still required to be addressed.

46.

On Page A-1, Appendix A should discuss that although Type E instrumentation is not required to be environmentally qualified, this instrumentation is expected to be designed to operate in the environment that it will see when it is needed to monitor its designated variable.

47.

On Page A-1, Appendix A should explain the significance of the BWR/4 columns in Table A-1. Are these columns supposed to indicate if there is a difference between BWR/6 and BWR/4 designs for an individual variable?

48.

On Page A-2, Table A-1 shows that for the Core Cooling function, Reactor Water Level, in the BWR/4 column is listed as a Type A Category 1 variable. Is this supposed to indicate that for BWR/4 reactors the Reactor Water Level is a Type A variable and not a Type B variable? If so, what BWR/4 variable fulfills the Core Cooling function? If not, why is Coolant Level in Reactor not listed as a Type B variable in the BWR/4 column? If reactor Water Level is not shown in Table A-1 as a key variable for a Type C function, why is it listed in the IEEE-497 Type column as A, B, C?

49.

On Pages A-2 and A-5, Table A-1 shows that for the Maintain RCS Integrity function, the Reactor Coolant Pressure Boundary function, and the Containment function, Reactor Pressure, in the BWR/4 column is listed as a Type A Category 1 variable. Is this supposed to indicate that for BWR/4 reactors the Reactor Pressure is a Type A variable and not a Type B or Type C variable? If so, what BWR/4 variables fulfill the Maintain RCS Integrity function, the Reactor Coolant Pressure Boundary function, and the Containment function? If not, why is RCS Pressure not listed as a Type B and Type C variable in the BWR/4 column?

50.

On Pages A-3, A-5, and A-6, Table A-1 shows that for the Maintain RCS Integrity function, the Reactor Coolant Pressure Boundary function, the Containment function, and the Primary Containment Related Systems function, Drywell Pressure, in the BWR/4 column is listed as a Type A Category 1 variable. Is this supposed to indicate that for BWR/4 reactors the Drywell Pressure is a Type A variable and not a Type B, Type C, or Type D variable? If so, what BWR/4 variables fulfill the Maintain RCS Integrity function, the Reactor Coolant Pressure Boundary function, the Containment function, and the Primary Containment Related Systems function? If not, why is Drywell Pressure not listed as a Type B, Type C, or Type D variable in the BWR/4 column?

51.

On Page A-6, Table A-1 shows that for the Primary Containment Related Systems function, Suppression Pool Temperature, in the BWR/4 column is listed as a Type A Category 1 variable. Is this supposed to indicate that for BWR/4 reactors the Suppression Pool Temperature is a Type A variable and not a Type D variable? If so, what BWR 4 variable fulfills the Primary Containment Related Systems function? If not, why is Suppression Pool Temperature not listed as a Type D variable in the BWR/4 column? If Suppression Pool Temperature is not shown in Table A-1 as a key variable for a Type B or C function, why is it listed in the IEEE-497 Type column as A, B, C?

52.

On Page A-5 and A-6, Table A-1 shows that for the Reactor Coolant Pressure Boundary function and the Primary Containment Related Systems function, Suppression Pool Water Level, in the BWR/4 column is listed as Type A Category 1 variable. Is this supposed to indicate that for BWR 4 reactors the Suppression Pool Water Level is a Type A variable and not a Type C or D variable? If so, what BWR 4 variables fulfill the Reactor Coolant Pressure Boundary function and the Primary Containment Related Systems function? If not, why is Suppression Pool Water Level not listed as a Type C and Type D variable in the BWR/4 column? If Suppression Pool Water Level is not shown in Table A-1 as a key variable for a Type B function, why is it listed in the IEEE-497 Type column as A, B, C?

53.

On Page A-7, Table A-1 should provide greater detail for the justification for Isolation Condenser System Shell Side Water Level no longer being a Type D key variable to monitor the operation of Safety Systems.

54.

On Page A-7, Table A-1 should provide greater detail for the justification for Isolation Condenser System Valve Position no longer being a Type D key variable to monitor status of Containment Related Systems.

55.

On Page A-8, Table A-1 should provide greater detail for the justification for how High Radioactivity Liquid Tank Level being a normal operating system would have an impact on it meeting the criteria for a Type D variable for monitoring the operation of Radwaste Systems.

56.

On Page A-8, Table A-1 should include greater detail for the justification for a generic alternate means for providing Emergency Ventilation Damper Position as a Type D key variable to monitor operation of the Ventilation Systems. Otherwise the review of alternate means should be plant specific.

57.

On Page A-9, Table A-1 should provide greater detail for the justification for Secondary Containment Release Point Radiation Level becoming a Type E variable.

58.

On Page A-9, Table A-1 should provide greater detail for the justification for Radiation Exposure Rate no longer a Type E variable for Area Radiation. RG 1.97 Revision 3 includes Radiation Exposure Rate as a Type E variable to provide detection of significant releases, release assessment, and long-term surveillance for Area Radiation. On Page 7-1, Section 7.1 identifies Radiation Exposure Rate as being granted a generic deviation to Category 3. Therefore, Radiation Exposure Rate should be included in Table A-1.

59.

On Page A-9, Table A-1 should provide greater detail for the justification for RG 1.97 Revision 3 Airborne Radiation variables no longer being Type E for Airborne Radioactive Materials Released from Plant variables for detection of significant releases and release assessment for Airborne Radioactive Materials Released from Plant. RG 1.97 Revision 3 includes (a) Noble Gases and Vent Flow Rate and (b) Particulates and Halogens as Type E variables to provide detection of significant releases, release assessment, and in some locations long term surveillance, for Airborne Radioactive Materials Released from Plant. On Page A-9, Table A-1 lists (a) Noble Gases and Vent Flow Rate and (b) Particulates and Halogens as Type D variables. Provide detailed justifications for these changes.

60.

RG 1.97 Revision 3 includes Estimation of Atmospheric Stability as a Type E variable.

On Page A-9, Table A-1 also lists Estimation of Atmospheric Stability as a Type E variable. However, on Pages 5-6 and 5-12, Tables 5-1 and 5-2 list Ambient Air Temperature as a Type E variable. Discuss the differences, if any, between Estimation of Atmospheric Stability and Ambient Air Temperature and provide appropriate detailed justification for any deviation.

61.

On Page A-9, Table A-1 should provide greater detail for the justification for Primary Coolant and Sump no longer being Type E for release assessment, verification, and analysis of Accident Sampling Capability.

62.

On Page A-9, Table A-1 should provide greater detail for the justification for Containment Air no longer being Type E for release assessment, verification, and analysis of Accident Sampling Capability.

Nuclear Security and Incident Response

1.

Section 1.3.6 should be expanded to acknowledge, and provide guidance on, situations in which the current licensing basis requirements are more restrictive than the requirements of Revision 4 to RG 1.97. Please add appropriate language or provide justification why the BWROG believes that this change is not necessary. For example:

NUREG-0737,Section II.F.1, Attachment 1, Noble Gas Effluent Monitors (high range), requires that the monitors are capable of performing their intended function in the environment to which they may be exposed during accidents, be powered from vital instrumentation bus or dependable backup power supply, and that their operability be addressed by TSs. This parameter is designated as a Type E variable.

NUREG-0737,Section II.F.1, Attachment 2, Sampling and Analysis of Plant Effluents, requires the preparation of TSs. This parameter is designated as a Type E variable.

NUREG-0737,Section II.F.1, Attachment 3, Containment High Range Radiation Monitor, requires a minimum of two Category 1 containment high-range monitors qualified to function in the accident environment, be powered from Category 1E power sources, and that their operability be addressed by TSs. This monitor is designated as a Type C and a Type E variable.

2.

Although the LTR has references to the current plant licensing basis and the need to perform evaluations against this licensing basis, the NRC staff believes that more specificity is needed, and requests that a clarification such as that suggested below be added to Section 1.4, Limitations." Please incorporate appropriate language, or provide justification why the BWROG believes that this change is not necessary.

Proposed changes to a plants accident monitoring variables, their classification under Regulatory Guide 1.97 Revision 4, and the associated treatment requirements (e.g., environmental qualification, technical specifications, etc.) must be evaluated within the context of the specific plants current licensing basis pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.59. In addition, proposed changes to any instrumentation, relied upon by the plants emergency plans to meet the planning standards of 10 CFR 50.47(b) and Appendix E to Part 50, must be evaluated pursuant to 10 CFR 50.54(q) to ascertain whether the proposed change would decrease the effectiveness of those plans. In this context, any change that would reduce the performance, reliability, or availability of such instruments, without compensatory measures, during an emergency condition will likely constitute a potential decrease in the effectiveness of the plans, requiring prior NRC approval.

3.

Section 1.5 appears to limit the applicability to plant-specific commitments with respect to accident monitoring that are documented in the UFSAR [update final safety analysis report] or other applicable license amendment documents. The NRC staff believes that the last sentence of this section should read:

include all plant-specific commitments with respect to accident monitoring that are documented in the current licensing basis, including but not limited to the UFSAR.

Please incorporate appropriate language, or provide justification why the BWROG believes that this clarification is not required.

4.

The last paragraph of Section 2.1 appears to inappropriately link the safety analyses with emergency procedure guidelines (EPGs) and emergency operation procedures (EOPs).

Section 1.5, provides that safety analysis is defined by anticipated operational occurrences [(AOOs)] and accidents or other equivalent nomenclature used in the safety or accident analysis section of the updated final safety analysis report (UFSAR). Yet, the safety analyses generally do not credit actions taken in accordance with EPGs or EOPs. Also, EPGs and EOPs can be generally characterized as taking credit for all available plant equipment and plant services (e.g., AC power) with response not obtained steps to address unavailability of that resource. However, the safety analyses only credit safety-related equipment, and offsite power is generally assumed to be lost at the accident onset. Is safety analysis as used here defined differently than in Section 1.5? If so, clarification is needed. Does the dichotomy in treatment impact the LTR methodology? Please reconsider this language and make necessary changes for clarity, or provide justification why the existing language is appropriate.

5.

In Section 2.4 the third portion of the definition of Safety System is not fully consistent with the regulatory definition used in 10 CFR Part 50. The definition should read:

The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in Section 50.34(a)(1),

Section 50.67(b)(2), or Section 100.11 of this chapter, as applicable.

Please revise the LTR definition accordingly.

6.

The discussion in Section 2.4, Pages 2 and 3 should be changed to acknowledge that a plants current licensing basis safety analyses may identify other BWR accidents for which there is a significant radioactivity release. Please incorporate appropriate changes, or provide justification why the BWROG believes that these changes are not required.

7.

The last paragraph of Section 2.4 appears to conflict with the definition of safety-related established earlier in Section 2.4. Whether or not a particular system, structure, or component is safety-related is established by the three-part definition of safety-related without regard to any particular AOOs or accidents in which the equipment may be credited. Please incorporate appropriate changes, or provide justification why the BWROG believes that these changes are not required.

8.

The parenthetical phrase in the first bullet of Section 2.5 should be expanded to include standby gas treat system and reactor building ventilation. The major portion of an accident release will be via these two pathways and should not be overlooked. Please incorporate appropriate changes, or provide justification why the BWROG believes that these changes are not required.

9.

In Section 2.5 there needs to be a bullet to reflect the need for monitoring the fission product inventory in the containment atmosphere as a means of assessing potential releases to the environment that (1) have not yet started, or (2) are via unmonitored pathways. Such releases would not be indicated by effluent monitors addressed by the first bullet. Risk studies have shown that the most severe releases may be via unmonitored pathways. Since these releases will generally involve the release of fission products from the containment, monitoring the containment inventory meets an assessment need.

10.

In Section 4.2.2 on Page 4-8 in the last sentence in the paragraph addressing the RCS fission product barrier is potentially misleading. A typical General Electric (GE) design basis analysis for the control rod drop accident postulates that the main steam isolation valves (MSIVs) remain open throughout the accident (In NEDO-31400, the main steam line radiation monitor activation of reactor trips and MSIV isolation was eliminated);

creating a bypass of both the RCS and containment barriers. Please incorporate appropriate changes, or provide justification why the BWROG believes that these changes are not required.

11.

Section 4 discusses the application methodology to be used to determine the accident monitoring variables consistent with the requirements of RG 1.97 Revision 4 and IEEE-497. The subsections for Section 4 then develop the accident monitoring variables associated with each of the variable types. Section 4.3 and Section 4.3.2 state that Type C variables are selected to represent the minimum set of parameters that provide the most direct indication of the integrity of the fission product barriers and provide a capability for monitoring beyond the normal operating range [emphasis added]. The NRC staff questions the implementation of this stipulation in arriving at the parameter listings for the fuel clad fission product barrier. In particular, Section 4.3.4.1 identifies the following two parameters as meeting the characterization for the fuel clad barrier:

Reactor water level Off gas activity (monitoring performed by normal operating systems)

The NRC staff is of the opinion that this listing inappropriately omits the containment high range radiation monitor -- most direct indication of the integrity of the fuel clad fission product barriers and the instrument most capable of monitoring beyond normal operating ranges. It is important to note that an upscale reading on this monitor is indicative of a breach of both the RCS barrier and the fuel clad barrier, and is therefore an indication for both barriers, differing only in the higher magnitude indication associated with fuel barrier failure as opposed to normal RCS activity associated with an RCS barrier failure.

a.

Although the NRC staff agrees that a decrease in reactor water level is a required precursor to fuel damage and is an indicator of potential fuel damage, it is not a direct indicator of that damage. As such, this parameter does not appear to meet the Section 4.3.2 basis as a direct indicator. Consider an accident sequence in which the emergency core coolant system (ECCS) is not initially successful, allowing water level to decrease below the minimum steam cooling reactor water level (MSCRWL), with the ECCS then restored and water level restored (1) prior to significant fuel clad damage, or (2) after fuel clad damage has occurred. In either of these sequences, the reactor water level is not likely a reliable indicator of fuel damage. Monitoring this Reactor water level is not a means for monitoring the integrity of the fuel clad once the water level has been restored.

b.

The NRC staff notes that the MSCRWL is a calculated parameter based on conservative analysis assumptions and that, depending on the actual transient conditions, may be uncertain. Since the containment high range monitors directly monitor the increase in radiation levels in the containment atmosphere, a direct consequence of the release of fission products from the fuel, the uncertainties are expected to be less.

c.

The off gas radiation monitors are typically designed to detect increases in main steam activity comparable to TS limiting conditions for operation. Although these monitors can be most the sensitive and timely indicators of increased RCS and main steam activity (e.g., clad defects, etc.) during normal operating conditions, they would likely be off scale for an incident involving substantial fuel damage.

As such, this parameter does not appear to meet the Section 4.3.2 basis as being capable of monitoring beyond normal operating ranges. Also, since these monitors are located downstream of the main condenser, isolation of the MSIVs following a design basis LOCA, or an accident in conjunction with a loss of circulating water to the main condenser (e.g., loss of offsite power), will effectively isolate the off gas monitors rendering them unusable as a direct indicator or monitor of fuel damage. There is a high degree of uncertainty associated with the transport and deposition of radioactive materials through the main steam piping, in the main condenser, and in the various components and filter media of the off gas system, making the process of equating the reading on the off gas radiation monitor to fuel clad status, uncertain.

For the reasons stated above, the NRC staff does not agree with the omission of the containment high range monitor from this list of Type C variables and from Tables 5-1 and 5-2 and requests that the BWROG reconsider its omission, or provide additional justification supporting the use of the reactor water level and off gas activity as meeting the parameter characteristics in Section 4.3.2. Such justification must address the NRC staffs concerns identified above and show that alternative variables meet the direct and beyond normal range characterization provided in Section 4.3.

12.

In Section 5 on Page 5-2, the text:...consistent with the provisions of 10 CFR 50.59 subject to plant reviews of their licensing commitments. must be revised to read:

consistent with the provisions of 10 CFR 50.59, and 10 CFR 50.54(q), as applicable, subject to plant reviews of their current plant licensing basis.

The NRC staff also requests that language be added to this section to clearly emphasize that not one size fits all and that, because of current licensing basis differences between facilities, what may be implemented under 10 CFR 50.59 and 10 CFR 50.54(q) at one facility may not be acceptable for another facility; that all such reviews shall be made against the plants current licensing basis.

Please make the requested changes or provide justification why the changes are not warranted.

13.

Section 4.3.4.1 identifies off gas activity as a Type C variable; yet Table 5-1 and Table 5-2 do not identify this parameter as a Type C variable. Please revise Tables 5-1 and 5-2, or Section 4.3.4.1, accordingly, or provide justification for this inconsistency.

14.

Consistent with the comments in Item 11 above, Table 5-1 and Table 5-2 need to be revised to include the containment radiation level as a Type C variable.

15.

Tables 5-1 and 5-2 identify containment radiation level as a Type E variable and indicate that EQ and SQ are not necessary. Although the NRC staff recognizes that under RG 1.97 Revision 4 and IEEE-497, the variables do need not be environmentally qualified to the requirements of RG 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, the monitors are required by NUREG-0737 Section II.F.1, Attachment 3, Containment High Range Radiation Monitor, to function in the accident environment. Components of these monitoring system (e.g., detectors, cabling) would need to be qualified for the post-accident environment (a design envelope is provided in NUREG-0737) within the containment since it under these conditions that the monitors will perform their design function.

Similar requirements apply to the high range noble gas effluent monitors addressed by NUREG-0737 Section II.F.1, Attachment 1, Noble Gas Effluent Monitors. The NUREG-0737 requirements were imposed on existing licensees via a generic letter and a confirming order. They are imposed on future license applications via 10 CFR Sections 50.34(f), 52.47(a)(ii), 52.79(b), and 52.83. Please revise the EQ categorization for the subject variables to Y* or something similar, and explain the requirement in the text or a footnote to the tables, or provide additional justification for the proposed treatment of these variables.

16.

The NRC staff finds the argument in Section 7.3.2 to be non-persuasive. The discussion notes that the high range containment monitors were included in RG 1.97 because of the requirements established in NUREG-0737,Section II.F.1. The requirement for the high range containment monitor arose out of the lessons-learned at TMI, where the existing monitors over-ranged or were otherwise unreliable causing difficulties for assessment of the plant status. The LTR argument then discusses some incremental changes in nomenclature between versions of RG 1.97. However, the intent of the LTR is to provide a methodology for establishing a technical basis for which parameters are required to be accident monitors; as an alternative to RG 1.97. What the argument does not provide is a basis for determining that the post-accident indications that the monitor would provide are not necessary or that they do not warrant the redundancy, quality, and TSs associated with a Type C variable. The NRC staff notes that the NUREG-0737 requirements were the subject of a generic letter and the licensee responses to that generic letter were accepted by means of confirming orders. The requirements were added to 10 CFR 50.34(f) for all pending Part 50 and future licensing under Part 52 where technically relevant. The NRC staff notes that that the containment radiation monitor is an indicator of the failures of both the RCS fission product barrier AND the fuel clad fission product barrier since the monitor measures the radiation from fission products released to the containment which requires BOTH barriers to fail. The NRC staff does not dispute that parameters such as reactor water level and pressure, drywell pressure, etc., are more direct and less uncertain indicators of a RCS barrier breach.

However, the same can not be said for the fuel clad barrier, for which Section 4.3.4.1 of this LTR does not provide a direct indication of breach as was discussed in Item 11 above. Please provide additional justification for your position.

17.

In Table A-1 on Page A-5, with regard to the primary containment area radiation entry under Type C variables, the note specifies that Not relied on in accident analysis or EPGs for breach of barrier. Only function is for EALs. The staff notes that the emergency action levels (EALs) are used to initiate the sites emergency response plan that provide for the protection of the public in those rare circumstances in which engineered design features and human capacity to take corrective actions have both failed to avert a serious mishap. The view that emergency planning is secondary to engineered design features and safe siting was refuted by the unexpected sequence of events that occurred at TMI. The Commission emphasizes the integration of safety, security, and emergency preparedness as the basis for the NRCs primary mission of protecting public health and safety. The Type C variables, which monitor fission product barriers, are particularly significant in that a General Emergency, the level at which public protective actions are necessary, is defined as the loss of two fission product barriers and a potential loss of the third barrier. As such, the staff does not find an argument based on only function is for EALs to be particularly persuasive as a justification for reducing the treatment requirements for the containment radiation instrumentation addressed in EALs. Please provide additional justification for your position.

Reactor Systems Branch

1.

The LTR identifies the requirements of IEEE-497 with respect to the five types of accident monitoring system variables. The five types of accident monitoring system variables are defined as follows: Type A variables provide the operators with the primary information necessary to take the normal actions credited in the safety analysis; Type B variables provide primary information to the control room operators to assess the plant critical safety functions; Type C variables provide extended range primary information to the control room operators to indicate the potential breach or the actual breach of the fission product barriers; Type D variables provide information to the control room operators to indicate both the performance of those required systems and auxiliary supporting features necessary for the mitigation of AOOs and accidents and the performance of other system necessary to achieve and maintain a safe shutdown condition, and to verify system status; and Type E variables provide information to be used in determining the magnitude of the release of radioactive material and continually assessing such releases.

Safety analysis events including AOOs and accidents are given in Table 4-1, systems assumed in the safety analysis including events, required action, and system assumed are given in Table 4-2, and required system, shutdown systems, and auxiliary support systems are given in Table 4-3.

It requires that approved methodologies should be used to analyze safety analysis events identified in Table 4-1 and to apply the results of the analysis to supporting the required action using system assumed for safety analysis events listed in the Table 4-2.

Please provide: (1) approved methodologies used to analyze the safety analysis events listed in Table 4-1; (2) identification of parameters and its acceptable criterion to be used for required action taken by operators using system assumed; (3) description of an updated version of BWR EPGs and the Severe Accident Guidelines (SAGs) used to perform the critical safety functions; and (4) description of an updated BWR EOPs to support shutdown systems.