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Category:MEETING MINUTES & NOTES--CORRESPONDENCE
MONTHYEARML20206S6461999-02-17017 February 1999
Summary of ACRS Reliability & Probabilistic Risk Assessment, Plant Operations & on Regulatory Policies & Practices Subcommittees 981119-20 Meeting in Rockville,Md Re Options Make 10CFR50 & 10CFR50.59 risk-informed
ML20206S4381999-01-25025 January 1999
Summary of 458th ACRS Meeting on 981203-05 in Rockville,Md Re Appropriate Action on Items Listed in Attached Agenda
ML20206S6611999-01-11011 January 1999
Summary of 981215-17 ACRS Thermal Hydraulic Phenomena Subcommittee Meeting in Rockville,Md Re T/H Code Activities/ NRC T/H Research Status
ML20206S6101998-12-23023 December 1998
Summary of ACRS Reliability & Probabilistic Risk Assessment Subcommittee 981029 Meeting in Rockville,Md Re Options to Make 10CFR50 risk-informed,NEI Whole Plant Study & Options for Developing risk-informed Approach,Revising 10CFR50.59
ML20206S8501998-11-13013 November 1998
Summary of 456th ACRS Meeting on 980930-1002 in Rockville,Md Re Appropriate Action on Items Listed in Attached Agenda
ML20206S2481998-11-12012 November 1998
Summary of 455th ACRS Meeting on 980902-04 in Rockville,Md Re Items Listed in Attached Agenda
ML20206S7621998-11-12012 November 1998
Summary of 980924 ACRS Submommittee on Reliability & PRA Open Meeting in Rockville,Md Re Proposed Options for Developing risk-informed Approach to Revising 10CFR50.59, Options for Broader Changes to 10CFR50
ML20206S4161998-09-29029 September 1998
Summary of 980826 Open ACRS Subcommittee on Reliability & PRA Meeting in Rockville,Md Re Issues in SRM on situation-specific Cases Where PRA Results & Insights Have Improved Existing Regulatory Sys
ML20206S3851998-09-0303 September 1998
Summary of 980729 Open ACRS Joint Meeting of Subcommittees on Plant Operations & Fire Protection in Atlanta,Ga Re Region II Activities & Other Items of Mutual Interest, Including Significant Operating Events & Fire Protection
ML20206S6831998-09-0101 September 1998
Summary of 453rd ACRS Meeting on 980603-05 in Rockville,Md Re Appropriate Action on Items Listed in Attached Agenda
ML20206S3761998-08-24024 August 1998
Summary of 980716 ACRS Plant License Renewal Subcommittee Meeting in Rockville,Md Re Presentations by & Holding Discussions with NRC Staff Re License Renewal,Proposed Staff Plans & Schedule for Reviewing License Renewal Applications
ML20206S3811998-08-0404 August 1998
Summary of ACRS Safety Research Program Subcommittee Meeting on 980717 in Rockville,Md Re Comments & Recommendations in 980616 ACRS Rept, Core Research Capabilities & Associated Staff Response
ML20206S3721998-07-29029 July 1998
Summary of 980707 ACRS Advanced Reactor Designs Subcommittee Meeting in Rockville,Md Re Info on AP600 Test & Analysis Program & Responses to ACRS Questions Asked During Previous Meetings
ML20206S6721998-07-0808 July 1998
Summary of ACRS Safety Research Program Subcommittee 980601 Meeting in Rockville,Md Re Discussion with NRC Concerning SECY-98-076, Core Research Capabilities
ML20206S3671998-06-30030 June 1998
Summary of 980619 Open ACRS Meeting of Subcommittee on Plant Operations in Rockville,Md Re Proposed Changes to 10CFR50.59,status of Resolution of Issues Identified in 980324 SRM Related to SECY-97-205 & Related Matters
ML20206S3501998-06-26026 June 1998
Summary of 980617-18 ACRS Advanced Reactor Designs Subcommittee Meeting in Rockville,Md Re Review of Advance FSER Chapters 4,5,7,8,11,13 & 18,level 1 AP600 Pra,Itaac & Associated ACRS Open Questions
ML20206S5991998-06-23023 June 1998
Summary of ACRS Reactor Fuels,Onsite Fuel Storage & Decommissioning Subcommittee 980423-24 Meeting in Rockville, MD Re Basis of Proposed NRC Fuel Failure Criterion for High Burnup Conditions & Adequacy of NRC Fuel Codes
ML20206S3361998-06-22022 June 1998
Summary of 980611-12 Open ACRS Thermal Hydraulic & Severe Accident Phenomena Subcommittee Meeting in Rockville,Md Re Continuance of Review of Results of W Test & Analysis Program in Support of AP600 Design Certification
ML20206S6411998-06-22022 June 1998
Summary of 980601 ACRS Matls & Metallurgy Subcommittee Meeting in Rockville,Md Re NRC Staff Concerns Related to ASME Boiler & Pressure Vessel Code,Section III Rule Revs
ML20206S2561998-06-17017 June 1998
Summary of 452nd ACRS Meeting on 980430-0502 in Rockville, MD Re Items Listed in Attached Agenda
ML20206S6301998-06-17017 June 1998
Summary of 980602 ACRS Thermal Hydraulic & Severe Accident Phenomena Subcommitteemeeting in Rockville,Md Re GE Extended Power Uprate Program for Operating BWRs & NSP lead-plant Application for Power Uprate
ML20206S5761998-06-0404 June 1998
Summary of ACRS Reliability & Probabilistic Risk Assessment Subcommittee 980416 Meeting in Rockville,Md Re Subcommittees Review of Matters Related to Elevation of CDF & Possible Rev to Commissions Safety Goal Policy Statement
ML20206S5671998-05-30030 May 1998
Summary of ACRS Advanced Reactor Designs Subcommittee 980513-15 Meeting in Rockville,Md Re Review of AP600 Standard SAR & Associated Advanced FSER Chapters 3,6,14,16 & 17
ML20206S2521998-05-14014 May 1998
Summary of 451st ACRS Meeting on 980402-04 in Rockville,Md Re Items Listed in Attached Agenda
ML20206S2801998-04-30030 April 1998
Summary of 980219-20 ACRS Open Meeting of Subcommittee on Reliability & Probabilistic Risk Assessment in Rockville,Md Re Continuance of Review of Proposed Final SRP Sections & RGs for risk-informed,performance-based Regulation
ML20206S2951998-04-15015 April 1998
Summary of 450th ACRS Meeting on 980305-07 in Rockville,Md Re Items Listed in Attached Agenda
ML20206S2871998-04-15015 April 1998
Summary of 449th ACRS Meeting on 980302-04 in Rockville,Md Re Items Listed in Attached Agenda
ML20206S6481998-03-26026 March 1998
Summary of ACRS Fire Protection Subcommittee 980122 Meeting in Rockville,Md Re Staff Actions Taken Related to Development of Revised Fire Protection Rule
ML20196L0531998-03-20020 March 1998
Summary of 980205-07 ACRS 448th Meeting in Rockville,Md Re Items Listed in Attached Agenda
ML20206S6561998-03-13013 March 1998
Summary of ACRS Plant License Renewal Subcommittee 980123 Meeting in Rockville,Md Re License Renewal Implementation Issues & Proposed Industry Guidelines
ML20206S2751998-03-0909 March 1998
Summary of 980218 ACRS Thermal Hydraulic & Severe Accident Phenomena Subcommittee Open Meeting in Rockville,Md Re Review of RES Program in Area of thermal-hydraulic Phenomena in Support of ACRS Planned Rept to Commission
ML20217A8561998-03-0505 March 1998
Summary of ACRS 448th Meeting on 980205-07 Re Several Matters & Completed Listed Rept & Ltr.Committee Authorized Larkins,Executive Director,To Transmit Memoranda Listed
ML20206S6791998-03-0505 March 1998
Summary of ACRS Plant Operations Subcommittee 980203 Meeting in Rockville,Md Re Subcommittee Review of Proposed Improvements to Senior Mgt Meeting Process
ML20206S2651998-02-23023 February 1998
Summary of 980203-04 ACRS Subcommittee on Advanced Reactor Designs Open Meeting in Rockville,Md Re Review of Chapters 1,4,5,7,8,11,13 & 18 of AP600 Ssar & AP600 Test & Analysis Program
IR 07100203/20064471998-02-0505 February 1998
Summary of 971203-06 447th ACRS Meeting in Rockville,Md Re Appropriate Action on Items Listed in Attached Agenda
ML20206S3931998-01-28028 January 1998
Summary of 980121 ACRS Subcommitee on Human Factors Meeting in Rockville,Md Re Human Performance & Reliability Plan & Integration of Human Factors Insights Into Insp Process
ML20217A8461998-01-0909 January 1998
Summary of ACRS Plant Operations Subcommittee 971202 Meeting in Rockville,Md Re SER on industry-sponsored Utility Resolution Guidance for Issue of Blockage of Suction Strainer in ECCS
ML20217A8771997-12-23023 December 1997
Summary of ACRS 446th Meeting on 971106-07 in Rockville,Md Re Appropriate Action on Items Listed on Attached Agenda.W/ Fr Notice,Meeting Schedule & Outline,List of Attendees, Future Agenda & List of Documents Provided to Committee
ML20217A9911997-12-23023 December 1997
Summary of 447th ACRS Meeting on 971203-06 Re Proposed Revs to 10CFR50.59 & Credit for Containment Overpressure to Provide Assurance of Sufficient Net Positive Suction Head for ECC & Containment Heat Removal
ML20217A8651997-12-15015 December 1997
Summary of ACRS Reliability & PRA Subcommittee 971112-13 Meeting in Rockville,Md Re Proposed Final SRP for risk- informed,performance-based Regulation
ML20217A9831997-12-0303 December 1997
Summary of 971203 ACRS Meeting in Rockville,Md Re Conduct of ACRS Business
ML20217A5901997-12-0202 December 1997
Summary of 445th ACRS Meeting in Rockville,Md on 971002-03 Re Appropriate Action on Items Listed in Attached Agenda.W/ Fr Notice,Meeting Schedule & Outline,List of Attendees, Future Agenda & List of Documents Provided to Committee
ML20217A8811997-12-0202 December 1997
Summary of ACRS Safety Research Program Subcommittee 971104- 05 Meeting in Rockville,Md Re NRC Safety Research Program & Draft Annual Rept to Congress
ML20217A8221997-12-0101 December 1997
Informs That During 446th Meeting on 971106-07,ACRS Discussed Several Matters & Completed Listed Ltr. Committee Authorized Larkins,Executive Director,To Transmit Memoranda,Listed
ML20217A8901997-11-21021 November 1997
Summary of ACRS Reliability & PRA 971021-22 Meeting in Rockville,Md Re Matter Included in Staff Requirements Memo,
ML20217A6381997-11-0606 November 1997
Summary of 444th ACRS Meeting in Rockville,Md on 970903-05 Re Appropriate Action Items Listed in Attached Agenda.W/Fr Notice,Meeting Schedule & Outline,List of Attendees & List of Documents Provided to Committee
ML20217B1921997-11-0606 November 1997
Summary of ACRS Planning & Procedures Subcommitte 971105 Meeting in Rockville,Md Re Conduct of ACRS Business
ML20217A8001997-10-21021 October 1997
Informs That During 445th Meeting on 971002-03,ACRS Discussed Several Matters & Completed Listed Rept & Ltrs. Committee Authorized Larkins,Executive Director,To Transmit Memoranda Listed
ML20217A7011997-10-20020 October 1997
Summary of 970826-27 ACRS Subcommittees on Matls & Metallurgy & on Severe Accidents Joint Meeting W/Nrc,Nei & Industry in Rockville,Md Re Review of Proposed Draft GL & Associated Draft Regulatory Guide Re SG Tube Integrity
ML20217A8141997-10-20020 October 1997
Summary of ACRS Subcommittees on Pra,Plant Operations & Fire Protection 970828-29 Meeting in Rockville,Md Re Review of Staff Requirements Memo
1999-02-17
[Table view]Some use of "" in your query was not closed by a matching "".
Category:MEETING SUMMARIES-INTERNAL (NON-TRANSCRIPT)
MONTHYEARML20206S6461999-02-17017 February 1999
Summary of ACRS Reliability & Probabilistic Risk Assessment, Plant Operations & on Regulatory Policies & Practices Subcommittees 981119-20 Meeting in Rockville,Md Re Options Make 10CFR50 & 10CFR50.59 risk-informed
ML20206S4381999-01-25025 January 1999
Summary of 458th ACRS Meeting on 981203-05 in Rockville,Md Re Appropriate Action on Items Listed in Attached Agenda
ML20206S6611999-01-11011 January 1999
Summary of 981215-17 ACRS Thermal Hydraulic Phenomena Subcommittee Meeting in Rockville,Md Re T/H Code Activities/ NRC T/H Research Status
ML20206S6101998-12-23023 December 1998
Summary of ACRS Reliability & Probabilistic Risk Assessment Subcommittee 981029 Meeting in Rockville,Md Re Options to Make 10CFR50 risk-informed,NEI Whole Plant Study & Options for Developing risk-informed Approach,Revising 10CFR50.59
ML20206S8501998-11-13013 November 1998
Summary of 456th ACRS Meeting on 980930-1002 in Rockville,Md Re Appropriate Action on Items Listed in Attached Agenda
ML20206S2481998-11-12012 November 1998
Summary of 455th ACRS Meeting on 980902-04 in Rockville,Md Re Items Listed in Attached Agenda
ML20206S7621998-11-12012 November 1998
Summary of 980924 ACRS Submommittee on Reliability & PRA Open Meeting in Rockville,Md Re Proposed Options for Developing risk-informed Approach to Revising 10CFR50.59, Options for Broader Changes to 10CFR50
ML20206S4161998-09-29029 September 1998
Summary of 980826 Open ACRS Subcommittee on Reliability & PRA Meeting in Rockville,Md Re Issues in SRM on situation-specific Cases Where PRA Results & Insights Have Improved Existing Regulatory Sys
ML20206S3851998-09-0303 September 1998
Summary of 980729 Open ACRS Joint Meeting of Subcommittees on Plant Operations & Fire Protection in Atlanta,Ga Re Region II Activities & Other Items of Mutual Interest, Including Significant Operating Events & Fire Protection
ML20206S6831998-09-0101 September 1998
Summary of 453rd ACRS Meeting on 980603-05 in Rockville,Md Re Appropriate Action on Items Listed in Attached Agenda
ML20206S3761998-08-24024 August 1998
Summary of 980716 ACRS Plant License Renewal Subcommittee Meeting in Rockville,Md Re Presentations by & Holding Discussions with NRC Staff Re License Renewal,Proposed Staff Plans & Schedule for Reviewing License Renewal Applications
ML20206S3811998-08-0404 August 1998
Summary of ACRS Safety Research Program Subcommittee Meeting on 980717 in Rockville,Md Re Comments & Recommendations in 980616 ACRS Rept, Core Research Capabilities & Associated Staff Response
ML20206S3721998-07-29029 July 1998
Summary of 980707 ACRS Advanced Reactor Designs Subcommittee Meeting in Rockville,Md Re Info on AP600 Test & Analysis Program & Responses to ACRS Questions Asked During Previous Meetings
ML20206S6721998-07-0808 July 1998
Summary of ACRS Safety Research Program Subcommittee 980601 Meeting in Rockville,Md Re Discussion with NRC Concerning SECY-98-076, Core Research Capabilities
ML20206S3671998-06-30030 June 1998
Summary of 980619 Open ACRS Meeting of Subcommittee on Plant Operations in Rockville,Md Re Proposed Changes to 10CFR50.59,status of Resolution of Issues Identified in 980324 SRM Related to SECY-97-205 & Related Matters
ML20206S3501998-06-26026 June 1998
Summary of 980617-18 ACRS Advanced Reactor Designs Subcommittee Meeting in Rockville,Md Re Review of Advance FSER Chapters 4,5,7,8,11,13 & 18,level 1 AP600 Pra,Itaac & Associated ACRS Open Questions
ML20206S5991998-06-23023 June 1998
Summary of ACRS Reactor Fuels,Onsite Fuel Storage & Decommissioning Subcommittee 980423-24 Meeting in Rockville, MD Re Basis of Proposed NRC Fuel Failure Criterion for High Burnup Conditions & Adequacy of NRC Fuel Codes
ML20206S6411998-06-22022 June 1998
Summary of 980601 ACRS Matls & Metallurgy Subcommittee Meeting in Rockville,Md Re NRC Staff Concerns Related to ASME Boiler & Pressure Vessel Code,Section III Rule Revs
ML20206S3361998-06-22022 June 1998
Summary of 980611-12 Open ACRS Thermal Hydraulic & Severe Accident Phenomena Subcommittee Meeting in Rockville,Md Re Continuance of Review of Results of W Test & Analysis Program in Support of AP600 Design Certification
ML20206S2561998-06-17017 June 1998
Summary of 452nd ACRS Meeting on 980430-0502 in Rockville, MD Re Items Listed in Attached Agenda
ML20206S6301998-06-17017 June 1998
Summary of 980602 ACRS Thermal Hydraulic & Severe Accident Phenomena Subcommitteemeeting in Rockville,Md Re GE Extended Power Uprate Program for Operating BWRs & NSP lead-plant Application for Power Uprate
ML20206S5761998-06-0404 June 1998
Summary of ACRS Reliability & Probabilistic Risk Assessment Subcommittee 980416 Meeting in Rockville,Md Re Subcommittees Review of Matters Related to Elevation of CDF & Possible Rev to Commissions Safety Goal Policy Statement
ML20206S5671998-05-30030 May 1998
Summary of ACRS Advanced Reactor Designs Subcommittee 980513-15 Meeting in Rockville,Md Re Review of AP600 Standard SAR & Associated Advanced FSER Chapters 3,6,14,16 & 17
ML20206S2521998-05-14014 May 1998
Summary of 451st ACRS Meeting on 980402-04 in Rockville,Md Re Items Listed in Attached Agenda
ML20206S2801998-04-30030 April 1998
Summary of 980219-20 ACRS Open Meeting of Subcommittee on Reliability & Probabilistic Risk Assessment in Rockville,Md Re Continuance of Review of Proposed Final SRP Sections & RGs for risk-informed,performance-based Regulation
ML20206S2871998-04-15015 April 1998
Summary of 449th ACRS Meeting on 980302-04 in Rockville,Md Re Items Listed in Attached Agenda
ML20206S2951998-04-15015 April 1998
Summary of 450th ACRS Meeting on 980305-07 in Rockville,Md Re Items Listed in Attached Agenda
ML20206S6481998-03-26026 March 1998
Summary of ACRS Fire Protection Subcommittee 980122 Meeting in Rockville,Md Re Staff Actions Taken Related to Development of Revised Fire Protection Rule
ML20196L0531998-03-20020 March 1998
Summary of 980205-07 ACRS 448th Meeting in Rockville,Md Re Items Listed in Attached Agenda
ML20206S6561998-03-13013 March 1998
Summary of ACRS Plant License Renewal Subcommittee 980123 Meeting in Rockville,Md Re License Renewal Implementation Issues & Proposed Industry Guidelines
ML20206S2751998-03-0909 March 1998
Summary of 980218 ACRS Thermal Hydraulic & Severe Accident Phenomena Subcommittee Open Meeting in Rockville,Md Re Review of RES Program in Area of thermal-hydraulic Phenomena in Support of ACRS Planned Rept to Commission
ML20217A8561998-03-0505 March 1998
Summary of ACRS 448th Meeting on 980205-07 Re Several Matters & Completed Listed Rept & Ltr.Committee Authorized Larkins,Executive Director,To Transmit Memoranda Listed
ML20206S6791998-03-0505 March 1998
Summary of ACRS Plant Operations Subcommittee 980203 Meeting in Rockville,Md Re Subcommittee Review of Proposed Improvements to Senior Mgt Meeting Process
ML20206S2651998-02-23023 February 1998
Summary of 980203-04 ACRS Subcommittee on Advanced Reactor Designs Open Meeting in Rockville,Md Re Review of Chapters 1,4,5,7,8,11,13 & 18 of AP600 Ssar & AP600 Test & Analysis Program
IR 07100203/20064471998-02-0505 February 1998
Summary of 971203-06 447th ACRS Meeting in Rockville,Md Re Appropriate Action on Items Listed in Attached Agenda
ML20206S3931998-01-28028 January 1998
Summary of 980121 ACRS Subcommitee on Human Factors Meeting in Rockville,Md Re Human Performance & Reliability Plan & Integration of Human Factors Insights Into Insp Process
ML20217A8461998-01-0909 January 1998
Summary of ACRS Plant Operations Subcommittee 971202 Meeting in Rockville,Md Re SER on industry-sponsored Utility Resolution Guidance for Issue of Blockage of Suction Strainer in ECCS
ML20217A9911997-12-23023 December 1997
Summary of 447th ACRS Meeting on 971203-06 Re Proposed Revs to 10CFR50.59 & Credit for Containment Overpressure to Provide Assurance of Sufficient Net Positive Suction Head for ECC & Containment Heat Removal
ML20217A8771997-12-23023 December 1997
Summary of ACRS 446th Meeting on 971106-07 in Rockville,Md Re Appropriate Action on Items Listed on Attached Agenda.W/ Fr Notice,Meeting Schedule & Outline,List of Attendees, Future Agenda & List of Documents Provided to Committee
ML20217A8651997-12-15015 December 1997
Summary of ACRS Reliability & PRA Subcommittee 971112-13 Meeting in Rockville,Md Re Proposed Final SRP for risk- informed,performance-based Regulation
ML20217A9831997-12-0303 December 1997
Summary of 971203 ACRS Meeting in Rockville,Md Re Conduct of ACRS Business
ML20217A5901997-12-0202 December 1997
Summary of 445th ACRS Meeting in Rockville,Md on 971002-03 Re Appropriate Action on Items Listed in Attached Agenda.W/ Fr Notice,Meeting Schedule & Outline,List of Attendees, Future Agenda & List of Documents Provided to Committee
ML20217A8811997-12-0202 December 1997
Summary of ACRS Safety Research Program Subcommittee 971104- 05 Meeting in Rockville,Md Re NRC Safety Research Program & Draft Annual Rept to Congress
ML20217A8221997-12-0101 December 1997
Informs That During 446th Meeting on 971106-07,ACRS Discussed Several Matters & Completed Listed Ltr. Committee Authorized Larkins,Executive Director,To Transmit Memoranda,Listed
ML20217A8901997-11-21021 November 1997
Summary of ACRS Reliability & PRA 971021-22 Meeting in Rockville,Md Re Matter Included in Staff Requirements Memo,
ML20217B1921997-11-0606 November 1997
Summary of ACRS Planning & Procedures Subcommitte 971105 Meeting in Rockville,Md Re Conduct of ACRS Business
ML20217A6381997-11-0606 November 1997
Summary of 444th ACRS Meeting in Rockville,Md on 970903-05 Re Appropriate Action Items Listed in Attached Agenda.W/Fr Notice,Meeting Schedule & Outline,List of Attendees & List of Documents Provided to Committee
ML20217A8001997-10-21021 October 1997
Informs That During 445th Meeting on 971002-03,ACRS Discussed Several Matters & Completed Listed Rept & Ltrs. Committee Authorized Larkins,Executive Director,To Transmit Memoranda Listed
ML20217A7011997-10-20020 October 1997
Summary of 970826-27 ACRS Subcommittees on Matls & Metallurgy & on Severe Accidents Joint Meeting W/Nrc,Nei & Industry in Rockville,Md Re Review of Proposed Draft GL & Associated Draft Regulatory Guide Re SG Tube Integrity
ML20217A8141997-10-20020 October 1997
Summary of ACRS Subcommittees on Pra,Plant Operations & Fire Protection 970828-29 Meeting in Rockville,Md Re Review of Staff Requirements Memo
1999-02-17
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atgepr DATE ISSUED: 2/10/87 a/n 7 PROPOSED MEETING
SUMMARY
FOR THE ADVANCED REACTOR DESIGNS SUBCOMMITTEE MEETING ON FEBRUARY 4, 1987 - WASHINGTON, DC PURPOSE:
The Subcommittee on Advanced Reactor Designs met on February 4, 1987 in Washington, DC, to review DOE advanced non-LWR designs regarding the use of proven technology and standardization. In addition, the Subcommittee discussed a draft Commission paper prepared by the NRC Staff regarding standardization of advanced reactor designs.
ATTENDEES:
ACRS NRC STAFF M. Carbon, Chairman R. Colman J. Ebersole, Member T. King C. Mark, Member P. Williams P. Shewmon, Member C. Siess, Member OTHERS M. El-Zeftawy, ACRS Staff A. Tabatabai, ACRS Fellow W. Bickford V. Boyer GA TECHNOLOGIES J. Cunliffe G. Davis A. Neylan D. Mears A. Mullinzi GENERAL ELECTRIC J. Recknagel J. Scarborough N. Brown K. Unnerstall B. Genetts ROCKWELL INT.
J. Brunings R. Lancet MEETING HIGHLIGHTS, AGREEMENTS, AND REQUESTS:
- 1. Dr. Carbon, Subcommittee Chainnan, introduced the members of the Subcommittee and stated the purpose of the meeting. He indicated that DOE and its subcontractors are currently developing the DESIGNATED ORIGINAL 33 PDR Coctified Dy ( d d[
2488
a Advanced Reactor Designs Minutes February 4, 1987 designs of three advanced reactors (one gas-cooled, and two liquid metals). In early 1984, DOE requested the NRC to provide guidance to the designers, early in the design, prior to any formal appli-cation regarding the requirements for the licensability of the design. The ACRS agreed and urged for such early interaction. The
- NRC agreed and a schedule for collaborative effort was established.
Preliminary Safety Information Documents (PSIDs) have been prepared by DOE and its contractors on each of the three designs and were submitted to the NRC for review in September (HTGR) and November 1986(LMRs). The PSID is basically a description of the conceptual design, including proposed licensing criteria and safety analysis to illustrate plant response to accident conditions. PRA and a description of the supporting R&D programs are also to be provided for each of the designs. The output of the NRC Staff's review of 4
the PSIDs would be a Safety Evaluation Report (SER) on each con-cept, giving guidance on the licensing criteria to be applied and the potential of the designs to meet those criteria.
The NRC Staff intends to prepare and submit Commission papers on three topics: (a) Standardization,(b)SevereAccidents,and(c)
Containment.
In accordance with the Commissioner's advanced reactor policy statement, which states that the ACRS should be involved early in the review, the Staff has asked the Subcommittee to review and l
provide comments on each of the three topics. In addition, the i
Staff has asked the Subcommittee to review other key issues (e.g.:
! non-safety grade control room, control of multimodular plants, and proposed use of metal fuel instead of oxide fuel for LMRs).
l Dr. Carbon emphasized that there are large differences between the new conceptual designs being presented and conventional LWRs, and
. r-Advanced Reactor Designs Minutes February 4, 1987 urged the Subcommittee to be prepared to accept the absence of some of the safety features present on other designs.
J II. Mr. Tom King, Section Leader-Safety Program Evaluation Branch /NRR, described the three DOE sponsored advanced reactor concepts. All three advanced reactor programs have as their objective the devel-
- opment of a standardized plant design which would be submitted to the NRC for design certification and approval. Mr. King indicated
- that the advanced reactors designs have unique characteristics, j namely
4
- The concentration of safety functions in the nuclear island of the plant, 3
- The use of modular reactor designs, including extensive shop fabrication of the modules, and provisions for staggered on-site module installation and operation, 4
- Less operating experience to support the designs as compared to LWR, and
- Varying degrees of design detail planned for submittal for certification.
Due to these unique characteristics, the NRC Staff is raising
. several key issues regarding what the NRC should require in order to certify a new reactor concept. These issues can be stated in the form of questions as follows:
(1) What plant systems, structures and components should be reviewed to be able to certify the design?
a
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Advanced Reactor Designs Minutes February 4, 1987 (2) What level of design detail should be provided for review on those systems, structures and components provided for certi-fication?
(3) What level of operating experience, existing technology and supporting R&D is required to support certification (i.e., is a prototype plant required to be built and operated prior to designcertification)?
(4) What information should be provided to allow flexibility in the design certification for variations in plant size (i.e.,
number of modules)?
(5) Is a manufacturing license (10 CFR 50, Appendix M) required prior to shop fabrication of reactor modules.
III. Mr. A. Neylan, GA Technologies, described the Modular High-Temperature Gas Cooled Reactor (MHTGR). The MHTGR plant design is a 350 Mwt standard reactor module being developed in conjunction with Gas Cooled Reactor Associates, Stone & Webster and Bechtel.
In March 1985, the side-by-side steel vessel concept was selected as the reference design to be further developed. In August 1985, the concept using prismatic fuel was selected for further develop-ment. The prismatic fuel option uses four modular, steel vessel reactors in the side-by-side configuration, each operating at a power level of 350 Mwt and supplying steam to two turbine genera-tors. The net plant electrical output is 558 Mwe. Each reactor module is housed in a vertical cylindrical concrete enclosure that is fully embedded and below grade. The nuclear island portion consists of four reactor enclosures and adjacent structures that house fuel handling, helium processing, and other essential reactor service systems. A common control room is used to operate all four reactors and the turbine plant. The design has no containment. A
5 Advanced Reactor Designs Minutes February 4, 1987 confinement system is used in the design. The design-utilizes active systems for nomal decay heat removal and reactor shutdown.
Passive means are provided as backup for accomplishing these functions.
Mr. A. Mullinzi, (DOE), stated that the major items that DOE considers safety related are the reactor vessel and its internals, the associated primary pressure boundary and the passive decay heat removal system. It is currently DOE's intent to request Design Certification on the entire Nuclear Island (which includes all safety related systems) and some of its key interfacing systems.
The remaining systems (which includes the balance of plant) would then be defined by interface requirements in the application for design certification. DOE's plan is not contingent upon a proto-4 type reactor module or the first commercial plant being built and tested prior to receiving design certification.
1 IV. Mr. N. Brown, General Electric Company, described a 425 ht modular liquid metal reactor called the Power Reactor Inherently Safe Module (PRISM). This concept emphasizes inherent safety charac-teristics and modularity. The reactor modules are a single stan-dard design that would be built in a factory and are shippable by rail as a unit. The plant uses nine PRISM reactors, with each 7
module producing 425 ht power. The plant combined power output is l 1245 Mwe. Each module is a pool type LMFBR design with its own I intermediate heat transport system and steam generator system. The PRISM reactor core is a homogeneous with U/Pu/Zr metal fuel similar I to that used in DOE's EBR-II reactor. The core lattice is being selected to be capable of breeding. The core and fuel design have the capability of mitigating anticipated transients without scram.
, The small size of each reactor module facilitates the use of i
i passive inherent self-shutdown and shutdown heat removal features.
J Advanced Reactor Designs Minutes February 4, 1987 The balance of plant (B0P) is completely disconnected from the primary loop safety considerations.
The containment vessel is located close to the reactor vessel to assure that primary coolant leaks from the reactor vessel do not result in loss of core cooling. GE's intent is to request design certification on only those portions of the plant which are con-sidered safety related. The remaining systems (e.g., control room, steam generator, B0P, etc.) would then be defined by interface requirements. The application for design certification would contain three options: (a) a three module plant, (b) a six module plant, and (c) a nine module plant. GE's overall plan includes construction and testing of a full scale prototype reactor module.
V. Mr. R. Lancet, Rockwell International (RI), described a 900 Mwt liquid metal standard reactor module called the Sodium Advanced Fast Reactor (SAFR). Sodium is the primary coolant, with a pool type primary system and passive decay heat removal. The reactor vessel is located above grade. The fuel is U/Pu/Zr metal similar to that used in DOE's EBR-II reactor. Each module is designed to produce 350 Mwe. RI envisions four 350 Mwe modules per site. Each SAFR module will use a building block approach with discrete increments of power generation called power packs. The SAFR plant will be designed to be comercially competitive with coal and LWR plants by the year 2000 and beyond. RI claims that radioactive releases during accident conditions will be low enough that no offsite evacuation plans are required.
RI's intent is to request design certification on all systems, structures and components (safety related as well as non-safety related)exceptsitespecificitems. The application for design certification would contain four options: (a)aonemodule,(b) two modules, (c) three modules, and (d) four modules plant. RI's l
2-Advanced Reactor Designs Minutes February 4, 1987 overall plan calls for the construction and testing of the first commercial unit prior to receiving design certification.
VI. Mr. T. King, NRR, pointed out that the standardization of advanced reactors poses several unique issues not faced in the standardi-zation of current generation of LWRs. Criteria for resolution of the five issues (described in item II above) are being proposed by the NRC Staff as follows:
Issue #1 - Extent of design to be certified -
- Prefer complete plant be submitted
- Staff could review for certification less than the complete plant provided the following were met:
(1) Sufficient infomation is included in the application to allow completion of a PRA and safety analysis.
(2) Compliance with interface requirements is verifiable through inspection, testing, previous experience or analysis. Reliability verification must be based on previous experience or testing.
(3) Certified portion of the design should include all systems, structures and components important to safety.
(4) Representative design for the non-certified portion of the plant should be provided as an example of how inter-face criteria can be met.
I
Advanced Reactor Designs Minutes February 4, 1987 Issue #2 - Level of design detail to be certified -
- Prefer final design information
- Staff could review for certification less than final design information provided the following were met:
(1) The level of design detail provided is sufficient to allow development and review of a PRA and safety analy-sis.
(2) The level of design detail provided is sufficient to support procurement, construction and operation of systems, structures and components that meet the perfor-mance and reliability characteristics assumed in the PRA and safety analysis.
(3) A representative design for those portions of the plant not finalized is submitted as an example,of how the final design will look to aid in the review of the PRA and safety analysis.
Issue #3 - Prototype Testing -
- A prototype plant should be built and tested prior to design certification by NRC unless the following can be demonstrated:
(1) The performance of each safety feature of the plant has been demonstrated via previous experience or full scale testing.
(2) Sufficient performance data exists on each safety feature of the plant to validate safety analysis analytical tools over a full range of operating and accident conditions, including plant lifetime.
O I
Advanced Reactor Designs Minutes February 4, 1987 (3) System interaction effects among the plant's safety features have been properly accounted for.
Issue #4 - Design Certification Options -
- A design certification with plant module size / options is acceptable provided that these options are described in the application, including: (1) variations in or sharing of common' systems,(2)variationsininterfacerequirements,and (3) variations in system interaction.
- The PRA and safety analysis should assess each of the options, including any restrictions during the construction and startup phase, to ensure safe operation of those modules already on line.
Issue #5 - Manufacturing License (ML) -
- Extent of proposed shop fabrication appears to be equivalent to fabrication of major components, not complete plant.
- ML is not required unless an essentially ready to operate plant is shop fabricated.
Mr. King stated that the ACRS review and feedback on the issues and proposed staff positions associated with advanced reactor stand-ardization are being sought at this time to allow consideration of ACRS comments prior to presenting a recommendation to the Comis-sion. Only verbal feedback is desired. Timing of a recommendation to the Comission is currently under review.
0-4 Advanced Reactor Designs Minutes February 4,1987 VII. As a ' result of the Subcommittee's discussion, the Subcommittee members raised some concerns regarding the following:
- Dr. Mark expressed some concern regarding the use and appli-cability of the same General Design Criteria (GDC) for LWRs on these non-LWRs advanced designs. He advised the Staff not to use the same philosophy as apparently in the case of Fort St.
Vrain.
- Dr. Siess is concerned regarding-the standardization versus the certification issue. He indicated that standardization is basically a business decision that relates to marketing and economics. Certification is not an essential feature of ~
standardization. Theoretically, a standard design could be developed without the intent to have it certified. Certifica-tion is one option offered by NRC.
- Some concern has been raised regarding the extent and depth of certification especially for the MHTGR and PRISM designs.
- Some concern has been raised regarding the need for a proto-type demonstration for the MHTGR design.
- Mr. Ebersole expressed some concern regarding the consid-eration of the steam generator as a non-safety grade system.
- Some concern has been raised regarding certification could limit any design changes that could be made.
- Dr. Siess mentioned that for these new designs and from DOE's and its contractor's presentations, there is a confusion 1
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Advanced Reactor Designs Minutes February 4, 1987 regarding the applicability of important to safety and safety-related.
- Dr. Shewmon likes to know if the PRA will be used to determine what is important to safety. The Staff is currently reviewing the PRA to determine if that is the case.
- Dr. Carbon is concerned regarding the definition and meaning of " interface requirements" for the new designs.
- Dr. Carbon indicated that the NRC Staff should encourage the technology of shop fabrication and installation of core assemblies in the factory, and consequently, there is no need to require manufacturing license (ML).
- Dr. Shewmon questioned the availability of high-temperature fuel codes for the new advanced designs. RI could demonstrate the availability of some codes, but GE and GA Technologies could not. The NRC Staff indicated they had difficulties dealing with the Clinch River reactor fuel codes.
FUTURE ACTIONS:
The Subcomittee Chainnan will brief the full Comittee regarding Subcorr11ttee activities at the February 1987(322nd)ACRSmeeting. The NRC Staff will also present a brief overview of a draft Comission Paper on Standardization.
NOTE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, NW, Washington, DC, or can be purchased from ACE-Federal Reporters, 444 North Capitol Street, Washington, DC 20001 (202) 347-3700.
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