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NRC Section XI Report - August 2005

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Presented By:  Mr. Wally Norris, United States Nuclear Regulatory Commission

NRC Report
Section XI
August 2005

1.    Amendments to 10 CFR 50.55a


A final amendment to 10 CFR 50.55a incorporating by reference the 2001 Edition up to and including the 2003 Addenda of the ASME Boiler and Pressure Vessel Code, was published on October 1, 2004 (69 FR 58804).  The rule became effective on November 1, 2004.


The staff has completed the technical bases for the amendment to 10 CFR 50.55a to endorse the 2004 Edition.  Development of the proposed rule is proceeding.


2.    ASME Code Cases - Rulemaking/Regulatory Guides


Three draft regulatory guides were published for public comment on August 3, 2004: Proposed Revision 33 to Regulatory Guide 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III [ADAMS Accession Number ML040850299]; Proposed Revision 14 to Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 [ML040850346]; and Proposed Revision 1 to Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use [ML040850236].”  The guides address Code Cases in Supplement 12 to the 1998 Edition through Supplement 6 to the 2001 Edition.  The Proposed Rule incorporating the Section III and Section XI regulatory guides was also published on August 3, 2004 [69 FR 46452].  The “Evaluation of Code Cases,” attached to the proposed rule, discussed the staff’s basis for any proposed conditions to Code Cases.  The public comment period for the regulatory guides closed on September 2, 2004, and on October 18, 2004, for the proposed rule.


The Committee to Review Generic Requirements and the Advisory Committee on Reactor Safeguards approved publication of the final guides.  Publication of the rulemaking/guides has been re-scheduled for the September/October 2005 time frame.


Draft Revision 34 to RG 1.84, Draft Revision 15 to RG 1.147, and Draft Revision 2 to RG 1.193 are nearly complete.  The guides address the Code Cases in Supplement 7 through Supplement 12 to the 2001 Edition.  These guides will be published for public comment shortly after the guides discussed in the above paragraph are published final.


The staff is currently reviewing Supplements 1, 2, and 3 to the 2004 Edition.


3.    Risk-Informed Activities


Plan for Phased Approach to PRA Quality Drafted


In mid‑November 2003, the ANS standard ANSI/ANS-58.21, "American National Standard - External Events in Probabilistic Risk Assessment Methodology," was issued.  The staff developed Appendix C to Regulatory Guide 1.200 to endorse ANSI/ANS-58.21.  The staff issued Appendix C for public comment on August 31, 2004.  The 60-day public comment period expired on October 29, 2004.  On November 9, 2004, the NRC staff conducted a meeting to solicit public comments on proposed staff positions on ANSI/ANS-58.21.  The NRC staff believes that its concerns will be addressed by a revision to the ANS standard scheduled to be published in January 2006.


Risk Management Coordinating Committee (NRMCC)


This Committee, which is jointly chaired by the Vice Chairman of the ASME BNCS and the Chairman of the American Nuclear Society Nuclear Standards Committee (ANS), was formed on February 20, 2004.  Its purpose is to coordinate codes and standards activities associated with risk management for current and new nuclear power plants, nuclear facilities, and the transportation and storage of nuclear material.  In addition to ASME and ANS, NRC, several consultants, the Westinghouse Owners Group, and NEI are represented on the committee.  Since the initial organization meeting of February 20, 2004, NRMCC has met several times both via teleconference and at the ASME Headquarters in Washington D.C.  Most recently NRMCC met on June 2, 2005, at ASME - Washington D.C.


At the June 2, 2005, meeting, the following were discussed and related action items assigned:


1)         Evaluation of NRMCC sponsoring a Potential Workshop on Risk Management Standards.  Some of the significant items that need to be addressed are:

·        Benefits of the PRA standards.

·        How to use the PRA standards (this also needs to be addressed in a prominent place in the standards).

·        Workshop needs to be structured to get the attention of management.


2)         Develop and distribute a schedule, with milestone dates, for all of the ANS and ASME PRA standards under development.


3)         To help communicate the charter and purpose of the NRMCC, ANS and ASME are expected to put the NRMCC Strategic Plan on their Websites.


4)         At the next meeting of the NRMCC in September, the NRMCC Strategic Plan will receive an in depth review and updating.


5)         A separate Task Group will be established that will address the issue of aggregation.  Several members of consensus committees have expressed wide-ranging opinions on how and whether to address aggregation of results from the several contributors to risk.  In this context "contributors" means the individual results of the several PRAs, such as internal events at full power, low power and shutdown, seismic, and external events (separately or together).


The next NRMCC meeting is being scheduled for September 2005 at the ASME Washington D.C. offices.


4.    Generic Activities on PWR Alloy 600/182/82 PWSCC


As discussed by the NRC staff at the Subcommittee on Nuclear Inservice Inspection and the Boiler and Pressure Vessel Committee meetings, NRC staff supported ASME approval of Code Case N-729, “Alternative Examination Requirements for PWR Closure Heads with Partial-Penetration Nozzle Welds.”  The NRC staff’s technical concerns relative to the implementation of the Code Case were resolved.  While a few longer-term issues have yet to be resolved, this is a safety significant issue requiring implementation at the earliest possible date.


Representatives from NEI and MRP met with the NRC staff at NRC headquarters on August 4, 2005, to discuss inspections for PWSCC in ASME Class 1 nickel-based components, other than vessel upper head penetrations and steam generator tubes.  A meeting summary will be developed.  NRC is preparing a regulatory action to address inspections and potential mitigative actions applicable to dissimilar metal butt welds in Class 1 components.  Regulatory action by the NRC staff on the remaining nickel-based alloy components is under consideration.


5.    New Reactor Licensing Activities


Meeting Between the NRC and Pebble Bed Modular Reactor Pty. LTD


On June 30, 2005, staff from RES, NRR, MSS, and NSIR conducted a public meeting with Pebble Bed Modular Reactor Pty. LTD (PBMR Pty.).  The meeting objective was for the staff and PBMR Pty., to begin planning for a potential pre-application review of the Pebble Bed Modular Reactor (PBMR) beginning in FY 2006.  Meeting discussion topics included pre-application review planning objectives, PBMR design and safety principles, PBMR development and testing program, the proposed focus topics for the pre-application review, and identification of potential new licensing policy issues.  PBMR Pty., indicated that it intended to coordinate the PBMR licensing approach, including event selection, defense-in-depth and classification of systems, structures and components (SSCs) with the principles and approach of the technology-neutral risk-informed and performance-based framework being developed by the staff.  The next planning meeting is expected to take place in September 2005.


Meeting with NEI, and EPRI’s Consultants on New Plant Seismic Issues


On June 22 and 23, 2005, staff members from RES and NRR attended a meeting with representatives of the Nuclear Energy Institute (NEI), the Electric Power Research Institute (EPRI), and EPRI’s consultants.  The topic of the meeting was the technical program being conducted by NEI to resolve seismic issues for new nuclear power plants.  An outline of the industry program was presented at a public meeting on May 25, 2005.  At this meeting, industry described the details of its technical program, including the technical basis for two separate efforts to reduce the effects of high frequency (10 to ~ 100 Hertz) ground motion and its influence on the seismic design spectra.


Other Information


The final safety evaluation report (FSER) and Final Design Approval (FDA) for the AP1000 design certification were issued on September 13, 2004.  The staff is scheduled to complete rulemaking to certify the AP1000 design in December 2005.  The staff is reviewing early site permit (ESP) applications for three sites: North Anna, Clinton, and Grand Gulf.  All three applications were accepted for docketing in late 2003, and the staff’s safety and environmental reviews of the applications are in progress.  The reviews are expected to take 23, 25, and 27 months, respectively, followed by a mandatory hearing.


Pre-application reviews for the ESBWR (New Simplified Boiling Water Reactor by General Electric), ACR‑700 (Advanced CANDU Reactor by Atomic Energy of Canada Limited), and IRIS (International Reactor Innovative and Secure by Westinghouse Electric Company) designs will continue.  One topic being addressed in the ACR-700 pre‑application review is unique features that may impact ASME Code Sections III and XI.  The design uses materials and fabrication techniques recognized by the Canadian Standards Association but not by the ASME code.  Design certification review of the ESBWR and ACR‑700 are expected to start in 2005 and 2006, respectively.  A design certification application for IRIS is also possible in 2006.


Documents related to future licensing activities can be found on the NRC web site at, http://www.nrc.gov:201/NRC/GENACT/GC/index.html .


6.    Public Meeting on the Mitigating Systems Performance Index (MSPI)


On July 20, 2005, staff from the Offices of Nuclear Regulatory Research (RES) and Nuclear Reactor Regulation, and the Regions participated in a public meeting on the implementation of the Mitigating Systems Performance Index (MSPI).  The purpose of the meeting was to discuss the results and follow-up actions from the June 20-21, 2005, MSPI Workshop.  Additionally, industry representatives discussed the status of proposed alternatives to the requirements for probabilistic risk analysis (PRA) quality issues related to MSPI implementation.  RES staff presented preliminary results from its confirmatory analysis of the industry’s PRA cross‑comparison effort, as well as comparisons between the NRC’s Standardized Plant Analysis Risk model results and industry MSPI PRA data.  The NRC has also developed an MSPI database to assist the staff in the identification of PRA model "outliers" to support staff inspection activities for MSPI implementation.  The MSPI is a more risk-informed performance indicator than the NRC’s current Safety System Unavailability performance indicator used as part of the Agency’s Reactor Oversight Process.  Industry-wide implementation of the MSPI is scheduled for January 2006.


Publication of NUREG‑1816, "Independent Verification of the Mitigating Systems Performance Index (MSPI) Results for the Pilot Plants"


On March 21, 2005, the Office of Nuclear Regulatory Research published NUREG-1816, "Independent Verification of the Mitigating Systems Performance Index (MSPI) Results for the Pilot Plants."  An electronic version of this NUREG was previously made publicly available through ADAMS in February of this year.  The primary goal of the research described in this report was to provide independent verification of the results of the MSPI pilot program.  On the basis of this research, the staff concluded that, in general, the MSPI is capable of differentiating risk‑significant changes in plant performance and addresses some problems associated with the safety system unavailability performance indicators currently used in the reactor oversight process.  This research activity supported the development of the MSPI and ongoing staff efforts to implement the MSPI on an industry-wide basis.


7.    Meeting of the Proactive Materials Degradation Assessment Expert Panel


On July 11-14, 2005, RES staff and contractors from Brookhaven National Laboratory (BNL), held the last formal meeting of the Proactive Materials Degradation Assessment (PMDA) Phenomena Identification and Ranking Table (PIRT) Expert Panel.  The purpose of the PMDA PIRT is to identify locations and components in light-water reactors (LWRs) that could be susceptible to future materials degradation phenomena.  This information will provide a basis for the development of proactive regulatory and industry actions to avoid, mitigate, and/or manage potential future degradation of reactor components.  The Panel is comprised of eight internationally recognized experts in materials degradation phenomena of components in nuclear and non-nuclear applications.  The experts are from industry, regulatory agencies, and academic institutions in the U.S., France, Sweden, Japan, and Canada.


During the meeting, the panel members reviewed and discussed degradation ranking for the remaining boiling-water reactor (BWR) plant components.  The panel used methodology developed in prior meetings to examine individual system components and assign values to these components for susceptibility to various degradation mechanisms, confidence in the susceptibility calls, and knowledge level of the degradation mechanisms.  In addition, the panel revised sections of the previous pressurized‑water reactor draft report to address comments from the RES program manager, and other domestic and international reviewers.  Assignments for the panel members and BNL staff were made for evaluating and discussing all BWR results.  A final draft report which combines the results from the BWR and pressurized‑water reactor systems will be developed by November 2005.  Results from this activity will be used to guide research and regulatory actions to address proactive assessment and management of materials degradation in operating reactors.


8.    International Atomic Energy Agency (IAEA) Meeting on Deployment of On‑Line Monitoring Systems in Nuclear Power Plants (NPPs)


On June 27-28, 2005, staff participated in a technical meeting to discuss deployment of on‑line monitoring (OLM) technologies at nuclear power plants.  The meeting was hosted by the IAEA, and provided a forum for presentations and discussions of experience with the development, validation, and implementation of OLM techniques for instrument calibration verification, equipment condition monitoring, and diagnostics of equipment and process anomalies.  Both current trends and potential for future activities were discussed with focus on implementation of OLM in current generation of reactors and its potential for the next generation of reactors.  This meeting was attended by representatives from 21 countries and provided insights into how this technology is being implemented at NPPs around the world.  The information provided at this meeting will be used to develop an IAEA TECDOC entitled "On-line Monitoring of Sensor Performance in Nuclear Power Plants", which provide implementation guidance for the use of on‑line monitoring techniques to extend the calibration intervals of instrument channels in nuclear power plants.  Also, the information obtained at this meeting will be helpful to NRC in completing on-going research in this area, and formulating review criteria for licensing OLM Systems in the US.

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