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NRC Section XI Report - December 2004

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


NRC Report
Section XI
December 2004

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 schedule for the 2004 Edition is not yet available.

 

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 staff has developed responses to the comments, and the draft final guide package is currently under review.  Publication is anticipated in the Spring 2005.

 

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 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.

 

3. Risk-Informed Activities

 

Plan for Phased Approach to PRA Quality Drafted

 

The objective of the phased approach to stabilizing the PRA quality expectations and requirements is to achieve an appropriate level of PRA quality for NRC’s risk‑informed regulatory decision making.  That is, the phased approach defines the needed PRA quality for all envisioned applications and the process for achieving this quality while the necessary guidance documents defining the PRA quality are developed and implemented.  It is expected that meeting the phased approach objective will result in the following:

 

·           Industry movement towards improved and more complete PRAs

·           Increased efficiencies in the staff’s review of risk‑informed applications

·           Clarification of expectations for 10CFR50.46 and 10CFR50.69 rulemakings

·           Continued near‑term progress in enhancing safety through the use of available risk‑informed methods while striving for increased effectiveness and efficiency in the longer term.

 

The plan describes the phased approach and what activities, on the part of both NRC and industry, are needed to achieve the program objectives.  In addition, the action plan discusses the resolution of the following technical issues: model uncertainty; treatment of seismic and other external events; and human performance issues.

 

The plan was issued to the Commission on July 13, 2004, and approved in an SRM dated October 6, 2004.  The staff has initiated the activities necessary to implement the plan.

 

The NRC website contains information at

http://www.nrc.gov/what-we-do/regulatory/rulemaking/risk-informed/reactor-safety.html.

relative to risk‑informed activities in the Reactor Safety Arena.

 

Probabilistic Risk Assessment (PRA) Standard

 

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, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities” 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 staff expects to publish Appendix C to Regulatory Guide 1.200 in final form in the June to August 2005 time frame.

 

Risk‑informing Special Treatment Requirements of 10 CFR Part 50 ‑ Option 2

 

Final 10 CFR 50.69, "Risk‑Informed Treatment of Structures, Systems, and Components" was approved by the Commission on October 7, 2004, subject to the changes denoted during the affirmation session and documented in the SRM (ADAMS ML042810516).  The staff revised the final rule package per the Commission direction, and the final rule was published in the federal register on November 22, 2004 (69 FR 68008).

 

The final rule can be accessed at http://ruleforum.llnl.gov/cgi-bin/rulelist?type=final.

 

NEI submitted revision 2 of final draft NEI 00-04 (industry implementation guidance for § 50.69) on October 29,  2004.  The staff is currently reviewing this latest version of NEI 00-04 and plans to meet with NEI in mid-December 2004 to discuss the revised guidance.  The objective is to remove, where appropriate, clarifications and exceptions from RG 1.201 (the NRC staff’s RG that endorses the NEI guidance) that are no longer needed.

 

Risk‑informing 10 CFR 50.46, ECCS ‑ Option 3

 

The staff has prepared a proposed rulemaking to add a new section to 10 CFR Part 50 to provide an alternative, risk-informed set of requirements for emergency core cooling systems.  The Commission SRMs of March 31, 2003, on SECY-02-0057, and July 1, 2004, on SECY-04-0037, provided guidance to the staff on the content of the alternative risk-informed ECCS rule.

 

The proposed rule would divide the current spectrum of loss of coolant accident (LOCA) break sizes into two regions.  The division between the two regions is delineated by a “transition” break size (TBS).  The first region includes small breaks up to and including the TBS.  The second region includes breaks larger than the TBS up to and including the double ended guillotine break (DEGB) of the largest reactor coolant system pipe.  The proposed rule could be voluntarily adopted by light-water reactor licensees.  The TBS could be used, for certain purposes, in lieu of  the double-ended rupture of the largest pipe in the reactor coolant system (DEGB).   Licensees would still need to demonstrate the plant’s ability to mitigate breaks up to the DEGB.

 

Pipe breaks in the smaller break size region are considered much more likely than pipe breaks in the larger break size region.  Consequently, each region will be subject to ECCS requirements commensurate with the relative likelihood of breaks in each region.  LOCAs in the smaller break size region will continue to be “design-basis accidents” and will continue to be analyzed by current methods, assumptions, and criteria.  In the design-basis region, licensees must perform analyses under current ECCS requirements to determine the limiting size and location for breaks up to and including the TBS.

 

Pipe breaks larger than the TBS, based on their lower likelihood, can be analyzed by the more realistic and less stringent methods established in the new §50.46a.  Although LOCAs for break sizes larger than the transition break will become “beyond design-basis accidents,” the NRC will include requirements ensuring that licensees maintain the ability to mitigate all LOCAs up to and including the DEGB of the largest reactor coolant system pipe.  Although these breaks would be required to be mitigated, the analysis methods and initial and boundary conditions used may be realistic.  Licensees would be allowed to take credit for sufficiently reliable non-safety-related systems without assuming other independent failures and must show that the core remains amenable to cooling.  The specific metrics for demonstrating "coolable core geometry" are not necessarily limited to a peak cladding temperature of 2200 degrees F and 17 percent local cladding oxidation as required for breaks smaller than the TBS.  Licensees would be able to propose other criteria for assuring coolable core geometry if an adequate technical basis was also provided to support the proposed criteria.

 

The Staff plans to send the proposed risk-informed ECCS rule to the Commission in December 2004.

 

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 convened via  teleconference on May 20, 2004, and met again on September 9, 2004.

 

At the September 9 meeting, NRMCC voted to: endorse a strategy for the development of a single PRA standard for requirements that would include: 1) at-power and low power/shutdown internal events - Level 1 and (LERF) (including internal fire) and external events (including external flooding, seismic, and wind); 2) a level of detail consistent with ASME Standard RA‑S‑2002; 3) requirements in the single standard based on requirements in existing standards (ASME - internal-events - full power PRA and ANS - external - events - full power PRA) or standards now in the final stages of development (ANS - low power shutdown (LPSD) PRA, and ANS - internal fire PRA).  Guidance on how to perform a PRA for the scope covered by the single requirements standard would continue to be provided in separate guides.

 

Also at the September 9 meeting, NRMCC voted to form a Joint Consensus Committee reporting to either the ANS Nuclear Standards Board or the ASME BNCS and proceed with the development of the single PRA standard.  Responsibility for the standard would be assigned to a single society (ASME or ANS) with support, recognition, and revenue-sharing by the other Society.

 

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

 

Preliminary comments were submitted to the ASME Task Group on Alloy 600/82/182 Issues by the NRC staff on October 18, 2004.  The staff is very supportive of the industry’s efforts regarding the development of Code inspection requirements and criteria for upper vessel heads and penetrations but expressed concerns with the format of the first draft of the Code Case.  The first draft closely mirrors an industry topical report.  The staff found that the transition from industry topical report to ASME provisions resulted in a Code Case that was complicated and difficult to follow.  In addition, some fundamental questions were raised  such as: (1) the determination of head temperature?  Industry practice is not consistent, and it appears that the current draft provisions of the Code Case would permit manipulation of this input parameter.  (2) is leakage permitted?  It appears that leakage would be acceptable provided nozzle ejection isn’t likely and structural integrity is maintained.  It also appears that corrective measures would be optional..  (3) the requirements for NDE qualification?  The staff believes that UT performance demonstration requirements for these exams should be stipulated in a new supplement to Appendix VIII (including eddy current).  (4) the basis for some of the reinspection frequencies?  For example, additional credit is granted for surface examinations which does not appear to be warranted given the highly accelerated crack growth rates for the weld materials.

 

5. New Reactor Licensing Activities

 

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.

 

On August 30, 2004, the Office of Nuclear Regulatory Research (RES) and the Office of Nuclear Reactor Regulation (NRR) staff met with Mr. Edward Wallace, Senior General Manager ‑ US Programs, PBMR (Pebble Bed Modular Reactor)  Pty. LTD, Republic of South Africa (RSA).  The purpose of the meeting was to update the staff on the status of the PBMR project, including the PBMR plant design and the company’s plan to:  (1) conduct pre-application interactions with the staff on selected key technical topics beginning in April 2005; and (2) submit the PBMR design for NRC certification in the first quarter of CY 2007.  According to PBMR Ltd, construction on the demonstration PBMR plant at the Koeberg site, RSA, will begin in April 2007 and operation is envisioned to commence in 2010.

 

The staff held a meeting with PBMR Ltd. on November 3, 2004, to discuss PMBRs plan for pre-application interactions, submitting the PBMR design certification application, and US industry involvement in PBMR project activities.  U.S. nuclear utility interests in PBMR are reflected in the PBMR being included as a potential plant design in each of the three early site permit applications; the establishment of a PBMR U.S. Utility Advisory Group with seven member utilities, and nuclear utility interest in the Department of Energy’s Next Generation Nuclear Plant (including the PBMR as a potential design option) with NRC licensing.  Estimates of required NRC staff resources to conduct pre-application reviews were discussed.  The NRC did not budget for the review of the PBMR in FY 2005‑2007, and Commission approval is needed to reprogram resources to accommodate the review.  PBMR Pty. plans to submit a letter to the NRC to formally request the proposed PBMR pre-application interactions.  NRC staff has been participating on ASME working groups regarding graphite, high temperature metals, and ISI of high temperature gas-cooled reactors (HTGRs) in anticipation of these interactions.  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  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.

 

On August 30, 2004, the Office of Nuclear Regulatory Research (RES) and the Office of Nuclear Reactor Regulation (NRR) staff met with Mr. Edward Wallace, Senior General Manager ‑ US Programs, PBMR (Pebble Bed Modular Reactor)  Pty. LTD, Republic of South Africa (RSA).  The purpose of the meeting was to update the staff on the status of the PBMR project, including the PBMR plant design and the company’s plan to:  (1) conduct pre-application interactions with the staff on selected key technical topics beginning in April 2005; and (2) submit the PBMR design for NRC certification in the first quarter of CY 2007.  According to PBMR Ltd, construction on the demonstration PBMR plant at the Koeberg site, RSA, will begin in April 2007 and operation is envisioned to commence in 2010.

 

The staff held a meeting with PBMR Ltd. on November 3, 2004, to discuss PMBRs plan for pre-application interactions, submitting the PBMR design certification application, and US industry involvement in PBMR project activities.  U.S. nuclear utility interests in PBMR are reflected in the PBMR being included as a potential plant design in each of the three early site permit applications; the establishment of a PBMR U.S. Utility Advisory Group with seven member utilities, and nuclear utility interest in the Department of Energy’s Next Generation Nuclear Plant (including the PBMR as a potential design option) with NRC licensing.  Estimates of required NRC staff resources to conduct pre-application reviews were discussed.  The NRC did not budget for the review of the PBMR in FY 2005‑2007, and Commission approval is needed to reprogram resources to accommodate the review.  PBMR Pty. plans to submit a letter to the NRC to formally request the proposed PBMR pre-application interactions.  NRC staff has been participating on ASME working groups regarding graphite, high temperature metals, and ISI of high temperature gas-cooled reactors (HTGRs) in anticipation of these interactions.  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 Mitigating Systems Performance Index (MSPI)

 

On September 15, 2004, RES, NRR, and the Regions held a public meeting on the Mitigating Systems Performance Index (MSPI), to discuss the revised draft Nuclear Energy Institute (NEI) MSPI guidance documents and implementation issues.  The NRC has been working with the industry to develop the MSPI as a potential replacement for safety system unavailability as a performance indicator.  Anticipated near‑term staff activities include the following: an Advisory Committee on Reactor Safeguards briefing on October 7; completion of a NUREG report on the results of the MSPI pilot plant implementation activities by January 2005; and the first workshop on industry‑wide implementation of the MSPI in February 2005.

 

7. Nuclear Safety Research Conference (NSRC)

 

The RES sponsored the Nuclear Safety Research Conference (NSRC) on October 25‑27, 2004.  This year’s conference was attended by over 400 persons, with representation from 17 different countries.  Next year, RES staff will participate in the Regulatory Information Conference (RIC) discussing some of the same research topics that have traditionally been highlighted through the NSRC.  Beginning in 2006, the NRC will integrate both the NSRC and RIC into an entirely new conference designed to focus on the agency’s current and future challenges in nuclear regulation.

 

At the October 25-27, 2004, conference, technical sessions were held on materials aging and degradation research, new reactors, materials fuel research, PRA infrastructure development, radiation protection, codes, and operating experience.  The various sessions highlighted the results of past research, insights from ongoing research activities, and where the NRC will direct its research attention in the future.

 

Guest speakers and panelists included NRC Chairman Nils J. Diaz , Commissioner Jeffrey S. Merrifield, representatives from organizations, industries, government, the research community and public interest groups in the United States and abroad.

 

8. Implementing Agreement between the USNRC and the Japan Nuclear Energy Safety Organization regarding participation in the USNRC Cooperative Project on Non‑Destructive Examination for Primary Water Stress Corrosion Cracking in Nickel‑Base Materials and Dissimilar Metal Welds

 

On November 16, 2004, the Executive Director for Operations signed the international "Implementing Agreement between the U.S. Nuclear Regulatory Commission (NRC) and the Japan Nuclear Energy Safety Organization (JNES) Regarding Participation in the USNRC Cooperative Project on Non‑Destructive Examination for Primary Water Stress Corrosion Cracking in Nickel‑base Materials and Dissimilar Metal Welds."

 

The NRC and JNES will work together to (1) compile data describing the morphology and NDE responses associated with primary water stress corrosion cracking, and (2) assess available and developing techniques for NDE of nuclear power plant components.  The insights gained through this cooperative project will enhance NRC’s understanding of the aging and environmental effects on materials in nuclear power plants, enable us to reduce uncertainty associated with non‑destructive examinations, and improve the quality of standards for in‑service inspections for structural integrity endorsed by 10 CFR 50.55a, "Codes and Standards," and Regulatory Guide 1.147 (Rev. 13) "In‑service Inspection Code Case Acceptability, ASME Section XI, Division 1."  This Implementing Agreement will remain in effect through December 2007.

 

9. Cooperative Irradiation‑Assisted Stress‑Corrosion Cracking Research Group Meeting

 

On November 7‑8, 2004, RES staff participated in a meeting of the Cooperative Irradiation‑Assisted Stress‑Corrosion Cracking (IASCC) Research (CIR) group.  The CIR group meets approximately every 6 ‑ 10 months to review the status of sponsored research and to discuss any adjustments that need to be made in the experimental programs related to the degradation of vessel internals.  Members of the group represent regulatory bodies from the U.S., Sweden, and Spain; research laboratories located in the U.S., France, and Finland; and plant staff from Ringhals/Barsebeck located in Sweden.  The meeting was attended by approximately 30 representatives of vendors, research laboratories and regulatory agencies with a common interest in IASCC.  Formal contributions included presentations of recent laboratory results, testing plans for the future, and a discussion of overall program goals for the 2005‑2007 time frame.  Results from the CIR support an NRR user need.

 

10. NUREG on Reliability of VT Issued

 

Utilities have recently proposed replacing current volumetric and surface examinations (UT) of certain components, as required by the ASME Boiler and Pressure Vessel Code, Section XI, with a simpler visual testing (VT) method to reduce examination times and personnel exposure.  The NRC’s Office of Nuclear Regulatory Research tasked Pacific Northwest National Laboratory to conduct a study of the capabilities of VT.  The results of this study show that VT as it is currently practiced has a low probability of detection of many service induced cracks in nuclear components.  The study determined that current industry guidelines for testing visual systems using two crossed wires as a performance demonstration standard is not reliable, and does not provide the level of accuracy that a combination resolution target test and a reading test chart can provide.  The report provides recommendations for a more reliable calibration standard, including the optimization of lighting angles and intensities for crack detection.  Further experimental work is needed to determine the magnification and resolution needed for camera systems to reliably detect very tight cracks under realistic conditions.  Printed copies will be available on 12/21/04.  The report is available at:

http://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6860/.


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