Federal Regulations, Codes, & Standards
Users Group ©
NRC Section XI Report - May 2007
Presented By: Mr. Wally Norris, United States Nuclear Regulatory Commission
1. Amendments to 10 CFR 50.55a
A proposed amendment to 10 CFR 50.55a that would incorporate by reference the 2004 Edition of the ASME Boiler and Pressure Vessel Code and Code for Operation and Maintenance of Nuclear Power Plant Components, was published on April 5, 2007 (72 FR 16731). The public comment period closes on June 19, 2007. The final rule is scheduled for publication in November 2007.
2. ASME Code Cases - Rulemaking/Regulatory Guides
The following draft final regulatory guides have been transmitted to the cognizant NRC offices and legal staff for review:
· Revision 34 to Regulatory Guide 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III
· Revision 15 to Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1
The final guides are scheduled to be published in September 2007.
No public comments were received on Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use,” Revision 2. No changes to this guide are required as a result of the public comments received on Regulatory Guides 1.84 and 1.147. Accordingly, the guide has been formatted for publication as a final guide and is currently being tech edited. This guide is also scheduled to be published as a final guide in September 2007.
The staff has initiated proposed Revision 35 to Regulatory Guide 1.84 and proposed Revision 16 to Regulatory Guide 1.147. The guides may include Code Cases through Supplement 12 to the 2004 Edition.
3. Risk-Informed Activities
Regulatory Guide1.200, “An approach for Determining the Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities”
A Federal Register Notice of availability of Revision 1 of
Regulatory Guide (RG) 1.200, “An Approach for Determining the Technical
Adequacy of Probabilistic Risk Assessment Results for Risk-Informed
Activities,” was published on February 8, 2007 (ADAMS No. ML070240001). The RG describes one acceptable approach
for determining whether the quality of a probabilistic risk assessment (PRA),
in total or the parts that are used to support an application, is sufficient
to provide confidence in the results, such that the PRA can be used in
regulatory decision‑making for light-water reactors. NRC Issued Regulatory Issue Summary (RIS)
2007-06 on March 22, 2007 (
10 CFR 50.69 - Risk Informed Special Treatment Requirements
A Federal Register Notice of the availability of RG 1.201, “Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance,” was published on January 27, 2006 (ADAMS No. ML060260361). Based on a public comment, Revision 1 of Regulatory Guide 1.201 was issued for trial use in May 2006 (ML061090627). In September 2006, the PWR owners group submitted, WCAP-16308-NP Revision 0 Pressurized Water Reactors Owners Group 10 CFR 50.69 Pilot Program - Categorization Process - Wolf Creek Generating Station. The topical is intended, in part, to simplify 50.69 applications by providing a template for the contents of the categorization process results and descriptions that should be included in a license amendment request to implement 50.69. The NRC has begun the review of this topical, including a determination of the extent that a standard template format can be developed and endorsed (e.g., quality of the PRA is not addressed).
10 CFR 50.46a - Option 3 Rulemaking (Risk-Informed Emergency Core Cooling System
The Advisory Committee on Reactor Safeguards (ACRS) issued a letter on November 16, 2006, recommending that the rule not be issued in its current form. The letter included three general recommendations: (1) the Rule to risk‑inform 10 CFR 50.46 should not be issued in its current form. It should be revised to strengthen the assurance of defense in depth for breaks beyond the transition break size (TBS); (2) the revision of draft NUREG-1829, "Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process," to include changes resulting from the resolution of public comments, should be completed before the revised Rule is issued; (3) the interpretation that the Rule limits the total increase in core damage frequency (CDF) resulting from all changes in a plant to be "small" (i.e., <10-5/yr) represents a significant departure from the current guidance for risk-informed regulation and should be reviewed for its implications. NRC staff is developing a SECY paper for the Commission providing its recommendation on how to proceed with the rulemaking.
Reactor Vessel Weld Inspection
The Topical report WCAP-16168-NP Rev 1, “Risk-informed Extension of the Reactor Vessel In-Service Inspection Interval,” requesting an extension of the weld inspection interval from 10 to 20 years is under review. A meeting at NRC to discuss draft requests for information is planned for the last week of May or first week of June 2007. The topical report relies extensively on work described in NUREG-1874, “Recommended Screening Limits for Pressurized Thermal Shock (PTS)” which the NRC intends to publish in the near future (ADAMs No. ML070740639).
NRC has approved several requests to extend the inspection interval on reactor vessels welds from 10 years to 10 years plus one operating cycle based on consistency with the letter from NRC to Westinghouse Electric Company, "Summary of Teleconference with the Westinghouse Owners Group Regarding Potential One Cycle Relief of Reactor Pressure Vessel Shell Weld Inspections at Pressurized Water Reactors Related to WCAP-16168-NP, "Risk Informed Extension of Reactor Vessel In-Service Inspection Intervals," dated January 27, 2005. NRC is finalizing the review of one request to extend the inspection interval an additional operating cycle (i.e., for a total extension of two operating cycles). A second such request has been received and is also under review.
Repair and Replacement
In September 2006, the PWR owners group submitted, WCAP-16308-NP Revision 0 Pressurized Water Reactors Owners Group 10 CFR 50.69 Pilot Program - Categorization Process - Wolf Creek Generating Station. The Topical includes, in part, an alternative methodology to the NRC endorsed Code case N-660 for categorization of passive components. The NRC is reviewing this Topical.
A licensee submitted a relief request under 50.55a(3)(I) to authorize the use of a risk-informed safety classification and treatment for repair/replacement activities in Class 2 and Class 3 moderate energy systems. A conference call was held with the licensee about this relief request on May 7, 2007. The NRC is determining whether the relief request includes sufficient information for the staff to begin its review.
On May 7, 2007, a meeting was held at NRC headquarters to discuss the draft responses from D.C. Cook and Grand Gulf Nuclear Station to the NRC Staff Requests for Additional Information on the application for the use of the methodology for applying and implementing the Risk-Informed Inservice Inspection Program in Code Case N-716. During the meeting, the possibility of identifying sufficient technical quality in Licensees’ flooding analyses based on the results of the review of the flooding analysis against the ASME RAS-S-2000 standard (currently Addendum B has been issued) was discussed. If it is practicable to define acceptable technical adequacy in terms of capability categories for each of the specific supporting requirements (SR's) that characterize a licensee's flooding analysis, it would simplify any NRC endorsement of a RI-ISI Code Case for use without prior review (or periodic relief request). The feasibility of this proposed course of action is under discussion.
During the review of the periodic, 10-year updates of the RI-ISI program, the NRC must develop confidence that the living program requirements are being appropriately implemented using a current PRA of technical adequacy. The potential impact of the recently issued RG 1.200 on PRA quality on RI-ISI relief requests is under discussion.
4. Generic Activities on Material Degradation/PWR Alloy 600/182/82 PWSCC
Representatives from the Nuclear Energy Institute (NEI),
Electric Power Research Institute (EPRI), and the EPRI Materials Reliability Project
(MRP) met with the staff on October 25, 2006, to discuss five circumferential
indications identified in three dissimilar metal (DM) welds on the
pressurizer at the Wolf Creek Generating Station. The inspection results raised safety
concerns based on the size and location of the indications. These findings also raised concerns
regarding the adequacy of the MRP-139, “Materials Reliability Program:
Primary System Piping Butt Weld Inspection and Evaluation Guideline,”
baseline inspection schedule for pressurizer welds, particularly the deferral
of the baseline inspections allowed by the industry’s NEI 03-08, “Guideline
for the Management of Materials Issues,” protocol. Three of the
It was the staffs desire to have these welds inspected by the end of 2007. Through efforts of the Nuclear Energy Institute, all PWR licensees submitted voluntary commitment letters pertaining to plans for inspection and mitigation of Alloy 82/182 welds in pressurizer locations. Plants generally fit into one of three categories, those that do not have Alloy 82/182 material in the pressurizer butt welds or had already mitigated the welds, those that planned on inspection activities by the end of 2007, and those that intended to inspect after 2007. The staff engaged each licensee that has Alloy 82/182 welds in pressurizer locations and discussed the requirements for inspection, mitigation, and contingency actions until inspections were completed. Based on these discussions and revised commitment letters, confirmatory action letters (CALs) were issued to all PWRs having alloy 82/182 weld material in pressurizer surge, spray, safety, and relief nozzle and safe end butt welds that have not mitigated the welds. These CALs were issued between March 12 and March 29.
Plants that did not receive a CAL will be receiving a letter acknowledging the information that they provided and a statement that their response adequately addresses the concerns that the NRC has at this time regarding primary water stress corrosion cracking (PWSCC) susceptibility of these welds. The first of these letters was sent April 18, and all letters should be issued by mid June.
The majority of licensees committed to inspect the welds by the end of 2007. Nine licensees intend to perform inspections in 2008, but have committed to perform the inspections by the end of 2007 or shut down by the end of 2007 if an adequate level of safety, supporting examination after 2007, has not been demonstrated to the NRC. The industry is currently working on an advanced finite element analyses to demonstrate this adequate level of safety. The NRC is monitoring the industry’s progress to be aware of the inputs and assumptions and perform confirmatory analyses and sensitivity analyses. Results of the industry analyses are expected in July.
5. New Reactor Licensing Activities
The New Reactor Licensing public web-site [http://www.nrc.gov/reactors/new-reactor-licensing.html] has a list of expected new nuclear power plant applications updated April 27, 2007, and an estimated schedule by fiscal year for new reactor licensing applications.
Meetings between NRC staff and Electric Power Research
Institute (EPRI) staff were held at the EPRI Office in
7. NUREG/CR-6923, “Expert Panel Report on Proactive Materials Degradation Assessment”
On March 16, 2007, the report was posted to the NRC public website. As part of NRC’s Proactive Materials Degradation Assessment (PMDA) program, Brookhaven National Laboratory used the knowledge of an international panel of eight experts and the Phenomena Identification and Ranking Table (PIRT) process to identify materials and components where future degradation may occur in specific light water reactor (LWR) systems. The panel was given information on the materials, fabrication process, and operational environment for hundreds of different boiling-water reactor and pressurized-water reactor components. Prediction methodologies for various phenomena, calibrated by past component failures in the global LWR fleet, aided in judgments of future degradation behavior. Also considered were the successes and limitations of mitigation/control approaches that have been used to date. This report includes the panel’s scoring, rationale, conclusions, as well as considerable documentation of issues relevant to materials degradation.
8. Memorandum of Understanding Signed by the NRC and the Electric Power Research Institute (EPRI)
On March 14, 2007, the Director of the NRC’s Office of Nuclear Regulatory Research and the Vice President and Chief Nuclear Officer of EPRI signed a Memorandum of Understanding (MOU) between the NRC and EPRI on cooperative nuclear safety research. This MOU (reference ADAMS accession number ML070740114) describes the parameters within which cooperative research programs between the NRC and EPRI will be considered and conducted, provides for the avoidance of unnecessary duplication of research efforts, and allows for the conservation of resources. Individual cooperative research programs are described in addenda to the MOU. This MOU supersedes and replaces the NRC-EPRI MOU dated November 25, 1997, and expires on September 30, 2016.
9. Summary of Public Meeting with Nuclear Energy Institute Regarding 10 CFR 50.55a
On March 7, 2007, the Nuclear Energy Institute (NEI) sent to the NRC for information, its white paper on “Improving the Effectiveness of 10 CFR 50.55a, Code and Standards,” (ADAMS Accession ML070940157). The white paper formed the basis for discussions held on April 3, 2007, at NRC headquarters between the staff and representatives from NEI, ASME, and utilities. The following options are discussed in the white paper: (1) revise 10 CFR 50.55a to require that the “Code of Record” be maintained in the Updated Final Safety Analysis Report by 10 CFR 50.71 and controlled by 10 CFR 50.59, (2) modify 10 CFR 50.55a to specify screening criteria to permit changes to the Code of Record without prior NRC approval, and (3) revise 10 CFR 50.55a to specify integrated risk-informed evaluation criteria to permit changes to risk-informed inservice inspection, inservice testing, and repair/replacement programs without prior NRC approval.
The majority of the discussion centered around the incorporation by reference of the ASME Code in 10 CFR 50.55a and the process of evaluating the acceptability of changes or alternatives to ASME Code requirements. In the meeting, the NRC staff asked for clarifications with respect to the proposed options and provided preliminary concerns and observations regarding the proposed revision. The NRC staff stated that it would respond to NEI formally regarding the proposed revision in the near future. NEI indicated that it plans to submit a petition for rulemaking in the Fall 2007, contingent upon favorable NRC response to the proposed revision.
A list of meeting attendees and the NEI’s presentation slides are available in ADAMS (ML071080146).