Federal Regulations, Codes, & Standards
Users Group ©
NRC Section XI Report - February 2007
Presented By: Mr. Wally Norris, United States Nuclear Regulatory Commission
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.
A proposed rulemaking to incorporate the 2004 Edition by reference is scheduled to be published in April/May 2007.
2. ASME Code Cases - Rulemaking/Regulatory Guides
The current final regulatory guides addressing ASME Code Cases are:
· Revision 33 to Regulatory Guide 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III”
· Revision 14 to Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1”
· Revision 1 to Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use.”
The guides address the Code Cases in Supplement 12 to the 1998 Edition through Supplement 6 to the 2001 Edition. The guides are available electronically in the Regulatory Guides Document Collection of the NRC's public Web site at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/power-reactors/active/index.html
The public comment period for Draft Revision 34 to RG 1.84, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III and Draft Revision 15 to Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, closed on January 2, 2007. Nine comment letters were received and are presently being reviewed. The guides are scheduled to be published final in August 2007.
The staff has completed its review of Supplements 2 through 8 to the 2004 Edition and has begun its review of Supplements 9 and 10.
3. Joint NRC/ASME Public Meeting
On January 10-11, 2007, a public meeting was held at NRC headquarters to discuss ASME Code issues specific to repair/replacement activities and to discuss a number of regulatory issues. The purpose of the meeting was to gain an understanding from the regulatory and industry perspectives so that an appropriate path forward could be developed. Issues discussed included: Code Case N-713 and the use of Section XI NDE acceptance criteria in lieu of those of the Construction Code for Repair/Replacement Activities (including the RG 1.147 restriction to Code Case 638-1); IWA-4461.4.2 and NRC concerns with thermal cutting without mechanical metal removal; IWA-4540(a) and NDE requirements regarding Pressure Testing; a framework of proposed modifications to Section XI adoption of Section IX temperbead qualification rules; elimination of the 48 hour hold for SMAW temper bead welding; and Code Case N-638-3 and the start of the 48 hour hold after layer 3. In addition, there were discussions regarding: NRC committee participation; timeliness of ASME Code and Code Case endorsement by the NRC; and the status of the envisioned rewrite of 10 CFR 50.55a.
Closing remarks by the participants were very positive relative to the outcomes of the meetings. It was agreed that a meeting should be routinely scheduled on a yearly basis.
4. Risk-Informed Activities
10 CFR 50.69 - Risk Informed Special Treatment Requirements
A Federal Register Notice of the availability of Regulatory Guide 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).
10 CFR 50.46a - Option 3 Rulemaking (Risk-Informed Emergency Core Cooling System
The 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 working to respond to these recommendations and other comments in the letter.
Reactor Vessel Weld Inspection
Several plants have requested one cycle extensions to the 10-year reactor vessel weld inspection interval. The submittals referred to Code case N-691. The NRC has not endorsed Code case N-691. NRC has approved these submittals 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.
A plant has submitted a request to extend the inspection interval an additional, second cycle. The NRC is reviewing this request.
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 in under review.
Repair and Replacement
A plant submitted a relief request under 50.55a(3)(i) to apply draft Code Case N-752 to develop alternative repair and replacement requirements for certain piping systems. The NRC is still determining whether to accept this request for review.
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 an alternative methodology to the NRC endorsed Code case N-660 for categorization of passive components. The NRC is reviewing this Topical.
Two plants have submitted relief request under 50.55a(3)(i) to incorporate a RI‑ISI program developed according to Code Case N‑716. The NRC is reviewing the requests.
During the review of the periodic,10-year updates of the risk-informed ISI program, the NRC must develop confidence that the living program requirements are being appropriately implemented using a current PRA of technical adequacy. Request for information have been increasing in response to limited information being included in the 10 year update relief requests.
5. Generic Activities on Material Degradation/PWR Alloy 600/182/82 PWSCC
On May 5, 2006, NRC staff met with representatives from NEI and the Electric Power Research Institute (EPRI) Materials Reliability Project (MRP) to discuss NRC staff comments on the industry guidance for the volumetric and visual inspection of dissimilar metal butt welds in pressurized water reactor (PWR) primary systems (MRP‑139). MRP‑139, was approved unanimously by the MRP Executive Committee and was issued to the PWR fleet as a "mandatory" action under the NEI 03‑08 Guideline for the Management of Materials Initiative. The NRR staff previously documented its comments and recommendations on the use of MRP‑139 in a letter dated October 12, 2005, from M. Mayfield (NRC) to A. Marion (NEI). During the meeting, the MRP representatives discussed proposed responses to address the concerns of the NRC staff. Discussions included interactions with the NRC, reporting of inspection findings with the NRC, and Spring 2006 inspection results and implications. MRP representatives indicated that a formal written response to the staff’s concerns documented in the October 12, 2005, letter would be forthcoming. Based on the MRP’s proposed resolution, 17 of 26 staff comments would be addressed satisfactorily.
Representatives from NEI, EPRI, and MRP met with the staff on October 25, 2006, to discuss the unresolved comments. Resolution of some additional items was achieved.
In October 2006 five circumferential indications were
identified in three DM welds on the pressurizer at the Wolf Creek Generating
Station. These 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 baseline inspection schedule for
pressurizer welds, particularly the deferral of the baseline inspections
allowed by the industry’s NEI 03-08 protocol.
Three of the
In the fall of 2006, the staff engaged in extensive discussions with a licensee regarding its request to leave defects in weld overlays that did not satisfy the welding acceptance criteria previously agreed to with the NRC staff (i.e., Appendix Q fabrication inspection criteria). The licensee proposed to employ the analytical evaluation methodology of ASME Section XI, IWB‑3640. The NRC staff rejected this proposal. At the November 2006 Section XI committee meetings, a proposed action was discussed at working groups and at the Subcommittee on Repair Replacement Activities. This action would have involved the use of the analytical evaluation methodology of Section XI, IWB-3640. The action was rejected.
6. New Reactor Licensing Activities
meeting with AREVA on
On October 24 - 25, 2006, NRC hosted two public meetings with AREVA, NP representatives regarding the US EPR standard design.
Public Meeting on 10 CFR Part 52 Final Rule
On October 25, 2006, NRC hosted a public meeting for the purpose of answering stakeholder questions about the draft final rule language for 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants," and for other 10 CFR Parts affected by the Part 52 rulemaking [proposed rule which was published on March 13, 2006 (71 FR 12781)]. The NRC received 19 comment letters from industry stakeholders, other Federal agencies, and individuals during the public comment period. SECY-06-0220 is publicly available. On November 1, 2006, the NRC hosted a public meeting for the purpose of answering stakeholder questions about the supplemental proposed rule for 10 CFR Part 52. On November 9, 2006, the staff and the industry met with the Commission to discuss the Part 52 draft final rule.
NUREG-1835, Supplement 1
On November 13, 2006, the NRC issued Supplement 1 to NUREG-1835, “Safety Evaluation Report for North Anna ESP Site” ahead of schedule. The ADAMS Accession number is ML063170371. The Environmental Impact Statement (EIS) is on schedule to be issued on December 29, 2006.
Standard Review Plan (SRP) Update
The New Reactor Licensing public web-site now has a link directly to the status of Standard Review Plan (NUREG-0800) Proposed Revisions. To access see:
http://www.nrc.gov/reactors/new-licensing/new-licensing-files/srp-table-of-contents.pdf. This Table of Contents illustrates which sections of the SRP have been updated and issued in final; updated SRP sections issued in draft for public comment; SRP section publicly available for preliminary use, and SRP sections that are not scheduled to be updated in March 2007. In March 2007, all SRP sections will be issued in final. This web-page is updated weekly.
Economic Simplified Boiling Water Reactor (ESBWR) Meeting
On January 9-10,
2007, the NRC held a public meeting at the Ramada Inn in
The report describes research to evaluate the effectiveness and determine the reliability of advanced low-frequency ultrasonic testing (UT) to penetrate relatively thick-walled sections of CSS primary piping. The Synthetic aperture focusing NDE technique was in used in conjunction with state-of-the-art phased array methods operating at low frequencies. The study shows that 500-kHz large-aperture phased arrays are capable of detecting inside diameter-connected cracking in heavy-walled CSS piping when inspected from the outside surface of the pipe, provided that access to the outside surface is sufficient for adequate transducer placement and coupling. The results show that for inside surface-breaking thermal and mechanical fatigue cracks greater than approximately 30% through-wall in depth, the 500‑kHz method detected 100% of the flaws. Further, cracks on the order of 15–30% through-wall could also be periodically detected with the 500-kHz phased array method. No through-wall sizing of flaws was performed due to an absence of tip-diffracted responses. Length sizing is possible, although the root mean square error (RMSE) is slightly higher than currently allowed by Section XI of the ASME Code.
· NUREG/CR-XXXX, “Assessment of Eddy Current Testing for the Detection of Cracks in Cast Stainless Steel Reactor Piping Components”
The study showed that if the inner-surface is accessible, an eddy-current (ET) method described in this report is feasible and very effective for detection of these types of surface-breaking cracks. The report describes a number of issues to be resolved before an ET approach can be employed for inservice inspection.
8. Office of Nuclear Regulatory Research (RES) Public Meeting with Electric Power Research Institute (EPRI) on Materials Engineering Research Programs
On October 17 and 18, 2006, the subject meeting was held to discuss primary system materials degradation mechanisms and research. Representatives from EPRI’s Materials Reliability Program presented their evaluations of primary system materials degradation mechanisms and described EPRI’s related research programs. RES staff presented an overview of the NRC’s Proactive Materials Degradation Assessment program. The meeting provided an opportunity to identify materials degradation research of mutual interest which may be amenable to collaboration or sharing. To this end, the RES and EPRI representatives agreed to meet on a semiannual basis, with February 2007 proposed for the next meeting.
9. 7th Meeting of International Cooperative Program for the Inspection of Nickel Alloy Components (PINC)
On October 25 ‑ 27, 2006, Office of Nuclear Regulatory Research (RES) staff participated in the subject meeting which was hosted by the Electric Power Research Institute’s (EPRI) Non‑destructive Evaluation (NDE) Center. The PINC program is focused on assessing the capabilities of current and emerging non‑destructive examination (NDE) techniques for detecting and sizing flaws associated with primary water stress corrosion cracking (PWSCC). The first of several round robin tests to quantify NDE effectiveness has begun, hosted by the Japan Power Engineering and Inspection Corporation (JAPEIC).
10. PBMR Pty Ltd Public Meeting
On October 27, 2006, Office of Nuclear Regulatory Research (RES) staff met with representatives from PBMR Pty Ltd as part of the pre-application review of the pebble bed modular reactor (PBMR) design. During the meeting, participants discussed issues related to the review of a set of white papers on PBMR Pty Ltd’s approach to perform a probabilistic risk assessment, the selection of licensing basis events, and the classification of structures systems and components. The staff was informed that PBMR Pty Ltd intends to submit a fourth white paper.
11. Second Meeting of the European Commission Joint Research Centre/Nuclear Energy Agency Benchmark Study Entitled "Risk Informed In‑Service Inspection Methodologies (RISMET) ‑ Scoping Study"
On November 7‑8,
2006, the NRC hosted the subject meeting, attended by representatives from
foreign regulatory agencies and research organizations and from the
12. November 2006 Publication of Brookhaven National Laboratory Report, "Recommendations for Revision of Seismic Damping Values in Regulatory Guide 1.61," NUREG/CR‑6919
The Office of Nuclear Regulatory Research (RES) published the subject report which provides recommendations for the selection of damping values used in the seismic analysis and design of nuclear power plants. The recommendations in this report considered recently published guidance included in the American Society of Chemical Engineers Standard 43‑05, "Seismic Design Criteria For Structures, Systems, and Components in Nuclear Facilities," and in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 1, Non‑Mandatory Appendix N, "Dynamic Analysis Methods", 2004 Edition.
13. Publication of NUREG/CR-6913, "Chemical Effects Head-Loss Research In Support of Generic Safety Issue 191"
On December 8, 2006, Office of Nuclear Regulatory Research staff published the subject report to the NRC public website. This report describes studies conducted at Argonne National Laboratory on the effects of chemical byproducts, formed in a simulated containment sump environment, on head loss across pressurized‑water reactor (PWR) containment sump screens. Results from this study are being used to support resolution of Generic Safety Issue‑191, "Assessment of Debris Accumulation on PWR Sump Performance."