Federal Regulations, Codes, & Standards
Users Group ©
NRC Section XI Report - August 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 closed on June 19, 2007. Responses to the comments are being developed. The final rule is scheduled for publication in January 2008.
2. ASME Code Cases - Rulemaking/Regulatory Guides
The following draft final regulatory guides have been approved by the cognizant NRC offices (including the Office of General Counsel) and have been transmitted to the Advisory Committee on Reactor Safeguards 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 October 2007.
No public comments were received on Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use,” Revision 2. However, changes to this guide are being considered as a result of public comments received on Regulatory Guide 1.84; specifically, Code Case N-659. As discussed in the Federal Register Notice (71 FR 62947, dated October 27, 2006), the NRC staff proposed conditional approval of N-659. Public comments were transmitted expressing concern with a number of the proposed conditions. The issues are complicated and addressing them is not straightforward. Potential approaches are being considered. One approach under consideration is to work with industry to develop acceptable performance criteria on the use of ultrasonic/radiographic testing and not endorse N-659 at this time. There are several Code actions under development that may be affected, i.e, BC04-247, Code Case N-713, “Use of Ultrasonic Examination in Lieu of Radiography, Section XI, Division 1,” and BC06-1092, “IWA-4520, revise to permit use of Section XI personnel qualifications, methods, and criteria for repair/replacement activities.” This guide is also scheduled to be published as a final guide in October 2007.
Proposed Revision 35 to Regulatory Guide 1.84 and proposed Revision 16 to Regulatory Guide 1.147 are under development. The guides will include Code Cases through Supplement 12 to the 2004 Edition. The draft guides are expected to be published for public comment early 2008..
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 (ECCS)
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 has provided SECY-07-0082 to the Commission recommending how to proceed with the rulemaking and providing several other options.
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 was held on May 30, 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. Additional information, including a pilot application was received by the NRC on August 6, 2007, requesting completion by April 17, 2008. The staff has begun its review.
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 and the staff is working together with NEI and EPRI to assess whether it is feasible to provide more directed PRA quality guidance that could be used in support of RI-ISI updates.
4. Generic Activities on Material Degradation/PWR Alloy 600/182/82 PWSCC
The circumferential indications identified in three dissimilar metal (DM) welds on the pressurizer at the Wolf Creek Generating Station 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. The pressurizer surge nozzle-to-safe end weld indications are of concern, as this is the first time that multiple circumferential indications have been identified in this weld. This condition calls into question the degree of safety margin present in past structural integrity evaluations for flawed DM welds, since multiple stress-corrosion cracking flaws may grow independently and ultimately grow together, significantly reducing the time from flaw initiation to leakage or rupture. The size of the relief nozzle-to-safe end flaw is also of concern, as this flaw has a much larger aspect ratio than those assumed in the estimates used to establish the basis for the industry-sponsored MRP-139 program. Larger aspect ratios could result in achieving a critical flaw size and rupture prior to the onset of detectable leakage. Various public meetings were held on these issues at the end of 2006 and early 2007.
Nine licensees with spring 2008 refueling outages have committed to inspect the welds by the end of 2007 if an adequate level of safety from an industry finite element analysis program has not been demonstrated to the NRC. The industry has been 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. Following is a summary of some of the recent meetings and activities.
A meeting between NRC staff and the Expert Review Panel for Advanced Finite Element Analysis (FEA) Crack Growth Calculations was held on May 1, 2007. A summary of the meeting is available in ADAMS (ML0713505380). The meeting was one of a series of meetings being held between industry representatives and NRC staff to discuss the industry’s advanced finite element fracture mechanics analysis program undertaken to address NRC staff’s question concerning the likelihood that rupture may occur in some cases without prior evidence of leakage in pressurizer nozzle welds susceptible to primary water stress corrosion cracking. The meeting specifically addressed weld residual stress inputs and validation. The industry’s and NRC’s analytical approaches provided similar results. The participants discussed the development of a list of weld residual stress distribution cases that the industry will need to calculate in the context of the overall analysis matrix.
A letter dated May 31, 2007, from Jay K. Thayer, NEI, to Luis
A. Reyes, Executive Director of Operations, NRC, (ML0715901240), provided
additional information on industry action associated with potential generic
implications of the
A letter from Michele G. Evans, Director, Division of
Component Integrity, Office of Nuclear Reactor Regulation (NRR), to Alex
Marion, Executive Director, Nuclear Energy Institute (NEI), dated June 20,
2007, is available in ADAMS (ML0715603750).
The letter responds to a February 14, 2007, letter from Jay Thayer,
NEI to J.E. Dyer, NRC, (ML070600672), proposing that industry perform
advanced finite element fracture mechanics analyses to address NRC staff concerns
that rupture could occur in pressurizer dissimilar metal nozzle welds without
prior evidence of leakage. The letter
from Ms. Evans provides a technical basis document with additional
information related to an NRC staff concern regarding the industry’s approach
for calculating crack stability. The
NRC staff’s final report on the
A summary of the June 19 - 20, 2007, meeting between NRC staff and the Expert Panel for the Wolf Creek Advanced Finite Element Analyses (FEA), is available at ML0718606720. The meeting was a continuation of ongoing discussions between industry representatives and NRC staff regarding the likelihood that pressurizer nozzle dissimilar metal welds, which are potentially susceptible to stress corrosion cracking, will leak before rupturing under certain conditions. The meeting specifically addressed several key technical issues such as: fabrication records, weld residual stress modeling results, secondary stresses, sensitivity case matrix results, probabilistic assessment, industry proposed acceptance criteria, safety factors, and scheduling issues. The NRC staff and contractor provided comments on industry’s results and proposals regarding acceptance criteria and safety factors.
A summary of a meeting between NRC staff and the Expert Panel for the Advanced Finite Element Analyses (FEA) of Pressurizer Nozzle Welds, held on July 17, 2007, is available at ML0720601230. The meeting was a continuation of ongoing discussions between industry representatives and NRC staff regarding the likelihood that pressurizer nozzle dissimilar metal welds, susceptible to stress corrosion cracking, will leak before rupturing under certain conditions. The meeting specifically provided an opportunity for the NRC staff to provide comments on Draft Report A (ADAMS Accession Number: ML071970073) submitted by industry on July 10, 2007. The meeting began with a discussion by representatives of Dominion Engineering of their final analytical results, including new evaluations addressing the effects of multiple flaws. Model validation was discussed further. The meeting concluded with a discussion on the final project schedule, milestones, and deliverable. A table of NRC comments was provided in the NRC presentation with the most significant comments highlighted in yellow.
NRC staff met with industry on August 9, 2007, to provide members of the public with an opportunity to obtain an overview of the entire project. NRC staff discussed the confirmatory analysis it has performed. NRC staff have received a copy of the final report entitled, “Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC Flaws in Pressurizer Nozzle Dissimilar Metal Welds Evaluations Specific to Nine Subject Plants,” Final Report Draft A, dated July 10, 2007, (ADAMs No. ML071970073), which is also available from the EPRI web site. A supporting draft report entitled, “Evaluation of Pressurizer Alloy 82/182 Nozzle Failure Probability (Including Effect of the Fall-06 Wolf Creek NDE Indications), by Structural Integrity Associates, Inc., dated July 14, 2007, is also available in ADAMs (ML071970083). NRC is preparing a safety assessment on the project which will be issued to the 9 affected utilities by the end of August 2007.
PINC Program Round-Robin on BMI Nozzles
Work has progressed on the NRC-led international
cooperative research program called the Program for the Inspection of
Nickel-Alloy Components (PINC). The
program is assessing the capabilities of current and emerging non-destructive
examination (NDE) techniques to detect and size flaws associated with primary
water stress corrosion cracking (PWSCC).
The most recent meeting of the program was hosted by
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.
A summary of the meetings between NRC staff and Electric
Power Research Institute (EPRI) staff held at the EPRI Office in
7. Publication of Regulatory Guide (RG) 1.206, “Combined License Applications for Nuclear Power Plants (LWR Edition)”
On June 20, 2007, the Office of Nuclear Regulatory
Research issued RG 1.206, “Combined License Applications for Nuclear Power
Plants (LWR Edition).” The RG,
prepared by the Office of New Reactors, provides guidance regarding the
information to be submitted in a combined license (
8. Pebble Bed Modular Reactor (PBMR) Meeting
On July 18, 2007, the staff held a public meeting with PBMR (Pty) Ltd to discuss the staff’s ongoing review of four white papers that have been submitted regarding a pebble-bed high temperature gas-cooled reactor design that is being proposed for submittal as a design certification application. The application is expected to be submitted in the later part of CY 2009. The subjects of the white papers are: (1) their approach to probabilistic risk assessment; (2) the identification of licensing basis events; (3) the classification of structures systems and components; and (4) the defense-in-depth approach. In addition to the technical discussions, PBMR (Pty) Ltd presented its plans to submit about 15 additional white papers before the end of 2008. The staff provided information about NRC’s current priorities and resource constraints as applicable to its pre-application review.