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Federal Regulations, Codes, &
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NRC Section XI Report - February 2008 |
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Presented By: Mr. Wally Norris, United States Nuclear Regulatory Commission
1. Amendments to 10 CFR 50.55a A proposed amendment to Part10 of the Code of Federal Regulations, Section 50.55a (10 CFR 50.55a) to 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 have been developed and are under review. It is anticipated that the final rule will be published in May 2008. 2. ASME Code Case - Rulemaking/Regulatory Guides The following final regulatory guides (RGs) were noticed in the Federal Register (72 FR 71750) on December 19, 2007: · 1.84, Revision 34, “Design, Fabrication, and Materials Code Case Acceptability, ASME Section III,” [NRC’s Agencywide Documents Access and Management System (ADAMS) No. ML072070407] ·
1.147,
Revision 15, “Inservice Inspection Code Case
Acceptability, ASME Section XI, Division 1,” ( ·
1.193,
“ASME Code Cases Not Approved for Use” ( The final rule incorporating RGs 1.84 and 1.147 by reference into 10 CFR 50.55a is also available in ADAMS (ML070360713). The effective date of the rule and hence, RGs 1.84 and 1.147, was January 18, 2008. RGs 1.84 and 1.147 list the new and revised Code Cases that the NRC has approved or conditionally approved for use. RG 1.193 lists unapproved Code Cases and is therefore not incorporated by reference into the regulations. Proposed Revision 35 to RG 1.84, proposed Revision 16 to RG 1.147, and proposed Revision 3 to RG 1.193 are currently in review. The guides address Code Cases from Supplement 2 through Supplement 0 to the 2007 Edition (Supplement 0, 2007 Edition also serves as Supplement 12 to the 2004 Edition). The draft guides are expected to be published for public comment in Spring 2008. The NRC staff has formally initiated review of Supplements 1 and 2 to the 2007 Edition. 3. Risk-Informed
Activities Alternative Risk-Informed Inservice Inspection (RI-ISI) Program – Code Case N-716 On September 28, 2007, the NRC approved an alternative RI-ISI program for Donald C. Cook Nuclear Plants 1 and 2. The methodology used by the licensee is based on Code Case N-716 but deviates and expands upon the Code Case. The NRC approved the proposed methodology with a number of conditions. Since the methodology deviates from that contained in the Code Case, the safety evaluation only approved the use of the licensee’s proposed methodology and not Code Case N-716. The safety evaluation is available in ADAMS (ML072620553). On September 21, 2007, the NRC issued a safety evaluation for Grand Gulf Nuclear Station Unit 1 regarding a request for an alternative risk-informed inservice inspection program based on Code Case N-716. Entergy also chose to modify the methodology described in CC N-716 while developing its proposed alternative program. Thus as with the DC Cook approval described above, the NRC staff’s approval of the licensee’s alternative program does not constitute approval of Code Case N-716. 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. The Commission’s August 10, 2007, Staff
Requirements Memorandum directed the staff to continue with the rule making
but to change the priority from high to medium and provided some addition
direction. The NRC staff is currently
developing a schedule to complete this rulemaking. The schedule is due to the Commission at
the end of March 2008. 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. The topical report relies extensively on
work described in NUREG-1874, “Recommended Screening Limits for Pressurized
Thermal Shock (PTS).” The staff has
begun resolving public comments and projects that the safety evaluation
report rule will be published Summer 2008. Protection Against Pressurized Thermal
Shock Events On October 3,
2007, the NRC published a proposed change 10 CFR 50.61, “Alternate Fracture Toughness Requirements
for Protection Against Pressurized Thermal Shock Events,” to provide updated requirements for
pressurized thermal shock (PTS) events for PWR reactor vessels (72 FRN
56275). The proposed rule is commonly
referred to as 10 CFR 50.61a. The
updated technical basis uses many different models and parameters to estimate
the yearly probability that a PWR will develop a through-wall crack as a
consequence of PTS loading. These new
requirements would be voluntarily utilized by any PWR licensee as an
alternative to complying with the existing requirements. The public comment period closed on December
17, 2007. The staff has begun
developing responses to the comments.
A schedule for the final rule process is under consideration. Repair and Replacement In September 2006, the Pressurized Water Reactor Owners Group (PWROG) 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 PWROG, through NEI, responded to NRC's August 27, 2007, request for additional information on October 22, 2007. The current review schedule includes an April 7, 2008, public meeting to discuss the staff's draft safety evaluation. 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. A number of
significant meetings have been held on these issues in 2007. The NRC staff is developing a temporary instruction (TI) on dissimilar metal butt welds. TIs are inspection procedures used by regional inspectors. The objective of this TI is to verify that licensees are implementing mitigation and inspection programs consistent with MRP-139. This TI is planned to take effect early in 2008. 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, and an estimated schedule by fiscal year for new reactor
licensing applications. Combined License
Application Review Process On December 18,
2007, the NRC issued NRC Regulatory Issue Summary 2007-25, “Combined
License Application Acceptance Review Process.” The purpose of issuing the regulatory issue summary (RIS) was
to clarify the procedural aspects of filing a Process for
Scheduling Acceptance Reviews On January 10,
2008, the NRC issued NRC Regulatory Issue Summary 2008-01, “Process for
Scheduling Acceptance Reviews Based on Notification of Applicant Submission Dates
for Early Site Permits (ESP), Combined Licenses (COL), and Design
Certifications (DC) and Process for Determining Budget Needs for Fiscal Year 2010. The NRC is issuing this RIS to communicate its process for scheduling
its acceptance reviews of ESP, DC, and 6. BWRVIP-100-A On November 1, 2007, the NRC approved with comment BWRVIP-100-A, “BWR Vessel and Internals Project, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds.” The comments were that the report was acceptable provided that the Boiling Water Reactor Vessels Internal Project (BWRVIP), in the future, provide the amount of delta ferrite in the stainless steel weld metal to facilitate an effective assessment of the synergistic effect of neutron embrittlement and thermal embrittlement on stainless steel welds. In addition, certain issues stated in the staff’s safety evaluation dated March 1, 2004, were reiterated, which require future actions, but these issues do not affect the acceptability of the BWRVIP-100-A report. The safety evaluation is available in ADAMS (ML073050135). 7. BWRVIP-96-A On November 1, 2007, the NRC approved BWRVIP-96-A, “BWR Vessel and Internals Project Sampling and Analysis Guidelines for Determining the Helium Content of Reactor Internals.” As a result of NRC staff inquiries, minimal revisions were made to the original submittal (BWRVIP-96) resulting in BWRVIP-96-A. The NRC has determined that the information in the BWRVIP-96-A report accurately incorporates all of the relevant information which was submitted by the BWRVIP to support NRC staff approval of the report. The safety evaluation is available in ADAMS (ML073005027). |
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