|
Federal Regulations, Codes, &
Standards Users Group © |
|
||
|
|
|||
|
NRC Section XI Report - November 2007 |
|||
|
|
|||
|
Presented By: Mr. Wally
Norris,
1. Amendment to 10 CFR 50.55a – ASME Code Edition/Addenda 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 Case - Rulemaking/Regulatory Guides A final rule to incorporate Revision 34 to Regulatory Guide 1.84,
“Design, Fabrication, and Materials Code Case Acceptability, ASME Section
III,” and Revision 15 to Regulatory Guide 1.147, “Inservice Inspection Code
Case Acceptability, ASME Section XI, Division 1,” into Part10 of the Code of
Federal Regulations, Section 50.55a (10 CFR 50.55a) by reference was signed
by the Executive Director of Operation on November 1, 2007. The final rule will be noticed in the Federal
Register and will be effective 30 days thereafter. The final regulatory guides listed above
have also been approved and will be noticed in the Federal Register on the
same day as the rulemaking. The effective
date of the guides is governed by the final rule. Final Revision 2 to Regulatory Guide 1.193, “ASME Code Cases Not
Approved for Use,” has also been approved and is available on the NRC’s
regulatory guide website. No public
comments were received on Regulatory Guide 1.193, “ASME Code Cases Not
Approved for Use,” Revision 2.
However, changes to this guide were made 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. Accordingly,
the NRC has decided to work with ASME International to develop acceptable
performance criteria on the use of ultrasonic/radiographic testing and will not
endorse N-659 or Revision 1 to the Code Case 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.” 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 from Supplement 2 through Supplement 0 to the 2007 Edition (also considered Supplement 12 to the 2004 Edition). The draft guides are expected to be published for public comment in Spring 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 is reviewing
the 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. 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. 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). Protection Against Pressurized Thermal
Shock Events On October 3, 2007, the NRC published a proposal to change 10 CFR
50.61 to provide updated requirements for pressurized thermal shock (PTS)
events for PWR reactor vessels (72 FRN 56275). 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. Public comments should
be submitted by December 17, 2007. 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. On July 11, 2007,
a public meeting was held between NRC staff and industry representatives to
discuss the details of the Title 10 of the Code of Federal Regulations (10 CFR) 50.69 passive categorization
process described in Topical Report (TR) WCAP-16308-NP, “Pressurized Water
Reactor Owners Group 10 CFR 50.69 Pilot Program - Categorization Process –
Wolf Creek Generating Station.” The
meeting summary is available in ADAMS (ML072010313). TR WCAP-16308-NP is available in ADAMS
(ML062770345). The meeting was held to
discuss several items arising from a May 17, 2007, NRC staff audit
(ML071640104) of the 10 CFR 50.69 pilot application documentation, related to
the ongoing review of the passive categorization process described in TR
WCAP-16308-NP. The NRC staff opened
the meeting by stating that a discussion of issues arising out of its audit
and TR WCAP-16308-NP would support the issuance of a request for additional
information (RAI) to NEI. The industry
representatives stated that their expectation for the meeting was to clarify
the manner in which the American Society of Mechanical Engineers (ASME) Code
Case N-660, “Risk-Informed Safety Classification for Use in Risk-Informed
Repair/Replacement Activities,” requirements were satisfied during the Wolf
Creek pilot categorization and to clarify the intent of any proposed changes
to N-660. Although the primary purpose
of the meeting was to discuss the 10 CFR 50.69 passive categorization process
described in the TR, there were a few questions regarding the treatment of
structures, systems, and components (ADAMS Accession No. ML071930256). At the meeting conclusion, the NRC staff
and industry representatives reached agreement regarding which comments would
be formalized as RAIs. A follow-up
meeting was held on September 28, 2007, (summary at ML072770519) for the
industry to discuss its proposed responses to the NRC staff’s request for
additional information (RAI) dated August 27, 2007 (ML072220129). The NRC letter dated August 27, 2007,
contained a total of eleven RAI questions.
The Nuclear Energy Institute (NEI) committed to provide RAI responses
to the NRC by October 18, 2007. The
September 28, 2007, meeting was specifically held to ensure that the proposed
responses would adequately address the NRC staff’s concern on the passive
categorization guidance contained in TR WCAP-16308-NP. During the meeting, the NRC staff also
identified an additional RAI that will require a response before the NRC
staff can complete its review. 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. ASME Addenda B
(2005) To Standard RA-S-2002 On July 18, 2007,
a public meeting was held between NRC staff and industry representatives to
present its proposal for satisfying the PRA quality guidelines for
risk-informed inservice inspection (RI-ISI) programs developed using ASME
Code Case N-716 “Alternative Piping Classification and Examination
Requirements, Section XI Division 1.”
The meeting summary is in ADAMS (ML072130491). Code Case N-716 has not yet been endorsed
for use by the NRC but has been used in two pilot applications under review
by the NRC staff. The industry
opened the meeting by providing a draft report (Enclosure 2 in 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. Nine licensees
with spring 2008 refueling outages had committed to inspect the welds by the
end of 2007 if an adequate level of safety from an industry finite element
analysis program was not demonstrated to the NRC. By letter dated February 14, 2007, the
Nuclear Energy Institute (NEI) indicated that the Electric Power Research
Institute Materials Reliability Program would be undertaking a task to refine
the crack growth analyses pertaining to the 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 had performed.
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). The NRC staff has
evaluated the MRP report and documentation provided by the industry (safety
assessment is available at ML072470394).
The principal conclusions resulting from this safety assessment are as
follows.
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. [Also see Number 8. below] 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. 6. Visual Testing NUREG NUREG/CR-6943, A Study of Remote Visual Methods to Detect Cracking in Reactor Components,” has been posted to the NRC public website. The study concludes that a significant fraction of the cracks that have been reported in nuclear power plant components are at the lower end of the capabilities of the VT equipment currently being used. The study also suggests that inspection conditions need to be nearly ideal to detect these cracks. The research has shown that the use of visual inspections in lieu of ultrasonic testing must be carefully considered taking into account factors such as surface conditions and expected crack opening displacement. 7. Treatment of Operational Leakage On August 30,
2007, a meeting was held between NRC staff and industry representatives to
discuss the development of NRC staff and industry guidance related to
operational leakage in ASME Code Class pressure boundary components. The meeting summary is available in The NRC interim
guidance indicates that American Society of Mechanical Engineers (ASME) Code
Class components with through wall leakage in high energy piping systems should
be declared inoperable immediately.
The draft revision to Part 9900 indicates that it is the NRC staff
view that for ASME Class 2 moderate or high energy (HE) components and Class
3 HE components with identified through-wall leakage, the staff considers
that it may not be feasible to make an immediate operability determination
that a reasonable expectation of operability exists. Industry representative discussed some
scenarios that could occur in ASME code Class 2 components and Class 3 HE
components to illustrate their view that such an immediate operability
determination may be feasible. As a result of
discussions, industry representatives indicated that they had a better
understanding of the NRC staff view and offered to prepare draft industry
guidance to address the NRC staff concerns in this area. NRC staff indicated it would review NEI
comments in meeting handouts and consider revising the draft Part 9900 Appendices. During this meeting, NRC staff and industry
representatives also discussed use of the ASME Code and ASME Code Cases for
evaluation of structural integrity and for structural integrity acceptance
criteria for pressure boundary components with through-wall leakage. NRC staff met with representatives from the industry on October 18, 2007, to continue discussions regarding industry concerns with Regulatory Issue Summary (RIS) 2005-20, “Revision to Guidance Formerly Contained in NRC Generic Letter 91-18, “Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and Operability,” as delineated in NEI White Paper dated May 2, 2006 (ML061320347), Revision 1 submitted October 24, 2006 (ML07240526), and Revision 2 submitted on May 11, 2007 (ML071590195). The summary of a meeting held on August 30, 2007, is available at ML072540764. During this meeting, NRC staff and industry representatives also discussed use of the ASME Code and ASME Code Cases for evaluation of structural integrity and for structural integrity acceptance criteria for pressure boundary components with through-wall leakage. 8. 10th Meeting of the International Cooperative Program for the Inspection of Nickel Alloy Components (PINC) On October 3-5,
2007, staff from the Office of Nuclear Regulatory Research (RES) conducted
this meeting that was and hosted by the Valtion Teknillinen Tutkimuskeskus
(VTT) Technical Research Centre of Finland.
The meeting focused on the reliability of non-destructive examinations
(NDE) for primary water stress corrosion cracking (PWSCC). Participants reviewed the progress of the
Atlas Database of PWSCC Morphology and NDE Responses and discussed round robin
tests that are assessing the performance of various NDE techniques to detect
and size PWSCC cracks in piping and dissimilar metal welds. Additional participants are sought for
trials on mockups simulating the seal weld of a bottom mounted instrumentation
(BMI) vessel penetration. Staff
visited the Finnish nuclear regulator (STUK) to discuss this program and the
codes and standards governing NDE in plant construction and in-service
inspection. 9. Public Meeting with Toshiba on
Requested Pre-application Review of the “Super-Safe, Small, and Simple” (4S)
Reactor Design On October 23, 2007, the staff held a public meeting with
representatives from the Toshiba Corporation (Toshiba) and its project partners,
Westinghouse Electric Company and |
|||
|
|
|||
|
|
|||
