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Federal Regulations, Codes, &
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NRC Section XI Report - August 2008 |
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Presented By: Mr.
1. Amendments to 10 CFR 50.55a – ASME Code Edition/Addenda 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
(ASME Code) and Code for Operation and Maintenance of Nuclear Power Plant
Components (OM Code) 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. The final rule
should be published in September 2008. The Executive Director for Operations approved a Lean Six Sigma Project for the purpose of reducing the rulemaking process cycle time to incorporate edition and addenda of the ASME Code and OM Code into 10 CFR 50.55a. The project started on May 27, 2008, and includes staff from the Office of the General Counsel, Office of Nuclear Reactor Regulation, Office of Regulatory Research, and Office of New Reactors. The project team has been reviewing the current process to determine how a more efficient process can be developed to improve NRC resource utilization. All other rulemaking activities are excluded from the scope of this project. However, lessons learned and identified process enhancements from this project will be evaluated for making improvements to other rulemakings (e.g., ASME Code Case regulatory guide rulemaking process). 2. ASME Code Case Rulemaking/Regulatory Guides Proposed Revision 35 to RG 1.84, “Design, Fabrication, and Materials
Code Case Acceptability, ASME Section III,” proposed Revision 16 to RG 1.147,
“Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1,”and
proposed Revision 3 to RG 1.193 “ASME Code Cases Not Approved for Use” are
currently in review. The guides
address Code Cases from Supplement 2 to the 2004 Edition 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 the September/October 2008
time frame. The NRC staff has completed its review of Supplements 1, 2, and 4 to the 2007 Edition (there were no nuclear Code Cases in Supplement 3). 3. Risk–Informed
Activities Alternative Risk-Informed Inservice
Inspection (RI-ISI) Program – Code Case N-716 On September 21,
2007, the NRC conditionally approved an alternative risk-informed inservice inspection
program based on Code Case N-716 for Grand Gulf Nuclear Station Unit 1 (the
safety evaluation is available in the NRC’s Agencywide Documents Access and
Management System (ADAMS) as Accession No. ML072430005). On September 28, 2007, the NRC conditionally
approved a similar alternative RI-ISI program for Donald C. Cook Nuclear
Plants 1 and 2 that is based on Code Case N-716 but deviates and expands upon
the Code Case (the safety evaluation is available in ADAMS [ML072620553]). Finally, a similar alternative RI-ISI
program was approved on April 28, 2008 [ML080980120] for Waterford Steam
Electric Station. The approvals were
specific to these units and relied on several changes to the methodology
described in Code Case N-716. The
Working Group Implementation of Risk-Based Examination is considering the
alternative RI-ISI programs as described in the safety evaluations. 10 CFR 50.46a - Option 3 Rulemaking (Risk-Informed Emergency Core
Cooling System - ECCS) The NRC staff prepared a proposed rule containing emergency core
cooling system evaluation requirements, that could be used as an alternative
to the current requirements in 10 CFR 50.46, “Acceptance Criteria for Emergency Core Cooling Systems ( The Advisory Committee on Reactor Safeguards (ACRS) issued a letter on
November 16, 2006, (ML063190465) with 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. As a result, the staff provided SECY-07-0082, “Rulemaking To Make
Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements,” to
the Commission recommending how to proceed with the rulemaking
(ML070180466). 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 additional
direction (ML072220595). In March 2008,
the staff provided the Executive Director for Operations its proposed
schedule for completing the final rule (ML080370354). It is projected that the staff will meet
with the ACRS to discuss the revised draft final rule in May 2009. Pressurized
Thermal Shock – 10 CFR 50.61 On October 3,
2007, the NRC published a proposed amendment to 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 NRC received over 40 comments. The staff is currently developing responses
to the comments. It is anticipated
that the final rule will be published in mid-2009. Reactor Vessel Weld Inspection By letter dated January 26, 2006, the
Westinghouse Owners Group, later the Pressurized-Water Reactor Owners Group
(PWROG), submitted topical report WCAP‑16168-NP, Revision 1,
“Risk-Informed Extension of the Reactor Vessel In-Service Inspection
Interval.” The The Final Safety
Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report
(TR) WCAP-16168-NP, Revision 2, “Risk-Informed Extension of the Reactor
Vessel In-service Inspection Interval,” was issued on May 8, 2008
(ML081060051). The topical report
relies extensively on work described in NUREG-1874, "Recommended
Screening Limits for Pressurized Thermal Shock (PTS)." The NRC staff has accepted TR
WCAP-16168-NP, Revision 2, based on the imposition of a condition related to
the augmented evaluation of in-service inspection (ISI) results taken from Section
(e) of the proposed rule 10 CFR 50.61a, published in the Federal Register on
October 3, 2007 (72 FR 56275). The NRC
staff is in the process of reviewing public comments on the proposed rule and
preparing the final rule. If the final
10 CFR 50.61a differs from the proposed 10 CFR 50.61a with regard to the
augmented ISI evaluation requirements, it is expected that the PWROG will
review the requirements in the final 10 CFR 50.61a and determine whether a
revision to the accepted TR WCAP-16168-NP, Revision 2, is required. Furthermore, licensees that choose to
implement 10 CFR 50.61a must perform the ISI required in Section (e) of the
rule, and must submit the required information for review and approval to the
Director, Office of Nuclear Reactor Regulation, at least three years before
the limiting RTPTS value calculated under 10 CFR 50.61 is projected to exceed
the PTS screening criteria in 10 CFR 50.61.
Licensees implementing Section (c) of 10 CFR 50.61a must perform the
inspections and analyses required by Section (e) of 10 CFR 50.61a prior to implementing
the extended interval. Phased Approach to Probabilistic Risk
Assessment Quality The increased use of probabilistic risk assessments (PRAs) in the
NRC’s regulatory decisionmaking process requires
consistency in the quality, scope, methodology, and data used in such
analyses. A key aspect of
implementing a phased approach to PRA quality is the development of PRA
standards and related guidance documents.
To achieve that objective, professional societies, the nuclear
industry, and the staff have undertaken initiatives to develop national
consensus standards and guidance on the use of PRA in regulatory
decisionmaking. ASME and ANS recently
published a joint PRA standard, “Level 1 and Large Early Release Frequency
(LERF) PRA Standard” (ASME/ANS RA-S-2008), which applies to at-power internal
events, internal fire events, and external events for operating reactors. The staff has
initiated work on Revision 2 of Regulatory Guide 1.200, “An Approach for Determining the Technical
Adequacy of Probabilistic Risk Assessment Results for Risk-Informed
Activities.” This revision will
endorse the joint ASME/ANS Level 1/LERF PRA standard. The staff recently held public meetings on
this topic and plans to issue Revision 2 for public comment in mid-2008. In November 2007,
the staff issued draft NUREG-1855, “Treatment of Uncertainties from PRAs in
Risk-Informed Decision-Making,” for public review and comment. It is being developed in collaboration with
the Electric Power Research Institute (EPRI) who has issued a draft report on
uncertainties, as part of the NRC/EPRI Memorandum of Understanding. These two documents are meant to be
complimentary. The NRC report along
with the EPRI report provides information and guidance on uncertainties
associated with PRA. They are meant to
provide guidance on meeting the requirements in the ASME/ANS PRA standard on
uncertainties, and provide guidance on how to treat the results from the
uncertainty analyses in decision making for risk-informed activities. The staff held two public meetings and
plans to issue a final version in late 2008. The regulatory
guide and NUREG report (including the EPRI report) will assist the staff in
establishing the technical acceptability of the PRA results to be used in
regulatory decision making. When used
in support of an application, these documents will obviate the need for an
in-depth review of the base PRA by NRC reviewers, and provide for a more
focused and consistent review process. 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”
[ML062770345]. The Topical includes,
in part, an alternative methodology to the NRC endorsed Code Case N-660,
“Risk-Informed Safety Classification for Use in Risk-Informed
Repair/Replacement Activities,” for categorization of passive components. On April 9, 2008,
a public meeting was held with industry to discuss the topical report
(TR). Specifically, industry
representatives met with the NRC staff to discuss responses (ML080780403) to
the NRC staff’s request for additional information (RAI) dated August 27,
2007 (ML072220129). Several key
technical issues in the TR were identified with which the staff had
disagreement. Regarding the final
document method, the NRC staff stated that it plans to write its safety
evaluation based on a review of TR WCAP-16308-NP, Revision 0, as supplemented
by the RAI responses. The NRC staff
identified several concerns with regards to the contents of Table A-2a and
A-2b of TR-WCAP-16308-NP. During the
meeting, Westinghouse representatives discussed proposed changes to these
tables, to make it easier for the NRC staff to compare/track changes between
the approved Code Case N-660 and the TR. The handouts that
Westinghouse representatives presented to facilitate the discussion are
available in 4. Generic Activities on Material Degradation/PWR Alloy 600/182/82 PWSCC Prior to 2005,
inspection of dissimilar metal (DM) butt welds was performed in accordance
with ASME Code, Section XI, requirements.
In late 2005, the industry implemented an initiative for more
aggressive inspection schedules than required by Code. This initiative is documented in MRP-139,
“Materials Reliability Program: Primary System Piping Butt Weld Inspection
and Evaluation Guideline.” The base
line inspection schedules in MRP-139 are based on temperature and size; for
example, pressurizer inspections were required to be completed first. The on-going inspection frequency is based
on the mitigation method applied to a weld (for example, the most frequent
inspections would occur in unmitigated pressurizer welds). The staff is monitoring the implementation
of the MRP-139 program for managing PWSCC in DM butt welds. NRC recently issued Temporary Instruction
(TI-172) for regional staff to verify that all PWRs with DM butt welds are
implementing MRP-139. For the longer
term, the NRC staff requested that ASME develop a Code Case for inspection of
Alloy 82/182 butt welds. ASME has been
actively working on the Code Case and has been responsive to NRC input on the
Code Case. The Code Case is nearing
completion. Based on progress to date,
the staff expects that the Code Case will be acceptable for referencing with
minimal comments/conditions and that the Code Case will be incorporated by
reference into 10 CFR 50.55a. In Oct 2006
inspections were performed at Recently, two
B&W plants experienced PWSCC in decay heat removal to hot leg welds. The staff evaluated the experience from
these plants and concluded that these experiences are within the assumptions
upon which the current inspection schedules are based. Recently a potential safety issue was identified related to inspections of nozzles from a retired pressurizer. These inspections caused the staff to question whether the initial flaw characterization was inconsistent with the advanced finite element analysis basis for pressurizer inspections scheduled in spring 2008. Based on quick turn around but more advanced inspection techniques used by industry and witnessed by the NRC staff, the staff concluded that the flaws were fabrication-induced and there was no structurally significant PWSCC in the welds. This was confirmed via destructive analysis. The NRC staff continues to monitor and evaluate operating experience to ensure the current inspection schedules are adequate. 5. New Reactor Licensing Activities The New Reactor Licensing public web-site [http://www.nrc.gov/reactors/new-reactors.html has a list of expected new nuclear power
plant applications, and an estimated schedule by fiscal year for new reactor
licensing applications. New Reactor License
Application Reviews On June 11, 2008,
the Office of New Reactors published a Notice of Availability of the Final
Interim Staff Guidance DC/COL-ISG-03 on Probabilistic Risk Assessment
Information to Support Design Certification and Combined License
Applications. The Final Interim Staff
Guidance (ISG) DC/COL-ISG-03 (ML081430087) supplements the guidance provided
to the staff in Section 19.0, “Probabilistic Risk Assessment and Severe
Accident Evaluation for New Reactors,” of NUREG-0800, “Standard Review Plan
for the Review of Safety Analysis Reports for Nuclear Power Plants,”
concerning the review of probabilistic risk assessment (PRA) information and
severe accident assessment submitted to support design certification (DC) and
combined license (COL) applications. Advanced Reactor Regulatory
Structure: NUREG-1860 The staff issued
NUREG-1860, “Feasibility Study for a Risk-Informed and Performance-Based
Regulatory Structure for Future Plant Licensing,” Volumes 1 and 2, in
December 2007. This NUREG establishes
the feasibility of developing a risk-informed and performance-based
regulatory structure for the licensing of future nuclear power plants. It documents a framework that provides an
approach, scope, and criteria that could be used to develop an alternative
set of requirements to 10 CFR Part 50, “Domestic Licensing of Production and
Utilization Facilities,” for future nuclear power plants. The staff will, as resources permit,
continue its pre-application discussions supporting a proposed risk-informed
licensing approach for the pebble-bed modular reactor (PBMR). The licensing strategy for the next
generation nuclear plant (NGNP) that is currently being prepared in
cooperation with the Department of Energy will also include recommendations
for the use of risk insights in the licensing of the NGNP prototype
facility. These experiences will
inform the staff's recommendation to the Commission in 2009 on the possible
use of risk-informed and performance-based approaches within rulemakings to support
the licensing of Generation IV reactors. 6. Periodic Meeting to Discuss Development of
Nuclear-related Codes and Standards at the National Institute of Standards
and Technology (NIST) in On July 9, 2008, staff of the NRC, National Institute of Standards and Technology, and Department of Energy held a public meeting with representatives of standards developing organizations (SDOs) and nuclear industry representatives to discuss priorities for developing nuclear-related codes and standards that may be endorsed in lieu of developing Government-unique standards. The discussion focused on identifying and addressing gaps and overlapping efforts among SDOs. The Government agencies recommended that SDOs increase their cooperative efforts, while the SDOs suggested that Government increase the attention given to international standards. 7. Preliminary Assessment of NDE Methods for the Inspection of HDPE The NRC has been
conducting research at Pacific Northwest National Laboratory to assess issues
related to the NDE of high density polyethylene (HDPE) piping. A Technical Letter Report, PNNL-17584,
“Preliminary Assessment of NDE Methods on Inspection of HDPE Butt Fusion
Piping Joints for Lack of Fusion,” is available in ADAMS (ML081780765). Issues currently
under consideration are PE pipe joining methods, NDE effectiveness in
detecting fabrication flaws, the degradation process to which PE pipe might
be susceptible, and the effectiveness of NDE to detect this degradation and
structural integrity. Some high-speed tensile impact tests of HDPE butt joints have been completed. The butt joints were fused under conditions expected to produce lack of fusion (LOF) in the joint. Time-of-flight diffraction and phased-array (PA) methods were used to evaluate the joints to select both “good” and “bad” areas (i.e., areas expected to contain LOF). The results are being evaluated, and additional testing will be performed. |
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