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Federal Regulations, Codes, & Standards Users Group © |
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NRC Section XI Report - April 2008 |
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Presented By: Mr. Wally Norris,
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. 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 Code Cases not
yet approved for use 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 the June/July 2008 time frame. The NRC staff has
nearly completed its review of Supplements 1 and 2 to the 2007 Edition. There were no nuclear Code Cases in
Supplement 3 to the 2007 Edition. 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 ADAMS [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]). Since the
methodologies deviated from that contained in the Code Case, the safety
evaluations only approved the use of the licensee’s proposed methodology and
not Code Case N-716. The Working Group
on Risk-Based Implementation has indicated that licensees have expressed
great interest in implementing the Code Case.
Accordingly, the staff is considering including the Code Case in draft
Revision 16 of Regulatory Guide 1.147 with conditions consistent with the
safety evaluations described above. 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 Topical Report WCAP-16168-NP Revisions 2,
"Risk-Informed Extension of Reactor Vessel In-Service Inspection
Interval," provides the basis for an extension of the reactor vessel
weld inspection interval from 10 to 20 years.
The staff has completed its review of this topical report and provided
a draft to the PWR Owners' Group. The
staff has resolved all the PWR Owners' Group comments and has placed the
final draft of the safety evaluation into concurrence. The topical report relies extensively on
work described in NUREG-1874, "Recommended Screening Limits for
Pressurized Thermal Shock (PTS)."
The safety evaluation requires licensees utilizing the topical report
to use the inspection criteria in the proposed PTS rule, 10 CFR 50.61a,
"Alternate Fracture Toughness Requirements for Protection against
Pressurized Thermal Shock Events." 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 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 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. 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-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. On April 9, 2008,
a public meeting was held in Rockville, MD, to discuss the 10 CFR Part 52 licensing process and the relationship
to the ASME Boiler and Pressure Vessel Code for the design and construction
of nuclear power plants. The purpose
of the meeting was to promote communication between the NRC and the ASME
relative to new reactor licensing activities and to identify potential
changes to ASME Code requirements to support the new Part 52 licensing
process. The NRC discussed Part 52 and
its relationship to other regulations, such as 10 CFR 50.55a. The ASME provided an overview of standards
actions and task teams related to new reactor developments (including
world-wide activities) and ASME Code structure and philosophy. The ASME has initiated a new joint Section
III / Section XI committee, the Task Group on New Reactors, that will hold
its first meeting in Information
Notice On April 7, 2008,
NRC Information Notice 2008-04: Counterfeit Parts Supplied to Nuclear Power
Plants, was issued to inform the industry of the potential for counterfeit
parts to enter supply chains. 6. Operating
Reactors Information Notice On March 19,
2008, the NRC issued NRC Information Notice 2008-02: Findings Identified
During Component Design Bases Inspections, to inform addressees of issues
identified during recent component design bases inspections (CDBIs) regarding
the capability of selected components to perform their design bases safety
functions. The NRC expects that recipients
will review the information for applicability to their facilities and
consider actions, as appropriate, to avoid similar problems. 7. NUREG/CR-6960 NUREG/CR-6960,
“Crack Growth Rates and Fracture Toughness of Irradiated Austenitic Stainless
Steels in BWR Environments,” was published in March 2008. Experimental data are presented on the
fracture toughness and crack growth rates (CGRs) of wrought and cast
austenitic stainless steels, including weld heataffected-zone materials, that
were irradiated to fluence levels as high as ≈ 2 x 1021 n/cm2 (E > 1
MeV) (≈ 3 dpa) in a boiling heavy water reactor at 288-300°C. The effects of material composition,
irradiation dose, and water chemistry on CGRs under cyclic and stress
corrosion cracking conditions were determined. The effects of neutron irradiation on the
fracture toughness of these steels, as well as the effects of material and
irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that
bounds the existing data has been defined.
The synergistic effects of thermal and radiation embrittlement of cast
austenitic SS internal components have also been evaluated. 8. NUREG/CR-6945 NUREG/DR-6945, “Fabrication Flaw Density and Distribution in Repairs
to Reactor Pressure Vessel and Piping Welds,” was published in April
2008. This report describes the
fabrication flaw distribution and characterization in the repair weld metal
of vessels and piping. Welds from four
different reactor pressure vessels and a collection of archived pipes were
studied to develop empirical estimates of fabrication flaw densities. This work indicates that large flaws occur
in these repairs, and that the repair flaws are complex in composition and
sometimes include cracks on the ends of the repair cavities. A generalized fabrication flaw distribution
for the population of nuclear reactor pressure vessels and for piping welds
in Construction records were evaluated relative to fabrication processes and product forms used. An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance. |
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