[Federal Register Volume 60, Number 243 (Tuesday, December 19, 1995)]
[Rules and Regulations]
[Pages 65456-65476]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-30665]
[[Page 65455]]
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Part IV
Nuclear Regulatory Commission
_______________________________________________________________________
10 CFR Part 50
Fracture Toughness Requirements for Light Water Reactor Pressure
Vessels; Final Rule
Federal Register / Vol. 60, No. 243 / Tuesday, December 19, 1995 /
Rules and Regulations
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[[Page 65456]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AD57
Fracture Toughness Requirements for Light Water Reactor Pressure
Vessels
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations for light-water-cooled nuclear power plants to clarify
several items related to the fracture toughness requirements for
reactor pressure vessels (RPV). The amendments will clarify the
pressurized thermal shock (PTS) requirements, make changes to the
Fracture Toughness Requirements and the Reactor Vessel Material
Surveillance Program Requirements, and provide new requirements for
thermal annealing of a reactor pressure vessel.
EFFECTIVE DATE: January 18, 1996.
FOR FURTHER INFORMATION CONTACT: Alfred Taboada, Division of
Engineering Technology, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-00001, telephone:
(301) 415-6014.
SUPPLEMENTARY INFORMATION: On October 4, 1994 (59 FR 50513), the NRC
published in the Federal Register a proposed amendment to clarify
several items related to fracture toughness requirements for reactor
pressure vessels (RPV) and to add a new section on thermal annealing of
a reactor vessel to 10 CFR Part 50.
Background
Maintaining the structural integrity of the reactor pressure vessel
of light-water-cooled reactors is a critical concern related to the
safe operation of nuclear power plants. To assure the structural
integrity of RPVs, NRC regulations and regulatory guides have been
developed to provide analysis and measurements methods and procedures
to establish that each RPV has adequate safety margin for continued
operation. Structural integrity of a RPV is generally assured through a
fracture mechanics evaluation, including measurement or estimation of
the fracture toughness of the materials which compose the RPV. However,
the fracture toughness of the RPV materials varies with time. As the
plant operates, neutrons escaping from the reactor core impact the
vessel beltline materials (e.g. the materials that surround the reactor
core), causing embrittlement of those materials. The NRC's regulations
and regulatory guides related to RPV integrity provide the criteria and
methods needed to estimate the extent of the embrittlement, to evaluate
the consequences of the embrittlement in terms of the structural
integrity of the RPV, and to provide methods to mitigate the
deleterious effects of the embrittlement.
The NRC has several regulations and regulatory guides that
establish criteria and procedures for assuring the structural integrity
of RPVs. With the addition of the thermal annealing requirements in
this rule and several regulatory guides, the regulatory documents
contribute to a comprehensive set of regulations and regulatory
guidance pertaining to RPV integrity.
This final rule adds requirements for thermal annealing of the RPV
as a method for mitigating the effects of neutron irradiation (10 CFR
50.66) and amends the following:
1. The Pressurized Thermal Shock (PTS) rule (10 CFR 50.61).
2. Appendix G of 10 CFR Part 50, ``Fracture Toughness
Requirements.''
3. Appendix H of 10 CFR Part 50, ``Reactor Vessel Material
Surveillance Program Requirements.''
Overview of the Final Rule
PTS Rule (10 CFR 50.61)
This amendment to the PTS rule makes three changes:
1. The rule incorporates in total, and therefore makes binding by
rule, the method for determining the reference temperature, RTNDT,
including treatment of the unirradiated RTNDT value, the margin
term, and the explicit definition of ``credible'' surveillance data,
which is currently described in Regulatory Guide 1.99, Revision 2.
2. The section is restructured to improve clarity, with the
requirements section giving only the requirements for the value for the
reference temperature for end of life fluence, RTPTS. The method
for calculating RTPTS is moved to a new paragraph of the rule.
3. Thermal annealing is identified as a method for mitigating the
effects of neutron irradiation, thereby reducing RTPTS.
Thermal Annealing Rule (10 CFR 50.66)
The thermal annealing rule, 10 CFR 50.66, provides a consistent set
of requirements for the use of thermal annealing to mitigate the
effects of neutron irradiation and replaces the requirements for
annealing in the current Appendix G of 10 CFR Part 50. The final rule
requires, prior to initiation of thermal annealing, submittal of a
Thermal Annealing Report containing: (1) A Thermal Annealing Operating
Plan, (2) a Requalification Inspection and Test Program, (3) a Fracture
Toughness Recovery and Reembrittlement Trend Assurance Program, and (4)
Identification of Unreviewed Safety Questions and Technical
Specifications Changes. The report must be submitted at least 3 years
before the date at which the limiting fracture toughness criteria in
50.61 and Appendix G to Part 50 would be exceeded. This 3-year period
is specified to provide the NRC staff with sufficient time to review
the thermal annealing program. Under Sec. 50.66(a), the NRC will,
within three years of submission of a licensee's Thermal Annealing
Report, document its views on the plan, including whether thermal
annealing constitutes an unreviewed safety question.
In order to provide for public participation in the regulatory
process, Section 50.66(f)(1) requires that the NRC hold a public
meeting a minimum of 30 days before the licensee starts to thermal
anneal the reactor vessel. The Commission will notify and solicit
comments from cognizant local and state governments, and will publish a
notice in the Federal Register and in a forum, such as local
newspapers, which is readily accessible to individuals in the vicinity
of the site, in order to solicit comments from the public.
The thermal annealing operating plan must include an evaluation of
the effects of temperature, and of mechanical and thermal stresses on
the reactor and associated equipment such as containment, the
biological shield, and attached piping, to demonstrate that the
operability of the reactor will not be detrimentally affected. The
bounding conditions of the temperatures and times used in this analysis
define the proposed annealing conditions. If these conditions are
exceeded during the vessel annealing, then the evaluation would no
longer be valid, and the acceptability of the actual vessel annealing
would have to be demonstrated as discussed below in the next paragraph.
Upon completion of the thermal annealing, the licensee must confirm
in writing to the Director, Office of Nuclear Reactor Regulation (NRR),
that the thermal annealing was performed in accordance with the Thermal
Annealing Operating Plan and the Requalification Inspection and Test
Program. Within 15 days of the licensee's written confirmation that the
thermal annealing was completed in accordance with the
[[Page 65457]]
Thermal Annealing Plan, and prior to restart, the NRC shall: (1)
Briefly document whether the thermal annealing was performed in
compliance with the licensee's Thermal Annealing Operating Plan and the
Requalification Inspection and Test Program, with the documentation to
be placed in the NRC public document room, and (2) hold a public
meeting to: (1) permit the licensee to explain the results of the
reactor vessel annealing to the NRC and the public, (2) allow the NRC
to discuss its inspection of the reactor vessel annealing, and (3)
provide an opportunity for the public to comment to the NRC on the
thermal annealing. The licensee may restart its reactor after the
meeting has been completed, unless the NRC orders otherwise. Within 45
days of the licensee's written confirmation that the thermal annealing
was completed in accordance with the Thermal Annealing Operating plan
and the Requalification Inspection and Test Program, the NRC staff
shall complete full documentation of the NRC's inspection of the
licensee's annealing process and place the documentation in the Public
Document Room.
If the thermal annealing was completed but not performed in
accordance with the Thermal Annealing Operating Plan and the
Requalification Inspection and Test Program, including the bounding
conditions of the temperature and times as discussed above, the
licensee must submit a summary of lack of compliance and a
justification for subsequent operations. The licensee must also
identify any changes to the facility which are attributable to the
noncompliances which constitute unreviewed safety questions and any
changes to the technical specifications which are required for
operation as a result of the noncompliances. This identification does
not relieve the licensee from complying with applicable requirements of
the Commission regulations and the operating license, and if, as a
result of the annealing operation, these requirements cannot be met,
the licensee must obtain the appropriate exemption per 10 CFR 50.12. If
unreviewed safety questions or changes to technical specifications are
not identified as necessary for resumed operation, the licensee may
restart after the NRC staff places a summary of its inspection of the
thermal annealing in the Public Document Room, and the NRC holds a
public meeting on the thermal annealing. On the other hand, if
unreviewed safety questions or changes to technical specifications are
identified as necessary for resumed operation, the licensee may restart
only after the Director of NRR authorizes restart, the summary of the
NRC staff inspection is placed in the public document room, and a
public meeting on the thermal annealing is held.
The final Thermal Annealing Rule also sets forth the requirements
that a licensee must follow if the thermal annealing was terminated
prior to completion. In general, the process and requirements for
partial annealing are analogous to the situations where the thermal
annealing was completed; viz., where the partial annealing was
otherwise performed in compliance with the Thermal Annealing Operating
Plan and relevant portions of the Requalification Inspection and Test
Program, the licensee submits written confirmation of such compliance
and may restart following, inter alia, holding of a public meeting on
the annealing. By contrast, where the partial annealing was not
performed in accordance with the Thermal Annealing Operating Plan and
relevant portions of the Requalification Inspection and Test Program,
the licensee is required to submit a summary of lack of compliance and
a justification for subsequent operations, and identify any changes to
the facility which are attributable to the noncompliances which
constitute unreviewed safety questions and changes to the technical
specifications which are required for operation as a result of the
noncompliances with the Thermal Annealing Operating Plan and relevant
portions of the Requalification Inspection and Test Program. If
Unreviewed Safety Questions and/or changes to technical specifications
are identified as necessary for resumed operation, the licensee may
restart only after the Director of NRR authorizes restart and the
public meeting on the thermal annealing is held.
Every licensee that either completes a thermal annealing or
terminates an annealing but elects to take full or partial credit for
the annealing shall provide a Thermal Annealing Results Report
detailing: (1) The time and temperature profile of the actual thermal
anneal, (2) the post-anneal RTNDT and Charpy upper shelf energy
values of the reactor material to be used in subsequent operations, (3)
the projected post-anneal reembrittlement trends for both RTNDT
and Charpy upper-shelf energy, and (4) the projected values of
RTPTS and Charpy upper-shelf energy at the end of the proposed
period of operation addressed in the application. The report must be
submitted within three months of completing the thermal anneal, unless
an extension is authorized by the Director, NRR.
Two items of particular importance to the overall annealing are the
recovery of fracture toughness and the degree of reembrittlement of the
RPV beltline materials. This final rule provides alternative methods
for determining these values, ranging from assessments using plant-
specific materials to an assessment using a generic computation.
Two methods provided for evaluating annealing recovery are
experimental methods to determine plant-specific annealing recovery,
and a third method is a generic computational method. Experimental
methods and the computational method are also provided for estimating
recovery of RTNDT and Charpy upper-shelf energy of the beltline
materials. The experimental methods for estimating recovery of
RTNDT and the Charpy upper-shelf energy utilize either
surveillance program specimens or material removed from the vessel
beltline. The experimental methods provide a plant-specific estimate of
recovery, rather than the generic value evaluated from the
computational method. This final rule requires that surveillance
specimens from ``credible'' surveillance programs must be used to
develop plant-specific recovery data, if such specimens are available.
This final rule does not require the removal of material from the RPV
beltline to permit plant-specific evaluation of recovery.
As described previously, the computational method requires
appropriate justification.
Post anneal reembrittlement trends of both the RTNDT and the
Charpy upper shelf energy must be estimated and monitored using a
surveillance program described in the Thermal Annealing Report.
The reactor pressure vessel is perhaps the most important single
component in the reactor coolant system. As such, ensuring its
integrity is a fundamental element of plant safety. Thermal annealing
is a positive action that could be taken to reduce the level of
embrittlement in the pressure vessel beltline and, thereby, improve the
ability of a pressure vessel to withstand accident loadings. While
thermal annealing is a positive action, there are numerous complex
technical questions regarding its application in the U.S. that are
unanswered.
Thermal annealing of a commercial reactor pressure vessel has never
been accomplished in the United States. Thermal annealing has been
successfully employed in Eastern Europe and Russia on Russian-designed
pressure vessels. However, there are significant differences between
the U.S. and Russian designs in terms of the
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geometry of the pressure vessels, the attached piping, and the
surrounding structures. The staff has observed one of these annealing
operations. While informative, the East European and Russian experience
does not provide answers to all of the potential questions related to
annealing of U.S. designed pressure vessels.
Research analyses performed previously indicated the potential for
plastic deformation of the main coolant piping for a typical U.S. plant
design and anticipated annealing conditions. There are also questions
regarding how thermal growth of the pressure vessel is treated, and the
adequacy of the thermal and stress analyses used to predict response of
the overall system under thermal annealing conditions. Additionally,
there may be questions in other areas such as temperature limits for
the concrete structures, and potential radiological hazards associated
with removing and storing the reactor internals during the annealing
process, and fire hazards associated with heating the vessel.
Recognition of the numerous complex technical questions related to
thermal annealing, and of the potential benefits for operating nuclear
power plants, has resulted in a cooperative effort, funded by the U.S.
Department of Energy and the industry, to perform Annealing
Demonstration Projects. Projects are planned to demonstrate two
different annealing processes, evaluating heater designs and vessel
designs. It is anticipated that the annealing demonstration projects
will answer many of the generic questions regarding thermal annealing
of U.S. pressure vessel and piping designs.
The thermal annealing report, required by the thermal annealing
rule, is designed to facilitate a detailed review by the licensee of
plant-specific questions and considerations in performing a thermal
annealing. The proposed rule specifically discusses the potential for
unreviewed safety questions and technical specification changes that
may result from or be related to thermal annealing of the reactor
pressure vessel. With completion of the demonstration projects and as
the staff and industry gain experience with thermal annealing, many of
the issues related to annealing will be better understood and related
questions will be answered. However, until this experience is realized,
the staff will critically review licensee determinations regarding
unreviewed safety questions and the need for technical specification
changes associated with each proposed thermal annealing.
The thermal annealing rule has been structured to provide time for
the staff to thoroughly review the licensee's annealing plan and
determination regarding unreviewed safety questions and the need for
technical specification changes. If the staff identifies an unreviewed
safety question or the need for a technical specification change, the
licensee would be so notified and the existing NRC regulatory practices
would be invoked to address the issues.
Appendix G of 10 CFR Part 50
Appendix G of 10 CFR Part 50 specifies fracture toughness
requirements for ferritic materials of pressure-retaining components of
the reactor coolant pressure boundary of light-water-cooled nuclear
power reactors. These requirements provide adequate margins of safety
during any condition of normal operation, including anticipated
operational occurrences and system hydrostatic tests. The amendments to
Appendix G are principally of a clarifying or a restructuring nature.
Requirements for ``volumetric inspection'' and ``additional evidence of
fracture toughness'' have been removed because they were unnecessary,
given the inspection and performance demonstration programs currently
required under 10 CFR 50.55a. The ``additional evidence of fracture
toughness'' requirement in Section V.C.2 is incorporated in the
``equivalent margins'' analysis in Section IV.A.1 as a provisional
method for developing fracture toughness data needed for that analysis.
The pressure-temperature and minimum permissible temperature
requirements in Section IV have been restructured. The principal
feature is the addition of a table which summarizes the pressure-
temperature limit requirements and minimum temperature requirements as
a function of the plant operating condition, the vessel pressure,
whether fuel is in the vessel, and whether the core is critical. In
addition, Section IV has been reworded to clarify the minimum
permissible temperature requirement by indicating the criteria for use
in determining the location in the component or material which must
satisfy the minimum temperature requirement. This minimum temperature
is defined in Section IV as the metal temperature of the controlling
material in the region which has the least favorable combination of
stress and temperature for the appropriate plant condition. An explicit
statement has been added to require that pressure and leak tests of the
reactor pressure vessel required by Section XI of the American Society
of Mechanical Engineers Boiler & Pressure Vessel (B&PV) Code (ASME
Code) must be completed before the core is critical.
The requirement that all pressure and leak tests of the RPV
required by Section XI of the ASME Code must be completed before the
core is critical is intended to prohibit the use of nuclear heat, i.e.,
core criticality, in the conduct of ASME, Section XI pressure and leak
tests. The use of nuclear heat before the completion of such tests is
not consistent with basic defense-in-depth nuclear safety principle for
several reasons, including the hindrance of finding leaks with the
vessel at such a high temperature and the potential for exacerbating
the consequences of a vessel rupture (in the extremely unlikely event
that it should occur) by having the core critical. The explicit
prohibition of nuclear heat in these cases was discussed in a letter to
Messrs. Reynolds and Stenger of the Nuclear Utility Backfitting and
Reform Group from James M. Taylor, Executive Director of Operations,
dated February 2, 1990.
The current requirements in 10 CFR Part 50, Appendix G, Section V.
D. with respect to reactor vessel thermal annealing are being replaced
by a sentence which references the new Thermal Annealing rule, 10 CFR
50.66.
Appendix H of 10 CFR Part 50
Appendix H of 10 CFR Part 50, ``Reactor Vessel Material
Surveillance Program Requirements'' provides the rules for monitoring
the changes in the fracture toughness properties of the RPV beltline
materials due to irradiation embrittlement using a surveillance
program. Appendix H references American Society for Testing and
Materials (ASTM) standard E 185 (``Standard Practice for Conducting
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor
Vessels'') for many of the detailed requirements of surveillance
programs, and permits the use of integrated surveillance programs,
wherein surveillance program capsules for one reactor are irradiated in
another reactor.
Integrated surveillance programs are permitted under Section II.C
of Appendix H of 10 CFR Part 50. One provision of this section is that
``the amount of testing may be reduced if the initial results agree
with predictions.'' This provision was deleted, although previous
authorizations granted by the Director, Office of Nuclear Reactor
Regulation, continue in effect.
A second change to Appendix H restructures Section II.C to clarify
the
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requirements for integrated surveillance programs.
The other principal change to Appendix H clarifies the version of
ASTM Standard E 185 that applies to the various portions of the
surveillance programs. Appendix H recognizes the need to separate
surveillance programs into two essential parts, specifically the design
of the program and the subsequent testing and reporting of results from
the surveillance capsules. Because the design of the surveillance
program cannot be changed once the program is in place, the
requirements for design of the surveillance program are static for each
plant. However, the testing and reporting requirements are updated
along with technical improvements made to ASTM standard E 185.
Request for Public Comments
At the request of the Commission, the proposed rule contained a
request for public comments on the following specific issues related to
the proposed regulation on thermal annealing:
1. The technical adequacy of the staff's guidance;
2. The sufficiency of the guidance and criteria to support a
certification that if satisfied, a plant with an annealed vessel can
safely resume operation;
3. Whether health and safety concerns are best served by approval
of the thermal annealing plan or of readiness for restart;
4. The preferred regulatory process (including opportunities for
public participation) and the commenter's basis for recommending a
particular process; and
5. Whether there are health and safety issues concerning thermal
annealing that cannot be addressed generically and would warrant plant-
specific consideration.
The supplementary information section of the proposed rule also
discussed the issue of opportunity for public participation in
regulating thermal annealing of pressure vessels.
The response to the request for public comments on these issues,
along with other items, are summarized below.
Summary of Comments
The following includes a summary of the comments received on the
proposed rule, on the five issues identified by the Commission, and on
the options for public participation in thermal annealing.
Comments were received from nine separate sources. These sources
consist of five utilities, the Nuclear Energy Institute (NEI), the
Nuclear Utility Backfitting and Reform Group (NUBARG) represented by
the firm Winston & Strawn, one public citizens group (Ohio Citizens for
Responsible Energy (OCRE)), and one nuclear steam system supplier
(NSSS).
NEI provided detailed comments on 10 CFR 50.61, 10 CFR 50.66,
Appendix G to 10 CFR Part 50, and Appendix H to 10 CFR Part 50,
responded to the request for comments on the five issues related to
thermal annealing and included detailed comments on the opportunities
for public participation. The five utilities and the NSSS endorsed the
NEI comments. Three of the five utilities provided additional comments
on 10 CFR 50.61; one of the five utilities provided additional comments
on 10 CFR Part 50, Appendix G; two of the utilities provided additional
comments on 10 CFR Part 50, Appendix H; and one of the five utilities
disagreed with the NEI position on the opportunity for public
participation and submitted a separate comment. OCRE provided comments
on the opportunity for public participation. NUBARG provided comments
on the backfitting aspects of the proposed rule and the staff's backfit
justification.
NEI and one of the utilities included comments on the Draft
Regulatory Guide DG-1027, ``Format and Content of Application for
Approval for Thermal Annealing of Reactor Pressure Vessels,'' that was
discussed in the proposed rule. These comments on Draft Regulatory
Guide DG-1027 are being reviewed by the NRC staff and will be addressed
separately in the resolution of comments on the regulatory guide.
The NRC reviewed the comments received on the proposed rule, the
comments on the five questions related to thermal annealing and the
issue of opportunities for public participation. The resolution of
these comments is presented below.
PTS Rule (10 CFR 50.61)
Sixteen specific comments in the submittals from NEI and three
utilities addressed 10 CFR 50.61. A general comment argued that both
the existing 10 CFR 50.61 and the proposed modifications contained an
excessive amount of prescriptive technical detail that limits licensee
compliance flexibility. The commenters proposed that these prescriptive
technical details be removed from the rule and placed in a regulatory
guide. These commenters suggested that the rule not be issued until it
has been written to contain only those requirements essential to
regulate reactor pressure vessel embrittlement. A number of comments
suggested changes that were clarifications to the proposed rule,
including proposals to clarify the procedure for calculating the
reference temperatures in the preservice condition, RTNDT, and, at
end of reactor life, RTPTS. One comment noted that the proposed
rule omitted part of the procedure in Regulatory Guide 1.99, presently
being applied by the NRC, that permits adjustments for differences in
chemistry between surveillance material and the vessel material when
using credible surveillance data to calculate a best fit chemistry
factor for transition temperature shifts due to irradiation. Several
comments proposed changes in the criteria for establishing whether
surveillance material data is credible that would result in a less
restrictive basis for using surveillance data in determining the
transition temperature shift. The comments argued that the proposed
rule is ambiguous with respect to the use of information from other
sources that contain limiting material for a specific plant and that
the NRC must have the flexibility to approve use of such information on
a case-by-case basis. Several comments proposed limiting the basis for
making changes of RTPTS subject to the approval of the Director,
NRR.
The NRC recognizes that 10 CFR 50.61 contains an unusual amount of
prescriptive material and that the comments proposing simplification
have merit. Some changes to the rule have been made to provide
flexibility, where appropriate. The NRC staff is evaluating subsequent
changes that would be more performance based. However, the NRC staff
believes that this rule, as written, is needed to ensure that plants
apply the appropriate method for determining RTPTS and that the
appropriate reference to the thermal annealing rule be applied for the
pressurized thermal shock situation.
A number of clarifications were made to the rule. The paragraphs
dealing with the determination of RTPTS were modified to make
clear that RTPTS is a unique, end of life, case of RTNDT and
to clarify the procedure for determining these values. As suggested,
the adjustment procedure was added to the rule to permit accounting for
differences in chemistry between surveillance materials and reactor
vessel materials when calculating chemistry factors. With respect to
the plant specific material surveillance data that is permitted to be
used in a surveillance program, the rule was modified to make clear
that such data includes results from other plant's surveillance
programs and test reactors. Several clarifications were made to the
criteria for determining credible material. The NRC determined that the
requirements for approval by the Director, NRR, for
[[Page 65460]]
changes in RTPTS are appropriate and should not be modified.
Thermal Annealing Rule (10 CFR 50.66)
Twelve individual comments were received on the proposed Thermal
Annealing Rule, 10 CFR 50.66. These comments included a number of
suggestions for clarification of details of the proposed rule. Three of
the comments addressed the requirements that, after the annealing
operation, the reembrittlement rate of the reactor vessel due to
neutron irradiation must be estimated and must be monitored using a
surveillance program which conforms to Appendix H of 10 CFR 50,
``Reactor Vessel Materials Surveillance Program.'' The comments are
summarized as follows:
(1) The supplementary information section for the proposed rule is
silent on what is acceptable if limiting material is not available. The
rule should provide appropriate requirements on the method for
monitoring reembrittlement after annealing for those plants that do not
have limiting material for their surveillance program and the
monitoring plans should be consistent with the preannealing
surveillance program approved by the NRC staff;
(2) Appendix H does not define an acceptable post-anneal
surveillance program, the reference to Appendix H should be deleted,
and the post-anneal surveillance program should be defined in the
annealing plan that is approved by the staff; and
(3) The term reembrittlement rate is unclear as to the period of
time to be used for its determination, and a wording change is proposed
for the requirement that would relate change in toughness to fluence
accumulated after the anneal.
Three of the comments addressed the requirements in the proposed
rule that the Thermal Annealing Operation Plan include time-temperature
profiles which represent the annealing conditions that may not be
exceeded during the annealing operation and are to be used for
determining the amount of recovery of the fracture toughness of the
material due to annealing. The comments suggested that, instead of a
single time-temperature profile, bounding time and temperature
conditions be established for the maximum values that would be used for
thermal and stress analysis and to verify the re-qualification
inspection and test program, and the minimum values that would be used
to establish the amount of recovery of fracture toughness and for
reembrittlement rate estimates. The bounding values would be based upon
the estimated uncertainties in the times and temperatures and the
actual annealing conditions should fall within these bounds.
Two comments addressed the section on Certification of Annealing
Effectiveness. One comment suggested deleting the requirement in the
proposed rule for certification of the annealing effectiveness and
instead adding a provision in the Thermal Annealing Operating Plan that
approval prior to subsequent power operation be required only if the
anneal was not performed in accordance with the approved plan. The
comment also suggested that, if the licensee terminates the annealing
before achieving the specified time but otherwise maintains the
annealing envelop such that no concern exists for stress or thermal
damage, no additional constraints be imposed on subsequent operations
and no credit be given for annealing. The second comment suggested that
(1) the staff's review of the annealing report (certification report)
need not be completed prior to reinitiating power operation if the
anneal was performed in accordance with the approved Thermal Annealing
Operating Plan, (2) reporting and quantification of the actual recovery
results need not be reported unless the vessel was at or above the PTS
screening criteria when annealing was started, and (3) the Thermal
Annealing Operating Plan should specify the minimum content and a
schedule for reporting the annealing results. The commenter provided a
proposed list of criteria, content, and schedule for reporting the
annealing results.
One comment stated that no guidance was provided in the proposed
rule on what constitutes components ``affected'' by the annealing
operation that are required to be reported in the Thermal Annealing
Operating Plan. The comment suggested alternative wording that
components to be reported should be structures and components that are
expected to experience significant temperature gradient or stress
variations during the thermal annealing operation. One comment
suggested qualifying the provision in the proposed rule that the
effects of localized high temperatures must be evaluated for changes in
thermal and mechanical properties of the reactor vessel insulation for
those cases where such changes may be negligible at annealing
conditions. One comment suggested that the use of applicable material
data, such as data from integrated surveillance programs, be an
optional part of the computational methods for determining fracture
toughness recovery.
The NRC reviewed the comments received on the proposed rule in
detail. After consideration, the NRC reached the conclusion that most
of the comments are not inconsistent with the intent of the proposed
rule and in some cases reflect a need for clarification of the rule. In
these cases, alternative wording that clarified the intent of the rule
was substituted in the text. With respect to the comments on the
requirement that reembrittlement rate after annealing must be monitored
using a surveillance program, the NRC is aware that some plants do not
have limiting materials for their existing preannealing surveillance
programs. For these situations the staff has approved alternative
surveillance plans on a case-by-case basis. Clearly, these plants will
not have limiting material for surveillance programs for use in
determining reembrittlement rates after annealing.
The NRC recognizes that Appendix H of 10 CFR Part 50, which is
referenced in this rule, does not specifically address the surveillance
of an annealed reactor vessel. However, the requirements of Appendix H
to 10 CFR Part 50 apply to all reactors including the specific case of
an annealed reactor vessel. To clarify the surveillance requirements of
an annealed plant, the final rule has been modified to include, as
suggested, that the post-anneal reembrittlement is to be monitored
using a surveillance program defined in the Thermal Annealing Report
and that the surveillance program must conform to the intent of
Appendix H to 10 CFR Part 50.
The term reembrittlement ``rate'' in the proposed rule was intended
to mean the projected amount of reembrittlement over a specific fluence
period. It is recognized that reembrittlement is not a straight line
function of fluence. Determination of reembrittlement rate is discussed
in more detail in Draft Regulatory Guide 1.162, ``Format and Content of
Report for Thermal Annealing of Reactor Pressure Vessels.'' In
Regulatory Guide 1.162, the approved method for estimating the
reembrittlement rate, the lateral shift method, results in the same
embrittlement trend as that used for the pre-anneal operating period.
To avoid confusion the term ``rate'' has been changed to ``trend'' in
the final rule and the regulatory guide.
The NRC agrees with the comments that the time and temperature
profile required in the annealing operating plan should be bounding
values. In this regard, Regulatory Guide DG-1027 calls for the thermal
annealing operating plan to include identification of the
[[Page 65461]]
limitations and permitted variations in temperature, time, heatup and
cooldown rate. For clarification, the final rule has been modified to
use the terms ``bounding conditions for times and temperatures and
heatup and cooldown schedules'' to describe conditions that may not be
exceeded during the annealing operation, and the lower limit time and
temperature of the actual anneal is used for determining the projected
recovery of fracture toughness by annealing.
The NRC considers that the intent of paragraphs (c), Completion or
Termination of Thermal Annealing, and (d), Thermal Annealing Results
Report, of the final rule to be consistent with the two comments on
that subject. The final rule does not require that the NRC approve
restart following the annealing operation if the Thermal Annealing
Operating Plan and the Requalification Inspection and Test Program was
complied with. The NRC accepts the suggestion that the rule should be
more specific on the items the licensee should include in the report
and has included the list in the final rule.
Finally, the NRC agrees with the suggestion to make clear that a
report is not required if:
(1) The licensee terminates the anneal prior to completion;
(2) The partial anneal was otherwise in accordance with the Thermal
Annealing Plan;
(3) The licensee does not elect to take credit for any recovery. A
statement was added to the Final Rule to cover the early termination
situation.
The NRC has accepted the suggested clarifications of what
constitutes an ``affected'' component and the qualification on the
requirement to evaluate changes in properties on reactor vessel
insulation if these are negligible. The NRC considers it unnecessary to
include a reference in the rule to data from integrated surveillance
programs as an optional part of the computational methods to determine
fracture toughness recovery. Generic computational methods for this
purpose are provided in the Regulatory Guide 1.162. However, the final
rule does not prohibit use of alternative methods if adequate
justification is provided.
Appendix G to 10 CFR Part 50
Two comments were received on the Appendix G to 10 CFR Part 50 of
the proposed rule. The NEI comment, which was endorsed by five
utilities and one NSSS organization, included a table with six items on
Appendix G. The other comment on Appendix G was received from one of
the five utilities. Two of the comments identified typographical errors
and suggested a change in organization to improve clarity. One of the
comments suggested revising the rule to change the definition of
reference temperature, RTNDT, for cases where plants do not have
data to comply with code procedures for determining RTNDT. One
comment suggested a change in the title of Table 1, ``Pressure and
Temperature Requirements,'' by adding to the title ``For the Reactor
Pressure Vessel'' to make clear that this table does not apply to other
components in the reactor coolant pressure system and proposed adding a
footnote to the table for the same purpose. One comment identified an
error in the minimum temperature requirements for the hydrostatic and
leak testing of the pressure vessel without fuel when the vessel
pressure is equal or below 20 percent of the vessel design pressure.
One of the comments suggested that two of the entries in the table were
new requirements when the table was intended to provide clarification.
The utility's comment disagreed with the proposed rule change to
prohibit the use of nuclear heat for the performance of vessel leak and
hydrostatic testing. The utility contended that using nuclear heat, by
providing a significant temperature margin above the pressure and
temperature limit curves, greatly reduces the probability of brittle
fracture and should be allowed.
The NRC corrected the typographical errors and corrected the
minimum temperature requirement for the hydrostatic and leak testing of
the pressure vessel at low vessel pressures and without fuel. The title
to Table 1 was changed, as suggested, for clarification.
The NRC does not agree with the proposal to change the definition
of RTNDT. The situation described in the comment, when data is not
available to comply with code procedures, is presently handled on a
case-by-case basis in accordance with MEB Branch position, MEB 5-2. The
NRC staff does not agree with the comment that the two requirements
cited are new requirements. Item 2.2.c. and Item 2.2.d of Table 1 are
in the existing ASME code requirement and in Paragraph IV.A.3. in the
rule. The NRC also does not agree with the utility's comment that using
nuclear heat greatly reduces the probability of brittle fracture. The
reasons for this are set forth in the February 2, 1990, letter to
Messrs. Reynolds and Stenger of NUBARG from James M. Taylor, Executive
Director for Operations.
Appendix H to 10 CFR Part 50
Three comments were received on Appendix H to 10 CFR 50. The
comment from NEI was endorsed by the five utilities and the NSSS. Two
of the five utilities submitted additional comments. NEI and one
utility commented that the proposed change to Paragraph III.B.1, which
establishes the applicable edition of ASTM standard E 185 for a reactor
surveillance program, constituted a backfit that would require a
substantial design change in the surveillance program for those plants
fabricated to a code edition prior to 1973. The other two commenters
suggested new changes to Appendix H to 10 CFR Part 50. One of the
commenters noted that an existing provision in Appendix H to 10 CFR
Part 50, not part of the proposed rule change, dealing with
requirements for attaching capsule holders to the vessel wall is a
reiteration of a requirement in the ASME Code and should be removed.
The other commenter suggested a new change to Appendix H to 10 CFR Part
50 to add a statement to the criteria for approval of an integrated
surveillance program that would permit the use of surveillance
specimens for extension of license purposes. The commenter also
suggested that there is an apparent conflict between Paragraph III.C.2.
and Paragraph III.C.3. that address requirements for an integrated
surveillance.
The provision in the proposed rule was changed and reference to
ASTM E 185 73 was deleted to make clear that the surveillance programs
must be designed to the edition of ASTM 185 that is current on the
issue date of the ASME Code to which the reactor vessel was purchased
or to a later edition through 1982. The Commission agrees with the
industry comments that imposing the ASTM E 185 1973 edition is
impractical because vessels purchased prior to 1973 could not
necessarily comply with all of the surveillance requirements in the
1973 edition of the ASTM standard. The NRC staff believes that the
provision in the present rule on requirements for attaching capsule
holders to the reactor vessel wall is required for clarity and should
not be deleted. The comments related to the requirements for an
integrated surveillance program were not persuasive to the NRC staff.
The existing provisions of the rule do not preclude the application of
the integrated surveillance program for extension of license purposes.
The two paragraphs purported to be in conflict address separate items;
one addresses the number of materials to be irradiated,
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specimen types, and number of specimens per reactor; the other
addresses amount of testing.
Request for Comments on Issues Related to Thermal Annealing
Comments were received from NEI on the five issues on thermal
annealing that were included in the proposed rule at the Commission's
direction. In addition, OCRE and one utility, Pacific Gas and Electric,
submitted comments on Issue 4, concerning the preferred regulatory
process (including opportunity for public participation). Public
Comments on the five issues are summarized below:
Issue 1: The technical adequacy of the NRC staff's guidance.
Comment: The detailed comments submitted on 10 CFR 50.66 are
summarized in the Summary of Comments section on the Thermal Annealing
Rule. In addition, NEI suggested that draft Regulatory Guide, DG-1027,
be revised to include acceptance criteria where an action is required,
but the acceptance criteria was not defined. NEI further commented that
the re-embrittlement rate equation (DG-1027, Equation 1) appeared to be
very conservative and would result in a post-anneal operating life that
is less than industry believes justified.
Response: The NRC is concurrently revising the noted draft
regulatory guide and will address this comment in the resolution of
comments for the guide.
Issue 2: The sufficiency of the guidance and criteria to support a
certification that if satisfied, a plant with an annealed vessel can
safely resume operation.
Comment: NEI noted that ``The reactor pressure vessel thermal
annealing rule and guide address appropriate issues to assure public
health and safety and that the annealed reactor pressure vessel may be
safely operated. The prior NRC staff approval of the reactor vessel
annealing plan assures a clear process and criteria to restart
following the vessel anneal. The licensee needs only to attest to
compliance with the approved plan prior to resuming operations. The
resumption of operations should not be needlessly delayed while a
report documenting performance of the vessel anneal and recovery of the
embrittled material properties is confirmed, because the vessel anneal
will only improve the material properties. The final report should be
submitted on a schedule that considers when the vessel would have
exceeded the RTPTS or uppershelf energy (USE) screening criteria
without an anneal. The material property recovery will document prior
to the time when the vessel would have exceeded the screening criteria,
thereby assuring that the vessel is safe to operate at restart and for
the duration justified by the material embrittlement recovery.''
Response: NRC agrees with the NEI comment, except NRC believes it
is necessary for the licensee to submit the final report within three
months of completing or terminating the anneal, unless an extension is
authorized by the Director, Office of Nuclear Reactor Regulation.
Issue 3: Whether health and safety concerns are best served by
approval of the thermal annealing plan or of readiness for restart.
Comment: NEI noted that ``The performance of a reactor pressure
vessel anneal in accordance with an approved annealing plan improves
the public health and safety by reducing the probability of core melt
frequency. This improvement occurs because of the increase in reactor
vessel material ductility. The amount of recovery achieved by a thermal
anneal will be documented prior to the original date when the reactor
vessel would have exceeded the PTS or USE screening limit. Therefore, a
demonstration for ``restart readiness'' is an extra burden that will
not provide any further improvement of the public health and safety.''
Response: The NRC's determination as to the procedures for NRC
review of the Thermal Annealing Operation Plan, Requalification
Inspection and Test Program and justification for restart discussed
below in further detail in the Opportunities for Public Participation
section.
Issue 4: The preferred regulatory process (including opportunities
for public participation) and the commenter's basis for recommending a
particular process.
Comment: NEI noted that ``The industry recommends that a hearing
opportunity be provided, but that it be a non-adjudicatory, 10 CFR Part
2, Subpart L type hearing on the docketed record. The essential
features of the hearing process proposed are as follows. The NRC would
at time of receiving the licensee proposed annealing plan issue a
Federal Register announcement that staff is performing the review per
10 CFR 50.66. A Subpart L hearing could be held, if requested by an
intervener, after the NRC staff has issued a safety evaluation report
on the licensee annealing plan, but prior to commencement of the
reactor vessel thermal annealing unless the NRC staff makes a ``no
significant hazards determination.'' Enclosure 4 provides additional
details that support this industry position.'' Additional detailed
comments by NEI and the comments on this subject by OCRE are discussed
under the Opportunities for Public Participation heading.
Response: The rule provides for public participation in the
regulatory process by incorporating a public meeting on the Licensee's
Thermal Annealing Report a minimum of 30 days before the start of
thermal annealing, and a public meeting after the licensee completes
the anneal but before the reactor is restarted. The opportunity for
public hearings in thermal annealing should be limited to those cases
where there is an unreviewed safety question or a change to the
Technical Specifications or where the licensee did not comply with the
Thermal Annealing Operating Plan and Requalification Inspection and
Test Program. Expanded discussion on this issue is provided below under
the Opportunities for Public Participation heading.
Issue 5: Whether there are health and safety issues concerning
thermal annealing that cannot be addressed generically and would
warrant plant-specific consideration.
Comment: NEI noted that ``Thermal annealing to reduce material
irradiation embrittlement is a well understood metallurgical
phenomenon. The supporting thermal and stress analysis used to
demonstrate that the vessel is not damaged during the anneal are
standard technologies used at nuclear plants. Because thermal annealing
uses well understood technology, public health and safety is reasonably
assured.''
Response: The NRC agrees with this comment.
Opportunities for Public Participation
The Supplementary Information section of the proposed rule
discussed the four options the Commission considered for structuring
the regulatory process related to public participation in the NRC's
review and approval of a licensee's proposal for thermal annealing of a
reactor vessel. The proposed rule, at the Commission's direction,
requested comments on the preferred regulatory process (including
opportunities for public participation). The four options included:
(1) No hearings under the rule as proposed;
(2) Discretionary opportunity for hearing under rule as proposed in
which situation the Commission would decide on a case-by-case basis to
determine whether a hearing should be held;
[[Page 65463]]
(3) Required opportunity for hearing under rule as proposed, but
work could commence if the NRC were to make a ``no significant hazard
determination'' on the proposed thermal annealing; and
(4) Modify the proposed rule to require suspension of license prior
and during the thermal annealing at which time no hearing would be
afforded and the license would only be reinstated if the licensee
demonstrates that it has addressed the reactor embrittlement such that
it is acceptable to operate the plant.
Three comments were submitted on the subject. OCRE and NEI
addressed all of the alternatives in detail and they, as well as one
utility, identified and discussed individual preferred alternatives.
NEI commented that each of the four alternatives has a sufficiently
serious flaw to prevent adoption. With respect to the no hearing
alternative, NEI agrees that annealing is presently subject to approval
by the Director of NRR in accordance with Part 50 Appendix G rather
than being the subject of a license amendment as an unreviewed safety
question under Sec. 50.59. However, NEI believes that annealing is an
important process from a regulatory standpoint and that public
participation, in the form of informal hearings, is appropriate. NEI
objected to a discretionary opportunity for a hearing because it
provides significant uncertainty in the process for licensees and
members of the public. NEI's objection to requiring a hearing, as
discussed in staff Option 3, is that it would allow those who object to
the resumption of operation, on other than technical grounds, to use
hearings to delay restart. Option 4 is objectionable to NEI because it
does not provide the licensee with any stability or predictability
since the licensee would be required to demonstrate compliance after
the annealing was performed, and does not provide the public with any
opportunity to express its views.
NEI further commented that a license amendment is not necessary to
approve a thermal annealing plan because annealing will not change the
reactor vessel or other components in a manner inconsistent with the
facility technical specifications nor will it require changes in the
FSAR, and further, that a licensee is not required to modify its
procedures to address or accommodate the annealing process. NEI noted
that, while there is an incentive for the licensee to obtain credit for
its improved P/T curves, and could seek a licensee amendment to do so,
the licensee's existing P/T curves could remain in force.
Despite the conclusion that a license amendment is not necessary
for thermal annealing, NEI recommended that a hearing opportunity be
provided, but that it be a non-adjudicatory, Subpart L type hearing on
the record. NEI gave the following advantages for this approach: (1)
The NRC would be provided with a clear understanding of the licensee's
annealing process, and the NRC's hearing process; (2) a Subpart L
hearing is held on the written record and typically does not include
the discovery or live testimony associated with adjudicatory hearings,
but allows the public to participate in a meaningful way without
consuming the vast NRC, licensee, and public resources required for an
adjudicatory hearing; and (3) it would provide predictability and
stability by ensuring that all issues which could be subject to a
hearing are addressed prior to restart. Any inspection or test
performed in order to restart would be for the purpose of confirming
compliance with the rule.
OCRE supported the proposed rule provided that the public hearing
rights were preserved with regard to reactor pressure vessel annealing.
It is OCRE's position on the request for public comment that, based on
the Sholly decision, the NRC must offer the opportunity for a formal
adjudicatory hearing on the application for annealing and on the
licensee's justification for subsequent operation where the licensee
cannot certify that the thermal annealing was performed in accordance
with the approved application. OCRE commented that approval by the
Director of NRR of the application for annealing and restart of the
reactor, if the licensee cannot certify that annealing was performed in
accordance with the approved application, will give the licensee the
authority to operate in ways in which they otherwise could not, and is
thus, a de facto license amendment. OCRE fully supported Option 3 which
requires opportunity for hearing under the rule as proposed. OCRE
suggested that the adequacy of the thermal annealing plan, as well as
the vessel's ability to perform its safety function after annealing,
could be raised in the hearing on the thermal annealing plan and that
the licensee's implementation of the thermal annealing plan could not
commence until any hearing is concluded or unless the NRC makes a ``no
significant hazards determination'' with respect to thermal annealing.
With respect to Option 1, OCRE concluded that the informal hearings
or public meetings proposed by the Commission for the initial thermal
annealing are not a substitute for adjudicatory hearings required by
the Atomic Energy Act (AEA) and do not give the interveners the same
rights as they would have in a Section 189a hearing. OCRE found Option
2 preferable to having no hearing. However, OCRE contended that this
option is flawed by the assumption that ``Section 189a of the AEA does
not afford an interested member of the public a right to request a
hearing.'' They contend that approval by the Director, NRR to anneal
the reactor pressure vessel or to restart after annealing does
constitute a de facto operating licensing amendment for which the
opportunity for a hearing is required. OCRE found Options 1 and 4
unacceptable in that they do not provide the opportunity for a formal
adjudicatory hearing.
The comment from the utility suggested that Option 1 is the
appropriate approach as long as the annealing process to be implemented
is approved in advance by the NRC staff and the utility certifies that
they have complied with the approved annealing process during the
annealing operation, as provided for in the proposed rule. The utility
further commented that if Technical Specifications changes or
amendments to the operating license are required in order to perform
the annealing then the opportunity for hearings would be required due
to the normal license amendment process and if the final safety
analysis report (FSAR) were required to be updated to reflect the
thermal annealing process, the provisions of 10 CFR 50.59 would apply.
The utility suggested that if those changes did not constitute an
``unreviewed safety question,'' no amendment would be needed and the
license amendment process should not be invoked and that if a member of
the public is concerned about a licensee's compliance with the NRC
approved thermal annealing plan, those concerns could be addressed
pursuant to the 10 CFR 2.206 petition process. The utility commented
that, under its proposal, existing regulatory provisions for public
participation would apply as appropriate and no new prescriptive
requirements would be necessary.
The Commission has considered the public comments and has modified
the proposed rule as follows. A licensee that seeks to utilize thermal
annealing to mitigate the effects of neutron irradiation of the nuclear
reactor vessel must, at least three years prior to the date at which
the limiting fracture toughness criteria in Sec. 50.61 or Appendix G to
Part 50 would be exceeded, submit a Thermal Annealing Report to the NRC
staff for review. The
[[Page 65464]]
report shall contain four sections: (i) Thermal Annealing Operating
Plan, (ii) Requalification Inspection and Test Program, (iii) Program
for determining Fracture Toughness Recovery and Reembrittlement Trend,
and (iv) a section identifying any changes to the description of the
facility as described in the updated final safety analysis report
(FSAR) which constitute unreviewed safety questions (USQs) under
Sec. 50.59, and changes to the facility's technical specifications,
which are necessary either to perform the thermal annealing, or to
operate following completion of the annealing. Section 50.66(a)
provides that the NRC will, within three years of submission of a
licensee's annealing report, document its views on whether the plan for
conducting thermal annealing constitutes an unreviewed safety question
or otherwise requires a change to the plant's technical specifications.
Such a determination is the threshold determination for whether NRC
approval is required before undertaking the activity. In the event the
NRC were to conclude, contrary to the licensee, that an unreviewed
safety question is present or a change to the technical specifications
is necessary, the NRC would, as a discretionary enforcement matter,
issue an appropriate order to the licensee prohibiting annealing prior
to issuance of a license amendment. An opportunity for formal
adjudicatory hearing would be provided in connection with the license
amendment; however, if the NRC makes a finding that the proposed change
to the FSAR description or technical specification constitutes a ``no
significant hazards consideration'' pursuant to Section 189.(a)(2)(A),
the licensee may conduct the thermal annealing prior to completion of
any hearing. In any event, at least 30 days before the licensee starts
to thermal anneal and before the NRC completes its review, the NRC will
hold a public meeting on the licensee's proposed Thermal Annealing Plan
and Requalification Inspection and Test Program.
Following the completion of the annealing operation, the licensee
must confirm in writing to the Director, Office of Nuclear Reactor
Regulation, that the thermal annealing was performed in accordance with
the Thermal Annealing Operating Plan and the Requalification and
Inspection Test Program. In support of this confirmation, the licensee
must submit a report, within three months of completion or termination
of the anneal, that presents the results of the annealing operation.
Within two weeks of the licensee's written confirmation that the
thermal annealing was completed in accordance with the Thermal
Annealing Plan, and prior to restart, the NRC shall: (1) Place in its
public document room a summary of the NRC staff's inspection of the
licensee's thermal annealing process to confirm that the thermal
annealing was completed in accordance with the Thermal Annealing
Operating Plan and the Requalification Inspection and Test Program, and
(2) hold a public meeting with the licensee to permit the licensee to
explain the results of the reactor vessel annealing to the NRC and the
public, for the NRC to discuss its inspection of the reactor vessel
annealing process, and to provide an opportunity for the public to
comment to the NRC on the annealing operation and the results of the
Staff's inspection.
Within 45 days of the licensee's written confirmation that the
thermal annealing was completed, the NRC shall complete full
documentation of the NRC's inspection of the licensee's annealing
process to confirm that the annealing was completed in accordance with
the Thermal Annealing Operating Plan and the Requalification Inspection
and Test Program.
The licensee may resume operation if: (1) The licensee concludes
that the thermal annealing operation was performed in compliance with
the Thermal Annealing Operating Plan, the Requalification Inspection
and Test Program, and the provisions of Section 50.66(b), (2) a summary
of the NRC's inspection of the thermal annealing is placed in the NRC
public document room as required by Section 50.66(c) (2) and (3) the
NRC holds the public meeting required by Section 50.66(f)(2), unless
the staff takes action against the licensee. Since NRC approval to
resume operation is not necessary, an opportunity for hearing would not
be provided in this situation. If, however, the licensee cannot
conclude that the thermal annealing was performed in compliance with
the Thermal Annealing Operating Plan or the Requalification Inspection
and Test Program, the licensee must submit a justification for
continued operation to the Director. If the noncompliance presents an
unreviewed safety question, as determined by the licensee or directed
by the NRC following its review of the report, then the plant may not
restart until the Director has approved restart. Those failures to
comply with the Thermal Annealing Operating Plan and the
Requalification Inspection and Test Program, which either (1) Are
considered to be ``unreviewed safety questions'' or (2) require changes
to the technical specifications as a result of the noncompliances,
would also be subject to an opportunity for a formal adjudicatory
hearing in accordance with the Commission's regulations governing
license amendments. However, the licensee may restart prior to
completion of the hearing if the Director makes a finding that such
restart constitutes a ``no significant hazards consideration,'' as
provided under Section 189.(a)(2)(A) of the Atomic Energy Act of 1954,
as amended.
The regulatory process for thermal annealing and the associated
hearing opportunities are consistent with long-standing NRC regulatory
practices defining those matters which present sufficient potential
effect on public health and safety (e.g., are unreviewed safety
questions) to justify both prior NRC review of the change, and an
opportunity for hearings (with the associated time and resource impacts
on both the licensee and the NRC). With respect to the thermal
annealing review process, the Commission reassessed the regulatory
requirements and processes for assuring safety. The Commission
determined that the most important safety matters are normally
addressed in license conditions, technical specifications, and the
FSAR. The regulatory process for NRC consideration of licensee-
initiated changes concerning these matters, and the associated
opportunities for hearings is in 10 CFR 50.59. In view of this well-
established regulatory process for important safety information, the
Commission determined that a regulatory process requiring NRC approval
of a thermal annealing plan is not necessary, because the licensee is
already required to comply with its license conditions, technical
specifications, and FSAR. Important changes to license conditions,
technical specifications, and FSAR from a safety standpoint are subject
to both prior NRC review and approval and an opportunity for hearing.
With respect to restart following completion of the annealing, the 15-
day delay period should be sufficient time for review of the licensee's
input given the NRC staff's understanding of the annealing operation
plan prior to implementation, ongoing resident inspections and
headquarters inspections of the implementation of thermal annealing
operating plan. The Commission did not adopt NEI's suggestion for
informal hearings where the Director must approve restart if the
Thermal Annealing Operating Plan and Requalification Inspection and
Test Program were not complied with, because the Commission does not
see
[[Page 65465]]
any distinction (in terms of safety implications) between the subject
matter of hearings under this rule, as compared with other actions
under Part 50 which would require formal hearings.
As discussed earlier in the supplementary information, previously
performed research analyses indicated the potential for plastic
deformation of the main coolant piping for a typical U.S. plant design
and anticipated annealing conditions. There are also questions
regarding how thermal growth of the pressure vessel is treated, and the
adequacy of the thermal and stress analyses used to predict response of
the overall system under thermal annealing conditions. Additionally,
there may be questions in other areas such as temperature limits for
the concrete structures, and potential radiological hazards associated
with removing and storing the reactor internals during the annealing
process, and fire hazards associated with heating the vessel.
Recognition of the numerous complex technical questions related to
4 thermal annealing and of the potential benefits for operating nuclear
power plants has resulted in a cooperative effort, funded by the U.S.
Department of Energy and the industry, to perform Annealing
Demonstration Projects. Projects are planned to demonstrate two
different annealing processes, evaluating heater designs and vessel
designs. It is anticipated that the annealing demonstration projects
will answer many of the generic questions regarding thermal annealing
of U.S. pressure vessel and piping designs.
The Thermal Annealing Report, required by the thermal annealing
rule, is designed to facilitate a detailed review by the licensee of
plant-specific questions and considerations in performing a thermal
annealing. The proposed rule specifically discusses the potential for
unreviewed safety questions and technical specification changes that
may result from or be related to thermal annealing of the reactor
pressure vessel. With completion of the demonstration projects and as
the staff and industry gain experience with thermal annealing, many of
the issues related to annealing will be better understood and related
questions will be answered. However, until this experience is realized,
the staff will critically review licensee determinations regarding
unreviewed safety questions and the need for technical specification
changes associated with each proposed thermal annealing. The level of
staff effort is expected to be significantly greater during its review
of the initial proposed vessel annealings than that which will be
required after experience is gained.
The thermal annealing rule has been structured to provide time for
the staff to thoroughly review the licensee's annealing plan and
determination regarding unreviewed safety questions and the need for
technical specification changes. If the staff identifies an unreviewed
safety question or the need for a technical specification change, the
licensee would be so notified and the existing NRC regulatory practices
would be invoked to address the issues.
Backfitting Issues
Comments were received on backfitting issues from the Nuclear
Utility Backfitting and Reform Group (NUBARG). NUBARG commented that
they do not object to the new NRC position in Appendix G to 10 CFR Part
50 which prohibits core criticality before completion of hydrostatic
pressure and leak tests as a conservative measure to enhance safety.
However, they are concerned that amending Appendix G on the basis of a
compliance exception may set a bad precedent for avoiding backfitting
analyses. NUBARG stated that ``The logic of the proposed rule would
seem to allow the NRC to avoid a backfitting analysis by (1) invoking
the intent of one requirement to override the explicit provisions of
another, (2) using the compliance exception when the practice being
eliminated seems specifically contemplated by and specified in the
pertinent regulation, and (3) overlooking the fact that the NRC has
apparently accepted this position in practice by some licensees * * *''
In NUBARG's view, this proposed amendment should be supported by a
backfit analysis. The Commission has reviewed this comment and has
concluded that use of the compliance exception under Sec. 50.109 for
the changes in Appendix G to 10 CFR Part 50 is appropriate. The Backfit
Analysis section contains further discussion on this subject. The issue
of explicitly prohibiting core criticality before completing pressure
and leak tests has been addressed previously (letter from J. M. Taylor,
EDO, to N. S. Reynolds and D. F. Stenger, NUBARG, dated February 2,
1990) and the NUBARG comment did not provide new information. The
Commission has concluded that any backfit requirements in this
amendment are necessary to bring the facilities into compliance with
licenses, or the rules and orders of the Commission, or into
conformance with written commitments by the licensees. Therefore, a
backfit analysis is not required pursuant to 10 CFR 50.109(a)(4)(i).
NUBARG also commented on the amendment to Appendix H to 10 CFR Part
50 regarding surveillance that would preclude reducing the amount of
testing if the initial test results agreed with predicted results.
Although NUBARG recognizes the change would be prospective, it believes
that NRC should provide flexibility to allow continued relief for any
licensee who lacks such an authorization but has relied on the
provision. The Commission believes that sufficient flexibility already
exists in that licensees who do not have an authorization may seek an
exemption under 10 CFR Part 50.12.
Another aspect of the backfitting concern raised by NUBARG
addresses the proposed amendment to Sec. 50.61 which, based on the
adequate protection exception, would impose a uniform methodology for
calculating the reference temperature. NUBARG contends that to rely on
the adequate protection exception is arguably erroneous because the
change in methodology is not likely an adequate protection issue (i.e.,
for most plants, the screening criteria will not be approached for many
years). As discussed further under Backfit Analysis, the Commission
believes that a new backfit analysis is not required for this
conforming change, which corrects an inadvertent omission from the
previous rulemaking. Therefore, the Commission concludes that the
adequate protection basis for the backfit continues to apply from the
previous rulemaking (56 FR 22300; May 15, 1991) to Sec. 50.61.
Criminal Penalties
For purposes of Section 223 of the Atomic Energy Act (AEA), the
Commission is issuing the final rule under one or more of Sections
161b, 161i or 161o of the AEA. Willful violations of the rule will be
subject to criminal enforcement.
Finding of No Significant Environmental Impact
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission's regulations in
Subpart A of 10 CFR Part 51, that this rule is not a major Federal
action significantly affecting the quality of human environment and,
therefore, an environmental impact statement is not required.
The individual actions covered in this final rule would either
serve to enhance safety of the reactor pressure vessel, thereby
decreasing the environmental impact of plant operation, or have no
[[Page 65466]]
impact on the environment. Therefore, in all cases these individual
actions will not have an adverse impact on the environment.
PTS Rule (10 CFR 50.61)
The inclusion of thermal annealing as an option for mitigating the
effects of neutron irradiation serves to decrease the environmental
impact of plant operation by enhancing the safety of the reactor
pressure vessel.
The incorporation of the Regulatory Guide 1.99, Revision 2, method
for determining RTNDT into the PTS rule has no impact on the
environment because this change will result in values of RTPTS
which are consistent with those currently used in plant operation.
The restructuring of the PTS rule is the type of action described
in categorical exclusion 10 CFR 51.22(c)(2). Therefore, an
environmental assessment is not necessary for this change.
Thermal Annealing Rule (10 CFR 50.66)
The thermal annealing rule (10 CFR 50.66) permits and provides
requirements for the thermal annealing of a reactor vessel to restore
fracture properties of the reactor vessel material which have been
degraded by neutron irradiation. This final rule only applies when a
licensee elects to use it. The final rule provides an alternative for
assuring compliance with the requirements in 10 CFR 50.61 and Appendix
G of 10 CFR Part 50.
The application of thermal annealing to a reactor vessel improves
the condition of the reactor vessel material. In addition, this rule
establishes requirements to avoid damaging the reactor system and to
protect against accidents during the annealing operation.
This rule is one of several regulatory requirements that will
function to ensure reactor vessel integrity. In that sense, this rule
has a positive impact on the environment by reducing the potential for
vessel failure. For these reasons, the Commission has determined that
there is no significant impact and, therefore, an environmental
statement is not required.
Appendix G to 10 CFR Part 50
The prohibition of core criticality before completion of the
required pressure and leak tests will serve to reduce the potential for
vessel failure, and thereby decrease the potential environmental impact
of plant operation.
The restructuring of Sections IV and V of Appendix G is clarifying
or corrective in nature, and is the type of action described in
categorical exclusion 10 CFR 51.22(c)(2). Therefore, an environmental
assessment is not necessary for this change.
The changing of the reference from Appendix G of Section III of the
ASME Code to Appendix G of Section XI of the ASME Code has no impact on
the environment because the requirements in the Appendices are
identical. Therefore, there is no adverse impact on the environment
from this change.
The referencing of the thermal annealing rule results in no adverse
impact on the environment because Appendix G currently permits the use
of thermal annealing to reduce fracture toughness loss of the RPV
materials due to irradiation embrittlement.
Appendix H to 10 CFR Part 50
Concerning the amendments to Appendix H to 10 CFR Part 50 in the
final rule, the requirement that all irradiation surveillance tests be
made (i.e., no reduction in testing is permitted) will have a positive
impact on the environment in helping to assure the integrity of the
reactor pressure vessel.
The restructuring of Section II.C is the type of action described
in categorical exclusion 10 CFR 51.22(c)(2). Therefore, an
environmental assessment is not necessary for this change.
The clarification of the applicable version of ASTM Standard E 185
will result in no adverse impact to the environment since there will be
no change to current surveillance programs. Changes to future
surveillance programs will make the programs more effective in
assessing irradiation embrittlement effects to the RPV materials,
thereby helping to assure the integrity of the reactor pressure vessel
Paperwork Reduction Act Statement
This final rule amends information collection requirements that are
subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et
seq.). These requirements were approved by the Office of Management and
Budget, approval number 3150-0011.
The public reporting burden for this collection of information is
estimated to average 6,000 hours per response, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
collection of information. Send comments regarding the burden estimate
or any other aspect of this collection of information, including
suggestions for reducing the burden, to the Information and Records
Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the Desk Officer, Office of
Information and Regulatory Affairs, NEOB-10202, (3150-0011), Office of
Management and Budget, Washington, DC 20503.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless it displays a currently
valid OMB control number.
Regulatory Analysis
The NRC staff has prepared a regulatory analysis for the amendments
to 10 CFR 50.61, Appendix G of 10 CFR Part 50, and Appendix H of 10 CFR
Part 50 that describes the factors and alternatives considered by the
Commission in deciding to issue these amendments. A copy of the
regulatory analysis is available for inspection and copying for a fee
at the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC 20555-0001. Single copies of the analysis may be
obtained from Alfred Taboada, Office of Nuclear Regulatory Research,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
telephone (301) 415-6014.
Regulatory Flexibility Act Certification
As required by the Regulatory Flexibility Act, 5 U.S.C. 605(b), the
Commission certifies that this final rule will not have a significant
economic impact on a substantial number of small entities. The rules
which are affected by the amendments will: (1) Preclude brittle
fracture of embrittled vessels during PTS events, (2) provide the
general fracture toughness requirements for RPVs, including ductile
fracture toughness requirements and pressure-temperature limits, (3)
provide the requirements for surveillance programs to monitor
irradiation embrittlement of RPV beltline materials, and (4) provide
for a method for restoring the fracture toughness of RPV beltline
materials used in nuclear facilities licensed under the provision of 10
CFR 50.21(b) and 10 CFR 50.22. The companies that own these facilities
do not fall within the scope of the definition of ``small entities'' as
set forth in the Regulatory Flexibility Act, the Small Business Size
Standards in regulations issued by the Small Business Administration at
13 CFR Part 121, or the size standards established by the NRC at 10 CFR
2.810 (60 FR 18344; April 11, 1995).
[[Page 65467]]
Backfit Analysis
PTS Rule (10 CFR 50.61)
The revision to Sec. 50.61 requires licensees to calculate
RTPTS using the same methodology specified in Regulatory Guide
1.99, Revision 2, for determining RTNDT. This change was logically
a requisite part of the previous rulemaking (56 FR 22300; May 15, 1991)
to Sec. 50.61 that set forth a unified method for calculating radiation
embrittlement of the reactor beltline materials in Part 50. However,
the Commission, at that time, inadvertently failed to make the
conforming change to Sec. 50.61. The Commission believes that the
backfit statement for the previous amendment, which determined that the
backfit was necessary to ensure that the facility continues to provide
adequate protection to the public health and safety, is applicable to
this conforming change to Sec. 50.61.
The restructuring of the PTS rule does not impose any backfits as
defined in 10 CFR 50.109(a)(1) because there is no change in
requirements due to this restructuring.
The inclusion of thermal annealing in Sec. 50.61 does not
constitute a backfit as defined in 10 CFR 50.109(a)(1) because the
decision to perform annealing is voluntary, no annealing has been
conducted in this country, and there are no staff positions or
Commission requirements relied upon by licensees that are being
changed.
Thermal Annealing Rule (10 CFR 50.66)
The final thermal annealing rule establishes requirements with
respect to applications for thermal annealing. However, the Commission
has determined that the rule does not impose a ``backfit'' as defined
in 10 CFR 50.109(a)(1). The thermal annealing rule does not require any
licensee to perform thermal annealing. Under existing requirements, all
licensees are required to evaluate whether they exceed the PTS
screening limits in 10 CFR 50.61 and the Charpy upper shelf screening
limits in Appendix G of CFR Part 50. However, these rules provide an
alternative means for meeting these screening limits (e.g., performing
thermal annealing). No licensee currently has pending before the NRC an
application for thermal annealing, nor has any current licensee been
granted permission to conduct thermal annealing. The rule does not
reflect any new or different NRC staff position which conflicts with a
prior NRC staff position or Commission rule. Thus, the final rule will
have a purely prospective effect on future applications for thermal
annealing. The Commission has stated in other rulemakings establishing
prospective requirements (10 CFR Part 52 and the License Renewal Rule,
10 CFR Part 54) that the Backfit Rule was not intended to protect the
future applicant from current changes in Commission requirements.
Accordingly, the Commission concludes that the rule does not impose
backfits and a backfit analysis need not be prepared for the final
thermal annealing rule.
Appendix G to 10 CFR Part 50
The restructuring of Sections IV and V of this appendix,
referencing of the thermal annealing rule, changing the reference from
Appendix G of Section III of the ASME Code to Appendix G of Section XI
of the ASME Code, and deleting the ``design to permit annealing''
requirement do not impose any backfits as defined in 10 CFR
50.109(a)(1), because they are either prospective in nature or are of a
clarifying nature.
10 CFR Part 50, Appendix G, Paragraph IV.2.d. of the final rule
explicitly prohibits core criticality before completion of ASME Code
hydrostatic pressure and leak tests. This is intended to make clear
that licensees may not use nuclear heat in order to perform ASME Code
hydrostatic tests. This amendment can be construed as a backfit,
inasmuch as the prior version of 10 CFR Part 50, Appendix G, Paragraph
IV.A.5 could be read to permit core criticality during ASME hydrostatic
tests and Section XI of the ASME Code does not explicitly prohibit core
criticality prior to completion of these tests. However, the Commission
never intended the disputed language in Paragraph IV.A.5 of Appendix G
to permit core criticality before successful completion of the required
ASME hydrostatic tests. The scope of Appendix G is ``fracture toughness
requirements'' only; that scope is stated clearly in the title of
Appendix G, and Appendix G was not intended to specify system
operational requirements. It is not correct, therefore, to interpret
paragraph IV.A.5. as permitting nuclear hydrotesting. The final phrase
in IV.A.5, ``depending on whether the core is critical during the
test,'' was included in the rule for the sake of completeness, to
specify appropriate fracture toughness requirements in the event that a
licensee for some reason wanted to have the core critical during
hydrotest, and was given approval to do so (e.g., as in the case of the
Hatch units, where nuclear hydrotesting was allowed one last time as an
approved exception.) The ASME Code's hydrostatic testing provisions for
the reactor coolant pressure boundary (RCPB) provides the necessary
assurance that GDC-14 is met. GDC-14 inter alia requires RCPB testing
in order to provide an extremely low probability of RCPB failure, in
terms of abnormal leakage, rapidly propagating failure, and gross
rupture. Using heat produced by a critical reactor core to perform such
testing essentially undercuts the basic safety principle embodied in
GDC-14 that testing should be completed prior to nuclear reactor
operation. It makes little sense to allow core criticality--thereby
allowing the reactor to be in an operational condition where a loss of
coolant could have significant consequences--prior to successful
completion of tests that are intended to ensure that the probability of
such coolant losses during such an operational condition are extremely
low.\1\ The ASME Code, Section XI, requires that the System Leakage
Test be performed prior to plant startup following each refueling
outage (Table-2500-1, Examination Category B-P, Note 2). The only way
to interpret the ASME Code as permitting core criticality prior to
completion of the hydrostatic tests is to read the term, ``plant
startup'' as referring to something other than reactor criticality.
This is neither the normal industry practice, nor has it been the NRC
staff's longstanding interpretation of this provision of the ASME code.
Indeed, it does not appear that the NRC staff has construed either
Appendix G, Paragraph IV.A.5 nor Section XI of the ASME Code as
permitting core criticality prior to successful completion of ASME Code
hydrostatic tests. Moreover, the vast majority of nuclear utility
licensees do not use nuclear heat to perform ASME code hydrostatic
tests. This suggests that most licensees hold the same interpretation
of Appendix G and Section XI of the ASME Code as the Commission. In
sum, the Commission believes Section XI of the ASME Code, which is
endorsed by 10 CFR 50.55a, implicitly prohibits core criticality prior
to successful completion of hydrostatic testing. Therefore, the
Commission concludes that the change in the language of Appendix G,
Paragraph IV.2.d. is necessary to assure compliance with 10 CFR 50.55a
and the ASME Code.
\1\ The Commission is aware that NUBARG has presented an
argument to the NRC that performance of ASME Code hydrostatic tests
are more effective at the higher temperatures achieved when using
nuclear heat, as compared with the heat sources normally employed by
utilities in performing the hydrostatic tests. However, for the
reasons set forth in the 1990 letter from James M. Taylor, EDO to N.
S. Reynolds and D.F. Stenger, NUBARG, the Commission rejects this
argument.
[[Page 65468]]
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The Commission has concluded that any backfit requirements in this
amendment are necessary to bring the facilities into compliance with
licenses, or the rules and orders of the Commission, or into
conformance with written commitments by the licensees. Therefore, a
backfit analysis is not required pursuant to 10 CFR 50.109(a)(4)(i).
Appendix H to 10 CFR Part 50
The amendments to Appendix H to 10 CFR Part 50 are either
prospective in nature or of a clarifying nature, and hence do not
involve any provisions which would impose backfits as defined in 10 CFR
50.109(a)(1).
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
record keeping requirements.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553; the NRC is adopting
the following amendments to 10 CFR Part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
1. The general authority citation for Part 50 is corrected to read
as set forth below, and the section-specific authority citations
continue to read as follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,
83 Stat. 1444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended 1244, 1246, (42 U.S.C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 101,
185, 68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub.
L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, and
50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as
amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56
also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections
50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L.
91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also
issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections
50.58, 50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat.
2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68
Stat. 939 (42 U.S.C. 2152). Sections 50.80-50.81 also issued under
sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also
issued under sec. 187, 68 Stat 955 (42 U.S.C. 2237).
2. In Sec. 50.8, paragraph (b) is revised to read as follows:
Sec. 50.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Secs. 50.30, 50.33, 50.33a, 50.34, 50.34a, 50.35,
50.36, 50.36a, 50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 50.60, 50.61,
50.63, 50.64, 50.65, 50.66, 70.71, 50.72, 50.73, 50.75, 50.80, 50.82,
50.90, 50.91, 50.120, and Appendices A, B, E, G, H, I, J, K, M, N, O,
Q, and R, to this part.
* * * * *
3. Section 50.61 is revised to read as follows:
Sec. 50.61 Fracture toughness requirements for protection against
pressurized thermal shock events.
(a) Definitions. For the purposes of this section:
(1) ASME Code means the American Society of Mechanical Engineers
Boiler and Pressure Vessel Code, Section III, Division I, ``Rules for
the Construction of Nuclear Power Plant Components,'' edition and
addenda and any limitations and modifications thereof as specified in
Sec. 50.55a.
(2) Pressurized Thermal Shock Event means an event or transient in
pressurized water reactors (PWRs) causing severe overcooling (thermal
shock) concurrent with or followed by significant pressure in the
reactor vessel.
(3) Reactor Vessel Beltline means the region of the reactor vessel
(shell material including welds, heat affected zones and plates or
forgings) that directly surrounds the effective height of the active
core and adjacent regions of the reactor vessel that are predicted to
experience sufficient neutron radiation damage to be considered in the
selection of the most limiting material with regard to radiation
damage.
(4) RTNDT means the reference temperature for a reactor vessel
material, under any conditions. For the reactor vessel beltline
materials, RTNDT must account for the effects of neutron
radiation.
(5) RTNDT(U) means the reference temperature for a reactor
vessel material in the pre-service or unirradiated condition, evaluated
according to the procedures in the ASME Code, Paragraph NB-2331 or
other methods approved by the Director, Office of Nuclear Reactor
Regulation.
(6) EOL Fluence means the best-estimate neutron fluence projected
for a specific vessel beltline material at the clad-base-metal
interface on the inside surface of the vessel at the location where the
material receives the highest fluence on the expiration date of the
operating license.
(7) RTPTS means the reference temperature, RTNDT,
evaluated for the EOL Fluence for each of the vessel beltline
materials, using the procedures of paragraph (c) of this section.
(8) PTS Screening Criterion means the value of RTPTS for the
vessel beltline material above which the plant cannot continue to
operate without justification.
(b) Requirements.
(1) For each pressurized water nuclear power reactor for which an
operating license has been issued, the licensee shall have projected
values of RTPTS, accepted by the NRC, for each reactor vessel
beltline material for the EOL fluence of the material. The assessment
of RTPTS must use the calculation procedures given in paragraph
(c)(1) of this section, except as provided in paragraphs (c)(2) and
(c)(3) of this section. The assessment must specify the bases for the
projected value of RTPTS for each vessel beltline material,
including the assumptions regarding core loading patterns, and must
specify the copper and nickel contents and the fluence value used in
the calculation for each beltline material. This assessment must be
updated whenever there is a significant 2 change in projected
values of RTPTS, or upon a request for a change in the expiration
date for operation of the facility.
\2\ Changes to RTPTS values are considered significant if
either the previous value or the current value, or both values,
exceed the screening criterion prior to the expiration of the
operating license, including any renewed term, if applicable, for
the plant.
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(2) The pressurized thermal shock (PTS) screening criterion is 270
deg.F for plates, forgings, and axial weld materials, and 300 deg.F
for circumferential weld materials. For the purpose of comparison with
this criterion, the value of RTPTS for the reactor vessel must be
evaluated according to the procedures of paragraph (c) of this section,
for each weld and plate, or forging, in the reactor vessel beltline.
RTPTS must be determined for each vessel beltline material using
the EOL fluence for that material.
(3) For each pressurized water nuclear power reactor for which the
value of RTPTS for any material in the beltline is projected to
exceed the PTS screening criterion using the EOL fluence, the licensee
shall implement those flux
[[Page 65469]]
reduction programs that are reasonably practicable to avoid exceeding
the PTS screening criterion set forth in paragraph (b)(2) of this
section. The schedule for implementation of flux reduction measures may
take into account the schedule for submittal and anticipated approval
by the Director, Office of Nuclear Reactor Regulation, of detailed
plant-specific analyses, submitted to demonstrate acceptable risk with
RTPTS above the screening limit due to plant modifications, new
information or new analysis techniques.
(4) For each pressurized water nuclear power reactor for which the
analysis required by paragraph (b)(3) of this section indicates that no
reasonably practicable flux reduction program will prevent RTPTS
from exceeding the PTS screening criterion using the EOL fluence, the
licensee shall submit a safety analysis to determine what, if any,
modifications to equipment, systems, and operation are necessary to
prevent potential failure of the reactor vessel as a result of
postulated PTS events if continued operation beyond the screening
criterion is allowed. In the analysis, the licensee may determine the
properties of the reactor vessel materials based on available
information, research results, and plant surveillance data, and may use
probabilistic fracture mechanics techniques. This analysis must be
submitted at least three years before RTPTS is projected to exceed
the PTS screening criterion.
(5) After consideration of the licensee's analyses, including
effects of proposed corrective actions, if any, submitted in accordance
with paragraphs (b)(3) and (b)(4) of this section, the Director, Office
of Nuclear Reactor Regulation, may, on a case-by-case basis, approve
operation of the facility with RTPTS in excess of the PTS
screening criterion. The Director, Office of Nuclear Reactor
Regulation, will consider factors significantly affecting the potential
for failure of the reactor vessel in reaching a decision.
(6) If the Director, Office of Nuclear Reactor Regulation,
concludes, pursuant to paragraph (b)(5) of this section, that operation
of the facility with RTPTS in excess of the PTS screening
criterion cannot be approved on the basis of the licensee's analyses
submitted in accordance with paragraphs (b)(3) and (b)(4) of this
section, the licensee shall request and receive approval by the
Director, Office of Nuclear Reactor Regulation, prior to any operation
beyond the criterion. The request must be based upon modifications to
equipment, systems, and operation of the facility in addition to those
previously proposed in the submitted analyses that would reduce the
potential for failure of the reactor vessel due to PTS events, or upon
further analyses based upon new information or improved methodology.
(7) If the limiting RTPTS value of the plant is projected to
exceed the screening criteria in paragraph (b)(2), or the criteria in
paragraphs (b)(3) through (b)(6) of this section cannot be satisfied,
the reactor vessel beltline may be given a thermal annealing treatment
to recover the fracture toughness of the material, subject to the
requirements of Sec. 50.66. The reactor vessel may continue to be
operated only for that service period within which the predicted
fracture toughness of the vessel beltline materials satisfy the
requirements of paragraphs (b)(2) through (b)(6) of this section, with
RTPTS accounting for the effects of annealing and subsequent
irradiation.
(c) Calculation of RTPTS. RTPTS must be calculated for
each vessel beltline material using a fluence value, f, which is the
EOL fluence for the material. RTPTS must be evaluated using the
same procedures used to calculate RTNDT, as indicated in paragraph
(c)(1) of this section, and as provided in paragraphs (c)(2) and (c)(3)
of this section.
(1) Equation 1 must be used to calculate values of RTNDT for
each weld and plate, or forging, in the reactor vessel beltline.
Equation 1: RTNDT=RTNDT(U)+M+RTNDT
(i) If a measured value of RTNDT(U) is not available, a
generic mean value for the class 3 of material may be used if
there are sufficient test results to establish a mean and a standard
deviation for the class.
\3\ The class of material for estimating RTNDT(U) is
generally determined for welds by the type of welding flux (Linde
80, or other), and for base metal by the material specification.
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(ii) For generic values of weld metal, the following generic mean
values must be used unless justification for different values is
provided: 0 deg.F for welds made with Linde 80 flux, and -56 deg.F for
welds made with Linde 0091, 1092 and 124 and ARCOS B-5 weld fluxes.
(iii) M means the margin to be added to account for uncertainties
in the values of RTNDT(U), copper and nickel contents, fluence and
the calculational procedures. M is evaluated from Equation 2.
[GRAPHIC][TIFF OMITTED]TR19DE95.003
(A) In Equation 2, U is the standard deviation for
RTNDT(U). If a measured value of RTNDT(U) is used, then
U is determined from the precision of the test method. If
a measured value of RTNDT(U) is not available and a generic mean
value for that class of materials is used, then U is the
standard deviation obtained from the set of data used to establish the
mean. If a generic mean value given in paragraph (c)(1)(i)(B) of this
section for welds is used, then U is 17 deg.F.
(B) In Equation 2, is the standard deviation for
RTNDT. The value of to be used is
28 deg.F for welds and 17 deg.F for base metal; the value of
need not exceed one-half of RTNDT.
(iv) RTNDT is the mean value of the transition
temperature shift, or change in RTNDT, due to irradiation, and
must be calculated using Equation 3.
Equation 3: RTNDT=(CF)f(0.28-0.10 log f)
(A) CF ( deg.F) is the chemistry factor, which is a function of
copper and nickel content. CF is given in Table 1 for welds and in
Table 2 for base metal (plates and forgings). Linear interpolation is
permitted. In Tables 1 and 2, ``Wt-% copper'' and ``Wt-% nickel'' are
the best-estimate values for the material, which will normally be the
mean of the measured values for a plate or forging. For a weld, the
best estimate values will normally be the mean of the measured values
for a weld deposit made using the same weld wire heat number as the
critical vessel weld. If these values are not available, the upper
limiting values given in the material specifications to which the
vessel material was fabricated may be used. If not available,
conservative estimates (mean plus one standard deviation) based on
generic data 4 may be used if justification is provided. If none
of these alternatives are available, 0.35% copper and 1.0% nickel must
be assumed.
\4\ Data from reactor vessels fabricated to the same material
specification in the same shop as the vessel in question and in the
same time period is an example of ``generic data.''
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(B) f is the best estimate neutron fluence, in units of 1019
n/cm2 (E greater than 1 MeV), at the clad-base-metal interface on
the inside surface of the vessel at the location where the material in
question receives the highest fluence for the period of service in
question. As specified in this paragraph, the EOL fluence for the
vessel beltline material is used in calculating KRTPTS.
(v) Equation 4 must be used for determining RTPTS using
equation 3 with EOL fluence values for determining RTPTS.
Equation 4: RTPTS=RTNDT(U)+M+RTPTS
(2) To verify that RTNDT for each vessel beltline material is
a bounding value for the specific reactor vessel, licensees shall
consider plant-specific information that could affect the level of
[[Page 65470]]
embrittlement. This information includes but is not limited to the
reactor vessel operating temperature and any related surveillance
program 5 results.
\5\ Surveillance program results means any data that
demonstrates the embrittlement trends for the limiting beltline
material, including but not limited to data from test reactors or
from surveillance programs at other plants with or without
surveillance program integrated per 10 CFR Part 50, Appendix H.
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(i) Results from the plant-specific surveillance program must be
integrated into the RTNDT estimate if the plant-specific
surveillance data has been deemed credible as judged by the following
criteria:
(A) The materials in the surveillance capsules must be those which
are the controlling materials with regard to radiation embrittlement.
(B) Scatter in the plots of Charpy energy versus temperature for
the irradiated and unirradiated conditions must be small enough to
permit the determination of the 30-foot-pound temperature
unambiguously.
(C) Where there are two or more sets of surveillance data from one
reactor, the scatter of RTNDT values must be less than
28 deg.F for welds and 17 deg.F for base metal. Even if the range in
the capsule fluences is large (two or more orders of magnitude), the
scatter may not exceed twice those values.
(D) The irradiation temperature of the Charpy specimens in the
capsule must equal the vessel wall temperature at the cladding/base
metal interface within 25 deg.F.
(E) The surveillance data for the correlation monitor material in
the capsule, if present, must fall within the scatter band of the data
base for the material.
(ii)(A) Surveillance data deemed credible according to the criteria
of paragraph (c)(2)(i) of this section must be used to determine a
material-specific value of CF for use in Equation 3. A material-
specific value of CF is determined from Equation 5.
[GRAPHIC][TIFF OMITTED]TR19DE95.004
(B) In Equation 5, ``n'' is the number of surveillance data points,
``Ai'' is the measured value of RTNDT and
``fi'' is the fluence for each surveillance data point. If there
is clear evidence that the copper and nickel content of the
surveillance weld differs from the vessel weld, i.e. differs from the
average for the weld wire heat number associated with the vessel weld
and the surveillance weld, the measured values of RTNDT
must be adjusted for differences in copper and nickel content by
multiplying them by the ratio of the chemistry factor for the vessel
material to that for the surveillance weld.
(iii) For cases in which the results from a credible plant-specific
surveillance program are used, the value of to be
used is 14 deg.F for welds and 8.5 deg.F for base metal; the value of
need not exceed one-half of DRTNDT.
(iv) The use of results from the plant-specific surveillance
program may result in an RTNDT that is higher or lower than those
determined in paragraph (c)(1).
(3) Any information that is believed to improve the accuracy of the
RTPTS value significantly must be reported to the Director, Office
of Nuclear Reactor Regulation. Any value of RTPTS that has been
modified using the procedures of paragraph (c)(2) of this section is
subject to the approval of the Director, Office of Nuclear Reactor
Regulation, when used as provided in this section.
Table 1.--Chemistry Factor for Weld Metals, deg.F
----------------------------------------------------------------------------------------------------------------
Nickel, wt-%
Copper, wt-% ---------------------------------------------------------------------
0 0.20 0.40 0.60 0.80 1.00 1.20
----------------------------------------------------------------------------------------------------------------
0......................................... 20 20 20 20 20 20 20
0.01...................................... 20 20 20 20 20 20 20
0.02...................................... 21 26 27 27 27 27 27
0.03...................................... 22 35 41 41 41 41 41
0.04...................................... 24 43 54 54 54 54 54
0.05...................................... 26 49 67 68 68 68 68
0.06...................................... 29 52 77 82 82 82 82
0.07...................................... 32 55 85 95 95 95 95
0.08...................................... 36 58 90 106 108 108 108
0.09...................................... 40 61 94 115 122 122 122
0.10...................................... 44 65 97 122 133 135 135
0.11...................................... 49 68 101 130 144 148 148
0.12...................................... 52 72 103 135 153 161 161
0.13...................................... 58 76 106 139 162 172 176
0.14...................................... 61 79 109 142 168 182 188
0.15...................................... 66 84 112 146 175 191 200
0.16...................................... 70 88 115 149 178 199 211
0.17...................................... 75 92 119 151 184 207 221
0.18...................................... 79 95 122 154 187 214 230
0.19...................................... 83 100 126 157 191 220 238
0.20...................................... 88 104 129 160 194 223 245
0.21...................................... 92 108 133 164 197 229 252
[[Page 65471]]
0.22...................................... 97 112 137 167 200 232 257
0.23...................................... 101 117 140 169 203 236 263
0.24...................................... 105 121 144 173 206 239 268
0.25...................................... 110 126 148 176 209 243 272
0.26...................................... 113 130 151 180 212 246 276
0.27...................................... 119 134 155 184 216 249 280
0.28...................................... 122 138 160 187 218 251 284
0.29...................................... 128 142 164 191 222 254 287
0.30...................................... 131 146 167 194 225 257 290
0.31...................................... 136 151 172 198 228 260 293
0.32...................................... 140 155 175 202 231 263 296
0.33...................................... 144 160 180 205 234 266 299
0.34...................................... 149 164 184 209 238 269 302
0.35...................................... 153 168 187 212 241 272 305
0.36...................................... 158 172 191 216 245 275 308
0.37...................................... 162 177 196 220 248 278 311
0.38...................................... 166 182 200 223 250 281 314
0.39...................................... 171 185 203 227 254 285 317
0.40...................................... 175 189 207 231 257 288 320
----------------------------------------------------------------------------------------------------------------
Table 2.--Chemistry Factor for Base Metals, deg.F
----------------------------------------------------------------------------------------------------------------
Nickel, wt-%
Copper, wt-% ---------------------------------------------------------------------
0 0.20 0.40 0.60 0.80 1.00 1.20
----------------------------------------------------------------------------------------------------------------
0......................................... 20 20 20 20 20 20 20
0.01...................................... 20 20 20 20 20 20 20
0.02...................................... 20 20 20 20 20 20 20
0.03...................................... 20 20 20 20 20 20 20
0.04...................................... 22 26 26 26 26 26 26
0.05...................................... 25 31 31 31 31 31 31
0.06...................................... 28 37 37 37 37 37 37
0.07...................................... 31 43 44 44 44 44 44
0.08...................................... 34 48 51 51 51 51 51
0.09...................................... 37 53 58 58 58 58 58
0.10...................................... 41 58 65 65 67 67 67
0.11...................................... 45 62 72 74 77 77 77
0.12...................................... 49 67 79 83 86 86 86
0.13...................................... 53 71 85 91 96 96 96
0.14...................................... 57 75 91 100 105 106 106
0.15...................................... 61 80 99 110 115 117 117
0.16...................................... 65 84 104 118 123 125 125
0.17...................................... 69 88 110 127 132 135 135
0.18...................................... 73 92 115 134 141 144 144
0.19...................................... 78 97 120 142 150 154 154
0.20...................................... 82 102 125 149 159 164 165
0.21...................................... 86 107 129 155 167 172 174
0.22...................................... 91 112 134 161 176 181 184
0.23...................................... 95 117 138 167 184 190 194
0.24...................................... 100 121 143 172 191 199 204
0.25...................................... 104 126 148 176 199 208 214
0.26...................................... 109 130 151 180 205 216 221
0.27...................................... 114 134 155 184 211 225 230
0.28...................................... 119 138 160 187 216 233 239
0.29...................................... 124 142 164 191 221 241 248
0.30...................................... 129 146 167 194 225 249 257
0.31...................................... 134 151 172 198 228 255 266
0.32...................................... 139 155 175 202 231 260 274
0.33...................................... 144 160 180 205 234 264 282
0.34...................................... 149 164 184 209 238 268 290
0.35...................................... 153 168 187 212 241 272 298
0.36...................................... 158 173 191 216 245 275 303
0.37...................................... 162 177 196 220 248 278 308
0.38...................................... 166 182 200 223 250 281 313
0.39...................................... 171 185 203 227 254 285 317
0.40...................................... 175 189 207 231 257 288 320
----------------------------------------------------------------------------------------------------------------
[[Page 65472]]
4. A new Sec. 50.66 is added under the center heading ``Issuance,
Limitations, and Conditions of Licenses and Construction Permits'' to
read as follows:
Sec. 50.66 Requirements for thermal annealing of the reactor pressure
vessel.
(a) For those light water nuclear power reactors where neutron
radiation has reduced the fracture toughness of the reactor vessel
materials, a thermal annealing may be applied to the reactor vessel to
recover the fracture toughness of the material. The use of a thermal
annealing treatment is subject to the requirements in this section. A
report describing the licensee's plan for conducting the thermal
annealing must be submitted in accordance with Sec. 50.4 at least three
years prior to the date at which the limiting fracture toughness
criteria in Sec. 50.61 or Appendix G to Part 50 would be exceeded.
Within three years of the submittal of the Thermal Annealing Report and
at least thirty days prior to the start of the thermal annealing, the
NRC will review the Thermal Annealing Report and place the results of
its evaluation in its Public Document Room. The licensee may begin the
thermal anneal after:
(1) Submitting the Thermal Annealing Report required by paragraph
(b) of this section;
(2) the NRC places the results of its evaluation of the Thermal
Annealing Report in the Public Document Room; and
(3) the requirements of paragraph (f)(1) of this section have been
satisfied.
(b) Thermal Annealing Report. The Thermal Annealing Report must
include: a Thermal Annealing Operating Plan; a Requalification
Inspection and Test Program; a Fracture Toughness Recovery and
Reembrittlement Trend Assurance Program; and Identification of
Unreviewed Safety Questions and Technical Specification Changes.
(1) Thermal Annealing Operating Plan.
The thermal annealing operating plan must include:
(i) A detailed description of the pressure vessel and all
structures and components that are expected to experience significant
thermal or stress effects during the thermal annealing operation;
(ii) An evaluation of the effects of mechanical and thermal
stresses and temperatures on the vessel, containment, biological
shield, attached piping and appurtenances, and adjacent equipment and
components to demonstrate that operability of the reactor will not be
detrimentally affected. This evaluation must include:
(A) Detailed thermal and structural analyses to establish the time
and temperature profile of the annealing operation. These analyses must
include heatup and cooldown rates, and must demonstrate that localized
temperatures, thermal stress gradients, and subsequent residual
stresses will not result in unacceptable dimensional changes or
distortions in the vessel, attached piping and appurtenances, and that
the thermal annealing cycle will not result in unacceptable degradation
of the fatigue life of these components.
(B) The effects of localized high temperatures on degradation of
the concrete adjacent to the vessel and changes in thermal and
mechanical properties, if any, of the reactor vessel insulation, and on
detrimental effects, if any, on containment and the biological shield.
If the design temperature limitations for the adjacent concrete
structure are to be exceeded during the thermal annealing operation, an
acceptable maximum temperature for the concrete must be established for
the annealing operation using appropriate test data.
(iii) The methods, including heat source, instrumentation and
procedures proposed for performing the thermal annealing. This shall
include any special precautions necessary to minimize occupational
exposure, in accordance with the As Low As Reasonably Achievable
(ALARA) principle and the provisions of Sec. 20.1206.
(iv) The proposed thermal annealing operating parameters, including
bounding conditions for temperatures and times, and heatup and cooldown
schedules.
(A) The thermal annealing time and temperature parameters selected
must be based on projecting sufficient recovery of fracture toughness,
using the procedures of paragraph (e) of this section, to satisfy the
requirements of Sec. 50.60 and Sec. 50.61 for the proposed period of
operation addressed in the application.
(B) The time and temperature parameters evaluated as part of the
thermal annealing operating plan, and supported by the evaluation
results of paragraph (b)(1)(ii) of this section, represent the bounding
times and temperatures for the thermal annealing operation. If these
bounding conditions for times and temperatures are violated during the
thermal annealing operation, then the annealing operation is considered
not in accordance with the Thermal Annealing Operating Plan, as
required by paragraph (c)(1) of this section, and the licensee must
comply with paragraph (c)(2) of this section.
(2) Requalification Inspection and Test Program. The inspection and
test program to requalify the annealed reactor vessel must include the
detailed monitoring, inspections, and tests proposed to demonstrate
that the limitations on temperatures, times and temperature profiles,
and stresses evaluated for the proposed thermal annealing conditions of
paragraph (b)(1)(iv) of this section have not been exceeded, and to
determine the thermal annealing time and temperature to be used in
quantifying the fracture toughness recovery. The requalification
inspection and test program must demonstrate that the thermal annealing
operation has not degraded the reactor vessel, attached piping or
appurtenances, or the adjacent concrete structures to a degree that
could affect the safe operation of the reactor.
(3) Fracture Toughness Recovery and Reembrittlement Trend Assurance
Program. The percent recovery of RTNDT and Charpy upper-shelf
energy due to the thermal annealing treatment must be determined based
on the time and temperature of the actual vessel thermal anneal. The
recovery of RTNDT and Charpy upper-shelf energy provide the basis
for establishing the post-anneal RTNDT and Charpy upper-shelf
energy for each vessel material. Changes in the RTNDT and Charpy
upper-shelf energy with subsequent plant operation must be determined
using the post-anneal values of these parameters in conjunction with
the projected reembrittlement trend determined in accordance with
paragraph (b)(3)(ii) of this section. Recovery and reembrittlement
evaluations shall include:
(i) Recovery Evaluations.
(A) The percent recovery of both RTNDT and Charpy upper-shelf
energy must be determined by one of the procedures described in
paragraph (e) of this section, using the proposed lower bound thermal
annealing time and temperature conditions described in the operating
plan.
(B) If the percent recovery is determined from testing surveillance
specimens or from testing materials removed from the reactor vessel,
then it shall be demonstrated that the proposed thermal annealing
parameters used in the test program are equal to or bounded by those
used in the vessel annealing operation.
(C) If generic computational methods are used, appropriate
justification must be submitted as a part of the application.
(ii) Reembrittlement Evaluations.
(A) The projected post-anneal reembrittlement of RTNDT must be
[[Page 65473]]
calculated using the procedures in Sec. 50.61(c), or must be determined
using the same basis as that used for the pre-anneal operating period.
The projected change due to post-anneal reembrittlement for Charpy
upper-shelf energy must be determined using the same basis as that used
for the pre-anneal operating period.
(B) The post-anneal reembrittlement trend of both RTNDT and
Charpy upper-shelf energy must be estimated, and must be monitored
using a surveillance program defined in the Thermal Annealing Report
and which conforms to the intent of Appendix H of this part, ``Reactor
Vessel Material Surveillance Program Requirements.''
(4) Identification of Unreviewed Safety Questions and Technical
Specification Changes. Any changes to the facility as described in the
updated final safety analysis report constituting unreviewed safety
questions, and any changes to the technical specifications, which are
necessary to either conduct the thermal annealing or operate the
nuclear power reactor following the annealing, must be identified. The
section shall demonstrate that the Commission's requirements continue
to be complied with, and that there is reasonable assurance of adequate
protection to the public health and safety following the changes.
(c) Completion or Termination of Thermal Annealing.
(1) If the thermal annealing was completed in accordance with the
Thermal Annealing Operating Plan and the Requalification Inspection and
Test Program, the licensee shall so confirm in writing to the Director,
Office of Nuclear Reactor Regulation. The licensee may restart its
reactor after the requirements of paragraph (f)(2) of this section have
been met.
(2) If the thermal annealing was completed but the annealing was
not performed in accordance with the Thermal Annealing Operating Plan
and the Requalification Inspection and Test Program, the licensee shall
submit a summary of lack of compliance with the Thermal Annealing
Operating Plan and the Requalification Inspection and Test Program and
a justification for subsequent operation to the Director, Office of
Nuclear Reactor Regulation. Any changes to the facility as described in
the updated final safety analysis report which are attributable to the
noncompliances and constitute unreviewed safety questions, and any
changes to the technical specifications which are required as a result
of the noncompliances, shall also be identified.
(i) If no unreviewed safety questions or changes to technical
specifications are identified, the licensee may restart its reactor
after the requirements of paragraph (f)(2) of this section have been
met.
(ii) If any unreviewed safety questions or changes to technical
specifications are identified, the licensee may not restart its reactor
until approval is obtained from the Director, Office of Nuclear Reactor
Regulation and the requirements of paragraph (f)(2) of this section
have been met.
(3) If the thermal annealing was terminated prior to completion,
the licensee shall immediately notify the NRC of the premature
termination of the thermal anneal.
(i) If the partial annealing was otherwise performed in accordance
with the Thermal Annealing Operating Plan and relevant portions of the
Requalification Inspection and Test Program, and the licensee does not
elect to take credit for any recovery, the licensee need not submit the
Thermal Annealing Results Report required by paragraph (d) of this
section but instead shall confirm in writing to the Director, Office of
Nuclear Reactor Regulation that the partial annealing was otherwise
performed in accordance with the Thermal Annealing Operating Plan and
relevant portions of the Requalification Inspection and Test Program.
The licensee may restart its reactor after the requirements of
paragraph (f)(2) of this section have been met.
(ii) If the partial annealing was otherwise performed in accordance
with the Thermal Annealing Operating Plan and relevant portions of the
Requalification Inspection and Test Program, and the licensee elects to
take full or partial credit for the partial annealing, the licensee
shall confirm in writing to the Director, Office of Nuclear Reactor
Regulation that the partial annealing was otherwise performed in
compliance with the Thermal Annealing Operating Plan and relevant
portions of the Requalification Inspection and Test Program. The
licensee may restart its reactor after the requirements of paragraph
(f)(2) of this section have been met.
(iii) If the partial annealing was not performed in accordance with
the Thermal Annealing Operating Plan and relevant portions of the
Requalification Inspection and Test Program, the licensee shall submit
a summary of lack of compliance with the Thermal Annealing Operating
Plan and the Requalification Inspection and Test Program and a
justification for subsequent operation to the Director, Office of
Nuclear Reactor Regulation. Any changes to the facility as described in
the updated final safety analysis report which are attributable to the
noncompliances and constitute unreviewed safety questions, and any
changes to the technical specifications which are required as a result
of the noncompliances, shall also be identified.
(A) If no unreviewed safety questions or changes to technical
specifications are identified, the licensee may restart its reactor
after the requirements of paragraph (f)(2) of this section have been
met.
(B) If any unreviewed safety questions or changes to technical
specifications are identified, the licensee may not restart its reactor
until approval is obtained from the Director, Office of Nuclear Reactor
Regulation and the requirements of paragraph (f)(2) of this section
have been met.
(d) Thermal Annealing Results Report. Every licensee that either
completes a thermal annealing, or that terminates an annealing but
elects to take full or partial credit for the annealing, shall provide
the following information within three months of completing the thermal
anneal, unless an extension is authorized by the Director, Office of
Nuclear Reactor Regulation:
(1) The time and temperature profiles of the actual thermal
annealing;
(2) The post-anneal RTNDT and Charpy upper-shelf energy values
of the reactor vessel materials for use in subsequent reactor
operation;
(3) The projected post-anneal reembrittlement trends for both
RTNDT and Charpy upper-shelf energy; and
(4) The projected values of RTPTS and Charpy upper-shelf
energy at the end of the proposed period of operation addressed in the
Thermal Annealing Report.
(e) Procedures for Determining the Recovery of Fracture Toughness.
The procedures of this paragraph must be used to determine the percent
recovery of RTNDT, Rt, and percent recovery of
Charpy upper-shelf energy, Ru. In all cases, Rt and Ru
may not exceed 100.
(1) For those reactors with surveillance programs which have
developed credible surveillance data as defined in Sec. 50.61, percent
recovery due to thermal annealing (Rt and Ru) must be
evaluated by testing surveillance specimens that have been withdrawn
from the surveillance program and that have been annealed under the
same time and temperature conditions as those given the beltline
material.
(2) Alternatively, the percent recovery due to thermal annealing
(Rt and Ru) may be determined from the results of
[[Page 65474]]
a verification test program employing materials removed from the
beltline region of the reactor vessel 6 and that have been
annealed under the same time and temperature conditions as those given
the beltline material.
\6\ For those cases where materials are removed from the
beltline of the pressure vessel, the stress limits of the applicable
portions of the ASME Code Section III must be satisfied, including
consideration of fatigue and corrosion, regardless of the Code of
record for the vessel design.
---------------------------------------------------------------------------
(3) Generic computational methods may be used to determine recovery
if adequate justification is provided.
(f) Public information and participation.
(1) Upon receipt of a Thermal Annealing Report, and a minimum of 30
days before the licensee starts thermal annealing, the Commission
shall:
(i) Notify and solicit comments from local and State governments in
the vicinity of the site where the thermal annealing will take place
and any Indian Nation or other indigenous people that have treaty or
statutory rights that could be affected by the thermal annealing,
(ii) Publish a notice of a public meeting in the Federal Register
and in a forum, such as local newspapers, which is readily accessible
to individuals in the vicinity of the site, to solicit comments from
the public, and
(iii) Hold a public meeting on the licensee's Thermal Annealing
Report.
(2) Within 15 days after the NRC's receipt of the licensee
submissions required by paragraphs (c)(1), (c)(2) and (c)(3)(i)-(iii)
of this section, the NRC staff shall place in the NRC Public Document
Room a summary of its inspection of the licensee's thermal annealing,
and the Commission shall hold a public meeting:
(i) For the licensee to explain to NRC and the public the results
of the reactor pressure vessel annealing,
(ii) for the NRC to discuss its inspection of the reactor vessel
annealing, and
(iii) for the NRC to receive public comments on the annealing.
(3) Within 45 days of NRC's receipt of the licensee submissions
required by paragraphs (c)(1), (c)(2) and (c)(3)(i)-(iii) of this
section, the NRC staff shall complete full documentation of its
inspection of the licensee's annealing process and place this
documentation in the NRC Public Document Room.
5. In 10 CFR Part 50, Appendix G is revised to read as follows:
Appendix G to Part 50--Fracture Toughness Requirements
I. Introduction and scope.
II. Definitions.
III. Fracture toughness tests.
IV. Fracture toughness requirements.
I. Introduction and Scope
This appendix specifies fracture toughness requirements for
ferritic materials of pressure-retaining components of the reactor
coolant pressure boundary of light water nuclear power reactors to
provide adequate margins of safety during any condition of normal
operation, including anticipated operational occurrences and system
hydrostatic tests, to which the pressure boundary may be subjected
over its service lifetime.
The ASME Code forms the basis for the requirements of this
appendix. ``ASME Code'' means the American Society of Mechanical
Engineers Boiler and Pressure Vessel Code. If no section is
specified, the reference is to Section III, Division 1, ``Rules for
Construction of Nuclear Power Plant Components.'' ``Section XI''
means Section XI, Division 1, ``Rules for Inservice Inspection of
Nuclear Power Plant Components.'' If no edition or addenda are
specified, the ASME Code edition and addenda and any limitations and
modifications thereof, which are specified in Sec. 50.55a, are
applicable.
The sections, editions and addenda of the ASME Boiler and
Pressure Vessel Code specified in Sec. 50.55a have been approved for
incorporation by reference by the Director of the Federal Register.
A notice of any changes made to the material incorporated by
reference will be published in the Federal Register. Copies of the
ASME Boiler and Pressure Vessel Code may be purchased from the
American Society of Mechanical Engineers, United Engineering Center,
345 East 47th Street, New York, NY 10017, and are available for
inspection at the NRC Library, 11545 Rockville Pike, Two White Flint
North, Rockville, MD 20852-2738.
The requirements of this appendix apply to the following
materials:
A. Carbon and low-alloy ferritic steel plate, forgings,
castings, and pipe with specified minimum yield strengths not over
50,000 psi (345 MPa), and to those with specified minimum yield
strengths greater than 50,000 psi (345 MPa) but not over 90,000 psi
(621 MPa) if qualified by using methods equivalent to those
described in paragraph G-2110 of Appendix G of Section XI of the
latest edition and addenda of the ASME Code incorporated by
reference into Sec. 50.55a(b)(2).
B. Welds and weld heat-affected zones in the materials specified
in paragraph I.A. of this appendix.
C. Materials for bolting and other types of fasteners with
specified minimum yield strengths not over 130,000 psi (896 MPa).
Note: The adequacy of the fracture toughness of other ferritic
materials not covered in this section must be demonstrated to the
Director, Office of Nuclear Reactor Regulation, on an individual
case basis.
II. Definitions
A. Ferritic material means carbon and low-alloy steels, higher
alloy steels including all stainless alloys of the 4xx series, and
maraging and precipitation hardening steels with a predominantly
body-centered cubic crystal structure.
B. System hydrostatic tests means all preoperational system
leakage and hydrostatic pressure tests and all system leakage and
hydrostatic pressure tests performed during the service life of the
pressure boundary in compliance with the ASME Code, Section XI.
C. Specified minimum yield strength means the minimum yield
strength (in the unirradiated condition) of a material specified in
the construction code under which the component is built under
Sec. 50.55a.
D. RTNDT means the reference temperature of the material,
for all conditions.
(i) For the pre-service or unirradiated condition, RTNDT is
evaluated according to the procedures in the ASME Code, Paragraph
NB-2331.
(ii) For the reactor vessel beltline materials, RTNDT must
account for the effects of neutron radiation.
E. RTNDT means the transition temperature shift, or change
in RTNDT, due to neutron radiation effects, which is evaluated
as the difference in the 30 ft-lb (41 J) index temperatures from the
average Charpy curves measured before and after irradiation.
F. Beltline or Beltline region of reactor vessel means the
region of the reactor vessel (shell material including welds, heat
affected zones, and plates or forgings) that directly surrounds the
effective height of the active core and adjacent regions of the
reactor vessel that are predicted to experience sufficient neutron
radiation damage to be considered in the selection of the most
limiting material with regard to radiation damage.
III. Fracture Toughness Tests
A. To demonstrate compliance with the fracture toughness
requirements of Section IV of this appendix, ferritic materials must
be tested in accordance with the ASME Code and, for the beltline
materials, the test requirements of Appendix H of this part. For a
reactor vessel that was constructed to an ASME Code earlier than the
Summer 1972 Addenda of the 1971 Edition (under Sec. 50.55a), the
fracture toughness data and data analyses must be supplemented in a
manner approved by the Director, Office of Nuclear Reactor
Regulation, to demonstrate equivalence with the fracture toughness
requirements of this appendix.
B. Test methods for supplemental fracture toughness tests
described in paragraph IV.A.1.b of this appendix must be submitted
to and approved by the Director, Office of Nuclear Reactor
Regulation, prior to testing.
C. All fracture toughness test programs conducted in accordance
with paragraphs III.A and III.B must comply with ASME Code
requirements for calibration of test equipment, qualification of
test personnel, and retention of records of these functions and of
the test data.
IV. Fracture Toughness Requirements
A. The pressure-retaining components of the reactor coolant
pressure boundary that are made of ferritic materials must meet the
requirements of the ASME Code, supplemented by the additional
requirements set forth below, for fracture toughness during system
hydrostatic tests and any condition of
[[Page 65475]]
normal operation, including anticipated operational occurrences.
Reactor vessels may continue to be operated only for that service
period within which the requirements of this section are satisfied.
For the reactor vessel beltline materials, including welds, plates
and forgings, the values of RTNDT and Charpy upper-shelf energy
must account for the effects of neutron radiation, including the
results of the surveillance program of Appendix H of this part. The
effects of neutron radiation must consider the radiation conditions
(i.e., the fluence) at the deepest point on the crack front of the
flaw assumed in the analysis.
1. Reactor Vessel Charpy Upper-Shelf Energy Requirements
a. Reactor vessel beltline materials must have Charpy upper-
shelf energy,1 in the transverse direction for base material
and along the weld for weld material according to the ASME Code, of
no less than 75 ft-lb (102 J) initially and must maintain Charpy
upper-shelf energy throughout the life of the vessel of no less than
50 ft-lb (68 J), unless it is demonstrated in a manner approved by
the Director, Office of Nuclear Reactor Regulation, that lower
values of Charpy upper-shelf energy will provide margins of safety
against fracture equivalent to those required by Appendix G of
Section XI of the ASME Code. This analysis must use the latest
edition and addenda of the ASME Code incorporated by reference into
Sec. 50.55a(b)(2) at the time the analysis is submitted.
\1\ Defined in ASTM E 185-79 and -82 which are incorporated by
reference in Appendix H to Part 50.
---------------------------------------------------------------------------
b. Additional evidence of the fracture toughness of the beltline
materials after exposure to neutron irradiation may be obtained from
results of supplemental fracture toughness tests for use in the
analysis specified in section IV.A.1.a.
c. The analysis for satisfying the requirements of section
IV.A.1 of this appendix must be submitted, as specified in
Sec. 50.4, for review and approval on an individual case basis at
least three years prior to the date when the predicted Charpy upper-
shelf energy will no longer satisfy the requirements of section
IV.A.1 of this appendix, or on a schedule approved by the Director,
Office of Nuclear Reactor Regulation.
2. Pressure-Temperature Limits and Minimum Temperature Requirements
a. Pressure-temperature limits and minimum temperature
requirements for the reactor vessel are given in Table 3, and are
defined by the operating condition (i.e., hydrostatic pressure and
leak tests, or normal operation including anticipated operational
occurrences), the vessel pressure, whether or not fuel is in the
vessel, and whether the core is critical. In Table 3, the vessel
pressure is defined as a percentage of the preservice system
hydrostatic test pressure. The appropriate requirements on both the
pressure-temperature limits and the minimum permissible temperature
must be met for all conditions.
b. The pressure-temperature limits identified as ``ASME Appendix
G limits'' in Table 3 require that the limits must be at least as
conservative as limits obtained by following the methods of analysis
and the margins of safety of Appendix G of Section XI of the ASME
Code.
c. The minimum temperature requirements given in Table 3 pertain
to the controlling material, which is either the material in the
closure flange or the material in the beltline region with the
highest reference temperature. As specified in Table 3, the minimum
temperature requirements and the controlling material depend on the
operating condition (i.e., hydrostatic pressure and leak tests, or
normal operation including anticipated operational occurrences), the
vessel pressure, whether fuel is in the vessel, and whether the core
is critical. The metal temperature of the controlling material, in
the region of the controlling material which has the least favorable
combination of stress and temperature, must exceed the appropriate
minimum temperature requirement for the condition and pressure of
the vessel specified in Table 1.
d. Pressure tests and leak tests of the reactor vessel that are
required by Section XI of the ASME Code must be completed before the
core is critical.
B. If the procedures of Section IV.A. of this appendix do not
indicate the existence of an equivalent safety margin, the reactor
vessel beltline may be given a thermal annealing treatment to
recover the fracture toughness of the material, subject to the
requirements of Sec. 50.66. The reactor vessel may continue to be
operated only for that service period within which the predicted
fracture toughness of the beltline region materials satisfies the
requirements of Section IV.A. of this appendix using the values of
RTNDT and Charpy upper-shelf energy that include the effects of
annealing and subsequent irradiation.
Table 1.--Pressure and Temperature Requirements for the Reactor Pressure Vessel
----------------------------------------------------------------------------------------------------------------
Vessel Requirements for pressure-
Operating condition pressure temperature limits Minimum temperature requirements
-------------------------------------------\1\------------------------------------------------------------------
1. Hydrostatic pressure and leak
tests (core is not critical):
1.a Fuel in the vessel.......... ASME Appendix G Limits.... (\2\)
20%
1.b Fuel in the vessel.......... >20% ASME Appendix G Limits.... (\2\) +90 deg.F (\6\)
1.c No fuel in the vessel ALL (Not Applicable).......... (\3\) +60 deg.F
(Preservice Hydrotest Only).
2. Normal operation (incl. heat-up
and cool-down), including
anticipated operational occurrences:
2.a Core not critical........... ASME Appendix G Limits.... (\2\)
20%
2.b Core not critical........... >20% ASME Appendix G Limits.... (\2\) +120 deg.F (\6\)
2.c Core critical............... ASME Appendix G Limits + Larger of [(\4\)] or [(\2\) + 40
20% 40 deg.F. deg.F]
2.d Core critical............... >20% ASME Appendix G Limits + Larger of [(\4\)] or [(\2\) +
40 deg.F. 160 deg.F]
2.e Core critical for BWR (\5\). ASME Appendix G Limits + (\2\) + 60 deg.F
20% 40 deg.F.
----------------------------------------------------------------------------------------------------------------
\1\ Percent of the preservice system hydrostatic test pressure.
\2\ The highest reference temperature of the material in the closure flange region that is highly stressed by
the bolt preload.
\3\ The highest reference temperature of the vessel.
\4\ The minimum permissible temperature for the inservice system hydrostatic pressure test.
\5\ For boiling water reactors (BWR) with water level within the normal range for power operation.
\6\ Lower temperatures are permissible if they can be justified by showing that the margins of safety of the
controlling region are equivalent to those required for the beltline when it is controlling.
[[Page 65476]]
6. In 10 CFR Part 50, Appendix H is revised to read as follows:
Appendix H to Part 50--Reactor Vessel Material Surveillance Program
Requirements
I. Introduction
II. Definitions
III. Surveillance Program Criteria
IV. Report of Test Results
I. Introduction
The purpose of the material surveillance program required by
this appendix is to monitor changes in the fracture toughness
properties of ferritic materials in the reactor vessel beltline
region of light water nuclear power reactors which result from
exposure of these materials to neutron irradiation and the thermal
environment. Under the program, fracture toughness test data are
obtained from material specimens exposed in surveillance capsules,
which are withdrawn periodically from the reactor vessel. These data
will be used as described in Section IV of Appendix G to Part 50.
ASTM E 185-73, -79, and -82, ``Standard Practice for Conducting
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor
Vessels,'' which are referenced in the following paragraphs, have
been approved for incorporation by reference by the Director of the
Federal Register. Copies of ASTM E 185-73, -79, and -82, may be
purchased from the American Society for Testing and Materials, 1916
Race Street, Philadelphia, PA 19103 and are available for inspection
at the NRC Library, 11545 Rockville Pike, Two White Flint North,
Rockville, MD 20852-2738.
II. Definitions
All terms used in this Appendix have the same meaning as in
Appendix G.
III. Surveillance Program Criteria
A. No material surveillance program is required for reactor
vessels for which it can be conservatively demonstrated by
analytical methods applied to experimental data and tests performed
on comparable vessels, making appropriate allowances for all
uncertainties in the measurements, that the peak neutron fluence at
the end of the design life of the vessel will not exceed 1017
n/cm2 (E > 1 MeV).
B. Reactor vessels that do not meet the conditions of paragraph
III.A of this appendix must have their beltline materials monitored
by a surveillance program complying with ASTM E 185, as modified by
this appendix.
1. The design of the surveillance program and the withdrawal
schedule must meet the requirements of the edition of ASTM E 185
that is current on the issue date of the ASME Code to which the
reactor vessel was purchased. Later editions of ASTM E 185 may be
used, but including only those editions through 1982. For each
capsule withdrawal, the test procedures and reporting requirements
must meet the requirements of ASTM E 185-82 to the extent
practicable for the configuration of the specimens in the capsule.
2. Surveillance specimen capsules must be located near the
inside vessel wall in the beltline region so that the specimen
irradiation history duplicates, to the extent practicable within the
physical constraints of the system, the neutron spectrum,
temperature history, and maximum neutron fluence experienced by the
reactor vessel inner surface. If the capsule holders are attached to
the vessel wall or to the vessel cladding, construction and
inservice inspection of the attachments and attachment welds must be
done according to the requirements for permanent structural
attachments to reactor vessels given in Sections III and XI of the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code). The design and location of the capsule holders
must permit insertion of replacement capsules. Accelerated
irradiation capsules may be used in addition to the required number
of surveillance capsules.
3. A proposed withdrawal schedule must be submitted with a
technical justification as specified in Sec. 50.4. The proposed
schedule must be approved prior to implementation.
C. Requirements for an Integrated Surveillance Program.
1. In an integrated surveillance program, the representative
materials chosen for surveillance for a reactor are irradiated in
one or more other reactors that have similar design and operating
features. Integrated surveillance programs must be approved by the
Director, Office of Nuclear Reactor Regulation, on a case-by-case
basis. Criteria for approval include the following:
a. The reactor in which the materials will be irradiated and the
reactor for which the materials are being irradiated must have
sufficiently similar design and operating features to permit
accurate comparisons of the predicted amount of radiation damage.
b. Each reactor must have an adequate dosimetry program.
c. There must be adequate arrangement for data sharing between
plants.
d. There must be a contingency plan to assure that the
surveillance program for each reactor will not be jeopardized by
operation at reduced power level or by an extended outage of another
reactor from which data are expected.
e. There must be substantial advantages to be gained, such as
reduced power outages or reduced personnel exposure to radiation, as
a direct result of not requiring surveillance capsules in all
reactors in the set.
2. No reduction in the requirements for number of materials to
be irradiated, specimen types, or number of specimens per reactor is
permitted.
3. After (the effective date of this section), no reduction in
the amount of testing is permitted unless previously authorized by
the Director, Office of Nuclear Reactor Regulation.
IV. Report of Test Results
A. Each capsule withdrawal and the test results must be the
subject of a summary technical report to be submitted, as specified
in Sec. 50.4, within one year of the date of capsule withdrawal,
unless an extension is granted by the Director, Office of Nuclear
Reactor Regulation.
B. The report must include the data required by ASTM E 185, as
specified in paragraph III.B.1 of this appendix, and the results of
all fracture toughness tests conducted on the beltline materials in
the irradiated and unirradiated conditions.
C. If a change in the Technical Specifications is required,
either in the pressure-temperature limits or in the operating
procedures required to meet the limits, the expected date for
submittal of the revised Technical Specifications must be provided
with the report.
Dated at Rockville MD, this 12th day of December, 1995.
For the Nuclear Regulatory Commission.
John C. Hoyle,
Secretary of the Commission.
[FR Doc. 95-30665 Filed 12-18-95; 8:45 am]
BILLING CODE 7590-01-P
