[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





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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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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, RT<INF>NDT, 
including treatment of the unirradiated RT<INF>NDT 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, RT<INF>PTS. The method 
for calculating RT<INF>PTS 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 RT<INF>PTS.

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 

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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 RT<INF>NDT 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 RT<INF>NDT 
and Charpy upper-shelf energy, and (4) the projected values of 
RT<INF>PTS 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 RT<INF>NDT and Charpy upper-shelf energy of the beltline 
materials. The experimental methods for estimating recovery of 
RT<INF>NDT 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 RT<INF>NDT 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, RT<INF>NDT, and, at 
end of reactor life, RT<INF>PTS. 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 RT<INF>PTS 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 RT<INF>PTS 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 RT<INF>PTS were modified to make 
clear that RT<INF>PTS is a unique, end of life, case of RT<INF>NDT 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 RT<INF>PTS 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, RT<INF>NDT, for cases where plants do not have 
data to comply with code procedures for determining RT<INF>NDT. 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 RT<INF>NDT. 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, 

[[Page 65462]]
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 RT<INF>PTS 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 RT<INF>NDT into the PTS rule has no impact on the 
environment because this change will result in values of RT<INF>PTS 
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 
RT<INF>PTS using the same methodology specified in Regulatory Guide 
1.99, Revision 2, for determining RT<INF>NDT. 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]]

---------------------------------------------------------------------------

    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) RT<INF>NDT means the reference temperature for a reactor vessel 
material, under any conditions. For the reactor vessel beltline 
materials, RT<INF>NDT must account for the effects of neutron 
radiation.
    (5) RT<INF>NDT(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) RT<INF>PTS means the reference temperature, RT<INF>NDT, 
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 RT<INF>PTS 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 RT<INF>PTS, accepted by the NRC, for each reactor vessel 
beltline material for the EOL fluence of the material. The assessment 
of RT<INF>PTS 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 RT<INF>PTS 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 <SUP>2 change in projected 
values of RT<INF>PTS, or upon a request for a change in the expiration 
date for operation of the facility.

    \2\ Changes to RT<INF>PTS 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.
---------------------------------------------------------------------------

    (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 RT<INF>PTS 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. 
RT<INF>PTS 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 RT<INF>PTS 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 
RT<INF>PTS 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 RT<INF>PTS 
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 RT<INF>PTS 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 RT<INF>PTS 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 RT<INF>PTS 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 RT<INF>PTS 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 
RT<INF>PTS accounting for the effects of annealing and subsequent 
irradiation.
    (c) Calculation of RT<INF>PTS. RT<INF>PTS must be calculated for 
each vessel beltline material using a fluence value, f, which is the 
EOL fluence for the material. RT<INF>PTS must be evaluated using the 
same procedures used to calculate RT<INF>NDT, 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 RT<INF>NDT for 
each weld and plate, or forging, in the reactor vessel beltline.

Equation 1: RT<INF>NDT=RT<INF>NDT(U)+M+<greek-D>RT<INF>NDT

    (i) If a measured value of RT<INF>NDT(U) is not available, a 
generic mean value for the class <SUP>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 RT<INF>NDT(U) is 
generally determined for welds by the type of welding flux (Linde 
80, or other), and for base metal by the material specification.
---------------------------------------------------------------------------

    (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 RT<INF>NDT(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, <greek-s><INF>U is the standard deviation for 
RT<INF>NDT(U). If a measured value of RT<INF>NDT(U) is used, then 
<greek-s><INF>U is determined from the precision of the test method. If 
a measured value of RT<INF>NDT(U) is not available and a generic mean 
value for that class of materials is used, then <greek-s><INF>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 <greek-s><INF>U is 17 deg.F.
    (B) In Equation 2, <greek-s><greek-D> is the standard deviation for 
<greek-D>RT<INF>NDT. The value of <greek-s><greek-D> to be used is 
28 deg.F for welds and 17 deg.F for base metal; the value of 
<greek-s><greek-D> need not exceed one-half of <greek-D>RT<INF>NDT.
    (iv) <greek-D>RT<INF>NDT is the mean value of the transition 
temperature shift, or change in RT<INF>NDT, due to irradiation, and 
must be calculated using Equation 3.

Equation 3: <greek-D>RT<INF>NDT=(CF)f<SUP>(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 <SUP>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.''
---------------------------------------------------------------------------

    (B) f is the best estimate neutron fluence, in units of 10<SUP>19 
n/cm<SUP>2 (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 KRT<INF>PTS.
    (v) Equation 4 must be used for determining RT<INF>PTS using 
equation 3 with EOL fluence values for determining <greek-D>RT<INF>PTS.

Equation 4: RT<INF>PTS=RT<INF>NDT(U)+M+<greek-D>RT<INF>PTS

    (2) To verify that RT<INF>NDT 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 <SUP>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.
---------------------------------------------------------------------------

    (i) Results from the plant-specific surveillance program must be 
integrated into the RT<INF>NDT 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 <greek-D>RT<INF>NDT 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 <plus-minus>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, 
``A<INF>i'' is the measured value of <greek-D>RT<INF>NDT and 
``f<INF>i'' 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 <greek-D>RT<INF>NDT 
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 <greek-s><greek-D> to be 
used is 14 deg.F for welds and 8.5 deg.F for base metal; the value of 
<greek-s><greek-D> need not exceed one-half of DRT<INF>NDT.
    (iv) The use of results from the plant-specific surveillance 
program may result in an RT<INF>NDT that is higher or lower than those 
determined in paragraph (c)(1).
    (3) Any information that is believed to improve the accuracy of the 
RT<INF>PTS value significantly must be reported to the Director, Office 
of Nuclear Reactor Regulation. Any value of RT<INF>PTS 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 RT<INF>NDT 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 RT<INF>NDT and Charpy upper-shelf energy provide the basis 
for establishing the post-anneal RT<INF>NDT and Charpy upper-shelf 
energy for each vessel material. Changes in the RT<INF>NDT 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 RT<INF>NDT 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 RT<INF>NDT 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 RT<INF>NDT 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 RT<INF>NDT 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 
RT<INF>NDT and Charpy upper-shelf energy; and
    (4) The projected values of RT<INF>PTS 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 <greek-D>RT<INF>NDT, R<INF>t, and percent recovery of 
Charpy upper-shelf energy, R<INF>u. In all cases, R<INF>t and R<INF>u 
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 (R<INF>t and R<INF>u) 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 
(R<INF>t and R<INF>u) may be determined from the results of 

[[Page 65474]]
a verification test program employing materials removed from the 
beltline region of the reactor vessel <SUP>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. RT<INF>NDT means the reference temperature of the material, 
for all conditions.
    (i) For the pre-service or unirradiated condition, RT<INF>NDT is 
evaluated according to the procedures in the ASME Code, Paragraph 
NB-2331.
    (ii) For the reactor vessel beltline materials, RT<INF>NDT must 
account for the effects of neutron radiation.
    E. RT<INF>NDT means the transition temperature shift, or change 
in RT<INF>NDT, 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 RT<INF>NDT 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,<SUP>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 
RT<INF>NDT 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..........  <ls-thn-eq>  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...........  <ls-thn-eq>  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...............  <ls-thn-eq>  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\).  <ls-thn-eq>  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 10<SUP>17 
n/cm<SUP>2 (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