[Federal Register Volume 59, Number 191 (Tuesday, October 4, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-24209]


[[Page Unknown]]

[Federal Register: October 4, 1994]


                                                   VOL. 59, NO. 191

                                           Tuesday, October 4, 1994

NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

RIN: 3150-AD57

 

Fracture Toughness Requirements for Light Water Reactor Pressure 
Vessels

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend 
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 proposed amendments would 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.

DATES: The comment period expires January 3, 1994. Comments received 
after this date will be considered if it is practical to do so, but the 
Commission is able to assure consideration only for comments received 
on or before this date.

ADDRESSES: Mail comments to: The Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, Attention: 
Docketing and Service Branch.
    Comments may be delivered to: One White Flint North, 11555 
Rockville Pike, Rockville, MD 20852, between 7:30 am and 4:15 pm on 
Federal workdays. Comments received on the proposed rules may be 
examined at the NRC Public Document Room, 2120 L Street NW. (Lower 
Level), Washington, DC.
    A free single copy of draft regulatory guides DG-1023, DG-1025, and 
DG-1027 may be requested by those considering public comment by writing 
to the U.S. Nuclear Regulatory Commission, ATTN: Distribution and Mail 
Services Section, Room P-130A, Washington, DC 20555. A copy is also 
available for inspection and/or copying in the NRC Public Document 
Room, 2120 L Street NW. (Lower Level), Washington, DC.

FOR FURTHER INFORMATION CONTACT: Alfred Taboada, Division of 
Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555, telephone: (301) 415-6014.

SUPPLEMENTARY INFORMATION:

Background

    Maintaining the structural integrity of the reactor pressure vessel 
(RPV) 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 measurement methods and procedures to 
assure that each RPV has adequate safety margins for continued 
operation. Structural integrity of a reactor pressure vessel 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 a proposed rule on thermal annealing and 
several draft regulatory guides, the existing and proposed regulatory 
documents contribute to a comprehensive set of regulations and 
regulatory guidance pertaining to RPV integrity, including:

1. The fracture toughness criteria that each RPV must satisfy (10 CFR 
50.60, 10 CFR 50.61, and 10 CFR Part 50, Appendix G).
2. Irradiation embrittlement surveillance requirements (10 CFR Part 50, 
Appendix H).
3. Guidance for estimating the fracture toughness of the RPV 
(Regulatory Guide 1.99 and a draft regulatory guide on dosimetry).
4. Guidance for cases in which the RPV is estimated to exceed specified 
screening criteria (Regulatory Guide 1.154 and a draft regulatory guide 
on evaluating RPVs with Charpy upper-shelf energy less than 50 ft-lb).
5. Requirements and guidance for using thermal annealing of the RPV as 
a method for mitigating the effects of neutron irradiation (proposed 10 
CFR 50.66 and a draft regulatory guide).

    This notice proposes changes to the following requirements:

a. The Pressurized Thermal Shock (PTS) rule (10 CFR 50.61).
b. Appendix G of 10 CFR Part 50, ``Fracture Toughness Requirements.''
c. Appendix H of 10 CFR Part 50, ``Reactor Vessel Material Surveil- 
lance Program Requirements.''

    This notice also proposes a thermal annealing requirement, 10 CFR 
50.66. In addition to the proposed amendments addressed in this 
document, the NRC will publish for public comment a draft regulatory 
guide on thermal annealing, DG-1027 (the availability of this draft 
regulatory guide for will be announced in the Federal Register at a 
later date comment.)
    Two related draft regulatory guides have been published for public 
comment (58 FR 51392; October 1, 1993). These draft regulatory guides 
are:

1. A draft regulatory guide that addresses evaluation of RPVs with 
Charpy upper-shelf energy levels less than 50 ft-lb (DG-1023).
2. A draft regulatory guide on dosimetry (DG-1025).

    Other regulatory guides pertaining to RPV integrity, Regulatory 
Guides 1.99 and 1.154, are under evaluation. Revisions to these guides, 
if any, will be addressed in the future.
    Other regulatory issues related to reactor pressure vessel 
integrity, such as low temperature over-pressure protection system 
setpoints, are being addressed as part of a broader scope evaluation of 
the pressure vessel regulations and are not part of this proposed 
amendment.

Reasons for the Proposed Changes

    The needs for these proposed amendments to the fracture toughness 
regulations have been identified from three sources:

1. The 1989 Nuclear Utility Backfitting and Reform Group (NUBARG) 
appeal concerning use of nuclear heat to warm the RPV for system 
leakage and hydrostatic pressure tests;
2. The 1990 review of the RPV integrity of the Yankee Nuclear Power 
Station; and
3. A comprehensive review of the regulations by NRC staff, resulting in 
the identification of the need for clarifications, corrections, and 
improved guidance in certain areas.

    The recognition in 1986 by NRC staff that certain boiling water 
reactor (BWR) units were using nuclear heat to warm the system prior to 
performing leakage and pressure tests led to an NRC staff initiative to 
amend Appendix G to 10 CFR Part 50. During the NRC staff review to 
determine if this use of nuclear heating was permissible under either 
the ASME Code or the NRC regulations, the Nuclear Utility Backfitting 
and Reform Group (NUBARG) filed a backfitting claim, and later an 
appeal of the determination that a backfit was not involved. Stemming 
from this claim and appeal process, the Committee to Review Generic 
Requirements recommended to the Executive Director for Operations that 
the affected portions of Appendix G be revised to clearly indicate that 
all required leakage and pressure tests of the reactor pressure vessel 
must be performed when the core is not critical.
    In 1990, the NRC began a review of the integrity of the Yankee 
Nuclear Power Station (YNPS) RPV. That review, along with stated plans 
by the licensee to consider thermal annealing of the RPV, highlighted 
the need for the NRC to amend its regulations and guidance pertaining 
to RPV integrity. The NRC staff proposed a plan to revise and clarify 
the pertinent regulations in SECY-91-333 (October 22, 1991) and SECY-
92-283 (August 14, 1992), including schedules and general descriptions 
of the changes contemplated. The proposed changes included 
clarifications and corrections planned prior to the YNPS review. 
However, the YNPS review identified the need to clarify the 
requirements in Sections IV and V of Appendix G to 10 CFR Part 50, and 
the need to provide more complete requirements and guidance for thermal 
annealing.
    The PTS rule, 10 CFR 50.61, was amended on May 15, 1991 (56 FR 
22300) to make the method for evaluating irradiation embrittlement 
consistent with the recommended procedures of Regulatory Guide 1.99, 
Revision 2, ``Radiation Embrittlement of Reactor Vessel Materials.'' 
Subsequent inquiries to the Commission concerning the appropriate 
margin terms and use of surveillance data indicated that the PTS rule 
required clarification. A recent review of the rule by NRC staff 
concluded that the PTS rule should also be modified to bring the 
procedures for evaluating RTPTS into complete agreement with the 
recommended procedures in Regulatory Guide 1.99, Revision 2.

Overview of the Proposed Changes

PTS Rule (10 CFR 50.61)
    The pressurized thermal shock rule, 10 CFR 50.61, was initially 
published in final form on July 23, 1985 (50 FR 29937) and amended on 
May 15, 1991 (56 FR 22300). This rule provides a screening criterion 
for irradiation embrittlement of RPV beltline materials, above which 
the plant cannot continue to be operated without justification. 
Historically, a value of reference temperature has been determined for 
each vessel beltline material for comparison to the PTS screening 
criteria. These values of reference temperature are termed RTPTS 
values. However, the method for evaluating RTPTS values has not 
been consistent with the embrittlement estimates used for other 
purposes, such as pressure-temperature limit calculations. The May 15, 
1991, amendment was a step towards unifying the embrittlement estimate 
methodology. The amendment included the procedures given in Regulatory 
Guide 1.99, Revision 2, for the evaluation of irradiation embrittlement 
of the RPV beltline materials. The 1991 amendment left two differences 
between the rule and Regulatory Guide 1.99, Revision 2. These two 
differences are:
    1. Values of unirradiated RTNDT are specified for general 
classes of material in the PTS rule, while greater flexibility in 
determining unirradiated values is permitted in Regulatory Guide 1.99, 
Revision 2; and
    2. The margin terms used in the PTS rule are based on assumptions 
which are not consistent with the method used in Regulatory Guide 1.99, 
Revision 2 for calculating the margin term.
    This proposed amendment is intended to make the evaluation of 
RTPTS consistent with the recommended methods of Regulatory Guide 
1.99, Revision 2, which are used to evaluate RTNDT. In this case, 
the RTPTS value for each vessel beltline material is simply the 
RTNDT value estimated for the projected end of license fluence.
    This proposed amendment to the PTS rule would make three changes:
    1. The Regulatory Guide 1.99, Revision 2, method for determining 
RTNDT, of which RTPTS is a unique value determined for the 
end of license fluence, would be incorporated in total, including 
treatment of the unirradiated RTNDT value, the margin term, and 
the explicit definition of ``credible'' surveillance data.
    2. The section would be restructured to improve clarity, with the 
requirements section giving only the requirements for the RTPTS 
value. The method for calculating RTPTS would be moved to a new 
paragraph of the rule.
    3. Thermal annealing would be introduced as a method for mitigating 
the effects of neutron irradiation, thereby reducing RTPTS.
    Additionally, it should be noted that evaluations of current 
surveillance data have indicated that the standard deviation of the 
differences between predicted and measured shifts in RTNDT, termed 
the residual, are higher than the margin values used in the PTS rule 
and in Regulatory Guide 1.99, Revision 2, particularly for plate 
materials. However, the mean embrittlement estimation equations in the 
rule and in Regulatory Guide 1.99, Revision 2, overestimate the actual 
surveillance data by less than 10 deg.F on average. The NRC staff 
considered amending the PTS rule to incorporate the revised margin 
terms and the overestimation bias, but decided against such an 
amendment due to the small number of plants that would be affected by 
such a change, and a related NRC research program that addresses the 
overall issue of irradiation embrittlement correlations. The number of 
plants which would have their RTPTS values change 
``significantly'' by such a change to the margin terms is not large; 
the impact of the revised margin terms on those plants is being 
addressed through other regulatory mechanisms. The effect of the 
revised margin on pressure-temperature limits is being handled in a 
similar manner.
    As noted, this proposed amendment to 10 CFR 50.61 introduces 
thermal annealing of the reactor pressure vessel beltline as a method 
for mitigating the effects of neutron irradiation and reducing 
RTPTS to levels below the screening criteria. As specified in 
Sec. 50.61(b)(7) of this proposed rule, the use of thermal annealing 
would be subject to the requirements of the proposed new section on 
thermal annealing (10 CFR 50.66).

Thermal Annealing Rule (10 CFR 50.66)

    The proposed thermal annealing rule, 10 CFR 50.66, would provide a 
consistent set of requirements for the use of thermal annealing to 
mitigate the effects of neutron irradiation. The proposed rule would 
replace the requirements for annealing in the current Appendix G of 10 
CFR Part 50 with the proposed consistent set of requirements in this 
proposed rule. Also, the PTS rule would be amended to add a new 
paragraph (b)(7) which would reference the proposed thermal annealing 
rule as a method for mitigating the effects of neutron irradiation, 
thereby reducing RTPTS. Therefore, the intent of the thermal 
annealing rule, and related changes in 10 CFR 50.61 and Appendix G of 
10 CFR Part 50, is to provide requirements for use of thermal annealing 
to mitigate the deleterious effects of neutron irradiation on reactor 
vessel material properties.
    Consistent with guidance in Section V.D of the current Appendix G, 
the proposed thermal annealing rule would specify that thermal 
annealing would be subject to the approval of the Director, Office of 
Nuclear Reactor Regulation (NRR). Section 50.66(a) of the proposed 
thermal annealing rule would require submittal of an application 
containing three sections: a thermal annealing operating plan, a 
requalification inspection and test program, and a fracture toughness 
recovery and reembrittlement rate assurance program. This application 
would be required to be submitted at least three years before the 
proposed date of the annealing operation. This three-year period is 
specified only to provide NRC staff with sufficient time to review the 
thermal annealing application. The licensee may initiate the thermal 
annealing program as soon as NRC approval is given, even if this 
approval is given before three years from the date of the application.
    The thermal annealing operating plan also must include an 
evaluation of the effects of temperature, and of mechanical and thermal 
stresses on the reactor and associated equipment to demonstrate that 
the operability of the reactor will not be detrimentally affected. 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.
    Upon completion of the thermal annealing and before subsequent 
operation of the plant, the licensee would be required to certify that 
the thermal annealing was performed in accordance with the approved 
application, that the approved criteria were satisfied, and that the 
proposed annealing conditions were not exceeded. However, in the event 
that the licensee cannot make this certification, a justification for 
subsequent operation would have to be submitted for approval by the 
Director, NRR. However, this provision does not relieve the licensee 
from obtaining 10 CFR 50.12 exemptions from any other requirements of 
this part that cannot be satisfied.
    Two items of particular importance to the overall annealing are the 
recovery of fracture toughness and the rate of reembrittlement of the 
RPV beltline materials. This proposed rule provides three alternative 
methods for determining these values, ranging from assessments using 
plant-specific materials to an assessment using a generic computation.
    Two methods for evaluating annealing recovery are experimental 
methods to determine plant-specific annealing recovery, and the third 
method is a generic computational method. The first method would be 
required for those plants with ``credible'' surveillance programs and 
where surveillance specimens are available. The second method would be 
an optional method, in which the licensee may remove material from the 
beltline of the RPV to evaluate annealing recovery. This method should 
provide the most accurate evaluation of annealing recovery. Presumably, 
it would be selected for those plants without credible surveillance 
programs or when surveillance specimens are not available. However, for 
this method to be acceptable, the vessel must be sufficiently thick so 
that the stress limits in Section III of the ASME Code can be 
satisfied, considering the effects of fatigue and corrosion.
    The third method would use generic computational methods, for which 
appropriate justification would be required.
    Paragraph (d) of Sec. 50.66 describes the experimental methods and 
the computational method for estimating recovery of RTNDT and 
Charpy upper-shelf energy of the beltline materials. The experimental 
methods for estimating recovery of RTNDT and the Charpy upper-
shelf energy utilize either surveillance program specimens or material 
removed from the vessel beltline. The experimental methods provide a 
plant-specific estimate of recovery, rather than the generic value 
evaluated from the computational method. It is the intent of this 
proposed rule to require that surveillance specimens from ``credible'' 
surveillance programs must be used to develop plant-specific recovery 
data, if such specimens are available. It is not the intent of this 
rule to require the removal of material from the RPV beltline to permit 
plant-specific evaluation of recovery.
    As described previously, the computational method would require 
appropriate justification.
    Reembrittlement trends are estimated, and monitored by continued 
surveillance in accordance with Appendix H of 10 CFR Part 50.
    Paragraph (b)(3)(ii) provides that the reembrittlement rate must be 
monitored using a surveillance program which conforms to Appendix H of 
this part. Some older plants conform to Appendix H by applying issues 
of ASTM Standard E 185 that do not require the use of the vessel 
``limiting materials'' in the surveillance program. Within this 
context, the term ``limiting materials'' refers to the materials 
predicted to have the highest RTNDT or the lowest Charpy upper 
shelf energy during the operational lifetime of the plant. It is the 
intent of this rule that, as required by later issues of ASTM Standard 
E 185, the vessel ``limiting materials'' should be used to monitor 
reembrittlement if the materials are available.

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 proposed 
amendments to Appendix G are principally of a clarifying or a 
restructuring nature. These amendments include:
    1. Sections IV and V of Appendix G which would be combined and 
rewritten to clarify the Charpy upper-shelf energy requirements, and 
the pressure-temperature and minimum permissible temperature 
requirements.
    2. An explicit statement that would be added to Section IV 
requiring 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.
    3. The proposed thermal annealing rule, 10 CFR 50.66, that would be 
referenced in lieu of the details on thermal annealing previously given 
in Section V.D.
    4. The reference to the ASME Code that would be changed from 
Appendix G of Section III to Appendix G of Section XI of the ASME Code.
    5. The ``design to permit annealing'' requirement (Section IV.B), 
which would be deleted.
    The restructuring of Sections IV and V is intended to promote 
clarity of the requirements in these sections. The procedures required 
for cases in which the Charpy upper-shelf energy of a RPV beltline 
material falls below 50 ft-lb also would be clarified by consolidating 
the requirements previously addressed in parts of Sections IV and V.
    The provisions in Section V.C concerning requirements for 
``volumetric inspection'' and ``additional evidence of fracture 
toughness'' would be removed. The volumetric examination requirement 
would be removed because it was 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 would be incorporated in the ``equivalent margins'' 
analysis in Section IV.A.1, as a provisional method for developing 
fracture toughness data needed for that analysis. At the present time 
there is an adequate generic fracture toughness data base available to 
perform these analyses, with appropriate bounding considerations. The 
modification would permit a licensee to develop plant-specific data. 
Generally, plant-specific data would result in a reduction in the 
margin applied to the fracture toughness data, to reflect the reduction 
in uncertainties due to the availability of plant-specific data. 
However, this must be evaluated on a case-by-case basis.
    The pressure-temperature and minimum permissible temperature 
requirements in Section IV would be restructured, with the principal 
feature being 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 would be 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.
    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, before the completion of these tests. The use of 
nuclear heat before the completion of such tests is considered unsafe 
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 recommended to the 
Executive Director for Operations by the Committee to Review Generic 
Requirements in a memorandum dated June 7, 1990.
    The requirements on thermal annealing contained in the current 
Appendix G (Section V.D) would be replaced by a reference to the 
proposed Thermal Annealing rule, 10 CFR 50.66.
    Changing the reference to Appendix G of the ASME Code from Section 
III to Section XI means that the requirements for operating plants will 
no longer come from the construction code (Section III of the ASME 
Code) but instead will come from Section XI, the in-service inspection 
code. Appendix G to Section XI and Appendix G to Section III are 
identical, so this amendment would not result in a change in technical 
requirements.
    Section IV.B of Appendix G requires that:
    ``Reactor vessels for which the predicted value of upper-shelf 
energy at end of life is below 50 ft-lb or for which the predicted 
value of adjusted reference temperature at end of life exceeds 200 
deg.F (93  deg.C) must be designed to permit a thermal annealing 
treatment * * *''
    This proposed rule would delete that requirement. This deletion 
conforms with Commission direction from 1985 and public comments to 
delete this section. An additional consideration to delete this 
requirement is that there should be no requirement to ensure the 
feasibility of a (future) voluntary activity.
    During the Commission review of the revision of Appendix G 
published final on May 27, 1983 (48 FR 24009), the requirement to 
``design to permit annealing'' was criticized because licensee response 
to the requirement was perfunctory and staff review of the responses 
was cursory, as detailed in SECY-83-254 (June 27, 1983). Further, there 
were no criteria to assess whether a design would permit annealing. An 
additional problem cited with the requirement was that it was 
misinterpreted to mean that plant operation with an RTNDT greater 
than 200 deg.F or a Charpy upper-shelf energy below 50 ft-lb was 
unsafe. The Commission indicated that it would seek public comments on 
the proposed deletion of the requirement, and this was done 
concurrently with the publication of the proposed PTS rule on February 
4, 1984 (49 FR 4498). All sixteen of the commenters on this item 
recommended deletion of the paragraph, although eight of them urged 
that the deletion should not in any way imply that annealing is no 
longer an option to increase safety margins. In the notice of final 
rulemaking for the PTS rule published on July 23, 1985 (50 FR 29944), 
the ``Supplementary Information'' noted that the Commission planned a 
separate rulemaking action to delete Section IV.B. That planned 
deletion was delayed so that it could be combined with other amendments 
to Appendix G.

Appendix H of 10 CFR Part 50

    Changes in the fracture toughness properties of the RPV beltline 
materials due to irradiation embrittlement are monitored using a 
surveillance program, as required in Appendix H of 10 CFR Part 50, 
``Reactor Vessel Material Surveillance Program Requirements''. 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. This proposed 
amendment would make the following changes:

1. End the provision for ``reducing the amount of testing'' for 
integrated surveillance programs,
2. Restructure the section on requirements for integrated surveillance 
programs (Section II.C), and
3. Clarify the version of ASTM Standard E 185 that applies to the 
surveillance program.

    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.'' It is proposed to discontinue this provision as of 
the effective date of the Appendix, although previous authorizations 
granted by the Director, Office of Nuclear Reactor Regulation, would 
continue in effect.
    A second proposed change to Appendix H restructures Section II.C to 
clarify the 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 subsequent testing and reporting of results from 
the surveillance capsules. Since 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. The clarification proposed in 
this revision indicates that the design of the program and the 
withdrawal schedule must meet the requirements of ASTM E 185-73, or the 
edition of ASTM E 185 that is current on the issue date of the ASME 
Code to which the reactor vessel was purchased, whichever is latest. 
Licensees could choose to comply with later editions of ASTM E 185, up 
through the 1982 edition. Further, specimen 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.''
    The NRC staff intended that this proposed amendment to Appendix H 
would incorporate by reference a version of ASTM standard E 185 updated 
from the currently available 1982 version. However, that 
standardization process has not been completed, and it was decided to 
proceed with this proposed amendment. A subsequent amendment to 
Appendix H will be considered after the NRC staff has reviewed the 
updated ASTM standard.

Request for Public Comments

    On June 13, 1994 (SECY-94-163) the staff requested Commission 
approval to publish for public comment these proposed revisions and 
provided a discussion of options for public participation related to 
thermal annealing. The Commission approved issuance of the proposed 
revisions but directed that the staff to (1) include with the proposed 
rule package a discussion of options the staff has considered for 
structuring of the regulatory process for the proposed thermal 
annealing rule (10 CFR 50.66), which is included in the following 
section, and (2) request comments on the following 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.

Options the Staff Has Considered for Structuring of the Regulatory 
Process Related to Public Participation in Thermal Annealing
    A significant issue with respect to thermal annealing, identified 
in SECY-92-283 (August 14, 1992), is the nature and timing of public 
participation related to the NRC's review and approval of a licensee's 
proposal for thermal annealing. The proposed rule does not address 
public participation per se, but instead provides the performance 
requirements that a licensee would have to meet to gain NRC approval of 
a thermal annealing application and to permit subsequent operation. 
Under the proposed rule, there are three circumstances that arguably 
require an opportunity for hearing pursuant to Section 189 of the 
Atomic Energy Act of 1954 (AEA) in connection with NRC review and 
approval of thermal annealing. First, a licensee seeking to anneal its 
reactor vessel must obtain NRC approval of the content of the thermal 
annealing plan prior to implementing the plan (see Sec. 50.66(a) of the 
proposed rule). Second, and apart from the NRC approval required under 
Sec. 50.66(a), the thermal annealing process as described in the 
licensee's plan may necessitate license amendments (including technical 
specification changes). License amendments may be required if the 
licensee's final safety analysis report (FSAR) needs to be revised to 
reflect the thermal annealing process, and the licensee is unable to 
conclude that such FSAR changes do not constitute ``unreviewed safety 
questions'' under 10 CFR 50.59. Implementation of the thermal annealing 
plan may also violate existing technical specifications, necessitating 
requests for changes to technical specifications. Any license amendment 
and technical specification change must be approved by the NRC before 
the licensee may implement the thermal annealing plan. Finally, after 
the licensee implements the annealing plan, if he determines that he 
cannot meet the criteria specified in Sec. 50.66(c)(1) of the proposed 
rule, then NRC approval is needed in order for the licensee to resume 
operation (see Sec. 50.66(c)(2) of the proposed rule).
    It is clear that any license amendments and technical specification 
changes necessitated by the thermal annealing plan would require an 
opportunity for hearing, in accordance with Section 189 of the AEA. 
However, the scope of such a hearing would normally be limited to 
consideration of whether the proposed license amendment and technical 
specification changes are in accordance with the Commission's rules, 
and therefore provide reasonable assurance of adequate protection to 
the public health and safety. Issues related to the more general matter 
of the acceptability of the thermal annealing plan proposed by the 
licensee would not fall within the scope of any hearing for license 
amendment or technical specification change, except as they fall 
directly in the scope of the requested amendment or technical 
specification change.
    However, there is some question whether the AEA requires an 
opportunity for hearing in connection with the NRC approval of the 
thermal annealing plan or the NRC decision approving resumption of 
operation under the proposed rule. There are four primary alternatives 
with respect to providing an opportunity for hearing in connection with 
thermal annealing. These alternatives are discussed in greater detail 
below:
Alternative 1. No Opportunity for Hearing Under Rule as Proposed
    Under this alternative, the contention is that Section 189 of the 
AEA does not afford an interested member of the public a right to 
request a hearing in connection with NRC approvals of thermal annealing 
plans and resumption of operation under Sec. 50.66(c)(2). This 
alternative is consistent with other provisions in 10 CFR Part 50 where 
approval by the Director of NRR is required and hearings are not 
routinely offered.
    Notwithstanding the lack of a requirement for a public hearing, the 
staff anticipates that, with respect to the initial or the first 
several applications for thermal annealing, several informal hearings 
or public meetings would be held by the staff to permit discussion of 
both the thermal annealing plan proposed by the licensee and the 
technical issues related to annealing. These hearings or meetings would 
ensure that all of the pertinent technical issues have been addressed 
by the licensee in its thermal annealing plan and by the staff in its 
review of the plan. These hearings or meetings would be noticed in the 
Federal Register.
Alternative 2. Discretionary Opportunity for Hearing Under Rule as 
Proposed
    Under this alternative, the contention is that Section 189 of the 
AEA does not afford an interested member of the public a right to 
request a hearing. However, as a matter of discretion, the Commission 
would determine on a case-by-case basis whether an opportunity for 
hearing will be provided in connection with the Director of NRR's 
determination on a thermal annealing application under Sec. 50.66(b) of 
the proposed rule. In the hearing, the Commission would consider issues 
related to the adequacy of the thermal annealing plan, as well as the 
vessel's ability to perform its safety function after being annealed.
    A case-by-case determination would also be made by the NRC with 
respect to providing an opportunity for hearing on the Director of 
NRR's determination on the licensee's justification for subsequent 
operation under the proposed Sec. 50.66(c)(2).
    In both cases, the Commission would publish a notice in the Federal 
Register announcing the NRC's approval of the licensee's thermal 
annealing plan or approval of resumed operation under Sec. 50.66(c)(2). 
Neither implementation of the thermal annealing plan nor resumption of 
operation, once approved by the NRC, would be contingent upon 
completion of any hearing; i.e., the Commission does not believe that 
it is required to make a Section 189 ``no significant hazards 
determination'' (``Sholly finding'') when it provides a discretionary 
hearing.
Alternative 3. Required Opportunity for Hearing Under Rule as Proposed
    Under this alternative, the contention is that a hearing is 
required by Section 189 of the AEA for both NRC's approval of the 
thermal annealing plan and any NRC approval of resumed operation 
following annealing. The adequacy of the thermal annealing plan, as 
well as the vessel's ability to perform its safety function after being 
annealed, could be raised in the hearing associated with approval of 
the thermal annealing plan. Licensee implementation of the thermal 
annealing plan could not commence until any hearing is concluded unless 
the NRC makes a ``no significant hazards determination'' with respect 
to the thermal annealing.
Alternative 4. Modify Proposed Rule to Require Suspension of License 
Prior to Thermal Annealing
    Under this alternative, the proposed rule's regulatory approach for 
thermal annealing would be modified to include a suspension of the 
operating license during thermal annealing. The suspension would be 
automatic under the rule, without the need for a suspension order, 
although a letter confirming the licensee's status under the annealing 
rule would be prepared. The rule itself, as is currently drafted, would 
specify the conditions for lifting of the suspension (Section 
50.66(b)). The licensee would anneal its reactor vessel without prior 
NRC approval of its program for conducting the annealing. Upon 
completion, the suspension would be lifted only if the licensee 
demonstrated that the thermal annealing has addressed the reactor 
enbrittlement such that it is acceptable to operate the plant. There 
would be no opportunity for hearing associated with the lifting of the 
suspension, and since there would be no prior NRC approval of the 
annealing program, a hearing opportunity under Section 189 would not be 
implicated by any such approval.
Submission of Comments in Electronic Format
    Commenters are encouraged to submit, in addition to the original 
paper copy, a copy of the letter in electronic format on a DOS-
formatted (IBM compatible) 5.25 or 3.5 inch computer diskette. Text 
files should be provided in WordPerfect format or unformatted ASCII 
code. The format and version should be identified on the diskette's 
external label.
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, if adopted, would not be a 
major Federal action significantly affecting the quality of human 
environment and, therefore, an environmental impact statement is not 
required.
    As discussed below, the individual actions covered in this proposed 
rulemaking would either serve to enhance safety of the reactor pressure 
vessel, thereby decreasing the environmental impact of plant operation, 
or have no 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 would serve to decrease the 
environmental impact of plant operation by enhancing the safety of the 
reactor pressure vessel.
    The incorporation of the Regulatory Guide 1.99, Revision 2, method 
for determining RTNDT into the PTS rule would have no impact on 
the environment because this change will result in values of RTPTS 
which are consistent with those currently used in plant operation.
    The restructuring of the PTS rule is the type of action described 
in categorical exclusion 10 CFR 51.22(c)(2). Therefore, an 
environmental assessment is not necessary for this change.

Thermal Annealing Rule (10 CFR 50.66)

    The proposed thermal annealing rule (10 CFR 50.66) would permit and 
provide 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 rule does not affect all 
plants but provides an alternative for assuring compliance with the 
requirements in 10 CFR 50.61 and Appendix G of 10 CFR Part 50, and 
would only apply when a licensee elects to use it.
    The application of thermal annealing to a reactor vessel would 
improve the condition of the reactor vessel material. In addition, this 
rule would establish requirements to avoid damaging the reactor system 
and to protect against accidents during the annealing operation, with 
attendant environmental consequences.
    This rule is one of several regulatory requirements that will 
function to ensure reactor vessel integrity. In that sense, this rule 
would have a positive impact on the environment by reducing the 
potential for vessel failure. For these reasons, the Commission has 
determined that there would be no significant impact and, therefore, an 
environmental statement is not required.

Appendix G of 10 CFR Part 50

    Concerning the amendments proposed to Appendix G of 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 environmental impact of plant 
operation.
    The restructuring of Sections IV and V of Appendix G is clarifying 
or corrective in nature, and hence 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 since 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 since Appendix G currently permits the use of 
thermal annealing to reduce fracture toughness loss of the RPV 
materials due to irradiation embrittlement.
    The deletion of the ``design to permit annealing'' requirement has 
no adverse impact on the environment. This assessment is based on the 
fact that annealing is a voluntary activity.

Appendix H of 10 CFR Part 50

    Concerning the amendments proposed to Appendix H of 10 CFR Part 50, 
the requirement that all irradiation surveillance tests be made (i.e., 
no reduction in testing is permitted) would 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 proposed rule amends information collection requirements that 
are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et 
seq.). This rule has been submitted to the Office of Management and 
Budget for review and approval of the paperwork requirements.
    The public reporting burden for this collection of information is 
estimated to average 6,000 hours per respondent, 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, D.C. 20555; and to the Desk Officer, Office of Information 
and Regulatory Affairs, NEOB-3019, (3150-0011), Office of Management 
and Budget, Washington, D.C. 20503.

Regulatory Analysis

    The NRC staff has prepared a regulatory analysis for the proposed 
amendments to 10 CFR 50.61, 10 CFR Part 50, Appendix G and 10 CFR Part 
50 Appendix H, which describes the factors and alternatives considered 
by the Commission in deciding to propose 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. Single copies of the analysis may be obtained 
from Alfred Taboada, Office of Nuclear Regulatory Research, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, telephone, (301) 
415-6014.
    Single copies of the regulatory analysis prepared for 10 CFR 50.66 
may be obtained from Alfred Taboada, Office of Nuclear Regulatory 
Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 
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, if adopted, the proposed amendments would 
not have a significant economic impact on a substantial number of small 
entities. The rules which would be affected by the proposed amendments: 
(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 or the Small Business Size Standards in regulations 
issued by the Small Business Administration at 13 CFR Part 121.

Backfit Analysis

PTS Rule (10 CFR 50.61)

    The proposed revision to Section 50.61 would require licensees to 
calculate RTPTS using the same methodology specified in Regulatory 
Guide 1.99, Revision 2 for determining RTNDT. This proposal is 
logically a requisite part of the 1991 revisions to Sec. 50.61, which 
addressed a unified method for calculating radiation enbrittlement of 
the reactor beltline materials. However, the Commission inadvertently 
failed to make the conforming change to Sec. 50.61. Therefore, the 
Commission believes that the backfit statement for the 1991 amendments, 
which determined that the backfits were necessary to ensure that the 
facility provides adequate protection to the public health and safety, 
are 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), since there is no change in 
requirements due to this restructuring.
    The inclusion of thermal annealing does not impose any backfits as 
defined in 10 CFR 50.109(a)(1), for the reasons set forth below in 
``Thermal Annealing Rule (10 CFR 50.66).''

Thermal Annealing Rule (10 CFR 50.66)

    The proposed thermal annealing rule would establish new 
requirements with respect to applications for thermal annealing. 
However, the Commission has determined that the proposed rule would not 
impose a ``backfit'' as defined in 10 CFR 50.109(a)(1). The proposed 
thermal annealing rule would 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 10 CFR Part 
50, Appendix G. However, these rules provide an alternative means to 
meet these screening limits, viz., 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. In addition, the proposed rule does not 
reflect any new or different Staff position which conflicts with a 
prior Staff position or Commission rule. Thus, the proposed rule would 
have a purely prospective effect on future applications for thermal 
annealing. The Commission has stated in other rulemakings establishing 
prospective requirements, e.g., 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 
when there are no prior NRC positions upon which the ``substantial 
increase in overall protection'' can be measured. Accordingly, the 
Commission concludes that the proposed rule does not impose backfits 
and a backfit analysis need not be prepared for the proposed thermal 
annealing rule.

10 CFR Part 50 Appendix G

    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 of a 
clarifying nature.
    The explicit prohibition on core criticality before the completion 
of pressure and leak tests can be construed as a backfit, although NRC 
staff intent was never to permit such a procedure (letter from J. M. 
Taylor, NRC, to N. S. Reynolds and D. F. Stenger, NUBARG, dated 
February 2, 1990). 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). This amendment underscores the prior intent of the 
Commission to prohibit the use of nuclear heat before the completion of 
leak and pressure tests that is implicit in 10 CFR 50.55a and Section 
XI of the ASME Code. The Commission's intent in this regard is 
demonstrated by the fact that only a very small minority of licensees 
actually used nuclear heat to conduct pressure and leak tests required 
by the ASME Code.

10 CFR Part 50 Appendix H

    The amendments to Appendix H of 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).

Criminal Penalties

    For purposes of Section 223 of the Atomic Energy Act (AEA), the 
Commission proposes to issue the proposed rule under one or more of 
Sections 161b, 161i or 161o of the AEA. Willful violations of the rule 
would be subject to criminal enforcement.

List of Subjects

10 CFR Part 50

    Antitrust, Classified information, Criminal penalties, Fire 
protection, Incorporation by reference, Intergovernmental relations, 
Nuclear power plants and reactors, Radiation protection, Reactor siting 
criteria, Reporting and recordkeeping 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. 553; the NRC is proposing to 
adopt the following amendments to 10 CFR Part 50.

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    1. The authority citation for Part 50 is revised 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. 1244, 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. 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 of this part.
    (2) Pressurized Thermal Shock Event means an event or transient in 
pressurized water reactors (PWRs) causing severe overcooling (thermal 
shock) concurrent with or followed by significant pressure in the 
reactor vessel.
    (3) Reactor Vessel Beltline means the region of the reactor vessel 
(shell material including welds, heat affected zones and plates or 
forgings) that directly surrounds the effective height of the active 
core and adjacent regions of the reactor vessel that are predicted to 
experience sufficient neutron radiation damage to be considered in the 
selection of the most limiting material with regard to radiation 
damage.
    (4) RTNDT means the reference temperature for a reactor vessel 
material, under any conditions. For the reactor vessel beltline 
materials, RTNDT must account for the effects of neutron 
radiation.
    (5) RTNDT(U) means the reference temperature for a reactor 
vessel material in the pre-service or unirradiated condition, evaluated 
according to the procedures in the ASME Code, Paragraph NB-2331.
    (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, the proposed expiration date if a change in the term 
of the operating license has been requested, or the end of a renewal 
term if an application for a renewed license under 10 CFR Part 54 has 
been submitted.
    (7) RTPTS means the reference temperature, RTNDT, 
evaluated for the EOL Fluence for each of the vessel beltline 
materials, using the procedures of paragraph (c) of this section.
    (8) PTS Screening Criterion means the value of RTPTS for the 
vessel beltline material above which the plant cannot continue to 
operate without justification.
    (b) Requirements.
    (1) For each pressurized water nuclear power reactor for which an 
operating license has been issued, the licensee shall have projected 
values of RTPTS, accepted by the NRC, for each reactor vessel 
beltline material for the EOL fluence of the material. The assessment 
of RTPTS must use the calculative procedures given in paragraph 
(c)(1) of this section, except as provided in paragraphs (c)(2) and 
(c)(3) of this section. The assessment must specify the bases for the 
projected value of RTPTS for each vessel beltline material, 
including the assumptions regarding core loading patterns, and must 
specify the copper and nickel contents and the fluence value used in 
the calculation for each beltline material. This assessment must be 
updated whenever there is a significant\1\ change in projected values 
of RTPTS, or upon a request for a change in the expiration date 
for operation of the facility.
---------------------------------------------------------------------------

    \1\Changes to RTPTS values are considered significant if 
either the previous value or the current value, or both values, 
exceed the screening criterion prior to the expiration of the 
operating license, including any renewed term, if applicable, for 
the plant.
---------------------------------------------------------------------------

    (2) The pressurized thermal shock (PTS) screening criterion is 
270 deg.F for plates, forgings, and axial weld materials, and 300 deg.F 
for circumferential weld materials. For the purpose of comparison with 
this criterion, the value of RTPTS for the reactor vessel must be 
evaluated according to the procedures of paragraph (c) of this section, 
for each weld and plate, or forging, in the reactor vessel beltline. 
RTPTS must be determined for each vessel beltline material using 
the EOL fluence for that material.
    (3) For each pressurized water nuclear power reactor for which the 
value of RTPTS for any material in the beltline is projected to 
exceed the PTS screening criterion using the EOL fluence, the licensee 
shall implement those flux reduction programs that are reasonably 
practicable to avoid exceeding the PTS screening criterion set forth in 
paragraph (b)(2) of this section. The schedule for implementation of 
flux reduction measures may take into account the schedule for 
submittal and anticipated approval by the Director, Office of Nuclear 
Reactor Regulation, of detailed plant-specific analyses, submitted to 
demonstrate acceptable risk with RTPTS above the screening limit 
due to plant modifications, new information or new analysis techniques.
    (4) For each pressurized water nuclear power reactor for which the 
analysis required by paragraph (b)(3) of this section indicates that no 
reasonably practicable flux reduction program will prevent RTPTS 
from exceeding the PTS screening criterion using the EOL fluence, the 
licensee shall submit a safety analysis to determine what, if any, 
modifications to equipment, systems, and operation are necessary to 
prevent potential failure of the reactor vessel as a result of 
postulated PTS events if continued operation beyond the screening 
criterion is allowed. In the analysis, the licensee may determine the 
properties of the reactor vessel materials based on available 
information, research results, and plant surveillance data, and may use 
probabilistic fracture mechanics techniques. This analysis must be 
submitted at least three years before RTPTS is projected to exceed 
the PTS screening criterion.
    (5) After consideration of the licensee's analyses, including 
effects of proposed corrective actions, if any, submitted in accordance 
with paragraphs (b)(3) and (b)(4) of this section, the Director, Office 
of Nuclear Reactor Regulation, may, on a case-by-case basis, approve 
operation of the facility with RTPTS in excess of the PTS 
screening criterion. The Director, Office of Nuclear Reactor 
Regulation, will consider factors significantly affecting the potential 
for failure of the reactor vessel in reaching a decision.
    (6) If the Director, Office of Nuclear Reactor Regulation, 
concludes, pursuant to paragraph (b)(5) of this section, that operation 
of the facility with RTPTS in excess of the PTS screening 
criterion cannot be approved on the basis of the licensee's analyses 
submitted in accordance with paragraphs (b)(3) and (b)(4) of this 
section, the licensee shall request and receive approval by the 
Director, Office of Nuclear Reactor Regulation, prior to any operation 
beyond the criterion. The request must be based upon modifications to 
equipment, systems, and operation of the facility in addition to those 
previously proposed in the submitted analyses that would reduce the 
potential for failure of the reactor vessel due to PTS events, or upon 
further analyses based upon new information or improved methodology.
    (7) If the limiting RTPTS value of the plant is projected to 
exceed the screening criteria in paragraph (b)(2), or the criteria in 
paragraphs (b)(3) through (b)(6) of this section cannot be satisfied, 
the reactor vessel beltline may be given a thermal annealing treatment 
to recover the fracture toughness of the material, subject to the 
requirements of Sec. 50.66. The reactor vessel may continue to be 
operated only for that service period within which the predicted 
fracture toughness of the vessel beltline materials satisfy the 
requirements of paragraphs (b)(2) through (b)(6) of this section, with 
RTPTS accounting for the effects of annealing and subsequent 
irradiation.
    (c) Calculation of RTPTS. RTPTS must be evaluated using 
the same procedures used to calculate RTNDT, as indicated in 
paragraph (c)(1) of this section, and as provided in paragraphs (c)(2) 
and (c)(3). RTPTS must be calculated for each vessel beltline 
material using a fluence value, f, which is the EOL fluence for the 
material.
    (1) Equation 1 must be used to calculate values of RTNDT for 
each weld and plate, or forging, in the reactor vessel beltline.

TP04OC94.000

    (i) RTNDT(U) is the reference temperature, RTNDT, of the 
material in the pre-service or unirradiated condition, evaluated 
according to the procedures in the ASME Code, Paragraph NB-2331.
    (A) If a measured value of RTNDT(U) is not available, a 
generic mean value for the class\2\ of material may be used if there 
are sufficient test results to establish a mean and a standard 
deviation for the class.
---------------------------------------------------------------------------

    \2\The class of material for estimating RTNDT(U) is 
generally determined for welds by the type of welding flux (Linde 
80, or other), and for base metal by the material specification.
---------------------------------------------------------------------------

    (B) For weld metals, 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.
    (ii) M means the margin to be added to account for uncertainties in 
the values of RTNDT(U), copper and nickel contents, fluence and 
the calculational procedures. M is evaluated from Equation 2.

TP04OC94.001

    (A) In Equation 2, U is the standard deviation for 
RTNDT(U). If a measured value of RTNDT(U) is used, then 
U is determined from the precision of the test method. If 
a measured value of RTNDT(U) is not available and a generic mean 
value for that class of materials is used, then U is the 
standard deviation obtained from the set of data used to establish the 
mean. If a generic mean value given in paragraph (c)(1)(i)(B) for welds 
is used, then U is 17 deg.F.
    (B) In Equation 2,  is the standard deviation 
for RTNDT. The value of  to be used 
is 28 deg.F for welds and 17 deg.F for base metal; the value of 
 shall not exceed one-half of RTNDT.
    (iii) RTNDT is the mean value of the transition 
temperature shift, or change in RTNDT, due to irradiation, and 
must be calculated using Equation 3.

TP04OC94.002

    (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\3\ may be used if justification is provided. If none of 
these alternatives are available, 0.35% copper and 1.0% nickel must be 
assumed.
---------------------------------------------------------------------------

    \3\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\19\ n/
cm\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 paragraph (c), the EOL fluence for the vessel 
beltline material is used in calculating RTPTS.
    (2) To verify that RTNDT for each vessel beltline material is 
a bounding value for the specific reactor vessel, licensees shall 
consider plant-specific information that could affect the level of 
embrittlement. This information includes but is not limited to the 
reactor vessel operating temperature and surveillance program results.
    (i) Results from the plant-specific surveillance program must be 
integrated into the RTNDT estimate if the plant-specific 
surveillance data has been deemed credible as judged by the following 
criteria:
    (A) The materials in the surveillance capsules must be those which 
are the controlling materials with regard to radiation embrittlement.
    (B) Scatter in the plots of Charpy energy versus temperature for 
the irradiated and unirradiated conditions must be small enough to 
permit the determination of the 30-foot-pound temperature 
unambiguously.
    (C) Where there are two or more sets of surveillance data from one 
reactor, the scatter of RTNDT values must be less than 
28 deg.F for welds and 17 deg.F for base metal. Even if the range in 
the capsule fluences is large (two or more orders of magnitude), the 
scatter may not exceed twice those values.
    (D) The irradiation temperature of the Charpy specimens in the 
capsule must equal the vessel wall temperature at the cladding/base 
metal interface within 25 deg.F.
    (E) The surveillance data for the correlation monitor material in 
the capsule must fall within the scatter band of the data base for the 
material.
    (ii) Surveillance data deemed credible according to the criteria of 
paragraph (c)(2)(i) 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 4.

TP04OC94.003

In Equation 4, ``n'' is the number of surveillance data points, 
``Ai'' is the measured value of RTNDT and 
``fi'' is the fluence for each surveillance data point.
    (iii) For cases in which the results from a credible plant-specific 
surveillance program are used, the value of  to be 
used is 14 deg.F for welds and 8.5 deg.F for base metal; the value of 
 may not exceed one-half of RTNDT.
    (iv) The use of results from the plant-specific surveillance 
program may result in an RTNDT that is higher or lower than those 
determined in paragraph (c)(1).
    (3) Any information that is believed to improve the accuracy of the 
RTPTS value significantly must be reported to the Director, Office 
of Nuclear Reactor Regulation. Any value of RTPTS that has been 
modified using the procedures of paragraph (c)(2) 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
.02..........            21             26            27            27            27            27            27
.03..........            22             35            41            41            41            41            41
.04..........            24             43            54            54            54            54            54
.05..........            26             49            67            68            68            68            68
.06..........            29             52            77            82            82            82            82
.07..........            32             55            85            95            95            95            95
.08..........            36             58            90           106           108           108           108
.09..........            40             61            94           115           122           122           122
.10..........            44             65            97           122           133           135           135
.11..........            49             68           101           130           144           148           148
.12..........            52             72           103           135           153           161           161
.13..........            58             76           106           139           162           172           176
.14..........            61             79           109           142           168           182           188
.15..........            66             84           112           146           175           191           200
.16..........            70             88           115           149           178           199           211
.17..........            75             92           119           151           184           207           221
.18..........            79             95           122           154           187           214           230
.19..........            83            100           126           157           191           220           238
.20..........            88            104           129           160           194           223           245
.21..........            92            108           133           164           197           229           252
.22..........            97            112           137           167           200           232           257
.23..........           101            117           140           169           203           236           263
.24..........           105            121           144           173           206           239           268
.25..........           110            126           148           176           209           243           272
.26..........           113            130           151           180           212           246           276
.27..........           119            134           155           184           216           249           280
.28..........           122            138           160           187           218           251           284
.29..........           128            142           164           191           222           254           287
.30..........           131            146           167           194           225           257           290
.31..........           136            151           172           198           228           260           293
.32..........           140            155           175           202           231           263           296
.33..........           144            160           180           205           234           266           299
.34..........           149            164           184           209           238           269           302
.35..........           153            168           187           212           241           272           305
.36..........           158            172           191           216           245           275           308
.37..........           162            177           196           220           248           278           311
.38..........           166            182           200           223           250           281           314
.39..........           171            185           203           227           254           285           317
.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
.02..........            20             20            20            20            20            20            20
.03..........            20             20            20            20            20            20            20
.04..........            22             26            26            26            26            26            26
.05..........            25             31            31            31            31            31            31
.06..........            28             37            37            37            37            37            37
.07..........            31             43            44            44            44            44            44
.08..........            34             48            51            51            51            51            51
.09..........            37             53            58            58            58            58            58
.10..........            41             58            65            65            67            67            67
.11..........            45             62            72            74            77            77            77
.12..........            49             67            79            83            86            86            86
.13..........            53             71            85            91            96            96            96
.14..........            57             75            91           100           105           106           106
.15..........            61             80            99           110           115           117           117
.16..........            65             84           104           118           123           125           125
.17..........            69             88           110           127           132           135           135
.18..........            73             92           115           134           141           144           144
.19..........            78             97           120           142           150           154           154
.20..........            82            102           125           149           159           164           165
.21..........            86            107           129           155           167           172           174
.22..........            91            112           134           161           176           181           184
.23..........            95            117           138           167           184           190           194
.24..........           100            121           143           172           191           199           204
.25..........           104            126           148           176           199           208           214
.26..........           109            130           151           180           205           216           221
.27..........           114            134           155           184           211           225           230
.28..........           119            138           160           187           216           233           239
.29..........           124            142           164           191           221           241           248
.30..........           129            146           167           194           225           249           257
.31..........           134            151           172           198           228           255           266
.32..........           139            155           175           202           231           260           274
.33..........           144            160           180           205           234           264           282
.34..........           149            164           184           209           238           268           290
.35..........           153            168           187           212           241           272           298
.36..........           158            173           191           216           245           275           303
.37..........           162            177           196           220           248           278           308
.38..........           166            182           200           223           250           281           313
.39..........           171            185           203           227           254           285           317
.40..........           175            189           207           231           257           288           320
----------------------------------------------------------------------------------------------------------------

    3. A new Sec. 50.66 is added 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 treatment may be applied to the reactor 
vessel to restore the fracture toughness to acceptable levels. The use 
of a thermal annealing treatment is subject to the approval of the 
Director, Office of Nuclear Reactor Regulation, and to the requirements 
in this section. The application for the Director's approval must be 
submitted in accordance with Sec. 50.4, and at least three years prior 
to the proposed date of the annealing operation.
     (b) Thermal Annealing Application. The content of the application 
for approval by the Director, Office of Nuclear Reactor Regulation, for 
thermal annealing of the reactor vessel must include: a thermal 
annealing operating plan that includes an evaluation of the effects of 
mechanical and thermal stresses and temperatures, an inspection and 
test program to requalify the annealed reactor vessel, and a program 
for demonstrating that the recovery of fracture toughness and the re-
embrittlement rate are adequate to permit subsequent safe operation of 
the reactor vessel for the period specified in the application.
    (1) Thermal Annealing Operating Plan.
    (i) The thermal annealing operating plan must include:
    (A) A detailed description of the pressure vessel and all 
structures and components that will be affected by the thermal 
annealing operation;
    (B) The methods, instrumentation and procedures proposed for 
performing the thermal annealing;
    (C) A description of the heat source to be used; and
    (D) The proposed thermal annealing operating parameters, including 
temperatures, times, and heatup and cooldown schedules.
    (ii) The annealing time and temperature parameters selected must be 
based on projecting sufficient recovery of fracture toughness, using 
the procedures of paragraph (d) 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. In addition, the operating plan 
must describe 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.
    (iii) An evaluation of the effects of mechanical and thermal 
stresses and temperatures on the vessel, attached piping and 
appurtenances, and adjacent equipment and components must demonstrate 
that operability of the reactor will not be detrimentally affected. A 
detailed thermal and structural analysis must be performed 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. The effects of 
localized high temperatures must be evaluated for degradation of the 
concrete adjacent to the vessel and changes in thermal and mechanical 
properties of the reactor vessel insulation. If the design temperature 
limitations for the adjacent concrete structure are to be exceeded 
during the annealing operation, an acceptable maximum temperature for 
the concrete must be established for the annealing operation using 
appropriate test data.
    (iv) The time and temperature profile evaluated as part of the 
annealing operating plan, and supported by the evaluation results of 
paragraph (b)(1)(iii) of this section, represents the proposed 
annealing conditions that may not be exceeded during the annealing 
operation. If these conditions are exceeded, then the licensee cannot 
certify that the annealing operation was performed in accordance with 
the approved application, as required by paragraph (c)(1) of this 
section, and must comply with paragraph (c)(2) of this section.
    (v) The projected percent recovery of both RTNDT and Charpy 
upper-shelf energy must be determined by the procedures described in 
paragraph (d) of this section, using the proposed annealing time and 
temperature described in the operating plan. The projected post-anneal 
RTNDT and Charpy upper-shelf energy must be determined from the 
projected percent recovery.
    (vi) The projected rate of reembrittlement of RTNDT must be 
calculated using the procedures in Sec. 50.61(c), or must be the same 
rate as that used for the pre-anneal operating period. The projected 
rate of reembrittlement for Charpy upper-shelf energy must be the same 
rate as that used for the pre-anneal operating period.
    (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 annealing conditions of 
paragraph (b)(1)(iv) of this section have not been exceeded, and to 
determine the annealing time and temperature to be used in quantifying 
the fracture toughness recovery. In addition, the program must 
demonstrate that the 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 Rate Assurance 
Program. The percent recovery of RTNDT and Charpy upper-shelf 
energy obtained by the thermal annealing treatment must be determined 
from the time and temperature of the actual vessel annealing. The 
recovery of RTNDT and Charpy upper-shelf energy provide the basis 
for establishing the post-anneal RTNDT and Charpy upper-shelf 
energy for each vessel material. Changes in the RTNDT and Charpy 
upper-shelf energy with subsequent plant operation must be determined 
using the post-anneal values of these parameters in conjunction with 
the projected reembrittlement rate determined in accordance with 
paragraph (b)(3)(ii) of this section.
    (i) The recovery of RTNDT and Charpy upper-shelf energy must 
be established using the procedures in paragraph (d) of this section, 
using the time and temperature of the actual vessel annealing.
    (A) 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 annealing parameters 
used in the test program are equal to or bounded by those used in the 
vessel annealing operation.
    (B) If generic computational methods are used, appropriate 
justification must be submitted as a part of the application.
    (ii) The reembrittlement rate of both RTNDT and Charpy upper-
shelf energy must be estimated, and must be monitored using a 
surveillance program which conforms to Appendix H of this part, 
``Reactor Vessel Material Surveil- lance Program Requirements.''
    (c) Certification of the Annealing Effectiveness.
    (1) Upon completion of the anneal and prior to re-start of the 
nuclear power plant, the licensee shall certify to the Director, Office 
of Nuclear Reactor Regulation, that the thermal annealing was performed 
in accordance with the approved application required by paragraph (a) 
of this section, and meets the provisions of paragraph (b) of this 
section. In this certification, the licensee shall establish the period 
for which the reactor vessel will satisfy the requirements of 
Sec. 50.60 and Sec. 50.61, and shall provide:
    (i) The post-anneal RTNDT and Charpy upper-shelf energy values 
of the reactor vessel materials for use in subsequent reactor 
operation;
    (ii) The projected reembrittlement trends for both RTNDT and 
Charpy upper-shelf energy; and
    (iii) The projected values of RTPTS and Charpy upper-shelf 
energy at the end of the proposed period of operation addressed in the 
application.
    (2) If the licensee cannot certify that the thermal annealing was 
performed in accordance with the approved application and the 
provisions of paragraph (b) of this section, the licensee shall submit 
a justification for subsequent operation for approval by the Director, 
Office of Nuclear Reactor Regulation.
    (d) Procedures for Determining the Recovery of Fracture Toughness. 
The procedures of this paragraph must be used to determine the percent 
recovery of NDT, Rt, and percent 
recovery of Charpy upper-shelf energy, Ru. In all cases, Rt 
and Ru may not exceed 100.
    (1) For those reactors with surveillance programs which have 
developed credible surveillance data as defined in Sec. 50.61, percent 
recovery due to annealing (Rt and Ru) must be evaluated by 
testing surveillance specimens that have been withdrawn from the 
surveillance program and that have been annealed under the same time 
and temperature conditions as those given the beltline material.
    (2) Alternatively, the percent recovery due to annealing (Rt 
and Ru) may be determined from the results of a verification test 
program employing materials removed from the beltline region of the 
reactor vessel\1\ and that have been annealed under the same time and 
temperature conditions as those given the beltline material.
---------------------------------------------------------------------------

    \1\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.
    4. In 10 CFR Part 50, Appendix G is revised to read as follows:

Appendix G to Part 50--Fracture Toughness Requirements

Table of Contents

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 St., New 
York, NY 10017 and are available for inspection at the NRC Library, 
11545 Rockville Pike, Two White Flint North, Rockville, Maryland 20852-
2738.
    The requirements of this appendix apply to the following materials:
    A. Carbon and low-alloy ferritic steel plate, forgings, castings, 
and pipe with specified minimum yield strengths not over 50,000 psi 
(345 MPa), and to those with specified minimum yield strengths greater 
than 50,000 psi (345 MPa) but not over 90,000 psi (621 MPa) if 
qualified by using methods equivalent to those described in paragraph 
G-2110 of Appendix G of Section XI of the latest edition and addenda of 
the ASME Code incorporated by reference into Sec. 50.55a(b)(2).
    B. Welds and weld heat-affected zones in the materials specified in 
paragraph I.A. of this appendix.
    C. Materials for bolting and other types of fasteners with 
specified minimum yield strengths not over 130,000 psi (896 MPa).

    Note: The adequacy of the fracture toughness of other ferritic 
materials not covered in this section must be demonstrated to the 
Director, Office of Nuclear Reactor Regulation, on an individual 
case basis.

II. Definitions

    A. Ferritic material means carbon and low-alloy steels, higher 
alloy steels including all stainless alloys of the 4xx series, and 
maraging and precipitation hardening steels with a predominantly body-
centered cubic crystal structure.
    B. System hydrostatic tests means all preoperational system leakage 
and hydrostatic pressure tests and all system leakage and hydrostatic 
pressure tests performed during the service life of the pressure 
boundary in compliance with the ASME Code, Section XI.
    C. Specified minimum yield strength means the minimum yield 
strength (in the unirradiated condition) of a material specified in the 
construction code under which the component is built under Sec. 50.55a.
    D. RTNDT means the reference temperature of the material, for 
all conditions.
    (i) For the pre-service or unirradiated condition, RTNDT is 
evaluated according to the procedures in the ASME Code, Paragraph NB-
2331.
    (ii) For the reactor vessel beltline materials, RTNDT must 
account for the effects of neutron radiation.
    E. RTNDT means the transition temperature shift, or 
change in RTNDT, due to neutron radiation effects, which is 
evaluated as the difference in the 30 ft-lb (41 J) index temperatures 
from the average Charpy curves measured before and after irradiation.
    F. Beltline or Beltline region of reactor vessel means the region 
of the reactor vessel (shell material including welds, heat affected 
zones, and plates or forgings) that directly surrounds the effective 
height of the active core and adjacent regions of the reactor vessel 
that are predicted to experience sufficient neutron radiation damage to 
be considered in the selection of the most limiting material with 
regard to radiation damage.

III. Fracture Toughness Tests

    A. To demonstrate compliance with the fracture toughness 
requirements of Section IV of this appendix, ferritic materials must be 
tested in accordance with the ASME Code and, for the beltline 
materials, the test requirements of Appendix H of this part. For a 
reactor vessel that was constructed to an ASME Code earlier than the 
Summer 1972 Addenda of the 1971 Edition (under Sec. 50.55a), the 
fracture toughness data and data analyses must be supplemented in a 
manner approved by the Director, Office of Nuclear Reactor Regulation, 
to demonstrate equivalence with the fracture toughness requirements of 
this appendix.
    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

    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 normal operation, including anticipated operational 
occurrences. Reactor vessels may continue to be operated only for that 
service period within which the requirements of this section are 
satisfied. For the reactor vessel beltline materials, including welds, 
plates and forgings, the values of RTNDT and Charpy upper-shelf 
energy must account for the effects of neutron radiation, including the 
results of the surveillance program of Appendix H of this part. The 
effects of neutron radiation must consider the radiation conditions 
(i.e., the fluence) at the deepest point on the crack front of the flaw 
assumed in the analysis.
1. Reactor Vessel Charpy Upper-Shelf Energy Requirements
    a. Reactor vessel beltline materials must have Charpy upper-shelf 
energy,\1\ in the transverse direction for base material and along the 
weld for weld material according to the ASME Code, of no less than 75 
ft-lb (102 J) initially and must maintain Charpy upper-shelf energy 
throughout the life of the vessel of no less than 50 ft-lb (68 J), 
unless it is demonstrated in a manner approved by the Director, Office 
of Nuclear Reactor Regulation, that lower values of Charpy upper-shelf 
energy will provide margins of safety against fracture equivalent to 
those required by Appendix G of Section XI of the ASME Code. This 
analysis must use the latest edition and addenda of the ASME Code 
incorporated by reference into Sec. 50.55a(b)(2) at the time the 
analysis is submitted.
---------------------------------------------------------------------------

    \1\Defined in ASTM E 185-79 and -82 which are incorporated by 
reference in Appendix H to Part 50.
---------------------------------------------------------------------------

    b. Additional evidence of the fracture toughness of the beltline 
materials after exposure to neutron irradiation may be obtained from 
results of supplemental fracture toughness tests, for use in the 
analysis specified in section IV.A.1.a.
    c. The analysis for satisfying the requirements of section IV.A.1 
of this appendix must be submitted, as specified in Sec. 50.4, for 
review and approval on an individual case basis at least three years 
prior to the date when the predicted Charpy upper-shelf energy will no 
longer satisfy the requirements of section IV.A.1 of this appendix, or 
on a schedule approved by the Director, Office of Nuclear Reactor 
Regulation.
2. Pressure-Temperature Limits and Minimum Temperature Requirements
    a. Pressure-temperature limits and minimum temperature requirements 
for the reactor vessel are given in Table 1, 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 or 
not the core is critical. In Table 1, 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 1 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 1 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 1, the minimum temperature 
requirements and the controlling material depend on the operating 
condition (i.e., hydrostatic pressure and leak tests, or normal 
operation including anticipated operational occurrences), the vessel 
pressure, whether fuel is in the vessel, and whether the core is 
critical. The metal temperature of the controlling material, in the 
region of the controlling material which has the least favorable 
combination of stress and temperature, must exceed the appropriate 
minimum temperature requirement for the condition and pressure of the 
vessel specified in Table 1.
    d. Pressure tests and leak tests of the reactor vessel that are 
required by Section XI of the ASME Code must be completed before the 
core is critical.
    B. If the procedures of Section IV.A. of this appendix do not 
indicate the existence of an equivalent safety margin, the reactor 
vessel beltline may be given a thermal annealing treatment to recover 
the fracture toughness of the material, subject to the requirements of 
Sec. 50.66. The reactor vessel may continue to be operated only for 
that service period within which the predicted fracture toughness of 
the beltline region materials satisfies the requirements of Section 
IV.A. of this appendix using the values of RTNDT and Charpy upper-shelf 
energy that include the effects of annealing and subsequent 
irradiation.

                                Table 1.--Pressure and Temperature Requirements                                 
----------------------------------------------------------------------------------------------------------------
                                        Vessel                                                                  
       Operating condition          pressure(\1\)      Requirements for pressure-         Minimum temperature   
                                                          temperature limits                 requirements       
----------------------------------------------------------------------------------------------------------------
1. Hydrostatic pressure and leak tests (core is not critical):                                                  
    1.a Fuel in the vessel........  20  ASME Appendix G Limits...........  (2) + 60  deg.F.          
                                               %                                                                
    1.b Fuel in the vessel........          >20%   ASME Appendix G Limits...........  (2)+90  deg.F.            
    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.........  20  ASME Appendix G Limits...........  (2).                      
                                               %                                                                
    2.b Core not critical.........          >20%   ASME Appendix G Limits...........  (2)+120  deg.F (6).       
    2.c Core critical.............  20  ASME Appendix G Limits+40  deg.F.  Larger of [(4)] or [(2)+40
                                               %                                        deg.F].                 
    2.d Core critical.............          >20%   ASME Appendix G Limits+40  deg.F.  Larger of [(4)] or        
                                                                                       [(2)+160  deg.F].        
    2.e Core critical for BWR       20  ASME Appendix G Limits+40  deg.F.  (2)+60  deg.F.            
     (\5\).                                    %                                                                
----------------------------------------------------------------------------------------------------------------
(\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.                  

    5. In 10 CFR Part 50, Appendix H is revised to read as follows:

Appendix H to Part 50--Reactor Vessel Material Surveillance Program 
Requirements

Table of Contents

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 this part.
    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 St., 
Philadelphia, PA 19103 and are available for inspection at the NRC 
Library, 11545 Rockville Pike, Two White Flint North, Rockville, 
Maryland 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\17\ n/cm\2\ (E >1 MeV).
    B. Reactor vessels that do not meet the conditions of paragraph 
II.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 ASTM E 185-73 or the edition of 
ASTM E 185 that is current on the issue date of the ASME Code to which 
the reactor vessel was purchased, whichever is later. 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 26th day of September 1994.

    For the Nuclear Regulatory Commission.
John C. Hoyle,
Acting Secretary of the Commission.
[FR Doc. 94-24209 Filed 10-3-94; 8:45 am]
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