[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]
BILLING CODE 7590-01-P
