[Federal Register Volume 60, Number 138 (Wednesday, July 19, 1995)]
[Rules and Regulations]
[Pages 36952-36959]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-17723]
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[[Page 36953]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 50
RIN 3150-AF06
Technical Specifications
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations pertaining to technical specifications for nuclear power
reactors. The rule codifies criteria for determining the content of
technical specifications. Each licensee covered by these regulations
may voluntarily use the criteria as a basis to propose the relocation
of existing technical specifications that do not meet any of the
criteria from the facility license to licensee-controlled documents.
The voluntary conversion of current technical specifications in this
manner is expected to produce an improvement in the safety of nuclear
power plants through a reduction in unnecessary plant transients and
more efficient use of NRC and industry resources.
EFFECTIVE DATE: August 18, 1995.
FOR FURTHER INFORMATION CONTACT: Christopher I. Grimes, Chief,
Technical Specifications Branch, Division of Project Support, Office of
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Telephone: (301) 415-1161.
SUPPLEMENTARY INFORMATION:
Background
Section 182a. of the Atomic Energy Act of 1954 (Act), as amended
(42 U.S.C. 2232), mandates the inclusion of technical specifications in
licenses for the operation of production and utilization facilities.
The Act requires that technical specifications include information
concerning the amount, kind, and source of special nuclear material;
the place of use; and the specific characteristics of the facility.
That section also states that technical specifications shall contain
information the Commission requires through regulation to enable it to
find that the utilization of special nuclear material will be in accord
with the common defense and security and will provide adequate
protection of public health and safety. Finally, that section requires
technical specifications to be made a part of any license issued.
The Commission promulgated Sec. 50.36, ``Technical
Specifications,'' which implements section 182a. of the Atomic Energy
Act on December 17, 1968 (33 FR 18610). This rule delineates
requirements for determining the contents of technical specifications.
Technical specifications, at a minimum, must set forth the specific
characteristics of the facility and the conditions for its operation
that are required to provide adequate protection of the health and
safety of the public. Specifically, Sec. 50.36 requires the following:
Each license authorizing operation of a production or
utilization facility of a type described in Sec. 50.21 or Sec. 50.22
will include technical specifications. The technical specifications
will be derived from the analyses and evaluation included in the
safety analysis report, and amendments thereto, submitted pursuant
to Sec. 50.34. The Commission may include such additional technical
specifications as the Commission finds appropriate.
Technical specifications cannot be changed by licensees without
prior NRC approval. However, since 1969, there has been a trend toward
including in technical specifications not only those requirements
derived from the analyses and evaluation in the safety analysis report
but also essentially all other Commission requirements governing the
operation of nuclear power reactors. This extensive use of technical
specifications was due in part to a lack of well-defined criteria (in
either the body of the rule or in some other regulatory document) for
what should be included in technical specifications. Since 1969, this
use has contributed to the volume of technical specifications and to
the several-fold increase in the number of license amendment
applications to effect changes to the technical specifications. It has
diverted both NRC staff and licensee attention from the more important
requirements in these documents to the extent that it has resulted in
an adverse but unquantifiable impact on safety.
On March 30, 1982 (47 FR 13369), the NRC published in the Federal
Register a proposed amendment to Part 50. The proposed rule would have
revised Sec. 50.36, ``Technical Specifications,'' to establish a new
system of specifications divided into two general categories. Only
those specifications contained in the first general category as
technical specifications would have become part of the operating
license and would have required prior NRC approval for any changes.
Those specifications contained in the second general category would
have become supplemental specifications and would not have required
prior NRC approval for most changes. The NRC review of the first
general category of specifications would have been the same as that
currently performed for technical specification changes, which are
amendments to the operating license. For the second category,
``supplemental specifications,'' the licensee would have been allowed
to make changes within specified conditions without prior NRC approval.
The NRC would have reviewed these changes when they were made and would
have done so in a manner similar to that currently used for reviewing
design changes, tests, and experiments performed under the provisions
of Sec. 50.59. Because of difficulties with defining the criteria for
dividing the technical specifications into the two categories of the
proposed rule and because of other higher priority licensing work, the
proposed amendment was deferred.
In the early 1980s, the nuclear industry and the NRC staff began
studying whether the existing system of establishing technical
specification requirements for nuclear power plants needed improvement.
During this period, an NRC task group known as the Technical
Specifications Improvement Project (TSIP) and a Subcommittee of the
Atomic Industrial Forum's (AIF's) Committee on Reactor Licensing and
Safety performed two studies of this issue.\1\ The overall conclusion
of these studies was that many improvements in the scope and content of
technical specifications were needed and that a joint NRC and industry
program should be initiated to implement these improvements. Both
groups made specific recommendations; these are summarized as follows:
\1\ SECY-86-10, ``Recommendations for Improving Technical
Specifications,'' January 13, 1986, contains both ``Recommendations
for Improving Technical Specifications,'' NRC Technical
Specifications Improvement Project, September 30, 1985, and
``Technical Specifications Improvements,'' AIF Subcommittee on
Technical Specifications Improvements, October 1, 1985.
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(1) The NRC should adopt the criteria for defining the scope of
technical specifications proposed in the AIF and TSIP reports. Those
criteria should then be used by the NRC and each of the nuclear steam
supply system vendor owners groups to completely rewrite and streamline
the existing standard technical specifications (STS). This process
would result in the transfer of many requirements from control by
technical specification requirements to control by other mechanisms
(e.g., the final safety analysis report (FSAR), operating procedures,
quality assurance (QA) plan) that would not require a license amendment
or prior NRC approval when changes were needed.
[[Page 36954]]
The new STS should place greater emphasis on human factors principles
in order to make the text of the STS clearer and easier to understand.
The new STS should also improve the bases section of technical
specifications, which gives the purpose for each requirement in the
specification.
(2) A parallel program of short-term improvements in both the scope
and substance of the existing technical specifications should be
initiated in addition to developing new STS as stated in recommendation
1.
On February 6, 1987 (52 FR 3788), the NRC published in the Federal
Register for public comment an ``Interim Policy Statement on Technical
Specification Improvements for Nuclear Power Reactors'' (interim policy
statement) containing proposed criteria in response to recommendation
1. These criteria were generally derived from the criteria proposed in
the AIF and TSIP reports and were modified slightly on the basis of
discussions between the NRC staff and the industry. The public comment
period for the interim policy statement expired on March 23, 1987.
The criteria were developed with the intention that they would
apply to limiting conditions for operation (LCOs). The NRC staff
believed that the safety limits needed to remain unchanged in the
technical specifications because of their more direct link to
protection of the physical barriers that guard against the uncontrolled
release of radioactivity. At the time the criteria were developed, the
industry did not wish to address administrative controls and design
features in the effort to improve the STS. Later, however, both the
industry and the NRC staff realized that it would be beneficial to
include upgraded administrative controls and design features in the
improved STS, and these were handled separately from the application of
the criteria to the LCOs.
The NRC has developed a program for short-term improvements as
described in recommendation 2 (above). These are known as ``line-item''
improvements and are generic improvements developed and promulgated by
the NRC staff for voluntary adoption by licensees.
Subsequently, improved vendor-specific STS were developed and
issued by the NRC in September 1992. The improved STS were published as
the following NRC reports:
NUREG-1430, ``Standard Technical Specifications, Babcock
and Wilcox Plants''
NUREG-1431, ``Standard Technical Specifications,
Westinghouse Plants''
NUREG-1432, ``Standard Technical Specifications,
Combustion Engineering Plants''
NUREG-1433, ``Standard Technical Specifications, General
Electric Plants, BWR/4''
NUREG-1434, ``Standard Technical Specifications, General
Electric Plants, BWR/6''
Copies of these NUREGs, as revised, may be purchased from the
Superintendent of Documents, U.S. Government Printing Office, by
calling (202) 275-2060 or by writing to the Superintendent of
Documents, U.S. Government Printing Office, PO Box 37082, Washington,
DC 20013-7082. Copies are also available from the National Technical
Information Service, 5825 Port Royal Road, Springfield, VA 22161.
These improved STS were the result of extensive technical meetings
and discussions among the NRC staff, industry owners groups, vendors,
and the Nuclear Management and Resources Council (NUMARC).
On July 22, 1993 (58 FR 39132), the Commission published a ``Final
Policy Statement on Technical Specifications Improvements for Nuclear
Power Reactors'' (final policy statement), which incorporated
experience and lessons learned since publication of the interim policy
statement. The Commission has decided not to withdraw the final policy
statement because it contains detailed discussions of the four criteria
and guidance on how the NRC staff and licensees should apply the
criteria.
The interim policy statement identified three criteria to be used
to define which of the current technical specification requirements
should be retained or included in technical specifications and which
LCOs could be relocated to licensee-controlled documents, as follows:
Criterion 1: Installed instrumentation that is used to detect, and
indicate in the control room, a significant abnormal degradation of the
reactor coolant pressure boundary.
Criterion 2: A process variable, design feature, or operating
restriction that is an initial condition of a design basis accident or
transient analysis that either assumes the failure of or presents a
challenge to the integrity of a fission product barrier.
Criterion 3: A structure, system, or component that is part of the
primary success path and which functions or actuates to mitigate a
design basis accident or transient that either assumes the failure of
or presents a challenge to the integrity of a fission product barrier.
The interim policy statement also stated that, in addition to
structures, systems, and components captured by the three criteria, it
was the Commission's policy that licensees retain in the technical
specifications LCOs for a specified list of systems that operating
experience and probabilistic risk assessment (PRA) had generally shown
to be important to public health and safety. In the final policy
statement, the Commission retained this thought as a fourth criterion
as follows:
Criterion 4: A structure, system, or component which operating
experience or probabilistic risk assessment has shown to be significant
to public health and safety.
As stated in the final policy statement, if a requirement meets any
one of the four criteria, it should be retained or included in
technical specifications.
The final policy statement also addressed comments received on the
interim policy statement and described the Commission's intent with
regard to use of the criteria and their codification through
rulemaking.
This final rule codifies the four criteria contained in the final
policy statement for defining the scope of LCOs in technical
specifications. These criteria are intended to be consistent with the
scope of technical specifications as stated in the Statement of
Consideration for the final rule issuing Sec. 50.36 (33 FR 18610,
December 17, 1968). The Statement of Consideration discussed the scope
of technical specifications as including the following:
In the revised system, emphasis is placed on two general classes
of technical matters: (1) Those related to prevention of accidents,
and (2) those related to mitigation of the consequences of
accidents. By systematic analysis and evaluation of a particular
facility, each applicant is required to identify at the construction
permit stage those items that are directly related to maintaining
the integrity of the physical barriers designed to contain
radioactivity. Such items are expected to be the subjects of
Technical Specifications in the operating license.
The first of these two general classes of technical matters to be
included in technical specifications is captured by Criteria 1, 4, and,
to some extent, Criterion 2, in that they address systems and process
variables that alert the operator to a situation when accident
initiation is more likely. The second general class of technical
matters is explicitly addressed and captured by Criteria 2, 3, and 4.
By applying the four criteria contained in this rule, a licensee should
capture the conditions for operation of its facility that are required
[[Page 36955]]
to meet the principal operative standard in Section 182a. of the Atomic
Energy Act, that is, that adequate protection is provided to the health
and safety of the public.
The Commission recognizes that the four criteria carry a theme of
focusing on the technical requirements for features of controlling
importance to safety. Since many of the requirements are of
significance to the health and safety of the public, this rule reflects
the subjective statement of the purpose of technical specifications
expressed by the Atomic Safety and Licensing Appeal Board in Portland
General Electric Company (Trojan Nuclear Plant), ALAB-531, 9 NRC 263
(1979). There, the Appeal Board interpreted technical specifications as
being reserved for those conditions or limitations upon reactor
operation necessary to obviate the possibility of an abnormal situation
or event giving rise to an immediate threat to the public health and
safety.
The Commission wishes to emphasize that this rule is intended to be
consistent with the language of section 182a. of the Atomic Energy Act,
the current Sec. 50.36 rule, and previous interpretations of the
regulations. This rule merely clarifies the scope and purpose of
technical specifications by identifying criteria which can be used to
establish, more clearly, the framework for LCOs in technical
specifications.
The Commission believes that amending Sec. 50.36 to include the
four criteria contained in the final policy statement will codify a
viable, potentially safety-enhancing and cost-saving method for
technical specification improvement. The Commission continues to
encourage licensees to use the improved STS as the basis for plant-
specific technical specifications. As stated in the final policy
statement, the Commission will place the highest priority on requests
based on the criteria for individual license amendments that are used
to evaluate all of the LCOs for an individual plant to determine which
LCOs should be included in the technical specifications. Related
surveillance requirements and actions would be retained for each LCO
that remains in the technical specifications. Each LCO, action, and
surveillance requirement should have supporting bases. Such requests
would constitute complete conversions to the improved STS.
In addition, the Commission will also entertain requests to adopt
portions of the improved STS, even if the licensee does not adopt all
STS improvements. These portions will include all related requirements
and will be developed as line-item improvements by the NRC staff when
they are clearly generic in nature, when there is evidence that a
significant number of licensees could benefit from the improvement, and
when the industry expresses interest in the improvement. The Commission
encourages all licensees who submit technical specification related
submittals based on these criteria to emphasize human factors
principles to the extent practical consistent with the format and
content of their current technical specifications.
LCOs that do not meet any of the criteria, and their associated
actions and surveillance requirements, may be proposed for relocation
from the technical specifications to licensee-controlled documents,
such as the FSAR. The criteria may be applied to either standard or
custom technical specifications. The Commission will also consider the
criteria in evaluating future generic requirements for inclusion in
technical specifications.
The Commission expects that licensees, in preparing their technical
specification submittals, will utilize any plant-specific PRA or risk
survey and any available literature on risk insights and PRAs. This
material should be employed to strengthen the technical bases for those
provisions that remain in technical specifications, when applicable,
and to indicate whether the provisions to be relocated contain
constraints of importance in limiting the likelihood or severity of the
accident sequences that are commonly found to dominate risk. Similarly,
the NRC staff has and will continue to employ risk insights in
evaluating technical specifications submittals.
In addition to the use of PRA in Criterion 4 to determine the scope
of technical specifications, PRA has been used as a basis for a number
of improvements to the content of technical specifications over the
last several years. The NRC staff has approved several relaxations in
technical specification allowed outage times and surveillance test
intervals which were based on PRA. In addition, the NRC staff used PRA
to develop screening criteria to evaluate all of the changes in allowed
outage times and surveillance test intervals that were made during the
development of the improved STS. The industry and the NRC staff have
used PRA to an even greater extent in the development and review of the
technical specifications for advanced reactor designs.
The industry and the NRC staff are currently exploring several new
approaches to utilizing PRA for technical specification improvements
including the use of on-line risk assessment tools. In addition, the
industry and the NRC staff are using PRA to explore further
improvements in technical specifications by examining the risks during
shutdown and during the transition between modes of operation. As a
part of this ongoing program of improving technical specifications, the
Commission will continue to consider methods to make better use of risk
and reliability information for defining future generic technical
specification requirements.
During technical specification conversions, the staff will apply
the backfit rule (Sec. 50.109) when adding new requirements from the
improved STS to individual plant technical specifications, provided the
licensee does not voluntarily accept the new requirements. If, however,
the staff suggested additional changes are needed to make the licensee
requested changes acceptable from the standpoint of adequate protection
or compliance with NRC regulations, Sec. 50.109(a)(2) and
Sec. 50.109(a)(3) do not apply and the request may be denied without
the additional items.
Summary of Public Comments
The Commission received three letters commenting on the proposed
rule. Each letter contained several comments.
One commenter representing the commercial nuclear industry
expressed concern that there is insufficient regulatory guidance on how
the NRC staff intends to implement this rule with respect to the fourth
criterion (Sec. 50.36(c)(2)(ii)(D)). The commenter believes that this
rule should not be modified until the NRC and the industry have reached
a common understanding of the application, threshold, and intent of
Criterion 4. The commenter stated, ``It is our view, and the Commission
apparently recognizes, that this criterion goes beyond the adequate
protection standard for public health and safety and license compliance
purposes embodied in the first three criteria.''
Similar to this comment on the proposed rule, the Advisory
Committee on Reactor Safeguards (ACRS) commented in a June 18, 1993,
letter to the Chairman that the NRC staff needs to provide more
detailed guidance on the definition of ``significant to public health
and safety,'' as it is used in Criterion 4.
Criterion 4 is intended to capture those constraints that
probabilistic risk assessment or operating experience show to be
significant to public health and safety, consistent with the
Commission's PRA Policies. The level of significance either would need
to be
[[Page 36956]]
such that it justified including the constraints in the technical
specifications to ensure adequate protection of the public health and
safety or that the addition of such constraints provides substantial
additional protection to the public health and safety.
The Commission identified four systems that meet Criterion 4 in the
final policy statement based on previous qualitative reviews of
operating experience and risk. They are reactor core isolation cooling/
isolation condenser, residual heat removal, standby liquid control, and
recirculation pump trip. The Commission recognizes, however, that other
structures, systems, or components may meet this criterion. Plant- and
design-specific PRAs have yielded valuable insight to unique plant
vulnerabilities not fully recognized in the safety, design basis
accident, or transient analyses.
The NRC's current regulatory requirements are largely based on
deterministic engineering criteria involving the use of multiple
barriers and defense in depth. Recently, the NRC staff has formulated a
comprehensive plan for the application of PRA technology and insights
throughout the agency. It is expected that the PRA Implementation Plan
will serve as the framework for continued and future applications of
PRA at the NRC. Implementation of this plan will increase the
systematic use of risk assessment techniques. To ensure consistent and
appropriate decision-making that incorporates PRA methods and results,
it is important that coherent and clear application guidelines are
applied. As part of the PRA Implementation Plan, such guidelines will
be established (incorporating safety goals and backfit rule
considerations) that address the interdependence of probabilistic risk
and deterministic engineering principles. The process of developing
these guidelines will involve communications among the NRC staff, the
nuclear industry, and the public to ensure that all parties understand
the role of PRA methods and results in NRC's risk management efforts.
The NRC staff anticipates that, as it gains experience with the
development and use of such PRA application guidelines, it will be
better able to refine such phrases as ``significant to public health
and safety,'' and other phrases that are used in many of the
Commission's regulations.
The Commission could delay publication of this final rule until the
PRA application guidelines are in place. However, the Commission
believes that the experience gained while using the criteria under the
interim and final policy statements combined with the limitations
imposed on the NRC staff by the backfit rule provide assurance that, in
the interim, the staff's use of Criterion 4 to apply PRA to technical
specification content will be properly controlled. The Commission has
concluded that it is appropriate to publish this final rule, which
provides the framework for technical specifications, at this time.
One commenter stated that the PRA portion of the fourth criterion
should be clarified to include only those equipment items important to
risk-significant sequences as defined in Generic Letter 88-20,
``Individual Plant Examination for Severe Accident Vulnerabilities,''
Appendix 2, and reported in licensees' individual plant examination
(IPE) reports.
The IPE program has resulted in commercial reactor licensees using
risk-assessment methods to identify plant-specific severe accident
vulnerabilities. Since submittal of their IPE reports, many licensees
have enhanced their plant-specific PRAs and have gained additional
insights into unique plant vulnerabilities. These additional insights
from PRAs are being used by licensees in such areas as implementation
of the maintenance rule.
As stated in the Commission's ``Proposed Policy Statement on the
Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory
Activities,'' the use of PRA technology should be increased in all
regulatory matters to the extent supported by the state of the art in
PRA methods and data and in a manner that complements the NRC's
deterministic approach and supports the NRC's traditional defense-in-
depth philosophy. The Commission will continue to apply PRA to
technical specifications in accordance with its proposed policy
statement on the use of PRA. In addition, guidance for specific
applications or classes of applications will be developed under the PRA
Implementation Plan. The Commission believes this is a more appropriate
means to define how Criterion 4 will be used in practice, rather than
to limit the structures, systems, and components captured by Criterion
4 to those items important to risk-significant sequences as defined in
Generic Letter 88-20, Appendix 2, and reported in licensees' IPE
reports. The Commission believes that this process will provide the NRC
staff and the industry with additional risk insights, beyond those
identified through the IPE program.
The same commenter said that the operating experience portion of
the fourth criterion should be deleted because it is subjective and
because no equipment would satisfy only that portion of the fourth
criterion and none of the other criteria.
While operating experience is an important part of PRA, not all PRA
models are sophisticated enough to capture all operating experience.
The Commission believes that operating experience can play an important
role in determining the safety significance of structures, systems, and
components and that there will be no adverse impact by including
operating experience as part of Criterion 4.
One commenter emphasized that the development of implementation
guidance, especially with respect to Criterion 4, should be consistent
with the implementation guidance of the maintenance rule.
As stated previously, the Commission believes that the improved
STS, the final policy statement, the backfit rule (Sec. 50.109), and
the statement of consideration for this rule contain sufficient
guidance on implementation of the criteria to proceed with rulemaking.
Supplementary guidance will continue to be provided to the NRC staff
that will support the process for implementing the four criteria on
both a generic and plant-specific basis, and will be publicly
available. The NRC staff will ensure that any guidance documents that
relate to the implementation of the four criteria will be consistent
with the implementation guidance of the maintenance rule along with the
guidance for other rules promulgated by the Commission.
One commenter expressed a concern with respect to the level of PRA
information necessary to support the relocation of existing technical
specifications which do not meet the first three criteria.
If a technical specification provision does not meet any of the
first three criteria, and if the current PRA knowledge or operating
experience does not identify the structure, system, or component as
risk significant, the NRC staff will not preclude relocating such
technical specifications. The level of PRA information necessary to
support relocation would be considered as part of the overall review of
the supporting basis for the proposed change. The Commission expects
that licensees will utilize PRA insights to indicate whether the
provisions to be relocated contain constraints of importance in
limiting the likelihood or severity of the accident sequences that are
commonly found to dominate risk.
One commenter stated that the implementing guidance needs to be
[[Page 36957]]
clear on how the proposed criteria would be used to determine if new
requirements are to be incorporated into technical specifications.
The Commission believes that the improved STS, the final policy
statement, the backfit rule (Sec. 50.109), and the statement of
consideration for this rule contain sufficient guidance on
implementation of the criteria. The staff will also ensure that
application of the criteria to new requirements is consistent with the
guidance in the draft ``Regulatory Analysis Guidelines,'' Revision 2,
published in August 1993 (NUREG/BR-0058), and the final version of
Revision 2 when it is approved by the Commission. In addition, the NRC
has recently published NUREG/CR-6141, ``Handbook of Methods for Risk-
Based Analyses of Technical Specifications,'' December 1994, which
summarizes systematic risk-based methods to improve various aspects of
technical specification requirements. The handbook was developed
through research sponsored by the NRC and will be used as a reference
document to assist the NRC staff in reviewing licensees' risk-based
analyses submitted as part of the bases for proposed changes in
facility technical specifications. This guidance will be updated
periodically to incorporate lessons learned and changes in the state of
the art, will help ensure the criteria are applied in a consistent and
controlled manner, and will be publicly available. As stated above, as
part of the PRA Implementation Plan, PRA application guidelines will be
established (incorporating safety goals and backfit rule
considerations) that address the interdependence of probabilistic risk
and deterministic engineering principles. As these application
guidelines develop, they will progressively be used to provide guidance
to the NRC staff on the use of the criteria contained in this rule and
the application of the backfit rule to new regulatory requirements.
One commenter stated that the same or similar criteria to those in
the rule should also be applied to 10 CFR 50.36(c)(3), (4), and (5), so
that surveillance requirements, design features, and administrative
controls which do not provide the necessary ``adequate protection of
the health and safety of the public'' can be relocated to other
licensee-controlled documents.
With respect to Sec. 50.36 (c)(3), ``Surveillance Requirements,''
the Commission stated in the final policy statement that appropriate
surveillance requirements and actions should be retained for each LCO
which remains or is included in the technical specifications.
The criteria in Sec. 50.36(c)(2) apply to safety functions.
Therefore, the Commission does not believe that these criteria can be
appropriately applied to the types of requirements found in the
``design features'' and ``administrative controls'' sections of the
technical specifications. The NRC staff has, however, been pursuing
separate improvements to these requirements, in cooperation with
industry, using the intent of the criteria to identify the optimum set
of requirements in each of these areas and to eliminate redundancy to
other regulations consistent with the minimum requirements of
Sec. 50.36 and the Atomic Energy Act, as amended.
One commenter stated that the removal of items from plant technical
specifications may decrease enforceability and licensee attention to
safety.
The Commission does not agree that the removal of items from plant
technical specifications will decrease licensee attention to safety. On
the contrary, the Commission believes that implementation of the
criteria contained in this rule will produce an improvement in the
safety of nuclear power plants through the use of more operator-
oriented technical specifications, improved technical specification
bases, reduced action statement induced plant transients, and more
efficient use of NRC and industry resources. Clarification of the scope
and purpose of technical specifications has provided useful guidance to
both the NRC and industry and has resulted in improved technical
specifications that are intended to focus licensee and plant operator
attention on those plant conditions most important to safety.
The Commission also does not agree that the removal of items from
plant technical specifications will have any adverse impact on the
NRC's ability to take enforcement action on safety-significant issues.
The improved STS are intended specifically to focus on the operating
plant parameters and associated surveillance criteria of safety
significance. The Commission requires compliance with technical
specifications, and expects adherence to commitments contained in
licensee-controlled documents. Violations and deviations will, as in
the past, be handled in accordance with the NRC enforcement policy in
10 CFR Part 2, Appendix C. Any changes to a licensee's technical
specifications to apply these criteria will be made by the license
amendment process prior to implementation.
When a licensee elects to apply these criteria, some requirements
are relocated from technical specifications to the FSAR or to other
licensee- controlled documents. Licensees are to operate their
facilities in conformance with the descriptions of their facilities and
procedures in their FSAR. Changes to the facility or to procedures
described in the FSAR are to be made in accordance with 10 CFR 50.59.
The Commission will take appropriate enforcement action to ensure that
licensees comply with 10 CFR 50.59. Changes made in accordance with the
provisions of other licensee-controlled documents (e.g., QA plan,
security plan) are subject to the specific requirements for those
documents. Nothing in this rule limits the authority of the NRC to
conduct necessary inspections and to take appropriate enforcement
action when regulatory requirements or commitments are not met.
The same commenter stated that the removal of items from plant
technical specifications will diminish public participation rights in
the regulation of operating nuclear power plants by diminishing the
universe of potential operating license amendment cases.
Any changes to a licensee's technical specifications to apply these
criteria will be made by the license amendment process before
implementation. The review of each license amendment will involve an
opportunity for public participation. One of the goals of the technical
specifications improvement program was to make more efficient use of
NRC and industry resources by focusing attention on those plant
conditions most important to safety and, in turn, reducing the number
of license amendment requests. Since 1969, there has been a trend
toward including in technical specifications not only those
requirements derived from the analyses and evaluations included in the
safety analysis report but also essentially all other Commission
requirements governing the operation of nuclear power reactors. This
extensive use of technical specifications is due in part to a lack of
well-defined criteria (in either the body of the rule or in some other
regulatory document) for what should be included in technical
specifications. This has contributed to the volume of technical
specifications and to the several-fold increase, since 1969, in the
number of license amendment applications to effect changes to the
technical specifications. It has diverted both NRC staff and licensee
attention from the more important requirements in these documents to
the extent that it has resulted in an adverse but unquantifiable impact
on safety.
[[Page 36958]]
The commenter found it curious that an industry and an agency that
claim to be able to quantify the risks of nuclear power are unable to
quantify this impact on safety, and stated, ``Perhaps if it is
unquantifiable, the alleged adverse impact does not really exist.''
The Commission agrees that there are limitations and uncertainties
in the ability to quantify the impact on safety described above.
Uncertainties exist in any regulatory approach and these uncertainties
are derived from knowledge limitations. A probabilistic approach has
exposed some of these limitations and yielded an improved framework to
better focus and assess their significance and assist in developing a
strategy to accommodate them in the regulatory process. The Commission
does not intend, however, to let these limitations prevent it from
taking steps to improve the regulations in a manner that will have
substantial safety benefits. The Commission believes the public will be
better served by focusing both NRC and industry attention on the most
safety-significant items.
The NRC staff has made three changes to this rule since it was
published in its proposed form. The first change was made in order to
maintain consistency with other NRC staff and Commission documents that
have been issued since this rule was published in its proposed form. In
Sec. 50.36(c)(2)(ii)(D), the term ``probabilistic safety assessment''
has been changed to ``probabilistic risk assessment.''
The second and third changes are in Sec. 50.36(c)(2)(iii). The
beginning of the first sentence was changed to read, ``A licensee is
not required to propose to modify technical specifications * * *''
rather than ``A licensee is not required to modify technical
specifications * * *'' This change was made to clarify that a licensee
would be required to modify their technical specifications if the
Commission determined that a new requirement was necessary in
accordance with the backfit rule and the new requirement met one of the
four criteria contained in Sec. 50.36(c)(2)(ii).
The third change is the deletion of the last sentence in
Sec. 50.36(c)(2)(iii). The sentence read, ``However, for technical
specification amendments a licensee proposes after August 18, 1995, the
criteria in paragraph (c)(2)(ii) of this section provide an acceptable
scope for limiting conditions for operation.'' This sentence was
deleted because it did not add or modify any requirements and the
thought is adequately expressed in this statement of consideration.
Finding of No Significant Environmental Impact: Availability
The Commission has determined under the National Environmental
Policy Act of 1969, as amended, and the Commission regulations in
Subpart A of Part 51, that this final rule is not a major Federal
action significantly affecting the quality of the human environment and
will not degrade the environment in any way. Therefore, the Commission
concludes that there will be no significant impact on the environment
from this rule. This discussion constitutes the environmental
assessment and finding of no significant impact for this rule; a
separate assessment has not been prepared.
Paperwork Reduction Act Statement
This final rule does not contain a new or amended information
collection requirement subject to the Paperwork Reduction Act of 1980
(44 U.S.C. 3501 et seq.). Existing requirements were approved by the
Office of Management and Budget, approval number 3150-0011.
Regulatory Analysis
The Commission has determined that a regulatory analysis is not
required for this rule. The Commission believes that the intent of the
regulatory analysis has been met through the extensive consideration
given to the development of the ``Final Policy Statement on Technical
Specifications Improvements for Nuclear Power Reactors'' and the
improved STS, both of which gave the public an opportunity for comment.
In addition, the determination that no regulatory analysis is necessary
was noted in the Federal Register Notice for the proposed rule, and the
NRC received no comments on this issue.
The criteria being added to Sec. 50.36 are the same as those
contained in the final policy statement and have been used by the NRC
and the nuclear power industry to define the content of technical
specifications since September 1992. The rule does not impose any
requirements but, rather, allows nuclear power reactor licensees to
voluntarily use the criteria to relocate existing technical
specifications that do not meet any of the criteria to licensee-
controlled documents. The NRC staff also uses these criteria to
determine whether technical specifications are appropriate to provide
regulatory control over new requirements or positions that have been
justified consistent with the backfit rule.
The Commission considered the need for and consequences of this
action when it made the decision not only to publish the criteria in
the final policy statement but also to codify the criteria through
rulemaking. Appropriate alternative approaches to this action have been
identified and analyzed over the life of the Technical Specifications
Improvement Program, beginning with an earlier attempt to define the
content of technical specifications through rulemaking. As described in
the background discussion, the Commission published a proposed
amendment to Sec. 50.36 (47 FR 13369) on March 30, 1982. However,
because of difficulties with defining criteria for technical
specifications and because of other higher priority licensing work, the
rule change was deferred. In February 1987, the Commission published an
``Interim Policy Statement on Technical Specification Improvements for
Nuclear Power Reactors,'' and in July 1993, published the final policy
statement. During its review of the final policy statement, the
Commission concluded that the four criteria should be codified in a
rule. Thus, alternative approaches to regulatory objectives have been
identified and analyzed, and the Commission has decided that there is
no preferable alternative to codifying the four criteria in a rule.
With regard to evaluation of values and impacts of alternatives, the
Commission believes there is no difference in the values or impacts of
applying the criteria under the final policy statement or through a
rule, except that the criteria are more readily available to future
users in a rule rather than in a policy statement.
Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C.
605(b)), the Commission certifies that this final rule does not have a
significant economic impact on a substantial number of small entities.
This rule affects only the licensing and operation of nuclear power
plants. The companies that own these plants do not fall within the
scope of the definition of ``small entities'' as given 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
The NRC has determined that the backfit rule, Sec. 50.109, does not
apply to this final rule and, therefore, a backfit analysis is not
required for this final rule because these amendments do not involve
any provisions that would impose backfits as defined in
Sec. 50.109(a)(1).
[[Page 36959]]
List of Subjects in 10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Fire
protection, Intergovernmental relations, Nuclear power plants and
reactors, Radiation protection, Reactor siting criteria, Reporting and
recordkeeping requirements.
For the reasons given in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended, the Energy Reorganization
Act of 1974, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting
the following amendment to Part 50.
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for Part 50 continues 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, 50.54(dd),
and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42
U.S.C. 2138). Sections 50.23. 50.35, 50.55, and 50.56 also issued
under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,
50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-190, 83
Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued
under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58-
50.91, and 50.92 also issued under Pub. L. 97-415, 96 Stat. 2073 (42
U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939
(42 U.S.C. 2152). Sections 50.80-50.81 also issued under sec. 184,
68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued
under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
2. In Sec. 50.36, paragraphs (c)(2) and (3) are revised to read as
follows:
Sec. 50.36 Technical specifications.
* * * * *
(c) * * *
(2) Limiting conditions for operation. (i) Limiting conditions for
operation are the lowest functional capability or performance levels of
equipment required for safe operation of the facility. When a limiting
condition for operation of a nuclear reactor is not met, the licensee
shall shut down the reactor or follow any remedial action permitted by
the technical specifications until the condition can be met. When a
limiting condition for operation of any process step in the system of a
fuel reprocessing plant is not met, the licensee shall shut down that
part of the operation or follow any remedial action permitted by the
technical specifications until the condition can be met. In the case of
a nuclear reactor not licensed under Sec. 50.21(b) or Sec. 50.22 of
this part or fuel reprocessing plant, the licensee shall notify the
Commission, review the matter, and record the results of the review,
including the cause of the condition and the basis for corrective
action taken to preclude recurrence. The licensee shall retain the
record of the results of each review until the Commission terminates
the license for the nuclear reactor or the fuel reprocessing plant. In
the case of nuclear power reactors licensed under Sec. 50.21(b) or
Sec. 50.22, the licensee shall notify the Commission if required by
Sec. 50.72 and shall submit a Licensee Event Report to the Commission
as required by Sec. 50.73. In this case, licensees shall retain records
associated with preparation of a Licensee Event Report for a period of
three years following issuance of the report. For events which do not
require a Licensee Event Report, the licensee shall retain each record
as required by the technical specifications.
(ii) A technical specification limiting condition for operation of
a nuclear reactor must be established for each item meeting one or more
of the following criteria:
(A) Criterion 1. Installed instrumentation that is used to detect,
and indicate in the control room, a significant abnormal degradation of
the reactor coolant pressure boundary.
(B) Criterion 2. A process variable, design feature, or operating
restriction that is an initial condition of a design basis accident or
transient analysis that either assumes the failure of or presents a
challenge to the integrity of a fission product barrier.
(C) Criterion 3. A structure, system, or component that is part of
the primary success path and which functions or actuates to mitigate a
design basis accident or transient that either assumes the failure of
or presents a challenge to the integrity of a fission product barrier.
(D) Criterion 4. A structure, system, or component which operating
experience or probabilistic risk assessment has shown to be significant
to public health and safety.
(iii) A licensee is not required to propose to modify technical
specifications that are included in any license issued before August
18, 1995, to satisfy the criteria in paragraph (c)(2)(ii) of this
section.
(3) Surveillance requirements. Surveillance requirements are
requirements relating to test, calibration, or inspection to assure
that the necessary quality of systems and components is maintained,
that facility operation will be within safety limits, and that the
limiting conditions for operation will be met.
* * * * *
Dated at Rockville, Maryland, this 13th day of July 1995.
For the Nuclear Regulatory Commission.
John C. Hoyle,
Secretary of the Commission
[FR Doc. 95-17723 Filed 7-18-95; 8:45 am]
BILLING CODE 7590-01-P
