[Federal Register Volume 64, Number 246 (Thursday, December 23, 1999)]
[Rules and Regulations]
[Pages 71990-72002]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-33283]
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NUCLEAR REGULATORY COMMISSION
10 CFR Parts 21, 50, and 54
RIN 3150-AG12
Use of Alternative Source Terms at Operating Reactors
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations to allow holders of operating licenses for nuclear power
plants to voluntarily replace the traditional source term used in
design basis accident analyses with alternative source terms. This
action will allow interested licensees to pursue cost beneficial
licensing actions to reduce unnecessary regulatory burden without
compromising the margin of safety of the facility. The NRC is
announcing the availability of a draft regulatory guide and a draft
Standard Review Plan section on this subject for public comment. The
NRC is also amending its regulations to revise certain sections to
conform with the final rule published on December 11, 1996, concerning
reactor site criteria.
EFFECTIVE DATE: January 24, 2000.
FOR FURTHER INFORMATION CONTACT: Mr. Stephen F. LaVie, Office of
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; telephone: (301) 415-1081; or by Internet
electronic mail to sfl@nrc.gov.
SUPPLEMENTARY INFORMATION:
I. Background
II. Analysis of Public Comments
III. Section-by-Section Analysis
IV. Draft Regulatory Guide; Issuance, Availability
V. Draft Standard Review Plan Section; Issuance, Availability
VI. Referenced Documents
VII. Finding of No Significant Environmental Impact; Availability
VIII. Paperwork Reduction Act Statement
IX. Regulatory Analysis
X. Regulatory Flexibility Act Certification
XI. Backfit Analysis
XII. Small Business Regulatory Enforcement Fairness Act
XIII. National Technology Transfer and Advancement Act
I. Background
A holder of an operating license (i.e., the licensee) for a light-
water power reactor is required by regulations issued by the NRC (or
its predecessor, the U.S. Atomic Energy Commission, (AEC)) to submit a
safety analysis report (or, for early reactors, a hazard summary
report) that contains assessments of the radiological consequences of
potential accidents and an evaluation of the proposed facility site.
The NRC uses this information in its evaluation of the suitability of
the reactor design and the proposed site as required by its regulations
contained in 10 CFR Parts 50 and 100. Section 100.11, which was adopted
by the AEC in 1962 (27 FR 3509; April 12, 1962), requires an applicant
to assume (1) a fission product release from the reactor core, (2) the
expected containment leak rate, and (3) the site meteorological
conditions to establish an exclusion area and a low population zone.
This fission product release is based on a major accident that would
result in substantial release of appreciable quantities of fission
products from the core to the containment atmosphere. A note to
Sec. 100.11 states that Technical Information Document (TID) 14844,
``Calculation of Distance Factors for
[[Page 71991]]
Power and Test Reactors,'' may be used as a source of guidance in
developing the exclusion area, the low population zone, and the
population center distance. Changes to the design of the facility and
the procedures for operating the facility are evaluated in part by
determining whether there are changes to the calculated fission product
release.
The fission product release from the reactor core into containment
is referred to as the ``source term'' and it is characterized by the
composition and magnitude of the radioactive material, the chemical and
physical properties of the material, and the timing of the release from
the reactor core. The accident source term is used to evaluate the
radiological consequences of design basis accidents (DBAs) in showing
compliance with various requirements of the NRC's regulations. Although
originally used for site suitability analyses, the accident source term
is a design parameter for accident mitigation features, equipment
qualification, control room operator radiation doses, and post-accident
vital area access doses. The measurement range and alarm setpoints of
some installed plant instrumentation and the actuation of some plant
safety features are based in part on the accident source term. The TID-
14844 source term was explicitly stated as a required design parameter
for several Three Mile Island (TMI)-related requirements.
The NRC's methods for calculating accident doses, as described in
Regulatory Guide 1.3, ``Assumptions Used for Evaluating the Potential
Radiological Consequences of a Loss of Coolant Accident for Boiling
Water Reactors''; Regulatory Guide 1.4, ``Assumptions Used for
Evaluating the Potential Radiological Consequences of a Loss of Coolant
Accident for Pressurized Water Reactors''; and NUREG-0800, ``Standard
Review Plan for the Review of Safety Analysis Reports for Nuclear Power
Plants,'' were developed to be consistent with the TID-14844 source
term and the whole body and thyroid dose guidelines stated in
Sec. 100.11. In this regulatory framework, the source term is assumed
to be released immediately to the containment at the start of the
postulated accident. The chemical form of the radioiodine released to
the containment atmosphere is assumed to be predominantly elemental,
with the remainder being small fractions of particulate and organic
iodine forms. Radiation doses are calculated at the exclusion area
boundary (EAB) for the first 2 hours and at the low population zone
(LPZ) for the assumed 30-day duration of the accident. The whole body
dose comes primarily from the noble gases in the source term. The
thyroid dose is based on inhalation of radioiodines. In analyses
performed to date, the thyroid dose has generally been limiting. The
design of some engineered safety features, such as containment spray
systems and the charcoal filters in the containment, the building
exhaust, and the control room ventilation systems, are predicated on
these postulated thyroid doses. Subsequently, the NRC adopted the whole
body and thyroid dose criteria in Criterion 19 of 10 CFR Part 50,
Appendix A (36 FR 3255; February 20, 1971).
The source term in TID-14844 is representative of a major accident
involving significant core damage and is typically postulated to occur
in conjunction with a large loss-of-coolant accident (LOCA). Although
the LOCA is typically the maximum credible accident, NRC experience in
reviewing license applications has indicated the need to consider other
accident sequences of lesser consequence but higher probability of
occurrence. Some of these additional accident analyses may involve
source terms that are a fraction of those specified in TID-14844. The
DBAs were not intended to be actual event sequences but, rather, were
intended to be surrogates to enable deterministic evaluation of the
response of the plant engineered safety features. These accident
analyses are intentionally conservative in order to address
uncertainties in accident progression, fission product transport, and
atmospheric dispersion. Although probabilistic risk assessments (PRAs)
can provide useful insights into system performance and suggest changes
in how the desired defense in depth is achieved, defense in depth
continues to be an effective way to account for uncertainties in
equipment and human performance. The NRC's policy statement on the use
of PRA methods (60 FR 42622; August 16, 1995) calls for the use of PRA
technology in all regulatory matters in a manner that complements the
NRC's deterministic approach and supports the traditional defense-in-
depth philosophy.
Since the publication of TID-14844, significant advances have been
made in understanding the timing, magnitude, and chemical form of
fission product releases from severe nuclear power plant accidents.
Many of these insights developed out of the major research efforts
started by the NRC and the nuclear industry after the accident at Three
Mile Island (TMI). In 1995, the NRC published NUREG-1465, ``Accident
Source Terms for Light-Water Nuclear Power Plants,'' which utilized
this research to provide more physically based estimates of the
accident source term that could be applied to the design of future
light-water power reactors. The NRC sponsored significant review
efforts by peer reviewers, foreign research partners, industry groups,
and the general public (request for public comment was published in 57
FR 33374; July 28, 1992).
The information in NUREG-1465 presents a representative accident
source term (``revised source term'') for a boiling-water reactor (BWR)
and for a pressurized-water reactor (PWR). These revised source terms
are described in terms of radionuclide composition and magnitude,
physical and chemical form, and timing of release. Where TID-14844
addressed three categories of radionuclides, the revised source terms
categorize the accident release into eight groups on the basis of
similarity in chemical behavior. Where TID-14844 assumed an immediate
release of the activity, the revised source terms have five release
phases that are postulated to occur over several hours, with the onset
of major core damage occurring after 30 minutes. Where TID-14844
assumed radioiodine to be predominantly elemental, the revised source
terms assume radioiodine to be predominantly cesium iodide (CsI), an
aerosol that is more amenable to mitigation mechanisms.
For DBAs, the NUREG-1465 source terms (up to and including the
early in-vessel phase) are comparable to the TID-14844 source term with
regard to the magnitude of the noble gas and radioiodine release
fractions. However, the revised source terms offer a more
representative description of the radionuclide composition and release
timing. The NRC has determined (SECY-94-302, December 19, 1994) that
design basis analyses will address the first three release phases--
coolant, gap, and in-vessel. The ex-vessel and late in-vessel phases
are considered to be inappropriate for design basis analysis purposes.
These latter releases could only result from core damage accidents with
vessel failure and core-concrete interactions.
The objective of NUREG-1465 was to define revised accident source
terms for regulatory application for future light water reactors
(LWRs). The NRC's intent was to capture the major relevant insights
available from severe accident research to provide, for regulatory
purposes, a more realistic portrayal of the amount of the postulated
accident source term. These source terms were derived from examining a
set of severe accident sequences for LWRs of current
[[Page 71992]]
design. Because of general similarities in plant and core design
parameters, these results are considered to be applicable to
evolutionary and passive LWR designs. The revised source term has been
used in evaluating the Westinghouse AP600 standard design certification
application. (A draft version of NUREG-1465 was used in evaluating
Combustion Engineering's (CE's) System 80+ design.)
The NRC considered the applicability of the revised source terms to
operating reactors and determined that the current analytical approach
based on the TID-14844 source term would continue to be adequate to
protect public health and safety, and that operating reactors licensed
under this approach would not be required to reanalyze accidents using
the revised source terms. The NRC concluded that some licensees may
wish to use an alternative source term in analyses to support
operational flexibility and cost-beneficial licensing actions and that
some of these applications could provide concomitant improvements in
overall safety and in reduced occupational exposure. The NRC initiated
several actions to provide a regulatory basis for operating reactors to
voluntarily amend their facility design bases to enable use of the
revised source term in design basis analyses. First, the NRC solicited
ideas on how an alternative source term might be implemented. In
November 1995, the Nuclear Energy Institute (NEI) submitted its generic
framework, Electric Power Research Institute Technical Report TR-
105909, ``Generic Framework for Application of Revised Accident Source
Term to Operating Plants.'' This report and the NRC response were
discussed in SECY-96-242 (November 25, 1996). Second, the NRC initiated
an assessment of the overall impact of substituting the NUREG-1465
source terms for the traditionally used TID-14844 source term at three
typical facilities. This was done to evaluate the issues involved with
applying the revised source terms at operating plants. SECY-98-154
(June 30, 1998) described the conclusions of this assessment. Third,
the NRC accepted license amendment requests related to implementation
of the revised source terms at a small number of pilot plants.
Experience has demonstrated that evaluation of a limited number of
plant-specific submittals improves regulation and regulatory guidance
development. The review of these pilot projects is currently in
progress. Insights from these pilot plant reviews have been
incorporated into the regulatory guidance that was developed in
conjunction with this rulemaking. Fourth, the NRC initiated an
assessment on whether rulemaking would be necessary to allow operating
reactors to use an alternative source term. This final rule and the
supporting regulatory guidance have resulted from this assessment.
This final rulemaking for use of alternative source terms is
applicable to holders of operating licenses issued prior to January 10,
1997, under 10 CFR Part 50, ``Domestic Licensing of Production and
Utilization Facilities,'' and to holders of renewed licenses under 10
CFR Part 54, ``Requirements for Renewal of Operating Licenses for
Nuclear Power Plants,'' whose initial operating license was issued
prior to January 10, 1997. The regulations of Part 50 are supplemented
by those in other parts of Chapter I of Title 10, including Part 100,
``Reactor Site Criteria.'' Part 100 contains language that
qualitatively defines a required accident source term and contains a
note that discusses the availability of TID-14844. With the exception
of Sec. 50.34(f), there are no explicit requirements in Chapter I of
Title 10 to use the TID-14844 accident source term. Section 50.34(f),
which addresses additional TMI-related requirements, is only applicable
to a limited number of construction permit applications pending on
February 16, 1982, and to applications under Part 52.
An applicant for an operating license is required by Sec. 50.34(b)
to submit a final safety analysis report (FSAR) that describes the
facility and its design bases and limits, and presents a safety
analysis of the structures, systems, and components of the facility as
a whole. Guidance in performing these analyses is given in regulatory
guides. In its review of the more recent applications for operating
licenses, the NRC has used the review procedures in NUREG-0800,
``Standard Review Plan for the Review of Safety Analysis Reports for
Nuclear Power Plants'' (SRP). These review procedures reference or
provide acceptable assumptions and analysis methods. The facility FSAR
documents the assumptions and methods actually used by the applicant in
the required safety analyses. The NRC's finding that a license may be
issued is based on the review of the FSAR, as documented in the
Commission's safety evaluation report (SER). Fundamental assumptions
that are design inputs, including the source term, were required to be
included in the FSAR and became part of the design basis 1
of the facility. From a regulatory standpoint, the requirement to use
the TID-14844 source term is expressed as a licensee commitment
(typically to Regulatory Guide 1.3 or 1.4) documented in the facility
FSAR, and is subject to the requirements of Sec. 50.59.
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\1\ As defined in Sec. 50.2, design bases means that information
which identifies the specific functions to be performed by a
structure, system, or component of a facility, and the specific
values or ranges of values chosen for controlling parameters as
reference bounds for design. These values may be (1) restraints
derived from generally accepted ``state of the art'' practices for
achieving functional goals, or (2) requirements derived from
analysis (based on calculation and/or experiments) of the effects of
a postulated accident for which a structure, system, or component
must meet its functional goals. The NRC considers the accident
source term to be an integral part of the design basis because it
sets forth specific values (or range of values) for controlling
parameters that constitute reference bounds for design.
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In 1996 (61 FR 65175; December 11, 1996), the NRC amended its
regulations in 10 CFR Parts 21, 50, 52, 54, and 100. That regulatory
action produced site criteria for future sites, presented a stable
regulatory basis for seismic and geologic siting and the engineering
design of future nuclear power plants to withstand seismic events, and
relocated source term and dose requirements for future plants into Part
50. Because these dose requirements tend to affect reactor design
rather than siting, they are more appropriately located in Part 50.
This decoupling of siting from design is consistent with the future
licensing of facilities using standardized plant designs, the design
features of which have been or will be certified in a separate design
certification rulemakings. This decoupling of siting from design was
directed by Congress in the 1980 Authorization Act for the NRC. Because
the revised criteria would not apply to operating reactors, the non-
seismic and seismic reactor site criteria for operating reactors were
retained as Subpart A and Appendix A to Part 100, respectively. The
revised reactor site criteria were added as Subpart B in Part 100, and
revised source term and dose requirements were moved to Sec. 50.34. The
existing source term and dose requirements of Subpart A of Part 100
will remain in place as the licensing bases for those operating
reactors that do not elect to use an alternative source term.
In relocating the source term and dose requirements for future
reactors to Sec. 50.34, the NRC retained the requirements for the
exclusion area and the low population zone, but revised the associated
numerical dose criteria to replace the two different doses for the
whole body and the thyroid gland with a single, total effective dose
equivalent (TEDE) value. The dose criteria for the whole body and the
thyroid, and the
[[Page 71993]]
immediate 2-hour exposure period were largely predicated by the assumed
source term being predominantly noble gases and radioiodines
instantaneously released to the containment and the assumed ``single
critical organ'' method of modeling the internal dose used at the time
that Part 100 was originally published. However, the current dose
criteria, by focusing on doses to the thyroid and the whole body,
assume that the major contributor to doses will be radioiodine.
Although this may be appropriate with the TID-14844 source term, as
implemented by Regulatory Guides 1.3 and 1.4, it may not be true for a
source term based on a more complete understanding of accident
sequences and phenomenology.
The postulated chemical and physical form of radioiodine in the
revised source terms is more amenable to mitigation and, as such,
radioiodine may not always be the predominant radionuclide in an
accident release. The revised source terms include a larger number of
radionuclides than did the TID-14844 source term as implemented in
regulatory guidance. The whole body and thyroid dose criteria ignore
these contributors to dose. The NRC amended its radiation protection
standards in Part 20 in 1991 (56 FR 23391; May 21, 1991) replacing the
single, critical organ concept for assessing internal exposure with the
TEDE concept that assesses the impact of all relevant nuclides upon all
body organs. TEDE is defined to be the deep dose equivalent (for
external exposure) plus the committed effective dose equivalent (for
internal exposure). The deep dose equivalent (DDE) is comparable to the
present whole body dose; the committed effective dose equivalent (CEDE)
is the sum of the products of doses (integrated over a 50-year period)
to selected body organs resulting from the intake of radioactive
material multiplied by weighting factors for each organ that are
representative of the radiation risk associated with the particular
organ.
The TEDE, using a risk-consistent methodology, assesses the impact
of all relevant nuclides upon all body organs. Although it is expected
that in many cases the thyroid could still be the limiting organ and
radioiodine the limiting radionuclide, this conclusion cannot be
assured in all potential cases. The revised source terms postulate that
the core inventory is released in a sequence of phases over 10 hours,
with the more significant release commencing at about 30 minutes from
the start of the event. The assumption that the 2-hour exposure period
starts immediately at the onset of the release is inconsistent with the
phased release postulated in the revised source terms. The final rule
adopts the future LWR dose criteria for operating reactors that elect
to use an alternative source term.
An accidental release of radioactivity can result in radiation
exposure to control room operators. Normal ventilation systems may draw
this activity into the control room where it can result in external and
internal exposures. Control room designs differ but, in general, design
features are provided to detect the accident or the activity and
isolate the normal ventilation intake. Emergency ventilation systems
are activated to minimize infiltration of contaminated air and to
remove activity that has entered the control room. Personnel exposures
can also result from radioactivity outside of the control room.
However, because of concrete shielding of the control room, these
latter exposures are generally not limiting. The objective of the
control room design is to provide a location from which actions can be
taken to operate the plant under normal conditions and to maintain it
in a safe condition under accident conditions. General Design Criterion
19 (GDC-19), ``Control Room,'' of Appendix A to 10 CFR Part 50 (36 FR
3255; February 20, 1971), establishes minimum requirements for the
design of the control room, including a requirement for radiation
protection features adequate to permit access to and occupancy of the
control room under accident conditions. The GDC-19 criteria were
established for judging the acceptability of the control room design
for protecting control room operators under postulated design basis
accidents, a significant concern being the potential increases in
offsite doses that might result from the inability of control room
personnel to adequately respond to the event.
The GDC-19 criteria are expressed in terms of whole body dose, or
its equivalent to any organ. The NRC did not revise the criteria when
Part 20 was amended (56 FR 23391; May 21, 1991) instead deferring such
action to individual facility licensing actions (NUREG/CR-6204,
``Questions and Answers Based on the Revised 10 CFR Part 20''). This
position was taken in the interest of maintaining the licensing basis
for those facilities already licensed. The NRC is replacing the current
dose criteria of GDC-19 for future reactors and for operating reactors
that elect to use an alternative source term with a criterion expressed
in terms of TEDE. The rationale for this revision is similar to the
rationale, discussed earlier in this preamble, for revising the dose
criteria for offsite exposures.
On January 10, 1997 (61 FR 65157), the NRC amended 10 CFR Parts 21,
50, 52, 54, and 100 of its regulations to update the criteria used in
decisions regarding power reactor siting for future nuclear power
plants. The NRC intended that future licensing applications in
accordance with Part 52 utilize a source term consistent with the
source term information in NUREG-1465 and the accident TEDE criteria in
Parts 50 and 100. However, during the final design approval (FDA) and
design certification proceeding for the Westinghouse AP600 advanced
light-water reactor design, the NRC staff and Westinghouse determined
that exemptions were necessary from Secs. 50.34(f)(2)(vii), (viii),
(xxvi), and (xxviii) and 10 CFR Part 50, Appendix A, GDC-19. This final
rule would eliminate the need for these exemptions for future
applicants under Part 52 by making conforming changes to Part 50,
Appendix A, GDC-19 and Sec. 50.34.
II. Analysis of Public Comments
The NRC published a proposed rule in the Federal Register (64 FR
12117, March 31, 1999); that would provide a regulatory framework for
the voluntary implementation of alternative source terms as a change to
the design basis at currently licensed power reactors, while retaining
the existing regulatory framework for currently licensed power reactor
licensees who choose not to implement an alternative source term. The
rule proposed relocating source term and dose requirements that apply
primarily to plant design into 10 CFR Part 50 for operating reactors
that choose to implement an alternative source term. The rule also
proposed conforming changes to Sec. 50.34(f) and Part 50, Appendix A,
GDC-19 to eliminate the need for exemptions for future applicants under
Part 52.
The NRC received seven letters commenting on the proposed rule. All
comments including those received by the NRC after the expiration of
the public comment period but before June 25, 1999, were considered.
The commenters included two State regulatory agencies, two nuclear
industry groups and three utilities. The State of Florida Department of
Community Affairs indicated that they had no comments on the proposed
rule. The State of New Jersey Department of Environmental Protection
concurred with the NRC's position on the use of an AST in emergency
preparedness applications and stated a desire to review the draft
regulatory guidance when issued. Winston & Strawn
[[Page 71994]]
submitted comments on behalf of the Nuclear Utility Backfitting and
Reform Group (NUBARG). The Nuclear Energy Institute (NEI) submitted
comments on behalf of the nuclear industry. Two of the utilities
provided comments, while the third endorsed the comments submitted by
NEI. Copies of these letters are available for public inspection and
copying for a fee at the NRC Public Document Room, 2120 L Street NW.
(Lower Level), Washington, DC.
1. NUBARG Comments
NUBARG supports the rule, noting that the rule as proposed defines
an acceptable regulatory process for implementing more realistic
accident source terms. NUBARG requested clarification in the final rule
of situations in which an alternative source term (AST) may be applied
in future backfitting 2 decisions. First, NUBARG suggests
that the NRC clarify the extent it intends to use the revised source
term in assessing whether new generic requirements provide a cost-
justified, substantial increase in safety in accordance with NRC's
backfitting rule, Sec. 50.109. NUBARG believes that continued use of
the source term in TID-14844 for this purpose in spite of its known
limitations would be inappropriate and could lead to overly
conservative estimates of the safety impact of proposed new
requirements. Second, NUBARG suggests a similar clarification for
plant-specific backfit decisions for plants that have not opted to
implement the revised source term. NUBARG believes that the NRC has
discretion to take all relevant factors into account in making its
safety benefit assessment of the proposed backfit, including the
current state of knowledge concerning the accident source term. NUBARG
suggested that the statements of considerations accompanying the final
rule address these issues. NUBARG also suggests that relevant NRC
guidance should also be revised to reflect NRC policy in these areas.
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\2\ As provided in Sec. 50.109, Backfitting is defined as the
modification of or addition to systems, structures, components, or
the design of a facility; or the design approval or manufacturing
license for a facility; or the procedures or organization required
to design, construct or operate a facility; any of which may result
from a new or amended provision in the Commission rules or the
imposition of a regulatory staff position interpreting the
Commission rules that is either new or different from a previously
applicable staff position.
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NRC Response. When radiological consequence analyses are involved,
the NRC expects to use a technically appropriate AST in evaluating
generic and plant-specific backfitting analyses, including those
proposed for facilities that have not implemented an AST. The NRC
agrees with the NUBARG position that the NRC has discretion to take all
new information on accident source terms into account. The NRC's
guidance for evaluating proposed NRC regulatory actions (including
backfitting) are contained in NUREG/BR-0058, ``Regulatory Analysis
Guidelines of the U.S. Nuclear Regulatory Commission,'' and NUREG/BR-
0184, ``Regulatory Analysis Technical Evaluation Handbook.'' These
documents state that value and impact (including adverse effects on
health and safety) parameters are to be best estimates, preferably mean
or expected values. These documents also provide that analyses are to
be based largely on risk considerations.
2. NEI Comment 1
NEI stated that the Section-by-Section Analysis in the proposed
rule notice is consistent with the NRC's intent to permit limited
application of the new research results. NEI noted that these limited
applications are of two types: (1) application of alternative source
term radiological composition and magnitude in a quantitative analysis
relative to the effect on the performance of a given engineered safety
feature; or (2) application of only the timing aspects in conjunction
with the original TID-14844 source term. NEI stated that proposed
Sec. 50.67 appears to apply to applications where a licensee would use
a completely new source term such as NUREG-1465 in all aspects of the
plant design. The NEI comment acknowledged that further guidance in a
subsequent regulatory guide and standard review plan is helpful and
necessary. Nonetheless, NEI is concerned that licensee pursuit of
either of these limited applications might ultimately require seeking
an exemption to Sec. 50.67, or require extensive analysis. NEI
recommended that the NRC should: (1) revise the proposed rule language
to accommodate limited application of an alternative source term as
done in the Section-By-Section Analysis; (2) provide clarification in
the Statement of Consideration (SOC) for the rule; and (3) for
applications that continue to use the TID source term but incorporate
attributes of newer technical insights such as timing of releases,
specify that the provisions of the proposed rule do not apply.
NRC Response. The language of Sec. 50.67(b) requires an evaluation
of the consequences of applicable design basis accidents. The NRC
believes that the use of the modifier applicable provides the basis for
processing selective implementations. Design basis accidents not
applicable to a particular selective implementation would not be
required to be evaluated. The NRC expects that the licensee will
evaluate all applicable impacts of the proposed AST implementation.
While a selective implementation may result in a reduced scope of
evaluation, the licensee must still demonstrate that the AST
implementation and any associated proposed modifications will not
result in accident conditions exceeding the criteria specified in
Sec. 50.67. Therefore, these criteria are applicable to full and
selective implementations alike. The scope of the required re-analyses
will depend on the specific application proposed by the licensee.
Guidance with regard to this scope is properly provided in the draft
regulatory guide prepared for this rule. Therefore, the NRC has decided
against revising the rule language as suggested by NEI. Consistent with
the second NEI recommendation, the NRC has modified paragraph D of the
section-by-section analysis to clarify this issue.
3. NEI Comment 2
In its second comment, NEI noted that the SOC provides that
licensees may need to perform additional evaluations of equipment
qualifications (Sec. 50.49). The SOC should discuss the circumstances
when such an evaluation may be necessary. NEI recommended that the SOC
should be amended to state that regardless of source term used, the
licensee would be required to re-evaluate the equipment qualification
only when a plant modification alters the plant configuration so that
the underlying assumptions, with respect to dose distribution and
effects, are materially altered. NEI summarized conclusions of several
references in support of its position. NEI stated that there is no
basis to require or expect additional analyses of equipment
qualification if a licensee applied the alternative source term in
limited scope applications, absent a plant configuration change that
materially alters the dose distribution and effects assumed in existing
analyses.
NRC Response. The re-baselining study prepared by the NRC staff
(SECY-98-154, June 30, 1998) considered the impact of an AST on
analyses of the postulated integrated radiation doses for plant
components exposed to containment atmosphere radiation sources and
those exposed to containment sump radiation sources. The staff's
conclusions regarding the atmosphere sources are consistent with those
identified by NEI in its comment. However, the re-baselining study also
concluded that the increased concentration of cesium in the containment
sump water could result in
[[Page 71995]]
an increase in the postulated integrated radiation doses for certain
plant components subject to equipment qualification. It is because of
this conclusion that the NRC included the discussion in the SOC
regarding re-evaluation of equipment environmental qualification. The
NEI comment provides no additional information that would cause the NRC
to change its position on this matter. Further, the NRC has determined
that it is necessary to consider the potential impact of the postulated
cesium concentration in the containment sump water as it applies to all
operating power reactors, not just to those licensees amending their
design basis to use an AST. Since the postulated increase in the
integrated dose occurs only following an accident, there is no adverse
effect on equipment relied upon to perform safety functions immediately
following an accident. Rather, this issue affects equipment that is
required to be operable longer than about 30 days to 4 months after an
accident. As such, the NRC determined that continued plant operation
does not pose an immediate threat to public health and safety. Also,
should such long-term equipment fail there will not be an undue threat
to public health and safety as protective actions for the public would
have already been implemented by the time the postulated failure could
occur. In addition, the time period between the onset of the event and
the projected failure allows compensatory measures to be taken to
prevent the equipment failure or to restore the degraded safety
function. The NRC will evaluate this issue as a generic safety issue to
determine whether further regulatory actions are justified. The final
regulatory guide, or subsequent revisions thereto, is expected to
reflect the resolution of this generic safety issue.
4. NEI Comment 3
NEI recommends that the definition of Source Term in Sec. 50.2 be
revised to ``Source term refers to the magnitude and mix of
radionuclides released from the fuel, their physical and chemical form,
and the timing of their release.'' NEI stated that the language in the
proposed rule would prohibit the use of Sec. 50.67 for accidents such
as the fuel handling accident.
NRC Response. The NRC agrees with the proposed revision. The
proposed definition was consistent with the definition of source term
as used in NUREG-1465, which was written primarily to address loss of
coolant accidents (LOCA). The regulatory guidance for this rule extends
the NUREG-1465 source terms to other accidents which involve core
damage. The definition suggested by NEI is consistent with the proposed
use of the AST. The Sec. 50.2 definition has been revised in the final
rule to reflect the change suggested by NEI and that suggested by
Arizona Public Service Comment 1 below.
5. NEI Comment 4
NEI stated that the proposed rule does not permit new test reactors
to use an alternative source term. New test reactors would have to use
the Part 100 Subpart A, ``Evaluation Factors for Stationary Power
Reactor Site Applications Before January 10, 1997, and for Testing
Reactors,'' even though their application for an operating license
would be filed after January 10, 1997. The use of Section 50.67,
``Accident Source Term,'' is limited to holders of operating licenses
issued before January 10, 1997. This wording prohibits new test
reactors from using the alternative source term. NEI recommended that
Sec. 50.67 be amended to allow new test reactors to use an alternative
source term.
NRC Response. Section 50.67 applies only to holders of licenses for
operating reactors, including test reactors, whose licenses were issued
before January 10, 1997. There is no regulatory requirement for a
specific source term for reactors to be licensed in the future,
including test reactors. Accordingly, no regulatory action is necessary
to accommodate the NEI recommendation.
6. Duke Energy Corporation Comment
Duke Energy Corporation (Duke) endorsed the comments submitted on
behalf of the industry by NEI. Duke stated that the proposed
Sec. 50.67(b)(1) was not clear regarding whether licensees will be
allowed to use a revised source term on a limited basis (e.g., for
analyses of a specific accident or function), or whether they will be
required to review the entire radiological consequence analyses to
apply for the new source term. Duke suggested that necessary guidance
be provided in the draft regulatory guidance to allow for limited use
of the new source terms where such use can be justified.
NRC Response. This comment is similar to NEI Comment 1 addressed
previously. As stated in the SOC, the NRC will consider justifiable
limited (i.e., selective) applications of an AST. Although a selective
implementation may result in a reduced scope of evaluation, the
licensee must still demonstrate that the AST implementation and any
associated proposed modifications will not exceed the criteria
specified in Sec. 50.67. The scope of the required re-analyses will
depend on the specific application proposed by the licensee. Regulatory
guidance on selective implements and the scope of required re-analyses
has been included in the draft guide and are available as announced in
this Federal Register notice.
7. Arizona Public Service Company Comment 1
Arizona Public Service Company (APS) noted that the SOC statement,
``a subsequent change to the source term must be made through a license
amendment'' could be interpreted as requiring prior NRC approval for
any change in the magnitude and mix of radionuclides released from the
reactor core. APS stated that this interpretation could place
additional restrictions on licensee efforts at economical fuel
management, including reload design.
NRC Response. The NRC agrees with the APS comment. The NRC had
intended the phrase ``magnitude and mix'' to refer to the fractions of
the fission product inventory of the radionuclides released from the
reactor fuel. The NRC intent for the provision in question was to
require approval for changes in the radioactivity release fractions,
the radionuclides released, their physical and chemical form, and the
timing of their release. Since ``magnitude and mix'' could be a source
of confusion, the NRC has modified the Sec. 50.2 definition of Source
Term in the final rule to read: ``Source term refers to the magnitude
and mix of the radionuclides released from the fuel, expressed as
fractions of the fission product inventory in the fuel, as well as
their physical and chemical form, and the timing of their release.''
This is consistent with NUREG-1465 when it refers to ``magnitude and
mix,'' since the NUREG-1465 presents these data in the form of tables
of release fractions and radionuclides. This revised language also
addresses NEI Comment 3 above.
8. Arizona Public Service Company Comment 2
In its second comment, APS noted that NUREG-1465 contains a
disclaimer that the accident source terms provided therein may not be
applicable to fuel irradiated in excess of 40 GWD/MTU. The NRC has
licensed core designs with fuel irradiations of up to 62 GWD/MTU. APS
questioned whether the NRC staff was going to address the affect of
high burnups on a generic basis, or on a facility-by-facility basis.
NRC Response. The AST tabulated in the draft regulatory guidance,
which
[[Page 71996]]
differs in some aspects from that provided in NUREG-1465, is applicable
to peak rod average irradiations up to 62 GWD/MTU. Attachment 1 to the
regulatory analysis for this rulemaking describes the bases of this
extension in fuel irradiation as it applies to the AST. There are some
facility-by-facility considerations. For example, the increase in core
inventory for some long-lived radionuclides and the change in isotopic
mix due to the increase in plutonium fission as the fuel ages is
addressed by the Draft Guide-1081 provision that licensees re-analyze
the core inventory based on current operating parameters, including
fuel burnup.
III. Section-by-Section Analysis
A. Section 50.2
The general ``definitions'' section for Part 50 is supplemented by
adding a definition of source term for the purpose of Sec. 50.67. In
NUREG-1465, the source term is defined by five projected
characteristics: (1) magnitude of radioactivity release, (2)
radionuclides released, (3) physical form of the radionuclides
released, (4) chemical form of the radionuclides released, and (5)
timing of the radioactivity release. The definition of source term in
Sec. 50.2 embodies the NUREG-1465 definition; however, the Sec. 50.2
definition includes the clarifying phrase, ``expressed as fractions of
the fission product inventory in the fuel,'' (see prior response to
Arizona Public Service Comment 1). Although all five characteristics
should be addressed in applications proposing the use of an alternative
source term, there may be technically justifiable applications in which
all five characteristics need not be addressed. The NRC intends to
allow licensees flexibility in implementing alternative source terms
consistent with maintaining a conservative, clear, logical, and
consistent plant design basis. The regulatory guidance that supports
this final rule describes an acceptable basis for defining the
characteristics of an alternative source term.
B. Section 50.67(a)
This paragraph defines the licensees that may seek to revise their
current radiological source term with an alternative source term. The
final rule is applicable to holders of operating licenses that were
issued under 10 CFR Part 50 before January 10, 1997, and to holders of
renewed licenses issued under 10 CFR Part 54 whose initial operating
license was issued prior to January 10, 1997. The final rule does not
require licensees to revise their current source term. The NRC
considered the acceptability of the TID-14844 source term at current
operating reactors and determined that the analytical approach based on
the TID-14844 source term would continue to be adequate to protect
public health and safety, and that operating reactors licensed under
this approach should not be required to reanalyze design basis
accidents using a new source term. The final rule does not explicitly
define an alternative source term. In lieu of an explicit reference to
NUREG-1465, Footnote 1 to the final rule identifies the significant
attributes of an accident source term. The regulatory guidance that is
being issued to support this final rule will identify ASTs (based on
the NUREG-1465 source terms) that are acceptable alternatives to the
source term in TID-14844, and will provide implementation guidance.
This approach will provide for future revised source terms if they are
developed and will allow licensees to propose additional alternatives
for NRC consideration.
C. Section 50.67(b)(1)
This paragraph of Sec. 50.67 identifies the information that a
licensee must submit as part of a license amendment application to use
an alternative source term. Because of the extensive use of the
accident source term in the design and operation of a power reactor and
the potential impact on postulated accident consequences and margins of
safety of a change of such a fundamental design assumption, the NRC has
determined that any change to the design basis to use an alternative
source term should be reviewed and approved by the NRC in the form of a
license amendment. Changes to the source term, by itself, would
ordinarily constitute a no significant hazards consideration. In
addition, generic analyses performed by the NRC staff in support of
this final rule have indicated that there are potential changes to the
facility as documented in the FSAR that will constitute a no
significant hazards consideration. However, these determinations will
have to be made for each proposed change based upon facility-specific
evaluations. The procedural requirements for processing a license
amendment are presented in Secs. 50.90 through 50.92.
The NRC's regulations provide a regulatory mechanism for a licensee
to effect a change in its design basis in Sec. 50.59 3 that
allows a licensee to make changes to the facility as described in the
final safety analysis report (FSAR) without prior NRC approval, if the
proposed change meets certain criteria specified in Sec. 50.59. If the
criteria are not met, the licensee must request NRC approval of the
change using the license amendment process detailed in Sec. 50.90.
Significant to this final rule is the criterion that NRC review is
required if the proposed change would result in a greater than minimal
increase in consequences of an accident or malfunction. In many
applications, alternative source terms may reduce the postulated
consequences of the accident or malfunction. For this reason, the NRC
determined that the regulatory framework of Sec. 50.59 might not
provide assurance that this change in the design basis would be
recognized by the licensee as needing review by the NRC staff.
---------------------------------------------------------------------------
\3\ Section 10 CFR 50.59 is being amended in a parallel, but
separate, rulemaking action. That rulemaking, when implemented is
expected to replace the unreviewed safety question (USQ) concept.
Further, the criteria for consequences are being revised from ``may
be increased'' to ``result in more than a minimal increase.'' Those
changes are not expected to invalidate the conclusions drawn in this
analysis.
---------------------------------------------------------------------------
After a licensee has been authorized to substitute an alternative
source term in its design basis, subsequent changes to the facility
that involve an alternative source term may be processed under
Sec. 50.59 or Sec. 50.90, as appropriate. However, a subsequent change
to the fractions of the fission product inventory of the radionuclides
released from the reactor fuel, their chemical and physical form, or
the timing of their release as tabulated in the regulatory guidance
(with deviations proposed by the licensee and approved by the NRC)
could not be implemented under Sec. 50.59. This provision applies only
to these tabulated parameters.
The final rule will require the applicant to perform analyses of
the consequences of applicable design basis accidents previously
analyzed in the safety analysis report and to submit a description of
the analysis inputs, assumptions, methodology, and results of these
analyses for NRC review. Applicable evaluations may include, but are
not limited to, those previously performed to show compliance with
Sec. 100.11, Sec. 50.49, Part 50 Appendix A GDC-19, Sec. 50.34(f), and
NUREG-0737, ``Clarification of TMI Action Plan Requirements,''
requirements II.B.2, II.B.3, III.D.3.4. The regulatory guidance that
supports this final rule will provide guidance on the scope and extent
of analyses used to show compliance with this rule and on the
assumptions and methods used therein. It is not the NRC's intent that
all of the design basis radiological analyses for a facility be
[[Page 71997]]
performed again as a prerequisite for approval of the use of an
alternative source term. Nor is it the NRC's intent that EAB, LPZ, and
control room dose calculations be performed for all applications under
Sec. 50.67. The NRC does expect that the applicant will perform
sufficient evaluations, supported by calculations as warranted, to
demonstrate the acceptability of the proposed amendment.
D. Sections 50.67(b)(2)(i),(ii), (iii)
These subparagraphs contain the three criteria for NRC approval of
the license amendment to use an alternative source term. A detailed
rationale for the use of 0.25 Sv (25 rem) TEDE as an accident dose
criterion and the use of the 2-hour exposure period resulting in the
maximum dose for future LWRs is provided at 61 FR 65157 (December 11,
1996). The same considerations that formed the basis for that rationale
are similarly applicable to operating reactors that elect to use an
alternative source term. The NRC believes that it is technically
appropriate and logical to extend the philosophy of decoupling of
design and siting, and the dose criteria established for future LWRs to
operating reactors that elect to use an alternative source term.
The NRC is replacing the current GDC-19 dose criteria for operating
reactors that elect to use an alternative source term with a criterion
of 0.05 Sv (5 rem) TEDE for the duration of the accident. This
criterion is included in Sec. 50.67 as well as in GDC-19 in order to
co-locate all of the dose requirements associated with alternative
source terms. The bases for the NRC's decision are: first, that the
criteria in GDC-19 and that in the final rule are based on a primary
occupational exposure limit. Second, the language in GDC-19: ``5 rem
whole body, or its equivalent to any part of the body'' is subsumed by
the definition of TEDE in Sec. 20.1003 and by the 0.05 Sv (5 rem) TEDE
annual limit in Sec. 20.1201(a). Although the weighting factors stated
in Sec. 20.1003 for use in determining TEDE differ in magnitude from
the weighting factors implied in the 0.3 Sv (30 rem) thyroid criteria
used for showing compliance with GDC-19, these differences are the
result of improvement in the science of assessing internal exposures
and do not represent a reduction in the level of protection. Third, as
discussed earlier, the use of TEDE in conjunction with alternative
source terms has been deemed appropriate and necessary. Fourth, the use
of TEDE for the control room dose criterion is consistent with the use
of TEDE in the accident dose criteria for offsite exposure.
The NRC has not included a ``capping'' limitation, an additional
requirement that the dose to any individual organ not be in excess of
some fraction of the total as provided for routine occupational
exposures. The bases for the NRC's decision are: first, that this non-
inclusion of a ``capping'' limitation is consistent with the final rule
published in December 11, 1996 (61 FR 65157), with regard to doses to
persons offsite. Second, the use of 0.05 Sv (5 rem) TEDE as the control
room criterion does not imply that this would be an acceptable exposure
during emergency conditions, or that other radiation protection
standards of Part 20, including individual organ dose limits, might not
apply. This criterion is provided only to assess the acceptability of
design provisions for protecting control room operators under
postulated DBA conditions. The DBA conditions assumed in these
analyses, although credible, generally do not represent actual accident
sequences but are specified as conservative surrogates to create
bounding conditions for assessing the acceptability of engineered
safety features. Third, Sec. 20.1206 permits a once-in-a-lifetime
planned special dose of five times the annual dose limits. Also,
Environmental Protection Agency (EPA) guidance sets a limit of five
times the annual dose limits for workers performing emergency services
such as lifesaving or protection of large populations.
Considering the individual organ weighting factors of Sec. 20.1003
and assuming that only the exposure from a single organ contributed to
TEDE, the organ dose, although exceeding the dose specified in
Sec. 20.1201(a), would be less than that considered acceptable as a
planned special dose or as an emergency worker dose. The NRC is not
suggesting that control room dose during an accident can be treated as
a planned special exposure or that the EPA emergency worker dose limits
are an alternative to GDC-19 or the final rule. However, the NRC does
believe that these provisions offer a useful perspective that supports
the conclusion that the organ doses implied by the 0.05 Sv (5 rem)
criterion can be considered to be acceptable due to the relatively low
probability of the events that could result in doses of this magnitude.
Although the dose criteria in the final rule supersede the dose
criteria in GDC-19, the other provisions of GDC-19 remain applicable.
There may be technically justifiable implementations of an AST that
would not require calculation of the EAB, LPZ, or control room doses.
For example, a proposed modification to change the closure time of a
containment isolation valve from 2 seconds to 5 seconds may be based on
the timing insights of the AST. Although a specific calculation might
not be necessary in this case, the licensee is still required to affirm
with reasonable assurance that the doses would comply with these stated
criteria.
E. 10 CFR Part 50, Appendix A, GDC-19
GDC-19 is changed to include the TEDE dose criterion for control
room design for applicants for construction permits, design
certifications, and combined licenses that submitted applications after
January 10, 1997 (the effective date of the 1996 rulemaking adopting
the TEDE criterion), and for those licenses using an alternative source
term under Sec. 50.67. The change to GDC-19 addresses the use of
alternative source terms at operating reactors and a deficiency
identified in the regulatory framework for early site permits, standard
design certifications, and combined licenses under Part 52. Sections
52.18, 52.48, and 52.81 establish that applications filed under Part
52, Subparts A, B, and C, respectively, will be reviewed according to
the standards given in 10 CFR Parts 20, 50, 51, 55, 73, and 100 to the
extent that those standards are technically relevant to the proposed
design. Therefore, GDC-19 is pertinent to applications under Part 52.
The final rule that became effective on January 10, 1997 (61 FR
65157; December 11, 1996), established accident TEDE criteria (in
Sec. 50.34) for applicants under Part 52 but did not change the
existing control room whole body (or equivalent) dose criterion in GDC-
19. Thus, exemptions from the dose criteria in the current GDC-19 were
necessary in the design certification process for the Westinghouse
AP600 advanced LWR in order to use the 0.05 Sv (5 rem) TEDE criterion
deemed necessary for use with alternative source terms. Exemptions will
arguably be necessary for future applicants for construction permits,
design certifications, and combined licenses. This amendment will
eliminate the need for these exemptions.
F. Sections 21.3, 50.2, 50.49(b)(1)(i)(C), 50.65(b)(1), and
54.4(a)(1)(iii)
These sections are revised to conform with the relocation of
accident dose criteria from Sec. 100.11 to Sec. 50.67 for operating
reactors that have amended their design bases to use an alternative
source term.
[[Page 71998]]
G. Section 50.34
A new footnote to Sec. 50.34 has been added to define what
constitutes an accident source term. This new footnote is identical to
the existing footnote 1 to Sec. 100.11, and was added to provide for
consistency between Parts 50 and 100.
H. Sections 50.34(f)(2)(vii), (viii), (xxvi) and (xxviii)
These paragraphs are revised to replace an explicit reference to
the ``TID-14844 source term'' with a more general reference to
``accident source term.'' These changes potentially affect three
classes of applicants. The first affected class is comprised of
applicants for design certification under Part 52, Subpart B. Section
52.47(a)(1)(ii) states that applications for combined licenses must
contain, inter alia, ``demonstration of compliance with any
technically-relevant portions of the Three Mile Island requirements set
forth in Sec. 50.34(f).'' Section 50.34(f) contains several references
to the TID-14844 source term. These references were modified to delete
the reference to TID-14844. This change makes it clear that applicants
for combined licenses should not use the TID-14844 source term but
should use the source term in the referenced design certification, or a
source term that is justified in the combined license application. The
second affected class is comprised of applicants for combined licenses
under Part 52, Subpart C. Section 52.79(b) makes the requirements of
52.47(a)(1)(i) applicable if a certified design is not referenced.
Thus, the combined license applicant is also subject to the
requirements of Section 50.34(f).
The third affected class is the small subset of plants that had
construction permits pending on February 16, 1982. With the proposed
change, these plants could use either the TID-14844 source term or an
alternative source term in their operating license applications.
IV. Draft Regulatory Guide; Issuance, Availability
The Nuclear Regulatory Commission is issuing for public comment a
draft of a guide planned for its Regulatory Guide Series. This series
has been developed to describe and make available to the public
information such as methods acceptable to the NRC staff for
implementing specific parts of the Commission's regulations, techniques
used by the staff in evaluating specific problems or postulated
accidents, and data needed by the NRC staff in its review of
applications for permits and licenses. Copies of the draft guide may be
obtained as described in Section VI, ``Referenced Documents,'' of these
statements of consideration. You may also download copies from the
NRC's interactive rulemaking forum website through the NRC home page
(http://ruleforum.llnl.gov/cgi-bin/rulemake).
The draft guide, temporarily identified by its task number DG-1081
(which should be mentioned in all correspondence concerning this draft
guide) is titled ``Alternative Radiological Source Terms for Evaluating
Design Basis Accidents at Nuclear Power Reactors.'' This guide is
intended for Division 1, ``Power Reactors.'' This draft guide is being
developed to provide regulatory guidance on the implementation of an
alternative source term at an operating reactor. The guide addresses
issues involving limited or selective implementation of an alternative
source term and probabilistic risk assessment (PRA) issues related to
plant modifications based on an alternative source term, and provides
guidance on the scope and extent of affected design basis accident
(DBA) radiological analyses and associated acceptance criteria. The
guide includes revised assumptions and methods for each affected DBA in
a series of appendices. These appendices supersede the guidance in
Regulatory Guides 1.3, 1.4, 1.5, 1.25, and 1.77, and supplement
guidance in Regulatory Guide 1.89 for those facilities using an
alternative source term.
The draft guide has not received complete NRC staff review and does
not represent an official NRC staff position.
Previous draft versions of DG-1081 have been made publicly
available to support technical interactions with the public. This
Federal Register announcement provides an opportunity for the public to
provide comments on the DG-1081 guidance. The NRC staff will consider
the public comments in its efforts to finalize the regulatory guidance.
The Commission invites advice and recommendations on the content of
the draft regulatory guide. Comments and suggestion are particularly
requested on the following questions.
A. Scope of Implementation
1. The guidance provided in the draft regulatory guide is intended
to allow licensees the maximum flexibility in pursuing technically
justifiable AST implementations provided that a clear, consistent, and
logical design basis is maintained. Comments are specifically requested
on the following questions.
A. Does the proposed guidance provide the desired flexibility while
providing reasonable assurance that a clear, consistent, and logical
design basis will be maintained?
B. Is there a less complex alternative approach that would provide
the desired flexibility while maintaining a clear, consistent, and
logical design basis?
C. Should the Commission allow licensees that have received
approval for a selective implementation to extend the AST and the TEDE
criteria to other design basis applications (that do not involve
reanalysis of the DBA LOCA) under Sec. 50.59 rather than under
Sec. 50.67 as currently proposed?
2. The guidance would allow selective implementation of the
characteristics (i.e., the fractions of fission product inventory of
the radionuclides released from the reactor fuel, their chemical and
physical form, and the timing of their release) of an AST. The
Commission believes that implementations based only on the timing
insights of an AST may be technically justifiable. The Commission
believes that the other combinations may be internally inconsistent.
Comments are specifically requested on the following questions.
A. What other combinations of AST characteristics are technically
consistent?
B. What plant modifications might be based on these combinations?
B. Scope of Re-Analyses
1. The draft regulatory guide provides guidance on the scope of the
re-analyses that should be performed to support an AST implementation.
Comments are requested on the following questions.
A. Is the proposed guidance on the scope of re-analyses technically
appropriate and clear? How could it be improved?
B. The guidance allows licensees to disposition certain impacts of
an AST on the basis of the NRC staff's re-baselining study. Does this
study or other documents provide a sufficient basis for the Commission
to generically disposition these impacts?
2. It may be possible for licensees to demonstrate that the doses
from certain affected analyses assessed using the prior source term and
dose methodology would be greater than the doses obtained using a
proposed AST and the TEDE methodology. The proposed guidance would
allow the licensee to disposition these affected analyses without re-
calculation. Nonetheless, the design basis would now include the
approved AST and TEDE criteria. The guidance in the draft regulatory
guide would require the licensee to update the calculation to be
consistent with the approved AST and dose methodology described in the
facility design basis in
[[Page 71999]]
the event of a subsequent re-calculation. Comments are requested on the
following questions.
A. Should the Commission allow licensees to continue to use the
prior source term and dose criteria for these analyses and not require
that they be updated on subsequent revisions?
B. If the analyses are not updated, how will licensees assure that
the earlier conclusion that the analyses are limiting remains valid
following subsequent revisions?
3. Analyses of the integrated radiation doses for environmental
qualification of certain equipment important to safety will be affected
by the increased concentration of radioactive cesium in the containment
sump water. The Commission has been considering the position that
licensees proposing to implement an AST must address all impacts of the
proposed implementation, including the impact of the increased cesium
concentration. However, the Commission now believes it may be necessary
for all operating power reactors to address the postulated increase in
the cesium concentration. The Commission will consider this issue as a
generic safety issue. Comments are requested on the following
questions.
A. Is there information that should be considered by the Commission
in resolving this generic issue?
B. If the Commission should conclude that there is safety
significance but that the costs of implementing corrective actions are
not justified on a generic basis, should licensees who are voluntarily
proposing to amend their design basis to use an AST be required to
address the impact of the increased cesium concentration?
C. If a licensee proposes a change in the plant configuration that
would result in an increase in the integrated dose for one or more
components and this licensee is also proposing, or has already
implemented an AST, should the re-analysis of the integrated dose be
based on that AST or on the prior TID14844 source term?
Comments may be accompanied by relevant information or supporting
data. Written comments may be mailed to: Secretary, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemakings and Adjudications Staff. Mail Stop O16C1. Copies of
comments received may be examined at the NRC Public Document Room, 2120
L Street NW., Washington, DC. Comments will be most helpful if received
by March 7, 2000.
You may also provide comments via the NRC's interactive rulemaking
website through the NRC home page (http://ruleforum.llnl.gov/cgi-bin/
rulemake). This site provides the availability to upload comments as
files (any format), if your web browser supports that function. For
information about the interactive rulemaking website, contact Ms. Carol
Gallagher, (301) 415-5905; or by internet electronic mail to
cag@nrc.gov. For information about the draft guide, contact Mr. Stephen
F. LaVie, (301) 415-1081; Internet electronic mail sfl@nrc.gov.
Although a time limit is given for comments on this draft guide,
comments and suggestions in connection with items for inclusion in
guides currently being developed or improvements in all published
guides are encouraged at any time.
V. Draft Standard Review Plan Section; Issuance, Availability
The Nuclear Regulatory Commission is issuing for public comment a
draft of a new section to NUREG-0800, ``Standard Review Plan.''
Standard review plan (SRP) sections are prepared for the guidance of
the Office of Nuclear Reactor Regulation staff responsible for the
review of applications to construct and operate nuclear power plants.
These documents are made available to the public as part of the
Commission's policy to inform the nuclear industry and the general
public of regulatory procedures and policies. The draft SRP Section
15.0.1, is titled ``Radiological Consequence Analyses Using Alternative
Source Terms.'' The SRP section complements draft regulatory guide DG-
1081. The draft SRP section has not received complete NRC staff review
and does not represent an official NRC staff position.
Copies of the draft SRP section may be obtained as described in
Section VI, ``Referenced Documents,'' of these statements of
consideration. You may also download copies from the NRC's interactive
rulemaking forum website through the NRC home page (http://
ruleforum.llnl.gov/cgi-bin/rulemake).
Comments on the content of the draft SRP section are invited.
Comments may be accompanied by relevant information or supporting data.
Comments should be submitted as described above for the draft
regulatory guide. Although a time limit is given for comments on this
draft SRP section, comments and suggestions in connection with items
for inclusion in SRP sections currently being developed or improvements
in all published SRP sections are encouraged at any time.
VI. Referenced Documents
Copies of NUREG-0737, NUREG-0800, NUREG-1465, NUREG/BR-0058, NUREG/
BR-184, and NUREG/CR-6204 may be purchased from the Superintendent of
Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington,
DC 20402-9328. Copies also are available from the National Technical
Information Service, 5285 Port Royal Road, Springfield, VA 22161. A
copy also is available for inspection and copying for a fee in the NRC
Public Document Room, 2120 L Street, NW (Lower Level), Washington, DC.
Single copies of regulatory guides, both active and draft may be
obtained free of charge by writing the Reproduction and Distribution
Services Section, OCIO, USNRC, Washington DC 20555-0001, or by fax to
(301) 415-2289, or by email to distribution@nrc.gov. Active guides may
also be purchased from the National Technical Information Service on a
standing order basis. Details of this service may be obtained by
writing NTIS, 5285 Port Royal Road, Springfield, VA 22161. Copies of
active and draft guides are available for inspection or copying for a
fee from the NRC Public Document Room at 2120 L Street NW., Washington
DC.
Copies of SECY-94-302, SECY-96-242, SECY-98-154, SECY-98-289, TID-
14844, and TR-105909 are available for inspection and copying for a fee
at the NRC Public Document Room, 2120 L Street, NW. (Lower Level),
Washington, DC.
VII. Finding of No Significant Environmental Impact: Availability
The NRC has determined under the National Environmental Policy Act
of 1969, as amended, and the NRC's regulations in Subpart A of 10 CFR
Part 51, that this regulation is not a major Federal action
significantly affecting the quality of the human environment and,
therefore, an environmental impact statement is not required. This
final rule allows operating reactors to replace the traditional TID-
14844 source term with a more realistic source term based on the
insights gained from extensive accident research activities. The actual
accident sequence and progression are not changed; it is the regulatory
assumptions regarding the accident that would be affected by the
change. The use of an alternative source term alone cannot increase the
core damage frequency (CDF) or the large early release frequency (LERF)
or actual offsite or onsite radiation doses. An alternative source term
could be used to justify changes in the plant design that might have an
impact on CDF or LERF or that might increase offsite or onsite doses.
Those plant changes that do not
[[Page 72000]]
require prior NRC review and approval pursuant to Sec. 50.59 are not
likely to involve any significant increase in environmental impacts.
The Sec. 50.59 criteria are sufficiently stringent that any potential
change in plant design that could have an adverse environmental impact
in all likelihood could not be made by the licensee without prior NRC
review and approval. Every plant change that requires NRC review and
approval under Sec. 50.59 requires a license amendment and, therefore,
the preparation of an environmental assessment to determine whether the
proposed change involves any significant environmental impact. Thus,
this final rule, by itself, will not result in plant changes that
involve any significant increase in environmental impacts. The final
rule does not affect non-radiological plant effluents.
The NRC requested public comments on any environmental justice
considerations that may be related to this rule. No public comments
relevant to the draft environmental assessment or environmental justice
considerations were received. The NRC requested the views of the States
on the environmental assessment for this rule. No comments relevant to
the draft environmental assessment or environmental justice
considerations were received.
The environmental assessment and finding of no significant impact
on which this determination is based are available for inspection at
the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC. Single copies of the environmental assessment and
finding of no significant impact are available from Mr. Stephen F.
LaVie, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory
NRC, Washington, DC 20555-0001, telephone: (301) 415-1081, or by
Internet electronic mail to sfl@nrc.gov.
VIII. Paperwork Reduction Act Statement
This final rule increases the burden on licensees by requiring that
when seeking to revise their current accident source term in design
basis radiological consequence analyses, they apply for an amendment
under Sec. 50.90. The public burden for this information collection is
estimated to average 609 hours per request. Because the burden for this
information collection is insignificant relative to the total burden
estimated, Office of Management and Budget (OMB) clearance is not
required. Existing requirements were approved by the Office of
Management and Budget, approval number 3150-0011.
Public Protection Notification
If an information collection does not display a currently valid OMB
control number, the NRC may not conduct or sponsor, and a person is not
required to respond to, the information collection.
IX. Regulatory Analysis
The Commission has prepared a regulatory analysis on this
regulation. Interested persons may examine a copy of the regulatory
analysis at the NRC Public Document Room, 2120 L Street NW. (Lower
Level), Washington, DC. Single copies of the analysis are available
from Mr. Stephen F. LaVie, Office of Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone:
(301) 415-1081, or by Internet electronic mail to sfl@nrc.gov.
X. Regulatory Flexibility Act Certification
As required by the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this regulation will not have a
significant economic impact on a substantial number of small entities.
This regulation will affect only the licensing and operation of nuclear
power plants. The companies that own these plants do not fall within
the definition of ``small entities'' found in the Regulatory
Flexibility Act or within the size standards established by the NRC
(April 11, 1995; 60 FR 18344).
XI. Backfit Analysis
The NRC has determined that the backfit rule in 10 CFR 50.109 does
not apply to this final rule, and that a backfit analysis is not
required for this rulemaking because these amendments do not involve
any provisions that would impose backfits as defined in 10 CFR
50.109(a)(1). This final rule amends the NRC's regulations by
establishing alternate requirements that may be voluntarily adopted by
licensees, and makes changes to the regulations to conform them to a
1996 rulemaking.
XII. Small Business Regulatory Enforcement Fairness Act
In accordance with the Small Business Regulatory Fairness Act of
1996, the NRC has determined that this action is not a major rule and
has verified this determination with the Office of Information and
Regulatory Affairs, Office of Management and Budget.
XIII. National Technology Transfer and Advancement Act
The National Technology Transfer Act of 1995, Pub. L. 104-113,
requires that Federal agencies use technical standards that are
developed or adopted by voluntary consensus standards bodies unless the
use of such a standard is inconsistent with applicable law or otherwise
impractical. In this final rule the NRC is establishing a government-
unique standard in Section 50.67(b)(2) by specifying accident radiation
dose criteria. These criteria were issued for use by future license
applicants by an earlier rulemaking (61 FR 65157, December 11, 1996)
and, by this final rule, are being applied to operating reactors that
voluntarily use an alternative source term. No voluntary consensus
standard has been identified that could be used instead of the
government-unique standard.
List of Subjects
10 CFR Part 21
Nuclear power plants and reactors, Penalties, Radiation protection,
Reporting and recordkeeping requirements.
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.
10 CFR Part 54
Administrative practice and procedure, Age-related degradation,
Backfitting, Classified information, Criminal penalties, Environmental
protection, Nuclear power plants and reactors, Reporting and
recordkeeping requirements.
For the reasons noted 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 the
following amendments to 10 CFR Parts 21, 50, and 54:
PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE
1. The authority citation for Part 21 continues to read as follows:
Authority: Sec. 161, 68 Stat. 948, as amended, sec. 234, 83
Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2953 (42 U.S.C.
2201, 2282, 2297f); secs. 201, as amended, 206, 88 Stat. 1242, as
amended, 1246 (42 U.S.C. 5841, 5846).
Section 21.2 also issued under secs. 135, 141, Pub. L. 97-425,
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
2. Section 21.3 is amended by republishing the introductory text
and revising paragraph (1)(i)(C) of the
[[Page 72001]]
definition of Basic Component to read as follows:
Sec. 21.3 Definitions.
As used in this part:
Basic component. (1)(i) * * *
(C) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or
Sec. 100.11 of this chapter, as applicable.
* * * * *
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
3. 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. 444, 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. 95-9601, 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. 91-9190, 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.
91-9190, 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. 97-9415, 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).
4. Section 50.2 is amended by republishing the introductory text
and revising paragraph (1)(iii) of the definition of Basic component,
and by adding in alphabetical order the definition for Source term to
read as follows:
Sec. 50.2 Definitions.
As used in this part,
* * * * *
Basic component * * *
(1) * * *
(iii) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or
Sec. 100.11 of this chapter, as applicable.
* * * * *
Source term refers to the magnitude and mix of the radionuclides
released from the fuel, expressed as fractions of the fission product
inventory in the fuel, as well as their physical and chemical form, and
the timing of their release.
* * * * *
5. Section 50.34 is amended by revising paragraphs (f)(2)(vii),
(viii), (xxvi), and (xxviii) to read as follows:
Sec. 50.34 Contents of applications; technical information.
* * * * *
(f) * * *
(2) * * *
(vii) Perform radiation and shielding design reviews of spaces
around systems that may, as a result of an accident, contain accident
source term \11\ radioactive materials, and design as necessary to
permit adequate access to important areas and to protect safety
equipment from the radiation environment. (II.B.2)
(viii) Provide a capability to promptly obtain and analyze samples
from the reactor coolant system and containment that may contain
accident source term \11\ radioactive materials without radiation
exposures to any individual exceeding 5 rems to the whole body or 50
rems to the extremities. Materials to be analyzed and quantified
include certain radionuclides that are indicators of the degree of core
damage (e.g., noble gases, radioiodines and cesiums, and nonvolatile
isotopes), hydrogen in the containment atmosphere, dissolved gases,
chloride, and boron concentrations. (II.B.3)
* * * * *
(xxvi) Provide for leakage control and detection in the design of
systems outside containment that contain (or might contain) accident
source term \11\ radioactive materials following an accident.
Applicants shall submit a leakage control program, including an initial
test program, a schedule for re-testing these systems, and the actions
to be taken for minimizing leakage from such systems. The goal is to
minimize potential exposures to workers and public, and to provide
reasonable assurance that excessive leakage will not prevent the use of
systems needed in an emergency. (III.D.1.1)
* * * * *
(xxviii) Evaluate potential pathways for radioactivity and
radiation that may lead to control room habitability problems under
accident conditions resulting in an accident source term \11\ release,
and make necessary design provisions to preclude such problems.
(III.D.3.4)
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\11\ The fission product release assumed for these calculations
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events, that would result in potential hazards not
exceeded by those from any accident considered credible. Such
accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
---------------------------------------------------------------------------
* * * * *
6. Section 50.49 is amended by revising paragraph (b)(1)(i)(C) to
read as follows:
Sec. 50.49 Environmental qualification of electric equipment important
to safety for nuclear power plants.
* * * * *
(b) * * *
(1) * * *
(i) * * *
(C) The capability to prevent or mitigate the consequences of
accidents that could result in potential offsite exposures comparable
to the guidelines in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or Sec. 100.11
of this chapter, as applicable.
* * * * *
7. Section 50.65 is amended by revising paragraph (b)(1) to read as
follows:
Sec. 50.65 Requirements for monitoring the effectiveness of
maintenance at nuclear power plants.
* * * * *
(b) * * *
(1) Safety-related structures, systems and components that are
relied upon to remain functional during and following design basis
events to ensure the integrity of the reactor coolant pressure
boundary, the capability to shut down the reactor and maintain it in a
safe shutdown condition, or the capability to prevent or mitigate the
consequences of accidents that could result in potential offsite
exposure comparable to the guidelines in Sec. 50.34(a)(1),
Sec. 50.67(b)(2), or Sec. 100.11 of this chapter, as applicable.
* * * * *
8. Part 50 is amended by adding Sec. 50.67 to read as follows:
Sec. 50.67 Accident source term.
(a) Applicability. The requirements of this section apply to all
holders of operating licenses issued prior to January 10, 1997, and
holders of renewed licenses under part 54 of this chapter whose initial
operating license was issued prior to January 10, 1997, who seek to
revise the current accident source term used in their design basis
radiological analyses.
(b) Requirements. (1) A licensee who seeks to revise its current
accident source term in design basis radiological
[[Page 72002]]
consequence analyses shall apply for a license amendment under
Sec. 50.90. The application shall contain an evaluation of the
consequences of applicable design basis accidents \1\ previously
analyzed in the safety analysis report.
---------------------------------------------------------------------------
\1\ The fission product release assumed for these calculations
should be based upon a major accident, hypothesized for purposes of
design analyses or postulated from considerations of possible
accidental events, that would result in potential hazards not
exceeded by those from any accident considered credible. Such
accidents have generally been assumed to result in substantial
meltdown of the core with subsequent release of appreciable
quantities of fission products.
---------------------------------------------------------------------------
(2) The NRC may issue the amendment only if the applicant's
analysis demonstrates with reasonable assurance that:
(i) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 0.25 Sv (25 rem) \2\ total effective dose equivalent
(TEDE).
---------------------------------------------------------------------------
\2\ 2 The use of 0.25 Sv (25 rem) TEDE is not intended to imply
that this value constitutes an acceptable limit for emergency doses
to the public under accident conditions. Rather, this 0.25 Sv (25
rem) TEDE value has been stated in this section as a reference
value, which can be used in the evaluation of proposed design basis
changes with respect to potential reactor accidents of exceedingly
low probability of occurrence and low risk of public exposure to
radiation.
---------------------------------------------------------------------------
(ii) An individual located at any point on the outer boundary of
the low population zone, who is exposed to the radioactive cloud
resulting from the postulated fission product release (during the
entire period of its passage), would not receive a radiation dose in
excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
(iii) Adequate radiation protection is provided to permit access to
and occupancy of the control room under accident conditions without
personnel receiving radiation exposures in excess of 0.05 Sv (5 rem)
total effective dose equivalent (TEDE) for the duration of the
accident.
9. Part 50, Appendix A, section II, ``Protection by Multiple
Fission Product Barriers,'' ``Criterion 19--Control room'' is revised
to read as follows:
Appendix A to Part 50--General Design Criteria for Nuclear Power
Plants
* * * * *
II. Protection by Multiple Fission Product Barriers
* * * * *
Criterion 19--Control room. A control room shall be provided
from which actions can be taken to operate the nuclear power unit
safely under normal conditions and to maintain it in a safe
condition under accident conditions, including loss-of-coolant
accidents. Adequate radiation protection shall be provided to permit
access and occupancy of the control room under accident conditions
without personnel receiving radiation exposures in excess of 5 rem
whole body, or its equivalent to any part of the body, for the
duration of the accident. Equipment at appropriate locations outside
the control room shall be provided (1) with a design capability for
prompt hot shutdown of the reactor, including necessary
instrumentation and controls to maintain the unit in a safe
condition during hot shutdown, and (2) with a potential capability
for subsequent cold shutdown of the reactor through the use of
suitable procedures.
Applicants for and holders of construction permits and operating
licenses under this part who apply on or after January 10, 1997,
applicants for design certifications under part 52 of this chapter
who apply on or after January 10, 1997, applicants for and holders
of combined licenses under part 52 of this chapter who do not
reference a standard design certification, or holders of operating
licenses using an alternative source term under Sec. 50.67, shall
meet the requirements of this criterion, except that with regard to
control room access and occupancy, adequate radiation protection
shall be provided to ensure that radiation exposures shall not
exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as
defined in Sec. 50.2 for the duration of the accident.
* * * * *
PART 54--REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR
POWER PLANTS
10. The authority citation for Part 54 continues to read as
follows:
Authority: Secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 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, 202, 206, 88 Stat. 1242,
1244, as amended (42 U.S.C. 5841, 5842), E.O. 12829, 3 CFR, 1993
Comp., p. 570; E.O. 12958, as amended, 3 CFR, 1995 Comp., p. 333;
E.O. 12968, 3 CFR, 1995 Comp., p. 391.
11. Section 54.4 is amended by revising paragraph (a)(1)(iii) to
read as follows:
Sec. 54.4 Scope.
(a) * * *
(1) * * *
(iii) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or
Sec. 100.11 of this chapter, as applicable.
* * * * *
Dated at Rockville, Maryland, this 17th day of December 1999.
For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 99-33283 Filed 12-22-99; 8:45 am]
BILLING CODE 7590-01-P
