[Federal Register Volume 64, Number 47 (Thursday, March 11, 1999)]
[Proposed Rules]
[Pages 12117-12126]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-6058]
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Proposed Rules
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 64, No. 47 / Thursday, March 11, 1999 /
Proposed Rules
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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: Proposed rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend
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 would 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 also
proposing to amend its regulations to revise certain sections to
conform with the final rule published on December 11, 1996, concerning
reactor site criteria.
DATES: The comment period expires on May 25, 1999. Comments received
after this date will be considered, if it is practical to do so, but
the NRC is able to assure consideration only for comments received on
or before this date.
ADDRESSES: Mail written comments to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, Mail Stop O16C1.
Deliver comments to: One White Flint North, 11555 Rockville Pike,
Rockville, Maryland, 20852, between 7:30 a.m. and 4:15 p.m. on Federal
workdays.
You may also submit comments via the NRC's interactive rulemaking
web site, ``Rulemaking Forum,'' through the NRC home page (http://
www.nrc.gov). This site enables people to transmit comments as files
(in any format, but WordPerfect version 6.1 is preferred), if your web
browser supports that function. Information on the use of the
Rulemaking Forum is available on the website. For additional assistance
on the use of the interactive rulemaking site, contact Ms. Carol
Gallagher, telephone: 301-415-5905; or by Internet electronic mail to
cag@nrc.gov.
Certain documents related to this rulemaking, including comments
received and the environmental assessment and finding of no significant
impact may be examined at the NRC Public Document Room, 2120 L Street,
NW. (Lower Level), Washington, DC. These same documents also may be
viewed and downloaded electronically via the interactive rulemaking
website established by NRC for this rulemaking.
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. Objectives
III. Alternatives
IV. Section-by-Section Analysis
V. Future Regulatory Action
VI. Referenced Documents
VII. Draft Finding of No Significant Environmental Impact;
Availability
VIII. Paperwork Reduction Act Statement
IX. Regulatory Analysis
X. Regulatory Flexibility Certification
XI. Backfit Analysis
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 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 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.
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
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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 known
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).
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 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, dated December
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 unduly conservative for design basis
analysis purposes. These latter releases could only result from core
damage accidents with vessel failure and core-concrete interactions.
The estimated frequencies of such scenarios are low enough that they
need not be considered for the purpose of meeting the requirements of
Sec. 100.11 or, as proposed herein, Sec. 50.67.
The objective of NUREG-1465 was to define revised accident source
terms for regulatory application for future light water reactors. 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 light water reactors (LWRs) of current 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 AP-600 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 also concluded that some licensees
may wish to use an alternative source term in analyses to support
operational flexibility and cost-beneficial licensing actions. 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 1996). Second, the NRC
initiated a comprehensive 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 1998) described the conclusions of
this assessment. Third, the NRC accepted license amendment requests
related to implementation of the revised source
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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 will be incorporated into the regulatory guidance
that will be 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. The
proposed rule and the supporting regulatory guidance that will be
developed as part of this rulemaking have resulted from this
assessment. The NRC plans to issue the supporting regulatory guidance
for public comment on the same day as it publishes the final rule.
This proposed rulemaking for use of alternative source terms is
applicable only to those facilities for which a construction permit was
issued before January 10, 1997, under 10 CFR Part 50, ``Domestic
Licensing of Production and Utilization Facilities.'' The regulations
of this part 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). By their inclusion in the
FSAR, the assumptions (including the source term) become 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 10 CFR Part 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 January 1997 (61 FR 65157), 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 plan designs, the design features of
which will be certified in a separate design certification rulemaking.
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 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
[[Page 12120]]
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 proposed rule would extend the future
LWR dose criteria to 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) instead deferring such action to
individual facility licensing actions (NUREG/CR-6204). This position
was taken in the interest of maintaining the licensing basis for those
facilities already licensed. The NRC is proposing to replace the
current GDC-19 dose criteria 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 AP-600 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 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. Objectives
The objectives of this proposed regulatory action are to--
1. Provide a regulatory framework for the voluntary implementation
of alternative source terms as a change to the design basis at
currently licensed power reactors, thereby enabling potential cost-
beneficial licensing actions while continuing to maintain existing
safety margins and defense in depth.
2. Retain the existing regulatory framework for currently licensed
power reactor licensees who choose not to implement an alternative
source term, but continue to comply with their existing source term.
3. Relocate 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, and
4. Implement 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.
III. Alternatives
The first alternative considered by the NRC was to continue using
current regulations for accident dose criteria and control room dose
criteria. This is not considered to be an acceptable alternative. As
discussed in the statements of consideration for the final siting rule
(61 FR 65157, 65159; December 11, 1996), the NRC determined that dose
criteria expressed in terms of whole body and thyroid doses were
inconsistent with the use of new source terms not based upon TID-14844.
With regard to the exclusion area dose guideline, the NRC had
previously determined (id. at 65160) that the dose criterion applies to
the 2-hour period resulting in the maximum dose.
The second alternative considered by the NRC was the replacement of
the existing guidelines in Sec. 100.11 and the existing criteria in 10
CFR Part 50 Appendix A, GDC-19 with revised dose criteria. This is not
considered to be a desirable alternative because the provisions of the
existing regulations form part of the licensing bases for many of the
operating reactors. Therefore, these provisions must remain in effect
for operating reactors that do not implement an alternative source
term. In addition, this alternative would also be inconsistent with the
NRC's philosophy of separating plant siting criteria and dose
requirements.
The approach of establishing the requirements for use of
alternative source terms in a new section to Part 50 while retaining
the existing regulations in Part 100 Subpart A and Part 50 Appendix A
GDC-19 was chosen as the best alternative.
The NRC considered alternatives with regard to providing regulatory
guidance to support the new section to Part 50. The first option was to
issue no additional regulatory guidance. This option was not considered
to be acceptable because in the absence of clear regulatory guidance,
licensee efforts in preparing applications and the NRC staff review of
submitted applications, could be hindered by differences in
interpretations and technical positions. This could result in the
inefficient use of licensee and NRC staff resources, could cause
licensing delays, and lead to less uniform and less consistent
regulatory implementation.
The second option was to replace the existing regulatory guides
that address the radiological consequences of accidents with new
revisions. This is not considered to be an acceptable choice because
the provisions of the existing regulatory guides form part of the
licensing bases for many of the operating reactors. Therefore, these
provisions must remain in effect for those operating reactors that do
not implement an alternative source term. The third option was to issue
a new regulatory guide on the implementation
[[Page 12121]]
of alternative source terms that would include revised assumptions and
acceptable analysis methods for each design basis accident in a series
of appendices. The approach of issuing a new regulatory guide was
determined to be the best option. To provide review guidance for the
NRC staff, a new section on design basis radiological analyses using
alternative source terms would be added to the Standard Review Plan.
IV. Section-by-Section Analysis
A. Section 50.2
The general ``definitions'' section for Part 50 would be
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. 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 guide that supports this
proposed rule will contain guidance on an acceptable basis for defining
the characteristics of an alternative source term.
B. Section 50.67(a)
This paragraph would define the licensees that may seek to revise
their current radiological source term with an alternative source term.
The proposed rule is applicable only to holders of nuclear power plant
operating licenses that were issued under 10 CFR Part 50 before January
10, 1997. The proposed rule would 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 proposed
rule does not explicitly define an alternative source term. In lieu of
an explicit reference to NUREG-1465, Footnote 1 to the proposed rule
identifies the significant characteristics of an accident source term.
The regulatory guide that will be issued to support this proposed rule
will identify the NUREG-1465 source terms as acceptable alternatives to
the source term in TID-14844, and will provide implementation guidance.
This approach would provide for future revised source terms if they are
developed and would allow licensees to propose additional alternatives
for NRC consideration.
C. Section 50.67(b)(1)
This paragraph of Sec. 50.67 would state 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 proposed rule have indicated that there are potential changes to
the facility as documented in the FSAR which would constitute a no
significant hazards consideration. However, such determinations would
have to be made for each proposed change based upon facility-specific
evaluations. The procedural requirements for processing a license
amendment are given 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. That section
allows a licensee to make changes to the facility as described in the
final safety evaluation report (FSAR) without prior NRC approval,
unless the proposed change is deemed to involve an unreviewed safety
question (USQ), or involves a change to the technical specifications
incorporated into the facility license. If a USQ is determined to exist
or if a change to the technical specifications is involved, the
licensee must request NRC approval of the change using the license
amendment process detailed in Sec. 50.90. The criteria for determining
that a USQ is involved appear in Sec. 50.59. Significant to this
proposed rule is the criterion that a USQ would exist if the proposed
change resulted in an 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
does not provide assurance that this change in the design basis would
be recognized by the licensee as needing review by the NRC staff. 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 source
term itself could not be implemented under Sec. 50.59; in all cases a
change to the source term must be made through a license amendment.
The proposed rule would 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 requirements II.B.2, II.B.3, III.D.3.4. The regulatory guide
that supports this proposed 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
performed again as a prerequisite for approval of the use of an
alternative source term. 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 would 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
[[Page 12122]]
for future LWRs to operating reactors that elect to use an alternative
source term.
The NRC is proposing to replace 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 would be included in Sec. 50.67 rather than
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 proposed 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 is not including 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 proposed rule. However, the NRC does believe that these
provisions offer a useful perspective that supports the conclusion that
the organ doses implied by the proposed 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 proposed rule would supersede the
dose criteria in GDC-19, the other provisions of GDC-19 remain
applicable.
E. 10 CFR Part 50, Appendix A, GDC-19
GDC-19 would be changed to include the TEDE dose criterion for
control room design for applicants for construction permits, design
certifications, and combined operating 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 proposed 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 AP-600 advanced LWR in order
to use the 0.05 Sv (5 rem) TEDE criterion deemed necessary for use with
alternative source terms. Exemptions would arguably be necessary for
future applicants for construction permits, design certifications, and
combined operating licenses. This proposed change would 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 would be 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.
G. Section 50.34
A new footnote to Sec. 50.34 would be added to define what
constitutes an accident source term. This new footnote is identical to
the existing footnote 1 to Sec. 100.11, and is being added to provide
for consistency between Parts 50 and 100.
H. Sections 50.34(f)(2)(vii), (viii), (xxvi) and (xxviii)
These paragraphs would be 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 two classes
of applicants. The first affected class is facilities that obtain
combined licenses under part 52. Section 52.47(a)(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 would be modified to delete the reference to
TID-14844. This would make it clear that applicants for combined
licenses would not use the TID-14844 source term but would 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 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.
[[Page 12123]]
V. Future Regulatory Action
The NRC is developing the following regulatory guides and Standard
Review Plan sections to provide prospective applicants with the
necessary guidance for implementing the proposed regulation. The draft
guide and draft Standard Review Plan section will be issued to coincide
with the publication of the final regulations that would implement this
proposed rulemaking. A notice of availability for these materials will
be published in the Federal Register at a future date.
1. Draft Guide DG-1081, ``Alternative Radiological Source Terms for
Evaluating the Radiological Consequences of Design Basis Accidents at
Boiling and Pressurized Water Reactors''
This guide is expected to present regulatory guidance on the
implementation of an alternative source term at an operating reactor.
The guide is expected to address 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 to provide guidance on the scope and
extent of affected DBA radiological analyses and associated acceptance
criteria. The guide is expected to include revised assumptions and
methods for each affected DBA in a series of appendices. These
appendices will supersede the guidance in Regulatory Guides 1.3, 1.4,
1.25, and 1.77, and will supplement guidance in Regulatory Guide 1.89
for those facilities using an alternative source term.
2. Standard Review Plan Section, 15.0.1, ``Radiological Consequence
Analyses Using Alternative Source Terms''
This SRP section presents guidance to NRC staff in the review of
the adequacy of licensee submittals requesting approval for use of an
alternative source term.
VI. Referenced Documents
Copies of NUREG-0737, NUREG-0800, NUREG-1465, 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.
Copies of issued regulatory guides may be purchased from the
Government Printing Office (GPO) at the current GPO price. Information
on current GPO prices may be obtained by contacting the Superintendent
of Documents, U.S. Government Printing Office, P.O. Box 37082,
Washington, DC 20402-9328. Issued guides also may be purchased from the
National Technical Information Service (NTIS) on a standing order
basis. Details on this service may be obtained by writing NTIS, 5826
Port Royal Road, Springfield, VA 22161.
Copies of SECY-94-302, SECY-96-242, SECY-98-154, TID14844, 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. Draft 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
proposed rule would allow 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 would not be 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. These potential changes are subject to existing
requirements in the NRC's regulations. Thus, the level of protection of
public health and safety provided in NRC regulations would not be
decreased by this proposed rule. The proposed rule would not affect
non-radiological plant effluents and would have no significant
environmental impact.
As discussed above, the determination of the environmental
assessment is that there would be no significant offsite impact on the
public from this action. However, the general public should note that
the NRC welcomes public participation. Also, the NRC has committed
itself to complying in all its actions with Executive Order (E.O.)
12898, ``Federal Actions to Address Environmental Justice in Minority
Populations and Low-Income Populations,'' dated February 11, 1994. In
accordance with that Executive Order, the NRC has determined that there
are no disproportionately high and adverse impacts on minority and low
income parties. In the letter and spirit of E.O. 12898, the NRC is
requesting public comments on any environmental justice considerations
or questions that the public thinks may be related to this proposed
rule, but that somehow were not addressed. The NRC uses the following
working definition of environmental justice: Environmental justice
means the fair treatment and meaningful involvement of all people,
regardless of race, ethnicity, culture, income, or educational level
with respect to the development, implementation and enforcement of
environmental laws, regulations, and policies. Comments on any aspect
of the environmental assessment, including environmental justice, may
be submitted to the NRC as indicated under the ADDRESSES heading.
The draft environmental assessment and the draft 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 proposed 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, 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.
[[Page 12124]]
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 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 proposed 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 proposed regulation and that a backfit analysis is
not required for this proposed regulation because these amendments do
not involve any provisions that would impose backfits as defined in 10
CFR 50.109(a)(1). This proposed regulation amends the NRC's regulations
by establishing alternate requirements that may be voluntarily adopted
by licensees.
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 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,
by 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 radionuclides
released from the reactor core to the reactor containment, 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)
---------------------------------------------------------------------------
\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.
---------------------------------------------------------------------------
(viii) Provide a capability to promptly obtain and analyze samples
from the reactor coolant system and containment that may contain
accident source term 12 radioactive materials without
radiation exposures to any individual exceeding 5 rems to the whole
body or 50 rems to
[[Page 12125]]
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)
---------------------------------------------------------------------------
\12\ See footnote 11 to paragraph (f)(2)(vii) of this section.
---------------------------------------------------------------------------
* * * * *
(xxvi) Provide for leakage control and detection in the design of
systems outside containment that contain (or might contain) accident
source term 13 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)
---------------------------------------------------------------------------
\13\ See footnote 11 to paragraph (f)(2)(vii) of this section.
---------------------------------------------------------------------------
* * * * *
(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 14
release, and make necessary design provisions to preclude such
problems. (III.D.3.4)
---------------------------------------------------------------------------
\14\ See footnote 11 to paragraph (f)(2)(vii) of this section.
---------------------------------------------------------------------------
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, 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 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\ 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, II., General Design Criterion 19, is
revised to read as follows:
Appendix A to Part 50--General Design
Criteria for Nuclear Power Plants
* * * * *
II. * * *
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 construction permits under this part or a design
certification or combined license under part 52 of this chapter who
apply on or after January 10, 1997, 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:
[[Page 12126]]
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 5th day of March 1999.
For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 99-6058 Filed 3-10-99; 8:45 am]
BILLING CODE 7590-01-U
