[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
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 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  

[[Page 12117]]


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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 
[email protected].
    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 [email protected].

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

[[Page 12118]]

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

[[Page 12119]]

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

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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 [email protected].

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 [email protected].

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