[Federal Register Volume 64, Number 246 (Thursday, December 23, 1999)]
[Rules and Regulations]
[Pages 71990-72002]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-33283]


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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 21, 50, and 54

RIN 3150-AG12


Use of Alternative Source Terms at Operating Reactors

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
regulations to allow holders of operating licenses for nuclear power 
plants to voluntarily replace the traditional source term used in 
design basis accident analyses with alternative source terms. This 
action will allow interested licensees to pursue cost beneficial 
licensing actions to reduce unnecessary regulatory burden without 
compromising the margin of safety of the facility. The NRC is 
announcing the availability of a draft regulatory guide and a draft 
Standard Review Plan section on this subject for public comment. The 
NRC is also amending its regulations to revise certain sections to 
conform with the final rule published on December 11, 1996, concerning 
reactor site criteria.

EFFECTIVE DATE: January 24, 2000.

FOR FURTHER INFORMATION CONTACT: Mr. Stephen F. LaVie, Office of 
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001; telephone: (301) 415-1081; or by Internet 
electronic mail to [email protected].

SUPPLEMENTARY INFORMATION:

I. Background
II. Analysis of Public Comments
III. Section-by-Section Analysis
IV. Draft Regulatory Guide; Issuance, Availability
V. Draft Standard Review Plan Section; Issuance, Availability
VI. Referenced Documents
VII. Finding of No Significant Environmental Impact; Availability
VIII. Paperwork Reduction Act Statement
IX. Regulatory Analysis
X. Regulatory Flexibility Act Certification
XI. Backfit Analysis
XII. Small Business Regulatory Enforcement Fairness Act
XIII. National Technology Transfer and Advancement Act

I. Background

    A holder of an operating license (i.e., the licensee) for a light-
water power reactor is required by regulations issued by the NRC (or 
its predecessor, the U.S. Atomic Energy Commission, (AEC)) to submit a 
safety analysis report (or, for early reactors, a hazard summary 
report) that contains assessments of the radiological consequences of 
potential accidents and an evaluation of the proposed facility site. 
The NRC uses this information in its evaluation of the suitability of 
the reactor design and the proposed site as required by its regulations 
contained in 10 CFR Parts 50 and 100. Section 100.11, which was adopted 
by the AEC in 1962 (27 FR 3509; April 12, 1962), requires an applicant 
to assume (1) a fission product release from the reactor core, (2) the 
expected containment leak rate, and (3) the site meteorological 
conditions to establish an exclusion area and a low population zone. 
This fission product release is based on a major accident that would 
result in substantial release of appreciable quantities of fission 
products from the core to the containment atmosphere. A note to 
Sec. 100.11 states that Technical Information Document (TID) 14844, 
``Calculation of Distance Factors for

[[Page 71991]]

Power and Test Reactors,'' may be used as a source of guidance in 
developing the exclusion area, the low population zone, and the 
population center distance. Changes to the design of the facility and 
the procedures for operating the facility are evaluated in part by 
determining whether there are changes to the calculated fission product 
release.
    The fission product release from the reactor core into containment 
is referred to as the ``source term'' and it is characterized by the 
composition and magnitude of the radioactive material, the chemical and 
physical properties of the material, and the timing of the release from 
the reactor core. The accident source term is used to evaluate the 
radiological consequences of design basis accidents (DBAs) in showing 
compliance with various requirements of the NRC's regulations. Although 
originally used for site suitability analyses, the accident source term 
is a design parameter for accident mitigation features, equipment 
qualification, control room operator radiation doses, and post-accident 
vital area access doses. The measurement range and alarm setpoints of 
some installed plant instrumentation and the actuation of some plant 
safety features are based in part on the accident source term. The TID-
14844 source term was explicitly stated as a required design parameter 
for several Three Mile Island (TMI)-related requirements.
    The NRC's methods for calculating accident doses, as described in 
Regulatory Guide 1.3, ``Assumptions Used for Evaluating the Potential 
Radiological Consequences of a Loss of Coolant Accident for Boiling 
Water Reactors''; Regulatory Guide 1.4, ``Assumptions Used for 
Evaluating the Potential Radiological Consequences of a Loss of Coolant 
Accident for Pressurized Water Reactors''; and NUREG-0800, ``Standard 
Review Plan for the Review of Safety Analysis Reports for Nuclear Power 
Plants,'' were developed to be consistent with the TID-14844 source 
term and the whole body and thyroid dose guidelines stated in 
Sec. 100.11. In this regulatory framework, the source term is assumed 
to be released immediately to the containment at the start of the 
postulated accident. The chemical form of the radioiodine released to 
the containment atmosphere is assumed to be predominantly elemental, 
with the remainder being small fractions of particulate and organic 
iodine forms. Radiation doses are calculated at the exclusion area 
boundary (EAB) for the first 2 hours and at the low population zone 
(LPZ) for the assumed 30-day duration of the accident. The whole body 
dose comes primarily from the noble gases in the source term. The 
thyroid dose is based on inhalation of radioiodines. In analyses 
performed to date, the thyroid dose has generally been limiting. The 
design of some engineered safety features, such as containment spray 
systems and the charcoal filters in the containment, the building 
exhaust, and the control room ventilation systems, are predicated on 
these postulated thyroid doses. Subsequently, the NRC adopted the whole 
body and thyroid dose criteria in Criterion 19 of 10 CFR Part 50, 
Appendix A (36 FR 3255; February 20, 1971).
    The source term in TID-14844 is representative of a major accident 
involving significant core damage and is typically postulated to occur 
in conjunction with a large loss-of-coolant accident (LOCA). Although 
the LOCA is typically the maximum credible accident, NRC experience in 
reviewing license applications has indicated the need to consider other 
accident sequences of lesser consequence but higher probability of 
occurrence. Some of these additional accident analyses may involve 
source terms that are a fraction of those specified in TID-14844. The 
DBAs were not intended to be actual event sequences but, rather, were 
intended to be surrogates to enable deterministic evaluation of the 
response of the plant engineered safety features. These accident 
analyses are intentionally conservative in order to address 
uncertainties in accident progression, fission product transport, and 
atmospheric dispersion. Although probabilistic risk assessments (PRAs) 
can provide useful insights into system performance and suggest changes 
in how the desired defense in depth is achieved, defense in depth 
continues to be an effective way to account for uncertainties in 
equipment and human performance. The NRC's policy statement on the use 
of PRA methods (60 FR 42622; August 16, 1995) calls for the use of PRA 
technology in all regulatory matters in a manner that complements the 
NRC's deterministic approach and supports the traditional defense-in-
depth philosophy.
    Since the publication of TID-14844, significant advances have been 
made in understanding the timing, magnitude, and chemical form of 
fission product releases from severe nuclear power plant accidents. 
Many of these insights developed out of the major research efforts 
started by the NRC and the nuclear industry after the accident at Three 
Mile Island (TMI). In 1995, the NRC published NUREG-1465, ``Accident 
Source Terms for Light-Water Nuclear Power Plants,'' which utilized 
this research to provide more physically based estimates of the 
accident source term that could be applied to the design of future 
light-water power reactors. The NRC sponsored significant review 
efforts by peer reviewers, foreign research partners, industry groups, 
and the general public (request for public comment was published in 57 
FR 33374; July 28, 1992).
    The information in NUREG-1465 presents a representative accident 
source term (``revised source term'') for a boiling-water reactor (BWR) 
and for a pressurized-water reactor (PWR). These revised source terms 
are described in terms of radionuclide composition and magnitude, 
physical and chemical form, and timing of release. Where TID-14844 
addressed three categories of radionuclides, the revised source terms 
categorize the accident release into eight groups on the basis of 
similarity in chemical behavior. Where TID-14844 assumed an immediate 
release of the activity, the revised source terms have five release 
phases that are postulated to occur over several hours, with the onset 
of major core damage occurring after 30 minutes. Where TID-14844 
assumed radioiodine to be predominantly elemental, the revised source 
terms assume radioiodine to be predominantly cesium iodide (CsI), an 
aerosol that is more amenable to mitigation mechanisms.
    For DBAs, the NUREG-1465 source terms (up to and including the 
early in-vessel phase) are comparable to the TID-14844 source term with 
regard to the magnitude of the noble gas and radioiodine release 
fractions. However, the revised source terms offer a more 
representative description of the radionuclide composition and release 
timing. The NRC has determined (SECY-94-302, December 19, 1994) that 
design basis analyses will address the first three release phases--
coolant, gap, and in-vessel. The ex-vessel and late in-vessel phases 
are considered to be inappropriate for design basis analysis purposes. 
These latter releases could only result from core damage accidents with 
vessel failure and core-concrete interactions.
    The objective of NUREG-1465 was to define revised accident source 
terms for regulatory application for future light water reactors 
(LWRs). The NRC's intent was to capture the major relevant insights 
available from severe accident research to provide, for regulatory 
purposes, a more realistic portrayal of the amount of the postulated 
accident source term. These source terms were derived from examining a 
set of severe accident sequences for LWRs of current

[[Page 71992]]

design. Because of general similarities in plant and core design 
parameters, these results are considered to be applicable to 
evolutionary and passive LWR designs. The revised source term has been 
used in evaluating the Westinghouse AP600 standard design certification 
application. (A draft version of NUREG-1465 was used in evaluating 
Combustion Engineering's (CE's) System 80+ design.)
    The NRC considered the applicability of the revised source terms to 
operating reactors and determined that the current analytical approach 
based on the TID-14844 source term would continue to be adequate to 
protect public health and safety, and that operating reactors licensed 
under this approach would not be required to reanalyze accidents using 
the revised source terms. The NRC concluded that some licensees may 
wish to use an alternative source term in analyses to support 
operational flexibility and cost-beneficial licensing actions and that 
some of these applications could provide concomitant improvements in 
overall safety and in reduced occupational exposure. The NRC initiated 
several actions to provide a regulatory basis for operating reactors to 
voluntarily amend their facility design bases to enable use of the 
revised source term in design basis analyses. First, the NRC solicited 
ideas on how an alternative source term might be implemented. In 
November 1995, the Nuclear Energy Institute (NEI) submitted its generic 
framework, Electric Power Research Institute Technical Report TR-
105909, ``Generic Framework for Application of Revised Accident Source 
Term to Operating Plants.'' This report and the NRC response were 
discussed in SECY-96-242 (November 25, 1996). Second, the NRC initiated 
an assessment of the overall impact of substituting the NUREG-1465 
source terms for the traditionally used TID-14844 source term at three 
typical facilities. This was done to evaluate the issues involved with 
applying the revised source terms at operating plants. SECY-98-154 
(June 30, 1998) described the conclusions of this assessment. Third, 
the NRC accepted license amendment requests related to implementation 
of the revised source terms at a small number of pilot plants. 
Experience has demonstrated that evaluation of a limited number of 
plant-specific submittals improves regulation and regulatory guidance 
development. The review of these pilot projects is currently in 
progress. Insights from these pilot plant reviews have been 
incorporated into the regulatory guidance that was developed in 
conjunction with this rulemaking. Fourth, the NRC initiated an 
assessment on whether rulemaking would be necessary to allow operating 
reactors to use an alternative source term. This final rule and the 
supporting regulatory guidance have resulted from this assessment.
    This final rulemaking for use of alternative source terms is 
applicable to holders of operating licenses issued prior to January 10, 
1997, under 10 CFR Part 50, ``Domestic Licensing of Production and 
Utilization Facilities,'' and to holders of renewed licenses under 10 
CFR Part 54, ``Requirements for Renewal of Operating Licenses for 
Nuclear Power Plants,'' whose initial operating license was issued 
prior to January 10, 1997. The regulations of Part 50 are supplemented 
by those in other parts of Chapter I of Title 10, including Part 100, 
``Reactor Site Criteria.'' Part 100 contains language that 
qualitatively defines a required accident source term and contains a 
note that discusses the availability of TID-14844. With the exception 
of Sec. 50.34(f), there are no explicit requirements in Chapter I of 
Title 10 to use the TID-14844 accident source term. Section 50.34(f), 
which addresses additional TMI-related requirements, is only applicable 
to a limited number of construction permit applications pending on 
February 16, 1982, and to applications under Part 52.
    An applicant for an operating license is required by Sec. 50.34(b) 
to submit a final safety analysis report (FSAR) that describes the 
facility and its design bases and limits, and presents a safety 
analysis of the structures, systems, and components of the facility as 
a whole. Guidance in performing these analyses is given in regulatory 
guides. In its review of the more recent applications for operating 
licenses, the NRC has used the review procedures in NUREG-0800, 
``Standard Review Plan for the Review of Safety Analysis Reports for 
Nuclear Power Plants'' (SRP). These review procedures reference or 
provide acceptable assumptions and analysis methods. The facility FSAR 
documents the assumptions and methods actually used by the applicant in 
the required safety analyses. The NRC's finding that a license may be 
issued is based on the review of the FSAR, as documented in the 
Commission's safety evaluation report (SER). Fundamental assumptions 
that are design inputs, including the source term, were required to be 
included in the FSAR and became part of the design basis 1 
of the facility. From a regulatory standpoint, the requirement to use 
the TID-14844 source term is expressed as a licensee commitment 
(typically to Regulatory Guide 1.3 or 1.4) documented in the facility 
FSAR, and is subject to the requirements of Sec. 50.59.
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    \1\ As defined in Sec. 50.2, design bases means that information 
which identifies the specific functions to be performed by a 
structure, system, or component of a facility, and the specific 
values or ranges of values chosen for controlling parameters as 
reference bounds for design. These values may be (1) restraints 
derived from generally accepted ``state of the art'' practices for 
achieving functional goals, or (2) requirements derived from 
analysis (based on calculation and/or experiments) of the effects of 
a postulated accident for which a structure, system, or component 
must meet its functional goals. The NRC considers the accident 
source term to be an integral part of the design basis because it 
sets forth specific values (or range of values) for controlling 
parameters that constitute reference bounds for design.
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    In 1996 (61 FR 65175; December 11, 1996), the NRC amended its 
regulations in 10 CFR Parts 21, 50, 52, 54, and 100. That regulatory 
action produced site criteria for future sites, presented a stable 
regulatory basis for seismic and geologic siting and the engineering 
design of future nuclear power plants to withstand seismic events, and 
relocated source term and dose requirements for future plants into Part 
50. Because these dose requirements tend to affect reactor design 
rather than siting, they are more appropriately located in Part 50. 
This decoupling of siting from design is consistent with the future 
licensing of facilities using standardized plant designs, the design 
features of which have been or will be certified in a separate design 
certification rulemakings. This decoupling of siting from design was 
directed by Congress in the 1980 Authorization Act for the NRC. Because 
the revised criteria would not apply to operating reactors, the non-
seismic and seismic reactor site criteria for operating reactors were 
retained as Subpart A and Appendix A to Part 100, respectively. The 
revised reactor site criteria were added as Subpart B in Part 100, and 
revised source term and dose requirements were moved to Sec. 50.34. The 
existing source term and dose requirements of Subpart A of Part 100 
will remain in place as the licensing bases for those operating 
reactors that do not elect to use an alternative source term.
    In relocating the source term and dose requirements for future 
reactors to Sec. 50.34, the NRC retained the requirements for the 
exclusion area and the low population zone, but revised the associated 
numerical dose criteria to replace the two different doses for the 
whole body and the thyroid gland with a single, total effective dose 
equivalent (TEDE) value. The dose criteria for the whole body and the 
thyroid, and the

[[Page 71993]]

immediate 2-hour exposure period were largely predicated by the assumed 
source term being predominantly noble gases and radioiodines 
instantaneously released to the containment and the assumed ``single 
critical organ'' method of modeling the internal dose used at the time 
that Part 100 was originally published. However, the current dose 
criteria, by focusing on doses to the thyroid and the whole body, 
assume that the major contributor to doses will be radioiodine. 
Although this may be appropriate with the TID-14844 source term, as 
implemented by Regulatory Guides 1.3 and 1.4, it may not be true for a 
source term based on a more complete understanding of accident 
sequences and phenomenology.
    The postulated chemical and physical form of radioiodine in the 
revised source terms is more amenable to mitigation and, as such, 
radioiodine may not always be the predominant radionuclide in an 
accident release. The revised source terms include a larger number of 
radionuclides than did the TID-14844 source term as implemented in 
regulatory guidance. The whole body and thyroid dose criteria ignore 
these contributors to dose. The NRC amended its radiation protection 
standards in Part 20 in 1991 (56 FR 23391; May 21, 1991) replacing the 
single, critical organ concept for assessing internal exposure with the 
TEDE concept that assesses the impact of all relevant nuclides upon all 
body organs. TEDE is defined to be the deep dose equivalent (for 
external exposure) plus the committed effective dose equivalent (for 
internal exposure). The deep dose equivalent (DDE) is comparable to the 
present whole body dose; the committed effective dose equivalent (CEDE) 
is the sum of the products of doses (integrated over a 50-year period) 
to selected body organs resulting from the intake of radioactive 
material multiplied by weighting factors for each organ that are 
representative of the radiation risk associated with the particular 
organ.
    The TEDE, using a risk-consistent methodology, assesses the impact 
of all relevant nuclides upon all body organs. Although it is expected 
that in many cases the thyroid could still be the limiting organ and 
radioiodine the limiting radionuclide, this conclusion cannot be 
assured in all potential cases. The revised source terms postulate that 
the core inventory is released in a sequence of phases over 10 hours, 
with the more significant release commencing at about 30 minutes from 
the start of the event. The assumption that the 2-hour exposure period 
starts immediately at the onset of the release is inconsistent with the 
phased release postulated in the revised source terms. The final rule 
adopts the future LWR dose criteria for operating reactors that elect 
to use an alternative source term.
    An accidental release of radioactivity can result in radiation 
exposure to control room operators. Normal ventilation systems may draw 
this activity into the control room where it can result in external and 
internal exposures. Control room designs differ but, in general, design 
features are provided to detect the accident or the activity and 
isolate the normal ventilation intake. Emergency ventilation systems 
are activated to minimize infiltration of contaminated air and to 
remove activity that has entered the control room. Personnel exposures 
can also result from radioactivity outside of the control room. 
However, because of concrete shielding of the control room, these 
latter exposures are generally not limiting. The objective of the 
control room design is to provide a location from which actions can be 
taken to operate the plant under normal conditions and to maintain it 
in a safe condition under accident conditions. General Design Criterion 
19 (GDC-19), ``Control Room,'' of Appendix A to 10 CFR Part 50 (36 FR 
3255; February 20, 1971), establishes minimum requirements for the 
design of the control room, including a requirement for radiation 
protection features adequate to permit access to and occupancy of the 
control room under accident conditions. The GDC-19 criteria were 
established for judging the acceptability of the control room design 
for protecting control room operators under postulated design basis 
accidents, a significant concern being the potential increases in 
offsite doses that might result from the inability of control room 
personnel to adequately respond to the event.
    The GDC-19 criteria are expressed in terms of whole body dose, or 
its equivalent to any organ. The NRC did not revise the criteria when 
Part 20 was amended (56 FR 23391; May 21, 1991) instead deferring such 
action to individual facility licensing actions (NUREG/CR-6204, 
``Questions and Answers Based on the Revised 10 CFR Part 20''). This 
position was taken in the interest of maintaining the licensing basis 
for those facilities already licensed. The NRC is replacing the current 
dose criteria of GDC-19 for future reactors and for operating reactors 
that elect to use an alternative source term with a criterion expressed 
in terms of TEDE. The rationale for this revision is similar to the 
rationale, discussed earlier in this preamble, for revising the dose 
criteria for offsite exposures.
    On January 10, 1997 (61 FR 65157), the NRC amended 10 CFR Parts 21, 
50, 52, 54, and 100 of its regulations to update the criteria used in 
decisions regarding power reactor siting for future nuclear power 
plants. The NRC intended that future licensing applications in 
accordance with Part 52 utilize a source term consistent with the 
source term information in NUREG-1465 and the accident TEDE criteria in 
Parts 50 and 100. However, during the final design approval (FDA) and 
design certification proceeding for the Westinghouse AP600 advanced 
light-water reactor design, the NRC staff and Westinghouse determined 
that exemptions were necessary from Secs. 50.34(f)(2)(vii), (viii), 
(xxvi), and (xxviii) and 10 CFR Part 50, Appendix A, GDC-19. This final 
rule would eliminate the need for these exemptions for future 
applicants under Part 52 by making conforming changes to Part 50, 
Appendix A, GDC-19 and Sec. 50.34.

II. Analysis of Public Comments

    The NRC published a proposed rule in the Federal Register (64 FR 
12117, March 31, 1999); that would provide a regulatory framework for 
the voluntary implementation of alternative source terms as a change to 
the design basis at currently licensed power reactors, while retaining 
the existing regulatory framework for currently licensed power reactor 
licensees who choose not to implement an alternative source term. The 
rule proposed relocating source term and dose requirements that apply 
primarily to plant design into 10 CFR Part 50 for operating reactors 
that choose to implement an alternative source term. The rule also 
proposed conforming changes to Sec. 50.34(f) and Part 50, Appendix A, 
GDC-19 to eliminate the need for exemptions for future applicants under 
Part 52.
    The NRC received seven letters commenting on the proposed rule. All 
comments including those received by the NRC after the expiration of 
the public comment period but before June 25, 1999, were considered. 
The commenters included two State regulatory agencies, two nuclear 
industry groups and three utilities. The State of Florida Department of 
Community Affairs indicated that they had no comments on the proposed 
rule. The State of New Jersey Department of Environmental Protection 
concurred with the NRC's position on the use of an AST in emergency 
preparedness applications and stated a desire to review the draft 
regulatory guidance when issued. Winston & Strawn

[[Page 71994]]

submitted comments on behalf of the Nuclear Utility Backfitting and 
Reform Group (NUBARG). The Nuclear Energy Institute (NEI) submitted 
comments on behalf of the nuclear industry. Two of the utilities 
provided comments, while the third endorsed the comments submitted by 
NEI. Copies of these letters are available for public inspection and 
copying for a fee at the NRC Public Document Room, 2120 L Street NW. 
(Lower Level), Washington, DC.

1. NUBARG Comments

    NUBARG supports the rule, noting that the rule as proposed defines 
an acceptable regulatory process for implementing more realistic 
accident source terms. NUBARG requested clarification in the final rule 
of situations in which an alternative source term (AST) may be applied 
in future backfitting 2 decisions. First, NUBARG suggests 
that the NRC clarify the extent it intends to use the revised source 
term in assessing whether new generic requirements provide a cost-
justified, substantial increase in safety in accordance with NRC's 
backfitting rule, Sec. 50.109. NUBARG believes that continued use of 
the source term in TID-14844 for this purpose in spite of its known 
limitations would be inappropriate and could lead to overly 
conservative estimates of the safety impact of proposed new 
requirements. Second, NUBARG suggests a similar clarification for 
plant-specific backfit decisions for plants that have not opted to 
implement the revised source term. NUBARG believes that the NRC has 
discretion to take all relevant factors into account in making its 
safety benefit assessment of the proposed backfit, including the 
current state of knowledge concerning the accident source term. NUBARG 
suggested that the statements of considerations accompanying the final 
rule address these issues. NUBARG also suggests that relevant NRC 
guidance should also be revised to reflect NRC policy in these areas.
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    \2\ As provided in Sec. 50.109, Backfitting is defined as the 
modification of or addition to systems, structures, components, or 
the design of a facility; or the design approval or manufacturing 
license for a facility; or the procedures or organization required 
to design, construct or operate a facility; any of which may result 
from a new or amended provision in the Commission rules or the 
imposition of a regulatory staff position interpreting the 
Commission rules that is either new or different from a previously 
applicable staff position.
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    NRC Response. When radiological consequence analyses are involved, 
the NRC expects to use a technically appropriate AST in evaluating 
generic and plant-specific backfitting analyses, including those 
proposed for facilities that have not implemented an AST. The NRC 
agrees with the NUBARG position that the NRC has discretion to take all 
new information on accident source terms into account. The NRC's 
guidance for evaluating proposed NRC regulatory actions (including 
backfitting) are contained in NUREG/BR-0058, ``Regulatory Analysis 
Guidelines of the U.S. Nuclear Regulatory Commission,'' and NUREG/BR-
0184, ``Regulatory Analysis Technical Evaluation Handbook.'' These 
documents state that value and impact (including adverse effects on 
health and safety) parameters are to be best estimates, preferably mean 
or expected values. These documents also provide that analyses are to 
be based largely on risk considerations.

2. NEI Comment 1

    NEI stated that the Section-by-Section Analysis in the proposed 
rule notice is consistent with the NRC's intent to permit limited 
application of the new research results. NEI noted that these limited 
applications are of two types: (1) application of alternative source 
term radiological composition and magnitude in a quantitative analysis 
relative to the effect on the performance of a given engineered safety 
feature; or (2) application of only the timing aspects in conjunction 
with the original TID-14844 source term. NEI stated that proposed 
Sec. 50.67 appears to apply to applications where a licensee would use 
a completely new source term such as NUREG-1465 in all aspects of the 
plant design. The NEI comment acknowledged that further guidance in a 
subsequent regulatory guide and standard review plan is helpful and 
necessary. Nonetheless, NEI is concerned that licensee pursuit of 
either of these limited applications might ultimately require seeking 
an exemption to Sec. 50.67, or require extensive analysis. NEI 
recommended that the NRC should: (1) revise the proposed rule language 
to accommodate limited application of an alternative source term as 
done in the Section-By-Section Analysis; (2) provide clarification in 
the Statement of Consideration (SOC) for the rule; and (3) for 
applications that continue to use the TID source term but incorporate 
attributes of newer technical insights such as timing of releases, 
specify that the provisions of the proposed rule do not apply.
    NRC Response. The language of Sec. 50.67(b) requires an evaluation 
of the consequences of applicable design basis accidents. The NRC 
believes that the use of the modifier applicable provides the basis for 
processing selective implementations. Design basis accidents not 
applicable to a particular selective implementation would not be 
required to be evaluated. The NRC expects that the licensee will 
evaluate all applicable impacts of the proposed AST implementation. 
While a selective implementation may result in a reduced scope of 
evaluation, the licensee must still demonstrate that the AST 
implementation and any associated proposed modifications will not 
result in accident conditions exceeding the criteria specified in 
Sec. 50.67. Therefore, these criteria are applicable to full and 
selective implementations alike. The scope of the required re-analyses 
will depend on the specific application proposed by the licensee. 
Guidance with regard to this scope is properly provided in the draft 
regulatory guide prepared for this rule. Therefore, the NRC has decided 
against revising the rule language as suggested by NEI. Consistent with 
the second NEI recommendation, the NRC has modified paragraph D of the 
section-by-section analysis to clarify this issue.

3. NEI Comment 2

    In its second comment, NEI noted that the SOC provides that 
licensees may need to perform additional evaluations of equipment 
qualifications (Sec. 50.49). The SOC should discuss the circumstances 
when such an evaluation may be necessary. NEI recommended that the SOC 
should be amended to state that regardless of source term used, the 
licensee would be required to re-evaluate the equipment qualification 
only when a plant modification alters the plant configuration so that 
the underlying assumptions, with respect to dose distribution and 
effects, are materially altered. NEI summarized conclusions of several 
references in support of its position. NEI stated that there is no 
basis to require or expect additional analyses of equipment 
qualification if a licensee applied the alternative source term in 
limited scope applications, absent a plant configuration change that 
materially alters the dose distribution and effects assumed in existing 
analyses.
    NRC Response. The re-baselining study prepared by the NRC staff 
(SECY-98-154, June 30, 1998) considered the impact of an AST on 
analyses of the postulated integrated radiation doses for plant 
components exposed to containment atmosphere radiation sources and 
those exposed to containment sump radiation sources. The staff's 
conclusions regarding the atmosphere sources are consistent with those 
identified by NEI in its comment. However, the re-baselining study also 
concluded that the increased concentration of cesium in the containment 
sump water could result in

[[Page 71995]]

an increase in the postulated integrated radiation doses for certain 
plant components subject to equipment qualification. It is because of 
this conclusion that the NRC included the discussion in the SOC 
regarding re-evaluation of equipment environmental qualification. The 
NEI comment provides no additional information that would cause the NRC 
to change its position on this matter. Further, the NRC has determined 
that it is necessary to consider the potential impact of the postulated 
cesium concentration in the containment sump water as it applies to all 
operating power reactors, not just to those licensees amending their 
design basis to use an AST. Since the postulated increase in the 
integrated dose occurs only following an accident, there is no adverse 
effect on equipment relied upon to perform safety functions immediately 
following an accident. Rather, this issue affects equipment that is 
required to be operable longer than about 30 days to 4 months after an 
accident. As such, the NRC determined that continued plant operation 
does not pose an immediate threat to public health and safety. Also, 
should such long-term equipment fail there will not be an undue threat 
to public health and safety as protective actions for the public would 
have already been implemented by the time the postulated failure could 
occur. In addition, the time period between the onset of the event and 
the projected failure allows compensatory measures to be taken to 
prevent the equipment failure or to restore the degraded safety 
function. The NRC will evaluate this issue as a generic safety issue to 
determine whether further regulatory actions are justified. The final 
regulatory guide, or subsequent revisions thereto, is expected to 
reflect the resolution of this generic safety issue.

4. NEI Comment 3

    NEI recommends that the definition of Source Term in Sec. 50.2 be 
revised to ``Source term refers to the magnitude and mix of 
radionuclides released from the fuel, their physical and chemical form, 
and the timing of their release.'' NEI stated that the language in the 
proposed rule would prohibit the use of Sec. 50.67 for accidents such 
as the fuel handling accident.
    NRC Response. The NRC agrees with the proposed revision. The 
proposed definition was consistent with the definition of source term 
as used in NUREG-1465, which was written primarily to address loss of 
coolant accidents (LOCA). The regulatory guidance for this rule extends 
the NUREG-1465 source terms to other accidents which involve core 
damage. The definition suggested by NEI is consistent with the proposed 
use of the AST. The Sec. 50.2 definition has been revised in the final 
rule to reflect the change suggested by NEI and that suggested by 
Arizona Public Service Comment 1 below.

5. NEI Comment 4

    NEI stated that the proposed rule does not permit new test reactors 
to use an alternative source term. New test reactors would have to use 
the Part 100 Subpart A, ``Evaluation Factors for Stationary Power 
Reactor Site Applications Before January 10, 1997, and for Testing 
Reactors,'' even though their application for an operating license 
would be filed after January 10, 1997. The use of Section 50.67, 
``Accident Source Term,'' is limited to holders of operating licenses 
issued before January 10, 1997. This wording prohibits new test 
reactors from using the alternative source term. NEI recommended that 
Sec. 50.67 be amended to allow new test reactors to use an alternative 
source term.
    NRC Response. Section 50.67 applies only to holders of licenses for 
operating reactors, including test reactors, whose licenses were issued 
before January 10, 1997. There is no regulatory requirement for a 
specific source term for reactors to be licensed in the future, 
including test reactors. Accordingly, no regulatory action is necessary 
to accommodate the NEI recommendation.

6. Duke Energy Corporation Comment

    Duke Energy Corporation (Duke) endorsed the comments submitted on 
behalf of the industry by NEI. Duke stated that the proposed 
Sec. 50.67(b)(1) was not clear regarding whether licensees will be 
allowed to use a revised source term on a limited basis (e.g., for 
analyses of a specific accident or function), or whether they will be 
required to review the entire radiological consequence analyses to 
apply for the new source term. Duke suggested that necessary guidance 
be provided in the draft regulatory guidance to allow for limited use 
of the new source terms where such use can be justified.
    NRC Response. This comment is similar to NEI Comment 1 addressed 
previously. As stated in the SOC, the NRC will consider justifiable 
limited (i.e., selective) applications of an AST. Although a selective 
implementation may result in a reduced scope of evaluation, the 
licensee must still demonstrate that the AST implementation and any 
associated proposed modifications will not exceed the criteria 
specified in Sec. 50.67. The scope of the required re-analyses will 
depend on the specific application proposed by the licensee. Regulatory 
guidance on selective implements and the scope of required re-analyses 
has been included in the draft guide and are available as announced in 
this Federal Register notice.

7. Arizona Public Service Company Comment 1

    Arizona Public Service Company (APS) noted that the SOC statement, 
``a subsequent change to the source term must be made through a license 
amendment'' could be interpreted as requiring prior NRC approval for 
any change in the magnitude and mix of radionuclides released from the 
reactor core. APS stated that this interpretation could place 
additional restrictions on licensee efforts at economical fuel 
management, including reload design.
    NRC Response. The NRC agrees with the APS comment. The NRC had 
intended the phrase ``magnitude and mix'' to refer to the fractions of 
the fission product inventory of the radionuclides released from the 
reactor fuel. The NRC intent for the provision in question was to 
require approval for changes in the radioactivity release fractions, 
the radionuclides released, their physical and chemical form, and the 
timing of their release. Since ``magnitude and mix'' could be a source 
of confusion, the NRC has modified the Sec. 50.2 definition of Source 
Term in the final rule to read: ``Source term refers to the magnitude 
and mix of the radionuclides released from the fuel, expressed as 
fractions of the fission product inventory in the fuel, as well as 
their physical and chemical form, and the timing of their release.'' 
This is consistent with NUREG-1465 when it refers to ``magnitude and 
mix,'' since the NUREG-1465 presents these data in the form of tables 
of release fractions and radionuclides. This revised language also 
addresses NEI Comment 3 above.

8. Arizona Public Service Company Comment 2

    In its second comment, APS noted that NUREG-1465 contains a 
disclaimer that the accident source terms provided therein may not be 
applicable to fuel irradiated in excess of 40 GWD/MTU. The NRC has 
licensed core designs with fuel irradiations of up to 62 GWD/MTU. APS 
questioned whether the NRC staff was going to address the affect of 
high burnups on a generic basis, or on a facility-by-facility basis.
    NRC Response. The AST tabulated in the draft regulatory guidance, 
which

[[Page 71996]]

differs in some aspects from that provided in NUREG-1465, is applicable 
to peak rod average irradiations up to 62 GWD/MTU. Attachment 1 to the 
regulatory analysis for this rulemaking describes the bases of this 
extension in fuel irradiation as it applies to the AST. There are some 
facility-by-facility considerations. For example, the increase in core 
inventory for some long-lived radionuclides and the change in isotopic 
mix due to the increase in plutonium fission as the fuel ages is 
addressed by the Draft Guide-1081 provision that licensees re-analyze 
the core inventory based on current operating parameters, including 
fuel burnup.

III. Section-by-Section Analysis

A. Section 50.2

    The general ``definitions'' section for Part 50 is supplemented by 
adding a definition of source term for the purpose of Sec. 50.67. In 
NUREG-1465, the source term is defined by five projected 
characteristics: (1) magnitude of radioactivity release, (2) 
radionuclides released, (3) physical form of the radionuclides 
released, (4) chemical form of the radionuclides released, and (5) 
timing of the radioactivity release. The definition of source term in 
Sec. 50.2 embodies the NUREG-1465 definition; however, the Sec. 50.2 
definition includes the clarifying phrase, ``expressed as fractions of 
the fission product inventory in the fuel,'' (see prior response to 
Arizona Public Service Comment 1). Although all five characteristics 
should be addressed in applications proposing the use of an alternative 
source term, there may be technically justifiable applications in which 
all five characteristics need not be addressed. The NRC intends to 
allow licensees flexibility in implementing alternative source terms 
consistent with maintaining a conservative, clear, logical, and 
consistent plant design basis. The regulatory guidance that supports 
this final rule describes an acceptable basis for defining the 
characteristics of an alternative source term.

B. Section 50.67(a)

    This paragraph defines the licensees that may seek to revise their 
current radiological source term with an alternative source term. The 
final rule is applicable to holders of operating licenses that were 
issued under 10 CFR Part 50 before January 10, 1997, and to holders of 
renewed licenses issued under 10 CFR Part 54 whose initial operating 
license was issued prior to January 10, 1997. The final rule does not 
require licensees to revise their current source term. The NRC 
considered the acceptability of the TID-14844 source term at current 
operating reactors and determined that the analytical approach based on 
the TID-14844 source term would continue to be adequate to protect 
public health and safety, and that operating reactors licensed under 
this approach should not be required to reanalyze design basis 
accidents using a new source term. The final rule does not explicitly 
define an alternative source term. In lieu of an explicit reference to 
NUREG-1465, Footnote 1 to the final rule identifies the significant 
attributes of an accident source term. The regulatory guidance that is 
being issued to support this final rule will identify ASTs (based on 
the NUREG-1465 source terms) that are acceptable alternatives to the 
source term in TID-14844, and will provide implementation guidance. 
This approach will provide for future revised source terms if they are 
developed and will allow licensees to propose additional alternatives 
for NRC consideration.

C. Section 50.67(b)(1)

    This paragraph of Sec. 50.67 identifies the information that a 
licensee must submit as part of a license amendment application to use 
an alternative source term. Because of the extensive use of the 
accident source term in the design and operation of a power reactor and 
the potential impact on postulated accident consequences and margins of 
safety of a change of such a fundamental design assumption, the NRC has 
determined that any change to the design basis to use an alternative 
source term should be reviewed and approved by the NRC in the form of a 
license amendment. Changes to the source term, by itself, would 
ordinarily constitute a no significant hazards consideration. In 
addition, generic analyses performed by the NRC staff in support of 
this final rule have indicated that there are potential changes to the 
facility as documented in the FSAR that will constitute a no 
significant hazards consideration. However, these determinations will 
have to be made for each proposed change based upon facility-specific 
evaluations. The procedural requirements for processing a license 
amendment are presented in Secs. 50.90 through 50.92.
    The NRC's regulations provide a regulatory mechanism for a licensee 
to effect a change in its design basis in Sec. 50.59 3 that 
allows a licensee to make changes to the facility as described in the 
final safety analysis report (FSAR) without prior NRC approval, if the 
proposed change meets certain criteria specified in Sec. 50.59. If the 
criteria are not met, the licensee must request NRC approval of the 
change using the license amendment process detailed in Sec. 50.90. 
Significant to this final rule is the criterion that NRC review is 
required if the proposed change would result in a greater than minimal 
increase in consequences of an accident or malfunction. In many 
applications, alternative source terms may reduce the postulated 
consequences of the accident or malfunction. For this reason, the NRC 
determined that the regulatory framework of Sec. 50.59 might not 
provide assurance that this change in the design basis would be 
recognized by the licensee as needing review by the NRC staff.
---------------------------------------------------------------------------

    \3\ Section 10 CFR 50.59 is being amended in a parallel, but 
separate, rulemaking action. That rulemaking, when implemented is 
expected to replace the unreviewed safety question (USQ) concept. 
Further, the criteria for consequences are being revised from ``may 
be increased'' to ``result in more than a minimal increase.'' Those 
changes are not expected to invalidate the conclusions drawn in this 
analysis.
---------------------------------------------------------------------------

    After a licensee has been authorized to substitute an alternative 
source term in its design basis, subsequent changes to the facility 
that involve an alternative source term may be processed under 
Sec. 50.59 or Sec. 50.90, as appropriate. However, a subsequent change 
to the fractions of the fission product inventory of the radionuclides 
released from the reactor fuel, their chemical and physical form, or 
the timing of their release as tabulated in the regulatory guidance 
(with deviations proposed by the licensee and approved by the NRC) 
could not be implemented under Sec. 50.59. This provision applies only 
to these tabulated parameters.
    The final rule will require the applicant to perform analyses of 
the consequences of applicable design basis accidents previously 
analyzed in the safety analysis report and to submit a description of 
the analysis inputs, assumptions, methodology, and results of these 
analyses for NRC review. Applicable evaluations may include, but are 
not limited to, those previously performed to show compliance with 
Sec. 100.11, Sec. 50.49, Part 50 Appendix A GDC-19, Sec. 50.34(f), and 
NUREG-0737, ``Clarification of TMI Action Plan Requirements,'' 
requirements II.B.2, II.B.3, III.D.3.4. The regulatory guidance that 
supports this final rule will provide guidance on the scope and extent 
of analyses used to show compliance with this rule and on the 
assumptions and methods used therein. It is not the NRC's intent that 
all of the design basis radiological analyses for a facility be

[[Page 71997]]

performed again as a prerequisite for approval of the use of an 
alternative source term. Nor is it the NRC's intent that EAB, LPZ, and 
control room dose calculations be performed for all applications under 
Sec. 50.67. The NRC does expect that the applicant will perform 
sufficient evaluations, supported by calculations as warranted, to 
demonstrate the acceptability of the proposed amendment.

D. Sections 50.67(b)(2)(i),(ii), (iii)

    These subparagraphs contain the three criteria for NRC approval of 
the license amendment to use an alternative source term. A detailed 
rationale for the use of 0.25 Sv (25 rem) TEDE as an accident dose 
criterion and the use of the 2-hour exposure period resulting in the 
maximum dose for future LWRs is provided at 61 FR 65157 (December 11, 
1996). The same considerations that formed the basis for that rationale 
are similarly applicable to operating reactors that elect to use an 
alternative source term. The NRC believes that it is technically 
appropriate and logical to extend the philosophy of decoupling of 
design and siting, and the dose criteria established for future LWRs to 
operating reactors that elect to use an alternative source term.
    The NRC is replacing the current GDC-19 dose criteria for operating 
reactors that elect to use an alternative source term with a criterion 
of 0.05 Sv (5 rem) TEDE for the duration of the accident. This 
criterion is included in Sec. 50.67 as well as in GDC-19 in order to 
co-locate all of the dose requirements associated with alternative 
source terms. The bases for the NRC's decision are: first, that the 
criteria in GDC-19 and that in the final rule are based on a primary 
occupational exposure limit. Second, the language in GDC-19: ``5 rem 
whole body, or its equivalent to any part of the body'' is subsumed by 
the definition of TEDE in Sec. 20.1003 and by the 0.05 Sv (5 rem) TEDE 
annual limit in Sec. 20.1201(a). Although the weighting factors stated 
in Sec. 20.1003 for use in determining TEDE differ in magnitude from 
the weighting factors implied in the 0.3 Sv (30 rem) thyroid criteria 
used for showing compliance with GDC-19, these differences are the 
result of improvement in the science of assessing internal exposures 
and do not represent a reduction in the level of protection. Third, as 
discussed earlier, the use of TEDE in conjunction with alternative 
source terms has been deemed appropriate and necessary. Fourth, the use 
of TEDE for the control room dose criterion is consistent with the use 
of TEDE in the accident dose criteria for offsite exposure.
    The NRC has not included a ``capping'' limitation, an additional 
requirement that the dose to any individual organ not be in excess of 
some fraction of the total as provided for routine occupational 
exposures. The bases for the NRC's decision are: first, that this non-
inclusion of a ``capping'' limitation is consistent with the final rule 
published in December 11, 1996 (61 FR 65157), with regard to doses to 
persons offsite. Second, the use of 0.05 Sv (5 rem) TEDE as the control 
room criterion does not imply that this would be an acceptable exposure 
during emergency conditions, or that other radiation protection 
standards of Part 20, including individual organ dose limits, might not 
apply. This criterion is provided only to assess the acceptability of 
design provisions for protecting control room operators under 
postulated DBA conditions. The DBA conditions assumed in these 
analyses, although credible, generally do not represent actual accident 
sequences but are specified as conservative surrogates to create 
bounding conditions for assessing the acceptability of engineered 
safety features. Third, Sec. 20.1206 permits a once-in-a-lifetime 
planned special dose of five times the annual dose limits. Also, 
Environmental Protection Agency (EPA) guidance sets a limit of five 
times the annual dose limits for workers performing emergency services 
such as lifesaving or protection of large populations.
    Considering the individual organ weighting factors of Sec. 20.1003 
and assuming that only the exposure from a single organ contributed to 
TEDE, the organ dose, although exceeding the dose specified in 
Sec. 20.1201(a), would be less than that considered acceptable as a 
planned special dose or as an emergency worker dose. The NRC is not 
suggesting that control room dose during an accident can be treated as 
a planned special exposure or that the EPA emergency worker dose limits 
are an alternative to GDC-19 or the final rule. However, the NRC does 
believe that these provisions offer a useful perspective that supports 
the conclusion that the organ doses implied by the 0.05 Sv (5 rem) 
criterion can be considered to be acceptable due to the relatively low 
probability of the events that could result in doses of this magnitude.
    Although the dose criteria in the final rule supersede the dose 
criteria in GDC-19, the other provisions of GDC-19 remain applicable.
    There may be technically justifiable implementations of an AST that 
would not require calculation of the EAB, LPZ, or control room doses. 
For example, a proposed modification to change the closure time of a 
containment isolation valve from 2 seconds to 5 seconds may be based on 
the timing insights of the AST. Although a specific calculation might 
not be necessary in this case, the licensee is still required to affirm 
with reasonable assurance that the doses would comply with these stated 
criteria.

E. 10 CFR Part 50, Appendix A, GDC-19

    GDC-19 is changed to include the TEDE dose criterion for control 
room design for applicants for construction permits, design 
certifications, and combined licenses that submitted applications after 
January 10, 1997 (the effective date of the 1996 rulemaking adopting 
the TEDE criterion), and for those licenses using an alternative source 
term under Sec. 50.67. The change to GDC-19 addresses the use of 
alternative source terms at operating reactors and a deficiency 
identified in the regulatory framework for early site permits, standard 
design certifications, and combined licenses under Part 52. Sections 
52.18, 52.48, and 52.81 establish that applications filed under Part 
52, Subparts A, B, and C, respectively, will be reviewed according to 
the standards given in 10 CFR Parts 20, 50, 51, 55, 73, and 100 to the 
extent that those standards are technically relevant to the proposed 
design. Therefore, GDC-19 is pertinent to applications under Part 52.
    The final rule that became effective on January 10, 1997 (61 FR 
65157; December 11, 1996), established accident TEDE criteria (in 
Sec. 50.34) for applicants under Part 52 but did not change the 
existing control room whole body (or equivalent) dose criterion in GDC-
19. Thus, exemptions from the dose criteria in the current GDC-19 were 
necessary in the design certification process for the Westinghouse 
AP600 advanced LWR in order to use the 0.05 Sv (5 rem) TEDE criterion 
deemed necessary for use with alternative source terms. Exemptions will 
arguably be necessary for future applicants for construction permits, 
design certifications, and combined licenses. This amendment will 
eliminate the need for these exemptions.

F. Sections 21.3, 50.2, 50.49(b)(1)(i)(C), 50.65(b)(1), and 
54.4(a)(1)(iii)

    These sections are revised to conform with the relocation of 
accident dose criteria from Sec. 100.11 to Sec. 50.67 for operating 
reactors that have amended their design bases to use an alternative 
source term.

[[Page 71998]]

G. Section 50.34

    A new footnote to Sec. 50.34 has been added to define what 
constitutes an accident source term. This new footnote is identical to 
the existing footnote 1 to Sec. 100.11, and was added to provide for 
consistency between Parts 50 and 100.

H. Sections 50.34(f)(2)(vii), (viii), (xxvi) and (xxviii)

    These paragraphs are revised to replace an explicit reference to 
the ``TID-14844 source term'' with a more general reference to 
``accident source term.'' These changes potentially affect three 
classes of applicants. The first affected class is comprised of 
applicants for design certification under Part 52, Subpart B. Section 
52.47(a)(1)(ii) states that applications for combined licenses must 
contain, inter alia, ``demonstration of compliance with any 
technically-relevant portions of the Three Mile Island requirements set 
forth in Sec. 50.34(f).'' Section 50.34(f) contains several references 
to the TID-14844 source term. These references were modified to delete 
the reference to TID-14844. This change makes it clear that applicants 
for combined licenses should not use the TID-14844 source term but 
should use the source term in the referenced design certification, or a 
source term that is justified in the combined license application. The 
second affected class is comprised of applicants for combined licenses 
under Part 52, Subpart C. Section 52.79(b) makes the requirements of 
52.47(a)(1)(i) applicable if a certified design is not referenced. 
Thus, the combined license applicant is also subject to the 
requirements of Section 50.34(f).
    The third affected class is the small subset of plants that had 
construction permits pending on February 16, 1982. With the proposed 
change, these plants could use either the TID-14844 source term or an 
alternative source term in their operating license applications.

IV. Draft Regulatory Guide; Issuance, Availability

    The Nuclear Regulatory Commission is issuing for public comment a 
draft of a guide planned for its Regulatory Guide Series. This series 
has been developed to describe and make available to the public 
information such as methods acceptable to the NRC staff for 
implementing specific parts of the Commission's regulations, techniques 
used by the staff in evaluating specific problems or postulated 
accidents, and data needed by the NRC staff in its review of 
applications for permits and licenses. Copies of the draft guide may be 
obtained as described in Section VI, ``Referenced Documents,'' of these 
statements of consideration. You may also download copies from the 
NRC's interactive rulemaking forum website through the NRC home page 
(http://ruleforum.llnl.gov/cgi-bin/rulemake).
    The draft guide, temporarily identified by its task number DG-1081 
(which should be mentioned in all correspondence concerning this draft 
guide) is titled ``Alternative Radiological Source Terms for Evaluating 
Design Basis Accidents at Nuclear Power Reactors.'' This guide is 
intended for Division 1, ``Power Reactors.'' This draft guide is being 
developed to provide regulatory guidance on the implementation of an 
alternative source term at an operating reactor. The guide addresses 
issues involving limited or selective implementation of an alternative 
source term and probabilistic risk assessment (PRA) issues related to 
plant modifications based on an alternative source term, and provides 
guidance on the scope and extent of affected design basis accident 
(DBA) radiological analyses and associated acceptance criteria. The 
guide includes revised assumptions and methods for each affected DBA in 
a series of appendices. These appendices supersede the guidance in 
Regulatory Guides 1.3, 1.4, 1.5, 1.25, and 1.77, and supplement 
guidance in Regulatory Guide 1.89 for those facilities using an 
alternative source term.
    The draft guide has not received complete NRC staff review and does 
not represent an official NRC staff position.
    Previous draft versions of DG-1081 have been made publicly 
available to support technical interactions with the public. This 
Federal Register announcement provides an opportunity for the public to 
provide comments on the DG-1081 guidance. The NRC staff will consider 
the public comments in its efforts to finalize the regulatory guidance.
    The Commission invites advice and recommendations on the content of 
the draft regulatory guide. Comments and suggestion are particularly 
requested on the following questions.

A. Scope of Implementation

    1. The guidance provided in the draft regulatory guide is intended 
to allow licensees the maximum flexibility in pursuing technically 
justifiable AST implementations provided that a clear, consistent, and 
logical design basis is maintained. Comments are specifically requested 
on the following questions.
    A. Does the proposed guidance provide the desired flexibility while 
providing reasonable assurance that a clear, consistent, and logical 
design basis will be maintained?
    B. Is there a less complex alternative approach that would provide 
the desired flexibility while maintaining a clear, consistent, and 
logical design basis?
    C. Should the Commission allow licensees that have received 
approval for a selective implementation to extend the AST and the TEDE 
criteria to other design basis applications (that do not involve 
reanalysis of the DBA LOCA) under Sec. 50.59 rather than under 
Sec. 50.67 as currently proposed?
    2. The guidance would allow selective implementation of the 
characteristics (i.e., the fractions of fission product inventory of 
the radionuclides released from the reactor fuel, their chemical and 
physical form, and the timing of their release) of an AST. The 
Commission believes that implementations based only on the timing 
insights of an AST may be technically justifiable. The Commission 
believes that the other combinations may be internally inconsistent. 
Comments are specifically requested on the following questions.
    A. What other combinations of AST characteristics are technically 
consistent?
    B. What plant modifications might be based on these combinations?

B. Scope of Re-Analyses

    1. The draft regulatory guide provides guidance on the scope of the 
re-analyses that should be performed to support an AST implementation. 
Comments are requested on the following questions.
    A. Is the proposed guidance on the scope of re-analyses technically 
appropriate and clear? How could it be improved?
    B. The guidance allows licensees to disposition certain impacts of 
an AST on the basis of the NRC staff's re-baselining study. Does this 
study or other documents provide a sufficient basis for the Commission 
to generically disposition these impacts?
    2. It may be possible for licensees to demonstrate that the doses 
from certain affected analyses assessed using the prior source term and 
dose methodology would be greater than the doses obtained using a 
proposed AST and the TEDE methodology. The proposed guidance would 
allow the licensee to disposition these affected analyses without re-
calculation. Nonetheless, the design basis would now include the 
approved AST and TEDE criteria. The guidance in the draft regulatory 
guide would require the licensee to update the calculation to be 
consistent with the approved AST and dose methodology described in the 
facility design basis in

[[Page 71999]]

the event of a subsequent re-calculation. Comments are requested on the 
following questions.
    A. Should the Commission allow licensees to continue to use the 
prior source term and dose criteria for these analyses and not require 
that they be updated on subsequent revisions?
    B. If the analyses are not updated, how will licensees assure that 
the earlier conclusion that the analyses are limiting remains valid 
following subsequent revisions?
    3. Analyses of the integrated radiation doses for environmental 
qualification of certain equipment important to safety will be affected 
by the increased concentration of radioactive cesium in the containment 
sump water. The Commission has been considering the position that 
licensees proposing to implement an AST must address all impacts of the 
proposed implementation, including the impact of the increased cesium 
concentration. However, the Commission now believes it may be necessary 
for all operating power reactors to address the postulated increase in 
the cesium concentration. The Commission will consider this issue as a 
generic safety issue. Comments are requested on the following 
questions.
    A. Is there information that should be considered by the Commission 
in resolving this generic issue?
    B. If the Commission should conclude that there is safety 
significance but that the costs of implementing corrective actions are 
not justified on a generic basis, should licensees who are voluntarily 
proposing to amend their design basis to use an AST be required to 
address the impact of the increased cesium concentration?
    C. If a licensee proposes a change in the plant configuration that 
would result in an increase in the integrated dose for one or more 
components and this licensee is also proposing, or has already 
implemented an AST, should the re-analysis of the integrated dose be 
based on that AST or on the prior TID14844 source term?
    Comments may be accompanied by relevant information or supporting 
data. Written comments may be mailed to: Secretary, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemakings and Adjudications Staff. Mail Stop O16C1. Copies of 
comments received may be examined at the NRC Public Document Room, 2120 
L Street NW., Washington, DC. Comments will be most helpful if received 
by March 7, 2000.
    You may also provide comments via the NRC's interactive rulemaking 
website through the NRC home page (http://ruleforum.llnl.gov/cgi-bin/
rulemake). This site provides the availability to upload comments as 
files (any format), if your web browser supports that function. For 
information about the interactive rulemaking website, contact Ms. Carol 
Gallagher, (301) 415-5905; or by internet electronic mail to 
[email protected]. For information about the draft guide, contact Mr. Stephen 
F. LaVie, (301) 415-1081; Internet electronic mail [email protected].
    Although a time limit is given for comments on this draft guide, 
comments and suggestions in connection with items for inclusion in 
guides currently being developed or improvements in all published 
guides are encouraged at any time.

V. Draft Standard Review Plan Section; Issuance, Availability

    The Nuclear Regulatory Commission is issuing for public comment a 
draft of a new section to NUREG-0800, ``Standard Review Plan.'' 
Standard review plan (SRP) sections are prepared for the guidance of 
the Office of Nuclear Reactor Regulation staff responsible for the 
review of applications to construct and operate nuclear power plants. 
These documents are made available to the public as part of the 
Commission's policy to inform the nuclear industry and the general 
public of regulatory procedures and policies. The draft SRP Section 
15.0.1, is titled ``Radiological Consequence Analyses Using Alternative 
Source Terms.'' The SRP section complements draft regulatory guide DG-
1081. The draft SRP section has not received complete NRC staff review 
and does not represent an official NRC staff position.
    Copies of the draft SRP section may be obtained as described in 
Section VI, ``Referenced Documents,'' of these statements of 
consideration. You may also download copies from the NRC's interactive 
rulemaking forum website through the NRC home page (http://
ruleforum.llnl.gov/cgi-bin/rulemake).
    Comments on the content of the draft SRP section are invited. 
Comments may be accompanied by relevant information or supporting data. 
Comments should be submitted as described above for the draft 
regulatory guide. Although a time limit is given for comments on this 
draft SRP section, comments and suggestions in connection with items 
for inclusion in SRP sections currently being developed or improvements 
in all published SRP sections are encouraged at any time.

VI. Referenced Documents

    Copies of NUREG-0737, NUREG-0800, NUREG-1465, NUREG/BR-0058, NUREG/
BR-184, and NUREG/CR-6204 may be purchased from the Superintendent of 
Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington, 
DC 20402-9328. Copies also are available from the National Technical 
Information Service, 5285 Port Royal Road, Springfield, VA 22161. A 
copy also is available for inspection and copying for a fee in the NRC 
Public Document Room, 2120 L Street, NW (Lower Level), Washington, DC.
    Single copies of regulatory guides, both active and draft may be 
obtained free of charge by writing the Reproduction and Distribution 
Services Section, OCIO, USNRC, Washington DC 20555-0001, or by fax to 
(301) 415-2289, or by email to [email protected]. Active guides may 
also be purchased from the National Technical Information Service on a 
standing order basis. Details of this service may be obtained by 
writing NTIS, 5285 Port Royal Road, Springfield, VA 22161. Copies of 
active and draft guides are available for inspection or copying for a 
fee from the NRC Public Document Room at 2120 L Street NW., Washington 
DC.
    Copies of SECY-94-302, SECY-96-242, SECY-98-154, SECY-98-289, TID-
14844, and TR-105909 are available for inspection and copying for a fee 
at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), 
Washington, DC.

VII. Finding of No Significant Environmental Impact: Availability

    The NRC has determined under the National Environmental Policy Act 
of 1969, as amended, and the NRC's regulations in Subpart A of 10 CFR 
Part 51, that this regulation is not a major Federal action 
significantly affecting the quality of the human environment and, 
therefore, an environmental impact statement is not required. This 
final rule allows operating reactors to replace the traditional TID-
14844 source term with a more realistic source term based on the 
insights gained from extensive accident research activities. The actual 
accident sequence and progression are not changed; it is the regulatory 
assumptions regarding the accident that would be affected by the 
change. The use of an alternative source term alone cannot increase the 
core damage frequency (CDF) or the large early release frequency (LERF) 
or actual offsite or onsite radiation doses. An alternative source term 
could be used to justify changes in the plant design that might have an 
impact on CDF or LERF or that might increase offsite or onsite doses. 
Those plant changes that do not

[[Page 72000]]

require prior NRC review and approval pursuant to Sec. 50.59 are not 
likely to involve any significant increase in environmental impacts. 
The Sec. 50.59 criteria are sufficiently stringent that any potential 
change in plant design that could have an adverse environmental impact 
in all likelihood could not be made by the licensee without prior NRC 
review and approval. Every plant change that requires NRC review and 
approval under Sec. 50.59 requires a license amendment and, therefore, 
the preparation of an environmental assessment to determine whether the 
proposed change involves any significant environmental impact. Thus, 
this final rule, by itself, will not result in plant changes that 
involve any significant increase in environmental impacts. The final 
rule does not affect non-radiological plant effluents.
    The NRC requested public comments on any environmental justice 
considerations that may be related to this rule. No public comments 
relevant to the draft environmental assessment or environmental justice 
considerations were received. The NRC requested the views of the States 
on the environmental assessment for this rule. No comments relevant to 
the draft environmental assessment or environmental justice 
considerations were received.
    The environmental assessment and finding of no significant impact 
on which this determination is based are available for inspection at 
the NRC Public Document Room, 2120 L Street NW. (Lower Level), 
Washington, DC. Single copies of the environmental assessment and 
finding of no significant impact are available from Mr. Stephen F. 
LaVie, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory 
NRC, Washington, DC 20555-0001, telephone: (301) 415-1081, or by 
Internet electronic mail to [email protected].

VIII. Paperwork Reduction Act Statement

    This final rule increases the burden on licensees by requiring that 
when seeking to revise their current accident source term in design 
basis radiological consequence analyses, they apply for an amendment 
under Sec. 50.90. The public burden for this information collection is 
estimated to average 609 hours per request. Because the burden for this 
information collection is insignificant relative to the total burden 
estimated, Office of Management and Budget (OMB) clearance is not 
required. Existing requirements were approved by the Office of 
Management and Budget, approval number 3150-0011.

Public Protection Notification

    If an information collection does not display a currently valid OMB 
control number, the NRC may not conduct or sponsor, and a person is not 
required to respond to, the information collection.

IX. Regulatory Analysis

    The Commission has prepared a regulatory analysis on this 
regulation. Interested persons may examine a copy of the regulatory 
analysis at the NRC Public Document Room, 2120 L Street NW. (Lower 
Level), Washington, DC. Single copies of the analysis are available 
from Mr. Stephen F. LaVie, Office of Nuclear Reactor Regulation, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 
(301) 415-1081, or by Internet electronic mail to [email protected].

X. Regulatory Flexibility Act Certification

    As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 
605(b), the Commission certifies that this regulation will not have a 
significant economic impact on a substantial number of small entities. 
This regulation will affect only the licensing and operation of nuclear 
power plants. The companies that own these plants do not fall within 
the definition of ``small entities'' found in the Regulatory 
Flexibility Act or within the size standards established by the NRC 
(April 11, 1995; 60 FR 18344).

XI. Backfit Analysis

    The NRC has determined that the backfit rule in 10 CFR 50.109 does 
not apply to this final rule, and that a backfit analysis is not 
required for this rulemaking because these amendments do not involve 
any provisions that would impose backfits as defined in 10 CFR 
50.109(a)(1). This final rule amends the NRC's regulations by 
establishing alternate requirements that may be voluntarily adopted by 
licensees, and makes changes to the regulations to conform them to a 
1996 rulemaking.

XII. Small Business Regulatory Enforcement Fairness Act

    In accordance with the Small Business Regulatory Fairness Act of 
1996, the NRC has determined that this action is not a major rule and 
has verified this determination with the Office of Information and 
Regulatory Affairs, Office of Management and Budget.

XIII. National Technology Transfer and Advancement Act

    The National Technology Transfer Act of 1995, Pub. L. 104-113, 
requires that Federal agencies use technical standards that are 
developed or adopted by voluntary consensus standards bodies unless the 
use of such a standard is inconsistent with applicable law or otherwise 
impractical. In this final rule the NRC is establishing a government-
unique standard in Section 50.67(b)(2) by specifying accident radiation 
dose criteria. These criteria were issued for use by future license 
applicants by an earlier rulemaking (61 FR 65157, December 11, 1996) 
and, by this final rule, are being applied to operating reactors that 
voluntarily use an alternative source term. No voluntary consensus 
standard has been identified that could be used instead of the 
government-unique standard.

List of Subjects

10 CFR Part 21

    Nuclear power plants and reactors, Penalties, Radiation protection, 
Reporting and recordkeeping requirements.

10 CFR Part 50

    Antitrust, Classified information, Criminal penalties, Fire 
protection, Intergovernmental relations, Nuclear power plants and 
reactors, Radiation protection, Reactor siting criteria, Reporting and 
recordkeeping requirements.

10 CFR Part 54

    Administrative practice and procedure, Age-related degradation, 
Backfitting, Classified information, Criminal penalties, Environmental 
protection, Nuclear power plants and reactors, Reporting and 
recordkeeping requirements.

    For the reasons noted in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 553; the NRC is proposing the 
following amendments to 10 CFR Parts 21, 50, and 54:

PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE

    1. The authority citation for Part 21 continues to read as follows:

    Authority: Sec. 161, 68 Stat. 948, as amended, sec. 234, 83 
Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2953 (42 U.S.C. 
2201, 2282, 2297f); secs. 201, as amended, 206, 88 Stat. 1242, as 
amended, 1246 (42 U.S.C. 5841, 5846).

    Section 21.2 also issued under secs. 135, 141, Pub. L. 97-425, 
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).

    2. Section 21.3 is amended by republishing the introductory text 
and revising paragraph (1)(i)(C) of the

[[Page 72001]]

definition of Basic Component to read as follows:


Sec. 21.3  Definitions.

    As used in this part:
    Basic component. (1)(i) * * *
    (C) The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures comparable 
to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or 
Sec. 100.11 of this chapter, as applicable.
* * * * *

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    3. The authority citation for Part 50 continues to read as follows:

    Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).

    Section 50.7 also issued under Pub. L. 95-9601, sec. 10, 92 
Stat. 2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 
101, 185, 68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, 
Pub. L. 91-9190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 
50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as 
amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 
also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 
50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 
91-9190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 
also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 
50.58, 50.91, and 50.92 also issued under Pub. L. 97-9415, 96 Stat. 
2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 
Stat. 939 (42 U.S.C. 2152). Sections 50.80-50.81 also issued under 
sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also 
issued under sec. 187, 68 Stat. 955 (42 U.S.C 2237).

    4. Section 50.2 is amended by republishing the introductory text 
and revising paragraph (1)(iii) of the definition of Basic component, 
and by adding in alphabetical order the definition for Source term to 
read as follows:


Sec. 50.2  Definitions.

    As used in this part,
* * * * *
    Basic component * * *
    (1) * * *
    (iii) The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures comparable 
to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or 
Sec. 100.11 of this chapter, as applicable.
* * * * *
    Source term refers to the magnitude and mix of the radionuclides 
released from the fuel, expressed as fractions of the fission product 
inventory in the fuel, as well as their physical and chemical form, and 
the timing of their release.
* * * * *
    5. Section 50.34 is amended by revising paragraphs (f)(2)(vii), 
(viii), (xxvi), and (xxviii) to read as follows:


Sec. 50.34  Contents of applications; technical information.

* * * * *
    (f) * * *
    (2) * * *
    (vii) Perform radiation and shielding design reviews of spaces 
around systems that may, as a result of an accident, contain accident 
source term \11\ radioactive materials, and design as necessary to 
permit adequate access to important areas and to protect safety 
equipment from the radiation environment. (II.B.2)
    (viii) Provide a capability to promptly obtain and analyze samples 
from the reactor coolant system and containment that may contain 
accident source term \11\ radioactive materials without radiation 
exposures to any individual exceeding 5 rems to the whole body or 50 
rems to the extremities. Materials to be analyzed and quantified 
include certain radionuclides that are indicators of the degree of core 
damage (e.g., noble gases, radioiodines and cesiums, and nonvolatile 
isotopes), hydrogen in the containment atmosphere, dissolved gases, 
chloride, and boron concentrations. (II.B.3)
* * * * *
    (xxvi) Provide for leakage control and detection in the design of 
systems outside containment that contain (or might contain) accident 
source term \11\ radioactive materials following an accident. 
Applicants shall submit a leakage control program, including an initial 
test program, a schedule for re-testing these systems, and the actions 
to be taken for minimizing leakage from such systems. The goal is to 
minimize potential exposures to workers and public, and to provide 
reasonable assurance that excessive leakage will not prevent the use of 
systems needed in an emergency. (III.D.1.1)
* * * * *
    (xxviii) Evaluate potential pathways for radioactivity and 
radiation that may lead to control room habitability problems under 
accident conditions resulting in an accident source term \11\ release, 
and make necessary design provisions to preclude such problems. 
(III.D.3.4)
---------------------------------------------------------------------------

    \11\ The fission product release assumed for these calculations 
should be based upon a major accident, hypothesized for purposes of 
site analysis or postulated from considerations of possible 
accidental events, that would result in potential hazards not 
exceeded by those from any accident considered credible. Such 
accidents have generally been assumed to result in substantial 
meltdown of the core with subsequent release of appreciable 
quantities of fission products.
---------------------------------------------------------------------------

* * * * *
    6. Section 50.49 is amended by revising paragraph (b)(1)(i)(C) to 
read as follows:


Sec. 50.49  Environmental qualification of electric equipment important 
to safety for nuclear power plants.

* * * * *
    (b) * * *
    (1) * * *
    (i) * * *
    (C) The capability to prevent or mitigate the consequences of 
accidents that could result in potential offsite exposures comparable 
to the guidelines in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or Sec. 100.11 
of this chapter, as applicable.
* * * * *
    7. Section 50.65 is amended by revising paragraph (b)(1) to read as 
follows:


Sec. 50.65  Requirements for monitoring the effectiveness of 
maintenance at nuclear power plants.

* * * * *
    (b) * * *
    (1) Safety-related structures, systems and components that are 
relied upon to remain functional during and following design basis 
events to ensure the integrity of the reactor coolant pressure 
boundary, the capability to shut down the reactor and maintain it in a 
safe shutdown condition, or the capability to prevent or mitigate the 
consequences of accidents that could result in potential offsite 
exposure comparable to the guidelines in Sec. 50.34(a)(1), 
Sec. 50.67(b)(2), or Sec. 100.11 of this chapter, as applicable.
* * * * *
    8. Part 50 is amended by adding Sec. 50.67 to read as follows:


Sec. 50.67  Accident source term.

    (a) Applicability. The requirements of this section apply to all 
holders of operating licenses issued prior to January 10, 1997, and 
holders of renewed licenses under part 54 of this chapter whose initial 
operating license was issued prior to January 10, 1997, who seek to 
revise the current accident source term used in their design basis 
radiological analyses.
    (b) Requirements. (1) A licensee who seeks to revise its current 
accident source term in design basis radiological

[[Page 72002]]

consequence analyses shall apply for a license amendment under 
Sec. 50.90. The application shall contain an evaluation of the 
consequences of applicable design basis accidents \1\ previously 
analyzed in the safety analysis report.
---------------------------------------------------------------------------

    \1\ The fission product release assumed for these calculations 
should be based upon a major accident, hypothesized for purposes of 
design analyses or postulated from considerations of possible 
accidental events, that would result in potential hazards not 
exceeded by those from any accident considered credible. Such 
accidents have generally been assumed to result in substantial 
meltdown of the core with subsequent release of appreciable 
quantities of fission products.
---------------------------------------------------------------------------

    (2) The NRC may issue the amendment only if the applicant's 
analysis demonstrates with reasonable assurance that:
    (i) An individual located at any point on the boundary of the 
exclusion area for any 2-hour period following the onset of the 
postulated fission product release, would not receive a radiation dose 
in excess of 0.25 Sv (25 rem) \2\ total effective dose equivalent 
(TEDE).
---------------------------------------------------------------------------

    \2\ 2 The use of 0.25 Sv (25 rem) TEDE is not intended to imply 
that this value constitutes an acceptable limit for emergency doses 
to the public under accident conditions. Rather, this 0.25 Sv (25 
rem) TEDE value has been stated in this section as a reference 
value, which can be used in the evaluation of proposed design basis 
changes with respect to potential reactor accidents of exceedingly 
low probability of occurrence and low risk of public exposure to 
radiation.
---------------------------------------------------------------------------

    (ii) An individual located at any point on the outer boundary of 
the low population zone, who is exposed to the radioactive cloud 
resulting from the postulated fission product release (during the 
entire period of its passage), would not receive a radiation dose in 
excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
    (iii) Adequate radiation protection is provided to permit access to 
and occupancy of the control room under accident conditions without 
personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) 
total effective dose equivalent (TEDE) for the duration of the 
accident.
    9. Part 50, Appendix A, section II, ``Protection by Multiple 
Fission Product Barriers,'' ``Criterion 19--Control room'' is revised 
to read as follows:

Appendix A to Part 50--General Design Criteria for Nuclear Power 
Plants

* * * * *

II. Protection by Multiple Fission Product Barriers

* * * * *
    Criterion 19--Control room. A control room shall be provided 
from which actions can be taken to operate the nuclear power unit 
safely under normal conditions and to maintain it in a safe 
condition under accident conditions, including loss-of-coolant 
accidents. Adequate radiation protection shall be provided to permit 
access and occupancy of the control room under accident conditions 
without personnel receiving radiation exposures in excess of 5 rem 
whole body, or its equivalent to any part of the body, for the 
duration of the accident. Equipment at appropriate locations outside 
the control room shall be provided (1) with a design capability for 
prompt hot shutdown of the reactor, including necessary 
instrumentation and controls to maintain the unit in a safe 
condition during hot shutdown, and (2) with a potential capability 
for subsequent cold shutdown of the reactor through the use of 
suitable procedures.
    Applicants for and holders of construction permits and operating 
licenses under this part who apply on or after January 10, 1997, 
applicants for design certifications under part 52 of this chapter 
who apply on or after January 10, 1997, applicants for and holders 
of combined licenses under part 52 of this chapter who do not 
reference a standard design certification, or holders of operating 
licenses using an alternative source term under Sec. 50.67, shall 
meet the requirements of this criterion, except that with regard to 
control room access and occupancy, adequate radiation protection 
shall be provided to ensure that radiation exposures shall not 
exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as 
defined in Sec. 50.2 for the duration of the accident.
* * * * *

PART 54--REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR 
POWER PLANTS

    10. The authority citation for Part 54 continues to read as 
follows:

    Authority: Secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 68 
Stat. 936, 937, 938, 948, 953, 954, 955, as amended, sec. 234, 83 
Stat. 1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 
2232, 2233, 2236, 2239, 2282); secs 201, 202, 206, 88 Stat. 1242, 
1244, as amended (42 U.S.C. 5841, 5842), E.O. 12829, 3 CFR, 1993 
Comp., p. 570; E.O. 12958, as amended, 3 CFR, 1995 Comp., p. 333; 
E.O. 12968, 3 CFR, 1995 Comp., p. 391.

    11. Section 54.4 is amended by revising paragraph (a)(1)(iii) to 
read as follows:


Sec. 54.4  Scope.

    (a) * * *
    (1) * * *
    (iii) The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures comparable 
to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or 
Sec. 100.11 of this chapter, as applicable.
* * * * *
    Dated at Rockville, Maryland, this 17th day of December 1999.

    For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 99-33283 Filed 12-22-99; 8:45 am]
BILLING CODE 7590-01-P