99-33283. Use of Alternative Source Terms at Operating Reactors  

  • [Federal Register Volume 64, Number 246 (Thursday, December 23, 1999)]
    [Rules and Regulations]
    [Pages 71990-72002]
    From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
    [FR Doc No: 99-33283]
    
    
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    NUCLEAR REGULATORY COMMISSION
    
    10 CFR Parts 21, 50, and 54
    
    RIN 3150-AG12
    
    
    Use of Alternative Source Terms at Operating Reactors
    
    AGENCY: Nuclear Regulatory Commission.
    
    ACTION: Final rule.
    
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    SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its 
    regulations to allow holders of operating licenses for nuclear power 
    plants to voluntarily replace the traditional source term used in 
    design basis accident analyses with alternative source terms. This 
    action will allow interested licensees to pursue cost beneficial 
    licensing actions to reduce unnecessary regulatory burden without 
    compromising the margin of safety of the facility. The NRC is 
    announcing the availability of a draft regulatory guide and a draft 
    Standard Review Plan section on this subject for public comment. The 
    NRC is also amending its regulations to revise certain sections to 
    conform with the final rule published on December 11, 1996, concerning 
    reactor site criteria.
    
    EFFECTIVE DATE: January 24, 2000.
    
    FOR FURTHER INFORMATION CONTACT: Mr. Stephen F. LaVie, Office of 
    Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 
    Washington, DC 20555-0001; telephone: (301) 415-1081; or by Internet 
    electronic mail to sfl@nrc.gov.
    
    SUPPLEMENTARY INFORMATION:
    
    I. Background
    II. Analysis of Public Comments
    III. Section-by-Section Analysis
    IV. Draft Regulatory Guide; Issuance, Availability
    V. Draft Standard Review Plan Section; Issuance, Availability
    VI. Referenced Documents
    VII. Finding of No Significant Environmental Impact; Availability
    VIII. Paperwork Reduction Act Statement
    IX. Regulatory Analysis
    X. Regulatory Flexibility Act Certification
    XI. Backfit Analysis
    XII. Small Business Regulatory Enforcement Fairness Act
    XIII. National Technology Transfer and Advancement Act
    
    I. Background
    
        A holder of an operating license (i.e., the licensee) for a light-
    water power reactor is required by regulations issued by the NRC (or 
    its predecessor, the U.S. Atomic Energy Commission, (AEC)) to submit a 
    safety analysis report (or, for early reactors, a hazard summary 
    report) that contains assessments of the radiological consequences of 
    potential accidents and an evaluation of the proposed facility site. 
    The NRC uses this information in its evaluation of the suitability of 
    the reactor design and the proposed site as required by its regulations 
    contained in 10 CFR Parts 50 and 100. Section 100.11, which was adopted 
    by the AEC in 1962 (27 FR 3509; April 12, 1962), requires an applicant 
    to assume (1) a fission product release from the reactor core, (2) the 
    expected containment leak rate, and (3) the site meteorological 
    conditions to establish an exclusion area and a low population zone. 
    This fission product release is based on a major accident that would 
    result in substantial release of appreciable quantities of fission 
    products from the core to the containment atmosphere. A note to 
    Sec. 100.11 states that Technical Information Document (TID) 14844, 
    ``Calculation of Distance Factors for
    
    [[Page 71991]]
    
    Power and Test Reactors,'' may be used as a source of guidance in 
    developing the exclusion area, the low population zone, and the 
    population center distance. Changes to the design of the facility and 
    the procedures for operating the facility are evaluated in part by 
    determining whether there are changes to the calculated fission product 
    release.
        The fission product release from the reactor core into containment 
    is referred to as the ``source term'' and it is characterized by the 
    composition and magnitude of the radioactive material, the chemical and 
    physical properties of the material, and the timing of the release from 
    the reactor core. The accident source term is used to evaluate the 
    radiological consequences of design basis accidents (DBAs) in showing 
    compliance with various requirements of the NRC's regulations. Although 
    originally used for site suitability analyses, the accident source term 
    is a design parameter for accident mitigation features, equipment 
    qualification, control room operator radiation doses, and post-accident 
    vital area access doses. The measurement range and alarm setpoints of 
    some installed plant instrumentation and the actuation of some plant 
    safety features are based in part on the accident source term. The TID-
    14844 source term was explicitly stated as a required design parameter 
    for several Three Mile Island (TMI)-related requirements.
        The NRC's methods for calculating accident doses, as described in 
    Regulatory Guide 1.3, ``Assumptions Used for Evaluating the Potential 
    Radiological Consequences of a Loss of Coolant Accident for Boiling 
    Water Reactors''; Regulatory Guide 1.4, ``Assumptions Used for 
    Evaluating the Potential Radiological Consequences of a Loss of Coolant 
    Accident for Pressurized Water Reactors''; and NUREG-0800, ``Standard 
    Review Plan for the Review of Safety Analysis Reports for Nuclear Power 
    Plants,'' were developed to be consistent with the TID-14844 source 
    term and the whole body and thyroid dose guidelines stated in 
    Sec. 100.11. In this regulatory framework, the source term is assumed 
    to be released immediately to the containment at the start of the 
    postulated accident. The chemical form of the radioiodine released to 
    the containment atmosphere is assumed to be predominantly elemental, 
    with the remainder being small fractions of particulate and organic 
    iodine forms. Radiation doses are calculated at the exclusion area 
    boundary (EAB) for the first 2 hours and at the low population zone 
    (LPZ) for the assumed 30-day duration of the accident. The whole body 
    dose comes primarily from the noble gases in the source term. The 
    thyroid dose is based on inhalation of radioiodines. In analyses 
    performed to date, the thyroid dose has generally been limiting. The 
    design of some engineered safety features, such as containment spray 
    systems and the charcoal filters in the containment, the building 
    exhaust, and the control room ventilation systems, are predicated on 
    these postulated thyroid doses. Subsequently, the NRC adopted the whole 
    body and thyroid dose criteria in Criterion 19 of 10 CFR Part 50, 
    Appendix A (36 FR 3255; February 20, 1971).
        The source term in TID-14844 is representative of a major accident 
    involving significant core damage and is typically postulated to occur 
    in conjunction with a large loss-of-coolant accident (LOCA). Although 
    the LOCA is typically the maximum credible accident, NRC experience in 
    reviewing license applications has indicated the need to consider other 
    accident sequences of lesser consequence but higher probability of 
    occurrence. Some of these additional accident analyses may involve 
    source terms that are a fraction of those specified in TID-14844. The 
    DBAs were not intended to be actual event sequences but, rather, were 
    intended to be surrogates to enable deterministic evaluation of the 
    response of the plant engineered safety features. These accident 
    analyses are intentionally conservative in order to address 
    uncertainties in accident progression, fission product transport, and 
    atmospheric dispersion. Although probabilistic risk assessments (PRAs) 
    can provide useful insights into system performance and suggest changes 
    in how the desired defense in depth is achieved, defense in depth 
    continues to be an effective way to account for uncertainties in 
    equipment and human performance. The NRC's policy statement on the use 
    of PRA methods (60 FR 42622; August 16, 1995) calls for the use of PRA 
    technology in all regulatory matters in a manner that complements the 
    NRC's deterministic approach and supports the traditional defense-in-
    depth philosophy.
        Since the publication of TID-14844, significant advances have been 
    made in understanding the timing, magnitude, and chemical form of 
    fission product releases from severe nuclear power plant accidents. 
    Many of these insights developed out of the major research efforts 
    started by the NRC and the nuclear industry after the accident at Three 
    Mile Island (TMI). In 1995, the NRC published NUREG-1465, ``Accident 
    Source Terms for Light-Water Nuclear Power Plants,'' which utilized 
    this research to provide more physically based estimates of the 
    accident source term that could be applied to the design of future 
    light-water power reactors. The NRC sponsored significant review 
    efforts by peer reviewers, foreign research partners, industry groups, 
    and the general public (request for public comment was published in 57 
    FR 33374; July 28, 1992).
        The information in NUREG-1465 presents a representative accident 
    source term (``revised source term'') for a boiling-water reactor (BWR) 
    and for a pressurized-water reactor (PWR). These revised source terms 
    are described in terms of radionuclide composition and magnitude, 
    physical and chemical form, and timing of release. Where TID-14844 
    addressed three categories of radionuclides, the revised source terms 
    categorize the accident release into eight groups on the basis of 
    similarity in chemical behavior. Where TID-14844 assumed an immediate 
    release of the activity, the revised source terms have five release 
    phases that are postulated to occur over several hours, with the onset 
    of major core damage occurring after 30 minutes. Where TID-14844 
    assumed radioiodine to be predominantly elemental, the revised source 
    terms assume radioiodine to be predominantly cesium iodide (CsI), an 
    aerosol that is more amenable to mitigation mechanisms.
        For DBAs, the NUREG-1465 source terms (up to and including the 
    early in-vessel phase) are comparable to the TID-14844 source term with 
    regard to the magnitude of the noble gas and radioiodine release 
    fractions. However, the revised source terms offer a more 
    representative description of the radionuclide composition and release 
    timing. The NRC has determined (SECY-94-302, December 19, 1994) that 
    design basis analyses will address the first three release phases--
    coolant, gap, and in-vessel. The ex-vessel and late in-vessel phases 
    are considered to be inappropriate for design basis analysis purposes. 
    These latter releases could only result from core damage accidents with 
    vessel failure and core-concrete interactions.
        The objective of NUREG-1465 was to define revised accident source 
    terms for regulatory application for future light water reactors 
    (LWRs). The NRC's intent was to capture the major relevant insights 
    available from severe accident research to provide, for regulatory 
    purposes, a more realistic portrayal of the amount of the postulated 
    accident source term. These source terms were derived from examining a 
    set of severe accident sequences for LWRs of current
    
    [[Page 71992]]
    
    design. Because of general similarities in plant and core design 
    parameters, these results are considered to be applicable to 
    evolutionary and passive LWR designs. The revised source term has been 
    used in evaluating the Westinghouse AP600 standard design certification 
    application. (A draft version of NUREG-1465 was used in evaluating 
    Combustion Engineering's (CE's) System 80+ design.)
        The NRC considered the applicability of the revised source terms to 
    operating reactors and determined that the current analytical approach 
    based on the TID-14844 source term would continue to be adequate to 
    protect public health and safety, and that operating reactors licensed 
    under this approach would not be required to reanalyze accidents using 
    the revised source terms. The NRC concluded that some licensees may 
    wish to use an alternative source term in analyses to support 
    operational flexibility and cost-beneficial licensing actions and that 
    some of these applications could provide concomitant improvements in 
    overall safety and in reduced occupational exposure. The NRC initiated 
    several actions to provide a regulatory basis for operating reactors to 
    voluntarily amend their facility design bases to enable use of the 
    revised source term in design basis analyses. First, the NRC solicited 
    ideas on how an alternative source term might be implemented. In 
    November 1995, the Nuclear Energy Institute (NEI) submitted its generic 
    framework, Electric Power Research Institute Technical Report TR-
    105909, ``Generic Framework for Application of Revised Accident Source 
    Term to Operating Plants.'' This report and the NRC response were 
    discussed in SECY-96-242 (November 25, 1996). Second, the NRC initiated 
    an assessment of the overall impact of substituting the NUREG-1465 
    source terms for the traditionally used TID-14844 source term at three 
    typical facilities. This was done to evaluate the issues involved with 
    applying the revised source terms at operating plants. SECY-98-154 
    (June 30, 1998) described the conclusions of this assessment. Third, 
    the NRC accepted license amendment requests related to implementation 
    of the revised source terms at a small number of pilot plants. 
    Experience has demonstrated that evaluation of a limited number of 
    plant-specific submittals improves regulation and regulatory guidance 
    development. The review of these pilot projects is currently in 
    progress. Insights from these pilot plant reviews have been 
    incorporated into the regulatory guidance that was developed in 
    conjunction with this rulemaking. Fourth, the NRC initiated an 
    assessment on whether rulemaking would be necessary to allow operating 
    reactors to use an alternative source term. This final rule and the 
    supporting regulatory guidance have resulted from this assessment.
        This final rulemaking for use of alternative source terms is 
    applicable to holders of operating licenses issued prior to January 10, 
    1997, under 10 CFR Part 50, ``Domestic Licensing of Production and 
    Utilization Facilities,'' and to holders of renewed licenses under 10 
    CFR Part 54, ``Requirements for Renewal of Operating Licenses for 
    Nuclear Power Plants,'' whose initial operating license was issued 
    prior to January 10, 1997. The regulations of Part 50 are supplemented 
    by those in other parts of Chapter I of Title 10, including Part 100, 
    ``Reactor Site Criteria.'' Part 100 contains language that 
    qualitatively defines a required accident source term and contains a 
    note that discusses the availability of TID-14844. With the exception 
    of Sec. 50.34(f), there are no explicit requirements in Chapter I of 
    Title 10 to use the TID-14844 accident source term. Section 50.34(f), 
    which addresses additional TMI-related requirements, is only applicable 
    to a limited number of construction permit applications pending on 
    February 16, 1982, and to applications under Part 52.
        An applicant for an operating license is required by Sec. 50.34(b) 
    to submit a final safety analysis report (FSAR) that describes the 
    facility and its design bases and limits, and presents a safety 
    analysis of the structures, systems, and components of the facility as 
    a whole. Guidance in performing these analyses is given in regulatory 
    guides. In its review of the more recent applications for operating 
    licenses, the NRC has used the review procedures in NUREG-0800, 
    ``Standard Review Plan for the Review of Safety Analysis Reports for 
    Nuclear Power Plants'' (SRP). These review procedures reference or 
    provide acceptable assumptions and analysis methods. The facility FSAR 
    documents the assumptions and methods actually used by the applicant in 
    the required safety analyses. The NRC's finding that a license may be 
    issued is based on the review of the FSAR, as documented in the 
    Commission's safety evaluation report (SER). Fundamental assumptions 
    that are design inputs, including the source term, were required to be 
    included in the FSAR and became part of the design basis 1 
    of the facility. From a regulatory standpoint, the requirement to use 
    the TID-14844 source term is expressed as a licensee commitment 
    (typically to Regulatory Guide 1.3 or 1.4) documented in the facility 
    FSAR, and is subject to the requirements of Sec. 50.59.
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        \1\ As defined in Sec. 50.2, design bases means that information 
    which identifies the specific functions to be performed by a 
    structure, system, or component of a facility, and the specific 
    values or ranges of values chosen for controlling parameters as 
    reference bounds for design. These values may be (1) restraints 
    derived from generally accepted ``state of the art'' practices for 
    achieving functional goals, or (2) requirements derived from 
    analysis (based on calculation and/or experiments) of the effects of 
    a postulated accident for which a structure, system, or component 
    must meet its functional goals. The NRC considers the accident 
    source term to be an integral part of the design basis because it 
    sets forth specific values (or range of values) for controlling 
    parameters that constitute reference bounds for design.
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        In 1996 (61 FR 65175; December 11, 1996), the NRC amended its 
    regulations in 10 CFR Parts 21, 50, 52, 54, and 100. That regulatory 
    action produced site criteria for future sites, presented a stable 
    regulatory basis for seismic and geologic siting and the engineering 
    design of future nuclear power plants to withstand seismic events, and 
    relocated source term and dose requirements for future plants into Part 
    50. Because these dose requirements tend to affect reactor design 
    rather than siting, they are more appropriately located in Part 50. 
    This decoupling of siting from design is consistent with the future 
    licensing of facilities using standardized plant designs, the design 
    features of which have been or will be certified in a separate design 
    certification rulemakings. This decoupling of siting from design was 
    directed by Congress in the 1980 Authorization Act for the NRC. Because 
    the revised criteria would not apply to operating reactors, the non-
    seismic and seismic reactor site criteria for operating reactors were 
    retained as Subpart A and Appendix A to Part 100, respectively. The 
    revised reactor site criteria were added as Subpart B in Part 100, and 
    revised source term and dose requirements were moved to Sec. 50.34. The 
    existing source term and dose requirements of Subpart A of Part 100 
    will remain in place as the licensing bases for those operating 
    reactors that do not elect to use an alternative source term.
        In relocating the source term and dose requirements for future 
    reactors to Sec. 50.34, the NRC retained the requirements for the 
    exclusion area and the low population zone, but revised the associated 
    numerical dose criteria to replace the two different doses for the 
    whole body and the thyroid gland with a single, total effective dose 
    equivalent (TEDE) value. The dose criteria for the whole body and the 
    thyroid, and the
    
    [[Page 71993]]
    
    immediate 2-hour exposure period were largely predicated by the assumed 
    source term being predominantly noble gases and radioiodines 
    instantaneously released to the containment and the assumed ``single 
    critical organ'' method of modeling the internal dose used at the time 
    that Part 100 was originally published. However, the current dose 
    criteria, by focusing on doses to the thyroid and the whole body, 
    assume that the major contributor to doses will be radioiodine. 
    Although this may be appropriate with the TID-14844 source term, as 
    implemented by Regulatory Guides 1.3 and 1.4, it may not be true for a 
    source term based on a more complete understanding of accident 
    sequences and phenomenology.
        The postulated chemical and physical form of radioiodine in the 
    revised source terms is more amenable to mitigation and, as such, 
    radioiodine may not always be the predominant radionuclide in an 
    accident release. The revised source terms include a larger number of 
    radionuclides than did the TID-14844 source term as implemented in 
    regulatory guidance. The whole body and thyroid dose criteria ignore 
    these contributors to dose. The NRC amended its radiation protection 
    standards in Part 20 in 1991 (56 FR 23391; May 21, 1991) replacing the 
    single, critical organ concept for assessing internal exposure with the 
    TEDE concept that assesses the impact of all relevant nuclides upon all 
    body organs. TEDE is defined to be the deep dose equivalent (for 
    external exposure) plus the committed effective dose equivalent (for 
    internal exposure). The deep dose equivalent (DDE) is comparable to the 
    present whole body dose; the committed effective dose equivalent (CEDE) 
    is the sum of the products of doses (integrated over a 50-year period) 
    to selected body organs resulting from the intake of radioactive 
    material multiplied by weighting factors for each organ that are 
    representative of the radiation risk associated with the particular 
    organ.
        The TEDE, using a risk-consistent methodology, assesses the impact 
    of all relevant nuclides upon all body organs. Although it is expected 
    that in many cases the thyroid could still be the limiting organ and 
    radioiodine the limiting radionuclide, this conclusion cannot be 
    assured in all potential cases. The revised source terms postulate that 
    the core inventory is released in a sequence of phases over 10 hours, 
    with the more significant release commencing at about 30 minutes from 
    the start of the event. The assumption that the 2-hour exposure period 
    starts immediately at the onset of the release is inconsistent with the 
    phased release postulated in the revised source terms. The final rule 
    adopts the future LWR dose criteria for operating reactors that elect 
    to use an alternative source term.
        An accidental release of radioactivity can result in radiation 
    exposure to control room operators. Normal ventilation systems may draw 
    this activity into the control room where it can result in external and 
    internal exposures. Control room designs differ but, in general, design 
    features are provided to detect the accident or the activity and 
    isolate the normal ventilation intake. Emergency ventilation systems 
    are activated to minimize infiltration of contaminated air and to 
    remove activity that has entered the control room. Personnel exposures 
    can also result from radioactivity outside of the control room. 
    However, because of concrete shielding of the control room, these 
    latter exposures are generally not limiting. The objective of the 
    control room design is to provide a location from which actions can be 
    taken to operate the plant under normal conditions and to maintain it 
    in a safe condition under accident conditions. General Design Criterion 
    19 (GDC-19), ``Control Room,'' of Appendix A to 10 CFR Part 50 (36 FR 
    3255; February 20, 1971), establishes minimum requirements for the 
    design of the control room, including a requirement for radiation 
    protection features adequate to permit access to and occupancy of the 
    control room under accident conditions. The GDC-19 criteria were 
    established for judging the acceptability of the control room design 
    for protecting control room operators under postulated design basis 
    accidents, a significant concern being the potential increases in 
    offsite doses that might result from the inability of control room 
    personnel to adequately respond to the event.
        The GDC-19 criteria are expressed in terms of whole body dose, or 
    its equivalent to any organ. The NRC did not revise the criteria when 
    Part 20 was amended (56 FR 23391; May 21, 1991) instead deferring such 
    action to individual facility licensing actions (NUREG/CR-6204, 
    ``Questions and Answers Based on the Revised 10 CFR Part 20''). This 
    position was taken in the interest of maintaining the licensing basis 
    for those facilities already licensed. The NRC is replacing the current 
    dose criteria of GDC-19 for future reactors and for operating reactors 
    that elect to use an alternative source term with a criterion expressed 
    in terms of TEDE. The rationale for this revision is similar to the 
    rationale, discussed earlier in this preamble, for revising the dose 
    criteria for offsite exposures.
        On January 10, 1997 (61 FR 65157), the NRC amended 10 CFR Parts 21, 
    50, 52, 54, and 100 of its regulations to update the criteria used in 
    decisions regarding power reactor siting for future nuclear power 
    plants. The NRC intended that future licensing applications in 
    accordance with Part 52 utilize a source term consistent with the 
    source term information in NUREG-1465 and the accident TEDE criteria in 
    Parts 50 and 100. However, during the final design approval (FDA) and 
    design certification proceeding for the Westinghouse AP600 advanced 
    light-water reactor design, the NRC staff and Westinghouse determined 
    that exemptions were necessary from Secs. 50.34(f)(2)(vii), (viii), 
    (xxvi), and (xxviii) and 10 CFR Part 50, Appendix A, GDC-19. This final 
    rule would eliminate the need for these exemptions for future 
    applicants under Part 52 by making conforming changes to Part 50, 
    Appendix A, GDC-19 and Sec. 50.34.
    
    II. Analysis of Public Comments
    
        The NRC published a proposed rule in the Federal Register (64 FR 
    12117, March 31, 1999); that would provide a regulatory framework for 
    the voluntary implementation of alternative source terms as a change to 
    the design basis at currently licensed power reactors, while retaining 
    the existing regulatory framework for currently licensed power reactor 
    licensees who choose not to implement an alternative source term. The 
    rule proposed relocating source term and dose requirements that apply 
    primarily to plant design into 10 CFR Part 50 for operating reactors 
    that choose to implement an alternative source term. The rule also 
    proposed conforming changes to Sec. 50.34(f) and Part 50, Appendix A, 
    GDC-19 to eliminate the need for exemptions for future applicants under 
    Part 52.
        The NRC received seven letters commenting on the proposed rule. All 
    comments including those received by the NRC after the expiration of 
    the public comment period but before June 25, 1999, were considered. 
    The commenters included two State regulatory agencies, two nuclear 
    industry groups and three utilities. The State of Florida Department of 
    Community Affairs indicated that they had no comments on the proposed 
    rule. The State of New Jersey Department of Environmental Protection 
    concurred with the NRC's position on the use of an AST in emergency 
    preparedness applications and stated a desire to review the draft 
    regulatory guidance when issued. Winston & Strawn
    
    [[Page 71994]]
    
    submitted comments on behalf of the Nuclear Utility Backfitting and 
    Reform Group (NUBARG). The Nuclear Energy Institute (NEI) submitted 
    comments on behalf of the nuclear industry. Two of the utilities 
    provided comments, while the third endorsed the comments submitted by 
    NEI. Copies of these letters are available for public inspection and 
    copying for a fee at the NRC Public Document Room, 2120 L Street NW. 
    (Lower Level), Washington, DC.
    
    1. NUBARG Comments
    
        NUBARG supports the rule, noting that the rule as proposed defines 
    an acceptable regulatory process for implementing more realistic 
    accident source terms. NUBARG requested clarification in the final rule 
    of situations in which an alternative source term (AST) may be applied 
    in future backfitting 2 decisions. First, NUBARG suggests 
    that the NRC clarify the extent it intends to use the revised source 
    term in assessing whether new generic requirements provide a cost-
    justified, substantial increase in safety in accordance with NRC's 
    backfitting rule, Sec. 50.109. NUBARG believes that continued use of 
    the source term in TID-14844 for this purpose in spite of its known 
    limitations would be inappropriate and could lead to overly 
    conservative estimates of the safety impact of proposed new 
    requirements. Second, NUBARG suggests a similar clarification for 
    plant-specific backfit decisions for plants that have not opted to 
    implement the revised source term. NUBARG believes that the NRC has 
    discretion to take all relevant factors into account in making its 
    safety benefit assessment of the proposed backfit, including the 
    current state of knowledge concerning the accident source term. NUBARG 
    suggested that the statements of considerations accompanying the final 
    rule address these issues. NUBARG also suggests that relevant NRC 
    guidance should also be revised to reflect NRC policy in these areas.
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        \2\ As provided in Sec. 50.109, Backfitting is defined as the 
    modification of or addition to systems, structures, components, or 
    the design of a facility; or the design approval or manufacturing 
    license for a facility; or the procedures or organization required 
    to design, construct or operate a facility; any of which may result 
    from a new or amended provision in the Commission rules or the 
    imposition of a regulatory staff position interpreting the 
    Commission rules that is either new or different from a previously 
    applicable staff position.
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        NRC Response. When radiological consequence analyses are involved, 
    the NRC expects to use a technically appropriate AST in evaluating 
    generic and plant-specific backfitting analyses, including those 
    proposed for facilities that have not implemented an AST. The NRC 
    agrees with the NUBARG position that the NRC has discretion to take all 
    new information on accident source terms into account. The NRC's 
    guidance for evaluating proposed NRC regulatory actions (including 
    backfitting) are contained in NUREG/BR-0058, ``Regulatory Analysis 
    Guidelines of the U.S. Nuclear Regulatory Commission,'' and NUREG/BR-
    0184, ``Regulatory Analysis Technical Evaluation Handbook.'' These 
    documents state that value and impact (including adverse effects on 
    health and safety) parameters are to be best estimates, preferably mean 
    or expected values. These documents also provide that analyses are to 
    be based largely on risk considerations.
    
    2. NEI Comment 1
    
        NEI stated that the Section-by-Section Analysis in the proposed 
    rule notice is consistent with the NRC's intent to permit limited 
    application of the new research results. NEI noted that these limited 
    applications are of two types: (1) application of alternative source 
    term radiological composition and magnitude in a quantitative analysis 
    relative to the effect on the performance of a given engineered safety 
    feature; or (2) application of only the timing aspects in conjunction 
    with the original TID-14844 source term. NEI stated that proposed 
    Sec. 50.67 appears to apply to applications where a licensee would use 
    a completely new source term such as NUREG-1465 in all aspects of the 
    plant design. The NEI comment acknowledged that further guidance in a 
    subsequent regulatory guide and standard review plan is helpful and 
    necessary. Nonetheless, NEI is concerned that licensee pursuit of 
    either of these limited applications might ultimately require seeking 
    an exemption to Sec. 50.67, or require extensive analysis. NEI 
    recommended that the NRC should: (1) revise the proposed rule language 
    to accommodate limited application of an alternative source term as 
    done in the Section-By-Section Analysis; (2) provide clarification in 
    the Statement of Consideration (SOC) for the rule; and (3) for 
    applications that continue to use the TID source term but incorporate 
    attributes of newer technical insights such as timing of releases, 
    specify that the provisions of the proposed rule do not apply.
        NRC Response. The language of Sec. 50.67(b) requires an evaluation 
    of the consequences of applicable design basis accidents. The NRC 
    believes that the use of the modifier applicable provides the basis for 
    processing selective implementations. Design basis accidents not 
    applicable to a particular selective implementation would not be 
    required to be evaluated. The NRC expects that the licensee will 
    evaluate all applicable impacts of the proposed AST implementation. 
    While a selective implementation may result in a reduced scope of 
    evaluation, the licensee must still demonstrate that the AST 
    implementation and any associated proposed modifications will not 
    result in accident conditions exceeding the criteria specified in 
    Sec. 50.67. Therefore, these criteria are applicable to full and 
    selective implementations alike. The scope of the required re-analyses 
    will depend on the specific application proposed by the licensee. 
    Guidance with regard to this scope is properly provided in the draft 
    regulatory guide prepared for this rule. Therefore, the NRC has decided 
    against revising the rule language as suggested by NEI. Consistent with 
    the second NEI recommendation, the NRC has modified paragraph D of the 
    section-by-section analysis to clarify this issue.
    
    3. NEI Comment 2
    
        In its second comment, NEI noted that the SOC provides that 
    licensees may need to perform additional evaluations of equipment 
    qualifications (Sec. 50.49). The SOC should discuss the circumstances 
    when such an evaluation may be necessary. NEI recommended that the SOC 
    should be amended to state that regardless of source term used, the 
    licensee would be required to re-evaluate the equipment qualification 
    only when a plant modification alters the plant configuration so that 
    the underlying assumptions, with respect to dose distribution and 
    effects, are materially altered. NEI summarized conclusions of several 
    references in support of its position. NEI stated that there is no 
    basis to require or expect additional analyses of equipment 
    qualification if a licensee applied the alternative source term in 
    limited scope applications, absent a plant configuration change that 
    materially alters the dose distribution and effects assumed in existing 
    analyses.
        NRC Response. The re-baselining study prepared by the NRC staff 
    (SECY-98-154, June 30, 1998) considered the impact of an AST on 
    analyses of the postulated integrated radiation doses for plant 
    components exposed to containment atmosphere radiation sources and 
    those exposed to containment sump radiation sources. The staff's 
    conclusions regarding the atmosphere sources are consistent with those 
    identified by NEI in its comment. However, the re-baselining study also 
    concluded that the increased concentration of cesium in the containment 
    sump water could result in
    
    [[Page 71995]]
    
    an increase in the postulated integrated radiation doses for certain 
    plant components subject to equipment qualification. It is because of 
    this conclusion that the NRC included the discussion in the SOC 
    regarding re-evaluation of equipment environmental qualification. The 
    NEI comment provides no additional information that would cause the NRC 
    to change its position on this matter. Further, the NRC has determined 
    that it is necessary to consider the potential impact of the postulated 
    cesium concentration in the containment sump water as it applies to all 
    operating power reactors, not just to those licensees amending their 
    design basis to use an AST. Since the postulated increase in the 
    integrated dose occurs only following an accident, there is no adverse 
    effect on equipment relied upon to perform safety functions immediately 
    following an accident. Rather, this issue affects equipment that is 
    required to be operable longer than about 30 days to 4 months after an 
    accident. As such, the NRC determined that continued plant operation 
    does not pose an immediate threat to public health and safety. Also, 
    should such long-term equipment fail there will not be an undue threat 
    to public health and safety as protective actions for the public would 
    have already been implemented by the time the postulated failure could 
    occur. In addition, the time period between the onset of the event and 
    the projected failure allows compensatory measures to be taken to 
    prevent the equipment failure or to restore the degraded safety 
    function. The NRC will evaluate this issue as a generic safety issue to 
    determine whether further regulatory actions are justified. The final 
    regulatory guide, or subsequent revisions thereto, is expected to 
    reflect the resolution of this generic safety issue.
    
    4. NEI Comment 3
    
        NEI recommends that the definition of Source Term in Sec. 50.2 be 
    revised to ``Source term refers to the magnitude and mix of 
    radionuclides released from the fuel, their physical and chemical form, 
    and the timing of their release.'' NEI stated that the language in the 
    proposed rule would prohibit the use of Sec. 50.67 for accidents such 
    as the fuel handling accident.
        NRC Response. The NRC agrees with the proposed revision. The 
    proposed definition was consistent with the definition of source term 
    as used in NUREG-1465, which was written primarily to address loss of 
    coolant accidents (LOCA). The regulatory guidance for this rule extends 
    the NUREG-1465 source terms to other accidents which involve core 
    damage. The definition suggested by NEI is consistent with the proposed 
    use of the AST. The Sec. 50.2 definition has been revised in the final 
    rule to reflect the change suggested by NEI and that suggested by 
    Arizona Public Service Comment 1 below.
    
    5. NEI Comment 4
    
        NEI stated that the proposed rule does not permit new test reactors 
    to use an alternative source term. New test reactors would have to use 
    the Part 100 Subpart A, ``Evaluation Factors for Stationary Power 
    Reactor Site Applications Before January 10, 1997, and for Testing 
    Reactors,'' even though their application for an operating license 
    would be filed after January 10, 1997. The use of Section 50.67, 
    ``Accident Source Term,'' is limited to holders of operating licenses 
    issued before January 10, 1997. This wording prohibits new test 
    reactors from using the alternative source term. NEI recommended that 
    Sec. 50.67 be amended to allow new test reactors to use an alternative 
    source term.
        NRC Response. Section 50.67 applies only to holders of licenses for 
    operating reactors, including test reactors, whose licenses were issued 
    before January 10, 1997. There is no regulatory requirement for a 
    specific source term for reactors to be licensed in the future, 
    including test reactors. Accordingly, no regulatory action is necessary 
    to accommodate the NEI recommendation.
    
    6. Duke Energy Corporation Comment
    
        Duke Energy Corporation (Duke) endorsed the comments submitted on 
    behalf of the industry by NEI. Duke stated that the proposed 
    Sec. 50.67(b)(1) was not clear regarding whether licensees will be 
    allowed to use a revised source term on a limited basis (e.g., for 
    analyses of a specific accident or function), or whether they will be 
    required to review the entire radiological consequence analyses to 
    apply for the new source term. Duke suggested that necessary guidance 
    be provided in the draft regulatory guidance to allow for limited use 
    of the new source terms where such use can be justified.
        NRC Response. This comment is similar to NEI Comment 1 addressed 
    previously. As stated in the SOC, the NRC will consider justifiable 
    limited (i.e., selective) applications of an AST. Although a selective 
    implementation may result in a reduced scope of evaluation, the 
    licensee must still demonstrate that the AST implementation and any 
    associated proposed modifications will not exceed the criteria 
    specified in Sec. 50.67. The scope of the required re-analyses will 
    depend on the specific application proposed by the licensee. Regulatory 
    guidance on selective implements and the scope of required re-analyses 
    has been included in the draft guide and are available as announced in 
    this Federal Register notice.
    
    7. Arizona Public Service Company Comment 1
    
        Arizona Public Service Company (APS) noted that the SOC statement, 
    ``a subsequent change to the source term must be made through a license 
    amendment'' could be interpreted as requiring prior NRC approval for 
    any change in the magnitude and mix of radionuclides released from the 
    reactor core. APS stated that this interpretation could place 
    additional restrictions on licensee efforts at economical fuel 
    management, including reload design.
        NRC Response. The NRC agrees with the APS comment. The NRC had 
    intended the phrase ``magnitude and mix'' to refer to the fractions of 
    the fission product inventory of the radionuclides released from the 
    reactor fuel. The NRC intent for the provision in question was to 
    require approval for changes in the radioactivity release fractions, 
    the radionuclides released, their physical and chemical form, and the 
    timing of their release. Since ``magnitude and mix'' could be a source 
    of confusion, the NRC has modified the Sec. 50.2 definition of Source 
    Term in the final rule to read: ``Source term refers to the magnitude 
    and mix of the radionuclides released from the fuel, expressed as 
    fractions of the fission product inventory in the fuel, as well as 
    their physical and chemical form, and the timing of their release.'' 
    This is consistent with NUREG-1465 when it refers to ``magnitude and 
    mix,'' since the NUREG-1465 presents these data in the form of tables 
    of release fractions and radionuclides. This revised language also 
    addresses NEI Comment 3 above.
    
    8. Arizona Public Service Company Comment 2
    
        In its second comment, APS noted that NUREG-1465 contains a 
    disclaimer that the accident source terms provided therein may not be 
    applicable to fuel irradiated in excess of 40 GWD/MTU. The NRC has 
    licensed core designs with fuel irradiations of up to 62 GWD/MTU. APS 
    questioned whether the NRC staff was going to address the affect of 
    high burnups on a generic basis, or on a facility-by-facility basis.
        NRC Response. The AST tabulated in the draft regulatory guidance, 
    which
    
    [[Page 71996]]
    
    differs in some aspects from that provided in NUREG-1465, is applicable 
    to peak rod average irradiations up to 62 GWD/MTU. Attachment 1 to the 
    regulatory analysis for this rulemaking describes the bases of this 
    extension in fuel irradiation as it applies to the AST. There are some 
    facility-by-facility considerations. For example, the increase in core 
    inventory for some long-lived radionuclides and the change in isotopic 
    mix due to the increase in plutonium fission as the fuel ages is 
    addressed by the Draft Guide-1081 provision that licensees re-analyze 
    the core inventory based on current operating parameters, including 
    fuel burnup.
    
    III. Section-by-Section Analysis
    
    A. Section 50.2
    
        The general ``definitions'' section for Part 50 is supplemented by 
    adding a definition of source term for the purpose of Sec. 50.67. In 
    NUREG-1465, the source term is defined by five projected 
    characteristics: (1) magnitude of radioactivity release, (2) 
    radionuclides released, (3) physical form of the radionuclides 
    released, (4) chemical form of the radionuclides released, and (5) 
    timing of the radioactivity release. The definition of source term in 
    Sec. 50.2 embodies the NUREG-1465 definition; however, the Sec. 50.2 
    definition includes the clarifying phrase, ``expressed as fractions of 
    the fission product inventory in the fuel,'' (see prior response to 
    Arizona Public Service Comment 1). Although all five characteristics 
    should be addressed in applications proposing the use of an alternative 
    source term, there may be technically justifiable applications in which 
    all five characteristics need not be addressed. The NRC intends to 
    allow licensees flexibility in implementing alternative source terms 
    consistent with maintaining a conservative, clear, logical, and 
    consistent plant design basis. The regulatory guidance that supports 
    this final rule describes an acceptable basis for defining the 
    characteristics of an alternative source term.
    
    B. Section 50.67(a)
    
        This paragraph defines the licensees that may seek to revise their 
    current radiological source term with an alternative source term. The 
    final rule is applicable to holders of operating licenses that were 
    issued under 10 CFR Part 50 before January 10, 1997, and to holders of 
    renewed licenses issued under 10 CFR Part 54 whose initial operating 
    license was issued prior to January 10, 1997. The final rule does not 
    require licensees to revise their current source term. The NRC 
    considered the acceptability of the TID-14844 source term at current 
    operating reactors and determined that the analytical approach based on 
    the TID-14844 source term would continue to be adequate to protect 
    public health and safety, and that operating reactors licensed under 
    this approach should not be required to reanalyze design basis 
    accidents using a new source term. The final rule does not explicitly 
    define an alternative source term. In lieu of an explicit reference to 
    NUREG-1465, Footnote 1 to the final rule identifies the significant 
    attributes of an accident source term. The regulatory guidance that is 
    being issued to support this final rule will identify ASTs (based on 
    the NUREG-1465 source terms) that are acceptable alternatives to the 
    source term in TID-14844, and will provide implementation guidance. 
    This approach will provide for future revised source terms if they are 
    developed and will allow licensees to propose additional alternatives 
    for NRC consideration.
    
    C. Section 50.67(b)(1)
    
        This paragraph of Sec. 50.67 identifies the information that a 
    licensee must submit as part of a license amendment application to use 
    an alternative source term. Because of the extensive use of the 
    accident source term in the design and operation of a power reactor and 
    the potential impact on postulated accident consequences and margins of 
    safety of a change of such a fundamental design assumption, the NRC has 
    determined that any change to the design basis to use an alternative 
    source term should be reviewed and approved by the NRC in the form of a 
    license amendment. Changes to the source term, by itself, would 
    ordinarily constitute a no significant hazards consideration. In 
    addition, generic analyses performed by the NRC staff in support of 
    this final rule have indicated that there are potential changes to the 
    facility as documented in the FSAR that will constitute a no 
    significant hazards consideration. However, these determinations will 
    have to be made for each proposed change based upon facility-specific 
    evaluations. The procedural requirements for processing a license 
    amendment are presented in Secs. 50.90 through 50.92.
        The NRC's regulations provide a regulatory mechanism for a licensee 
    to effect a change in its design basis in Sec. 50.59 3 that 
    allows a licensee to make changes to the facility as described in the 
    final safety analysis report (FSAR) without prior NRC approval, if the 
    proposed change meets certain criteria specified in Sec. 50.59. If the 
    criteria are not met, the licensee must request NRC approval of the 
    change using the license amendment process detailed in Sec. 50.90. 
    Significant to this final rule is the criterion that NRC review is 
    required if the proposed change would result in a greater than minimal 
    increase in consequences of an accident or malfunction. In many 
    applications, alternative source terms may reduce the postulated 
    consequences of the accident or malfunction. For this reason, the NRC 
    determined that the regulatory framework of Sec. 50.59 might not 
    provide assurance that this change in the design basis would be 
    recognized by the licensee as needing review by the NRC staff.
    ---------------------------------------------------------------------------
    
        \3\ Section 10 CFR 50.59 is being amended in a parallel, but 
    separate, rulemaking action. That rulemaking, when implemented is 
    expected to replace the unreviewed safety question (USQ) concept. 
    Further, the criteria for consequences are being revised from ``may 
    be increased'' to ``result in more than a minimal increase.'' Those 
    changes are not expected to invalidate the conclusions drawn in this 
    analysis.
    ---------------------------------------------------------------------------
    
        After a licensee has been authorized to substitute an alternative 
    source term in its design basis, subsequent changes to the facility 
    that involve an alternative source term may be processed under 
    Sec. 50.59 or Sec. 50.90, as appropriate. However, a subsequent change 
    to the fractions of the fission product inventory of the radionuclides 
    released from the reactor fuel, their chemical and physical form, or 
    the timing of their release as tabulated in the regulatory guidance 
    (with deviations proposed by the licensee and approved by the NRC) 
    could not be implemented under Sec. 50.59. This provision applies only 
    to these tabulated parameters.
        The final rule will require the applicant to perform analyses of 
    the consequences of applicable design basis accidents previously 
    analyzed in the safety analysis report and to submit a description of 
    the analysis inputs, assumptions, methodology, and results of these 
    analyses for NRC review. Applicable evaluations may include, but are 
    not limited to, those previously performed to show compliance with 
    Sec. 100.11, Sec. 50.49, Part 50 Appendix A GDC-19, Sec. 50.34(f), and 
    NUREG-0737, ``Clarification of TMI Action Plan Requirements,'' 
    requirements II.B.2, II.B.3, III.D.3.4. The regulatory guidance that 
    supports this final rule will provide guidance on the scope and extent 
    of analyses used to show compliance with this rule and on the 
    assumptions and methods used therein. It is not the NRC's intent that 
    all of the design basis radiological analyses for a facility be
    
    [[Page 71997]]
    
    performed again as a prerequisite for approval of the use of an 
    alternative source term. Nor is it the NRC's intent that EAB, LPZ, and 
    control room dose calculations be performed for all applications under 
    Sec. 50.67. The NRC does expect that the applicant will perform 
    sufficient evaluations, supported by calculations as warranted, to 
    demonstrate the acceptability of the proposed amendment.
    
    D. Sections 50.67(b)(2)(i),(ii), (iii)
    
        These subparagraphs contain the three criteria for NRC approval of 
    the license amendment to use an alternative source term. A detailed 
    rationale for the use of 0.25 Sv (25 rem) TEDE as an accident dose 
    criterion and the use of the 2-hour exposure period resulting in the 
    maximum dose for future LWRs is provided at 61 FR 65157 (December 11, 
    1996). The same considerations that formed the basis for that rationale 
    are similarly applicable to operating reactors that elect to use an 
    alternative source term. The NRC believes that it is technically 
    appropriate and logical to extend the philosophy of decoupling of 
    design and siting, and the dose criteria established for future LWRs to 
    operating reactors that elect to use an alternative source term.
        The NRC is replacing the current GDC-19 dose criteria for operating 
    reactors that elect to use an alternative source term with a criterion 
    of 0.05 Sv (5 rem) TEDE for the duration of the accident. This 
    criterion is included in Sec. 50.67 as well as in GDC-19 in order to 
    co-locate all of the dose requirements associated with alternative 
    source terms. The bases for the NRC's decision are: first, that the 
    criteria in GDC-19 and that in the final rule are based on a primary 
    occupational exposure limit. Second, the language in GDC-19: ``5 rem 
    whole body, or its equivalent to any part of the body'' is subsumed by 
    the definition of TEDE in Sec. 20.1003 and by the 0.05 Sv (5 rem) TEDE 
    annual limit in Sec. 20.1201(a). Although the weighting factors stated 
    in Sec. 20.1003 for use in determining TEDE differ in magnitude from 
    the weighting factors implied in the 0.3 Sv (30 rem) thyroid criteria 
    used for showing compliance with GDC-19, these differences are the 
    result of improvement in the science of assessing internal exposures 
    and do not represent a reduction in the level of protection. Third, as 
    discussed earlier, the use of TEDE in conjunction with alternative 
    source terms has been deemed appropriate and necessary. Fourth, the use 
    of TEDE for the control room dose criterion is consistent with the use 
    of TEDE in the accident dose criteria for offsite exposure.
        The NRC has not included a ``capping'' limitation, an additional 
    requirement that the dose to any individual organ not be in excess of 
    some fraction of the total as provided for routine occupational 
    exposures. The bases for the NRC's decision are: first, that this non-
    inclusion of a ``capping'' limitation is consistent with the final rule 
    published in December 11, 1996 (61 FR 65157), with regard to doses to 
    persons offsite. Second, the use of 0.05 Sv (5 rem) TEDE as the control 
    room criterion does not imply that this would be an acceptable exposure 
    during emergency conditions, or that other radiation protection 
    standards of Part 20, including individual organ dose limits, might not 
    apply. This criterion is provided only to assess the acceptability of 
    design provisions for protecting control room operators under 
    postulated DBA conditions. The DBA conditions assumed in these 
    analyses, although credible, generally do not represent actual accident 
    sequences but are specified as conservative surrogates to create 
    bounding conditions for assessing the acceptability of engineered 
    safety features. Third, Sec. 20.1206 permits a once-in-a-lifetime 
    planned special dose of five times the annual dose limits. Also, 
    Environmental Protection Agency (EPA) guidance sets a limit of five 
    times the annual dose limits for workers performing emergency services 
    such as lifesaving or protection of large populations.
        Considering the individual organ weighting factors of Sec. 20.1003 
    and assuming that only the exposure from a single organ contributed to 
    TEDE, the organ dose, although exceeding the dose specified in 
    Sec. 20.1201(a), would be less than that considered acceptable as a 
    planned special dose or as an emergency worker dose. The NRC is not 
    suggesting that control room dose during an accident can be treated as 
    a planned special exposure or that the EPA emergency worker dose limits 
    are an alternative to GDC-19 or the final rule. However, the NRC does 
    believe that these provisions offer a useful perspective that supports 
    the conclusion that the organ doses implied by the 0.05 Sv (5 rem) 
    criterion can be considered to be acceptable due to the relatively low 
    probability of the events that could result in doses of this magnitude.
        Although the dose criteria in the final rule supersede the dose 
    criteria in GDC-19, the other provisions of GDC-19 remain applicable.
        There may be technically justifiable implementations of an AST that 
    would not require calculation of the EAB, LPZ, or control room doses. 
    For example, a proposed modification to change the closure time of a 
    containment isolation valve from 2 seconds to 5 seconds may be based on 
    the timing insights of the AST. Although a specific calculation might 
    not be necessary in this case, the licensee is still required to affirm 
    with reasonable assurance that the doses would comply with these stated 
    criteria.
    
    E. 10 CFR Part 50, Appendix A, GDC-19
    
        GDC-19 is changed to include the TEDE dose criterion for control 
    room design for applicants for construction permits, design 
    certifications, and combined licenses that submitted applications after 
    January 10, 1997 (the effective date of the 1996 rulemaking adopting 
    the TEDE criterion), and for those licenses using an alternative source 
    term under Sec. 50.67. The change to GDC-19 addresses the use of 
    alternative source terms at operating reactors and a deficiency 
    identified in the regulatory framework for early site permits, standard 
    design certifications, and combined licenses under Part 52. Sections 
    52.18, 52.48, and 52.81 establish that applications filed under Part 
    52, Subparts A, B, and C, respectively, will be reviewed according to 
    the standards given in 10 CFR Parts 20, 50, 51, 55, 73, and 100 to the 
    extent that those standards are technically relevant to the proposed 
    design. Therefore, GDC-19 is pertinent to applications under Part 52.
        The final rule that became effective on January 10, 1997 (61 FR 
    65157; December 11, 1996), established accident TEDE criteria (in 
    Sec. 50.34) for applicants under Part 52 but did not change the 
    existing control room whole body (or equivalent) dose criterion in GDC-
    19. Thus, exemptions from the dose criteria in the current GDC-19 were 
    necessary in the design certification process for the Westinghouse 
    AP600 advanced LWR in order to use the 0.05 Sv (5 rem) TEDE criterion 
    deemed necessary for use with alternative source terms. Exemptions will 
    arguably be necessary for future applicants for construction permits, 
    design certifications, and combined licenses. This amendment will 
    eliminate the need for these exemptions.
    
    F. Sections 21.3, 50.2, 50.49(b)(1)(i)(C), 50.65(b)(1), and 
    54.4(a)(1)(iii)
    
        These sections are revised to conform with the relocation of 
    accident dose criteria from Sec. 100.11 to Sec. 50.67 for operating 
    reactors that have amended their design bases to use an alternative 
    source term.
    
    [[Page 71998]]
    
    G. Section 50.34
    
        A new footnote to Sec. 50.34 has been added to define what 
    constitutes an accident source term. This new footnote is identical to 
    the existing footnote 1 to Sec. 100.11, and was added to provide for 
    consistency between Parts 50 and 100.
    
    H. Sections 50.34(f)(2)(vii), (viii), (xxvi) and (xxviii)
    
        These paragraphs are revised to replace an explicit reference to 
    the ``TID-14844 source term'' with a more general reference to 
    ``accident source term.'' These changes potentially affect three 
    classes of applicants. The first affected class is comprised of 
    applicants for design certification under Part 52, Subpart B. Section 
    52.47(a)(1)(ii) states that applications for combined licenses must 
    contain, inter alia, ``demonstration of compliance with any 
    technically-relevant portions of the Three Mile Island requirements set 
    forth in Sec. 50.34(f).'' Section 50.34(f) contains several references 
    to the TID-14844 source term. These references were modified to delete 
    the reference to TID-14844. This change makes it clear that applicants 
    for combined licenses should not use the TID-14844 source term but 
    should use the source term in the referenced design certification, or a 
    source term that is justified in the combined license application. The 
    second affected class is comprised of applicants for combined licenses 
    under Part 52, Subpart C. Section 52.79(b) makes the requirements of 
    52.47(a)(1)(i) applicable if a certified design is not referenced. 
    Thus, the combined license applicant is also subject to the 
    requirements of Section 50.34(f).
        The third affected class is the small subset of plants that had 
    construction permits pending on February 16, 1982. With the proposed 
    change, these plants could use either the TID-14844 source term or an 
    alternative source term in their operating license applications.
    
    IV. Draft Regulatory Guide; Issuance, Availability
    
        The Nuclear Regulatory Commission is issuing for public comment a 
    draft of a guide planned for its Regulatory Guide Series. This series 
    has been developed to describe and make available to the public 
    information such as methods acceptable to the NRC staff for 
    implementing specific parts of the Commission's regulations, techniques 
    used by the staff in evaluating specific problems or postulated 
    accidents, and data needed by the NRC staff in its review of 
    applications for permits and licenses. Copies of the draft guide may be 
    obtained as described in Section VI, ``Referenced Documents,'' of these 
    statements of consideration. You may also download copies from the 
    NRC's interactive rulemaking forum website through the NRC home page 
    (http://ruleforum.llnl.gov/cgi-bin/rulemake).
        The draft guide, temporarily identified by its task number DG-1081 
    (which should be mentioned in all correspondence concerning this draft 
    guide) is titled ``Alternative Radiological Source Terms for Evaluating 
    Design Basis Accidents at Nuclear Power Reactors.'' This guide is 
    intended for Division 1, ``Power Reactors.'' This draft guide is being 
    developed to provide regulatory guidance on the implementation of an 
    alternative source term at an operating reactor. The guide addresses 
    issues involving limited or selective implementation of an alternative 
    source term and probabilistic risk assessment (PRA) issues related to 
    plant modifications based on an alternative source term, and provides 
    guidance on the scope and extent of affected design basis accident 
    (DBA) radiological analyses and associated acceptance criteria. The 
    guide includes revised assumptions and methods for each affected DBA in 
    a series of appendices. These appendices supersede the guidance in 
    Regulatory Guides 1.3, 1.4, 1.5, 1.25, and 1.77, and supplement 
    guidance in Regulatory Guide 1.89 for those facilities using an 
    alternative source term.
        The draft guide has not received complete NRC staff review and does 
    not represent an official NRC staff position.
        Previous draft versions of DG-1081 have been made publicly 
    available to support technical interactions with the public. This 
    Federal Register announcement provides an opportunity for the public to 
    provide comments on the DG-1081 guidance. The NRC staff will consider 
    the public comments in its efforts to finalize the regulatory guidance.
        The Commission invites advice and recommendations on the content of 
    the draft regulatory guide. Comments and suggestion are particularly 
    requested on the following questions.
    
    A. Scope of Implementation
    
        1. The guidance provided in the draft regulatory guide is intended 
    to allow licensees the maximum flexibility in pursuing technically 
    justifiable AST implementations provided that a clear, consistent, and 
    logical design basis is maintained. Comments are specifically requested 
    on the following questions.
        A. Does the proposed guidance provide the desired flexibility while 
    providing reasonable assurance that a clear, consistent, and logical 
    design basis will be maintained?
        B. Is there a less complex alternative approach that would provide 
    the desired flexibility while maintaining a clear, consistent, and 
    logical design basis?
        C. Should the Commission allow licensees that have received 
    approval for a selective implementation to extend the AST and the TEDE 
    criteria to other design basis applications (that do not involve 
    reanalysis of the DBA LOCA) under Sec. 50.59 rather than under 
    Sec. 50.67 as currently proposed?
        2. The guidance would allow selective implementation of the 
    characteristics (i.e., the fractions of fission product inventory of 
    the radionuclides released from the reactor fuel, their chemical and 
    physical form, and the timing of their release) of an AST. The 
    Commission believes that implementations based only on the timing 
    insights of an AST may be technically justifiable. The Commission 
    believes that the other combinations may be internally inconsistent. 
    Comments are specifically requested on the following questions.
        A. What other combinations of AST characteristics are technically 
    consistent?
        B. What plant modifications might be based on these combinations?
    
    B. Scope of Re-Analyses
    
        1. The draft regulatory guide provides guidance on the scope of the 
    re-analyses that should be performed to support an AST implementation. 
    Comments are requested on the following questions.
        A. Is the proposed guidance on the scope of re-analyses technically 
    appropriate and clear? How could it be improved?
        B. The guidance allows licensees to disposition certain impacts of 
    an AST on the basis of the NRC staff's re-baselining study. Does this 
    study or other documents provide a sufficient basis for the Commission 
    to generically disposition these impacts?
        2. It may be possible for licensees to demonstrate that the doses 
    from certain affected analyses assessed using the prior source term and 
    dose methodology would be greater than the doses obtained using a 
    proposed AST and the TEDE methodology. The proposed guidance would 
    allow the licensee to disposition these affected analyses without re-
    calculation. Nonetheless, the design basis would now include the 
    approved AST and TEDE criteria. The guidance in the draft regulatory 
    guide would require the licensee to update the calculation to be 
    consistent with the approved AST and dose methodology described in the 
    facility design basis in
    
    [[Page 71999]]
    
    the event of a subsequent re-calculation. Comments are requested on the 
    following questions.
        A. Should the Commission allow licensees to continue to use the 
    prior source term and dose criteria for these analyses and not require 
    that they be updated on subsequent revisions?
        B. If the analyses are not updated, how will licensees assure that 
    the earlier conclusion that the analyses are limiting remains valid 
    following subsequent revisions?
        3. Analyses of the integrated radiation doses for environmental 
    qualification of certain equipment important to safety will be affected 
    by the increased concentration of radioactive cesium in the containment 
    sump water. The Commission has been considering the position that 
    licensees proposing to implement an AST must address all impacts of the 
    proposed implementation, including the impact of the increased cesium 
    concentration. However, the Commission now believes it may be necessary 
    for all operating power reactors to address the postulated increase in 
    the cesium concentration. The Commission will consider this issue as a 
    generic safety issue. Comments are requested on the following 
    questions.
        A. Is there information that should be considered by the Commission 
    in resolving this generic issue?
        B. If the Commission should conclude that there is safety 
    significance but that the costs of implementing corrective actions are 
    not justified on a generic basis, should licensees who are voluntarily 
    proposing to amend their design basis to use an AST be required to 
    address the impact of the increased cesium concentration?
        C. If a licensee proposes a change in the plant configuration that 
    would result in an increase in the integrated dose for one or more 
    components and this licensee is also proposing, or has already 
    implemented an AST, should the re-analysis of the integrated dose be 
    based on that AST or on the prior TID14844 source term?
        Comments may be accompanied by relevant information or supporting 
    data. Written comments may be mailed to: Secretary, U.S. Nuclear 
    Regulatory Commission, Washington, DC 20555-0001, Attention: 
    Rulemakings and Adjudications Staff. Mail Stop O16C1. Copies of 
    comments received may be examined at the NRC Public Document Room, 2120 
    L Street NW., Washington, DC. Comments will be most helpful if received 
    by March 7, 2000.
        You may also provide comments via the NRC's interactive rulemaking 
    website through the NRC home page (http://ruleforum.llnl.gov/cgi-bin/
    rulemake). This site provides the availability to upload comments as 
    files (any format), if your web browser supports that function. For 
    information about the interactive rulemaking website, contact Ms. Carol 
    Gallagher, (301) 415-5905; or by internet electronic mail to 
    cag@nrc.gov. For information about the draft guide, contact Mr. Stephen 
    F. LaVie, (301) 415-1081; Internet electronic mail sfl@nrc.gov.
        Although a time limit is given for comments on this draft guide, 
    comments and suggestions in connection with items for inclusion in 
    guides currently being developed or improvements in all published 
    guides are encouraged at any time.
    
    V. Draft Standard Review Plan Section; Issuance, Availability
    
        The Nuclear Regulatory Commission is issuing for public comment a 
    draft of a new section to NUREG-0800, ``Standard Review Plan.'' 
    Standard review plan (SRP) sections are prepared for the guidance of 
    the Office of Nuclear Reactor Regulation staff responsible for the 
    review of applications to construct and operate nuclear power plants. 
    These documents are made available to the public as part of the 
    Commission's policy to inform the nuclear industry and the general 
    public of regulatory procedures and policies. The draft SRP Section 
    15.0.1, is titled ``Radiological Consequence Analyses Using Alternative 
    Source Terms.'' The SRP section complements draft regulatory guide DG-
    1081. The draft SRP section has not received complete NRC staff review 
    and does not represent an official NRC staff position.
        Copies of the draft SRP section may be obtained as described in 
    Section VI, ``Referenced Documents,'' of these statements of 
    consideration. You may also download copies from the NRC's interactive 
    rulemaking forum website through the NRC home page (http://
    ruleforum.llnl.gov/cgi-bin/rulemake).
        Comments on the content of the draft SRP section are invited. 
    Comments may be accompanied by relevant information or supporting data. 
    Comments should be submitted as described above for the draft 
    regulatory guide. Although a time limit is given for comments on this 
    draft SRP section, comments and suggestions in connection with items 
    for inclusion in SRP sections currently being developed or improvements 
    in all published SRP sections are encouraged at any time.
    
    VI. Referenced Documents
    
        Copies of NUREG-0737, NUREG-0800, NUREG-1465, NUREG/BR-0058, NUREG/
    BR-184, and NUREG/CR-6204 may be purchased from the Superintendent of 
    Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington, 
    DC 20402-9328. Copies also are available from the National Technical 
    Information Service, 5285 Port Royal Road, Springfield, VA 22161. A 
    copy also is available for inspection and copying for a fee in the NRC 
    Public Document Room, 2120 L Street, NW (Lower Level), Washington, DC.
        Single copies of regulatory guides, both active and draft may be 
    obtained free of charge by writing the Reproduction and Distribution 
    Services Section, OCIO, USNRC, Washington DC 20555-0001, or by fax to 
    (301) 415-2289, or by email to distribution@nrc.gov. Active guides may 
    also be purchased from the National Technical Information Service on a 
    standing order basis. Details of this service may be obtained by 
    writing NTIS, 5285 Port Royal Road, Springfield, VA 22161. Copies of 
    active and draft guides are available for inspection or copying for a 
    fee from the NRC Public Document Room at 2120 L Street NW., Washington 
    DC.
        Copies of SECY-94-302, SECY-96-242, SECY-98-154, SECY-98-289, TID-
    14844, and TR-105909 are available for inspection and copying for a fee 
    at the NRC Public Document Room, 2120 L Street, NW. (Lower Level), 
    Washington, DC.
    
    VII. Finding of No Significant Environmental Impact: Availability
    
        The NRC has determined under the National Environmental Policy Act 
    of 1969, as amended, and the NRC's regulations in Subpart A of 10 CFR 
    Part 51, that this regulation is not a major Federal action 
    significantly affecting the quality of the human environment and, 
    therefore, an environmental impact statement is not required. This 
    final rule allows operating reactors to replace the traditional TID-
    14844 source term with a more realistic source term based on the 
    insights gained from extensive accident research activities. The actual 
    accident sequence and progression are not changed; it is the regulatory 
    assumptions regarding the accident that would be affected by the 
    change. The use of an alternative source term alone cannot increase the 
    core damage frequency (CDF) or the large early release frequency (LERF) 
    or actual offsite or onsite radiation doses. An alternative source term 
    could be used to justify changes in the plant design that might have an 
    impact on CDF or LERF or that might increase offsite or onsite doses. 
    Those plant changes that do not
    
    [[Page 72000]]
    
    require prior NRC review and approval pursuant to Sec. 50.59 are not 
    likely to involve any significant increase in environmental impacts. 
    The Sec. 50.59 criteria are sufficiently stringent that any potential 
    change in plant design that could have an adverse environmental impact 
    in all likelihood could not be made by the licensee without prior NRC 
    review and approval. Every plant change that requires NRC review and 
    approval under Sec. 50.59 requires a license amendment and, therefore, 
    the preparation of an environmental assessment to determine whether the 
    proposed change involves any significant environmental impact. Thus, 
    this final rule, by itself, will not result in plant changes that 
    involve any significant increase in environmental impacts. The final 
    rule does not affect non-radiological plant effluents.
        The NRC requested public comments on any environmental justice 
    considerations that may be related to this rule. No public comments 
    relevant to the draft environmental assessment or environmental justice 
    considerations were received. The NRC requested the views of the States 
    on the environmental assessment for this rule. No comments relevant to 
    the draft environmental assessment or environmental justice 
    considerations were received.
        The environmental assessment and finding of no significant impact 
    on which this determination is based are available for inspection at 
    the NRC Public Document Room, 2120 L Street NW. (Lower Level), 
    Washington, DC. Single copies of the environmental assessment and 
    finding of no significant impact are available from Mr. Stephen F. 
    LaVie, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory 
    NRC, Washington, DC 20555-0001, telephone: (301) 415-1081, or by 
    Internet electronic mail to sfl@nrc.gov.
    
    VIII. Paperwork Reduction Act Statement
    
        This final rule increases the burden on licensees by requiring that 
    when seeking to revise their current accident source term in design 
    basis radiological consequence analyses, they apply for an amendment 
    under Sec. 50.90. The public burden for this information collection is 
    estimated to average 609 hours per request. Because the burden for this 
    information collection is insignificant relative to the total burden 
    estimated, Office of Management and Budget (OMB) clearance is not 
    required. Existing requirements were approved by the Office of 
    Management and Budget, approval number 3150-0011.
    
    Public Protection Notification
    
        If an information collection does not display a currently valid OMB 
    control number, the NRC may not conduct or sponsor, and a person is not 
    required to respond to, the information collection.
    
    IX. Regulatory Analysis
    
        The Commission has prepared a regulatory analysis on this 
    regulation. Interested persons may examine a copy of the regulatory 
    analysis at the NRC Public Document Room, 2120 L Street NW. (Lower 
    Level), Washington, DC. Single copies of the analysis are available 
    from Mr. Stephen F. LaVie, Office of Nuclear Reactor Regulation, U.S. 
    Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 
    (301) 415-1081, or by Internet electronic mail to sfl@nrc.gov.
    
    X. Regulatory Flexibility Act Certification
    
        As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 
    605(b), the Commission certifies that this regulation will not have a 
    significant economic impact on a substantial number of small entities. 
    This regulation will affect only the licensing and operation of nuclear 
    power plants. The companies that own these plants do not fall within 
    the definition of ``small entities'' found in the Regulatory 
    Flexibility Act or within the size standards established by the NRC 
    (April 11, 1995; 60 FR 18344).
    
    XI. Backfit Analysis
    
        The NRC has determined that the backfit rule in 10 CFR 50.109 does 
    not apply to this final rule, and that a backfit analysis is not 
    required for this rulemaking because these amendments do not involve 
    any provisions that would impose backfits as defined in 10 CFR 
    50.109(a)(1). This final rule amends the NRC's regulations by 
    establishing alternate requirements that may be voluntarily adopted by 
    licensees, and makes changes to the regulations to conform them to a 
    1996 rulemaking.
    
    XII. Small Business Regulatory Enforcement Fairness Act
    
        In accordance with the Small Business Regulatory Fairness Act of 
    1996, the NRC has determined that this action is not a major rule and 
    has verified this determination with the Office of Information and 
    Regulatory Affairs, Office of Management and Budget.
    
    XIII. National Technology Transfer and Advancement Act
    
        The National Technology Transfer Act of 1995, Pub. L. 104-113, 
    requires that Federal agencies use technical standards that are 
    developed or adopted by voluntary consensus standards bodies unless the 
    use of such a standard is inconsistent with applicable law or otherwise 
    impractical. In this final rule the NRC is establishing a government-
    unique standard in Section 50.67(b)(2) by specifying accident radiation 
    dose criteria. These criteria were issued for use by future license 
    applicants by an earlier rulemaking (61 FR 65157, December 11, 1996) 
    and, by this final rule, are being applied to operating reactors that 
    voluntarily use an alternative source term. No voluntary consensus 
    standard has been identified that could be used instead of the 
    government-unique standard.
    
    List of Subjects
    
    10 CFR Part 21
    
        Nuclear power plants and reactors, Penalties, Radiation protection, 
    Reporting and recordkeeping requirements.
    
    10 CFR Part 50
    
        Antitrust, Classified information, Criminal penalties, Fire 
    protection, Intergovernmental relations, Nuclear power plants and 
    reactors, Radiation protection, Reactor siting criteria, Reporting and 
    recordkeeping requirements.
    
    10 CFR Part 54
    
        Administrative practice and procedure, Age-related degradation, 
    Backfitting, Classified information, Criminal penalties, Environmental 
    protection, Nuclear power plants and reactors, Reporting and 
    recordkeeping requirements.
    
        For the reasons noted in the preamble and under the authority of 
    the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
    Act of 1974, as amended; and 5 U.S.C. 553; the NRC is proposing the 
    following amendments to 10 CFR Parts 21, 50, and 54:
    
    PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE
    
        1. The authority citation for Part 21 continues to read as follows:
    
        Authority: Sec. 161, 68 Stat. 948, as amended, sec. 234, 83 
    Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2953 (42 U.S.C. 
    2201, 2282, 2297f); secs. 201, as amended, 206, 88 Stat. 1242, as 
    amended, 1246 (42 U.S.C. 5841, 5846).
    
        Section 21.2 also issued under secs. 135, 141, Pub. L. 97-425, 
    96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
    
        2. Section 21.3 is amended by republishing the introductory text 
    and revising paragraph (1)(i)(C) of the
    
    [[Page 72001]]
    
    definition of Basic Component to read as follows:
    
    
    Sec. 21.3  Definitions.
    
        As used in this part:
        Basic component. (1)(i) * * *
        (C) The capability to prevent or mitigate the consequences of 
    accidents which could result in potential offsite exposures comparable 
    to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or 
    Sec. 100.11 of this chapter, as applicable.
    * * * * *
    
    PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
    FACILITIES
    
        3. The authority citation for Part 50 continues to read as follows:
    
        Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 
    Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 
    83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 
    2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 
    Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).
    
        Section 50.7 also issued under Pub. L. 95-9601, sec. 10, 92 
    Stat. 2951 (42 U.S.C. 5851). Section 50.10 also issued under secs. 
    101, 185, 68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, 
    Pub. L. 91-9190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 
    50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as 
    amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 
    also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 
    50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 
    91-9190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 
    also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 
    50.58, 50.91, and 50.92 also issued under Pub. L. 97-9415, 96 Stat. 
    2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 
    Stat. 939 (42 U.S.C. 2152). Sections 50.80-50.81 also issued under 
    sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also 
    issued under sec. 187, 68 Stat. 955 (42 U.S.C 2237).
    
        4. Section 50.2 is amended by republishing the introductory text 
    and revising paragraph (1)(iii) of the definition of Basic component, 
    and by adding in alphabetical order the definition for Source term to 
    read as follows:
    
    
    Sec. 50.2  Definitions.
    
        As used in this part,
    * * * * *
        Basic component * * *
        (1) * * *
        (iii) The capability to prevent or mitigate the consequences of 
    accidents which could result in potential offsite exposures comparable 
    to those referred to in Sec. 50.34(a)(1), Sec. 50.67(b)(2), or 
    Sec. 100.11 of this chapter, as applicable.
    * * * * *
        Source term refers to the magnitude and mix of the radionuclides 
    released from the fuel, expressed as fractions of the fission product 
    inventory in the fuel, as well as their physical and chemical form, and 
    the timing of their release.
    * * * * *
        5. Section 50.34 is amended by revising paragraphs (f)(2)(vii), 
    (viii), (xxvi), and (xxviii) to read as follows:
    
    
    Sec. 50.34  Contents of applications; technical information.
    
    * * * * *
        (f) * * *
        (2) * * *
        (vii) Perform radiation and shielding design reviews of spaces 
    around systems that may, as a result of an accident, contain accident 
    source term \11\ radioactive materials, and design as necessary to 
    permit adequate access to important areas and to protect safety 
    equipment from the radiation environment. (II.B.2)
        (viii) Provide a capability to promptly obtain and analyze samples 
    from the reactor coolant system and containment that may contain 
    accident source term \11\ radioactive materials without radiation 
    exposures to any individual exceeding 5 rems to the whole body or 50 
    rems to the extremities. Materials to be analyzed and quantified 
    include certain radionuclides that are indicators of the degree of core 
    damage (e.g., noble gases, radioiodines and cesiums, and nonvolatile 
    isotopes), hydrogen in the containment atmosphere, dissolved gases, 
    chloride, and boron concentrations. (II.B.3)
    * * * * *
        (xxvi) Provide for leakage control and detection in the design of 
    systems outside containment that contain (or might contain) accident 
    source term \11\ radioactive materials following an accident. 
    Applicants shall submit a leakage control program, including an initial 
    test program, a schedule for re-testing these systems, and the actions 
    to be taken for minimizing leakage from such systems. The goal is to 
    minimize potential exposures to workers and public, and to provide 
    reasonable assurance that excessive leakage will not prevent the use of 
    systems needed in an emergency. (III.D.1.1)
    * * * * *
        (xxviii) Evaluate potential pathways for radioactivity and 
    radiation that may lead to control room habitability problems under 
    accident conditions resulting in an accident source term \11\ release, 
    and make necessary design provisions to preclude such problems. 
    (III.D.3.4)
    ---------------------------------------------------------------------------
    
        \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
    
    
    

Document Information

Effective Date:
1/24/2000
Published:
12/23/1999
Department:
Nuclear Regulatory Commission
Entry Type:
Rule
Action:
Final rule.
Document Number:
99-33283
Dates:
January 24, 2000.
Pages:
71990-72002 (13 pages)
RINs:
3150-AG12: Revised Source Term Use at Operating Reactors
RIN Links:
https://www.federalregister.gov/regulations/3150-AG12/revised-source-term-use-at-operating-reactors
PDF File:
99-33283.pdf
CFR: (20)
10 CFR 50.34)
10 CFR 20.1201(a)
10 CFR 50.67(b)(1)
10 CFR 50.67(b)(2)
10 CFR 184
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