Use of Alternative Source Terms at Operating Reactors : LUSENET : TimeBomb 2000 (Y2000) : One Thread


[Federal Register: December 23, 1999 (Volume 64, Number 246)] [Rules and Regulations] [Page 71990-72002] From the Federal Register Online via GPO Access [] [DOCID:fr23de99-4]

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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.


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


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

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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

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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. ---------------------------------------------------------------------------

\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. ---------------------------------------------------------------------------

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

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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

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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. ---------------------------------------------------------------------------

\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

-- stillwatching@fed.reg (, December 23, 1999


stillwatching.... don't suppose you would care to provide your assessment of the importance of this for those of us who are still a quart low on our coffee requirement?

-- Linda (, December 23, 1999.

This seems to be saying if you need to make changes to your reactor, you no longer have to supply a forest worth of paper filled with analysis.

-- David Holladay (, December 23, 1999.

Incorrect. You still need all the analysis, but you're allowed to do it with a different set of assumptions.

-- former nuke (, December 23, 1999.

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