Nuclear Power Reactor Source Term
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Introduction
This web page provides a selected collection of information about nuclear power reactor source term for stakeholders considering pre-application engagements with the NRC or developing applications for new nuclear power reactors under Parts 50 and 52. Applicants are solely responsible for providing to the NRC for approval, their specific reactor design source term associated with the analysis and evaluation of the performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the proposed nuclear power reactor facility. Although this web page references NRC guidance documents, the web page itself is only a tool to aid in locating information on development of nuclear power reactor source terms. The web page is not a guidance document that provides an acceptable method for applicants and licensees to comply with regulatory requirements. This repository of information is intended to facilitate user synergies and insights that could be valuable to applicants in the development of the source term for specific reactor designs. An applicant can best use the available information on this web page in concert with pre-application interactions with the NRC to be better prepared to develop high quality licensing submittals associated with a nuclear power reactor source term.
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. Determining source terms is a critical component in the NRC's licensing process, which is based on independent layers of nuclear safety established by the concept of defense in depth. This approach considers design basis accidents (DBAs) to determine the effectiveness of each line of defense. The DBAs establish and confirm the design bases of the nuclear facility to ensure that the specific plant design meets the safety and radiological criteria set forth in the Code of Federal Regulations (CFR). Therefore, specific safety requirements associated with source terms are reflected in regulations, regulatory guides, standard review plans, technical specifications, and license conditions, WASH reports, Technical Information documents, and NUREG documents.
The NRC's mission is to ensure that radioactive material is used in a safe and secure manner. For nuclear power plants, the NRC requires licensees demonstrate that the reactor's engineered safety features will prevent or mitigate a release of radioactive material from the reactor for a range of accidents. To provide an additional measure of safety, the NRC requires the reactor be housed in a containment building or, for non-light-water reactors (non-LWRs) approved functional containment for the unlikely event of a release of radioactive material from the reactor fuel. The NRC further requires licensees demonstrate that, if there were a release from the reactor fuel into the containment building, the resulting doses would be below the NRC's regulatory criteria. For this demonstration, licensees have historically used a release provided in NRC regulatory guidance. For advanced reactor designs, license applicants are developing design-specific releases to use for this demonstration instead of an NRC-guidance-provided release. Additionally, for non-power reactors, information associated with radiation source terms is provided in documents such as NUREG-1537, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors," Safety Reports Series No. 53, "Derivation of the Source Term and Analysis of the Radiological Consequences of Research Reactor Accidents," and at the NRC’s "Non-Power Facilities" web page.
History and evolution of Light Water Reactor (LWR) source term
In 1962, the AEC issued TID-14844, "Calculation of Distance Factors for Power and Test Reactors." In this document, a release of fission products from the core of an LWR into the containment atmosphere ("source term") was postulated for the purpose of calculating off-site doses in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 100, "Reactor Site Criteria." This source term was postulated from an accident that resulted in substantial meltdown of the reactor core, and the fission products assumed released into the containment were based on an understanding of fission product behavior up to that time. In addition to site suitability, the regulatory applications of this source term (in conjunction with the dose calculation methodology) affected the design of a wide range of plant systems. Over the subsequent 30 years, substantial information was developed updating the knowledge about severe LWR accidents and the resulting behavior of the released fission products. Using this information, a more mechanistic source term was developed, and the NRC issued Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design-Basis Accidents at Nuclear Power Reactors," which provided guidance to licensees of operating nuclear power reactors on acceptable applications of alternative source terms other than those based on TID-14844.
The evaluation of LWR source terms to the environment and off-site consequences have evolved to utilize computer models based on mechanistic models of fission product release and transport. These mechanistic models have been integrated into computer codes such as the NRC's RADTRAD and MELCOR codes or the Electric Power Research Institute's Modular Accident Analysis Program (MAAP). The MELCOR and MAAP codes have enabled the development of methods that calculate a mechanistic source term for LWRs. This capability has been utilized in a range of regulatory evaluations and rulemakings, including the recent Containment Protection and Release Reduction rulemaking by both the NRC and industry.
SMRs and non-LWRs
For new reactor design applications, the industry is using mechanistic models of fission product release and transport. In SECY-16-0012, the NRC staff stated that non-LWR applicants can use modern analysis tools to demonstrate quantitatively the safety features of new reactor designs. For example, one small modular reactor (SMR) vendor used the MELCOR code to estimate source terms for its safety analysis report. More recently, non-LWR reactor vendors have submitted topical reports describing their planned use of mechanistic models to estimate source terms.
In the NRC staff requirements memorandum (SRM) to SECY-93-092, the Commission approved the NRC staff's recommendation that source terms for non-LWRs be based upon a mechanistic analysis and that the acceptability of an applicant's analysis will rely on the NRC staff's assurance that the following conditions are met:
- The performance of the reactor and fuel under normal and off-normal conditions is sufficiently well understood to permit a mechanistic analysis. Sufficient data should exist on the reactor and fuel performance through the research, development, and testing programs to provide adequate confidence in the mechanistic approach.
- The transport of fission products can be adequately modeled for all barriers and pathways to the environs, including the specific consideration of containment design. The calculations should be as realistic as possible so that the values and limitations of any mechanism or barrier are not obscured.
- The events considered in the analyses to develop the set of source terms for each design are selected to bound severe accidents and design-dependent uncertainties.
- The design-specific source terms for each accident category would constitute one component for evaluating the acceptability of the design.
SECY-93-092 described a mechanistic source term as "the result of an analysis of fission product release based on the amount of cladding damage, fuel damage, and core damage resulting from the specific accident sequences being evaluated. It is developed using best-estimate phenomenological models of the transport of the fission products from the fuel through the reactor coolant system, through all holdup volumes and barriers, taking into account mitigation features, and finally, into the environs."
Accident consequence related regulation activities
There are a number of ongoing regulatory activities related to source term development for both operating and new reactors. The NRC staff is currently considering changes to dose criteria for the exclusion area boundary, the low population zone distance, and the control room. Other current activities related to source term considerations include a rulemaking on increased enrichment of conventional and accident tolerant fuel designs for LWRs, and a limited-scope rulemaking on alternative physical security requirements for advanced reactors.
For non-LWR designs, the NRC staff has developed a technology-inclusive, risk-informed, performance-based framework for licensing and regulating non-LWRs. This framework is described in Regulatory Guide 1.233. In addition, the NRC is developing a set of regulations called "Part 53" as an alternative regulatory framework for non-LWR designs and other new commercial nuclear plants.
Guidance and information for developing advanced reactor source term
The following are some useful examples of topics, information, videos, presentations, and references associated with nuclear reactor source terms:
| Description |
Date |
| History and Evolution of LWR Source Term |
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March 23, 1962 |
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February 1995 |
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July 2000 |
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October 2023 |
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November 2008 |
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| NRC Analytical Tools |
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January 31, 2020 |
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| SCALE/MELCOR non-LWR source term demonstration project |
- Heat-pipe reactor workshop
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June 29, 2021 |
- High-temperature gas-cooled reactor workshop
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July 20, 2021 |
- Fluoride-salt-cooled high-temperature reactor workshop
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September 14, 2021 |
- Molten-salt-fueled reactor workshop
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September 13, 2022 |
- Sodium-cooled fast reactor workshop
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September 20, 2022 |
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Coming in 2025 |
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| SCALE/MELCOR non-LWR fuel cycle demonstration project |
- High-temperature gas-cooled reactor fuel cycle workshop
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February 28, 2023 |
- Sodium-cooled fast reactor fuel cycle workshop
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September 20, 2023 |
- Molten salt reactor fuel cycle workshop
- Slides
- Video Recording
- SCALE Report
- MELCOR Report
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July 11, 2024 |
- Microreactor fuel cycle workshop
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Coming in 2025 |
- Non-LWR Fuel Cycle Scenarios for SCALE and MELCOR Modeling Capability Demonstration
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December 15, 2023 |
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| Light Water SMR Design Certification Source Term Approach |
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December 19, 1994 |
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March 2007 |
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February 2020 |
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August 16, 2019 |
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| Source Term Approach for Non-LWR |
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April 8, 1993 |
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July 30, 1993 |
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June 1995 |
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January 1994 |
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February 15, 2012 |
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| Guidance for Developing Advanced Reactor Source Term (long-term) |
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June 2020 |
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November 2023 |
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| Information on Developing Source for Non-Light Water Reactors |
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January 2020 |
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June 2020 |
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February 2021 |
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| Reports on Reactor Source Term |
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November 2013 |
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February 2012 |
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September 2020 |
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July 21, 2010 |
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June 2011 |
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2008 |
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February 28, 2015 |
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April 29, 2019 |
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October 2016 |
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June 26 – 29, 2017 |
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April 2022 |
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| Presentations on Non-LWR Source Term |
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February 17, 2022
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February 20, 2020 |
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March 27, 2019 |
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November 16, 2018 |
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