Regulatory Technology Development Plan Sodium Fast Reactor. Mechanistic Source Term Development
- Argonne National Lab. (ANL), Argonne, IL (United States)
Construction and operation of a nuclear power installation in the U.S. requires licensing by the U.S. Nuclear Regulatory Commission (NRC). A vital part of this licensing process and integrated safety assessment entails the analysis of a source term (or source terms) that represents the release of radionuclides during normal operation and accident sequences. Historically, nuclear plant source term analyses have utilized deterministic, bounding assessments of the radionuclides released to the environment. Significant advancements in technical capabilities and the knowledge state have enabled the development of more realistic analyses such that a mechanistic source term (MST) assessment is now expected to be a requirement of advanced reactor licensing. This report focuses on the state of development of an MST for a sodium fast reactor (SFR), with the intent of aiding in the process of MST definition by qualitatively identifying and characterizing the major sources and transport processes of radionuclides. Due to common design characteristics among current U.S. SFR vendor designs, a metal-fuel, pool-type SFR has been selected as the reference design for this work, with all phenomenological discussions geared toward this specific reactor configuration. This works also aims to identify the key gaps and uncertainties in the current knowledge state that must be addressed for SFR MST development. It is anticipated that this knowledge state assessment can enable the coordination of technology and analysis tool development discussions such that any knowledge gaps may be addressed. Sources of radionuclides considered in this report include releases originating both in-vessel and ex-vessel, including in-core fuel, primary sodium and cover gas cleanup systems, and spent fuel movement and handling. Transport phenomena affecting various release groups are identified and qualitatively discussed, including fuel pin and primary coolant retention, and behavior in the cover gas and containment. Radionuclides released from a primary sodium fire are also considered as potential sources. Any available experimental data and pertinent results relevant to the aforementioned phenomena are discussed, and operating incidents at domestically operated facilities are also examined. Considering the extensive range of phenomena affecting the release of radionuclides, the existing state of knowledge generally appears to be substantial, and may be sufficient in most areas. For core damage accidents, high retention rates should be expected within the fuel matrix and primary sodium coolant for all radionuclides other than the noble gases. These factors greatly reduce the magnitude of possible radionuclide release to the environment. Several possible gaps within the knowledgebase were identified during this effort. First, there are uncertainties with regard to radionuclide release from metal fuel in the molten state. Another knowledge gap appears in the available thermodynamic data regarding the behavior of lanthanides and actinides in liquid sodium. While not necessarily a phenomenological knowledge gap, a determination of the data requirements for MST development should be formally made prior to the expenditure of significant future research efforts. That is, if additional experimentation is performed in support of MST development, it is important to identify the proper quality assurance requirements for licensing
- Research Organization:
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE), Office of Nuclear Reactor Technologies, Advanced Reactor Technologies (ART)
- DOE Contract Number:
- AC02-06CH11357
- OSTI ID:
- 1179443
- Report Number(s):
- ANL-ART-3; 79581
- Country of Publication:
- United States
- Language:
- English
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