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By around 2025, thorium-based fuel cycles are likely to be deployed internationally. States such as China and India are pursuing research, development, and deployment pathways toward a number of co...

A study on the Fukushima Daiichi nuclear power station spent-fuel pool (SFP) at Unit 4 (SFP4) is presented in this paper. We discuss the design characteristics of SFP4 and its decay heat load in detail and provide a model that we developed to estimate the SFP evaporation rate based on the SFP temperature. The SFP level of SFP4 following the March 11, 2011, accident is predicted based on the fundamental conservation laws of mass and energy. Our predicted SFP level and temperatures are in good agreement with measured data and are consistent with Tokyo Electric Power Company evaluation results.

ITER (Latin for the “way”) is an experimental tokamak fusion energy reactor that is being built in Cadarache, France, in collaboration with seven agencies representing China, the European Union, India, Japan, South Korea, the Russian Federation, and the United States. The main objective of ITER is to demonstrate the scientific and technical feasibility of a controlled fusion reaction that will allow the production of approximately 500 MW of fusion power for durations of several hundred seconds. As an experimental facility, ITER is intended to allow the exploration of physics scenarios, to conduct the technological tests essential to the preparation of a fusion reactor, and to demonstrate the favorable safety characteristics of fusion. The ITER tokamak cooling water system (TCWS) consists of several separate systems to cool the major ITER components—the divertor/limiter, the first wall blanket, the vacuum vessel, and the neutral beam injector. The ex-vessel part of the TCWS provides a confinement function for tritium and activated corrosion products in the cooling water. The vacuum vessel system also has a functional safety requirement regarding the residual heat removal from in-vessel components. A preliminary hazards assessment (PHA) was performed for a better understanding of the hazards, initiating events, and defense-in-depth mechanisms associated with the TCWS. The PHA was completed using the following steps. (1) Hazard identification . Hazards associated with the TCWS were identified including radiological/chemical/electromagnetic hazards and physical hazards (e.g., high voltage, high pressure, high temperature, and falling objects). (2) Hazard categorization . Hazards identified in the first step were categorized as to their potential for harm to the workers, the public, and/or the environment. (3) Hazard evaluation . The design was examined to determine initiating events that might occur and that could expose the public, the environment, or workers to the hazard. In addition, the system was examined to identify barriers that prevent exposure. Finally, consequences to the public or workers were qualitatively assessed should the initiating event occur and one or more of the barriers fail. Frequency of occurrence of the initiating event and subsequent barrier failure was qualitatively estimated. (4) Accident analyses . A preliminary hazards analysis was performed on the conceptual design of the TCWS. As the design progresses, a detailed accident analysis will be performed in the form of a failure modes and effects analysis. The results of the PHA indicated that the principal hazards associated with the TCWS were those associated with radiation. These were low compared to hazards associated with nuclear fission reactors and were limited to potential exposure to the on-site workers if appropriate protective actions were not used. However, the risk to the general public off-site was found to be negligible even under worst case accident conditions.

This Resource Letter is intended to summarize the status of nuclear power in the world today, prospects of significant expansion of nuclear power over the next several decades, the planning of and forecasts for the addition of new power reactors, and issues surrounding the addition of these new reactors. Owing to the breadth of this subject, the list of references includes journal articles, web pages, and reports to guide the reader on the subject. The subject of nuclear power and its related issues are dynamic, so the most current information is likely to be found on reputable websites.

A hydrodynamics model has been developed to study extreme deformation of the space reactor system impacting on the ground with a high velocity. Two-dimensional geometry models for a monolithic core and a pinned core reactor have been developed with dynamic material models, including the material constitutive models and the equation-of-state models. Calculations have been performed for the reactor impacting onto dry sand at 230 and 150 m/s. A pinned core has a much larger fraction of gas volume in the reactor core and thus collapses faster than a monolithic core. The 150-m/s impact velocity case reveals that the gas coolant channels survive in a monolithic core even though the reactor is massively deformed. In a pinned core, however, most of the gas coolant region collapses with intact or partially collapsed fission product gas cores that are protected by solid UO2 fuel. Sand density varies as it is being compressed. Generally, sand beneath the impacting reactor has a higher density as it is compressed. In addition to consideration of global criticality, it is necessary to investigate local criticality. Because of nonuniform distribution of the gas coolant channels in a deformed monolithic core for the 230-m/s impact velocity case, it may bemore » possible to induce criticality locally in those regions where collapse is more severe. It is not straightforward to make an engineering judgment based solely on impact analysis regarding which core concept is more susceptible to criticality events. The current impact study reveals that a pinned core reactor collapses faster than a monolithic core reactor. A reactor that collapses faster is thought to be more susceptible to producing a criticality. However, a monolithic core reactor with much higher mass and kinetic energy develops much higher compaction in the dry sand beneath the reactor. This means that it is expected to better reflect fast neutrons from the bottom boundary where the sand density for a monolithic core impact becomes much higher than for a pinned core impact. It is strongly recommended that neutronics calculations be performed to determine the susceptibility of criticality for the massively deformed nuclear reactors including appropriate reflecting boundary conditions.« less

In summary, this preliminary WBS serves as an initial basis for the capital cost component of the economic analysis of SMRs. This preliminary WBS comes from the known WBS for existing, large nuclear power plants and develops the methodology for accounting for the anticipated differences between the current large plants and the projected SMR designs.

Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

This technical report was generated as a product of the Supervisory Control for Multi-Modular Small Modular Reactor (SMR) Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (AdvSMR) Research and Development Program of the US Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular AdvSMR plants. This research activity advances the state of the art by incorporating real-time, probabilistic-based decision-making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides background information on the state of the art of automated decision-making, including the description of existing methodologies. It then presents a description of a generalized decision-making framework, upon which the supervisory control decision-making algorithm is based. The probabilistic portion of automated decision-making is demonstrated through a simple hydraulic loop example.

An integrated simulation platform that couples probabilistic analysis-based tools with model-based simulation tools can provide valuable insights for reactive and proactive responses to plant operating conditions. The objective of this work is to demonstrate the benefits of a partial implementation of the Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Framework Specification through the coupling of advanced PRA capabilities and accurate multi-physics plant models. Coupling a probabilistic model with a multi-physics model will aid in design, operations, and safety by providing a more accurate understanding of plant behavior. This represents the first attempt at actually integrating these two types of analyses for a control system used for operations, on a faster than real-time basis. This report documents the development of the basic communication capability to exchange data with the probabilistic model using Reliability Workbench (RWB) and the multi-physics model using Dymola. The communication pathways from injecting a fault (i.e., failing a component) to the probabilistic and multi-physics models were successfully completed. This first version was tested with prototypic models represented in both RWB and Modelica. First, a simple event tree/fault tree (ET/FT) model was created to develop the software code to implement the communication capabilities between the dynamic-link librarymore » (dll) and RWB. A program, written in C#, successfully communicates faults to the probabilistic model through the dll. A systems model of the Advanced Liquid-Metal Reactor–Power Reactor Inherently Safe Module (ALMR-PRISM) design developed under another DOE project was upgraded using Dymola to include proper interfaces to allow data exchange with the control application (ConApp). A program, written in C+, successfully communicates faults to the multi-physics model. The results of the example simulation were successfully plotted.« less

Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

The purpose of this document is to provide an initial, focused reference to the safety characteristics of and a licensing approach for Fluoride-Salt-Cooled High-Temperature Reactors (FHRs). The document does not contain details of particular reactor designs nor does it attempt to identify or classify either design basis or beyond design basis accidents. Further, this document is an initial attempt by a small set of subject matter experts to document the safety and licensing characteristics of FHRs for a larger audience. The document is intended to help in setting the safety and licensing research, development, and demonstration path forward. Input from a wider audience, further technical developments, and additional study will be required to develop a consensus position on the safety and licensing characteristics of FHRs. This document begins with a brief overview of the attributes of FHRs and then a general description of their anticipated safety performance. Following this, an overview of the US nuclear power plant approval process is provided that includes both test and power reactors, as well as the role of safety standards in the approval process. The document next describes a General Design Criteria (GDC)–based approach to licensing an FHR and provides an initial draft setmore » of FHR GDCs. The document concludes with a description of a path forward toward developing an FHR safety standard that can support both a test and power reactor licensing process.« less

This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOEmore » associated with Applied Technology and Knowledge Management.« less

During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

Biomass-based ethanol and nuclear power are two viable elements in the path to U.S. energy independence. Numerous studies suggest nuclear power could provide a practical carbon-free heat source alternative for the production of biomass-based ethanol. In order for this coupling to occur, it is necessary to examine the interfacial requirements of both nuclear power plants and bioethanol refineries. This report describes the proposed characteristics of a small cogeneration nuclear power plant, a biochemical process-based cellulosic bioethanol refinery, and a thermochemical process-based cellulosic biorefinery. Systemic and interfacial issues relating to the co-location of either type of bioethanol facility with a nuclear power plant are presented and discussed. Results indicate future co-location efforts will require a new optimized energy strategy focused on overcoming the interfacial challenges identified in the report.

This paper summarizes an assessment of the regulatory framework and requirements for licensing a liquid metal reactor (LMR) for use in transmuting actinides, which was performed for the U.S. Department of Energy (DOE) Advanced Fuel Cycle Initiative (AFCI). Since the LMR designs currently under consideration are sodium-cooled, the assessment identifies and discusses requirements, issues, and topics important to the licensing process in general and those specific to sodium-cooled LMRs, as well as licensing options and associated recommendations. The goal of the regulatory framework assessment was to clarify and evaluate requirements that support the development of safe and cost-effective LMR designs. The scope of the assessment included an analysis of past and present licensing practices as well as an examination of possible future regulatory activities needed to support licensing LMR designs. Because this assessment included the identification of potentially problematic areas, a review of the past LMR licensing efforts was performed. Both technical and regulatory issues were identified and recommendations were made to address important issues. A review of the current regulatory framework for licensing a commercial reactor and the associated licensing schedules was performed as part of the assessment. In addition, specific options proposed by the U.S. Nuclear Regulatory Commission (NRC) for licensing an LMR were also assessed with regard to their potential impacts on different stakeholders, which include the NRC, DOE, industry, and the public. In addition to the licensing of a commercial LMR, the assessment also identifies and evaluates licensing options for an LMR prototype. The regulatory assessment supports a conclusion that a safe, licensable LMR design is fully feasible. The knowledge applied in the LMR design will be reinforced by past experience and available technology. The licensing of an LMR is expected to be manageable, notwithstanding the uncertainties associated with regulatory, technical, and other issues. With forward-looking planning, effective management, and adequate resources, the process of obtaining a license for an LMR would be greatly facilitated.

The neutron detection efficiency is a parameter required in the measurement of reactivity by the modified source technique. The direct solution of the detection efficiency at a perturbed state is costly. To solve for this, a particular variational functional, the Lewins' type variational functional, is presented. The functional is a ratio of two other functionals, each dealing with a reaction rate. The evaluation of this particular functional was done by treating the numerator and the denominator functionals separately. This leads to three flux equations, one for forward flux, and two for adjoint fluxes. The advantages of this formulation over, and the equivalence of this formulation to, the conventional functional presented in the literature are described in detail. The flexibility of the proposed functional is demonstrated by using it to estimate the detection efficiency with four different methods: variational interpolation, conventional variational, variational extrapolation, and multi- reference-state variational. Results are presented for one-dimensional and two- dimensional problems. All results are compared with direct calculations. In all cases, the results show that the variational interpolational method and the multi- reference-state variational method are efficient and practically acceptable.

Scanning and transmission electron microscopy were used to study the progressive disruption of Merino wool during the vigorous agitation of the fibres in aqueous 10J0 (w Iv) solutions of sodium dodecylsulfate (SDS). In contrast to the general disruption observed when wool was vigorously agitated in formic acid, the cuticle was slowly stripped from the fibre with virtually no release of cortical material unless prolonged periods of agitation were used. A similar type of disruption took place in aqueous 10J0 (w Iv) solutions of cetyltrimethylammonium bromide (CETAB) and Triton X-lOO. After the agitation in 10J0 (w/v) SDS solution, the released cuticle fragments and the remaining fibres were examined. Only a minority of the cell portions constituting the cuticle fragments had been cleaved within the endocuticle. Often, the fragments included portions from more than one cuticle cell, with the cell junctions still intact. An understanding of the disruptive process was facilitated by the frequent observation, on residual fibres, of low ridges on exposed underlying cuticle cells. These low ridges corresponded with the distal edges of the originally overlying cuticle cells. Amino-acid analysis and scanning electron microscopy performed on preparations of cuticle obtained in solutions of the above detergents and in formic acid indicated close similarities between all of the cuticle preparations.