Due to the urgency around climate change and associated goals for clean energy transition, as well as BEA's role described above, numerous stakeholders have asked for BEA's thoughts

This work provides a summary of selected experimental capabilities being developed to support nonnuclear testing and demonstration of technology in support of microreactors under the U.S. Department of Energy's (DOE's) Microreactor Program. Major capabilities include the Single Primary Heat Extraction and Removal Emulator (SPHERE) and the Microreactor Agile Non-nuclear Experimental Test Bed (MAGNET). The SPHERE facility allows for controlled testing of the steady-state and transient heat rejection capabilities of a single heat pipe using electrical heaters that simulate nuclear heating. The facility is capable of monitoring axial temperature profiles along the heat pipe and surrounding test articles during startup, steady-state operation, and transients. Instrumentation includes noncontact infrared thermal imaging, surface thermocouples, spatially distributed fiber optic temperature and strain sensors, electrical power meters, and a water-cooled, gas-gap calorimeter for quantifying heat rejection from the heat pipe. The facility can be operated under both vacuum and inert-gas conditions. The MAGNET facility is a large-scale, 250-kW electrically heated microreactor test bed to enable nonnuclear experimental evaluation of a variety of microreactor concepts. It can be supplied to electrically heat a scaled section of a microreactor and further test the capabilities of heat rejection systems. The initial MAGNET experiments will support technology maturation and reduce uncertainty and risk associated with the design, operation, and deployment of monolithic heat pipe-based reactors. However, this test bed can broadly be applied to multiple microreactor concepts to evaluate a wide range of thermal-hydraulic and structural phenomena such as interface coupling with power conversion units and other collocated systems. MAGNET can evaluate integral thermomechanical effects during electrical heating of an array of heat pipes in a larger test article. Examples of initial testing will include thermal stresses in the monolith and the impact of debonding of a heat pipe from the core block and how that failure could impact surrounding heat pipes, i.e., understanding the potential for cascading failure. This work also discusses some modeling capabilities that can support experiment design, analysis, and interpretation, including the heat pipe code Sockeye and a comparison of thermal-structural simulations performed using ABAQUS and STAR-CCM+.

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Various nuclear fuel cycle options have been proposed by developers of advanced nuclear energy systems. Motivations for these options, compared to the current light-water-cooled reactors used in the United States, have included more efficient use of resources, waste reduction, greater safety and security, and lower costs. To provide indications on which of the fuel cycle options are promising for future use in order to yield significant benefits in performance and to identify potential challenges, an assessment of options has been conducted using a consistent set of criteria and associated metrics. Findings of the comprehensive assessment of nuclear fuel cycle options are summarized in this chapter. R&D requirements for implementing the most promising options are identified.

Demand for clean and resilient energy has led to new and advancing frontiers of energy development in nuclear technology, specifically in the development of microreactors. These miniaturised modular reactors are generally <20 megawatts thermal (MWt) or 10 megawatts electric (MWe) and offer new opportunities to meet energy needs in remote locations and mobile operations. As with the slightly larger small modular reactors (<300 MWe), microreactor development must demonstrate security and safety, as well as economic competitiveness, to be seen as potential opportunities for new applications. Current research focuses on passive safety features, capital costs, reliability, semi-autonomous or autonomous control, cyber informed design, digital twins and non-proliferation. This paper focuses specifically on microreactor cyber informed design and cyber risk. An overview of microreactor technology provides a basis for examining the cyber nuclear playing field, with an emphasis on the USA. Frameworks for evaluating cyber security threats, and thereby designing for them, are reviewed. Recommendations follow with ideas for future research.

The accurate modeling and simulation of nuclear reactor designs depends greatly on the ability to couple differing sets of physics together. Current coupling techniques most often use a fixed-point, or Picard, iteration scheme in which each set of physics is solved separately, and the resulting solutions are passed between each solver. In the work presented here, two different coupling techniques are investigated: a Jacobian-Free Newton-Krylov (JFNK) approach and a new methodology called Coarse Mesh Finite Difference Coupling (CMFD-Coupling). In this work, both techniques were applied to the low-order CMFD system of equations. This allows for the multiphysics feedback effects to be captured on the low-order system without having to perform a neutron transport solve. The JFNK and CMFD-Coupling approaches were implemented in the MPACT neutron transport code, which is being developed for the Consortium for Advanced Simulation of Light Water Reactors (CASL). These methods were tested on a wide range of practical reactor physics problems, from a single 3D fuel pin to a massively parallel 3D full core problem. When coupled neutronics-thermal hydraulics problems were investigated with both JFNK and CMFD-Coupling, it was concluded that CMFD-Coupling outperformed JFNK in terms of both accuracy and runtime for every problem. When applied to large full core problems with multiple sources of strong feedback enabled, CMFD-Coupling reduced the overall number of transport sweeps required for convergence. (C) 2019 Elsevier Ltd. All rights reserved.

Since commercialization in the mid-1950s, nuclear power has been widely deployed and provides an important clean source of energy, primarily through the generation of electricity. There are currently 449 commercial nuclear power reactors operating in 31 countries with over 390 GWe total capacity, with 99 operating units in the United States with a total generating capacity of about 100 GWe. Nuclear power plants provide over 11% of the world’s electricity production and, as of 2016, thirteen countries rely on nuclear energy to supply at least one-quarter of their total electricity. Virtually all of that power is generated with water-cooled reactor systems. A number of advanced nuclear energy system concepts have been developed to provide improved efficiency, greater fissile fuel utilization, reduced high level waste generation, and increased margins of safety over today’s water-cooled systems. However, there are a number of technical challenges that must be addressed to enable the broad deployment of these systems from concept to the electrical grid. This Technical Document is written to provide workshop participants with an overview of the current status of a number of advanced reactor systems: Key issues for advanced nuclear energy, including coolants, materials in extremes (irradiation, coolant, and/or high temperature) and interfaces; Nuclear fuel technologies for key reactor types and limitations in testing and design; State of science and understanding for nuclear systems, including characterization, testing, and modeling and simulation tools. For each of these technical areas, this document describes the technological challenge, current state of the art for the technology, and barriers to advanced nuclear concepts.

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclide transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. These results provide confidence in VERA-CS's capability to perform high-fidelity calculations for practical PWR reactor problems.

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the U.S. Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions, and advanced reactor technology needs were identified based on recent DOE and international studies. A broad team of stakeholders from industry, academia, and government was assembled to develop a comprehensive set of goals, criteria, and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study's evaluation criteria and metrics, an independent group of experts from industry, universities, and national laboratories scored each of the point designs. Pathways to deployment for concepts of varying technical maturities were estimated for the different demonstration systems with regard to cost, schedule, and possible licensing approaches. This study also presents the trade-offs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development and to serve potential secondary missions.The main findings of the study indicate the following: (1) for industrial process heat supply, a high-temperature gas-cooled reactor is the best choice because of the high outlet temperature of the reactor and its strong passive and inherent safety characteristics; (2) for resource utilization and waste management, a sodium-cooled fast reactor (SFR) is best because of the use of a fast flux to destroy actinides; (3) to realize the advantages of a promising but less-mature technology, a fluoride salt-cooled high-temperature reactor and a lead-cooled fast reactor fare about the same; (4) for fulfilling the needs of a materials test reactor, a SFR is considered best because of its ability to produce high fast flux, incorporate test loops, and provide additional large volumes for testing.

The applicability of a Jacobian-Free Newton-Krylov (JFNK) method for use in coupled reactor problems is examined. The ability for JFNK to simultaneously solve coupled systems of equations offers the potential for significant computational speedup over traditional solver methods. Simplified 1D and infinite homogeneous medium problems were devised to test the effectiveness of JFNK on coupled problems. These problems demonstrated that updating the absorption cross section from within the JFNK solver can decrease computational times significantly. Additionally, a JFNK eigenvalue solver was fully implemented into the neutron transport code MPACT. Problems tested using this solver showed that JFNK offers, at worst, comparable performance as long as an appropriate preconditioner is used.