Development Of Advanced Reactors Research Articles

Multiphysics Analysis of PLOFC and DLOFC Accidents in the HTGR-TR

High-temperature, gas-cooled reactors (HTGRs) are promising advanced reactor designs. Leveraging passive safety features, higher core outlet temperatures, previous commercial operational experience, and potential coupling with nonelectrical systems, HTGRs present many advantages over traditional light water reactors and other advanced reactor designs. The High-Temperature, Gas-Cooled Test Reactor (HTGR-TR) is a point design developed by Idaho National Laboratory as a response to the need for the expansion of U.S. test reactor capabilities. The HTGR-TR was designed to serve as a test reactor for Generation-IV small modular prismatic block HTGRs and to provide irradiation data, two crucial contributions to the development of advanced reactors. As in any reactor design, it is necessary to understand the behavior of the reactor during accident scenarios to determine the overall safety of the design. Pressurized loss of forced cooling (PLOFC) and depressurized loss of forced cooling (DLOFC) are two accidents that can occur in HTGRs and that present challenges to decay heat removal. In order to understand the behavior of the fuel during these accidents, the resulting mechanical behavior and fission product migration in the fuel are evaluated. Using the HTGR-TR design, which utilizes tristructural isotropic (TRISO) fuel, a multiphysics analysis during the PLOFC and DLOFC transients was performed. This study utilizes neutronics, thermal hydraulics, and fuel performance modeling to analyze the behavior of TRISO particles during steady-state operation and the described transients. This analysis aims to characterize the hoop stress and strain during the transients, as well as predict fission product migration and radiological release. Due to the startup conditions of the HTGR-TR, the silicon carbide (SiC) layer for both of the transients and steady-state conditions are initially in compression for both the hoop stress and strain. The DLOFC transient at the beginning of cycle was the most challenging overall to the fuel, having the closest to tensile behavior and the closest to tensile strain. Because the integrity of the SiC layer was not challenged during any of the transients, the release of 90Sr and 137Cs was found to be negligible. The PLOFC transient at the end of cycle resulted in the largest 110mAg release and subsequent radioactivity increase. However, the increase in radioactivity was minor when compared to typical operating conditions due to the relatively large diffusivity of 110mAg though SiC at normal operating temperatures. Highlights include Neutronics and thermal-hydraulic modeling generating steady-state and transient conditions for use in fuel performance analysis.2. Detailed BISON analysis demonstrating the excellent performance of TRISO fuel during PLOFC and DLOFC transients.3. Fuel behavior showing negligible fission product release.

A new design of composite fuel with Zr-coated TRISO particles dispersed in metal matrix

With the development of advanced reactors, it is critical to improve the safety of nuclear reactors. Safety primarily depends on the ability to block fission products and the ability to discharge decay heat of fuel elements. To enhance the ability to retain the release of fission products and achieve high thermal conductivity, a novel fuel pellet was developed, consisting of zirconium (Zr)-coated TRISO particles embedded in a zirconium metal matrix. The zirconium coating was deposited via fluidized-bed chemical vapor deposition (FB-CVD), and the deposition mechanism of Zr coating via FB-CVD was proposed. The consolidation process of Zr-coated TRISO particles in the Zr matrix was achieved via spark plasma sintering (SPS). The microstructure and mechanical properties of Zr coating and the pellet were investigated via SEM and nanoindenter. The results proved the effectiveness of zirconium coating in preventing TRISO fuel particles from contact and cracking during sintering.

Extraction of Neptunium, Plutonium, Americium, Zirconium, and Technetium by Di-(2-Ethylhexyl)-Iso-Butyramide (DEHiBA) at High Metal Loadings

ABSTRACT Increased focus on carbon neutral energy has generated a resurgence of interest in nuclear power, and in particular advanced reactors which are likely to utilize high assay low enriched uranium (HALEU). This in turn could increase the economic attractiveness of recovering still partially enriched uranium from used nuclear fuel. Concomitant to development of advanced reactors, advanced reprocessing schemes should be developed which address the disadvantages to well established reprocessing schemes. The present study focuses on using di-(2-ethylhexyl)-iso-butyramide (DEHiBA) under high metal loading conditions for the reprocessing of used nuclear fuel. The elements examined in the study include the dominant transuranic actinides (Np, Pu, Am) as well as the often-problematic Tc and Zr. By increasing the concentration of the extractant from the more commonly reported 1.0 M – 1.5 M, the extraction of hexavalent actinides is substantially increased, while maintaining effective rejection of tri, tetra, and pentavalent actinides, particularly in the presence of high loadings of uranium. The extraction of Zr by 1.5 M DEHiBA is noted to be negligible by comparison to tributyl phosphate (TBP), however the coextraction of Tc with U is observed to be nominally twice the quantity that is extracted by TBP indicating a need for effective Tc management.

Oh, My Darling Clementine: A Detailed History and Data Repository of the Los Alamos Plutonium Fast Reactor

The Los Alamos Plutonium Fast Reactor, Clementine, provided the first fast neutron spectrum using a metallic plutonium fuel system and liquid mercury coolant. Clementine is often mentioned in the history of fast, metal-fueled, or metal-cooled designs, with little public information regarding the specifications or timeline. This paper contains a synopsis of each report regarding the development, operation, experimental results, failures, and subsequent disassembly of Clementine to explore this unique reactor within the context of modern advanced reactor development.

Intrinsic value of the international benchmark projects, ICSBEP and IRPhEP, for advanced reactor development

The International Criticality Safety Benchmark Evaluation Project, a sanctioned program under the auspices of the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development, has been a highly successful and productive collaboration, now encompassing over 5,000 evaluated experimental benchmarks trusted and relied upon throughout the international nuclear communities. The success of this project led to the development of the International Reactor Physics Experiment Evaluation Project, which is dedicated to the evaluation of benchmark experiment data to sustain current and future reactor physics validation needs. These exemplary programs, and their widely utilized handbooks, serve as gold standards to which other databases strive to emulate. The purpose of the two projects is to preserve modern and legacy experimental data and evaluate it in a standardized handbook format to provide quality benchmarks to support modern and future criticality safety and reactor physics validation. These two projects have often served as the mechanism through which historic and modern neutronics experiments are evaluated and shared across international borders, to best provide unique, high-quality peer-reviewed, and often otherwise unavailable, benchmark data. The contents of these handbooks are utilized not only in validating criticality safety, reactor physics, and advanced reactor calculations, but are used to validate neutronics calculations and nuclear data for most other nuclear applications. This manuscript discusses both international programs and available content to enable advanced reactor design validation.

Considerations for Hydride Moderator Readiness in Microreactors

The emergence of microreactor technology has helped to drive supporting nuclear materials qualification and acceptance processes. One essential component in these small reactors is a solid moderator, which typically consists of metal hydride and cladding. While the behavior and performance of metal-hydride moderators go back to early advanced reactor development for nuclear-powered aviation and space propulsion, there remains a knowledge gap in the understanding of hydrogen transport–related phenomena and irradiation performance for hydride moderators. This impacts the acceptance/qualification of hydride moderators for microreactors. The goal of this technical note is to lay out a potential path forward for advanced moderator qualification and acceptance for designers and developers of microreactors. The proposed approach has benefited from a model microreactor core with the design parameters of a hydride moderator. Based on the model core and design parameters, a simple chart was developed for the major challenges of hydride moderators where potential incidents, causes, effects, and resolutions are described. The relation between the offered resolutions and the maturity of the metal-hydride moderator technology was emphasized using technological readiness. Technological readiness levels (TRLs) were clustered to three sets: physical phenomena related, reactor irradiations, and system demonstration. Some essential needs to fill the knowledge gaps are discussed for physical phenomena–related TRLs. For reactor irradiations, the importance of identifying goals and priorities is stressed to reach certain TRLs. For system demonstration, it is noted that metal-hydride moderator qualification must comply with the overall microreactor design.

Preliminary Siting, Operations, and Transportation Considerations for Licensing Fission Batteries in the United States

Nuclear energy is currently in the spotlight as a future energy source all over the world amid the global warming crisis. In the current state of miniaturization, through the development of advanced reactors, such as small modular reactors (SMRs) and micro-reactors, a fission battery is inspired by the idea that nuclear energy can be used by ordinary people using the “plug-and-play” concept, such as chemical batteries. As for design requirements, fission batteries must be economical, standardized, installed, unattended, and reliable. Meanwhile, the commercialization of reactors is regulated by national bodies, such as the United States (U.S.) Nuclear Regulatory Commission (NRC). At an international level, the International Atomic Energy Agency (IAEA) oversees the safe and peaceful use of nuclear power. However, regulations currently face a significant gap in terms of their applicability to advanced non-light water reactors (non-LWRs). Therefore, this study investigates the regulatory gaps in the licensing of fission batteries concerning safety in terms of siting, autonomous operation, and transportation, and suggests response strategies to supplement them. To figure out the applicability of the current licensing framework to fission batteries, we reviewed the U.S. NRC Title 10, Code of Federal Regulations (CFR), and IAEA INSAG-12. To address siting issues, we explored the non-power reactor (NPR) approach for site restrictions and the permit-by-rule (PBR) approach for excessive time burdens. In addition, we discussed how the development of an advanced human-system interface augmented with artificial intelligence and monitored by personnel for fission batteries may enable successful exemptions from the current regulatory operation staffing requirements. Finally, we discovered that no transportation regulatory challenge exists.

Artificial dissipation model applied to Euler equations for analysis of supersonic flow around a geometric configurations ramp and diffusor type

Very High Temperature Gas Cooled Reactors - VHTGRs are studied by several research groups for the development of advanced reactors that can meet the world's growing energy demand. The analysis of the flow of helium coolant around the various geometries at the core of these reactors through computational fluid dynamics techniques is an essential tool in the development of conceptual designs of nuclear power plants that provide added security. This analysis suggests a close analogy with aeronautical cases widely studied using computational numerical techniques to solve systems of governing equations for the flow involved. The present work consists in using the DISSIPA2D_EULER code, to solve the Euler equations in a conservative form, in two-dimensional space employing a finite difference formulation for spatial discretization using the Euler method for explicit marching in time. The physical problem of supersonic flow of helium gas along a ramp and diffusor configurations is considered. For this, the Jameson and Mavriplis algorithm and the linear artificial dissipation model of Pulliam was implemented. A spatially variable time step is employed aiming to accelerate the convergence to the steady state solution. The main purpose of this work is obtain computational tools for flow analysis through the study the cited dissipation model and describe their characteristics in relation to the overall quality of the solution, as well as obtain preliminary results for the development of computational tools of dynamic analysis of helium gas flow in gas-cooled reactors.

Latest trends of advanced reactor development supporting nuclear innovation (3)

全世界でカーボンニュートラルに向けた動きが加速しているなか，2021年10月にエネルギー基本計画が策定され，安全の確保を大前提として再生可能エネルギー(再エネ)の最大限の導入や水素社会の実現に向けた取組の強化などが掲げられている。三菱重工業では，固有の安全性を有し炉心溶融がなく，水素を大量かつ安定的に供給でき，発電量と水素製造量の調整によって再エネ大量導入時の電力負荷変動に対応できる高温ガス炉コジェネプラントの開発を進めており，今回はその状況について紹介する。

原子力イノベーションを支える最新の新型炉開発の状況

原子力イノベーションを支える最新の新型炉開発の状況

Latest trends of advanced reactor development supporting nuclear innovation (1)

近年，国内外において革新炉導入に向けた活発な動きが見られる。米国では，2020年に「新型炉実証プログラム(ARDP)」を開始し，高速炉，高温ガス炉，マイクロ炉，溶融塩炉，軽水型小型モジュール炉(SMR)のプロジェクトが進められている。カナダ，英国，ロシア，中国でも同様に新型炉開発が盛んになってきている。わが国においても，経済産業省により，2019年度から「社会的要請に応える革新的な原子力技術開発支援事業」が開始され，民間の活力を活かした革新炉の開発に繋がる研究開発の促進に向けて事業成立性に関する調査(フィージビリティスタディ)が進められている。新型炉開発は原子力研究・開発の将来を開拓する技術分野であり，現在，原子力業界に求められているイノベーションの実現を目標としていることから，若い世代を原子力研究・開発へと惹きつけるもっとも魅力的な分野のひとつである。本稿では，主要な開発国を対象に最新の国外の開発動向を紹介するとともに，国内においては2050年カーボンニュートラルに向けた原子力イノベーションに係る革新炉開発の動向を主に紹介する。

Latest trends of advanced reactor development supporting nuclear innovation (2)

Latest trends of advanced reactor development supporting nuclear innovation (2)

原子力イノベーションを支える最新の新型炉開発の状況

原子力イノベーションを支える最新の新型炉開発の状況

Latest trends of advanced reactor development supporting nuclear innovation (5)

Latest trends of advanced reactor development supporting nuclear innovation (5)

Nanoscale 2D semi-conductors – Impact of structural properties on light propagation depth and photocatalytic performance

The performance of photocatalytic materials are largely dictated by the crystalline and optical properties of the semi-conductors. The density of photo-generated electron and hole pairs is greatly influenced by the light penetration in photocatalytic material, leading to specific degradation kinetics. In this work, the relationship between the metal oxide film thickness and the overall materials optical and photocatalytic performances are systematically established for the first time. Thin films of semiconductor metal oxides such as TiO2 and ZnO were prepared by atomic layer deposition (ALD) on stainless steel sputtered silicon wafers. The thickness of the metal oxide thin films was controlled by varying the number of deposition cycles (50–1000 cycles). The fabricated films were fully characterized to examine the change in morphology, roughness, crystallinity, optical and structural properties with varying thickness by several techniques such as Scanning Electron Microscopy (SEM), Atomic Force Microscopy (AFM), X-Ray Diffraction (XRD), Diffuse Reflectance Spectroscopy (DRS) and X-Ray Photo-electro Spectroscopy (XPS). The films were generated to yield very consistent crystallinity, roughness and light absorption properties. Critical thicknesses were observed when a plateau in photocatalytic efficiency was reached at the thickness of 31 nm in the case of the TiO2 and 89 nm thickness for ZnO films. The dependency of the thickness of nanometric ALD films on their photocatalytic efficiency results from the light diffusion and penetration within the material which was investigated through fundamentals and modelling of light-matter interaction in photocatalytic processes. This work establishes a new fundamental understanding of the operation and performance of photocatalysts for further development of advanced reactors and their scale-up.

Looking ahead at reactor development

This paper provides a historical overview of the development of advanced reactors, with a focus on Generation IV reactors and the unique international cooperative R&D framework that was put in place within the Generation IV International Forum. Drawing on the expertise developed at the Nuclear Energy Agency, the paper analyses the challenges for deploying advanced reactors in future energy markets, including evolving market requirements and economic considerations, regulatory challenges, research infrastructure needs and human resource issues. The paper concludes on the role of nuclear research and innovation to ensure the conditions for successful deployment of advanced reactors and competition with alternative technologies.

Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 ) flow loop operating at low pressure (0.2 MPa), at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift) in measuring operational data for extended periods of times, as data collected will be used for code and model verification and validation, one of the key purposes for the loop. The experimental facility will provide a much-needed database for successful development of advanced reactors and provide insight into the needs and challenges in instrumentation for advanced high-temperature reactors.

Management of actinide waste inventories in nuclear phase-out scenarios

The improvement of the “radiological cleanliness” of nuclear energy is a primary goal in the development of advanced reactors and fuel cycles. The multiple recycling of actinides in advanced nuclear systems with fast neutron spectra represents a key option for reducing the potential hazard from high-level waste, especially when the fuel cycle is fully closed. Such strategies, however, involve large inventories of radiotoxic, transuranic (TRU) nuclides in the nuclear park, both in-pile and out-of-pile. The management of these inventories with the help of actinide burners is likely to become an important issue, if nuclear energy systems are eventually phased out, i.e. replaced by other types of energy systems. The present paper compares phase-out scenarios for two transmutation strategies involving fast reactors (FRs) and accelerator-driven systems (ADSs), respectively, operating in symbiosis with conventional light water reactors (LWRs). Particular objectives are to evaluate and compare the TRU reduction performance of the systems as a function of the phase-out time and to determine the appropriate phase-out length for different phase-out criteria. In this connection, an interesting aspect concerns the continuous optimisation of the fuel cycle to counterbalance the reactivity decrease due to the depletion of the fissile isotopes in the fuel. It will be shown that both FRs and ADSs can achieve the goal, provided that the phase-out operation can be continued for about a hundred years.

Thermophysical properties affecting safety and performance of nuclear fuel

Improvement of the nuclear fuel exploitation has been one of the main objectives of reactor technology during the last decades. Today, in view of a sustainable nuclear energy production, development of advanced reactors re-proposes the choice of innovative fuel cycle concepts, in a context of greater expectations and more stringent requirements. From the experience gained in the past, a fuel research and development strategy can be devised, by which selected physical properties are taken as in-pile fuel performance indices.

6.4.3 Risk Informed Design for System Life Cycle

AbstractMany facilty life cycle activities including design, construction, fabrication, inspection and maintenance are evolving from a deterministic to a risk‐informed basis. The risk informed approach uses probabilistic methods to evaluate the contribution of individual system components to total system performance. Total system performance considers both safety and cost considerations including system failure, reliability, and availability. By necessity, a risk‐informed approach considers both the component's life cycle and the life cycle of the system.In the nuclear industry, risk‐based approaches, namely probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA), have become a standard tool used to evaluate the safety of nuclear power plants. Recent studies pertaining to advanced reactor development have indicated that these new power plants must provide improved safety over existing nuclear facilities and be cost‐competitive with other energy sources. Risk‐based approaches, beyond traditional PRA, offer the opportunity to optimize design while considering the total life cycle of the plant in order to realize these goals.The use of risk‐based design approaches in the nuclear industry is only beginning, with recent promulgation of risk‐informed regulations and proposals for risk‐based codes. This paper briefly summarizes the current state of affairs regarding the use of risk‐based approaches in design. Key points to fully realize the benefit of applying a risk‐based approach to nuclear power plant design are then presented and illustrated in a case study. These points are equally applicable to non‐nuclear facilities where optimization for cost competitiveness and/or safety is desired.