China ADS sub-critical experimental assembly—Venus-1 and preliminary experiment

China s accelerator-driven sub-critical system (ADS) sub-critical experimental assembly Venus-1 and the preliminary experiment is presented. The core of Venus-1 is a coupled one of a fast neutron zone and a thermal neutron zone. The fast neutron zone is at the centre of the core and formed by natural uranium fuel. A fast neutron spectrum field can be produced in the fast neutron zone and used for the transmutation of minor actinides (MAs). The thermal neutron zone surrounds the fast neutron zone and is formed by low-enriched uranium fuel. It is a fission zone. An epithermal neutron zone between the fast neutron zone and the thermal neutron zone can be established for the transmutation of long-lived fission products (LLFP). On July 18, 2005, the first fuel element was loaded into the Venus-1 sub-critical assembly and some preliminary experiments about the sub-critical neutronics were performed. The Venus-1 can be driven by an Am-Be source or other steady neutron source (Cf-252, D-D reaction and D-T reaction) to study the effect of the external neutron source with different energies or a D-T pulsed neutron source on the dynamic characteristics.     

China ADS sub-critical experimental assembly-Venus-1 and preliminary experiment

China’s accelerator-driven sub-critical system (ADS) sub-critical experimental assembly—Venus-1 and the preliminary experiment is presented. The core of Venus-1 is a coupled one of a fast neutron zone and a thermal neutron zone. The fast neutron zone is at the centre of the core and formed by natural uranium fuel. A fast neutron spectrum field can be produced in the fast neutron zone and used for the transmutation of minor actinides (MAs). The thermal neutron zone surrounds the fast neutron zone and is formed by low-enriched uranium fuel. It is a fission zone. An epithermal neutron zone between the fast neutron zone and the thermal neutron zone can be established for the transmutation of longlived fission products (LLFP). On July 18, 2005, the first fuel element was loaded into the Venus-1 sub-critical assembly and some preliminary experiments about the sub-critical neutronics were performed. The Venus-1 can be driven by an Am-Be source or other steady neutron source (Cf-252, D-D reaction and D-T reaction) to study the effect of the external neutron source with different energies or a D-T pulsed neutron source on the dynamic characteristics.     

A potential fission electric cell for space reactors

Fission Electric Cells (FEC) able to deliver significant power at high neutron fluxes and to operate without supply were tested in the nuclear reactor of the Central Institute of Physics (CIP) in Bucharest, Romania, at a 3 × 10⁹ neutrons/cm²s thermal neutron flux. The experimental results would make the design of efficient FEC for space reactors feasible.     

第十一届中国核电技术发展高峰论坛会议综述

为促进核电产业安全高效发展,加快落实中国核电“走出去”战略,由上海市核电办公室和中国能源建设集团广东省电力设计研究院有限公司(简称中国能建广东院)联合主办、中国核能行业协会和中国核工业勘测设计协会协办,诺本集团承办的第十一届中国核电技术发展高峰论坛于2016年8月26日在中国能建广东院召开。本次论坛高屋建瓴地探讨了我国核电发展的实践与展望,强调了加强安全监管对助力核电发展的重要性,探索了加强协同创新和技术创新对推动核电产业实现新突破的重要作用,介绍了我国在建的多种堆型核电厂的最新技术特点及工程建设的进展情况,包括具有自主知识产权的三代技术“华龙一号”、我国先进快堆技术、国家科技重大专项高温气冷堆示范工程等,同时就核电装备智能制造应用及通过技术创新提高电站的经济效率等进行了深入地探讨,本次论坛对推动中国核电技术的发展具有重要的意义。     

Computation of the neutron multiplicity moments for research reactor fuels using MCNP6 and SOURCES4c

Fissile material detection and quantification are often necessary for safeguards, nuclear security, and fuel management. Nondestructive assay, neutron, and gamma measurements are reliable means, which can facilitate the detection and estimation of the mass of fissile materials in a broad range of material matrix. Various flavours of neutron measurement are routinely used by facilities (like nuclear reactors, enrichment, and fabrication plants) to quantify fissile material mass and inventory lists. The Monte Carlo code, MCNP6, is used to model several neutron multiplicity measurements. A simulation scenario is set up in MCNP6 using the JCC71 neutron slab counter to obtain the multiplicity moments for fresh and irradiated fuel assemblies from the UMass Lowell Research Reactor (UMLRR) and Worcester Polytechnic Research Reactor (WPIRR). An MCNP6 burnup is initially performed on the fuel types under study to generate used fuel isotopic. The fresh and or used fuel isotopic is then used to produce independent SOURCES4c input tape1 files. SOURCES4c is used to generate (α, n), spontaneous fission spectrum, and the associated neutron emission rates necessary for the various fixed fuel source definitions in MCNP6 calculations. Under the comprehensive safeguards agreements, the International Atomic Energy Agency has the right and obligation to verify that no nuclear material is diverted from peaceful use to nuclear weapons or other nuclear explosive devices. Research reactors are required to be safeguarded facilities under the comprehensive safeguards. Several research efforts have studied the various flavours of neutron measurement for commercial power reactor operating at high power and long burnups; however, not nearly as many studies have been performed with neutron measurements for research reactors operating at relatively lower power and have significantly lower burnup. This work looks to establish the relevant isotopes to overall neutron source rate as well as the process involved in performing a typical neutron multiplicity measurement simulation for a research reactor fuel. The results demonstrate that the single and double moments for Worcester Polytechnic Institute (WPI) and UMLRR fuels can be measured reliably using two JCC71 slab detectors. The moment for the UMLRR and WPIRR fuel (in both fresh and used states) was estimated with a relative error below 0.031 for singles and 0.081 for doubles. The two fresh fuel types cannot be differentiated from each other on the sole basis of neutron analysis. However, fresh and irradiated fuel can be distinguished based on neutron multiplicity measurements.     

Latest research progress for LBE coolant reactor of China initiative accelerator driven system project

China’s accelerator driven subcritical system (ADS) development has made significant progress during the past decade. With the successful construction and operation of the international prototype of ADS superconducting proton linac, the lead-based critical/subcritical zero-power facility VENUS-II and the comprehensive thermal-hydraulic and material test facilities for LBE (lead bismuth eutectic) coolant, China is playing a pivotal role in advanced steady-state operations toward the next step, the ADS project. The China initiative Accelerator Driven System (CiADS) is the next facility for China’s ADS program, aimed to bridge the gaps between the ADS experiment and the LBE cooled subcritical reactor. The total power of the CiADS will reach 10 MW. The CiADS engineering design was approved by Chinese government in 2018. Since then, the CiADS project has been fully transferred to the construction application stage. The subcritical reactor is an important part of the whole CiADS project. Currently, a pool-type LBE cooled fast reactor is chosen as the subcritical reactor of the CiADS. Physical and thermal experiments and software development for LBE coolant were conducted simultaneously to support the design and construction of the CiADS LBE-cooled subcritical reactor. Therefore, it is necessary to introduce the efforts made in China in the LBE-cooled fast reactor to provide certain supporting data and reference solutions for further design and development for ADS. Thus, the roadmap of China’s ADS, the development process of the CiADS, the important design of the current CiADS subcritical reactor, and the efforts to build the LBE-cooled fast reactor are presented.     

Preliminary safety analysis for the SCWR-FQT facility

In recent years, a substantial number of theoretical, numerical and experimental R&D activities are carried out on the supercritical water-cooled reactor (SCWR), which proposed as one of the Generation IV nuclear power plants by the Generation IV International Forum (GIF). A research plan has been proposed by GIF on the designing and licensing of a SCWR prototype, which is planned to be constructed and operated in the near future. In the preliminary stage of this research plan a fuel assembly, with its experimental loop, will be constructed and tested in SCWR operating conditions. This article reviews the research activities carried out in the Supercritical Water Reactor Fuel Qualification Test (SCWR-FQT) project in Europe and Super Critical Reactor In-Pipe Test Preparation (SCRIPT) project in China. These research activities studied both neutronic and thermal–hydraulic behavior of the test fuel assembly and auxiliary systems of SCWR fuel test facility. CFD simulations, subchannel analysis and system simulations coupled with neutronics code are performed to study the performance of the tested fuel assembly, especially safety related aspects.     

Conceptual study of an accelerator‐driven ceramic fast reactor with long‐term operation

To improve both safe operation and high resource utilization in nuclear power, we propose and investigate the concept of an accelerator‐driven ceramic fast reactor (ADCFR). This reactor type has the potential to operate continuously throughout a 40‐year core life, without fuel shuffling or supplementation. The ADCFR consists of a high‐power superconducting linear accelerator, a gravity‐driven dense granular spallation‐target, and a ceramic fast reactor. The performance of the ADCFR was assessed by using a neutron‐physics simulation, thermal calculations, and a characteristic analysis. The results show that the peak position for the neutron spectrum in the ADCFR is at about 0.1 MeV. This means that it falls with the fast neutron spectrum, and it can convert loaded nuclear fertile material into fissile fuel. Using a burnup simulation, the ideal effective multiplication‐factor (K_eff) was calculated by using a combination of subcritical (accelerator‐driven) and critical modes. In 40 year of operation, K_eff is obtained from the initial 0.98 to the peak ~1.02 and then to ~0.99. Different granular coolant materials were selected to compare neutron performance. In breeding, the differences are relatively small. The thermal calculation indicates that heat transfer performance of granular makes it possible to meet the required specifications in theory. Finally, the corresponding characteristics, with regard to the 2‐phase coolant, ceramic materials, nuclear safety performance, operation modes, economics, and range of applications were analyzed. Accelerator‐driven ceramic fast reactors can achieve very high levels of inherent safety, good breeding performance, high power‐generation efficiency, and high flexibility in wide range of applications.     

Simulating Experimental Investigation on the Safety of Nuclear Heating Reactor in Loss—of —Coolant Accidents

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Using next generation nuclear power reactors for development of a techno‐economic model for hydrogen production

Nuclear and hydrogen are considered to be the most promising alternatives energy sources in terms of meeting future demand and providing a CO?‐free environment, and interest in the development of more cost‐effective hydrogen production plants is increasing—and nuclear‐powered hydrogen generation plants may be a viable alternative. This paper is a report on investigating the application of new generation nuclear power plants to hydrogen production and development of an associated techno‐economic model. In this paper, theoretical and computational assessments of generations II, III+, and IV nuclear power plants for hydrogen generation scenarios have been reported. Technical analyses were conducted on each reactor type—in terms of the design standard, fuel specification, overnight capital cost, and hydrogen generation. In addition, a theoretical model was developed for calculating various hydrogen generation parameters, and it was then extended to include an economic assessment of nuclear power plant‐based hydrogen generation. The Hydrogen Economic Evaluation Program originally developed by the International Atomic Energy Agency was used for calculating various parameters, including hydrogen production and storage costs, as well as equity, operation and maintenance (O&M), and capital costs. The results from each nuclear reactor type were compared against reactor parameters, and the ideal candidate reactor was identified. The simulation results also verified theoretically proven results. The main objective of the research was to conduct a prequalification assessment for a cogeneration plant, by developing a model that could be used for technical and economic analysis of nuclear hydrogen plant options. It was assessed that high‐temperature gas‐cooled reactors (HTGR‐PM and PBR200) represented the most economical and viable plant options for hydrogen production. This research has helped identify the way forward for the development of a commercially viable, nuclear power‐driven, hydrogen generation plant.     

Modeling representative Gen‐IV molten fuel reactivity effects in the ZEPHYR fast/thermal coupled ZPRs. Part I—Assembly level

The comprehension of severe criticality accident is a key issue in Gen‐IV neutronics and safety. Within the future zero‐power experimental physics reactor (ZEPHYR), to be built in Cadarache in the next decade, innovative approaches to reproduce high temperature partially degraded Gen‐IV cores into a critical facility is being investigated. This work presents the first attempt to represent a fuel assembly of sodium‐cooled fast reactor severe criticality accident based on surrogate models. One identified way to construct such representative configuration is to use MASURCA plates stockpile (MOX, UOx, Na, U, and Pu metal) in a fast/thermal coupled core to model a stratified molten assembly. The present study is the first step in a more global approach to full core analysis. The approach is based on a nature‐inspired metaheuristic algorithm, the particle swarm optimization algorithm, to find relevant ZEPHYR configuration at 20°C that exhibits characteristics of (2000‐3000°C) molten MOX assembly in a stratified metal arrangement in a reference sodium‐cooled fast reactor core. Thus, the underlying research question of this study is whether it is possible to represent temperature‐related reactivity effects occurring at fuel meltdown temperatures in a power reactor as density‐related reactivity effects at the operation temperature of a zero‐power reactor, and if so, how should it be done? The calculations are based on a Serpent‐2 Monte Carlo sensitivity and representativity analysis using the Cadarache's cross sections covariance data (COMAC). The single fuel assembly studies show that it is possible to represent the multiplication factor with a representativity factor greater than 0.98. As for reactivity variations, it is possible to achieve a satisfactory representativity factor of above 0.85 in all the presented cases. The representativity process demonstrates that temperature effects could be translated into density effects with good confidence. A complementary analysis on modified nuclear data covariance matrix demonstrates the importance of selecting consistent and robust uncertainties in the particle swarm optimization algorithm. This work provides insights on the behavior of the representativity scheme in different core states and shades some light on the problem in hand.     

Nuclear‐based hydrogen production with a thermochemical copper–chlorine cycle and supercritical water reactor: equipment scale‐up and process simulation

Issues related to equipment scale‐up and process simulation are described for a thermochemical cycle driven by nuclear heat from Canada's proposed Generation IV reactor (Super‐Critical Water‐Cooled Reactor; SCWR), which is a CANDU derivative using supercritical water cooling. The copper–chlorine (Cu‐Cl) cycle has been identified by Atomic Energy of Canada Limited as the most promising cycle for thermochemical hydrogen production with SCWR. Water is decomposed into hydrogen and oxygen through intermediate Cu‐Cl compounds. This article outlines the challenges and design issues of hydrogen production with a Cu‐Cl cycle coupled to Canada's nuclear reactors. The processes are simulated using the Aspen Plus process simulation code, allowing the cycle efficiency and possible efficiency improvements to be examined. The results are useful to assist the development of a lab‐scale cycle demonstration, which is currently being undertaken at the University of Ontario Institute of Technology in collaboration with numerous partners. Copyright © 2010 John Wiley & Sons, Ltd.     

Some developments in direct nuclear energy conversion: Neutron flux dependence of the fission electric cell output parameters

The output parameters of different types of vacuum (VAFEC) and gas filled (GAFEC) fission electric cells (FECs) have been measured during their irradiation in the nuclear reactor of the Central Institute of Physics in Bucharest, Romania, at different thermal neutron flux (?_nth) values in the 10⁸…10¹² neutrons/cm² s range. These measurements allowed us to estimate to what degree the FEC output parameters depend on the ?_nth variation and pointed out the differences in the operation of the VAFEC and GAFEC-type devices. The main results for some of the tested FECs are presented.     

十年风雨 沐浴阳光——国家发改委能源所可再生能源中心任东明主任专访

国家发改委/世界银行/GEF中国可再生能源规模化发展项目(CRESP项目)一期将要结束,该项目的实施对我国可再生能源的发展产生了很大的影响。为此,本刊就此项目实施情况以及将要进行的项目二期等采访了国家发改委能源研究所可再生能源发展中心主任任东明先生。     

Five Solar-Energy Desalination Systems

This paper gives an overview of the Solar Energy Storage Program at the Solar Energy Research Institute. The program provides research, systems analyses, and economic assessments of thermal and thermochemical energy storage and transport. Current activities include experimental research into very high temperature (above 800^°C) thermal energy storage and assessment of novel thermochemical energy storage and transport systems.

The applications for such high-temperature storage are thermochemical processes, solar thermal-electric power generation, regeneration of heat and electricity, industrial process heat, and thermally regenerative electrochemical systems.

The research results for five high-temperatuTe thermal energy storage technologies and two thermochemical systems are described.     

Estimated output parameters at high neutron flux values for an experimental fission electric cell

The output parameters of an experimental fission electric cell (FEC) operating in the current generator mode have been estimated for the values of the neutron flux existing in the nuclear reactors. The estimation has been performed on the basis of the experimental values obtained in an investigation carried out at the Central Institute of Physcis (CIP) in Bucharest, Romania, concentrating on direct nuclear fission energy conversion into electrical energy. These experimental values are higher by orders of magnitude than those reported by other laboratories. The results of the estimation indicate that hundreds of microamperes, hundreds of kilovolts and hundreds of watts could be delivered by an FEC containing 1g of ²³⁵U and irradiated in a thermal neutron flux of 1.4 × 10¹³ neutron cm^?2s^?1. The research demonstrates the feasibility of efficient fission electric cells as potentially usable components in nuclear reactors.     

Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

In this paper, three‐dimensional (3D) power distribution of newly designed small nuclear reactor core has been achieved by using neutron kinetic/thermal hydraulic (NK/TH) coupling. This is pressurized water reactor‐based small nuclear reactor in which plate type fuel element has been used and the core of the reactor has hexagonal type geometry. This paper depicts the design of the reactor core by using coupling approach of neutronics(Neutron Kinetic) and thermal hydraulic studies. For this purpose, neutronic analysis has been obtained by using lattice physics code, i.e. HELIOS and neutron kinetic code, i.e. REMARK. HELIOS code gives the cross‐section data which is being used as input to the REMARK code. At the same time, THEATRe code was used for the thermal hydraulic analysis of the reactor core. In the coupling process, some data (fuel temperature, moderator temperature, void fraction, etc.) from THEATRe code has been used in conjunction with HELIOS and REMARK codes. After finalizing the NK/TH coupling, 3D evaluation of the power distribution of the reactor core has been achieved and is included in the paper. The purpose of this paper is to evaluate the design and get the normal operational behavior of the reactor core by NK/TH coupling approach. Copyright © 2012 John Wiley & Sons, Ltd.     

A preliminary evaluation of irradiation damage in model alloys by electric properties based techniques

The most important effect of the degradation by neutron irradiation is a decrease in the ductility of reactor pressure vessel (RPV) ferritic steels. The main way to determine the mechanical behaviour of the RPV steels is tensile and impact tests, from which the ductile to brittle transition temperature and its increase due to neutron irradiation can be calculated. These tests are destructive and are regularly applied to surveillance specimens to assess the integrity of the RPV. The possibility of applying validated non-destructive aging monitoring techniques would however facilitate the surveillance of the materials that form the reactor vessel.The Institute for Energy of the Joint Research Centre has developed two devices, focussed on the measurement of the electrical properties which prove to give a good non-destructive assessment of the embrittlement state of ferritic steels. The first technique, called Seebeck and Thomson Effects on Aged Material (STEAM), is based on the measurement of the Seebeck coefficient, characteristic of the material and related to the microstructural changes induced by irradiation embrittlement. With the same aim, the second technique, named Resistivity Effects on Aged Material (REAM), measures instead the resistivity of the material.This paper explains (i) preliminary STEAM and REAM results and (ii) results compared with Charpy impact energy temperature shifts due to neutron irradiation. These results will make possible the improvement of such techniques based on the measurement of material electrical properties for their application to non-destructive irradiation embrittlement assessment.     

On the burn-up theory of fast soliton reactors

The basic reactor physics of a completely new nuclear fast fission reactor—the Soliton Reactor—is presented. In such a fast reactor, based either on the U/Pu- or the Th/U-233 fuel cycle, an auto-catalytically driven flux-wave, similar to a water wave on shallow water, propagates through initially fertile regions thus burning it up and producing power.Hence, a qualitatively new kind of harnessing nuclear fission energy may become possible: without transports of irradiated fuel elements, reprocessing and—according to Edward Teller—by co-location of the reactor itself and the subterranean final disposal site in a sand bed about 100m underground. Combined with the idea of an “energy island” soliton reactors could form the basis for a practically inexhaustible source of hydrogen for a climate-neutral energy source.