Steam-Chemical Reactivity Studies for Irradiated Beryllium

This paper reports the results of an experimental study to determine the influence of neutron irradiation effects on the chemical reactivity of beryllium exposed to steam. The study entailed measurements of the following: (1) swelling of irradiated Be specimens annealed at temperatures ranging from 450°C to 1200°C, (2) hydrogen generation rates for unirradiated Be control specimens exposed to steam at temperatures from 450°C to 1200°C, and (3) hydrogen generation rates and tritium mobilization rates for irradiated Be exposed to steam at temperatures from 450°C to 700°C. For irradiated Be, swelling occurred at temperatures above 600°C and it increased to about 56% for an anneal temperature of 1200°C. Tritium and ⁴He were released concurrently from specimens that were annealed at 800°C and above. Steam-Be reactivity measurements for the control specimens were consistent with previous work at temperatures above 700°C, and the new measurements extended the reactivity database down to 450°C. Steam-reactivity measurements for irradiated Be were comparable to control specimens for 600°C and below, but, they indicated a significant enhancement in the chemical reactivity at 700°C.     

Reducing Effective Liquid Wall Thickness in a HYLIFE-II Fusion Breeder

One of the major inertial fusion energy reactor designs is HYLIFE-II which uses protective flowing liquid wall between fusion plasma and solid first wall. The most attractive aspect of this reactor is that protective liquid wall eliminates the frequent replacement of the first wall structure during reactor lifetime. Liquid wall thickness must be at least the thickness required for supplying sufficient tritium for the deuterium–tritium (DT) driver and satisfying radiation damage on the first wall below the limits. Reducing this thickness results less pumping power requirements and cost of electricity. In this study, investigation on potential of utilizing refractory alloys (W-5Re, TZM and Nb-1Zr) as first wall to reduce effective liquid wall thickness in HYLIFE-II reactor using liquid wall of Flibe + 10 mol % UF₄ mixture. Neutron transport calculations were carried out with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈-P₃ approximation. Numerical results showed that using W-5Re or TZM as first wall was effective in decreasing liquid wall thickness in contrast to Nb-1Zr.     

聚变堆安全特性评价研究

确保核安全是未来聚变堆设计、建造和运行过程中必须坚持的最高原则,是聚变堆获得建造和运行许可的前提条件,也是聚变能得以吸引公众的主要理由之一。聚变堆具有高能中子、大量放射性氚、复杂结构、极端服役环境等特点,具有独特的潜在安全问题,因而必须开展针对性研究。本文将从聚变中子与放射性源项、热流体与能量传输、氚安全与环境影响、可靠性与风险管理、安全理念与公众接受度五个方面分别总结其安全特性,系统梳理其关键技术挑战,为建立聚变安全评价体系提供技术支持,进而服务于未来聚变堆的设计与建造。     

Overview of Safety and Environmental Issues for Inertial Fusion Energy

This paper summarizes safety and environmental issues of Inertial Fusion Energy (IFE): inventories, effluents, maintenance, accident safety, waste management, and recycling. The fusion confinement approach among inertial and magnetic options affects how the fusion reaction is maintained and which materials surround the reaction chamber. The target fill technology has a major impact on the target factory tritium inventory. IFE fusion reaction chambers usually employ some means to protect the first structural wall from fusion pulses. This protective fluid or granular bed also moderates and absorbs most neutrons before they reach the first structural wall. Although the protective fluid activates, most candidate fluids have low activation hazard. Hands-on maintenance seems practical for the driver, target factory, and secondary coolant systems; remote maintenance is likely required for the reaction chamber, primary coolant, and vacuum exhaust cleanup systems. The driver and fuel target facility are well separated from the main reaction chamber.     

国内外聚变核安全监管与许可初步分析研究

中国政府高度重视聚变发展,安全是聚变能发展的生命线,而核安全监管和相关许可制度是确保聚变能安全发展的必要手段。聚变堆具有其独特的安全特性,无法完全照搬目前基于裂变堆建立起来的法律法规等监管制度。本文初步梳理了国际(含ITER、国际原子能机构、国际能源署、欧盟、美国、韩国等)关于聚变核安全监管和许可的研究进展和相关经验,总结了我国目前在聚变核安全监管与许可方面的现状与存在的问题,为我国聚变核安全监管提出了发展建议。     

Plasma Technology Critical Issues and Opportunities

Key elements of the U.S. Virtual Laboratory for Technology (VLT) plasma technology portfolio and the advances from the state-of-the-art which would contribute significantly to the world-wide fusion program's objective of developing a practical and attractive fusion product are discussed.     

Fusion Material Studies Relating to Safety in Russia in 2002

The paper is a summary of Russiau material studies performed in frames of activities aiming at substantiation of safety of the International Thermonuclear Experimental Reactor (ITER) after 2001. Subthreshold sputtering of tungsten by 5 eV deuterons was revealed at temperatures above 1150℃. Mechanism of globular films formation was further studied. Computations of tritium permeation into vacuum vessel coolant confirmed the acceptability of vacuum vessel cooling system for removal of the decay heat. The most dangerous accident with high-current arc in toroidal superconducting magnets able to burn out a bore up to 0.6 m in diameter in the cryostat vessel was determined. Radiochemical reprocessing of V-Cr-Ti alloy and its purification from activation products down to a contact dose rate of ∽10 μSv/h was developed.     

Fusion Safety Research and Development in JAERI

This paper describes the current status and future plans of the fusion safety research and development regarding to the developments of the dust removal system and safety analysis code and the thermofluid experiments in the Japan Atomic Energy Research Institute (JAERI) for a fusion experimental reactor. The containment of the radioactive material is the key to achieve fusion safety. In the event of accidents, the source terms need to be evaluated with sufficient accuracy. Therefore, in JAERI, the dust characterization have been investigated and the dust removal system using electric force has been developed and tested. A safety analysis code including both thermal and plasma transient analyses under the various event sequences has been developed. Moreover, the preliminary experiments of thermofluid transients in the vacuum vessel such as Ingress of Coolant Event (ICE) and Loss of Vacuum Event (LOVA) have been started and the experimental results using preliminary LOVA/ICE apparatus during 1995–1996 are summarized in this paper.     

Development of Safety Assessment Method for Plasma Anomaly Events in Fusion Reactors

The objective of the study is to provide a safety assessment method for plasma transients including thermal response of in-vessel components. We developed a plasma physics model for safety analysis which has been implemented in a safety analysis code (SAFALY). The SAFALY code consists of a 0-D plasma dynamics model and a 1-D thermal behavior model of in-vessel components in the thickness direction. The code can treat hydraulic accidents using the results from a hydraulic code and analyze a passive plasma shutdown due to the impurity release from the wall. The overpower events in International Thermonuclear Experimental Reactor (ITER) were investigated, when the fueling rate and confinement improvement changes. The results show no significant damage to the confinement boundary of ITER is expected, as long as the cooling system works normally.     

我国聚变堆结构材料表面阻氚涂层的研究进展

阻氚涂层是聚变堆实现氚自持及氚安全的关键科学与技术问题之一。我国通过国家磁约束聚变能发展研究专项依托国内优势单位部署了阻氚涂层基础问题及工程化技术研发工作。本文介绍了国内外聚变堆结构材料表面阻氚涂层研究进展,重点评述了近几年我国在阻氚涂层的材料选择、制备技术及阻滞氢渗透机制三个科学技术问题的研究进展,提出今后的研究方向。目前我国阻氚涂层材料类型以氧化物涂层为主,涂层制备工艺技术在不断优化和更新。Al2O3/FeAl阻氚涂层的电化学沉积铝(ECA)、粉末包埋渗铝(PC)及热浸铝(HDA)等方法的工艺处理规模及涂层阻氚性能在国际上均相对领先。发展了研究阻氚涂层阻滞氢渗透作用机理的方法,将通常基于Fick定律的表象研究方法向原子级方法前推了一步。未来需在考虑涂层制备工艺与基体材料成分、性能的关系及其在复杂形状结构件的适用性基础上,开发长寿命、高阻氚性能的阻氚涂层材料及制备工艺。     

Investigation on the Neutronic Performance of a Fusion Reactor Using Flibe with Heavy Metal Fluorides

Using liquid wall between the plasma and solid first wall in a fusion reactor allows to use high neutron wall loads and could eliminate frequent replacement of the first wall structure during reactor’s lifetime. Liquid wall should have a certain effective or optimum thickness to extend solid first wall lifetime to reactor’s lifetime and supply sufficient tritium for deuterium–tritium (DT) fusion driver. This study presents the effect of thickness of flowing liquid wall containing 90 mol % Flibe+10 mol % UF₄ or ThF₄ on the neutronic performance of a magnetic fusion reactor design called APEX. Neutron transport calculations were carried out with the aid of code Scale4.3. Numerical results brought out that optimum liquid wall thickness of ∼38 cm was found for the blankets using Flibe+10% UF₄ whereas, 56 cm for that with Flibe+10% ThF₄. Significant amount of high quality fissile fuel was produced by using heavy metal salt.     

Tritium Safety-Related Studies at TPL of JAERI

Activities regarding tritium safety technology in the Tritium Process Laboratory (TPL) at Tokai Establishment of Japan Atomic Energy Research Institute are reviewed. Research and development of a new tritium removal system is being carried out by using a gas separation membrane which enable to make the ITER atmosphere detritiation system more compact and cost-effective. Techniques of gas flowing calorimetry and laser Raman spectroscopy are applied to develop new tritium accountancy methods. Studies of tritium-material interaction, such as plasma material interactions, radiochemical reaction of tritium in gas phase, radiolysis of tritiated water, and waste processing are being carried out under ITER/EDA and U.S.-Japan collaboration. Tritium release experiments for research of tritium behavior in confinements and environment and demonstration of safety related components are planned.     

Overview of the Preliminary Safety Analysis of the National Ignition Facility

The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, New Mexico, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 individual laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a low hazard, radiological facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis report be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A Preliminary Safety Analysis Report (PSAR) has been approved, which documents and evaluates the safety issues associated with the construction, operation, and decommissioning of the NIF.     

Recent progress on the Tokamak Fusion Test Reactor

The deuterium-tritium (D-T) experiments on the Tokamak Fusion Test Reactor (TFTR) have yielded unique information on the confinement, heating and alpha particle physics of reactor scale D-T plasmas as well as the first experience with tritium handling and D-T neutron activation in an experimental environment. The D-T plasmas produced and studied in TFTR have peak fusion power of 10.7 MW with central fusion power densities of 2.8 MWm^–3 which is similar to the 1.7 MWm^–3 fusion power densities projected for 1,500 MW operation of the International Thermonuclear Experimental Reactor (ITER). Detailed alpha particle measurements have confirmed alpha confinement and heating of the D-T plasma by alpha particles as expected. Reversed shear, highl _i and internal barrier advanced tokamak operating modes have been produced in TFTR which have the potential to double the fusion power to 20 MW which would also allow the study of alpha particle effects under conditions very similar to those projected for ITER. TFTR is also investigating two new innovations, alpha channeling and controlled transport barriers, which have the potential to significantly improve the standard advanced tokamak.     

Report of the Second Fusion Energy Sciences Committee of Visitors

This report summarizes the findings and recommendations of the second Committee of Visitors (COV) whose charge was to review the manner in which the U. S. Department of Energy’s Office of Fusion Energy Science (OFES) manages certain programs under its charter. The specific programs reviewed by this COV involve confinement innovation and basic plasma sciences. The charge letter from the Department of Energy is included as Appendix A.     

Report of the FEAC Inertial Fusion Energy Review Panel: July 1996

This report presents the results and recommendations of the U. S. Department of Energy Fusion Energy Advisory Committee (FEAC) review of its Inertial Fusion Energy (IFE) program. The subpanel charged with the review was chaired by John Sheffield of Oak Ridge National Laboratory. The FEAC, to which the subpanel reported, was chaired by Robert Conn of the University of California at San Diego.     

Measurement of the surface morphology of plasma facing components on the EAST tokamak by a laser speckle interferometry approach

The laser speckle interferometry approach provides the possibility of an in situ optical noncontacted measurement for the surface morphology of plasma facing components(PFCs), and the reconstruction image of the PFC surface morphology is computed by a numerical model based on a phase unwrapping algorithm. A remote speckle interferometry measurement at a distance of three meters for real divertor tiles retired from EAST was carried out in the laboratory to simulate a real detection condition on EAST. The preliminary surface morphology of the divertor tiles was well reproduced by the reconstructed geometric image. The feasibility and reliability of this approach for the real-time measurement of PFCs have been demonstrated.     

Fusion Specific Features in ITER Accident Analysis

Fusion specific features like inherent plasma shutdown, low decay heat densities, cryogenic temperatures, and limited source terms were considered during the safety design process of ITER. Uncertainties in plasma disruptions motivates a robust design to cope with multiple failures of in-vessel cooling piping. A vacuum vessel pressure suppression system mitigates pressure transients and effectively captures mobilized radioactivity. In case of pump trips or ex-vessel coolant losses in the divertor the plasma needs to be actively terminated in a few seconds. Failure to do so might damage the divertor but radiological consequences will be minor due to the intact first confinement barrier. Tritium plant inventories are protected by several layers of confinement. Uncontrolled release of magnet energy will be prevented by design. Postulated damage from magnets to confinement barriers causes fluid ingress (air, water, helium) into the cryostat. The cold environment limits pressurization. Most tritium and dust is captured by condensation.     

Safety Analysis of ITER EDA Design by GEMSAFE

General Methodology of Safety Analysis and Evaluation for Fusion Systems (GEMSAFE) was applied to the International Thermonuclear Experimental Reactor (ITER) design in the stage of Engineering Design Activities (EDA) to identify Design Basis Events (DBEs) and the related safety features, which were compared with those of the ITER design in the stage of Conceptual Design Activities (CDA). As a result, 18 DBEs for the EDA design were selected in comparison with 25 DBEs for the CDA design. DBEs related to the fuel area were categorized in higher event category than those of the CDA design due to the increase of the mobile tritium contained in some components. It was necessary to reduce the inventory of the tritium absorbed in the tokamak dust in the EDA design as well as in the CDA design. Some measures were recommended to reduce mobile tritium dissolved in the coolant in the single cooling loop due to the increase of this estimated inventory.     

On the Tritium Breeding Capability of Flibe, Flinabe, and Li20Sn80 in a Fusion-Fission (Hybrid) Reactor

In a commercial (DT) driven fusion reactor, the tritium breeding ratio per incident fusion neutron must be greater than 1.05 to maintain tritium self-sufficiency for the driver. In this study tritium breeding capability of three different coolants, namely Flibe (LiF·BeF₂), Flinabe (LiF·NaF·BeF₂), and Li₂₀Sn₈₀ in a (DT) driven fusion-fission (hybrid) reactor was investigated for different refractory alloys (W-5Re, TZM, T111, and Nb-1Zr) as structural material. Neutron transport calculations were conducted with the help of SCALE 4.3 SYSTEM by solving the Boltzmann transport equation with code XSDRNPM. The contribution of Flibe, Flinabe, and Li₂₀Sn₈₀ with respect to ⁶Li enrichment in their lithium content to overall TBR was investigated. In addition, the effect of structural material type on TBR was examined.