Evaluation of irradiation facility options for fusion materials research and development

Successful development of fusion energy will require the design of high-performance structural materials that exhibit dimensional stability and good resistance to fusion neutron degradation of mechanical and physical properties. The high levels of gaseous (H, He) transmutation products associated with deuterium–tritium (D–T) fusion neutron transmutation reactions, along with displacement damage dose requirements up to 50–200 displacements per atom (dpa) for a fusion demonstration reactor (DEMO), pose an extraordinary challenge. One or more intense neutron source(s) are needed to address two complementary missions: (1) scientific investigations of radiation degradation phenomena and microstructural evolution under fusion-relevant irradiation conditions (to provide the foundation for designing improved radiation resistant materials), and (2) engineering database development for design and licensing of next-step fusion energy machines such as a fusion DEMO.A wide variety of irradiation facilities have been proposed to investigate materials science phenomena and to test and qualify materials for a DEMO reactor. Some of the key technical considerations for selecting the most appropriate fusion materials irradiation source are summarized. Currently available and proposed facilities include fission reactors (including isotopic and spectral tailoring techniques to modify the rate of H and He production per dpa), dual- and triple-ion accelerator irradiation facilities that enable greatly accelerated irradiation studies with fusion-relevant H and He production rates per dpa within microscopic volumes, D–Li stripping reaction and spallation neutron sources, and plasma-based sources.The advantages and limitations of the main proposed fusion materials irradiation facility options are reviewed. Evaluation parameters include irradiation volume, potential for performing accelerated irradiation studies, capital and operating costs, similarity of neutron irradiation spectrum to fusion reactor conditions, temperature and irradiation flux stability/control, ability to perform multiple-effect tests (e.g., irradiation in the presence of a flowing coolant, or in the presence of complex applied stress fields), and technical maturity/risk of the concept. Ultimately, it is anticipated that heavy utilization of ion beam and fission neutron irradiation facilities along with sophisticated materials models, in addition to a dedicated fusion-relevant neutron irradiation facility, will be necessary to provide a comprehensive and cost-effective understanding of anticipated materials evolution in a fusion DEMO and to therefore provide a timely and robust materials database.     

Re-evaluation of neutron displacement cross sections for silicon carbide by a Monte Carlo approach

Neutron displacement cross sections for SiC are re-evaluated by a Monte Carlo approach, with damage energies of primary recoils calculated by the stopping and range of ions in matter (SRIM) code. The validity of the Monte Carlo model is examined by the case of iron, and the results show good agreement with the reference values. Neutron displacement cross sections for SiC at energies up to 100 MeV are calculated, and averaged over the neutron spectra of a fusion DEMO reactor, the high flux test module of the International Fusion Materials Irradiation Facility, and typical fission test reactors. Gas production is also calculated for those neutron irradiation facilities. Finally, the suitability of the displacement cross sections is discussed. The results on comparison among neutron irradiation of different facilities by the current displacement cross sections are similar to those by results of the previous work. Moreover, since neutron displacement cross sections in this study are calculated with damage energies of primary recoils calculated by SRIM, neutron damage evaluated by our displacement cross sections is suitable for correlation with damage by heavy ions calculated by SRIM.     

Recent applications of small-angle neutron scattering in the characterization of irradiated steels for nuclear technologies

Small-angle neutron scattering (SANS) is a powerful experimental tool to investigate the microstructural evolution under irradiation in steels for fission and future fusion reactor systems. We present recent SANS results concerning the modelling of helium bubble growth in F82H-mod. steel implanted with α-particles and the dose dependence of microstructural radiation damage in Eurofer-97 steel for fusion reactors irradiated at 250 °C. The discussion of these results is focussed on the quality of the metallurgical information obtained by such SANS measurements and consequently on their usefulness also for engineering and design purposes.     

Damage parameters of structural materials in fusion environment compared to fission reactor irradiation

Calculations were performed to quantify the damage parameters in the leading candidate structural and plasma facing materials when used in magnetic and inertial confinement fusion systems and when irradiated in fission reactors. The structural materials considered are ferritic steel, austenitic steel, vanadium alloy and SiC/SiC composite. Plasma facing materials included beryllium, tungsten, and carbon fiber composites. Atomic displacement damage and gas production rates are greatly influenced by the neutron energy spectrum. For the same neutron wall loading, atomic displacement damage is slightly lower in inertial fusion systems than in magnetic fusion systems but gas production is about a factor of 2 lower. In addition, much lower gas production is obtained in samples irradiated in fission reactors. The results help guide irradiation experiments in fission reactors to properly simulate the damage environment in fusion systems and facilitate extrapolating to the expected material performance in fusion systems.     

Basic design guideline for the preliminary engineering design of PIE facilities in IFMIF/EVEDA

The international fusion materials irradiation facility (IFMIF) is an accelerator-based intense 14 MeV neutron source for testing fusion reactor materials. Under broader approach (BA) agreement between EURATOM and Japan, the engineering validation and engineering design activity (EVEDA) were started from 2007. The IFMIF needs the post irradiation examination (PIE) facilities to generate a materials irradiation database for the design and licensing of fusion DEMO reactors. In this study we examined and discussed about the safety such as remote handling, hot cell design, and the equipments and apparatus of hot cells, and we summarized a basic design guideline for the preliminary engineering design of the PIE facilities.     

Neutron Excess Generation by Fusion Neutron Source for Self-Consistency of Nuclear Energy System

The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities.     

The need for improved temperature control during reactor irradiation

The irradiation conditions of materials irradiated in fission reactors are nearly always insufficiently described in the literature. A carful inspection of temperature-reactor power histories reveals a deficiency in the conventional control of irradiation temperatures. In particular, a short transient irradiation at a lower temperature commonly occurs during the startup of the reactor. A large difference in the final defect structure is expected to be caused by this transient irradiation from the mechanism of the defect structure development. An electron irradiation with a similar transient irradiation leads to a remarkably different defect microstructure. The defect structures introduced by a fission neutron irradiation with this transient was compared with those produced by a fusion neutron irradiation with perfect temperatrue control without transient. The difference in structures was found to be much greater than what is normally expected between fission and fusion neutron irradiations. An essential improvement in the control of reactor irradiations is proposed.     

Computational tracking of BN-600 operation

Computational tracking of BN-600 operation is described. The high quality of computational tracking is largely due to the nature of a fast reactor, in this case BN-600. Unlike reactors with a thermal neutron spectrum, in a fast reactor, because the prompt and delayed fission neutrons as well as the absorbed neutrons are almost in the same energy range as the fast neutrons, a computational cell can be confidently homogenized and the reactor is strongly coupled to the neutron field. These are the reasons why the behavior of the reactor can be successfully predicted by means of computational programs which are based on the diffusion approximation neglecting the anisotropy of the interaction of the neutrons and the heterogeneity of the medium.     

Characterization of displacement cascade damage produced in Cu3Au by fast-particle irradiation

Zones of reduced long-range order created at displacement cascade sites in well-ordered Cu₃Au may be directly imaged in the transmission electron microscope so that quantitative information can be obtained on individual cascade events. This technique has been used to characterise the cascade damage created by three fast particles (3.5 MeV protons, a source of moderated fission neutrons and a source of fusion neutrons with energies peaking at 14.8 MeV) with the aim of comparing the experimental observations with the relevant collision models. In each case, disordered zone number densities, sizes and shapes were determined, and were found to be characteristic of each irradiation, with the sizes of disordered zones and the proportion of zones of complex shape increasing on going from 3.5 MeV protons to fission neutrons to fusion neutrons. The quantitative results are largely consistent with the different calculated primary recoil spectra, although in the fusion neutron case some discrepancies are found which cannot readily be explained by limitations in the experimental technique. More specifically, more and larger disordered zones are found than expected from the calculated recoil spectrum. Subcascade formation was observed only in the neutron irradiations, with the distributions of sizes and shapes of individual sub-cascades being very similar in the two cases (in marked contrast to those obtained from sizing total cascade events). Finally, the production of point-defect clusters at cascade sites was studied. The efficiency of cascade collapse increased on going from 3.5 MeV protons to fission neutrons to fusion neutrons.     

Design of coated fuel particles for a hybrid fusion-fission system

Designs have been developed for coated ThO₂ fuel particles to be used in a hybrid fusion-fission system that could be operated without reprocessing. The fresh fertile fuel particle would first be cycled through the blanket of a fusion reactor to breed ²³³U, which would then be ‘burned’ in a thermal fission reactor. The depleted fuel would then be refreshed in a second pass through the fusion reactor, and the process above repeated as many times as feasible. Designs of coated particles for up to three cycles through the hybrid system of reactors have been developed. The outer structural layer for these particles is made from vapor-deposited silicon carbide, because of its remarkable dimensional stability under fast neutron irradiation, and an inner layer of porous pyrocarbon is used to accommodate the buildup of gaseous reaction products inside the particle. The production of gaseous emission products from the interaction of high-energy fusion neutrons with coating materials and with the oxygen in the kernel contributes significantly to pressure vessel stresses in these coatings, whereas gaseous fission products alone dominate in conventional thermal reactors. The most stringent design for the three-cycle particle is identical in fuel loading to the reference fertile particle for an HTGR, which would constitute an ideal hybrid partner for the fusion reactor. Consideration is also given to coated-particle designs for the containment of the bred tritium used to fuel the D-T fusion reactor.     

核反应堆内中子能谱测量技术

随着我国核能产业的迅速发展,各种类型的核反应堆设施相继研发和投入使用,掌握堆内的中子能谱信息对其性能诊断和安全运行具有重要的意义。本文针对裂变和聚变两种类型反应堆的特点,详细阐述了适用的中子能谱测量方法以及研发的中子谱仪,为未来相关研究工作提供参考。     

Neutronics study for the installation of the LBVM in the medium flux area of IFMIF

IFMIF (International Fusion Materials Irradiation Facility) will be a fusion dedicated facility producing a large amount of neutrons with the appropriate energy spectrum to test materials and subcomponents for DEMO and future Fusion Power Plants.While the high flux area of IFMIF will be devoted to reduced activation structural materials for first wall and blanket, the medium flux area will be dedicated to functional materials for breeder blankets. In particular, the Liquid Breeder Validation Module (LBVM), will host experiments related with functional materials for liquid breeder blankets. Since IFMIF neutron spectra have been intended to fit the most irradiated areas of a fusion reactor in the high flux area, the irradiation conditions in the LBVM placed in the medium flux area of IFMIF have been assessed. The effect of some neutron shifter/reflector components to optimize the neutron spectra have been evaluated in order to find out the proper irradiation conditions for functional materials for liquid breeder blankets.Therefore, the objective of this report is to summarize the neutronic calculations developed to evaluate the viability of IFMIF neutron source to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for future nuclear fusion power reactors (ITER, DEMO). The irradiation parameters evaluated for this purpose are: the tritium production for liquid breeder material (Pb–17Li) and the damage dose (dpa) and gas production to damage dose ratios for Al₂O₃ and SiC functional materials.The main conclusion is that, it is possible to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for the future nuclear fusion reactor DEMO. Nevertless, the use of some shifter components will be needed to optimize some irradiation parameters.     

Proton simulation of displacement effects induced in metals by 14 mev neutrons

In designing a D-T fusion reactor, one must know the effect of a high flux of 14 MeV neutrons on structural materials. Available laboratory sources of 14 MeV neutrons are not intense enough to expose samples to the expected flux. Bombardment with other particles is one way of simulating the anticipated neutron environment. The energy spectrum of atoms recoiling from collisions with bombarding particles can be calculated from elastic-scattering and nonelastic-reaction data for the incident species. This analysis shows that 16 MeV protons closely simulate the displacement effects caused by 14 MeV neutrons. In niobium the average atom recoiling from a 14 MeV neutron interaction has 65 keV of damage energy. The mean damage energy deposited per cm³ of niobium by a fluence of one 14 MeV neutron per cm² is 14 keV. The equivalent quantity for 16 MeV protons incident on niobium is 33 keV.     

55Fe effect on enhancing ferritic steel He/dpa ratio in fission reactor irradiations to simulate fusion conditions

This study evaluated methods for increasing the helium production rate in ferritic steel irradiation in a fission reactor neutron spectrum in order to increase the helium to atomic displacement ratio to values typical of fusion reactor first wall conditions. An early experiment showed that the accelerated He(appm)/dpa ratio of about 2.3 was achieved for 96% enriched ⁵⁴Fe in iron in the High Flux Isotope Reactor (HFIR), ORNL. In the current work, the ferritic steel He(appm)/dpa ratio was studied in the neutron spectrum of HFIR with the ⁵⁵Fe thermal neutron helium production taken into account. A benchmark calculation for the same sample, as used in the aforementioned experiment, was then used to adjust and evaluate the ⁵⁵Fe (n, a) cross section values in TALYS-based Evaluated Nuclear Data Library (TENDL). The analysis showed that a decrease of a factor of 6700 for the TENDL ⁵⁵Fe (n, a) cross section in the intermediate and low energy regions was required in order to fit the experimental results. The best fit to the cross section value at thermal neutron energy was about 27 mb. With the adjusted ⁵⁵Fe (n, a) cross sections, calculation showed that the ⁵⁴Fe and ⁵⁵Fe isotopes could be enriched by the isotopic tailoring technique in a ferritic steel sample irradiated in HFIR to significantly enhance the helium production rate. This new calculation can be used to guide future isotopic tailoring experiments designed to increase the He(appm)/dpa ratio in fission reactors. A benchmark experiment is suggested to be performed to evaluate the ⁵⁵Fe (n, a) cross section at thermal energy.     

Radiation Damage Study on Various Structural Refractory Alloys of a Multi-Purpose Reactor

Radiation damage properties of structural materials play a key role in design of a fusion–fission (hybrid) reactor. Refractory alloys offer a significant advantage of high neutron wall load capability under fusion neutron environment. In this study, main radiation damage parameters (displacement per atom (DPA) and helium production) on three different refractory alloys, namely W-5Re, TZM (Mo alloy) and Nb–1Zr used as structural material in a hybrid reactor were found. Neutron transport calculations were conducted with the aid of SCALE4.3 System by solving the Boltzmann transport equation with code XSDRNPM. The lowest radiation damage values were obtained for W-5Re alloy. Moreover, all investigated materials will require to be replaced frequently due to their radiation damage values during reactor life (~ 30 years).     

Mixed Ne/Ni beams for radiation damage studies on the harwell variable-energy cyclotron

Ion irradiation facilities which use metal ions beams concurrently with beams of He or He+H have been developed extensively in recent years for the simulation of radiation effects which occur in metals during neutron irradiation in fission or fusion reactors. Helium ions are considered necessary to simulate the effect of helium on cavity nucleation and the metal atom is used to create an adequate atomic displacement rate. Many of these dual-beam facilities require the provision of two accelerators, one to produce a gas atom and the other a metal atom beam, and commonly employ low-energy metal ions (eg. 4 MeV ⁵⁸Ni⁺) which have a limited range (0.75 μm) in metals of practical interest. Damage can be created at greater depths by use of high energy ions (eg 45 MeV ⁵⁸Ni) in a cyclotron, but charge-to-mass constraints make it difficult to produce a mixed beam of concurrent He and metal ions. It is noted that the heavier inert gas atom neon has similarities in atomistic behaviour in metals to those of helium and could possibly be used as an analogue for helium. In this report we consider this aspect and show than by suitable matching of $\text{e}\text{m}$ ratio a mixed beam of 15 MeV ²⁰Ne²⁺ and 45 MeV ⁶⁰Ni⁶⁺ can be obtained from the Harwell VEC with gas/metal atom ratios and beam currents that are suitable for use in radiation damage simulation experiments in metals. It is proposed that exploratory studies should be carried out with mixed Ne/Ni beams into nickel to investigate the possibility of using neon as an analogue for helium in radiation damage simulation studies.     

聚变-裂变混合堆水冷包层中子物理性能研究

研究直接应用国际热核聚变实验堆（ITER）规模的聚变堆作为中子驱动源,采用天然铀为初装核燃料,并采用现有压水堆核电厂成熟的轻水慢化和冷却技术,设计聚变-裂变混合堆裂变及产氚包层的技术可行性。应用MCNP与Origen2相耦合的程序进行计算分析,研究不同核燃料对包层有效增殖系数、氚增殖比、能量放大系数和外中子源效率等中子物理性能的影响。计算分析结果显示,现有核电厂广泛使用的UO₂核燃料以及下一代裂变堆推荐采用的UC、UN和U₉₀Zr₁₀等高性能陶瓷及合金核燃料作为水冷包层的核燃料,都能满足以产能发电为设计目标的新型聚变 裂变混合堆能量放大倍数的设计要求,但只有UC和U₉₀Zr₁₀燃料同时满足聚变燃料氚的生产与消耗自持的要求。研究结果对进一步研发满足未来核能可持续发展的新型聚变-裂变混合堆技术具有潜在参考价值。     

基于蒙特卡罗方法的固态氚增殖剂聚变中子辐照损伤行为分析

实验包层模块(TBM)是聚变反应堆最重要的组件之一,作用是产氚和能量提取。锂陶瓷具有良好的化学稳定性、热机械性能、产氚性能以及可在更高温度下使用等特点,被认为是聚变堆包层最具吸引力的氚增殖剂材料。中国ITER-TBM设计方案采用了氦冷固态氚增殖剂(HCCB)TBM结构,其聚变环境下的辐照损伤行为可为中国HCCB TBM结构设计提供支持。针对固态氚增殖剂聚变中子辐照损伤问题,利用蒙特卡罗模拟,对比分析了Li_4SiO_4和Li_2TiO_3的中子辐照离位损伤和嬗变气体损伤。结果表明:在相同的服役时间下,Li_4SiO_4比Li_2TiO_3将产生更多的嬗变气体,且在高6 Li丰度情况下,其中子辐照损伤也更严重,会产生更高的损伤剂量和更大的损伤截面。但是,嬗变气体所造成的空位损伤Li_2TiO_3要比Li_4SiO_4严重;对两种陶瓷材料来讲,氦损伤效应均强于氚损伤效应。     

Assessment of Radiation Damage to the Structural Material of EAST Tokamak

Radiation damage to structural material of fusion facilities is of high concern for safety. The superconducting tokamak EAST will conduct D-D plasma experiments with the neutron production of 1015 neutrons per second. To evaluate the material radiation damage a programme system has been devised with the Monte Carlo transport code MCNP-4C, the inventory code FISPACT99, a specific interface, and the fusion evaluated nuclear data library FENDL-2. The key nuclear responses, i.e. fast neutron flux, displacement per atom, and the helium and hydrogen production, are calculated for the structural material SS-316L of the first wall, and the vacuum vessel, using this programme. The results demonstrate that the radiation damage to the structural material is so little that it will not lead to any significant change of material properties according to the reference design. This indicates that there is a large potential space for EAST to test advanced operation regime from the viewpoint of structural material safety.