Radiation facilities for fusion reactor first wall and blanket structural materials development

Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed.     

Radiation damage in silicon carbide and graphite for fusion reactor first wall applications

Silicon carbide and graphite materials were exposed to fast neutron fluences of 2 × 10²³ to 2 × 10²⁴n/m² (E > 1 MeV) and a study was made of changes in fracture strength, Weibull modulus and electrical resistivity. Silicon carbide (Norton NC-430) exhibits a decrease in fracture strength (25%) at the higher fluence if the temperature is kept at 298 K, while at 1473 K the decrease in fracture strength is only 10% indicative of recovery due to thermal annealing. The fracture strength of the graphite (POCO AXF-5Q) tested at 298 K increases rapidly by ~20% after 2 × 10²³n/m² and remains constant at higher fluence. Analyses of the data using the Weibull weakest link model were given, in addition to annealing and swelling results.     

Safety study of primary coolants in fusion power reactor first wall design

This paper examines potential safety problems associated with the various primary coolant candidates currently considered for the EPR fusion blanket designs. The basic concern is the possibility of overheating and melting of the first wall and the blanket, induced by a malfunction in the primary coolant system. These accidents include the loss-of-coolant flow, the loss-of-heat removal, overpower transients, and the loss of coolant. Following a mechanistic safety for these four types of accident sequences and comparing helium and liquid metal cooling, it was found that helium has a more adverse effect on the first-wall heat up in the event of a loss-of-heat removal or a loss-of-coolant because its lack of thermal inertia.     

Thermohydraulic analysis of high-Prandtl-number fluid in complex duct simulating first wall in fusion reactor

For fusion reactors, molten salt is one of the candidates for coolant materials. Molten salt is a high-Prandtl-number fluid; thus, it is necessary to enhance the heat transfer coefficient. It is proposed that rods are inserted into a duct to enhance the heat transfer coefficient. The flow field behind the rod in the duct is visualized to compare experimental data with simulation results. The trends and distributions in the numerical simulation are the same as those in the experiment, and furthermore, the magnitudes of the time and space scales in the numerical simulation are of the same order as those in the experiment. Thermohydraulic numerical analysis confirmed that the heat transfer coefficient is improved by inserting the rod when the fluid is a high-Prandtl-number fluid and the flow field is in the turbulent region. However, it is necessary for the rods to be arranged in the streamwise direction.     

On the radiation damage characterization of candidate first wall materials in a fusion reactor using various molten salts

Evaluating radiation damage characteristics of structural materials considered to be used in fusion reactors is very crucial. In fusion reactors, the highest material damage occurs in the first wall because it will be exposed to the highest neutron, gamma ray and charged particle currents produced in the fusion chamber. This damage reduces the lifetime of the first wall material and leads to frequent replacement of this material during the reactor operation period. In order to decrease operational cost of a fusion reactor, lifetime of the first wall material should be extended to reactor’s lifetime. Using a protective flowing liquid wall between the plasma and first wall can decrease the radiation damage on first wall and extend its lifetime to the reactor’s lifetime. In this study, radiation damage characterization of various low activation materials used as first wall material in a magnetic fusion reactor blanket using a liquid wall was made. Various coolants (Flibe, Flibe + 4% mol ThF₄, Flibe + 8% mol ThF₄, Li₂₀Sn₈₀) were used to investigate their effect on the radiation damage of first wall materials. Calculations were carried out by using the code Scale4.3 to solve Boltzmann neutron transport equation. Numerical results brought out that the ferritic steel with Flibe based coolants showed the best performance with respect to radiation damage.     

Neutronic study on the effect of first wall material thickness on tritium production and material damage in a fusion reactor

In this study, the effects of changing first wall materials and their thicknesses on a reactor were investigated to determine the displacement per atom（DPA） and gas production（helium and hydrogen） in the first wall, as well as the tritium breeding ratio（TBR） in the coolant and tritium breeding zones. Therefore, the modeling of the magnetic fusion reactor was determined based on the blanket parameters of the International Thermonuclear Experimental Reactor（ITER）. Stainless steel（SS 316 LNIG）, Oxi...     

Modeling hydrogen and helium entrapment in flowing liquid metal surfaces as plasma-facing components in fusion devices

In a fusion reactor, the ability to use liquids as plasma-facing components (PFCs) depends on their interaction with the plasma and the magnetic field. One important issue for the moving liquid is the ability to entrain particles that strike the PFC surface (helium and hydrogen isotopes) while accommodating high heat loads. To study this problem, an analytical model and a two-dimensional comprehensive numerical model have been developed and implemented in the HEIGHTS computer simulation package. The models take into account the kinetics of particle injection, motion and interactions with the liquid lattice, and the ultimate release from the surface. The models were used to investigate an important issue, whether He particles can be pumped by the PFC liquid rather than requiring a standard vacuum system. Hydrogen isotope (DT) particles that strike the surface will likely be trapped in the liquid-metal surface (e.g., lithium) due to the high chemical solubility of hydrogen. The impinging He particles in the established low-recycling regime at PFCs could be harder to pump using the standard vacuum pumping techniques. The analysis results indicate a reasonable chance of adequate helium self-trapping in flowing lithium as PFC without active pumping.     

Deuterium and helium retentions of V–4Cr–4Ti alloy used as first wall of breeding blanket in a fusion reactor

The deuterium and helium retention properties of V–4Cr–4Ti alloy were investigated by thermal desorption spectroscopy (TDS). Ion energies of deuterium and helium were taken at 1.7 and 5 keV, respectively. The retained amount of deuterium in the sample irradiated at 380 K increased with the ion fluence and was not saturated to fluence of up to 1 × 10²³ D/m². For the irradiation at 773 K, 0.1% of implanted deuterium was retained at the highest fluence. For the helium ion irradiation at room temperature, three groups of desorption peaks appeared at around 500, 850, and 1200 K in the TDS spectrum. In the lower fluence region (<1 × 10²¹ He/m²), the retained helium desorbed mainly at around 1200 K. With increasing fluence, the amount desorbed at 500 K increased. Total amount of retained helium in the samples saturated at fluence up to 5 × 10²¹ He/m² and saturation level was 2.7 × 10²¹ He/m².     

Proof pressure tests of the first reactor vessels at Hartlepool and Heysham nuclear power stations

The nuclear stations currently nearing completion at Hartlepool and Heysham are the world's first design of Advanced Gas Cooled Reactor to use podded boiler construction. The proof pressure tests on the first reactor pressure vessels at both stations were carried out during the early part of 1980. The object of the proof pressure test is to satisfy statutory and contractual requirements by demonstrating the integrity of the completed concrete pressure vessel, liners and pressurised penetrations when subjected to a pneumatic test pressure of 740 lbf/in² g, equivalent to 1.15 times the design pressure of 644 lbf/in² g. Testing was carried out in the latter period of the construction phase of each station with the majority of the plant and buildings in an advanced state of construction and the reactor internals essentially complete. This paper examines the results of both reactor 1 vessel tests and compares the behaviour of the vessels with one another and with analytical predictions. Possible sources of instrumentation error are identified and discussed. The conclusions are drawn from correlation of the test results with theoretical predictions.The vessel pressure was raised and lowered in six stages to and from the test pressure with a hold period between each stage to read instruments, examine the vessel surfaces and leak check the penetrations and closures. To obtain the relevant information at each pressure hold, the vessels' permanent instrumentation, consisting of vibrating wire strain gauges and their associated electrical resistance thermometers, and the liner and reactor internal thermocouples were monitored. In addition, strain gauges fixed to the concrete face of the liner during construction were monitored and liner strains measured during the test were compared with predictions. Vessel deflections were monitored by special equipment mounted on the reactor building to enable vessel dilation to be measured. The vessel top and bottom caps were instrumented by manometric systems in order that the cap profiles and deflections could also be determined.The measurements and observations made during the proof pressure tests adequately demonstrated that the vessels behaved in accordance with predictions. The concrete strains and vessel external deflections confirmed that the vessels behaved in a linear and elastic manner throughout the tests, with no cracking being observed in the concrete during the tests. Liner strains compared favourably with predicted values, exhibiting a linear behaviour under increasing pressure. The strain levels recorded gave complete confidence in the liner design.The tests confirmed the integrity of the vessels, thus enabling them to go forward to the engineering run stage in the commissioning programme leading ultimately to the raising of power.     

Temporal and spatial dynamics of optical emission from laser ablation of the first wall materials of fusion device

Laser-induced breakdown spectroscopy(LIBS) has been developed to in situ diagnose the chemical compositions of the first wall in the EAST tokamak. However, the dynamics of optical emission of the key plasma-facing materials, such as tungsten, molybdenum and graphite have not been investigated in a laser produced plasma(LPP) under vacuum. In this work, the temporal and spatial dynamics of optical emission were investigated using the spectrometer with ICCD.Plasma was produced by an Nd:YAG laser(1064 nm) with pulse duration of 6 ns. The results showed that the typical lifetime of LPP is less than 1.4 μs, and the lifetime of ions is shorter than atoms at ~10~(-6)mbar. Temporal features of optical emission showed that the optimized delay times for collecting spectra are from 100 to 400 ns which depended on the corresponding species. For spatial distribution, the maximum LIBS spectral intensity in plasma plume is obtained in the region from 1.5 to 3.0 mm above the sample surface. Moreover, the plasma expansion velocity involving the different species in a multicomponent system was measured for obtaining the proper timing(gate delay time and gate width) of the maximum emission intensity and for understanding the plasma expansion mechanism. The order of expansion velocities for various species is V_C~+ V_H V_(Si)~+ V_(Li) V_(Mo) V_W.These results could be attributed to the plasma sheath acceleration and mass effect. In addition, an optimum signal-to-background ratio was investigated by varying both delay time and detecting position.     

Gas-transfer kinetics and regulation of the gas content of coolant in the first loop of a water-cooled-water-moderated reactor

OKBM. Translated from Atomnaya Énergiya, Vol. 72, No. 6, pp. 554-559, June, 1992.     

Gas entrainment allowance level at free surface and gas dynamic behavior of sodium-cooled fast reactor

In a sodium-cooled fast reactor (SFR), inert gases exist in the primary coolant system either in a state of dissolved gas or free gas bubbles. The sources of the gas bubbles are entrainment and dissolution of the reactor cover gas (argon) at the vessel free surface and emission of the helium gas that is produced as a result of disintegration of B₄C control rod material. The gas in the primary system may cause disturbance in reactivity, nucleation site for boiling, etc. Therefore, it is a key issue from the design and safety viewpoint and the allowance level is necessary regarding the gas entrainment at the free surface and the gas bubble concentration in the primary system. In the present study, a gas entrainment allowance level at the free surface is discussed and rationalized for the Japanese SFR (JSFR) design. The influence of the gas entrainment is evaluated using the void fraction at the core inlet. Design criteria for the acceptable level of the gas entrainment and gas concentration are proposed in consideration of the background level of gasses in the coolant. For the purpose, a plant dynamics code VIBUL has been developed to apply to the JSFR design to evaluate the concentration distribution of the dissolved gas and the free gas bubble in the JSFR system. Using the plant dynamics code for the bubble behavior, the background level of the free gas (void fraction at the core inlet) has been obtained. Assuming that the total void fraction should be kept below 105% of the background level, a preliminary design allowance level of gas entrainment is proposed as the map in terms of the entrainment rate and the entrained bubble radius. Furthermore, the possibility of bubble removal and design requirement of the device is investigated to satisfy the allowance level. It is noted that the background level is already very low in comparison with the induced void reactivity by the void passing the reactor core.     

An overview of the development of the first wall and other principal components of a laser fusion power plant

This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0-100 keV) and ions (0.1-4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations.     

Static and buckling analysis of a cryostat outer wall (Bell-Jar) for the whole magnetic system of the intor-net fusion reactor

In the framework of the studies performed for the Experimental Fusion Reactor INTOR/NET, the proposal was considered of enclosing the Torus and the whole magnetic system in a single large metallic containment under vacuum, constituting the outer boundary of the cryostat. Such a big containment (called the Bell-Jar) poses severe engineering problems of feasibility and stability, which are considered in the present acticle. The conceptual design was analysed and developed in detail with extensive use of the Finite Element Method. The studies enable us to identify a configuration which is able to sustain the applied loads in stable equilibrium conditions within the allowable stress limits.     

聚变驱动次临界堆双冷嬗变包层第一壁结构热应力计算与分析

使用有限元程序对聚变次临界堆双冷嬗变包层第一壁进行数值模拟 ,给出不同载荷条件下的温度场和应力场分布 ,结果证明典型氦气系统设计满足热工要求。依据数值模拟结果对第一壁氦气载热能力进行分析 ,并考虑了流道形状对结构热应力的影响。     

Modeling of a DC glow discharge in a neon–xenon gas mixture at low pressure and with metastable atom densities

The physical properties of Ne–Xe DC glow discharges at low pressure are reported for a gap length of 1 cm for the first time in the literature. The model deals specifically with the first three moments of Boltzmann's equation and includes the radiation processes and metastable atom densities. The spatio-temporal distributions of the electron and neon and xenon ion densities, the neon and xenon metastable atom densities, the electric potential and the electric field as well as the mean electron energy are presented at 1.5 Torr and 250 V. The current–voltage characteristic is shown at 3 Torr, and it is compared with previous work for pure neon gas. The model is validated theoretically and experimentally in the case of pure gas.