Tritium Breeding Blanket Research Articles

Neutronics analysis and assessment of shielding options of pipe forest and Bioshield-Plug design for ITER TBSs

In ITER [1], it is foreseen to have two Test Blanket Systems (TBS) installed in two dedicated Equatorial Ports (EP) for the purpose of testing and validating different concepts of tritium breeding blankets [2]. Each TBS consists of a Test Blanket Module Port Plug (TBM PP) and its associated systems; the Pipe Forest (PF) in the Port Interspace (PI) area and the corresponding systems in the Port Cell (PC) area. While analyses have been conducted before [3], the design has since gone through major evolution with a redesign of the PF, a new BP based on the butterfly doors concept, and an optional activated water delay tank. In this paper, the results of a comprehensive nuclear analyses that aim at assessing the neutronics performance of the revised PF and BP, will be discussed. These include calculations of neutron flux distributions and Shutdown Dose Rate (SDDR), following SA2 [4] scenario, in a baseline configuration and exploring the option of adding additional shielding plates in the PF. It has been found that the additional shielding plates reduces the neutron flux in the PI by ∼40–60% and the SDDR in the PI, due to activated components, by ∼50%.

Vapor adsorption testing of ambient molecular sieve bed in coolant purification system

A helium cooling system (HCS) circulates helium coolant to remove heat from a tritium breeding blanket in a fusion reactor. A small amount of tritium can permeate into the HCS from the blanket, and it should be removed as per safety requirements in the system. A coolant purification system (CPS) is connected to the HCS, and one percent of the helium flow is bypassed into the CPS from the HCS for purification of the coolant. An ambient molecular sieve bed (AMSB) will play a role in capturing the tritiated water, Q2O. In order to develop and confirm the AMSB design and its function, a test facility is constructed especially for the experimental validation of the adsorption characteristics under a very low concentration of vapor. An experiment will be conducted under a helium flow. However, before the main helium test, air tests are performed at room temperature and 0.33 MPa. The facility can be divided into two parts, one part containing a device generating a small amount of moisture and the other containing the test section including the AMSB. The vapor generator can continuously supply vapor of 1 to hundreds of ppm to the dry gas. Various AMSBs can be installed in the test section. The adsorption performance and saturation characteristics according to the diameter, length, and flow rate are compared. The experiment was conducted on two columns of different lengths. Specifically, when the performance validation test with the 64 mm AMSB was conducted over a period of more than 10 days, it showed very stable adsorption performance. The length of the AMSB estimated in the experiment was several times shorter than that obtained through a correlation equation. If these results are reflected in the design, the design can be more compact or the operation cycle can be increased. These experiments can contribute to the development of AMSB designs.

Assessment of a bolted-flange connection for ITER Test Blanket Module

In ITER, the ‘Pipe Forest (PF)’ is a network of pipes that connects the Ancillary Equipment Unit to the Test Blanket Module (TBM) at the level of the equatorial port-plug of the reactor prototype. The goal of ITER's TBM program is to validate concepts to be adopted for the DEMO tritium Breeding Blankets. Different types of pipes are mounted on the port-plug among which those used for cooling or tritium processing. During plasma operation, the PF has to accommodate severe thermomechanical loads. It shall also provide protection from abnormal contamination and be designed to ease connection/disconnection operations. With numerical simulation, this work assesses the use of bolted flanges as an alternative to the welded solution to connect the PF on the TBMs inside the port cell environment. The proposed designs integrate metallic seals as well as an embedded cooling system in the flanges when necessary, to limit the temperature of the seals and the bolts and prevent irreversible damage. It is shown that bolted-flange junctions for the PF are a credible solution. Still, important stresses in the socket of the flanges are present but do not represent a problematic issue for dimensioning purposes, as they shall be accommodated with a small amount of plastic strain.

Effect of Pebble Size Distribution and Wall Effect on Inner Packing Structure and Contact Force Distribution in Tritium Breeder Pebble Bed

In the tritium breeding blanket of nuclear fusion reactors, the heat transfer behavior and thermal-mechanical response of the tritium breeder pebble bed are affected by the inner packing structure, which is crucial for the design and optimization of a reliable pebble bed in tritium breeding blanket. Thus, the effect of pebble size distribution and fixed wall effect on packing structure and contact force in the poly-disperse pebble bed were investigated by numerical simulation. The results show that pebble size distribution has a significant influence on the inner packing structure of pebble bed. With the increase of the dispersion of pebble size, the average porosity and the average coordination number of the poly-disperse pebble bed gradually decrease. Due to the influence of the fixed wall, the porosity distribution of the pebble bed shows an obvious wall effect. For poly-disperse pebble bed, the influenced region of the wall effect gradually decreases with the increase of the dispersion of pebble size. In addition, the gravity effect and the pebble size distribution have an obvious influence on the contact force distribution inside the poly-disperse pebble bed. The majority of the contact force are weak contact force that is less than the average contact force. Only a few of pebbles have strong contact force that is greater than average contact force. This investigation can help in analyzing the pebble crushing characteristics and the thermal hydraulic analysis in the poly-disperse tritium breeder pebble bed.

Different Oxygen Desorption Durabilities of Lithium Titanium Oxides as Confirmed by Transmission Electron Energy-Loss Spectroscopy Analysis of Li4Ti5O12 and β-Li2TiO3 Biphase Specimens

Spinel-type Li4Ti5O12 and monoclinic β-Li2TiO3 phases are well known as typical functional materials in lithium titanium oxide. The former phase is practically applied as a negative electrode material for Li-ion battery systems, while the latter phase is used as a tritium breeder blanket material for thermonuclear reactors; thus, the material stability of these phases is essential for their functions. In this study, we investigate the oxygen desorption durability of oxide materials with different structures and chemical formulas by transmission electron microscopy-based electron energy-loss spectroscopy. To compare the different properties of these two phases, biphase specimens that consisted of Li4Ti5O12 and β-Li2TiO3 phases with a homogeneous thickness were successfully prepared by a two-step thermal process to perform suitable analysis by transmission electron microscopy. Irradiation with a scanning transmission electron microscopy probe, followed by electron energy-loss spectroscopy analysis, was performed in the Li4Ti5O12/β-Li2TiO3 interphase region of the biphase specimen, revealing that the oxygen desorption feature of β-Li2TiO3 was approximately one-third that of the Li4Ti5O12 phase. The stability of the cation–oxygen bond in each oxide crystal was also simulated by reaction enthalpy analysis based on density functional theory calculations, and the lattice stability of the Li4Ti5O12 phase was estimated to be three times higher than that of the β-Li2TiO3 phase. The oxygen vacancy formation energy of the β-Li2TiO3 lattice was approximately 0.8 eV higher than that of the Li4Ti5O12 phase. Thus, the reason for the different oxygen desorption durabilities of these oxide phases is not the stability of the cation–oxygen bond but the ease of reduction phase formation via oxygen desorption.

Fabrication technologies implemented for the European test blanket modules: Status for HCPB and WCLL

Abstract Within the framework of the European fusion strategy, two tritium Breeder Blankets concepts are developed to be tested in ITER as Test Blanket Modules (TBM): the Water-Cooled Lithium-Lead (WCLL) which uses the liquid Pb-16Li as a breeder, neutron multiplier and tritium carrier, and the Helium-Cooled Pebble-Bed (HCPB) with lithiated ceramic and beryllium pebbles as breeder and neutron multiplier, respectively, and helium purge gas as tritium carrier. Both concepts consider as structural material reduced activation ferritic martensitic steel, the EUROFER97 (X10CrWVTa9-1). Pressurized water (15.5 MPa, 295−328 °C for WCLL) and pressurized helium (8 MPa, 300−500 °C for HCPB) are foreseen for heat removal. The TBM is constituted of a BOX (side caps, first wall), stiffened by stiffening plates (SP) and closed on its back, in the manifold area, with back plates with passing-through elements, like stiffening/tie rods, SP nozzles and inlet/outlet water/He pipes. Internally, breeder units, delimited by SP, are cooled-down by cooling plates for the HCPB and double-wall tubes for the WCLL. This paper describes manufacturing technologies for the TBM and deals with the definition of the preliminary Welding Procedure Specifications (pWPS) and feasibility mock-ups for the manifold area and TBM BOX assembly. The transfer and validation of the pWPS developed for the TBM, from P91 to EUROFER97 steel, is demonstrated. Consolidation of the pWPS for the TBM BOX assembly made with HIP process and development for the HCPB TBM manifold area elements, based on GTAW technique, are discussed. Manufacturing tests are performed using the EUROFER97 steel (RCC-MRx, AFCEN Code rules [ 1 ]). As part of the WCLL conceptual design development activities, an assembly and manufacturing welding sequence is detailed, focusing on design changes and the impact on the selection of manufacturing technologies. The article will present as well the future fabrication developments and the associated design proposals envisaged on TBM.

Effects of Temperature and Pressure on Purge Gas Chemical Compositions in Tritium Breeding Blanket

The effects of temperature and pressure conditions on the equilibrium chemical compositions of purge gas at the outlet of the test blanket module (TBM) in the helium-cooled ceramic reflector (HCCR) are studied. As the chemical species in the equilibrium states, nine chemical species are considered: H, T, O, H2, HT, T2, H2O, HTO, and T2O. The mole fractions of these chemical species are calculated using a Gibbs free energy minimization method starting from the initial state of a H2-HTO mixture. The standard Gibbs free energies for the tritium species used in the study are calculated from the molecular constants obtained by a coupled-cluster calculation. The effects of pressure variations on the equilibrium compositions are shown to be negligible. The effects of temperature variations are also insignificant when the temperature exceeds 800 K. When the initial H2/HTO ratio is more than 10, more than 90% of tritium is expected to be recovered in the form of HT.

An adaptive optimization reverse regulation method for neutron spectrum based on standardized response module

• An adaptive optimization reverse regulation method for neutron spectrum based on standardized response module was developed. • 50 standardized response modules with different thickness and materials were designed for neutron spectrum regulation. • The differential evolution and artificial neural network algorithms are used to optimize the arrangement of response modules. • Verifications on ITER TBM and RBEC-M have confirmed the accuracy and universality of the developed method. The neutron spectrum is one of the essential parameters in the design of advanced nuclear energy systems. Neutron spectra of different nuclear energy systems have wide differences so that accurate regulation of a specific neutron spectrum is of great significance to its development. In this research, an adaptive optimization reverse regulation method for neutron spectrum based on standardized response module was developed. The proposed method designed different types of standardized response modules and calibrated it with their corresponding multi-dimensional response matrix calculated by the Monte Carlo program. Combined with an intelligent optimization algorithm and artificial neural network algorithm, the regulation scheme of the target neutron spectrum could be constructed and modified accurately and efficiently. The 14 MeV single-energy fusion neutron source was used to verify the regulated results of the proposed method on the tritium breeder blanket of a fusion reactor (ITER) whose results showed that the accuracy of the proposed method was improved by 7.69 % compared to the forward manual experiment. Also, a verification of the regulated spectrum on a fission reactor RBEC-M was done to ensure universal application of the method.

Effects of bed dimension, friction coefficient and pebble size distribution on the packing structures of the pebble bed for solid tritium breeder blanket

• Effects of bed width on the packing structures of pebble bed were investigated by DEM simulation. • Effects of pebble properties (friction coefficient between pebbles, pebble size distribution and pebble density) on the packing fraction of pebble bed were simulated and analyzed. • Packing structures of both the tritium breeder (Li 4 SiO 4 , Li 2 TiO 3 ) and the neutron multiplier (Be, Be 12 Ti) pebble beds were simulated and analyzed by DEM. In solid tritium breeder blanket, the packing structures of the tritium breeder pebble bed and the neutron multiplier pebble bed are very important to analyze the tritium breeder ratio and the heat and mass transfer process in the pebble bed. In this study, the numerous simulations were conducted to investigated the effects of bed dimension, friction coefficient between pebbles, pebble size distribution and pebble density on the packing structures of the ceramic tritium breeder (Li 4 SiO 4 , Li 2 TiO 3 ) pebble beds and the neutron multiplier (Be, Be 12 Ti) pebble bed by discrete element method (DEM), respectively. The simulation results reveal that the pebble bed dimension, friction coefficient and pebble size distribution have a significant impact on the pebble bed packing fraction. As the width of the pebble bed container increases, the average packing fraction of the pebble bed gradually increases, meanwhile the influence of volume fraction of the wall effect gradually decreases. In addition, a lower friction coefficient between pebbles (smoother surface of the tritium breeder and the neutron multiplier pebbles) and wider pebble size distribution will result in a higher packing fraction of the pebble bed. However, the material density of the ceramic tritium breeder (Li 4 SiO 4 , Li 2 TiO 3 ) pebbles and the neutron multiplier (Be, Be 12 Ti) pebble seems to have no obvious effect on the packing fraction of the wider pebble bed for the solid tritium breeder blanket.

In-situ oxidation of aluminized stainless-steel to form alumina as tritium permeation barrier coating

Tritium permeation barrier (TPB) coating technology is one of the most effective methods to solve the problem of permeation and leakage of tritium through the structural materials to the coolant, environment, and other functional components in tritium breeder blanket modules (TBMs). Alumina coatings prepared by in-situ oxidation of aluminized steel shows a high permeation reduced factor (PRF) and self-healing ability. However, to eliminate the Al5Fe2 phase in the aluminizing process, the stainless-steel substrate is often heat treated under high temperature conditions for a long time which leads to the degeneration of steel. In this work, a gradient alumina TPB coating was prepared on 316 L stainless steel by aluminizing process at low temperature. The influence of Al5Fe2 on the formation of Al2O3 is studied. The mechanism for the different growth behavior of Al2O3 in the polished and unpolished samples is also discussed. Compared to the polished samples, the existence of porous Al5Fe2 layer in unpolished sample changes the diffusion behavior of Al and leads to the formation of an un-continuous and rough Al2O3 layer.

Neutronics analysis of thorium-uranium based molten salt blanket in stellarator-type FFHR

• Detailed CAD model has been designed for thorium-uranium based molten salt blanket. • Complex geometry structure has been analyzed in coupling with Monte Carlo calculation. • Neutronics features have been simulated in consideration of the burnup effect. • The thickness and the 6 Li enrichment have been optimized in the simulation process. The stellarator-type Fusion-Fission Hybrid Reactor (FFHR) is an alternative conceptual design for the tokamak-based devices in order to realize the stable power output. It can bring the viable energy source for the fusion technology in a foreseeable future, and has been widely studied with a lot of advanced concepts being proposed. However, there are some challenges in the neutronics design of the FFHR blanket for the purpose of obtaining the appropriate Tritium Breeding Ratio (TBR) and blanket energy multiplication factor, since the blanket system has a complicated three-dimensional structure in helical shape, which can not be easily translated into native geometry representation syntax of most Monte Carlo neutron transport codes. In this paper, a liquid fuel, F-Li-Be molten salt, with uranium and thorium dissolved has been proposed for the preliminary design of stellarator-type FFHR blanket. Besides, the CAD-PSFO code [ 19 ] has been employed for the simulation of complex geometry structure in the helical blanket, and the OMCB package [ 22 ] has been used in the burnup analysis of the neutronics features in the molten salt blanket. Moreover, detailed geometry design of the blanket arrangement and thickness have been studied in order to meet the requirement of energy multiplication factor in the stellarator-type FFHR. Also, the composition and concentration of the fuel material and 6 Li in molten salt have been discussed for the purpose of tritium self-sufficiency with steady energy supply in the whole lifetime.

Development of a Thin Film Sensor for Electrochemical Analysis of Molten Fluoride Salts

Due to their favorable heat transfer and neutronic characteristics, as well as chemical stability at high temperatures, molten fluoride salts have been proposed as a coolant in fluoride salt-cooled high-temperature reactors (FHR), a fuel solvent in molten salt reactors (MSR), and as a tritium breeding blanket in fusion reactors. A fundamental understanding of the chemistry of molten salts is extremely important for advances in these fields. Electrochemical techniques encompass a suite of tools that can be used to probe corrosion mechanisms as well as study the chemical composition and various chemical and physicochemical properties. The tools available to investigate the chemistry of molten salts are limited, however, due to the challenges in instrument design associated with high-temperature operation, material compatibility and chemical stability of all sensor construction materials. Here we report the fabrication and characterization of a thin film sensor for electrochemical analysis of molten fluoride salts. Photolithographic patterning and etching, as well as other standard deposition techniques are used to deposit electrode and supporting materials onto a planar substrate. A typical three-electrode configuration, comprising a working, counter, and quasi-reference electrode, is employed, and incorporation of a fourth auxiliary electrode permits the use of the dynamic reference electrode procedure to provide a common reference potential. Optimized device design and material selection ensure stability of the probe under the demanding chemical and thermal conditions of molten salt systems. By minimizing the required quantity of analyte salt and providing a standardized electrode architecture, this fully integrated device enables high-throughput, reproducible experimentation.

Oxide Sensor Development By Square Wave Voltammetry for Molten Fluoride Salts

A fundamental understanding of oxide concentration is critical for operation of advanced nuclear reactors which use molten salts. Molten fluoride salts have been proposed as fuel solvents, coolants, and tritium breeding blankets in fission and fusion nuclear reactors due to their high temperature operation and favorable neutron-interaction properties. An electrochemical sensor which can measure the real-time concentration of oxide anions in the salt melt could be used to identify air ingress to the system. Additionally, similar electrochemical analyses used in the development of a sensor could be used to study the fundamental properties and behavior of oxides in molten fluoride salts. These properties include the activity coefficient, diffusion coefficient, and solubility of oxides. Understanding of these properties could, in turn, inform sensor development.A further understanding salt chemistry can also help predict and control the effects of oxide concentration on salt corrosion. Material integrity under exposure to high-temperature molten salts is an impactful technical challenge for the design of nuclear reactors that employ molten salts. Oxide anion concentration and complexation of cations and anions in the melt can influence corrosion. Therefore, electrochemical oxide sensors for molten fluoride salts would be instrumental in corrosion studies. Fundamental understanding of oxide chemistry in fluorides would also advance understanding of corrosion mechanisms in fluoride salts.Cyclic voltammetry, square wave voltammetry, and chronopotentiometry have been previously used to study the electrochemical behavior of oxides in fluoride salt melts. Square wave voltammetry (SWV) is particularly useful in this application because it limits the noise caused by oxygen bubbling from the oxidation of O2- anions, as found by Massot et al.. Titrations of known additions of oxide to the melt have previously been used to show that the intensity of the SWV oxide oxidation peak is linear with oxide concentration.Oxides have yet to be studied electrochemically in 2LiF-BeF2 (FLiBe) or LiF-NaF-KF (FLiNaK) molten salts, which are of interest for nuclear reactors. FLiBe is a likely coolant and solvent candidate for molten salt nuclear reactors. FLiNaK has been studied as a surrogate salt with similar thermo-physical properties to FLiBe. It is easier to handle experimentally since it presents no beryllium health hazard, but its chemical and neutronic behavior are different. FLiNaK is often used in salt loops for reactor design, safety analysis, component testing, and corrosion testing. Therefore, availability of oxide electrochemical sensors for FLiBe and FLiNaK would be instrumental in reactor operation as well as reactor development.This talk will summarize the application of SWV to quantification of oxides in FLiBe and FLiNaK melts. It will also discuss the available chemical and thermodynamic information for oxides in FLiBe and FLiNaK. Ultimately, an understanding of oxide behavior and a quantification of oxide concentration in molten fluoride salts will aide in the development of an electrochemical sensor for detecting leaks in molten salt reactors.

Ab initio study of hydrogen diffusion in Be and Be12Ti for fusion applications

Beryllium and Be12Ti are considered as neutron multiplier materials in various tritium breeding blanket designs for future fusion reactors. Tritium and helium are formed due to neutron-induced nuclear transmutation and accumulated in beryllium-based materials. As a β-radioactive isotope, tritium represents additional safety concern with respect to sudden release during operation or hindering consequent recycling. Therefore, detailed study of its accumulation and release is indispensable. In particular, it is important to reveal which process controls tritium release: binding of tritium with traps or diffusion out of material bulk. Ab initio calculations of solution energies at all non-equivalent hydrogen interstitial sites confirm easier dissolution of tritium in Be12Ti in comparison with pure beryllium. The 17 considered diffusion barriers between adjacent hydrogen interstitial sites are calculated using a dimer method instead of a computationally more expensive nudged elastic band approach. Three possible diffusion paths in Be12Ti were revealed: two of them along the c-axis and another one along the a- and b-axes. Considering calculated hydrogen diffusion barriers in Be12Ti, one can conclude that at low hydrogen concentrations and low temperatures, diffusion occurs preferentially along the c-axis. The rate limiting diffusion barriers in Be12Ti are found to be in the range of 0.4–0.5 eV, which are comparable to 0.42 eV in pure beryllium suggesting that tritium diffusion in both materials proceeds at comparable rates. Therefore, binding of tritium atoms with vacancies and other lattice imperfections are expected to play a crucial role in tritium retention.

Design and properties of FeAl/Al2O3/TiO2 composite tritium-resistant coating prepared through pack cementation and sol–gel method

Structural steel materials exist in tritium breeding blankets and auxiliary tritium handling systems in fusion reactors. Applying a tritium permeation barrier coating on the surface of the structural material is necessary to prevent material corrosion and embrittlement. In this study, FeAl/Al2O3/TiO2 composite tritium-resistant coating on 316 L stainless steel substrate was successfully prepared by pack cementation and sol–gel method. The surface and cross-section morphologies of the composite coatings were systematically studied. The hydrogen permeation properties and adhesion between the different composite coatings were investigated. Results show that the FeAl/Al2O3/TiO2 composite tritium resistant coating is uniform, dense, and preferably connected to the substrate without defects such as cracks and pores. Compared with FeAl/Al2O3 coating, FeAl/Al2O3/TiO2 coating has better bonding strength, toughness, and ductility. Hydrogen permeation test shows that the design of the FeAl/Al2O3/TiO2 composite coating improves the hydrogen resistance of the coating at room temperature.

Bulk porous materials as an alternative to pebble beds in tritium breeding blankets

• A new porous Li 4 SiO 4 bulk breeder was prepared by prefabricated in mold followed by removal of the organic binder. • The porous Li 4 SiO 4 bulk breeder contains an interconnected tritium release channel structure (K g ∼107 m 3 /m 2 h kpa). • The new porous Li 4 SiO 4 bulk breeding material has substantially higher density (∼72 %) compared with pebble beds. • Mechanical strength tests demonstrated good structure integrity and low elastic modulus (0.575 GPa). In this work, a new porous Li 4 SiO 4 bulk breeder was prepared by a novel synthesis method of compression molding-sintering technique based on organic binder. This process results in a denser bulk configuration breeder with an interconnected channel structure for tritium release, following prefabricated in mold the organic binder is removed by volatilization. we also studied in detail the effects of sintering temperature on the morphology, porosity, permeability and mechanical properties of porous bulk breeder. The results show that with the sintering temperature increasing from T = 0.7 T m to T = T m , the porosity of the bulk breeder decreases continuously, accompanied with an increase of compressive strength except for in the range of T ≥ 0.9T m . The main reason is that thermal stress leads to micro-cracks in Li 4 SiO 4 material under higher cooling rate. In summary, the as-prepared porous bulk material at 0.9 T m , aside from the increased density relative to pebble beds (Packing factor ∼71.62 % vs ∼ 63 %), it also has high phase purity, satisfactory air permeability (107 m 3 /m 2 h kPa), the excellent mechanical properties (2.89 MPa) and increased thermal durability, which will be a good candidate for the fusion reactor breeding blanket.

Particle-resolved simulation of randomly packed pebble beds with a novel fluid-solid coupling method

The Helium Cooled Pebble Bed (HCPB) is a common type of fusion reactor. An in-depth understanding of the helium flow behaviors in the interstitial void space between pebbles and the influence of hydrodynamics on the efficiency of purging tritium is important for the better design of the tritium breeding blanket. Particle-resolved simulation strategy has been widely used to model the HCPB system but generally suffers a tedious mesh generation process due to the complex geometries. To overcome this difficulty, this study develops a novel fluid-solid coupling strategy, p-IBM, which combines the porosity-based mesh adaption and the immersed boundary method. A small pebble bed consisting of 158 particles is first simulated with this method and good agreement with the traditional particle-resolved simulation using body-fitted grids is achieved. The proposed p-IBM method is then applied to simulate a larger pebble bed containing 3190 particles, where boundary wall effects on key parameters such as the pressure drop, fluid channeling and purging process of tritium are evaluated and discussed.

Evaluation of fatigue properties of reduced activation ferritic/martensitic steel, F82H for development of design criteria

Reduced activation ferritic/martensitic (RAFM) steels are promising candidate for structural material of tritium breeding blanket in a fusion reactor. Database accumulation and definition of design criteria for RAFM have been intensively studied together with the progress of blanket design activities. As a part of database accumulation for fusion blanket, a RAFM steel, F82H was fatigue-tested at 573 and 673 K in the air. Axial strain-controlled fatigue tests were carried out with a cylindrical specimen with 8 mm of diameter with -1 of strain ratio condition in accordance with Japanese Industrial Standard, JIS Z 2279, “Method of high-temperature low cycle fatigue testing for metallic materials.” For high cycle tests, the maximum test cycles exceeded 106 cycles. Fatigue lifetime of F82H at temperatures ranging from room temperature to 673 K fell into the factor of 2 of the empirical fitting curve for 673 K. It was studied using experimental results that fatigue-related design limit based on RCC-MRx, such as fatigue design curves, half-life cyclic curves, and related coefficients.

Preliminary performance analysis and optimization based on 1D neutronics model for Indian DEMO HCCB blanket

India, under its breeding blanket R&D program for DEMO, is focusing on the development of two tritium breeding blanket concepts; namely the lead-lithium-cooled ceramic breeder and the helium-cooled ceramic breeder (HCCB). The study presented in this paper focuses on the neutronic design analysis and optimization from the tritium breeding perspective of the HCCB blanket. The Indian concept has an edge-on configuration and is one of the variants of the helium-cooled solid breeder blanket concepts proposed by several partner countries in ITER. The Indian HCCB blanket having lithium titanate (Li2TiO3) as the tritium breeder and beryllium (Be) as the neutron multiplier with reduced-activation ferritic/martensitic steel structure aims at utilizing the low-energy neutrons at the rear part of the blanket. The aim of the optimization study is to minimize the radial blanket thickness while ensuring tritium self-sufficiency and provide data for further neutronic design and thermal-hydraulic layout of the HCCB blanket. It is found that inboard and outboard blanket thicknesses of 40 cm and 60 cm, respectively, can give a tritium breeding ratio (TBR) >1.3, with 60% 6Li enrichment, which is assumed to be sufficient to cover potential tritium losses and associated uncertainties. The results also demonstrated that the Be packing fraction (PF) has a more profound impact on the TBR as compared to 6Li enrichment and the PF of Li2TiO3.

Oxide Sensor Development By Square Wave Voltammetry for Molten Fluoride Salts

A fundamental understanding of oxide concentration is critical for operation of advanced nuclear reactors which use molten salts. Molten fluoride salts have been proposed as fuel solvents, coolants, and tritium breeding blankets in fission and fusion nuclear reactors due to their high temperature operation and favorable neutron-interaction properties. An electrochemical sensor which can measure the real-time concentration of oxide anions in the salt melt could be used to identify air ingress to the system. Additionally, similar electrochemical analyses used in the development of a sensor could be used to study the fundamental properties and behavior of oxides in molten fluoride salts. These properties include the activity coefficient, diffusion coefficient, and solubility of oxides. Understanding of these properties could, in turn, inform sensor development.A further understanding salt chemistry can also help predict and control the effects of oxide concentration on salt corrosion. Material integrity under exposure to high-temperature molten salts is an impactful technical challenge for the design of nuclear reactors that employ molten salts. Oxide anion concentration and complexation of cations and anions in the melt can influence corrosion. Therefore, electrochemical oxide sensors for molten fluoride salts would be instrumental in corrosion studies. Fundamental understanding of oxide chemistry in fluorides would also advance understanding of corrosion mechanisms in fluoride salts.Cyclic voltammetry, square wave voltammetry, and chronopotentiometry have been previously used to study the electrochemical behavior of oxides in fluoride salt melts. Square wave voltammetry (SWV) is particularly useful in this application because it limits the noise caused by oxygen bubbling from the oxidation of O2- anions, as found by Massot et al. Titrations of known additions of oxide to the melt have previously been used to show that the intensity of the SWV oxide oxidation peak is linear with oxide concentration.Oxides have yet to be studied electrochemically in 2LiF-BeF2 (FLiBe) or LiF-NaF-KF (FLiNaK) molten salts, which are of interest for nuclear reactors. FLiBe is a likely coolant and solvent candidate for molten salt nuclear reactors. FLiNaK has been studied as a surrogate salt with similar thermo-physical properties to FLiBe. It is easier to handle experimentally since it presents no beryllium health hazard, but its chemical and neutronic behavior are different. FLiNaK is often used in salt loops for reactor design, safety analysis, component testing, and corrosion testing. Therefore, availability of oxide electrochemical sensors for FLiBe and FLiNaK would be instrumental in reactor operation as well as reactor development.This talk will summarize the application of SWV to quantification of oxides in FLiBe and FLiNaK melts. It will also discuss the available chemical and thermodynamic information for oxides in FLiBe and FLiNaK. Ultimately, an understanding of oxide behavior and a quantification of oxide concentration in molten fluoride salts will aide in the development of an electrochemical sensor for detecting leaks in molten salt reactors.