Measurement of tritium production rate distribution in natural LiAlO2/HDPE assembly irradiated by D-T neutrons

A neutronics experiment was performed to measure the tritium production rate (TPR) profile in the breeder assembly with LiAlO₂ as breeder and high density polyethylene (HDPE) as neutron reflector. The breeder assembly was irradiated with 14 MeV neutrons from DT neutron generator at IPR Neutronics Laboratory. The objective of the experiment was to validate the tritium production prediction capability of the Monte-Carlo code MCNP and FENDL 2.1 data library. The tritium production rate profile in the breeding assembly was measured by irradiating Li₂CO₃ pellets kept at various locations and then tritium counting liquid scintillation technique. Experiment was analyzed with 3D Monte-Carlo code MCNP with FENDL 2.1 cross-section data library. The calculation results were found to agree with the measured tritium production rates except one point near to the source. This experiment is a starting experiment in the series of benchmark experiments for the Indian Demo breeding blanket.     

Design optimisation and measuring techniques for the neutronics experiment on a HCLL–TBM mock-up

A neutronics experiment on a mock-up of the EU Test Blanket Module (TBM), helium cooled lithium lead concept, is in preparation with the objective to validate the capability of the neutronics codes and nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. Three independent measurements of the TPR will be performed using Li₂CO₃ pellets. Other measurement techniques have been developed using thermo-luminescence detectors, and diamond detectors covered with ⁶LiF. Neutron flux spectra will also be measured from fast energies down to thermal energies, relevant for TPR. Comparison of measured quantities (E) with the same calculated quantities (C) will be provided, together with the related uncertainties.The paper presents the results of development of the measurement techniques and their relevance for tritium measurements in TBM in ITER. It presents also the pre-analyses conducted to optimise the mock-up configuration so that the neutron spectra are as similar as possible to those in the TBM in ITER. Sensitivity/uncertainty assessments of the TPR show that the calculation uncertainty due to the uncertainties of the neutron cross sections amounts to a few %, depending on position. The largest uncertainties are due to the elastic scattering (n,2n), and (n,3n) reactions on Pb.     

Neutronic analysis of an inertial fusion energy breeder with SiCf/SiC

Neutronic performance of a blanket driven ICF (Inertial confinement fusion) neutron based on SiC_f/SiC composite material is investigated for fissile fuel breeding. The investigated blanket is fueled with ThO₂ and cooled with natural lithium or (LiF)₂BeF₂ or Li₁₇Pb₈₃ or ⁴He coolant. MCNP4B Code is used for calculations of neutronic data per DT neutron. Calculations have show that values of TBR (tritium breeding ratio) being one of the main neutronic paremeters of fusion reactors are greater than 1.05 in all type of coolant, and the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Calculations show that natural lithium coolant blanket has the highest TBR (1.298) and M (fusion energy multiplication) (2.235), Li₁₇Pb₈₃ coolant blanket has the highest FFBR (fissile fuel breeding ratio) (0.3489) and NNM (net neutron multiplication) (1.6337). ⁴He coolant blanket has also the best Γ (peek-to-average fission power density ratio) (1.711). Values of neutron leakage out of the blanket in all type of coolants are quite low due to SiC reflector and B₄C shielding.     

Application of a 6LiF Small Neutron Detector with an Optical Fiber to Tritium Production Rate Measurement in D-T Neutron Fields

⁶LiF small neutron detectors with an optical fiber have been used to measure ⁶Li(n,α)T reaction rate distributions at thermal research reactors and accelerator facilities. In the present study, we developed an experimental method for the measurement of tritium production rate (TPR) of ⁶Li using this small detector in deuterium-tritium (D-T) neutron fields. Reaction rate measurements with the detector were conducted in the D-T neutron fields at the Fusion Neutronics Source (FNS) facility. From the results, we determined that this detector can be used to measure the TPR distribution in soft neutron spectrum fields such as in a Be assembly. It is difficult to obtain ⁶Li(n,α)T reaction rate separately in hard neutron spectrum fields such as in a Li₂O assembly, because many kinds of charged particle production reactions need to be taken into consideration. However, a time-dependent reaction rate measurement method combined with the ⁶LiF detector and the ZnS detector is effective to separate the ⁶Li(n,α)T reaction from other reactions even in a hard spectrum field, and it can be applied to the measurement of the TPR distribution accurately.     

CFETR水冷陶瓷增殖剂包层模块产氚率实验验证

水冷陶瓷增殖剂（WCCB）包层作为中国聚变工程试验堆（CFETR）候选包层之一，承担着氚增殖、核热提取、屏蔽等重要涉核功能，其中子学设计的可靠性直接影响CFETR氚自持目标的实现。为验证中子学设计工具，即MCNP和FNEDL3.0数据库，在WCCB包层中子学设计中的可靠性，基于研制出的WCCB包层模块，在DT中子环境下开展中子学实验，对以产氚率（TPR）为代表的中子学参数进行了模拟值（C）和实验值（E）对比分析。结果表明，模块中轴线位置处TPR的C/E为0.97?1.08，而模块边缘位置处TPR的C/E为0.65?0.82；模块钛酸锂层边缘区197Au（n，γ）198Au反应率的C/E为0.72?0.90，表明模块边缘区存在非期望的散射中子，导致该区TPR模拟值和实验值偏离较大。     

The TFTR Lithium Blanket Module program

The Lithium Blanket Module (LBM) is an approximately 80×80×80 cm cubic module, representative of a helium-cooled lithium oxide fusion reactor blanket module, that will be installed on the TFTR (Tokamak Fusion Test Reactor) in late 1986. The principal objective of the LBM Program is to perform a series of neutron transport and tritium-breeding measurements throughout the LBM when it is exposed to the TFTR toroidal fusion neutron source, and to compare these data with the predictions of Monte Carlo (MCNP) neutronics codes. The LBM consists of 920 2.5-cm diameter breeder rods constructed of lithium oxide (Li₂O) pellets housed in thin-walled stainless steel tubes. Procedures for mass-producing 25,000 Li₂O pellets with satisfactory reproducibility were developed using purified Li₂O powder, and fabrication of all the breeder rods was completed in early 1985. Tritium assay methods were investigated experimentally using both small lithium metal samples and LBM-type pellets. This work demonstrated that the thermal extraction method will be satisfactory for accurate evaluation of the minute concentrations of tritium expected in the LBM pellets (0.1–1 nCi/g).     

Neutronic analysis of PROMETHEUS reactor fueled with various compounds of thorium and uranium

In this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H (heavy ion) fueled with different fuels, namely, ThO₂, ThC, UO₂, UC, U₃Si₂ and UN is investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission products from contaminating coolant and direct contact fuel with coolant in the blanket. Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3 Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN has the highest value of ²³⁹Pu breeding capability among the uranium fuels whereas UO₂ has the lowest one. ²³⁹Pu production ratio changes from 0.119 to 0.169 according to the uranium fuel types, and ²³³U production values are 0.125 and 0.140 in the blanket fueled with ThO₂ and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron load varies from 1.30 to 7.89 W/cm³ in the first row of fissile fuel breeding zone depending on the fuel types. Heat production attains the maximum value in the blanket fueled with UN. Values of TBR (tritium breeding ratio) being one of the most important parameters in a fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is maintained for DT fusion driver. Values of peak-to-average fission power density ratio, Γ, are in the range of 1.390 and ∼1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum neutron leakage is only ∼0.025. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction.     

Preliminary spectrum shifter design for intermediate energy nuclear data benchmark experiments with DT neutrons

Integral benchmark experiments with DT neutrons are not always sufficient for nuclear data benchmarking in the MeV region, below 10 MeV. A neutron spectrum shifter, which will be placed between a sample and a DT neutron source, is effective to moderate DT neutrons incident to the sample. In order to estimate effects of the spectrum shifter, the ratio of the contribution of 14 MeV neutrons in the leakage neutron and gamma-ray spectra was calculated with MCNP-4C for an experimental configuration at FNS of JAEA, Japan. The calculations were carried out for a Li₂TiO₃ sample with a Be, D₂O, or ⁷LiD spectrum shifter. It was found out that the Be shifter was superior to others and the Be shifter was effective to decrease the contribution of 14 MeV neutrons especially for secondary gamma-ray spectrum measurements.     

Effect of Nuclear Data Libraries on Tritium Breeding in a (D–T) Fusion Driven Reactor

In design a Deuterium–Tritium (D–T) fusion driven hybrid reactor, neutronics and nuclear data libraries have an essential role for reliable neutronics calculations. Therefore, nuclear data libraries are very important to calculate of the neutronic parameters and selection of tritium breeder materials to be used in the blanket. In this study tritium breeding performances of candidate tritium breeding materials, namely, Li₂O, LiH, Li₂TiO₃, Li₂ZrO₃ and Li₄SiO₄ in a (D–T) driven fusion–fission (hybrid) reactor is investigated based on three dimensional (3-D) and one dimensional (1-D) neutronic calculations. 3-D and 1-D neutron transport calculations are performed with Monte Carlo transport code (MCNP 4C), SCALE 5 and ANISN nuclear data codes to determine the tritium breeding ratio (TBR) of the blanket. The effects of different nuclear data libraries on TBR are examined and TBR calculation results are comparatively investigated.     

Isotope Exchange Capacity on Li4SiO4 and Comparison of Tritium Inventory in Various Solid Breeder Blankets

It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of a breeder material, absorption of water vapor into bulk of the grain, adsorption of water on surface of the grain, and exchange capacity of tritium to be trapped to surface of the grain together with two types of isotope exchange reactions for evaluation of the tritium inventory in a solid breeder blanket under various conditions. The isotope exchange capacity on the Li₄SiO₄ surface is experimentally obtained in this study. Most of the properties required for evaluation of the tritium inventory for various blanket materials have been already quantified by the present authors. Then it has become possible to compare the tritium inventory in solid breeder blankets packed with either Li₂O, LiAlO₂, Li₂ZrO₃, Li₂TiO₃ or Li₄SiO₄ using the calculation model previously presented by the present authors.     

Neutronics Calculations in Pellets and Blankets of Laser Fusion Reactor Concept SENRI-I

The neutronic properties of SENRI-I, a reference design of laser fusion reactor proposed by Institute of Engineering, Osaka University, are discussed on the basis of the one-dimensional neutron transport calculations in burning DT plasmas and blankets. The softening of the fusion neutron energy spectrum, the neutron heating and the neutron multiplication are studied and discussed for the compressed DT pellets with various thickness of fuel plasmas and lead or lead-polyethylene tampers.

The neutronic and thermal features in the blanket of the SENRI-I design are also examined. The tritium breeding ratio is high enough (~1.6), depending on the neutron energy spectrum from a pellet. The maximum temperature increase per 1,000 MJ DT fusion reactions is ~3°C in the inner liquid Li layer and ~1.5°C in the stainless steel first wall. A parametric study is also presented on the effect of varying the thickness of the inner Li blanket ΔRⁱ to examine the thickness required for the enough tritium breeding ratio and energy deposition.     

R&D status on Water Cooled Ceramic Breeder Blanket Technology

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li₂TiO₃ pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.     

Benchmark experiment on titanium with DT neutron at JAEA/FNS

Titanium is contained in lithium titanate which is a tritium breeding material candidate. In the nuclear design, accurate nuclear data are needed. However, few benchmark experiments had been performed for titanium. We performed a benchmark experiment with a titanium assembly and a DT neutron source at JAEA/FNS. The titanium assembly was covered with Li₂O blocks in order to reduce background neutrons. Dosimetry reaction rates were measured with niobium, indium and gold foils inside the assembly. And fission rates of ²³⁵U were measured by using micro fission chambers. This experiment was analyzed by using the Monte Carlo neutron transport code MCNP5-1.40 with recent nuclear data libraries of ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0 and JENDL-4.0u1. The calculation results were compared with the measured one in order to validate the nuclear data libraries of titanium. The calculated results with ENDF/B-VII.1 agreed with the measured one the best because the (n,2n) and (n,n′_cont) reaction cross section data and resonance parameters were improved.     

Overview of design and thermal–hydraulic analysis of Indian solid breeder blanket concept

India has developed two concepts of breeding blanket for the DEMO reactor: one is Lead Lithium Ceramic Breeder (LLCB), and the other one is Helium-cooled Ceramic Breeder (HCCB) concept. Indian HCCB concept is having edge on configuration of helium-cooled solid breeder with RAFMS structure. Li₂TiO₃/Li₄SiO₄ and beryllium are used as the tritium breeder and neutron multiplier, respectively. 2D thermal–hydraulic simulation studies using ANSYS have been performed based on the heat load obtained from neutronics calculations to confirm heat removal under ITER pulsed operation. Transient thermal analysis has been simulated in ANSYS for the ITER relevant operational conditions. Thermal analysis provides important information about the temperature distribution in different materials used and their temperature–time histories. Result of thermal–hydraulic simulations shows that in each cycle, the maximum temperature of all materials remains same. The peak temperatures of all materials are well within their limiting value. Concept designs of HCCB blanket and its thermal hydraulic analysis will be presented in this paper.     

Effect of sweep gas species on tritium release behavior from lithium titanate packed bed during 14 MeV neutron irradiation

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li₂TiO₃) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.     

Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor’s operational duration, low failure percentage, short maintenance time and the inclusion of the system’s simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B–V–VI nuclear data. Around the fusion chamber, molten salts Flibe (LI₂BeF₄), lead–lithium (PbLi), Li–Sn, thin-lityum (Li₂₀Sn₈₀) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI₂BeF₄) and Li–Sn, for blanket layers. In case of usage of the Flibe (LI₂BeF₄), PbLi, and thin-lityum (Li₂₀Sn₈₀) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.     

Preparation and characterization of lithium-titanate pebbles by solid-state reaction extrusion and spherodization techniques for fusion reactor

For the development of TBM for fusion reactors, lithium containing ceramics as against the metal are preferred as tritium breeding material. Lithium titanate (Li₂TiO₃) is one such chosen ceramic tritium breeder. Li₂TiO₃ pebbles are conventionally prepared by sol-gel process and wet process. Solid state reaction of lithium carbonate with titanium dioxide is preferred route for the bulk production of Li₂TiO_3. Thermo-gravimetric and differential thermal analysis (TG-DTA) techniques have been used in the present study to understand the solid state reaction of intimate mixture of lithium carbonate and titanium dioxide. It was found out that single phase lithium titanate (Li₂TiO₃) is produced at 750 °C and the reaction is completed in 6 h. Fine powders of lithium titanate obtained after milling and classification were mixed with aqueous solution of PVA to prepare green pebbles of desired size and shape. The pebbles were subsequently sintered at 900 °C and the effect of sintering time on the properties of sintered pebbles was studied. The reaction mechanisms and the product qualities obtained by the solid state reaction, extrusion and spherodization techniques are discussed in this paper.     

Neutronics and photonics of inertial confinement fusion pellets with internal breeding

The neutronics and photonics performance of a pellet with a small DT core spark trigger, surrounded by a large volume of D to enable tritium and He-3 breeding, is examined. The response to a 70% DD and 30% DT composite neutron spectrum is calculated using either W, Be, or Pb as structural materials at core density radius products ranging from 9.42 to 94.2 kg/m². At a core density-product of 94.2, the DT neutron source leads to an excess particle multiplication of 0.43 neutrons per source neutron. The percentage of energy leakage from the pellet in the form of escaped neutrons is 42.3% of the source energy for the DT source, and 28.8% for the DD source. The gamma-ray energy percentage deposited in the pellet is 26.7% for the DT source and 106.6% for the DD source. For the pellet with the composite source, the energy multiplication factor is 1.27. Thus the large DD contribution to the composite neutron source results in the pellet performing many of the functions normally reserved for the blanket such as spectral softening, breeding, and neutron and energy multiplication. The neutron energy leakage is 38.4% of the source energy for the composite source. It is estimated that the neutron energy leakage amounts to 10% of the fusion energy, compared with 70% as neutron energy in a DT pellet. These results are significantly different from those encountered in conventional DT inertial confinement designs, and thus lower tritium inventories, higher power densities, reduced radiation damage, and materials activation of the reactor coolant and structure may be achievable.     

Application of Alpha-Track Method to Tritium Production Rate Measurements

The α-track method with a solid state nuclear track detector was used for the measurement of tritium production rates (TPR). Cellulose nitrate track detectors (Kodak LR-115) with Li-containing radiators were placed in a simulated fusion blanket of a Li₂O slab assembly with a Be layer and irradiated by D-T neutrons. After chemical etching, the etched-through tracks in the detectors were counted. Lithium metal was the most suitable radiator to distinguish tracks formed by α-particles of the LI(n, α)T reaction from the background. The experimental results agreed well with the calculations. This shows that the indirect α-track method is reliable in estimating TPR distribution.     

Octalithium plumbate as breeding blanket ceramic: Neutronic performances,synthesis and partial characterization

A neutronic assessment of the performances of a helium-cooled Li₈PbO₆ breeding blanket (BB) for the conceptual design of a DEMO fusion reactor is given. Different BB configurations have been considered in order to minimize the amount of beryllium required for neutron multiplication, including the use of graphite as reflector material. The calculated neutronic responses: tritium breeding ratio (TBR), power deposition in TF coils and power amplification factor, indicate the feasibility of Li₈PbO₆ as breeding material. Furthermore, the synthesis and characterization of Li₈PbO₆ by X-ray phase analysis are also discussed.