Synthesis of Li2O Single Crystal by Vacuum Fusion Technique

Detailed neutronics analysis on a tokamak fusion experimental reactor was performed. Neutronics responses of components in the vacuum vessel of a swimming pool type reactor were calculated by Monte Carlo method considering toroidal geometry, asymmetrical torus cross section, and neutron source distribution in plasma. It was found that the neutronics responses in the first wall and the blanket vary significantly in the poloidal direction. Effect of penetrations of neutral beam injectors on the tritium breeding performance was also evaluated in the analysis. The result shows that the effect is rather small in the present structure. As a result of the analysis, it becomes clear that modification of blanket structure and material composition should be done in order to improve the tritium breeding performance and to reduce the nuclear heating rate of the vacuum vessel in the divertor zone.     

CFETR氦冷陶瓷增殖包层中子学分析

为满足中国聚变工程实验堆(CFETR)包层的应用要求,本文提出氦冷陶瓷增殖(HCCB)包层方案。为验证HCCB包层设计方案的合理性与可行性,采用三维蒙特卡罗粒子输运程序MCNP,计算和分析了HCCB包层方案的氚增殖比、中子壁负载、中子通量密度、核热、辐照损伤等中子学特性。结果表明,HCCB包层方案满足氚自持要求,中子通量密度和核热分布合理,屏蔽性能良好,基本满足设计要求。     

First wall surface problems for a D ↽ T tokamak reactor

The effects of deuterium, tritium, helium and neutron bombardment on surface degradation of the first wall of a 5000 MW_th D-T reactor have been analyzed. The effects of both sputtering and blistering have been analyzed and the results applied to 316 stainless steel wall operating at temperatures from 300 to 500°C. It has been calculated that the total wall erosion rate is 0.22 mm/year and that 14 MeV neutron sputtering accounts for two thirds of this number. Sputtering from all neutrons results in $\text{≈0.17}\text{mm/year}$ erosion. The calculated erosion rate is 2–3 times that which would be allowable for a 30 year first wall lifetime.     

Neutronics Experimental Study on Tritium Production in Solid Breeder Blanket Mockup with Neutron Reflector

Neutronics experiments have been performed for the solid breeder blanket using a DT neutron source at the FNS facility in JAEA. We have applied the blanket mockup composed of two enriched Li₂TiO₃ and three beryllium layers, and measured the detailed spatial distribution of the tritium production rate (TPR) using enriched Li₂CO₃ pellets. TPRs in the pellets have been measured by a liquid scintillation counter. Experiments have been done under a condition with a neutron reflector surrounding the DT neutron source. Numerical simulations have been performed using the MCNP-4C with the FENDL-2.0 and JENDL-3.3. The ranges of ratios of calculation results to experimental ones (C/Es) are 0.97–1.17 concerning with local TPR, and 1.04–1.09 for the integrated tritium production. It is found that the total integrated tritium production, which corresponds to tritium breeding ratio, can be predicted within uncertainty of 10% using the Monte Carlo calculation code and latest nuclear data libraries.     

The dynomak: An advanced spheromak reactor concept with imposed-dynamo current drive and next-generation nuclear power technologies

A high-β spheromak reactor concept has been formulated with an estimated overnight capital cost that is competitive with conventional power sources. This reactor concept utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt (FLiBe) blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER-developed cryogenic pumping systems were implemented in this concept from the basis of technological feasibility. A tritium breeding ratio (TBR) of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%.     

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5otorus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models,shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1,the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined.The results indicate that the global TBR of no less than 1.2 will be a big challenge for the watercooled ceramic breeder blanket for CFETR.     

Neutron Total Cross Section for Iron around Resonance Valley near 24 keV

A neutron leakage spectrum from a nickel sphere surrounding a 14-MeV neutron source is measured and analyzed in order to verify the accuracy of nickel cross sections. Measurement is done by means of a time-of-flight technique in the range of 2–15 MeV using an NE213 scintillator, and compared with calculations carried out with MCNP, a continuous energy Monte Carlo transport code, using JENDL-3PR1 and ENDFIB-IV neutron libraries.

In spite of an overestimation of neutron flux near 13 MeV, the calculated result employing JENDL-3PR1 shows generally better agreement with the measured spectrum. In ENDF/B-IV usage, there is disagreement between measured and calculated spectra between 5 and 12 MeV. Problems in evaluated nuclear data for nickel are also described.     

Estimation of time-dependent neutron transport from point source based on Monte Carlo power iteration

This study proposes a method for calculating time-dependent neutron transport from a point source with a continuous-energy Monte Carlo code. To deal with neutron multiplication and attenuation in orders of magnitude, the power iteration method conventionally used to estimate the effective multiplication factor k_eff was utilized. The time of a neutron flying in a cycle from emission of its ancestor at the point source was estimated. In the estimation, the decay time of the delayed neutron precursor was considered. The neutron flux was tallied in time bins in each cycle. The source strength in the cycle was considered as the product of k_eff estimators from the first to the previous cycle. By summing up the tallied flux multiplied by the strength, the neutron flux variation with time was obtained. This method was verified against a UO₂ fuel lattice moderated and reflected by light water.     

Pu-breeding feasibility in PWR

This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO₂ or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived.     

Neutronics analysis for the moving ring fusion reactor,KARIN-1

Monte Carlo calculations on the tritium breeding ratio in the blanket and the heating rate in the first wall were carried out for moving ring reactor, KARIN-1. The blanket adopted in this calculation consists of a 2-cm thick silicon carbide (SiC) first wall, a 25-cm thick lithium (Li) layer, and a 10-cm thick lithium oxide (Li₂O) layer. It was found that the tritium breeding ratio in the blanket and the heating rate in the first wall are changing cyclically with a 1 s period. The maximum tritium breeding ratio obtained is 1.12, the minimum is 1.01, and the average value over a 1 s period is 1.10. As for the heating rate at the center of the burning section, the maximum is ～26 W/cc, the minimum ～10 W/cc, the average ～ 15 W/cc, and the ratio of the maximum to the minimum is ～2.6. The corresponding values at the entrance of the burning section are ～24 W/cc, ～4 W/cc, ～12 W/cc, and ～6.7, respectively. These results provide the basic information for thermal and mechanical design of the reactor system.     

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.     

球谐函数有限元程序NECP-FISH的开发及其在聚变堆包层中子学分析中的应用

球谐函数有限元方法采用非结构网格求解中子输运方程,具备处理复杂几何的能力;同时又可避免离散纵标方法所造成的射线效应。本文从一阶中子输运方程出发,通过方程的弱形式推导了球谐函数多尺度有限元方法,并基于此方法开发了中子学分析程序NECP-FISH。通过在前后处理平台SALOME中开发接口程序,实现了程序的建模可视化和计算结果可视化。应用此程序计算了氦冷陶瓷包层,数值结果表明NECP-FISH对中子通量密度、氚增殖比和中子释热的计算结果与蒙特卡罗程序NECP-MCX吻合良好。氚增殖比相对偏差为0.56%,所有区域的中子释热偏差均在6%以内。     

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.     

Improvement of Neutron Source Introduction Method for Absolute Measurements of Low Reactor Power

The neutron source introduction method was applied to absolute measurements of low reactor power at the Static Experiment Critical Facility STACY. To obtain the effective neutron source intensity more accurately, which is a key parameter for the source introduction method, the neutron source is newly defined as fission neutrons from the first fission reaction caused by neutrons emitted from the external neutron source. To obtain the newly defined effective neutron source intensity, the probability that a neutron from the external neutron source causes a fission reaction is calculated using the Monte Carlo code MCNP. This calculation took into consideration the three-dimensional complicated core structures. Furthermore, the fission reaction distribution, fundamental mode forward and adjoint flux distribution in a critical state were calculated using the three-dimensional transport code THREEDANT. Following the principle of the neutron source introduction method, an external neutron source was inserted near the STACY core tank and the reactor power was measured. The reactor powers by the neutron source introduction method were in good agreement with the ones from the analyses of the FP activity generated by high power operation.     

Three-dimensional neutronic calculations for a fusion breeder APEX reactor using some libraries

The effects of evaluated nuclear data files on neutronics characteristics of a fusion–fission hybrid reactor have been analyzed; three-dimensional calculations have been made using the MCNP4C Monte Carlo Code for ENDF/B-VII T = 300 K, JEFF-3.0 T = 300 K, and CENDL-2 T = 300 K evaluated nuclear data files. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, fissile fuel breeding and nuclear heating in a first wall, blanket and shield have been investigated for the mixture components of 90% Flibe (Li₂BeF₄) and 10% UF₄ for a blanket layer thickness of 50 cm. The contributions of each isotope of Flibe (⁶Li, ⁷Li, ¹⁹F, ⁹Be) and UF₄ (²³⁵U, ²³⁸U) to the integrated parameter values were calculated. The neutron wall load is assumed to be 10 MW/m².     

Neutronic analysis of ITER radial x-ray camera

The radial x-ray camera(RXC) is designed to measure the poloidal profile of plasma x-ray emission with high spatial and temporal resolution. The RXC diagnostic system consists of internal camera module and external camera module that view the core region and outer region through the vertical slots of the diagnostic first wall and diagnostics shield module of the equatorial port plug. To ensure the normal performance of the silicon photodiode array detectors of the cameras in the hard neutron irradiation environment in ITER tokamak, it is necessary to calculate neutron flux, radiation damage and the nuclear heating of the silicon photodiode array detectors and simulate the radiation maps of the range of RXC. This work estimated the nuclear environment of RXC based on Monte Carlo N-particle transport code, plasma scenarios of ITER tokamak and the RXC-integrated ITER CLITE model. The neutron flux of silicon photodiode array detectors and the lifetime of the silicon photodiode detector in the camera were calculated. The neutronic analysis results show that the shielding design has achieved the effect as expected and is able to guarantee the normal work of the detector during the ITER deuterium–deuterium phase without replacement, three detectors of the external camera can be operated during the whole deuterium–tritium phase without replacement.     

On the proliferation issues of a fusion fission fuel factory using a molten salt

The fusion fission fuel factory (FFFF) is a hybrid fusion fission reactor using a neutron source, which is in this case taken similar to the source of the Power Plant Conceptual Study - Water Cooled Lithium Lead (PPCS-A) design, for fissile material production instead of tritium self-sufficiency. As breeding blanket the first wall of the ITER design is attached to a molten salt zone, in which ThF₄ and UF₄ solute salts are transported by a LiF-BeF₂ solvent salt. For this blanket design, the fissile material is assessed in quantity and quality for both the Th-U and the U-Pu fuel cycle.The transport of the initial D-T fusion neutrons and the reaction rates in this breeding blanket are simulated with the Monte Carlo code MCNP4c2. The isotopic evolution of the actinides is calculated with the burn-up code ORIGEN-S.For the Th-U cycle the bred material output remains below 10 g/h with a ²³²U impurity level of 30 ppm, while for the U-Pu cycle supergrade material is produced at a rate up to 100 g/h.     

Preliminary Experiments on Accelerator-Driven Subcritical Reactor with Pulsed Neutron Generator in Kyoto University Critical Assembly

A series of preliminary experiments on an accelerator-driven subcritical reactor (ADSR) with 14 MeV neutrons were conducted at Kyoto University Critical Assembly (KUCA) with the prospect of establishing a new neutron source for research. A critical assembly of a solid-moderated and -reflected core was combined with a Cockcroft-Walton-type accelerator. A neutron shield and a beam duct were installed in the reflector region for directing as large a number as possible of the high-energy 14MeV neutrons generated by deuteron-tritium (D-T) reactions to the fuel region, since the tritium target is located outside the core. And then, neutrons (14MeV) were injected into a subcritical system through a polyethylene reflector. The objectives of this paper are to investigate the neutron design accuracy of the ADSR with 14MeV neutrons and to examine experimentally the neutronic properties of the ADSR with 14MeV neutrons at KUCA. The reaction rate distribution and the neutron spectrum were measured by the foil activation method for investigating the neutronic properties of the ADSR with 14 MeV neutrons. The eigenvalue and fixed-source calculations were executed using a continuous-energy Monte Carlo calculation code MCNP-4C3 with ENDF/B-VI.2 for the subcriticality and the reaction rate distribution, respectively; the unfolding calculation was done using the SAND-II code coupled with JENDL Activation Cross Section File 96 for the neutron spectrum. The values of the calculated subcriticality and the reaction rate distribution were in good agreement with those of the experiments. The results of the experiments and the calculations demonstrated that the installation of the neutron shield and the beam duct was experimentally valid and that the MCNP-4C3 calculations were accurately carried out for analyzing the neutronic properties of the ADSR with 14MeV neutrons at KUCA.     

Monte Carlo calculations of the nuclear effects of certain fluids in a hybrid reactor

This study analyzes the effects of certain heavy-metal-salt fluids on nuclear parameters in a fusion–fission hybrid reactor. Calculated parameters include the tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate, and fissile fuel breeding in the reactor's liquid first wall, blanket, and shield zones; gas production rates in the structural material of the reactor were calculated, as well. The fluid mixtures consisted of 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% UO₂, 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% NpO₂, and 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% UCO. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion–fission hybrid reactor system. A 3 cm wide beryllium (Be) zone was used for neutron multiplier between the liquid first wall and the blanket. The structural material used was 9Cr2WVTa ferritic steel, measuring 4 cm in width. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and the ENDF/B-VII.0 nuclear data library.     

Monte Carlo uncertainty quantification of the effective delayed neutron fraction

The applicability of Monte Carlo techniques, namely the Monte Carlo sensitivity method and the random-sampling method, for uncertainty quantification of the effective delayed neutron fraction β_eff is investigated using the continuous-energy Monte Carlo transport code, MCNP, from the perspective of statistical convergence issues. This study focuses on the nuclear data as one of the major sources of β_eff uncertainty. For validation of the calculated β_eff, a critical configuration of the VENUS-F zero-power reactor was used. It is demonstrated that Chiba's modified k-ratio method is superior to Bretscher's prompt k-ratio method in terms of reducing the statistical uncertainty in calculating not only β_eff but also its sensitivities and the uncertainty due to nuclear data. From this result and a comparison of uncertainties obtained by the Monte Carlo sensitivity method and the random-sampling method, it is shown that the Monte Carlo sensitivity method using Chiba's modified k-ratio method is the most practical for uncertainty quantification of β_eff. Finally, total β_eff uncertainty due to nuclear data for the VENUS-F critical configuration is determined to be approximately 2.7% with JENDL-4.0u, which is dominated by the delayed neutron yield of ²³⁵U.