聚变-裂变混合能源堆一维计算模型燃耗分析

聚变-裂变混合能源堆包括聚变中子源和以天然铀为燃料、水为冷却剂的次临界包层,主要目标是生产电力。利用输运燃耗耦合程序系统MCORGS计算了混合能源堆一维模型的燃耗,给出了中子有效增殖因数k_eff、能量放大倍数M、氚增殖比TBR等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。本文给出的结果可作为混合堆中子输运、燃耗分析程序校验的参考数据,为混合堆概念研究提供了基础数据。     

Infinite multiplication factor of low-enriched UO2–concrete system

The possibility of criticality of fuel debris in a form of uranium dioxide (UO₂)–concrete mixture is evaluated by calculating the infinite multiplication factor (k _∞) for a study of criticality control on the fuel debris generated through the molten core concrete interaction in a severe accident of a light water reactor. The infinite multiplication factor can be greater than unity, which means that handling of the mixture is subject to criticality control. This paper shows that concrete provides efficient neutron moderation and points out the necessity of further investigations on the criticality of UO₂–concrete system for actual handling of fuel debris.     

Preliminary design of hybrid energy reactor and integral neutron experiments

We propose a preliminary design for a fusion–fission hybrid energy reactor (FFHER), based on current fusion science and technology (with some extrapolations forward from ITER) and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. We study the interaction between the fusion neutron and the uranium fuel with the aim of to achieving greater energy multiplication and tritium sustainability. We also review other concept hybrid reactor designs. We design integral neutron experiments in order to verify the credibility of our proposed physical design. The combination of this program of research with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology foundation for the future development of the FFHER concept in China.     

The burn-up of natural uranium when it is moved axially in a reactor

We investigate the effect produced in the distribution of the neutron flux and the bum-up of the natural uranium in a heavy-water reactor when the fuel is moved contirmously in an axial direction. Using a simplified model and taking into consideration the variations of the neutron multiplication factor as a function of uranium bum-up, we show that when the uranium is moved in one direction, the bum-up is about 20% higher than when the fuel elements are continuously recharged [1] and that the heat-transfer conditions of fuel-element operation are improved. When the fuel is moved simultaneously in two opposite directions, the uranium bum-up is increased by 40%, and the conditions of heat removal from the fuel dements remain almost unchanged.Translated from Atomnaya Énergiya, Vol. 17, No. 5, pp. 380–384, November, 1964     

Application of CANDLE burnup to block-type high temperature gas cooled reactor

The CANDLE burnup strategy, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes, is applied to the block-type high temperature gas cooled reactor. If it is successful, a burnup control rod can be eliminated, and several merits are expected. This burnup may be realized by enriched uranium and burnable poison with large neutron absorption cross-section. With the fuel enrichment of 15%, gadolinium concentration of 3.0%, and fuel cell pitch of 6.6 cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half-width of power density distribution of 1.5 m. When the concentration of natural gadolinium is higher, the burning region moving speed becomes slower and the burnup becomes higher, though the effective neutron multiplication factor becomes smaller. When U-235 enrichment is higher, the effective neutron multiplication factor becomes larger, the speed becomes slower, and the burnup becomes higher. When the pitch is wider, the effective neutron multiplication factor becomes larger, the speed becomes faster, and the burnup becomes higher.     

MOX燃料贮存水池核临界安全分析

利用MONK程序对MOX热室项目燃料贮存水池进行了核临界安全分析。针对给定的水池尺寸和燃料棒数量,确定燃料以分区方式贮存。选取国际公布的临界基准实验数据,验证并确定MONK程序计算分析类似物料形态时的偏倚和次临界限值,其次进行保守假设,确定贮存水池在正常及事故工况下其中子有效增殖因数,评价贮存水池的安全性。计算结果表明,贮存水池在最危险事故工况下,其最大中子增殖因数小于次临界限值,系统处于临界安全状态。     

Accuracy of nuclear design of fast and thermal neutron coupled core by SRAC

Supercritical-pressure light water cooled fast reactor adopts the blanket fuel assemblies with depleted uranium fuel and zirconium hydride layer in the core for negative coolant void reactivity. Thermal neutrons are generated in the core of fast reactor. It is called “fast and thermal neutron coupled core”. The purpose of the present study is to examine the accuracy of assembly and core calculations including preparation of the macroscopic cross sections with the SRAC code system for “fast and thermal neutron coupled core” in comparison with the Monte Carlo codes, MVP and MVP-BURN. Accuracy of the neutron multiplication factor and coolant void reactivity calculation has been evaluated in four types of cores of different fractions of the blanket fuel assembly with zirconium hydride rods. The conventional analysis is based on the macroscopic cross sections which are prepared with infinite lattice. The conventional SRAC calculation underestimates the neuron multiplication factor for all types of cores. Other findings are that the conventional SRAC calculation overestimates coolant void reactivity for the cores without zirconium hydride rods, and underestimates coolant void reactivity for the core of all blanket fuel assemblies having zirconium hydride rods. To overcome these problems, it has been proposed that the macroscopic cross sections of seed fuel assembly are prepared with the model that a seed fuel assembly is surrounded by blanket fuel assemblies in order to take into account the effects of the surrounding fuel assemblies. Evaluations show that accuracy of the neutron multiplication factor by the SRAC calculation can be improved by the proposed method.     

Validation of the nuclear design code system for the HTTR using the criticality assembly VHTRC

The high temperature engineering test reactor is the first block-type HTGR designed for a 950 °C outlet gas temperature which uses low-enriched uranium fuel with burnable poison rod. For validation of the nuclear design code system for the HTTR, a critical assembly of VHTRC had been constructed. The calculation uncertainties of effective multiplication factor, neutron flux distribution, burnable poison reactivity worth, and control rod worth, temperature coefficients were evaluated. Calculation accuracy of a Monte Carlo code is also evaluated.     

钍基柱状高温气冷堆不同启动燃料特性初步分析

钍是一种可转换材料,将其转换成²³³U能极大提高现有核燃料资源的储量。为实现对钍的合理利用,以模块式柱状高温气冷堆GT-MHR的燃料组件作为研究对象,选取低浓缩铀、武器级钚、核反应堆级钚等作为其启动燃料。利用栅格输运计算程序DRAGON对这3种启动燃料下的钍基柱状燃料组件的寿期初中子能谱、无限增殖系数、燃耗、转换比以及²³³U和²³²Th的含量等参数进行了分析。结果表明,在易裂变物质初装量约为9%时,与低浓缩铀和武器级钚相比,核反应堆级钚作为启动燃料时组件寿期初中子能谱较硬、转换比较高;其燃耗达90 GW•d/tHM;其无限增殖系数在寿期内的波动最小;燃耗为75 GW•d/tHM时组件中²³³U存余量与²³²Th消耗量之比达0.566。     

核临界安全中的监督现场测量技术--源倍增方法的某些问题

给出了核临界安全中监督现场的测量技术——源倍增法的实验理论和实验方法。源倍增法实际测量的是有源次临界中子有效增殖系数k2而不是中子有效增殖系数Keff。在铀溶液核临界装置上进行了实验研究用源倍增法测量了次临界系统在外中子源作用下铀溶液不同液位的有源次临界中子有效增殖系数k2;用周期法测量了单位铀溶液位的反应性系数,然后用临界液位与次临界液位之差乘以单位铀溶液位的反应性系数,给出系统次临界液位时的反应性．由反应性给出传统观念上的中子有效增殖系数keff 。讨论了它们的差别及对核临界安全的影响。     

The feasibility study on perfect burning reactor system

The system of 100% natural uranium burning with once-through fuel cycle is defined as the Perfect Burning Reactor System (PBRS). This kind of nuclear system can be expected to have some good characteristic such as resource efficiency, radiotoxicity reduction, proliferation and nuclear safety. Therefore, the feasibility of the concept is studied in this paper. The preliminary results show that the system of 100% natural uranium burning with once-through fuel cycle is physically possible with a plenty supply of external neutron, and that the system demands no activities concerning with fuel cycle such as uranium enrichment, fuel fabrication, spent fuel reprocessing and radioactive waste treatment. The study also quantitatively clarifies the external neutron source strength, the nuclear criticality safety, the demanded accelerator performance and the energy balance. In addition, the more precise analysis is requested for well understanding and improving the characteristic and economical rationality of the system.     

小型中子源14MeV中子慢化体结构的研究（Ⅱ）

对天然铀和聚乙烯的各种组合进行了计算，结果可见：用２—２．５ｃｍ厚的天然铀紧包１４ＭｅＶ中子源，在离源面２—３ｃｍ范围内使用热中子为最佳，用浓缩铀作增强物，只有当离源一定距离放置时，才能比天然铀有较好的效果。讨论了在由高浓缩铀和聚乙烯组成的、几何尺寸与上面计算相同的次临界系统中使用天然铀或铅增强物对１４ＭｅＶ中子源的倍增情况，得出天然铀和铅在该次临界系统中仍有明显的增强作用；在合适的结构中，设ｋ为该次临界系统中的有效增殖因数，则可使增强因子Ｍ接近甚至大于（１－ｋ）－１。     

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.     

次临界能源堆物理性能初步分析

次临界能源堆（SER）是由托卡马克聚变源驱动的聚变裂变混合堆。SER以天然铀为燃料、水为冷却剂,主要目标是生产电能。本工作建立了次临界能源堆环形圆柱模型,利用蒙特卡罗输运和燃耗计算程序,比较了燃料区不同构型对k_eff、M、TBR和燃料增殖比等参数的影响,针对均匀模型进行中子源效率与聚变源强、功率分布与能谱、初步燃耗、寿期末停堆衰变热和卸载燃料放射性等物理性能分析。计算结果表明,该模型能满足能量倍增大于6、氚自持、较长时间不换料等设计目标。研究结果为下一步开展SER安全分析提供了基础。     

Conditions for criticality by uranium deposition in water-saturated geological formations

This study focuses on neutronic analysis to examine the criticality conditions for uranium depositions in geological formations resulting from geological disposal of damaged fuels from Fukushima Daiichi reactors. MCNP models are used to evaluate the neutron multiplication factor (k_eff) and critical mass for various combinations of host rock and geometries. It has been observed that the k_eff for the deposition become greater with (1) smaller concentrations of neutron-absorbing materials in the host rock, (2) larger porosity of the host rock, (3) heterogeneous geometry of the deposition, and (4) greater mass of uranium in the deposition. This study has revealed that the planar fracture geometry applied in the previous criticality safety assessment for geological disposal would not necessarily yield conservative results against the homogeneous uranium deposition.     

Neutron Behavior in Cluster-Type Fuel Lattices, (II)

A multigroup method of calculation is presented for describing the neutron behavior in a cluster-type fuel lattice. It solves the integral transport equation by a semi-analytical method proposed in a previous paper for calculating collision probabilities in the lattice of a clustered fuel element. Using only fundamental nuclear data, it gives space and energy dependent neutron flux. The method has been programmed for HITAC-5020F (computer code named CLUSTER-III).

The accuracy of the method has been tested by comparing the calculation with the experiment described in Part (I) of this paper. The lattices are 28-pin clusters of UO₂ or PuO²+UO² fuel pins, with heavy-or light-water moderators and with light-water coolant containing varying void ratios. The quantities studied are micro-parameters, reaction distributions in energy and space, thermal disadvantage factors and the multiplication factors. It is found that the calculated results are generally in good agreement with experiment, typically within 10% for micro-parameters and thermal disadvantage factor, and within 1% for the multiplication factor.     

Plant designing of ion exchange chemical uranium enrichment and its non-proliferation aspects

The chemical uranium enrichment using ion exchanger had been developed mainly in Japan, but now only the academic studies have been continued at Tokyo Tech. For the purpose of revaluation as the alternative production process of the low enriched uranium for nuclear power plants, the plant design and the non-proliferation aspects of redox ion exchange chemical uranium enrichment (Redox Ion Exchange Method) were studied.The conceptual design of a commercial enrichment plant with a scale of 1400 tSWU/Y was newly performed based on data reported. It is composed of seven enrichment units with two enrichment columns with an inner diameter of 6.5 m and a height of 11 m and redox systems. It is evaluated as having the site area of 97,200 m² including the vacancy to double the production with subsidiary sections of pre-treatment and post-treatment of uranium or the like, a construction cost is 218 billion Japanese yen(JPY), and a cost per enrichment work is 18,000 JPY/kgSWU. As for the nuclear proliferation resistance of this process, it is suggested that the production of highly enriched uranium for weapon use is difficult from the view point of nuclear fission criticality and the lengthy equilibrium time mentioned as following. It is difficult to operate the enrichment plant for the production of highly enriched uranium of 50 wt% of ²³⁵U or higher, because the effective neutron multiplication factor (k_eff) in the enrichment columns becomes 1 or above depending on the enrichment conditions. And the required time for the production of highly enriched uranium of 90 wt% of ²³⁵U is estimated 10 thousands days or more, this means substantially impossible to get highly enriched uranium under the safeguard of IAEA by the Redox Ion Exchange Method.The Redox Ion Exchange Method is suggested to be a promising candidate for an alternative enrichment process to obtain low enriched uranium in the range 3-5 wt% ²³⁵U for nuclear power plant in the world. Furthermore, for the future prospective fuel cycle without UF₆, new front-end system coupled with uranium from seawater and chemical enrichment is proposed.     

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.     

Conversion of high enriched uranium in thorium-232-based oxide fuel for light and heavy water reactors: MOX-T fuel

This paper discusses the possibility of using military high enriched uranium and plutonium in thorium oxide fuel for light and heavy water reactors (LWRs and HWRs). It is shown that such a fuel has several important advantages: (i) ²³⁹Pu and other long-living actinides are generated in quantities which are at least 100 times less than in conventional fuel; (ii) neutron emission is lower by a factor of more than 100; (iii) ²³³U is generated and burnt (the conversion factor for LWRs is 0.64–0.68 and for HWRs about 0.88); (iv) thorium is utilized and the total available amount of nuclear fuel is increased. The problem of non-proliferation of fissile material is also discussed and it is shown that the supervision of such fuel does not differ essentially from the supervision of low enriched uranium fuel with plutonium generation.