核燃料溶液系统瞬态特性分析研究

在核反应堆乏燃料后处理主工艺流程中,核燃料通常以溶液状态存在,可能发生核临界事故。研究核临界事故的产生机理和事故源项,对预防事故发生、缓解事故后果、事故应急响应与医学诊治均具有十分重要的意义。本文采用点堆动力学方程结合二维热传导方程,开发了基于圆柱形溶液贮存容器的瞬态特性分析程序GETAC,利用该程序计算了法国SILENE瞬态实验装置模拟临界事故功率随时间的变化,得到了功率振荡在不同反应性引入大小、方式和有无外中子源等情况下的变化规律,计算分析结果与瞬态实验测量数据以及国外其他程序计算结果较一致。     

先进三代核电机组反应堆达临界方式及特性分析

介绍了先进三代核电机组如何在低中子注量率的情况下通过堆外核测量系统源量程探测器监视反应堆达临界,并对其达临界过程中探测器的计数率变化进行比照、分析。通过分析发现,在低中子注量率情况下,利用反应堆启动率(或周期)的变化能够实现对反应堆临界实现与否的判断。同时,利用相对中子源不同位置的探测器计数率的变化规律,能够监测反应堆逼近临界的程度。这一反应堆达临界方式可以在诸如无源启动等低中子注量率情况下得到应用。     

Submersion criticality safety of fast spectrum space reactors: Potential spectral shift absorbers

Compact, fast spectrum, nuclear reactors are being considered to support NASA's future space exploration sometime in the next decade. In order to secure launch approval, these reactors should remain sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. In such an accident, the neutron spectrum in the reactor is thermalized, typically increasing reactivity, and potentially making the reactor supercritical. Incorporating “Spectral Shift Absorbers” (or SSAs), which have significantly higher absorption cross-sections for thermal versus fast neutrons, could offset the reactivity increase. It has always been the assertion that the worst-case submersion accident involves a fully flooded reactor; however, this work shows that, depending on the type and amount of SSA in the reactor, a submerged but unflooded reactor could be more reactive. A screening of the existing nuclear database for potential SSAs yielded 28 elements and nuclides, which are examined in detail as additives to a representative homogenous space reactor core by varying the SSA-to-U²³⁵ atom ratio. The effect of placing a thin coating of different SSA materials on the outside surface of the reactor core is also investigated. Nine SSAs (boron-10, cadmium, cadmium-113, samarium-149, europium-151, gadolinium, gadolinium-155, gadolinium-157, and iridium) are recommended for further consideration in actual space reactor designs.     

Adapting the deep burn in-core fuel management strategy for the gas turbine – modular helium reactor to a uranium–thorium fuel

In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine – modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium–thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: ²³⁵U, which represents the 20% of the fresh uranium, ²³³U, which is produced by the transmutation of fertile ²³²Th, and ²³⁹Pu, which is produced by the transmutation of fertile ²³⁸U. In order to compensate the depletion of ²³⁵U with the breeding of ²³³U and ²³⁹Pu, the quantity of fertile nuclides must be much larger than that one of ²³⁵U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of ²³⁵U. At the same time, the amount of ²³⁵U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k_eff and mass evolution, reaction rates, neutron flux and spectrum at the equilibrium of the fuel composition, highlights the features of a deep burn in-core fuel management strategy for a uranium–thorium fuel.     

Critical neutron heating in the control drums of a dual purpose thermionic space reactor for power and propulsion

The critical neutron heating in the reflector control drums is investigated for a fast incore thermionic space craft reactor for power and nuclear propulsion. The reactor is fueled with uranium carbide (UC) and controlled with the help of rotating B₄C drums imbedded into the beryllium reflector. While the neutron heating in the drums would not require a cooling mechanism in the power phase, the heat generation during the thrust phase obliges cooling for a nuclear thermal thrust around F = 5000 N by a specific impulse of 670 s⁻¹ at an hydrogen exit temperature around 1900°K. With a beryllium reflector without extra cooling measures, thermal thrust must be kept F < 2500 N to relieve the thermal load in the reflector. On the other hand, a reflector made of BeO may withstand a thermal load for a nuclear thermal thrust of F = 5000 N. The neutronic analysis has been conducted in S₁₆-P₃ and S₈-P₃ approximation with the help of one- and two-dimensional neutron transport codes ANISN and DORT, respectively. A reactor control with boronated reflector drums (drum diameter = 14 cm) at the outer periphery of the radial reflector of 16 cm thickness would make possible reactivity changes of Δk_{eff = 13.55%}—amply sufficient for a fast reactor—without a significant distortion of the fission power profile during all phases of the space mission. Calculations are conducted for a reactor with a core radius of 22 cm and core height of 35 cm leading to power levels around 50 kW_el.     

Characteristics of Neutron Guide Tubes in Upgraded JRR-3

Abstract

The JRR-3 has been upgraded to be a new high performance research reactor JRR-3M with neutron guide tubes on a large scale and a cold neutron source. The neutron fluxes and spectra were measured at the end of the two thermal and three cold neutron guide tubes. The gain of the cold neutron source is also found from these spectra. The neutron fluxes of thermal neutron guide tubes with characteristic wavelength 2 Å are 1.2x10⁸ n/cm².s at a reactor power of 20 MW. The neutron fluxes of cold guide tubes are 2.0x 10⁸ n/cm².s with characteristic wavelength 4 Å and 1.4x10⁸ n/cm².s with 6 A when the cold neutron source is operated. The neutron spectra measured by the time-of-flight method agree well with their designed ones. The gains of the cold neutron source are 8 for 4 Å and 20 for 6 Å at a reactor power of 20 MW.     

星球表面用核反应堆临界安全分析

核反应堆电源具有寿命长、可全天候工作等特点,可作为星球表面及其他深空探测任务的电源。针对星球表面用核反应堆电源在发射过程中重返地面的临界安全问题,提出了星球表面用核反应堆的临界安全分析要求、分析假设与模型,并对反应堆临界安全特性及采取的临界安全措施进行了计算分析。计算结果表明,不同假设掉落环境下的星球表面用核反应堆的有效增殖因数均小于0.98,满足临界安全要求。反应堆通过采用Mo-14%Re合金结构材料、设置相对较厚的堆芯反射层以及在反射层包壳和堆芯外围涂覆Gd₂O₃涂层等措施有利于确保反应堆在事故时处于次临界状态。     

SiO2热中子散射截面在空间堆事故分析中的应用

SiO₂热化效应可能对核废料地质储存库分析和空间反应堆坠落湿沙情况下的临界安全造成一定影响。本文结合最新的ENDF/B Ⅶ-1的TSL库,制作了ACE格式的SiO₂热中子截面数据库。分析了不同温度对SiO₂热中子散射截面的影响,比较了采用声子谱模型的SiO₂热中子散射截面数据与采用自由气体模型的SiO₂热中子散射截面数据的差异,并采用本文制作的截面库,对空间堆坠落湿沙情况下的临界安全特性进行了分析,给出了反应堆最易重返临界的湿沙成分比例。     

Optimization of the steady neutron source technique for absorption cross section measurement by using an 124Sb–Be neutron source

An improved experimental approach has been developed to determine thermal neutron absorption cross sections. It uses an ¹²⁴Sb–Be neutron source which has an average neutron energy of only about 12 keV. It can be moderated in either a water tank or a paraffin filled box and can be used for aqueous or powder samples. This new design is first optimized by MCNP simulation and then benchmarked and calibrated with experiments to verify the simulations and realize the predicted improved measurement sensitivity and reproducibility. The ¹²⁴Sb–Be source device is from 1.35 to 1.71 times more sensitive than the previous method based on the use of a ²⁵²Cf source.     

MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination

The IPR-R1 TRIGA is a research nuclear reactor managed and located at the Nuclear Technology Development Center (CDTN) a research institute of the Brazilian Nuclear Energy Commission (CNEN). It is mainly used to radioisotopes production, scientific experiments, training of nuclear engineers for research and nuclear power plant reactor operation, experiments with materials and minerals and neutron activation analysis. In this work, criticality calculation and reactivity changes are presented and discussed using two modelings of the IPR-R1 TRIGA in the MCNP5 code. The first model (Model 1) analyzes the criticality over the reactor. On the other hand, the second model (Model 2) includes the possibility of radial and axial neutron flux evaluation with different operation conditions. The calculated results are compared with experimental data in different situations. For the two models, the standard deviation and relative error presented values of around 4.9 × 10^?4. Both models present good agreement with respect to the experimental data. The goal is to validate the models that could be used to determine the neutron flux profiles to optimize the irradiation conditions, as well as to study reactivity insertion experiments and also to evaluate the fuel composition.     

A function to provide neutron spectrum produced from the 9Be + p reaction with protons of energy below 12 MeV

A function to give the total neutron production cross section, angular distribution, and energy spectrum via the ⁹Be + p reaction has been created by fitting experimental data to characterize compact neutron sources with thick Be targets bombarded by protons with energy below 12 MeV. To examine the suitability of the function, calculations of the angle-dependent neutron energy spectra produced in thick Be targets with 4- and 12-MeV protons using the function were compared with corresponding experiments and calculations using the nuclear data libraries of ENDF/B-VII.0 and JENDL4.0/HE. The function was in better agreement with the experiments than the calculations using the libraries except for at backward angles. The ¹¹⁵In(n,n’)^115mIn reaction rates calculated using GEANT4 with source neutrons given by both the function and ENDF/B-VII.0 were compared with that measured at the RIKEN Accelerator-Driven Compact Neutron Source to evaluate the neutron spectrum above 1 MeV. The function slightly overestimated the measurement by 14% and the calculation with ENDF/B-VII.0 underestimated by 35%. It was concluded that the function can be applied in compact neutron source designs.     

Monte Carlo simulation of a conceptual thermal column in the Syrian miniature neutron source reactor using MCNP-4C

Based on probabilistic approach, the MCNP-4C code has been used effectively to simulate the Syrian MNSR reactor core and all its surrounding components in three dimensions, including a preliminary conceptual design of a thermal column to be installed later. For verification and validation purposes, reactor calculations include: criticality and control rod worth. Values of these parameters are 1.00517 and 6.54 mk, respectively. The thermal column is to be installed in the water of the reactor pool. Optimal conditions for this thermal column were tested using the already developed model. Optimization focused on the most suitable position for placement of the column in the water pool, dimensions, and material. The aim was to have a thermal neutron flux of 1 × 10⁹ n cm⁻² s⁻¹ in the center of thermal column, and resonant and fast neutron fluxes to be as low as possible as well.     

Conceptual design study of fusion DEMO plant at SWIP

The basic definition and development strategy of the DEMO plant based on the Chinese fusion power plant (FPP) program are presented briefly. A conceptual design study of fusion HCSB-DEMO reactor with a fusion power of 2550 MW and a neutron wall loading of 2.3 MW/m² is performed recently. Three sets parameters of core plasma for different DEMO design objectives are proposed. A helium-cooled blanket system with ceramic breeder (Li₄SiO₄), the structure material of low-activation ferritic steel (LAF/M) and Be neutron multiplier based on Chinese ITER HCSB-TBM design foundation are considered. The design parameters, preliminary analyses and the basic structure as well as development strategy of HCSB-DEMO reactor are introduced.     

New approach to evaluate lattice expansion of light water reactor fuel elements on criticality safety of transport packages under impact accidents

Abstract

In order to safely transport packages containing light water reactor fuel assemblies, it is essential to maintain the fuel assemblies in a subcritical state in accidents during transport. To evaluate nuclear criticality safety, an estimator is required to determine an absolutely safe level based not only on hypothetical accidents but also on practical accident levels which, to some extent, are based on actual accidents. The purpose of the present study is to suggest the arrangement of the deformation range of the fuel assembly after an actual accident, and to obtain the maximum value of the neutron effective multiplication factor based on the criticality safety assessment for the transport cask. In the present study, two kinds of criticality calculations for the package were considered: large scale pin pitch shift and small scale pin pitch shift. For the large scale pin pitch shift, a parameter which determines the location of each fuel pin which constitutes the fuel assembly was introduced so that the criticality calculation for the fuel assembly with non-uniform lattice pitch can be performed parametrically. The result of the criticality calculation using the parameter made it clear that the fuel pin pitch is sensitive to the neutron reactivity because each of the fuel pin pitches is related to a ratio of the fissile to the moderator, and that the relationship of the ratio to the neutron reactivity depends on the type of the fuel assembly involved, i.e. the type of a nuclear reactor in which a fuel assembly is used. For the small scale pin pitch shift, the study focused on the small displacement of each fuel pin. The small displacement of each fuel pin pitch can be described probabilistically using the stochastic geometry routine in MCNP code. Using the scheme in combination with the scheme for the large scale pin pitch shift, the maximum value of the neutron effective multiplication factor of the package after an accident can be obtained. This scheme is useful to determine the maximum neutron effective multiplication factor for the criticality safety evaluation.     

棱柱式高温气冷空间核反应堆初步方案设计与中子物理分析

空间核反应堆电源将核裂变能转换为电能,与太阳能、化学燃料电池等其他形式的电源相比,具有电功率大、系统比功率高、使用寿命长等优点,在太空探索中具有广阔的应用前景。以高温气冷堆技术为基础,提出了以氦氙混合气体作冷却剂,直接布雷顿循环的空间核反应堆电源方案。核反应堆是采用包覆颗粒燃料的小型棱柱式高温气冷堆,热功率为5 MW。采用蒙特卡罗方法进行了中子物理分析。结果表明,设计的反应堆满足10a以上的满功率运行寿期,具有负的反应性温度系数。通过布置B4C安全棒,使反应堆在发射失败引起的堆芯进水事故中能保证次临界。     

Pre-equilibrium Emission Spectra Calculations of <Superscript>232</Superscript>Th(n,xn) Using New Evaluated Method

The knowledge of the thorium (²³²Th) cycle potentialities are required for the design of a fusion-fission (hybrid) reactor. Pre-equilibrium nuclear reactions have been used to investigate the effect of initial exciton numbers on the nucleon emission spectra. In this study, the initial exciton numbers for the target nucleus of ²³²Th were calculated through a method of offered by Tel et al. and then were used to obtain the effect cross section of the neutron emission spectra. Using this new method, a different way from the literature, the initial exciton numbers calculated with the theoretical neutron and proton densities have been obtained with SKM* on the ²³²Th(n,xn) reaction at 14.1 and 18.0 MeV incident neutron energies. The results were analyzed by comparing the empirical results in the literature.     

Neutron flux distribution in the irradiation channels of Am–Be neutron source irradiation facility

Monte Carlo (MCNP-5) simulations of the neutron fluxes were performed to determine the radial and axial neutron fluxes of the two irradiation sites of the 20 Ci ²⁴¹Am–Be neutron irradiation facility at NNRI. The geometry of the ²⁴¹Am–Be source as well as the irradiator design, constituted one cylindrical neutron source at the center of a cylindrical barrel with water as moderator. In the far and the near irradiation sites that were 13.1 cm and 4.2 cm, respectively, from the source, the average thermal, epithermal and fast neutron fluxes axially increase exponentially from the bottom and peak at the center of the source 3.0 cm from the bottom of the source and decrease to a very low value at the end of the tube. The percentage of the average thermal flux increases as the distance from the source increases, while the percentages of the epithermal and fast fluxes decrease as the distance from source increases. In the far and near irradiation sites the average radial thermal neutron flux decreases at the rates of 307.02 n cm⁻² s⁻¹ and 961.54 n cm⁻² s⁻¹ per cm along the diameter, respectively. The average radial, epithermal and fast neutron fluxes were fairly uniform along the diameter in the two irradiation sites.