不同类型核燃料对热管冷却反应堆燃耗性能的影响

热管冷却反应堆采用固态反应堆设计理念,具有功率密度高、结构紧凑、固有安全性高等特点,在深空探索、深海勘探、偏远地区等场景中具有广阔的应用前景。核燃料作为热管冷却反应堆的重要组成部分,不同类型核燃料在堆芯燃耗分析时会呈现不同的中子学性能。基于美国爱达荷国家实验室(INL)提出的热管冷却反应堆INL Design A,利用清华大学蒙特卡罗中子输运程序RMC (Reactor Monte Carlo code)建立堆芯物理模型,选取UO₂,(U_0.9Pu_0.1)O₂,U-10Zr,U-8Pu-10Zr,UN,UC这6种核燃料开展燃耗计算,分析了不同核燃料、不同功率水平对热管冷却反应堆堆芯燃耗性能的影响。计算结果表明:在堆芯燃耗深度相同情况下(20.8 GW·d·t?1),装载U-8Pu-10Zr燃料的堆芯所需235U富集度最低(9.8%),具有较好的U-Pu增殖性能。堆芯功率处于5 MW的热管冷却反应堆,燃料中241Pu的存在不仅没起到增大堆芯燃耗深度的作用,反而导致堆芯剩余反应性和堆芯寿期末次锕系核素(MAs)的产量增大,影响反应堆的安全性与经济性。因此,对于装载含有Pu燃料的小功率长寿期热管冷却反应堆,需重点关注241Pu对堆芯燃耗性能的影响。     

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.     

Pu或Am含量对铀基混合氧化物的结构和能量及力学性质的影响

U-Pu和U-Am混合氧化物中的Pu或Am含量对核反应堆燃料的高效循环利用至关重要.研究铀基混合氧化物中不同Pu或Am的含量对其结构、力学性质和能量的影响有助于理解和预测提高反应堆中燃料的行为以及与包层的化学或力学相互作用.本文通过DFT+U方法首先探索UO₂、PuO₂和AmO₂的结构和能量随U的变化关系,然后研究UO₂结构中不同Pu或Am含量对其结构和力学性质以及能量的影响.结果表明在UO₂结构中掺入不同Pu或Am的含量均使得体系晶格参数收缩,且与实验观测(U, Pu)O₂中Pu的含量结论是一致的.从能量角度观察,UO₂结构中掺入不同Pu或Am的含量使得体系形成能随掺入量的变化趋势明显不同.结果显示当UO₂结构中掺入Pu为25%时,U-Pu混合氧化物体系的形成能最低,而当UO₂结构中掺入Am为75%时,U-Am混合氧化物体系的形成能最低.此外,我们也探讨和分析了在UO₂     

铅冷快堆M2LFR-1000堆芯燃料管理方案设计

堆芯燃料管理是反应堆设计中极为重要而且复杂的工作,直接影响着堆芯的经济性。目前国内外对于压水堆等传统热堆已有了较为丰富和成熟的燃料管理计算方法,但对于快堆,由于其中子能谱硬,与传统热堆相比有着不同的控制方式和功率分布,快堆的堆芯燃料管理缺乏系统研究。针对中国科学技术大学自主研发的强迫循环冷却的铅基快堆M2LFR-1000,应用SRAC/COREBN软件包进行堆芯燃耗计算,根据燃耗深度提取核素核子密度,计算伪平衡循环参数进行燃料管理预估,然后进行首循环装料、过渡循环和平衡循环燃料管理方案设计。结果表明:对M2LFR-1000堆芯外区燃料换料组件Pu的富集度进行优化,可以延长换料周期到540 d,提高平均卸料燃耗深度;伪平衡循环结果与平衡循环基本一致,伪平衡循环可以用于燃料管理预估。     

不同燃耗计算模型对商用压水堆乏燃料组件核素成分的影响分析

燃耗计算精度对提高乏燃料贮存效率有着重要影响,在应用燃耗信用制时,燃耗计算得到的核素成分偏差决定了乏燃料贮存的临界安全裕量。不同燃耗计算模型所得到的核素成分偏差各不相同,为提高燃耗计算精度,提出了一种装载不同燃料富集度的多组件燃耗计算模型,并使用不同燃耗计算模型分别对TMI-1反应堆NJ07OG组件中的6个样本进行了计算、对比和分析。结果表明,相比其他模型,考虑不同燃料富集度的多组件模型得到的²³⁵U、²³⁸U和²³⁹Pu等核素平均相对偏差更接近于零且6个样本的相对偏差分布更为平均。     

Closed Thorium-Uranium-Plutonium Cycle Concept in Nuclear Power Engineering

The use of 232Th instead of 238U as a fertile isotope, 233U instead of 239Pu as the main fissile isotope, heavy water instead of light water as a coolant, and its dilution with light water in the VVER reactor campaign make possible self-enrichment of fuel with fissile isotopes, including the time upon achieving the balanced isotopic abundance ratio of actinides, and also provide conditions for closing the Th-U-Pu fuel cycle. This allows increasing the fuel lifetime by around two orders of magnitude, making it much easier to handle radioactive waste, reducing the nuclear hazard of PWE reactors, and providing a technological barrier to prevent the distribution of fissile materials and nuclear technologies.     

乏燃料中Pu同位素含量数值模拟

利用燃耗计算程序MCORGS模拟反应堆燃耗与乏燃料中Pu同位素含量之间的关系,通过对轴向上分为20段的重复栅元模型和组件模型进行的燃耗计算,得到压水堆中乏燃料中轴向不同位置燃耗的分布和Pu-239同位素含量的变化,模拟发现Pu-239同位素含量随着燃料棒在堆芯中的位置不同变化很大。同时,对VVER1000组件和压水堆1717组件也进行了燃耗计算,计算发现组件径向不同位置的燃耗有一定差别。轴向上和径向上不同位置的燃耗差别会导致同一批卸载的乏燃料中含有很多低燃耗的燃料区间,这种乏燃料给国际核不扩散带来了巨大的风险,应该加强监管。     

Determination of Actinide Traces in Reactor Water

In reactor water from primary coolant ²³⁹Np, ²³⁸U, ²³⁸Pu and ^239,240Pu was determined using isotope dilution and isotope dilution activation analysis. An example of crud and reactor water from a power station was also investigated by gamma spectroscopy and activation analysis. For localization and characterization methods of cladding failure the usefulness of analyti?al determination of actinides in reactor water is emphasized.     

Flux and spectrum of reactor antineutrinos

In order to perform reactor experiments aimed at studying the nature of the neutrino and measurements in the realms of geo- and astrophysical neutrinos and to meet practical requirements in this field, it is highly desirable to obtain deeper insight into the operation of nuclear reactors as a source of antineutrinos. The fluxes and spectra of neutrinos from a reactor in the on and off modes and from a reservoir intended for storing a spent reactor fuel and situated near the reactor being considered are calculated. Features that are peculiar to the flux and spectrum of reactor antineutrinos and which are of importance for implementing and interpreting experiments, but which were disregarded previously, are analyzed here.     

Gaseous Fuel Reactor Research

Gaseous Fuel Nuclear Reactors are externally moderated and contain the fissile material inside a cavity where it is suspended by fluid mechanics forces. The gaseous phase of the nuclear fuel permits operation of the reactor at temperatures much higher than the melting point of all materials. NASA has originally supported relevant research for space propulsion. The continuation of this work includes now research on power generation on Earth for improved economy and environmental acceptability. In reactor experiments with enriched uranium hexafluoride, UF6, a critical mass of 6 kg is determined. Pressurized UF6 remains chemically stable at temperatures up to 2000 kelvins. The interaction of fission fragments with their gaseous environment causes preferential excitation and ionization, leading to non-equilibrium optical radiation. Powerful fluxes of photons are expected to become a superior mechanism of energy extraction from the fissioning gas or plasma in the reactor. The pumping of lasers solely by fission fragments is realized in a variety of lasants. A near term objective of the NASA gaseous fuel reactor program is a benchmark experiment at 100 kw power and at a gas temperature of 1600 kelvins, demonstrating the feasibility of major advances in reactor technology. A concerted research effort is leading to this experiment. A plasma core cavity reactor for high specific impulse propulsion in space reminas a long range goal.     

铀-钍混合燃料反应堆的可行性分析

分析了以铀为燃料的核电系统的弊端、钍燃料反应堆的理论技术依据和世界范围内钍燃料反应堆的研究状况。提出在我国开发利用钍资源,建立铀．钍混合燃料反应堆具有的独特优势,建议应加大钍资源开发人力物力投入,改变我国核电利用水平落后和钍资源流失之现状。Nuclear energy is a preferred option for electric power generation. The disadvantages of the current uranium-dioxide （UO2 ） fuel in nuclear power were presented and the reactor using the mixed thorium dioxide and uranium dioxide fuel （ ThO2-UO2 ） in the near future was foretold. A proposal to strengthen the research cooperation on the use of the thorium mineral resources in china was put forward.     

加速器驱动快/热耦合次临界系统的概念设计

对加速器驱动快／热耦合次临界系统进行了概念设计研究。在该系统中,内区的快包层和外区的热包层是相互独立的,快、热包层之间为空腔和B4C包层以实现单向耦合。快包层装以合金（MA＋Pu）Zr为燃料,热包层初始循环装以氧化物（Th＋Pu）O₂为燃料,平衡循环装以（Th＋^233 U＋Pu）O2为燃料。^99Tc,^129I和^135Cs分别以单质、NaI和CsCl的形式装入热包层。该系统具有较高的能量放大倍数、嬗变效率和燃料转换比:系统能量放大系数不低于320;锕系元素（MA）和裂变产物（FP）的嬗变支持比分别为1个和2个压水堆;热包层的燃料转换比为0．715。 Accelerator driven coupled fast/thermal subcritical system is conceptually designed. In the system, the inner/fast blanket and the outer/thermal blanket are separated each other by large vacuum and B4C coating for on edirection coupling. The metal type fuel （MA ＋ Pu）Zr is loaded into the fast blanket. The oxide type fuels （Th ＋ Pu） O2 and （Th ＋ ^233U ＋ Pu）O2 are loaded into the thermal blanket during the initial cycle and the equilibrium cycle, respectively. ^99Tc, ^129I and ^135Cs are loaded respectively in the form of pure technetium metal, sodium iodide and cesium chlorine into the thermal blanket. The system has good transmutation efficiency, high energy amplification factor and good fuel conversion ability: the energy amplification factor is above 320; the transmutation support ratios of MA and FP are about 1.0 and 2.0 PWRs respectively; the fuel conversion ratio in the thermal blanket is about 0. 715.     

Components of antineutrino emission in nuclear reactor

New ?_ee scattering experiments aimed at sensitive searches for the ν_e magnetic moment and projects to explore small mixing angle neutrino oscillations at reactors require a better understanding of the reactor antineutrino spectrum. Six components which contribute to the total ?_e spectrum generated in a nuclear reactor are considered. They are beta decays of the fission fragments of ²³⁵U, ²³⁹Pu, ²³⁸U, and ²⁴¹Pu and decays of beta emitters produced as a result of neutron capture in ²³⁸U and in accumulated fission fragments which perturb the spectrum. For antineutrino energies of less than 3.5 MeV and for each of the four fissile isotopes, the time evolution of ?_e spectra is given during fuel irradiation and after the irradiation is stopped. The relevant uncertainties are estimated. Small corrections to the ILL spectra are considered.     

高温气冷堆堆芯球床流动与温度分布均匀性研究

由于燃料球的随机分布和球床的壁面效应,球床式高温气冷堆堆芯孔隙率分布会有一定的不均匀性。深入认识壁面漏流、随机孔隙率对球床温度分布均匀性的影响对进一步提高高温气冷堆冷却剂出口温度及其安全性具有重要意义。本文采用多孔介质模型实现了对堆芯球床壁面漏流、随机孔隙率效应的数值模拟。结果表明,由于壁面漏流效应,壁面附近局部区域冷却剂最大速度会比中心高50%,对球床温度影响则不大。中心区域局部极小、极大孔隙率只对很小区域内流速和温度有影响,但温度变化幅值很小。球床中心随机孔隙率使冷却剂速度波动小于13%,对球床温度影响很小。     

Studies of the radionuclide content of fuel particles sampled from the chernobyl nuclear power plant’s No. 4 reactor unit

Spectroscopic studies are performed of the L _x , K _x , and ?? emanations from fuel particles sampled in 2011 inside the Chernobyl Nuclear Power Plant??s (ChNPP??s) No. 4 reactor unit. The isotope ratios for ^134,135Cs, ^154,155Eu, Pu isotopes, ^241,243Am, and ²⁴³Cm are measured. The data on ?? emitters for all radionuclides above ²⁴¹Am exhibit considerable inconsistency with the theoretically calculated values. A systematic deviation of the ⁹⁰Sr and ¹³⁷Cs ratios for the fuel component from the 1986 data is observed. Zirconium is shown to be the main radionuclide in the fuel particles.     

热管堆在不同基体材料下的燃耗特性分析北大核心CSCD

热管冷却核反应堆具有非能动传热、模块化和固有安全性高等特点,在航空探索、深海作业和偏远地区电力市场上有广泛的应用。以洛斯阿拉莫斯国家实验室开发的5 MWth热管堆为研究对象,选择SS-316,Mo-14Re和SiC作为基体候选材料,采用反应堆蒙特卡罗中子输运分析程序对比分析了以上三种基体堆芯的反应性、中子能谱、增殖性能和燃耗演化。结果表明:为了维持堆芯的10年运行,SS-316,Mo-14Re和SiC三种基体堆芯所需的初始燃料235U富集度分别约为19.35%,28.80%和17.10%,SiC基体堆芯所需的初始燃料235U富集度最小;10年后,SiC基体堆芯产生的易裂变核素(239Pu和241Pu)和次锕系核素(通过分离嬗变可被再次利用)的量最高,分别约为11.91 kg和92.08 g。综合以上研究结果,推荐SiC作为热管冷却核反应堆的基体。     

On the transmutation of Am in a fast lead-cooled system

Characteristics of the equilibrium fuel cycle for the core or a blanket of ADS having the structure of the core of a fast lead-cooled reactor of type BREST (Russian abbreviation for ‘Bystryy Reaktor so Svintsovym Teplonositelem’) in a mode of americium transmutation are calculated. Americium loading was taken 5% of heavy atoms. Keeping the average multiplication factor the same as in a standard equilibrium cycle, reactivity swing over 1 year's microcycle is about 1%, that demands partial fuel reloading with a periodicity of about one month. For one year of operation, 61 kg of americium is destroyed, and due to increased ²³⁸Pu content, americium is mainly converted to fission products. Thus in a system of 1 GWt (thermal), 87 kg of americium can be transmuted yearly. The estimate of the reactivity void effect has shown that it increases to 0.6% almost linearly with the void fraction increasing up to 25% and reaches its maximum of 0.7% at a void fraction of about 50%. Application of similar strategy for ADS with a sub-criticality level ≈0.96−0.98 can essentially relax safety problems related to positive void effects.      

CMRR堆内高温高压辐照考验回路典型事故分析

基于中国绵阳研究堆(CMRR)高温高压辐照考验回路初步设计方案,就回路失水事故(LOCA)及失流事故(LOFA)两类典型事故进行分析。结果表明:回路在冷管段及热管段失水事故下包壳热点温度最高为880.6 ℃及367.6 ℃,均远低于1204 ℃;全部失流事故下最小偏离泡核沸腾比(MDNBR)大于1.5,不会发生偏离泡核沸腾;卡轴事故中包壳最高温度为734.1 ℃,低于1482 ℃。上述结果均满足验收准则,符合安全法规要求。     

Repeatedly using the decomposition product of struvite by ultrasound stripping to remove ammonia nitrogen from landfill leachate

In this study, the decomposition of struvite by ultrasound stripping and the recycle use of the decomposition product for the treatment of landfill leachate were investigated. The results indicated that when the decomposition of struvite by ultrasound stripping was performed at 55 °C for 40 min, the ammonium in the struvite could be almost completely eliminated from the solution system. The characterization analysis showed that magnesium phosphate and the dissolved phosphate ions were the main active derivatives. Approximately 90% of the total ammonia nitrogen (TAN) in landfill leachate can be removed by reusing the decomposition product at pH 9 for 60 min. Repeated use of the struvite decomposition product revealed that the TAN removal efficiency decreased with an increase in the number of recycles. However, in the process of multiple recycling, about 90% of TAN removal could be maintained by supplementing a certain amount of the preformed struvite to the solution for every recycle. An economic analysis demonstrated that 79.3% of the treatment cost could be saved by the proposed process compared to the non-recycling process.     

核爆发电问题中核燃料循环的中子学研究

研究核爆聚变电站中的一类中子学问题——如何利用核爆炸产生的大量中子生产核燃料.首先,根据核爆中子场的特点,确定可以利用的中子能量范围.用慢化理论对慢化材料的厚度进行估计,并用MCNP程序进行数值计算.研究如何利用中子反照效应减少造材料层的厚度,最后提出一个较合理的造材料技术路线.