高功率研究堆低浓化物理特性研究(英文)

应用FG2DB两维两群扩散燃耗程序和带69群中子截面库的CELL栅元少群参数计算程序,对高功率研究堆低浓化堆芯进行了物理计算。LEU燃料元件的铀密度为3.6～7.2g/cm~3,包壳厚度为0.38～0.56 mm。结果表明:改变燃料芯体铀密度或厚度在物理上相当;各堆芯方案的控制棒价值等运行安全有关参数都可以接受,部分计算结果被拟合成线性或二次关系式以便于应用。给出了各堆芯的最小临界值、剩余反应性、运行寿期、快热中子通量和积分通量等物理参数。分析这些参数后指出,当~(235)U含量提高20％或更多时,LEU堆芯与HEU堆芯的主要物理性能相近,这时快中子通量几乎不受影响,热中子通量的下降率近似正比于元件~(235)U含量增加率,但由于LEU堆芯运行寿期的延长,对一般同位素生产与燃料元件辐照考验不会有太大影响。     

微型中子源堆低浓化堆芯的探讨

通过对２３５Ｕ富集度为１９．９％的ＵＯ２和Ｕ３Ｓｉ２－Ａｌ的弥散体２种燃料进行物理计算，从中筛选出了优化的堆芯方案，并对其静态物理参数，诸如有效倍增因子、绝对中子通量密度、上铍反射层反应性价值、反应性温度系数、控制棒价值等进行了计算。     

10MW高温气冷实验堆反射层石墨毒物对平衡态堆芯特性影响分析

对１０ＭＷ高温气冷实验堆（ＨＴＲ－１０）反射层石墨毒物对平衡态堆芯特性的影响进行了敏感性分析计算，并且研究了反射层毒物浓度为５．２ｍｇ／Ｌ硼当量的情况下反应性的补偿手段。结果表明：毒物的存在，致使反应性下降，为了补偿这种效应，需要增大燃料中＾２３５Ｕ的富集度或者增大堆芯装料体积。本文工作可为ＨＴＲ－１０燃料中＾２３５Ｕ的富集度以及其它参数的选取提供参考依据。     

动态连续电解还原制备四价铀的研究

研究了ＵＯ２（ＮＯ３）２－ＨＮＯ３－Ｎ３Ｈ５ＮＯ３－Ｈ２Ｏ体系中，用三级串联电解槽动态连续电解还原制备四价铀的方法，在确定的工艺条件下，当进料液Ｕ（ＶＩ）浓度为１８０ｇ／Ｌ，硝酸浓度为３．１４ｍｏｌ／Ｌ，肼浓度为０．２１０ｍｏｌ／Ｌ，加料速度为２００ｍＬ／ｈ时，可连续制备得到了四价铀浓度大于１５０ｇ／Ｌ的产品溶液。     

硅化铀的肿胀特性和发展前景综述

本文介绍了辐照引起Ｕ３Ｓｉ破裂性肿胀和Ｕ３Ｓｉ２稳定性肿胀的规律及其与显微结构变化的关系。实验证实了Ｕ３Ｓｉ２－Ａｌ板型燃料在超高通量堆中使用的可行性与制造当量铀密度７ｇ·ｃｍ＾－３燃料板的可能性。     

便携式铀丰度仪的研制

研制了用于测量新燃料组件或新燃料元件￣（２３５）Ｕ丰度的便携式铀丰度仪。仪器采用带￣（２４１）Ａｍａ源的φ２０×２０Ｎａｌγ探头，具有高压自控稳定的特性，仪器的电子学线路全部密集安装在３２６×１０６×１７６的枪式箱体内，具有小型、轻便的特点。仪器在测新燃料组件中￣（２３５）Ｕ丰度时，测量的精度好于±２．０％，同时可扫描测出组件中的装料长度。     

低浓化微堆拓展应用可行性研究

对现有微型中子源反应堆（微堆）采用低浓铀燃料并对引出热中子束装置进行了物理可行性研究，给出堆芯核特性参数，并对不同中子束装置的结构方案进行了分析，为微堆燃料元件低浓化并拓宽应用提供了有益的结果。     

研究试验堆用于同位素生产的物理分析

本文对研究试验堆开展同位素生产进行了物理分析。分析了控制棒提棒顺序对同位素产量的影响,提出了提棒因子的概念。依据点堆模型和反应性-燃耗线性公式,得到了同位素的转换比和产量公式。最后根据这些公式,分析了高通量工程试验堆(HFETR)在高浓铀和低浓铀堆芯装载下,堆芯炉的运行寿期、燃料元件装载数量、燃料元件初始平均燃耗和堆芯功率对同位素转换比和产量的影响。结果显示,从小到大提棒、增加堆芯燃料组件盒数和功率水平均会增加堆芯同位素产量,而全年运行段数（运行段间检修时间不变）和堆芯平均初始燃耗增加则起到相反的作用。这些结果已经用于指导反应堆的堆芯装载设计。     

ASDEX—U结果综述

闭合式ＡＳＤＥＸ－Ｕ偏滤器Ⅱ“ＬＹＲＡ”有能力处理高达２０ＭＷ的加热功率或１２ＭＷ／ｍ的Ｐ／Ｒ,因了与开放式偏滤器Ｉ相比到靶板上的最大热通量减小１／２以上。这种减小是由偏滤器区域内碳和氢的强辐射损失引起的,而且与Ｂ２－ＥＩＲＥＮＥ模拟预计是一致的。在中等密度的Ｈ模式下,Ｉ型ＥＬＭ行为表明与加热方法（ＮＢＩ,ＩＣＲＨ）无关。ＡＳＤＥＸ－Ｕ－ＪＥＴ无量纲同一性实验表明了Ｌ－Ｈ转变与堆芯物理限制的相容性     

JT—60U的高密度ELMy H模式等离子体中能量和粒子约束的退降

在ＪＴ－６０Ｕ，用１８－１９ＭＷ的高功率中性约束注入，研究了在ＥＬＭｙＨ模式等离子体中有高再循环滤器时能量和粒子约束的退降，等离子体参数固定在Ｉｐ＝１－１．２ＭＡ，Ｂｔ＝２－２．１Ｔ和ｑ９５＝３．３－３．５。总体能量约束时间τＥ的减小是由于快离子的储能随等离子体密度增加而减小，而热能约束时间τ１ｈ在附着偏滤器条件下从０．０８８－０．０９２ｓ稍微降低到０．０８３ｓ。另一方面，由于自偏滤器注入的中性粒     

原型微堆低浓化初步研究

利用蒙特卡罗计算程序,对高浓铀为燃料的原型微堆的有效增殖因数、控制棒价值、上铍反射层价值以及辐照座内的中子注量率等参数进行了计算。将计算值与实验结果进行了比较,两者基本相符。在原型微堆堆芯尺寸保持不变的情况下,将堆芯燃料元件芯体用富集度为12.5%UO₂替换UAl和用锆包壳替换铝包壳,对堆芯燃料低浓化方案进行了计算,给出了方案的计算结果。并利用RELAP5程序计算了原型微堆低浓铀堆芯阶跃引入4.0 mk反应性情况下反应堆的相关参数。     

Sensitivity analyses of the use of different reflector materials on the neutronics parameters of Tehran research reactor

In this paper, the effect of changes in neutron reflector type on neutronics parameters of Tehran research reactor is analyzed. In this study, at first, calculations for the HEU and LEU fuel configurations of the reactor core in which the light water is used as a neutron reflector in the core is done. Then, by using the reflectors such as graphite, beryllium and heavy water, changes on the neutronic parameters are examined. The results show that by altering the reflector, at HEU core configuration (compared with LEU), more changes appear in parameters such as neutron multiplication factor, average fast and thermal neutron flux, excess reactivity and shut down margin. Moreover, at LEU core configuration, changes are tangible specifically on parameters of cycle length and power peaking factor. In addition, the evaluated fuel temperature coefficient of reactivity is greater at HEU than LEU, while the temperature alteration of fuels represented greater influence on reactivity at LEU configuration.     

Neutronic analysis for core conversion (HEU–LEU) of Pakistan research reactor-2 (PARR-2)

Neutronic analyses for the core conversion of Pakistan research reactor-2 (PARR-2) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel has been performed. Neutronic model has been verified for 90.2% enriched HEU fuel (UAl₄–Al). For core conversion, UO₂ fuel was chosen as an appropriate fuel option because of higher uranium density. Clad has been changed from aluminum to zircalloy-4. Uranium enrichment of 12.6% has been optimized based on the design basis criterion of excess reactivity 4 mk in miniature neutron source reactor (MNSR). Lattice calculations for cross-section generation have been performed utilizing WIMS while core modeling was carried out employing three dimensions option of CITATION. Calculated neutronic parameters were compared for HEU and LEU fuels. Comparison shows that to get same thermal neutron flux at inner irradiation sites, reactor power has to be increased from 30 to 33 kW for LEU fuel. Reactivity coefficients calculations show that doppler and void coefficient values of LEU fuel are higher while moderator coefficient of HEU fuel is higher. It is concluded that from neutronic point of view LEU fuel UO₂ of 12.6% enrichment with zircalloy-4 clad is suitable to replace the existing HEU fuel provided that dimensions of fuel pin and total number of fuel pins are kept same as for HEU fuel.     

Effect of high-density fuel loading on criticality of low enriched uranium fueled material test research reactors

The effect of high-density fuel loading on the criticality of low enriched uranium fueled material test reactors was studied using the standard reactor physics simulation codes WIMS-D/4 and CITATION. Three strategies were considered to increase the fuel loading per plate: (1) by substituting the high-density fuel in place of low-density fuel keeping meat thickness and water channel width constant, (2) by substituting the high-density fuel in place of low-density fuel keeping fuel meat thickness fixed and optimizing the water channel width between the fuel plates and (3) by increasing the fuel meat thickness of fixed density fuel and optimizing the water channel width between the fuel plates. The fuel requirements for critical and first high power cores were determined in each case for higher fuel loadings per plate. It has been found that in the first case, core volume reduces with increasing fuel loadings per plate but requirement of fuel also increases. In the second and third case, core volume as well as fuel requirement decreases with increasing fuel loadings per plate. However in the second case, core volume reduces more rapidly than in case 3 with increasing fuel loadings per plate. Employing standard computer code PARET, steady state thermal hydraulic analysis of all these cores was performed. The thermal hydraulic analysis reveals that cores with higher densities and fixed water channel width are better from thermal hydraulic point of view and have fuel and clad temperatures within the acceptable limits. But the core with higher densities and optimum water channel width is a better choice in terms of core compaction, less ²³⁵U loading and higher neutron fluxes. Finally, the core was compacted in three steps to exploit the benefits of both types of cores. The strategy resulted in 36% reduction in the core volume, 50% increase in thermal neutron flux for irradiation and isotope production and a slight reduction in ²³⁵U loading. All this was achieved with acceptable peak clad and peak fuel centerline temperatures.     

铀水栅零功率反应堆的功率刻度

反应堆功率的测量,在堆功率高时一般用热工方法,功率低时,可用各种堆物理方法,如中子源引进法、中子统计法和全堆总裂变率法。 中子源引进法误差较大,中子统计法需知探测器在堆内的效率和堆的β_(aff)值,此二者都较难测量。全堆总裂变率法是由测量堆的总裂变率来求得堆功率,它可避免前面两种方法的缺点,但需依赖裂变率相对分布的     

γ能谱法在快堆新燃料235U富集度核实测量中的应用

对快堆新燃料组件铀富集度进行了非破坏性核实测量,γ能谱法是测量铀富集度首选方法之一,快堆新燃料235U富集度真实值为64.4％【1】,235U富集度越高测量分析需要时间相对越长,本次核实测量工作量大,环境本底高,精确测量十分困难,对系统硬件的要求很高,能谱解析和数据处理过程更复杂。本次对多根燃料单棒实施了γ能谱法测量,利用专业的软件分析得到235U富集度与真实值绝大部分偏差在3%以内。     

Long-Cycle and High-Burnup Fuel Assembly for the VHTR

For a prismatic VHTR fuel assembly, a physics study has been performed to maximize the fuel performance in terms of the cycle length and the discharge burnup for a given fuel enrichment. The relationship between the fuel performance and the fuel configurations has been investigated in terms of the TRISO packing fraction, diameter of the fuel kernel, fuel management, and moderating power of the fuel block. Both a typical low-enrichment uranium fuel (LEU) and a fuel made of transuranics (TRU) from LWR spent fuel are considered in this paper. It is shown that in order to obtain a long refueling cycle and a high burnup at the same time, the fuel loading needs to be increased together with the moderating power of the fuel block. Three ways are considered for a higher moderation of the fuel block: a larger pitch of the coolant hole pattern, an extra graphite thickness in the fuel block, and a higher graphite density. The impact of the increased pitch on the fuel temperature is also evaluated with a thermal analysis code. We have shown that long refueling cycles and high burnups can be achieved simultaneously for both LEU and TRU fuels.     

Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% ²³⁵U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.     

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity (ρ_ex), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU).