高温气冷堆和轻水堆超铀元素管理特性研究

乏燃料中长寿命锕系元素对环境造成长期潜在危害,本文研究球床高温气冷堆不同燃料循环中超铀元素的产生和焚烧特性。在250 MW球床模块式高温气冷堆示范电站HTR-PM铀钚循环的乏燃料中提取铀和钚作为核燃料,设计了PuO₂和MOX燃料元件,将新设计的燃料元件重新装入与HTR-PM相同结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。采用高温气冷堆物理设计程序VSOP,研究了高温气冷堆一次通过燃料循环和不同闭式燃料循环的超铀元素焚烧特性,并与轻水堆燃料循环结果进行比较和分析。结果表明：高温气冷堆一次通过燃料循环超铀元素生成率约为轻水堆的1/2;高温气冷堆闭式燃料循环能有效嬗变超铀元素。     

基于精细燃耗历史及精细燃耗链的球床高温气冷堆燃耗不确定性分析

球床高温气冷堆的燃料管理具有燃料球多次通过堆芯的特点,使得燃料元件经历的燃耗历史十分复杂。球床高温气冷堆堆芯物理设计程序VSOP可以提供燃料元件的精细燃耗历史,但仅包含少量燃耗链和核素种类。而清华大学自主开发的燃耗计算程序NUIT可实现精细燃耗计算,且包含完整燃耗链和核素信息,但不具备精细燃耗历史跟踪功能。本文基于NUIT,结合VSOP提供的球床高温气冷堆精细燃耗历史,开发了球床高温气冷堆堆芯的精细燃耗计算功能,搭建了带有精细燃耗历史模拟和精细燃耗链核素的燃耗分析流程,并实现燃耗不确定性分析功能。在此基础上研究了裂变产额不确定性对球床高温气冷堆燃耗计算不确定性的贡献,并与VSOP的计算结果进行对比。计算分析结果显示,基于NUIT的精细燃耗计算结果和VSOP的燃耗计算结果得到了相互验证,且可以得到更多的核素浓度信息,该计算结果是开展球床高温气冷堆衰变热不确定性研究的基础。     

压水堆内钍-铀增殖循环研究——堆芯设计

在全UOX（铀氧化物）堆芯平衡循环的基础上,研究了UOX/PuThOX（钚钍混合氧化物）混合堆芯和UOX/U3ThOX（工业级²³³U-钍混合氧化物）混合堆芯的燃料管理方案设计,实现了钍 铀增殖循环。U3ThOX燃料组件在堆内停留6个燃料循环,平均循环长度较参考的全UOX堆芯增加5 EFPD;U3ThOX燃料组件卸料后冷却1年时易裂变核素存量较装料时增加了7%。为比较分析,设计了UOX/MOX（钚铀混合氧化物）混合堆芯的燃料管理方案。核特性分析结果表明：1）装载PuThOX燃料对堆芯核特性产生的影响与装载MOX燃料类似,硼微分价值和控制棒价值减小、临界硼浓度变大、慢化剂温度系数更负、停堆裕量减小、多普勒亏损更大;2） UOX/U3ThOX混合堆芯和参考的全UOX堆芯具备相似的核特性。     

聚变-裂变混合堆设计研究

利用MCNP5和MONK9A程序对聚变驱动裂变混合堆进行了初步研究,在等离子体第1壁外侧依次包覆长方体形状的燃料组件和产氚组件,形成裂变堆芯包层和产氚区.对分别装载贫铀、天然铀、贫铀MOX和天然铀MOX等4种燃料的混合堆进行了研究分析,其中,后两种燃料在整个运行寿期内的功率放大倍数和氚增殖比满足设计要求.通过随燃耗变化的同位素含量分析,初步探讨了混合堆的铀-钚燃耗循环策略.     

深燃耗球床式气冷快堆研究

气冷快堆是未来发展的第四代先进核能系统候选堆型之一,它可以满足核能的可持续性、安全可靠性和经济性要求.从反应堆物理和热工水力学的角度出发,设计了热功率300 MW的球床式气冷快堆,选择了碳化物燃料作为气冷快堆的燃料.用耦合燃耗计算程序COUPLE2.0模拟得到了深燃耗气冷快堆的铀燃料循环的平衡态.平衡态研究结果表明基于深燃耗的300 MW球床式气冷快堆可以提高铀资源的利用率同时降低乏燃料中的次锕系核素的含量.当燃料球直径为6 cm,燃料区的直径为5.5 cm,燃料占燃料区的体积的70%,燃料形式为UC,其中235U的初始富集度为12%时,燃料球通过堆芯的时间可以达到12 600 d,重金属燃耗深度为164.38 GWd/t,总的铀资源的利用率可以达到为28.03%.     

球床式高温气冷堆球流混流的影响分析

研究球床式高温气冷堆球流存在的混流对堆芯关键参数的影响。开发了能模拟球流混流过程与效果的MFVSOP程序。选择球床模块式高温气冷堆核电站示范工程(HTR-PM)平衡堆芯为研究对象,对比分析不同的混流程度对堆芯功率峰值、功率密度等参数的影响及其不确定性。分析发现,混流对球床式高温气冷堆关键参数的不确定性影响不大,多次通过的燃料循环方式可降低不确定性。     

冷却衰变时间对球床模块式高温气冷堆在线燃耗测量的影响分析

球床模块式高温气冷堆(HTR-PM)需要对球形燃料元件进行在线燃耗测量,以决定其是否退出燃料循坏.在燃料元件卸出堆芯到达测量位之前,要经过一段时间的冷却与衰变;这段时间对于燃耗测量过程有较大影响.利用同位素燃耗与衰变分析软件KORIGEN和粒子输运模拟计算软件MCNP相结合的方式,分析了燃料元件冷却衰变时间对燃耗测量过程的影响.结果表明,只要采取适当的γ谱分析方法,冷却衰变时间大于50 h,就可以满足HTR-PM对燃耗测量系统的要求.     

杂质硼在高温气冷堆中的燃耗特性

在球床式高温气冷堆的堆芯和石墨反射层中,不可避免地含有少量杂质硼。硼杂质的存在及其燃耗会对反应堆的反应性产生影响。对于多次通过的球床堆芯,根据燃料元件的运行历史计算所有元件的硼燃耗,对于中子注量率差别较大的反射层,分区计算了硼燃耗。再采用微扰理论,计算燃耗过程中硼反应性价值的变化。计算结果表明,硼杂质燃耗很快,因此,硼杂质对反应性的影响降低很快。     

球床高温气冷堆卸料燃耗阈值对平均卸料燃耗的影响

基于蒙特卡罗方法开发了球床高温气冷堆燃料球运行历史模拟程序,分析不同卸料燃耗阈值对平均卸料燃耗、卸料燃耗分布的影响,并分析了不同球流速度模型下的差别。结果表明,卸料燃耗阈值对于平均卸料燃耗、卸料燃耗分布很大程度上受到各流道燃料增量的特性的影响。     

基于MOX燃料组件的177混合堆芯装料方案数值研究

通过计算华龙一号（HPR1000）压水堆平均卸料燃耗得到乏燃料中钚（Pu）同位素的含量,以此成分比例来设计铀钚混合氧化物（MOX）燃料。采用离散型燃料组件设计,通过不同Pu含量的MOX燃料棒离散型布置来降低与UO₂燃料组件间的功率梯度。采用程序MCNP和COSLATC模拟堆芯功率分布和热中子注量率分布,采用分区分层的低泄漏装料方案,降低不同燃料组件间的功率梯度,展平堆芯的功率分布。在不考虑可燃毒物的前提下,利用3种Pu含量的MOX组件将混合堆芯的功率峰因子控制在1.77左右,明显优于原堆芯的功率峰因子,为国产三代压水堆引入MOX燃料提供了具有参考价值的装料方案。      

Core Burnup Calculation of Plutonium Burning Inert-Matrix Fueled High Temperature Gas Cooled Reactor

For the efficient reduction of excess plutonium amount, Japan Atomic Energy Research Institute (JAERI, now Japan Atomic Energy Agency) has studied a concept of rock-like oxide (ROX) fuel as a kind of inert matrix fuel (IMF). In the JAERI study, ROX fuel is burnt in existing light water reactors (LWRs), while in this study, pebble bed type high temperature gas cooled reactor (HTGR) is studied, mainly because of its high neutron economy and softer neutron spectrum than LWRs. Here, PuO₂-yttria stabilized zirconia (YSZ: (Zr,Y)O_2-x) particles are dispersed in graphite matrix. In the ROX fueled LWR study, it was necessary to improve fuel temperature reactivity coefficients by adding such additives as ²³⁸U and Er. Here in HTGR, although the negative temperature coefficient is much larger than that in LWR without any improvements, temperature coefficient was improved as large as possible to the level of UO₂ HTGR case by adding Er in the fuel. Burnup calculations on PuO_2-YSZ fueled HTGR, by simulating the continuous refueling of fuel pebbles with the batch fuel loading, showed almost complete transmutation for ²³⁹Pu and more than 80% for the total plutonium. As the maximum power density of the fuel pebble obtained by the core burnup calculation was very large in comparison with the UO₂ HTGR, the maximum temperature in YSZ fuel particle was also evaluated. Despite the low thermal conductivity of YSZ, the evaluated YSZ temperature was well below the melting point, thanks to the high thermal conductivity of graphite and small YSZ particle size. Here, the high power density in the Pu-YSZ fueled core might become a problem, but is possible to be reduced by adjusting the initial plutonium enrichment.     

Development of an innovative detection system for pebble bed kinematic studies

Pebble bed reactors enable the circulation of pebble fuel elements when the reactors are in operation.This unique design helps to optimize the burnup and power distribution, reduces the excessive reactivity of the reactor,and provides a mean to identify and segregate damaged fuel elements during operation. The movement of the pebbles in the core, or the kinematics of the pebble bed,significantly affect the above features and is not fully understood. We designed and built a detection system that can measure 3-axis acceleration, 3-axis angular velocity,3-axis rotation angles, and vibration and temperature of multiple pebbles anywhere in the pebble bed. This system uses pebble-shaped detectors that can flow with other pebbles and does not disturb the pebble movement. We used new technologies to enable instant response, precise measurement, and simultaneous collection of data from a large number of detectors. Our tests show that the detection system has a negligible zero drift and the accuracy is better than the designed value. The residence time of the pebbles in a moving pebble bed was also measured using the system.     

Burnup characteristics and fuel cycle economics of mixed uranium–thorium fuel in a simplified small pebble bed reactor

The limitation of natural uranium resources and the improvement of economic values of nuclear reactors are important issues to be solved in the future development of these reactors. In our previous study, we presented an innovative design for simplifying a pebble bed reactor, and the optimization of this design showed that burnup values could be increased and natural uranium uses could be reduced. The purposes of the current study were to design a simplified pebble bed reactor by removing the unloading device from the reactor system and to further optimize the burnup characteristics of this reactor with a peu à peu fuel-loading scheme by introducing thorium in the fuel configuration as a fertile material. Another goal was to optimize the fuel composition so that the system could achieve even better burnup characteristics and use scarce uranium resources more efficiently. Using a specially developed computer code, we analyzed and optimized the performance of a 110-MWt simplified pebble bed reactor using a peu à peu fuel-loading scheme. An optimized design using 30% of fertile thorium mixed with uranium fuel with 15% ²³⁵U enrichment and a 7% packing fraction calculated to achieve a high burnup of 140 GWD/T for more than 21 years' operation time that could save 13 to 33% of natural uranium use compared with the savings noted in our previous study. Neutronic, burnup and fuel economic analysis for this optimized design are discussed in this study.     

Advanced mixed oxide fuel assemblies with higher plutonium contents for pressurized water reactors

Optimizing fuel cycle costs by increasing the final burnup leads to reduced generation of plutonium. Under properly defined boundary conditions thermal recycling in mixed oxide (MOX) fuel assemblies (FAs) reduces further the amount of plutonium which has to be disposed of in final storage. Increasing the final burnup requires higher initial enrichments of uranium fuel to be matched by an advanced design of MOX FAs with higher plutonium contents. The neutronic design of these MOX FAs has to consider the licensing status of nuclear power plants concerning the use of MOX fuel. The Siemens Nuclear Fuel Cycle Division, with more than 20 years' experience in the production of MOX fuel, has designed several advanced MOX FAs of different types (14 × 14 to 16 × 16) with fissile plutonium contents up to 4.60 w/o.     

Study of the radiotoxicity of actinides recycling in boiling water reactors fuel

In this paper the production and destruction, as well as the radiotoxicity of plutonium and minor actinides (MA) obtained from the multi-recycling of boiling water reactors (BWR) fuel are analyzed. A BWR MOX fuel assembly, with uranium (from enrichment tails), plutonium and minor actinides is designed and studied using the HELIOS code. The actinides mass and the radiotoxicity of the spent fuel are compared with those of the once-through or direct cycle. Other type of fuel assembly is also analyzed: an assembly with enriched uranium and minor actinides; without plutonium. For this study, the fuel remains in the reactor for four cycles, where each cycle is 18 months length, with a discharge burnup of 48 MWd/kg. After this time, the fuel is placed in the spent fuel pool to be cooled during 5 years. Afterwards, the fuel is recycled for the next fuel cycle; 2 years are considered for recycle and fuel fabrication. Two recycles are taken into account in this study. Regarding radiotoxicity, results show that in the period from the spent fuel discharge until 1000 years, the highest reduction in the radiotoxicity related to the direct cycle is obtained with a fuel composed of MA and enriched uranium. However, in the period after few thousands of years, the lowest radiotoxicity is obtained using the fuel with plutonium and MA. The reduction in the radiotoxicity of the spent fuel after one or two recycling in a BWR is however very small for the studied MOX assemblies, reaching a maximum reduction factor of 2.     

Design of an overmoderated fuel and a full MOX core for plutonium consumption in boiling water reactors

The use of uranium–plutonium mixed oxide fuel (MOX) in light water reactors (LWR) is nowadays a current practice in several countries. Generally 1/3 of the reactor core is loaded with MOX fuel assemblies and the other 2/3 with uranium assemblies. Nevertheless the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this paper the design of a boiling water reactor (BWR) core fully loaded with an overmoderated MOX fuel design is investigated. The design of overmoderated BWR MOX fuel assemblies based on a 10×10 lattice are developed, these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. In order to increase the moderator to fuel ratio two approaches are followed: in the first approach, 8 or 12 fuel rods are replaced by water rods in the 10×10 lattice; in the second approach, an 11×11 lattice with 24 water rods is designed with an active fuel length very close to the standard MOX assembly. The results of the depletion behavior and the main steady state core parameters are presented. The feasibility of a full core loaded with the 11×11 overmoderated MOX fuel assembly is verified. This design take advantage of the softer spectrum comparable to the 10×10 lattice with 12 water rods but with thermal limits comparable to the standard MOX fuel assembly.     

Small high temperature gas-cooled reactors with innovative nuclear burning

Since the innovative concept of CANDLE (Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety, easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles.     

Thermal hydraulic calculation of the HTR-10 for the initial and equilibrium core

The thermal hydraulic calculations of the 10 MW high temperature gas-cooled-test module (HTR-10) are among the most important indications to judge the reactor performance under design conditions. The power distribution, the temperature distribution and the flow distribution of the HTR-10 are calculated for initial and equilibrium core in this paper. The temperature distribution includes the temperature parameters of fuel elements, the helium coolant and the main components in the reactor. In the temperature calculation of fuel elements, several uncertain factors are considered carefully, including non-uniform burnup, power distribution deviation, manufacture deviation of fuel elements, graphite balls mixed with fuel balls in the core, calculation deviation of heat transfer and so on. In the flow distribution calculation, the conservative pebble bed core flow value is selected. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core.