轻水堆乏燃料和钍燃料在ACR-700利用的探索

轻水堆乏燃料和钍燃料的利用是解决乏燃料后处理问题和核燃料短缺的有效途径之一。本工作以ACR-700标准燃料为参考,研究了4种不同混合比例的轻水堆乏燃料及钍燃料在ACR-700中的k_∞和燃耗。研究结果表明,将裂变产物分离后,轻水堆乏燃料的重锕系核素在ACR-700中可作为一很好的燃料;只要加入足够的启动燃料,钍燃料也可作为很好的转换燃料,使反应堆内生成²³³U的速率大于易裂变燃料的消耗速率,²³³U的生成对反应堆运行后期维持临界起重要作用。     

小型低慢化轻水堆的设计

低慢化轻水堆是革新型轻水堆之一，是在现有轻水堆技术的基础上，通过钚的多次循环，实现铀一钚资源的有效利用．通过燃料的高燃耗长期运行，以降低放射性废物的发生量等：．日本原子能研究所在实现轻水堆利用的长期化这一背景下，作为未来型轻水堆     

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。     

轻水堆用超5wt％铀氧化物燃料的实用化技术研究

1 所有实施计划的制定 鉴于全球性能源需求的增加和环境问题的考虑，为了控制放射性废物的产生量，进一步有效利用铀资源，正不断开展以铀富集度超过5wt％燃料（以下称作超5wt％燃料）为前提条件的革新型先进堆以及在现有轻水堆与下一代反应堆中实现高燃耗深度的超5wt％燃料的研究。但是，在燃料制造、动力堆中的应用、燃料的运输与贮藏及后处理等轻水堆燃料循环的各阶段，其前提条件是富集度不到5wt％。要满足超5wt％燃料的要求，重要的课题就是对设备进行变更。然而，在其安全设计方面，最重要的就是临界安全设计问题，富集度5wt％-10wt％这一范围燃料的临界实验数据极少。     

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%.     

高性能包壳管材料的开发

1 前言 到2030年，轻水堆将迎来对现有核电站进行更新换代的时期。提高安全性、经济性和减少废物，以及提高Pu的有效利用对先进的下一代轻水堆开发的需求日益高涨。由于核不扩散方面的制约，现有轻水堆UO2燃料的235^U富集度被限定在5％以内，这对于大幅提高燃耗深度及实现高转换比有很大的制约。MOX燃料使用30％Pu（Pu f20％）级的高富集度的Pu燃料，与现有液态金属快堆（LMFBR）基本相同，能为许多革新型水冷堆所选择。     

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。     

聚变—裂变混合能源堆燃料管理方案设计研究

本文根据聚变-裂变混合能源堆方案设计和燃料组件功率分布的特点,利用自主开发的蒙卡-燃耗耦合程序,开展了详细的燃料管理方案设计研究,分别设计了整体后处理的燃料管理方案、双循环燃料管理方案以及分批燃料管理方案,针对这些类型的燃料管理方案,进行了燃耗分析计算,研究了各种燃料管理方案下各区燃耗及主要裂变核素成分随燃耗的变化。根据各燃料管理方案的主要特点和计算分析结果,对比总结了它们的优点和缺点。本文为今后的聚变-裂变混合能源堆提供了燃料管理上的建议,也为进一步的经济性分析优化研究打下了基础。     

普雷克斯流程也适用于快堆后处理

【西德《原子经济》1983年第9期第434页报道】西德卡尔斯鲁厄核研究中心热化学研究所成功地处理了一批快中子增殖堆的辐照燃料元件。这是在名为“米利”的实验装置中进行的,这批燃料元件共7公斤,最高燃耗达100,000兆瓦·日/吨。从而证明了用于轻水堆辐照燃料的后处理方法——普雷克斯流程完全适用于快堆燃料元件的后处     

快堆金属燃料的发展

国外早期快堆发展的燃料集中在金属燃料上,但金属燃料辐照肿胀严重,只能实现较低的燃耗深度,且较低的固相线温度和与包壳间的共晶温度又制约了金属燃料的实际应用。文章回顾国外金属燃料的发展和主要问题的解决方法,并总结金属燃料改进后可行的设计方案。随后整理早期、后期金属燃料的辐照经验,给出已验证的最大燃耗深度。     

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li₂₀Sn₈₀ coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.     

用切伦科夫计数法确定反应堆燃料元件破损

介绍了利用液闪谱仪进行切伦科夫计数确定反应堆燃料元件破损的原理和方法。以２００ＭＷ核供热堆为例，分析了反应堆主回路水中产生切伦科夫辐射核素的放射性特性，计算了其中活化产物的活性及燃料元件裂变产物的活性，探讨了用本方法监测燃料元件破损的可行性。本方法的特点是操作简单，测量迅速。     

乏燃料溶液嬗变堆焚烧锕系核素能力分析

针对焚烧锕系核素的目标,选择不同的乏燃料成分和堆芯功率,构造了7种乏燃料溶液嬗变堆( HSTR)堆芯模型,采用溶液堆堆芯燃料管理程序FMCHR计算了堆芯内Pu、Np及其他长寿命锕系核素的燃耗变化,分析了HSTR焚烧锕系核素的能力.结果表明:HSTR可以有效实现焚烧239pu的目标,同时嬗变可观数量的237Np;若要实现...     

Basic study on characteristics of some important equilibrium fuel cycles of PWR

Equilibrium fuel cycle characteristics of a light water reactor (LWR) with enriched uranium supply were evaluated. In this study, five kinds of fuel cycles of 3000 MWt pressurized water reactor (PWR) were investigated, and a method to determine the uranium enrichment in order to achieve their criticality was presented. The results show that the enrichment decreases considerably with increasing number of confined heavy nuclides when U is discharged from the reactor. The required natural uranium was also evaluated for two different enrichment processes. The amount of required natural uranium also decreases as well. On the other hand, when U is totally confined, the enrichment becomes unacceptably high. Furthermore, Pu and minor actinides (MA) confining seem effective to incinerate the discharged radio-toxic wastes.     

Equilibrium characteristics of typical fuel cycles of PWR

We have performed equilibrium analysis of light water reactor (LWR) with enriched uranium supply. In this study, five kinds of fuel cycles of 3000 MWt pressurized water reactor (PWR) were investigated, and a method to determine the uranium enrichment in order to achieve their criticality was presented. The results indicated that the enrichment decreases significantly with increasing number of confined heavy nuclides when U is discharged from the reactor. The required natural uranium was also evaluated for two different enrichment processes. The amount of required natural uranium also decreases as well, which may agree with the systematic comparison of typical fuel cycles of PWR on the same condition for resource requirements and discharged radioactive wastes. On the other hand, when U is totally confined, the enrichment becomes uncceptably high.     

Impact of Partitioning and Transmutation on High-Level Waste Disposal for the Fast Breeder Reactor Fuel Cycle

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large storage capacity because the number of waste packages produced is significantly reduced by a factor of 5 from that of the glass waste package in the FBR fuel cycle without PT.     

Fuel cycle analysis of GFR using advanced fuels

The potential benefits of a synergistic light-water reactor (LWR) and gas-cooled fast reactor (GFR) fuel cycle were evaluated for its impact on the front-end and back-end of the fuel cycle. Comparisons were made with conventional once-through cycle (OTC) and MOX fuel cycle. Variations in the synergistic LWR/GFR fuel cycles were based on the degree of recycle in the LWR including both plutonium and reprocessed uranium with concomitant impact on used LWR fuel inventory. This provided a wide range in composition of GFR feed from low to high plutonium content with impact on transmutation/incineration within the GFR fuel cycle. Self-recycle of all actinides was modeled for the GFR with analyses demonstrating that the GFR can be sustained on and consequently accept a wide range of feed materials. Analyses were done using Monteburns along with MCNP and Origen2.2 to model a 60-year period corresponding to the anticipated lifetime of supposed contemporary LWRs and GFRs. All cycles were evaluated based on actinide total mass and isotopic inventory, radiotoxicity, heatload, and resource requirements including natural uranium and SWU. For comparison, all fuel cycles were normalized based on 1 TWHe output. Improvements in fuel cycle performance are dictated by the production and incineration of minor actinides in the GFR and their continued recycle may not be feasible due to the buildup of troublesome isotopes such as ²⁴⁴Cm and ²⁵²Cf. But where uranium and plutonium continue to be recycled beyond the 60-year period, the LWR/GFR cycles demonstrated order of magnitude reductions in used fuel inventories, heatload, and radiotoxicity on a per TWHe basis over LWR only cycles. The full details of the advanced fuel cycle methodology and results are presented.     

熔盐实验堆中燃料核素的灵敏度系数计算分析

以熔盐实验堆为模型,采用MCNP5和SCALE5.1中的TSUNAMI-3D-K5对燃料核素的灵敏度系数进行计算与分析。结果表明,灵敏度系数与核素在MSRE中的含量、位置和核素的中子反应截面有关,得到灵敏度系数最大的核素²³⁵U的宏观裂变截面和宏观俘获截面的灵敏度系数分别为0.267和0.110。MCNP5和TSUNAMI-3D-K5计算不同能区下²³²Th宏观总截面和俘获截面的灵敏度系数曲线一致,曲线在0.1 eV附近有一小峰,振荡区域同截面共振区范围相同。