压水堆内钍-铀增殖循环研究——乏燃料特性分析

利用ORIGENS程序对压水堆钍基乏燃料的特性进行分析,揭示了钍基乏燃料在放射性毒性、衰变热、γ射线等方面的特性,相关结果可为钍基乏燃料的贮存、后处理和地质处置提供必要的参考。研究的乏燃料是压水堆内钍-铀增殖循环堆芯设计方案中的4种,包括UOX（铀氧化物）、MOX（钚铀混合氧化物）、PuThOX（钚钍混合氧化物）和U3ThOX（工业级²³³U-钍混合氧化物）。研究结果表明：1）由于超铀核素的含量极低,在卸料后1 000年内,U3ThOX的放射性毒性显著低于超铀核素含量高的乏燃料;2）由于²³²U衰变链中²⁰⁸Tl的贡献,钍基乏燃料中2.6 MeV能量附近的γ射线强度明显高于铀基乏燃料,而这一能量附近的γ射线强度在卸料后约10年达到局部峰值,所以,钍基乏燃料的后处理最好避开此时间。     

球床式高温气冷堆MOX燃料循环优化

与压水堆相比,球床式高温气冷堆能在堆芯结构不做明显改变的情况下采用全堆芯装载混合氧化物（MOX）燃料元件。基于250 MW球床模块式高温气冷堆堆芯结构,设计了4种球床式高温气冷堆下MOX燃料循环方式,包括铀钚混合的燃料球和独立的钚球与铀球混合装载的等效方式,采用高温气冷堆设计程序VSOP进行分析,比较了初装堆的有效增殖因数、燃料元件在堆芯内滞留时间、卸料燃耗、温度系数等主要物理特性。结果表明：采用纯铀和纯钚两种分离燃料球且铀燃料球循环时间更长的方案,平均卸料燃耗较高,总体性能较其他循环方式优越。     

灵活燃料循环革新型水冷反应堆（FLWR）的概念设计和循环特性

提出了一种基于轻水反应堆（LWR）技术丰富经验、应用灵活的燃料循环的革新型水冷反应堆（FLWR）概念设计。该设计的目的是通过两个阶段的钚多次循环,实现有效和灵活地利用铀和钚的资源。在第一阶段中,FLWR堆芯是实现高转换型堆芯的概念设计,基本上平稳地保持现有轻水堆和来自轻水堆铀-钚混合氧化物（MOX-LWR）燃料技术的技术连续性,从技术的观点看没有重大的差异;第二阶段的堆芯将是一种慢化剂-减少型水冷反应堆（RMWR）堆芯的概念设计,达到大于1．0的高转换率。钚（Pu）的多次循环,对于长期持续的能量供应是有利的。FLWR是一种沸水堆型（BWR）反应堆,其堆芯设计特点为：堆芯呈短粗状,装载以三角形的栅格排列的燃料棒组成的六角形燃料组件,装有高富集度的混合金属氧化物（MOX）燃料和Y形控制棒。堆芯在两个阶段中使用一致的和相同尺寸的燃料组件,因此在反应堆运行寿期内,在同一个反应堆系统中,前一个反应堆堆芯概念设计可以过渡到后一个堆芯概念设计,这样就可以灵活地响应天然铀资源未来情况的预期变化,或建立金属氧化物乏燃料的经济的后处理技术。 完成了堆芯设计的详细研究,结合其他有关的研究,迄今为止所获得的结果已经表明所提出的这种反应堆概念设计是可行并具有发展前景的。     

压水堆平衡堆芯钍铀燃料循环初步研究

建立WIMSD5-SN2-CYCLE3D和CASMO3-CYCLE3D物理分析系统作为钍铀燃料循环研究工具.以大亚湾第1机组压水堆为参考堆型,不改变反应堆栅元、组件和堆芯的结构与几何尺寸,设计出含36根钍棒、4.2#5U富集度的新型含钍组件,并对含钍组件和3.2%富集度的铀组件进行中子学计算和分析.模拟并分析了大亚湾压水堆12个月换料从初始循环到铀钚平衡循环的换料过程.再从平衡铀堆芯出发,逐步加入含钍组件代替铀组件,对铀钚平衡循环到钍铀平衡循环的换料过程进行了模拟与分析.计算结果表明:钍铀平衡循环比铀钚平衡循环每天节省裂变核素质量约18.4%,并减少了长寿命放射性核废料的产生.不利因素是使得循环长度减少90EFPD,缩短了换料周期,增加运行费用,并给燃料管理、安全控制以及乏燃料的处理带来困难.建议提高组件的235U富集度,在压水堆上进行钍利用研究.     

球床高温堆钍-铀混合氧化物装载方案中子学与事故分析

根据钍-铀混合氧化物燃料在高温气冷堆核电站示范工程(HTR-PM)框架下的中子学与瞬态事故特性,基于铀原子份额和燃料碳/重金属比例2种参数的参数分析,寻找混合氧化物(MOX)装载的优化方案。分析结果表明,随着碳/重金属比例的减小,单位产能对应的天然铀需求量降低,同时以失冷失压事故后燃料温度为代表的安全特性参数都逐渐恶化;最优化方案相比于HTR-PM实际燃料装载方案,可节省约8.5%的U3O8需求量[~20 kg/(GW·d)];同时混合氧化物方案对钍燃料的利用率很低,仅为6%左右,必须进一步探索提高钍燃料在线利用率的途径。     

压水堆中使用分立型铀、钍燃料组件的堆芯物理特性研究

通过对分立型铀、钍燃料组件 ,使用在秦山 30 0MW电功率压水堆核电厂中堆芯物理特性的探讨 ,寻找2 3 2 Th在PWR中可能利用的途径。为此 ,特采用铀、钍燃料组件分立的双进料系统的装卸料方法 ,其堆芯寿期分别为铀组件 3个循环 ;钍组件 1 0个循环。并以秦山核电厂为参考电厂 ,进行了 1 0个循环的燃耗计算 ,每一循环装料时均有 4个钍组件进堆。计算结果表明 :到第 1 0循环寿期末 ,堆芯中 40个钍组件所含的2 3 3 U总量已达到 2 1 2 6kg ,可直接参与堆芯的链式反应 ,从而达到利用2 3 2 Th的目的。并可同全铀组件堆芯比较中看出 ,分立型铀、钍组件混装堆芯每一循环 (第 1 0循环后 )可少装 2 0 0多kg2 3 5U ,这样就为钍 铀燃料循环展示了光明的前景。当然如果要达到实际应用 ,仍有许多工程技术问题亟待解决     

使用离散钍铀燃料组件的CANDU6堆物理特性初步研究

为研究钍铀燃料在CANDU6堆中的应用,采用DRAGON/DONJON程序,对使用离散型钍铀燃料37棒束组件的CANDU6堆进行时均堆芯分析。结果表明,组件采用235U富集度为2.5%的铀棒以及第1、2、3圈布置钍棒的37棒束组件,堆芯在8棒束换料、3个燃耗分区的方案下,组件的冷却剂空泡反应性较使用天然铀的37棒束组件（NU-37组件）与采用混合钍铀元件棒的37棒束组件更负;堆芯最大时均通道/棒束功率满足小于6700?kW/860?kW的限值;燃料转化能力比采用NU-37组件时更高;卸料燃耗可到达13400?MW·d/t(U)。研究表明,所设计的离散型钍铀燃料37棒束组件可用于现有CANDU6堆芯,且无需对堆芯结构及控制机构作重大改造;燃料组件和堆芯设计方案可为钍铀燃料在CANDU6堆芯的应用提供参考。      

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.     

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。      

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。     

Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated.Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin.The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core.Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities.The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B₄C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution. The temperature reactivity coefficients of the TOX core were found to be always negative. The TOX core has a slightly reduced, as compared to UOX core, but still sufficient shutdown margin.In the TOX core β_eff is smaller by about a factor of two in comparison to the UOX core and even lower than that of the MOX core. The combination of small β_eff and reduced control materials worth may potentially deteriorate the performance under RIA conditions and requires an additional examination. The behavior of the considered cores during the most limiting RIAs, such as rod ejection, main steam line break, and boron dilution, is further investigated and reported in Part II of the paper.     

Core physics analysis of 100% MOX core in IRIS

International Reactor Innovative and Secure (IRIS) is an advanced small-to-medium-size (1000 MWt) Pressurized Water Reactor (PWR), targeting deployment around 2015. Its reference core design is based on the current Westinghouse UO₂ fuel with less than 5% ²³⁵U, and the analysis has been previously completed confirming good performance for that case. The full MOX fuel core is currently under evaluation as one of the alternatives for the second wave of IRIS reactors. A full 3-D neutronic analysis has been performed to examine main core performance and safety parameters, such as critical boron concentration, peaking factors, discharge burnup, reactivity coefficients, shut-down margin, etc. In addition, the basis to perform load follow maneuvers via the Westinghouse innovative strategy MSHIM has been established. The enhanced moderation of the IRIS fuel lattice facilitates MOX core design, and all the obtained results are within the operational and safety limits considered thus confirming viability of this option from the reactor physics standpoint.     

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.     

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.     

Neutronics and material attractiveness for PWR thorium systems using monte carlo techniques

Thorium (Th) oxide fuel offers a significant advantage over traditional low-enriched uranium and mixed uranium/plutonium oxide (MOX) fuel irradiated in a Light Water Reactor. The benefits of using thorium include the following: 1) unlike depleted uranium, thorium does not produce plutonium, 2) thorium is a more stable fuel material chemically than LEU and may withstand higher burnups, 3) the materials attractiveness of plutonium in Th/Pu fuel at high burnups is lower than in MOX at currently achievable burnups, and 4) thorium is three to four times more abundant than uranium. This paper quantifies the irradiation of thorium fuel in existing Light Water Reactors in terms of: 1) the percentage of plutonium destroyed, 2) reactivity safety parameters, and 3) material attractiveness of the final uranium and plutonium products. The Monte Carlo codes MCNP/X and the linkage code Monteburns were used for the calculations in this document, which is one of the first applications of full core Monte Carlo burnup calculations. Results of reactivity safety parameters are compared to deterministic solutions that are more traditionally used for full core computations.Thorium is fertile and leads to production of the fissile isotope ²³³U, but it must be mixed with enriched uranium or reactor-/weapons-grade plutonium initially to provide power until enough ²³³U builds in. One proposed fuel type, a thorium-plutonium mixture, is advantageous because it would destroy a significant fraction of existing plutonium while avoiding the creation of new plutonium. ²³³U has a lower delayed neutron fraction than ²³⁵U and acts kinetically similar to ²³⁹Pu built in from ²³⁸U. However, as with MOX fuel, some design changes may be required for our current LWR fleet to burn more than one-third a core of Th/Pu fuel and satisfy reactivity safety limits. The calculations performed in this research show that thorium/plutonium fuel can destroy up to 70% of the original plutonium per pass at 47 GWd/MTU, whereas only about 30% can be destroyed using MOX. Additionally, the materials attractiveness of the final plutonium product of irradiated plutonium/thorium fuel is significantly reduced if high burnups (∼94 GWD/MTU) of the fuel can be attained.     

Qualification of the neutronic efficiency of erbium at zero burnup

The main asset of erbium as an alternative burnable poison to gadolinium is that it has a much lower thermal absorption efficient cross section that contributes to giving it much slower consumption kinetics than gadolinium, and also helps to generate much lower perturbation in the power distribution. The calculations performed with the APOLLO code and its associated library must be qualified and validated with an experiment in order to obtain a sufficient degree of confidence to envisage an industrial application of this poison. For this purpose the MIRTE UOX and MOX experiments were performed in the critical reactor EOLE at Cadarache within the framework of the EROÏNE programme. These experiments concern the neutronic assessment of a (U,Er)O₂ rod in a representative core of a pressurised water reactor lattice with an enhanced moderation ratio. The purpose of this paper is to show that the APOLLO2 code associated with its APOLLIB CEA 93 library is perfectly qualified at time zero to calculate erbium reactivity worth.     

Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR) has been investigated. If increasing quantities of ²³⁸U, the fertile isotope in standard UO₂ fuel, are replaced by ²³²Th, then a greater yield of new fissile material (²³³U) is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 10³ years. Two innovative fuel types named S90 and S20, ThO₂ mixed with 90% and 20% enriched UO₂ respectively, are compared as an alternative to standard uranium oxide (UOX) and uranium/plutonium mixed oxide (MOX) fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated. Finally, we discuss the economics of such strategies.