快堆核燃料循环模式的经济性评价

以快堆核电厂的核燃料循环过程及核燃料循环模型为基础,利用注销法对2种核燃料循环方式进行经济性计算和分析;同时,也将快堆燃料循环经济性与压水堆(PWR)燃料"一次通过"的经济性进行对比。按目前价格水平计算,PWR"一次通过"的核燃料循环方式比快堆核燃料循环模式的经济性好,但随着天然铀价格的上涨以及燃料后处理技术水平的进步,快堆核燃料循环费用有望达到或低于PWR"一次通过"的核燃料循环费用。     

基于快堆的钍资源利用展望

由于钍首先在反应堆内经过转换或增殖后变成易裂变核素。^233U才能得以真正利用，因此，选择合适的堆型和燃料循环方式来生产和燃烧。^233U是切实有效利用钍资源的关键问题。本文就基于快堆来分析几种由不同燃料驱动和不同堆型匹配方案形成的钍铀／钚燃料循环模式，探讨我国通过快堆利用钍资源比较合理的燃料循环路线。     

英国当里原型快堆燃料燃耗创纪录,但面临资金问题

【英国《国际核工程》1988年7月号第4页报道】今年5月,苏格兰北部当里原型快堆的一束试验燃料已烧掉20％。原型快堆的原设计目标是燃料烧掉7.5％,现在已差不多达到它的三倍。管理局说,如果商用快堆整个堆芯的燃料能烧掉20％,那么燃料循环费用将减少40％,而且发电成本会象现有的热堆一样低廉。又说,这不包括新燃料中的后处理回收的钚的再利     

中国先进反应堆评价模型及分析

面向21世纪的核能发展,从快中子增殖堆、高温气冷堆及混合堆三种堆型中,中国将研究开发一种能大幅度提高核燃料利用率、安全性与经济性好的堆型。考虑到先进堆评价涉及到很多不确定和较难量化的因素,采用了层次分析法(AHP)和专家咨询法。在层次分析法的指标体系中,选取了4个主要指标:安全性、技术成熟度、经济性和适用性。在广泛征求了专家们的意见和大量调研了国内外文献基础上,选取了氧化物燃料快堆、金属燃料快堆、铀燃料高温气冷堆、铀-钍循环高温气冷堆和混合堆的评价参数,并作了综合评价和排序。同时邀请了130位在政府部门和学术研究部门工作的专家和权威,对先进堆的发展进行了咨询。咨询答卷的回收率达到86％。并对咨询数据进行了计算机处理。从上述两种方法的评价结果中可以得到如下结论:快堆,特别是金属燃料快堆应成为中国21世纪核能发展的主要堆型。     

中国先进反应堆评价模型及分析

面向21世纪的核能发展,从快中子增殖堆、高温气冷堆及混合堆三种堆型中,中国将研究开发一种能大幅度提高核燃料利用率、安全性与经济性好的堆型。考虑到先进堆评价涉及到很多不确定和较难量化的因素,采用了层次分析法(AHP)和专家咨询法。在层次分析法的指标体系中,选取了4个主要指标:安全性、技术成熟度、经济性和适用性。在广泛征求了专家们的意见和大量调研了国内外文献基础上,选取了氧化物燃料快堆、金属燃料快堆、铀燃料高温气冷堆、铀-钍循环高温气冷堆和混合堆的评价参数,并作了综合评价和排序。同时邀请了130位在政府部门和学术研究部门工作的专家和权威,对先进堆的发展进行了咨询。咨询答卷的回收率达到86％。并对咨询数据进行了计算机处理。从上述两种方法的评价结果中可以得到如下结论:快堆,特别是金属燃料快堆应成为中国21世纪核能发展的主要堆型。     

我国钍燃料循环发展研究

调研分析了钍燃料循环的优缺点和国内外研究现状.通过详细分析研究各种堆型的钍资源利用潜力,从核能可持续发展的角度出发,提出了我国钍燃料循环发展的有关结论和建议:(1)当前的核电堆型除高温堆外都不适合进行钍利用;(2)建议采用快堆/热中子堆联合钍燃料循环的方式进行钍资源利用;(3)先进反应堆研究应集中于其堆型本身的研发;(...     

熔盐快堆Th-U燃料循环增殖性能分析

熔盐快堆是当前国际上关注的热点之一,本文基于堆芯结构双流体方案,即裂变熔盐燃料和增殖熔盐介质各自独立冷却循环,利用氟化或氯化熔盐中钍铀重金属盐高温下的高溶解度特性,获得熔盐快堆的高增殖。通过比较钍铀燃料循环熔盐快堆的三种可行性熔盐燃料方案(LiF+ThF_4+UF_4、NaF+ThF_4+UF_4和NaCl+ThCl_3+UCl_3),采用基于反应堆安全分析和设计的综合性模拟程序SCALE(Standardized Computer Analyses for Licensing Evaluation),计算了中子能谱、反应性温度系数,分析了增殖比BR(breeding ratio)受反应堆裂变区、增殖区和ZrC中子反射层的尺寸影响、熔盐中~6Li和~(35)Cl同位素丰度的影响,以及熔盐密度误差对BR计算值的准确性影响、易裂变核素随反应堆运行时间演化等。在钍铀燃料循环熔盐快堆中,通过优化处理得到三种熔盐燃料方案的增殖比BR约为1.2。     

快堆闭式燃料循环对提高铀资源利用率的分析研究

快堆结合闭式燃料循环提高铀资源利用率需对乏燃料进行回收和再循环。对工业钚在大型MOX(混合铀钚)燃料钠冷增殖快堆中多次循环的特性进行了计算分析,结果表明,钚成分经多次循环后可达平衡,其中易裂变核维持在约74%的较高比例。从成分品质看,工业钚在增殖快堆中的循环次数不受限制。构建模型并分析了快堆闭式燃料循环对于铀资源利用率的提高。快堆闭式循环策略下,回收铀、钚多次循环后可大幅度提高铀资源利用率。提高燃料燃耗和乏燃料后处理回收率能显著提升铀利用率;但在最初的几次循环中后处理回收率的影响较小,循环次数增加后,将会对利用率有明显提升。较低的燃料燃耗和回收率情况下,将存在较低的无限次循环铀利用率上限。     

加速器驱动洁净能系统中的燃耗行为分析

研究了加速器驱动洁净核能系统（ADS）次临界反应堆内核素的演化。分析结果表明：ADS具有嬗变长寿命核废物的能力。从快堆和热堆的比较可知,ADS的快堆具有输出功率大、长寿命超铀放射性废物的累积水平低、裂变产物对反应堆反应性和能量增益影响小等优点。这些优点在利用U－Pu燃料循环的次临界堆中十分明显。对于利用Th-U燃料循环的次临界堆,热堆和快堆都是可以工作的;而对于U－Pu燃料循环的系统,快堆则是较好的选择。     

行波堆平准化发电成本分析

本文通过平准化发电成本的方法,以燃料循环作为研究对象,对行波堆一次通过式燃料循环和二次通过式燃料循环的经济性进行了研究,并选取10个重要的经济和技术参数进行成本敏感性分析。研究结果表明,行波堆的平准化发电成本低于现有压水堆和快堆,其中,行波堆一次通过式燃料循环方式的平准化发电成本最低。敏感性分析表明,贴现率、燃耗深度、隔夜价和反应堆热效率是影响行波堆经济性最重要的参数,而燃料价格和废物处置的价格由于占成本的比例较小,对行波堆经济性的影响不大。     

The potential of the thorium cycle in PWRs

This parametric study has been made to determine the optimum moderator to fuel volume ratio, pin diameter and burnup of thorium fuel in PWRs. Under optimum conditions a substantial reduction in uranium requirements can be obtained without adversely affecting fuel cycle costs. The development of the thorium cycle in light water reactors forms an alternative to the LMFBR development.     

Fuel element design for the 300 MW(e) gas-cooled fast breeder reactor

The fuel element design for a 300 MW(e) gas cooled fast breeder reactor (GCFR) is presented. The design is the result of a program sponsored by Kernforschungsanlage, Julich (KFA) to develop and fabricate a full size fuel element model under extension of an agreement between General Atomic (GA), Kraftwerk Union (KWU), and KFA to exchange information from GCFR irradiation experiments. The resulting fuel element model design was achieved by joint participation between GA and KWU and relies on the experience and knowledge of the two companies. The model, which will be manufactured by KWU using prototypical materials and specifications, except for dummy fuel pellets, will establish manufacturing feasibility and identify areas for future cost reduction improvements. The evolved designs, particularly the fuel rods, are very similar to those employed in the liquid metal fast breeder reactor (LMFBR) programs. These similarities enable the GCFR to use the vast amount of data being generated for the LMFBR programs, with only an incremental development plan needed to verify certain unique features inherent to the use of helium as the primary coolant.     

Thermal performances and melting risk assessment in a LMFBR fuel pin

A reliable evaluation of fuel temperature is a key safety requirement in the design of the fuel assembly of a nuclear reactor, especially in the case of a LMFBR whose efficient operation requires high thermal performance fuel.The physico-chemical properties such as density, oxygen to metal ratio and thermal conductivity of a typical LMFBR mixed-oxide fuel, which are known to change in a remarkable way under irradiation, strongly affect the temperature profile within the fuel pellet.A statistical analysis of the temperature values in the fuel of the Italian Fast Reactor PEC, has been performed by means of the RSM code (Response Surface Methodology) coupled to a Monte-Carlo Technique (MUP code), in order to demonstrate that the melting risk is substantially negligible.     

钠冷快堆分析程序ATHAS-LMR的子通道模型

以子通道模型和绕丝分布式阻力模型为基础,研发了液态金属快中子增殖堆热工水力子通道分析程序ATHAS-LMR,以对液态金属快中子增殖堆燃料组件中的热工水力现象进行分析。与国外知名实验和类似子通道分析程序比较,结果表明：ATHAS-LMR与实验结果及其他子通道分析程序的结果相近,能够完成包括堵流工况的各种工况下液态金属快中子增殖堆组件的热工水力性能分析。     

Analytical Method of Fuel Pin Deformations in LMFBR Assembly

A new analytical method is presented for analyzing, in three dimensions, the mechanical response of fuel pins with wire spacers, to their thermal and neutronic environment in a fuel assembly of a LMFBR. It analyzes the mechanical interactions between fuel pins in the assembly in each of three directions, which form an angle of π/3 radians with one another, based on the mathematical relationship between the displacements at the contact points and the associated contact forces with respect to all fuel pins forming a line in one of the three directions.

Based on this method, a new computational code, the Subchannel Deformation Analysis Code for Wire-Wrap Assemblies (SHADOW) has been developed, and is applied to a fuel assembly of a prototype fast breeder reactor in order to analyze the deformation of 169 fuel pins due to thermal bowing.

Conclusions drawn from the study confirm that the SHADOW code can be an effective tool for analyzing or evaluating thermal and structural designs of a LMFBR fuel assembly.     

Simulation of the transient behaviour of LMEBR fuel pins under consideration of special burnup phenomena using the BREDA-II model

A new fuel pin model was developed to describe the influence of specific burnup phenomena on the behaviour of fuel pins under transient overpower conditions in a liquid metal fast breeder reactor (LMFBR). It has been used for transient fuel pin deformation analysis during hypothetical core disruptive accidents (HCDA) and for the purpose of interpreting fuel pin failure tests. The fuel pin model, designated as BREDA-II, is based on the equations of the quasi-static theory of thermal elasticity. The fuel is regarded as elastic and the cladding as elasto-plastic material. The equations for the stress-strain analysis are based on the plane strain approximation. A multiregion fuel pin model allows to simulate long-time and transient burnup phenomena. The long-time effects taken into account are the steady state swelling of fuel, the change in fuel porosity and the production and partial release of fission gases. During a power excursion transient fuel swelling and pressure increase due to transient fission gas behaviour are included in the deformation analysis. Potential fuel pin failure is indicated by the application of various criteria of failure. In subsequent model calculations the behaviour of an irradiated LMFBR fuel pin during an overpower transient corresponding to a reactivity ramp of $5/sec is simulated and interpreted from the point of view of reactor safety.     

Studies of core disruptive accidents and licensing aspects of fast breeder reactors

This paper discusses the role of the core disruptive accident (CDA) in the safety evaluations and licensing of Liquid Metal Fast Breeder Reactors (LMFBR). Parametric studies of transient overpower (TOP) accidents based on calculations for SNR-300 using the HOPE computer code are presented. Major uncertainties in TOP analysis are identified and discussed with emphasis on the need for reliable fuel failure criteria. A series of calculations illustrating the possible behavior of the U.S. LMFBR demonstration plant following a loss-of-flow (LOF) accident without scram using the SAS-IIIA computer code are described. It is shown that for a beginning of life (BOL) core and end of equilibrium cycle (EOEC) core, the reactivity effects from sodium voiding and clad motion can lead to further sustained reactivity additions from subsequent fuel motion and FCI driven sodium voiding. In these calculations we have used the fuel enthalpy criterion which predicts clad failure around the core midplane. For the EOEC case these effects can add sufficient reactivity to take the system above prompt-critical (LOF driven TOP) and into hydrodynamic disassembly. For the BOL case the sodium void may not be sufficient to bring the system near sustained prompt-critical. However, clad motion appears to be effective in raising the reactivity to prompt-criticality. These results are based on clad failure dynamics modeling in SAS-IIIA. Further work is needed in the area of fuel-clad behavior under severe transients before definitive conclusions can be drawn regarding the applicability of current clad failure models at high clad temperatures (>1000°C). The potential significance of a new concept in CDA analysis called the “transition phase” is briefly mentioned.     

压水堆核燃料循环的经济分析程序介绍

所开发的压水堆核燃料循环分析计算机经济程序包括12个子程序,代表着压水堆整个核燃料循环各种不同的工艺过程。本程序能算出压水堆核电站核燃料循环中燃料费用对发电成本的影响,给出各工艺过程对燃料成本的敏感度分析,并尽可能给出燃料循环中燃料材料及服务的价格数据.     

Hydrodynamics of post-disassembly fuel expansion

The early expansion of the fuel following disassembly in an LMFBR core disruptive accident is modeled. Spherical expansion in the sodium is assumed. A Lagrangian, finite-difference hydrodynamic code (FEXPAN) describes the motion. Disassembly employs VENUS-II, and a consistent equation of state for fuel was used throughout disassembly and FEXPAN. Time-dependent mechanical work and fuel vaporized without fuel mixing are obtained. FEXPAN is compared with time-independent expansion for the effect of fuel mixing. For example, for a particular accident analysed 75 MW sec mechanical work was calculated for expansion with no mixing versus 10 MW sec with complete core mixing.