Sensitivity Study of Fuel Cost in Extended Burnup BWR Core

A sensitivity study on the fuel cost of an extended burnup BWR core has been carried out on the basis of a realistic model of discharge burnup extension. Full power operating length in months in a refueling cycle and the number of refueling batches are chosen as independent variables in the model to describe extended burnup cores of various types. The reference core for the sensitivity study adopts 9-month full power operation and 4-batch refueling scheme. The difference in the plant cost between the extended burnup core and the reference core, which is referred to as plant capacity factor (PCF) credit, is estimated and combined with the fuel cost to calculate the fuel cost with PCF credit.

The results show that the fuel cost with PCF credit decreases for the extended burnup core with stretched operating length as the burnup extends in cases of constant non-operating length in a cycle, and that it may increase for the extended burnup core with decreased batch number in cases of constant plant capacity factor. It is also suggested that the cost minimum combination of the independent variables can be found to a given discharge burnup for the extended burnup core with decreased batch number in an intermediate case between these two extreme cases. Extended burnup cores with fixed batch number tend to have a lower natural uranium requirement, but larger separative work requirement.

The economic break-even condition for the extended burnup core with decreased batch number is discussed based on the fraction of fixed part in the non-operating length, which is insensitive to the cycle length stretch-out.     

Thermal Resistance between Hexagonally Arrayed Channels

A new power-flattening method has been proposed for boiling water reactors (BWRs) which have an axially skewed power distribution caused by the void fraction distribution. In present BWRs, the skewed power distribution is avoided by using shallow control rods and/or axially distributed gadolinia fuel bundles. These means are effective for the axial power shape control, but perturb the self-power-flattening effect due to fuel burnup. The power-flattening method proposed here extensively utilizes this effect in the equilibrium cycle core. Based on this method, a new BWR core design and operating strategy, the WNS core concept, has been realized for reactor operation with no shallow control rod insertion and no fuel bundle shuffling. Studies of the WNS core has shown that the proposed power-flattening method has the potential to improve capacity factors, increase operating thermal margins and simplify reactor operations in comparison with current BWR cores.     

Optimization of BWR fuel lattice enrichment and gadolinia distribution using genetic algorithms and knowledge

An optimization methodology based on the Genetic Algorithms (GA) method was developed for the design of radial enrichment and gadolinia distributions for boiling water reactor (BWR) fuel lattices. The optimization algorithm was linked to the HELIOS code to evaluate the neutronic parameters included in the objective function. The goal is to search for a fuel lattice with the lowest average enrichment, which satisfy a reactivity target, a local power peaking factor (PPF), lower than a limit value, and an average gadolinia concentration target. The methodology was applied to the design of a 10 × 10 fuel lattice, which can be used in fuel assemblies currently used in the two BWRs operating at Mexico. The optimization process showed an excellent performance because it found forty lattice designs in which the worst one has a better neutronic performance than the reference lattice design. The main contribution of this study is the development of an efficient procedure for BWR fuel lattice design, using GA with an objective function (OF) which saves computing time because it does not require lattice burnup calculations.     

百万千瓦级环形MOX燃料堆芯设计

对环形UO2燃料及环形MOX燃料组件参数的计算方法进行了研究。设计了包含193盒环形UO2和MOX燃料组件的混合型长周期(18个月)堆芯方案。对设计的堆芯的重要物理参数进行了分析,并对各循环进行了燃耗计算。结果表明,装载约30%MOX组件的堆芯可在百万千瓦功率下实现长周期换料。堆芯从初装载可安全过渡到平衡循环,各循环的重要物理参数均满足设计要求,说明设计的堆芯及燃料管理方案是安全可行的。     

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.     

Azcaxalli: A system based on Ant Colony Optimization algorithms,applied to fuel reloads design in a Boiling Water Reactor

This paper presents some results of the implementation of several optimization algorithms based on ant colonies, applied to the fuel reload design in a Boiling Water Reactor. The system called Azcaxalli is constructed with the following algorithms: Ant Colony System, Ant System, Best–Worst Ant System and MAX–MIN Ant System. Azcaxalli starts with a random fuel reload. Ants move into reactor core channels according to the State Transition Rule in order to select two fuel assemblies into a 1/8 part of the reactor core and change positions between them. This rule takes into account pheromone trails and acquired knowledge. Acquired knowledge is obtained from load cycle values of fuel assemblies. Azcaxalli claim is to work in order to maximize the cycle length taking into account several safety parameters. Azcaxalli’s objective function involves thermal limits at the end of the cycle, cold shutdown margin at the beginning of the cycle and the neutron effective multiplication factor for a given cycle exposure. Those parameters are calculated by CM-PRESTO code. Through the Haling Principle is possible to calculate the end of the cycle. This system was applied to an equilibrium cycle of 18 months of Laguna Verde Nuclear Power Plant in Mexico. The results show that the system obtains fuel reloads with higher cycle lengths than the original fuel reload. Azcaxalli results are compared with genetic algorithms, tabu search and neural networks results.     

田湾核电站18个月换料燃料管理策略

为实现田湾核电站1、2号机组长周期换料项目,制定了过渡到18个月换料的燃料管理策略。参考俄罗斯核电站管理经验,长周期换料项目需采用TVS-2M新型燃料组件。在正式向长周期燃料循环过渡前通过TVS-2M先导组件运行验证了混合堆芯相容性。TVS-2M组件入堆替换AFA组件将分成两个阶段,即前两个过渡循环装入带有包覆层的TVS-2M组件,后继循环装入不带包覆层的TVS-2M组件。田湾核电站两台机组经历4个过渡循环,逐步延长运行时间,最终达到480 EFPD的循环长度。过渡循环和平衡循环均采用部分低泄漏堆芯装载,降低了对反应堆压力容器的中子辐照。田湾核电站18个月换料燃料管理策略提高了机组能力因子和经济性并具有灵活的循环长度。     

轻水堆含钆燃料棒稳态辐照行为分析

国内外的压水堆燃料组件最新设计中,广泛采用钆燃料（UO₂-Gd₂O₃）作为可燃毒物来控制初始反应性和展平堆芯功率分布。钆燃料棒的性能与普通燃料棒存在较大差异,本文利用燃料元件性能分析程序FRAPCON-3.5对BR3堆内含钆燃料棒性能进行计算,并与实验测量值进行比较。结果表明：FRAPCON-3.5对含钆燃料棒的计算结果与实验测量值符合较好;含钆燃料棒在辐照初期强化了燃料棒自屏效应,对燃料的径向功率分布影响显著;在平均功率密度相同的情况下,燃料中加入钆会导致热导率降低,芯块温度升高;钆含量不同,裂变气体释放及燃料和包壳的变形略有差异。     

Optimized gadolinia concepts for advanced in-core fuel management in PWRs

Utilities operating LWRs require fuel assemblies and in-core fuel management service, which ensure safe, flexible and cost-effective production of electricity. Because the reliability of the fuel has always been the most important requirement, advanced measures to minimize fuel cycle costs are receiving increasing attention in the light of the pressure on costs within the deregulated power generation markets. The role of in-core fuel management in supporting the goal to minimize fuel cycle costs consists in the development of more demanding core loading strategies, i.e. in the first place, more advanced low leakage loading patterns. A prerequisite for this type of loading pattern is the use of an optimized burnable absorber design. Gadolinia (Gd) as integrated burnable absorber is a very effective means for limiting the critical boron concentration and power peaking factors. Current development efforts for optimizing Gd-fuel are focused on the reduction of the inherent penalties of today's Gd-FA designs, i.e. reduced average fuel assembly (FA) enrichment and heavy metal content, as well as the residual reactivity binding. The most effective way to overcome these drawbacks is the reduction of the Gd₂O₃ concentration to values of ≈2 w/o, for which, according to recent measurements of the heat conductivity of modern Gd-fuels, the reduction of the fissile content in the Gd-rods is no longer necessary. Various feasibility studies have been performed to evaluate the consequences of FA designs with low Gd-concentrations (low-Gd designs) for Siemens PWRs and non-Siemens PWRs, for which more restrictive boundary conditions with respect to critical boron concentration and peaking factors have to be fulfilled. These studies, as well as operation experience of reactor cycles using low Gd-FA reload designs, confirm that the in-core fuel management can handle the different Gd burnout characteristics without problems. The economical benefits of low-Gd designs compared to conventional Gd designs are comparable to those achievable by distinctly more costly and complex alternatives, like the use of enriched gadolinia.     

Recommended strategy and limitations of burnable absorbers used in VVER fuel assemblies

There is an obvious effort to increase the burn up of used fuel assemblies in order to improve fuel utilization.A more effective operation can be realized by extending the fuel cycles or by increasing the number of reloadings.This change is nevertheless connected with increasing the uranium enrichment even above 5% of 235 U. Burnable absorbers are widely used to compensate for the positive reactivity of fresh fuel. With proper optimization, burnable absorbers decrease the reactivity excess at the beginning of the cycle, and they can help with stabilization of power distribution. This paper describes properties of several materials that can be used as burnable absorbers. The change in concentration or position of the pin with a burnable absorber in a fuel assembly was analyzed by the HELIOS transport lattice code. The multiplication factor and power peaking factor dependence on fuel burn up were used to evaluate the neutronic properties of burnable absorbers. The following four different materials are discussed in this paper: Gd_2O_3, IFBA, Er_2O_3,and Dy_2O_3.Gadolinium had the greatest influence on fuel characteristics. The number of pins with a burnable absorber was limited in the VVER-440 fuel assembly to six. In the VVER-1000 fuel assembly, 36 pins with a burnable absorber can be used as the assembly is larger. The erbium depletion rate was comparable with uranium burn up.Dysprosium had the largest parasitic absorption after depletion.     

百万千瓦级压水堆核电站长燃耗堆芯钆可燃毒物优化研究

对百万千瓦级参考核电站长燃耗堆芯（１８个月换料）采用的可燃毒物（钆）含量与堆芯燃料管理主要结果进行了分析研究。该研究采用先进的燃料管理程序系统,对不同可燃毒物含量和不同可燃毒物棒根数的燃料组件进行了计算,给出了组件无限增殖因子（ｋｉｎｆ）随燃耗的变化关系,据此对参考堆芯采用相同的装载进行了４种方案燃料管理计算。计算结果表明,对于堆芯燃料管理,采用低可燃毒物含量、含可燃毒物棒数多的装载方案明显优于高可燃毒物含量、含可燃毒物棒少的堆芯装载方案。     

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。      

Operation of VVR-SM with 36% Enrichment Uranium-Dioxide Fuel

VVR-SM after conversion to IRT-3M fuel assemblies with 36% fuel enrichment is operating at 10 MW with a core load of 18 fuel assemblies. The safety of the operation is ensured by neutrons and thermohydraulic calculations of the operating regimes of the core after each reloading of fuel assemblies. The experimental channels of nine fuel assemblies with flux density of neutrons with energy >0.821 MeV up to 1.1⋅10¹⁴ sec⁻¹⋅cm⁻² are used for producing ³²P and ³³P. The flux density of thermal neutrons in the channels of the beryllium reflector reaches 1.6⋅10¹⁴ sec⁻¹⋅cm⁻². Nine of the horizontal channels and the channel of the thermal column are used for fundamental and applied studies and for neutron-activation analysis. __________ Translated from Atomnaya Energiya, Vol. 99, No. 2, pp. 147–152, August 2005.     

Optimization of Axial Enrichment and Gadolinia Distributions for BWR Fuel under Control Rod Programming

The axial enrichment and gadolinia distributions of BWR (boiling water reactor) fuel are optimized under control rod programming. The objective of the problem is to minimize the average enrichment required to reach a planned EOC (end-of-cycle) with criticality condition and axial power peaking constraint.

A method of approximation programming is employed as the basis for the solution method. Resulting linear programming problem at each iteration step is solved by means of goal programming algorithm. The method is applied to the initial fuel for a typical BWR/5 represented by an axial one-dimensional core model

Two-region analysis leads to the conclusion that the core bottom should be depleted during the cycle so that the power shifts to the core top at EOC. The enrichment and gadolinia distributions are determined to maximize EOC power peaking within a limit. The optimal solution of a 24-region fuel with a power peaking limit of 1.4 saves 10.6% in uranium ore compared with a uniform fuel depleted with a Haling power shape. Half the saving comes from an optimal natural uranium blanket implementation.     

Expansion of the method of correlated temperatures to clad and fuel design analyses

In the thermal design of nuclear reactor cores, specified design limits (temperatures and linear power rating) should not be exceeded by the operating values of certain elements (coolant, clad and fuel). However, a certain number of channels or fuel pins could be permitted to exceed the specified limits without affecting the reactor's safety while still allowing reliable operation. An expansion of the method of correlated temperatures, developed for coolant temperature analysis, was performed to enable clad temperature and fuel centerline melting analyses for reactor core reliability studies. Since generation of random numbers is involved, calculational procedures, tailored to designer needs, were developed in order to reduce computational time. The method is applied to a typical LMFBR core and results are presented for various assumed clad and fuel design limits.     

Neutron-physical and thermohydraulic calculations of VVR-SM with high-and low-enrichment uranium fuel assemblies

The VVR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences of Uzbekistan is being converted from fuel assemblies with high-enrichment uranium (36% ²³⁵U) to fuel assemblies with low-enrichment uranium (19.7% ²³⁵U). During the conversion process consisting of nine cycles, the IRT-3M fuel assemblies with high-enrichment uranium, which are removed at the end of each cycle, will be replaced with IRT-4M fuel assemblies with low-enrichment uranium. This will require increasing the core size up to 20 fuel assemblies and increasing the power of the reactor to 11 MW. The methods used for and the results of neutron-physical calculations (burnup, power distribution, subcriticality), thermohydraulic analysis, and calculations of the kinetic parameters of a stable state are described for a core with high-enrichment uranium, a mixed core, and the first full core with low-enrichment uranium. __________ Translated from Atomnaya énergiya, Vol. 104, No. 5, pp. 269–273, May, 2008.     

核燃料组件无损检测探测系统设计

燃料组件是反应堆的核心部分,在高温、高压及强中子辐射场等复杂环境条件下,燃料棒中芯块会出现肿胀、变形甚至包壳破裂,严重威胁反应堆的安全运行。为了更好地了解燃料组件在反应堆内的变化,研究高燃耗的燃料组件中燃料棒的中心空洞形成和燃料棒的变形情况,高能X射线无损检测是一种有效的技术手段。由于辐照后核燃料组件自身具有强放射性,探测系统设计中必须考虑减弱燃料组件自身辐射对探测采集的影响,因此组件探测系统中探测器阵列及准直器的优化设计十分必要。经过建模及相关模拟计算,得到了探测器单元最佳尺寸,优化了后准直器的结构设计,为提高燃料组件无损检测系统重建图像的质量提供帮助。     

Improvement of LWR thermal margins by introducing thorium

The use of thorium in pressurized water reactor fuel assemblies is investigated in this paper. The novelty of the reported work is to study a fuel design primarily intended to control the excess of reactivity at beginning of life, and flatten the intra-assembly power distribution rather than converting fertile Th-232 into fissile U-233. The fuel assembly is a traditional 17 × 17 pressurized water reactor fuel design. The majority of the fuel pins contain a mixture of uranium and thorium oxides, while a few fuel pins contain a mixture between uranium and gadolinium oxides. The calculation were performed by two-dimensional transport calculations with the Studsvik Scandpower CASMO-4E code in order to determine the main neutronic properties of the new fuel design, compared with the traditional uranium-based fuel assembly containing gadolinium used as reference. The majority of the neutronic properties of the uranium-thorium-based fuel assembly were similar to the reference fuel assembly. The Doppler and the moderator temperature coefficients of reactivity were found to be appreciably more negative in the uranium-thorium-based design, but still within acceptable limits. One advantage of this new uranium-thorium-based design is a reduction of the pin peak power at beginning of life, because of smaller amount of gadolinium being used. This is important from an operational and safety viewpoint, since the margin to departure from nucleate boiling becomes larger. Consequently, this new type of thorium-based fuel assembly shows advantageous properties for use in power-uprated cores.     

Detection System for the Fuel Assembly Nondestructive Testing

Fuel assemblies are the central components of a reactor. The core fuel pellets in the fuel pins will swell and deform and the fuel cladding may even break under the complex environment of high temperature, high pressure and intense neutron radiation field, which threats the safety of the reactor. To better understand the changes in the behavior of the fuel assembly in the reactor and study the central void formations and deformations of fuel pins in fuel assemblies to high burn-up, high-energy X-ray non-destructive testing is an effective technical means. Irradiated nuclear fuel assembly has a strong radioactivity, it is necessary to optimize the design of the detector system and the collimator to reduce the effect from gamma rays emitted from the irradiated fuel assembly during detection system designing phase. Through modeling, estimating and optimization, the optimal size of the detector unit is obtained and the collimator design is optimized which can lay the foundation to improve the quality of the reconstructed images of the fuel assembly nondestructive system.     

Design of a gadolinia bearing mixed-oxide fuel assembly for pressurized water reactors

A study on neutronics design of a gadolinia (Gd₂O₃) bearing mixed-oxide (MOX) fuel assembly (MOX-UO₂ (Gd₂O₃) assembly) was performed for the purpose of suppressing the use of fresh lumped burnable poison rods (BPRs). The MOX-UO₂ (Gd₂O₃) assembly investigated consists of MOX and UO₂ (Gd₂O₃) fuel rods, which have already been verified through both fabrication and irradiation experiences. In all, 16 UO₂ (10 wt% Gd₂O₃) fuel rods are located at every corner and the peripheral region of the MOX-UO₂ (Gd₂O₃) assembly in order to reduce the power peaking of MOX fuel rods due to the thermal neutron inflow, and to reduce the reactivity penalty at the end of cycle (EOC). Since fresh BPRs are not expected to be inserted and UO₂ (Gd₂O₃) fuel rods are located at every corner of the assembly, the number of splits in plutonium (Pu) content can be only two, which is less than three splits required for a standard MOX assembly. Core characteristics of an equilibrium core loaded with MOX-UO₂ (Gd₂O₃) assemblies are evaluated and it is verified that adoption of the MOX-UO₂ (Gd₂O₃) assembly is effective to avoid the use of fresh BPRs with securing both the core safety and cycle length. The simplication of the splits in Pu content is also supposed to be beneficial, since it has the possibility of reduce MOX fuel fabrication costs.