FOCUS-type fuel assembly for PWRs

In recent years the development efforts for Siemens PWR fuel assemblies were mainly concentrated on reducing the fuel cycle costs and increasing the operational reliability of the fuel assemblies.The first objective was aimed at increasing the average discharge burnup to > 50 MWd/kgU and increasing the critical heat flux. The high envisaged burnup required to develop a corrosion resistant cladding tube outside the Zry-4 range. The decision was made to use a Duplex cladding tube consisting of a corrosion optimized outer layer on a Zry-4 base material. A ZrSnFeCr alloy with reduced tin content was chosen for the outer layer. The critical heat flux could be increased by introducing mixing vanes on the spacer grids within the active length.To reach the second objective, reliable avoidance of spacer grid damage during core loading and unloading and reduction of fuel rod defects by debris fretting, the spacer grid corners were improved and a debris separation grid was developed.These design improvements were introduced into the new FOCUS-type fuel assembly. The name FOCUS stands for “Fuel assembly with Optimized Cladding and Upgraded Structure”.     

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.     

大型压水堆堆芯燃料管理策略灵活性研究

本文对两环路大型压水堆开展燃料管理策略灵活性研究,设计12个月、18个月、16/20个月交替以及24个月换料的堆芯方案,建立并完成方案的安全性限值与燃料经济性评价,分析高功率高燃耗堆芯换料灵活性的关键限制条件。结果表明,延长循环长度可以采用提高换料富集度或者增加换料量以满足后备反应性的要求,但会增加燃料成本。为在燃料成本与电站收益间达到较好的平衡,须同时将提高卸料燃耗作为设计目标。在现有高燃耗性能燃料组件的技术条件下,通过在堆芯1/4~1/2范围内选取适当的换料量,可以实现12至24个月灵活换料,并具备较好燃料经济性。     

Thermal-hydraulics analysis during fuel element failure in an operating PWR

The fuel element failure in an operating pressurized water reactor (PWR), including fuel element breaks, has an effect on the operation safety of PWR. In this paper, the RELAP5 model of the fuel element failure is established for the safety analysis. The RELAP5 time step sensitivity analyses for the element pre-break steady and post-break transient simulation are carried out. And the variations of main thermal-hydraulics parameters related to the fuel element break are quantitatively studied, which include the internal gap pressure and the maximum fuel pellet temperature as well as the releasement of noncondensables in the gap. It is found that (1) the results by the RELAP5 code is very sensitive to the time step in a volume system with the noncondensables, and the time step sensitivity analysis is necessary if the effective time step range is unknown, (2) the larger the break area is, the more quickly the gap pressure increases and the maximum pellet temperature reaches to the stable value, (3) when the gap pressure increases and reaches to the coolant pressure, at the break the liquid inflow from coolant to gap will be turned to the vapor outflow from gap to coolant, (4) during the failure transient, the gap thermal conductivity experiences a sharp decrease in the break instant, which results in the decrease of heat transferred to cladding and the sharp decrease of cladding temperature as well as the sharp increase of minimum departure from nucleate boiling ratio (MDNBR). These conclusions can provide the basic for the operation safety analysis of PWR during the fuel element failure.     

Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead–bismuth, Pb–Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.     

Optimization layer by layer networks for in-core fuel management optimization computations in PWRs

The optimization layer by layer (OLL) learning algorithm is applied to prediction of assembly-wise power and burnup distribution, the critical soluble boron concentration, and the pin power peaking factor (PPPF) with core burnup in the pressurized water reactor (PWR) of the Korean Nuclear Unit (KNU) 11. It is shown that the OLL trained neural networks can predict core depletion characteristics as accurately as the high-precision modern nodal method codes and that the OLL trained neural networks can compute core depletion characteristics about 40 times faster than the modern nodal method code. The OLL networks are then utilized for determining the optimum fuel assembly (FA) loading pattern (LP) of the equilibrium cycle KNU 11 PWR core by a simulated annealing (SA) scheme. By demonstrating that the FA LP optimization by the SA scheme can be carried out within 10 to 15 min thanks to the speedy neutronics evaluation of the OLL networks, it is proposed that the OLL networks can make a satisfactory substitute for core evaluation codes based on modern nodal methods in in-core fuel management optimization computations where neutronics analysis for a large number of trial loading patterns has to be carried out.     

Small PWRs using coated particle fuel for district heating, PFPWR50

It has been said that nuclear energy is an important option for especially developing countries to satisfy their increasing energy demand. However, it will be difficult to deploy first of a kind nuclear power plant in developing countries because extensive safety demonstration has to be conducted in industrialized countries. On the other hand, it will be essential to present rigid proof of reliable operational experience to develop proper understanding of the safety features of new reactor systems among the people around the demonstration plant sites. One of the ways to solve the issue is to integrate existing technologies supported by a great deal of data and experience into a new reactor design. Based on the consideration, a small-sized district heating reactor system based on the pressurized water reactor (PWR) technologies combined with the fuel concept of high temperature gas cooled reactors (HTGRs) has been studied. The purpose of the combination of these two existing concepts is to take the best advantages of both excellent operational experience of PWRs and the integrity of HTGR fuel, coated particle fuel, against fission products release even at high temperature. We expect that this approach will help create a breakthrough to the current stagnation of nuclear power deployment.     

Operational feedback on internal structure vibration in 54 French PWRs during 300 fuel cycles

EDF has acquired extensive feedback on vibration of reactor vessel internals by analysing ex-core neutron noise on its 54 pressurized water reactors during the course of over 300 fuel cycles.

This feedback has been built up by processing more than 3,000 vibratory signatures acquired since the startup of its reactors. These signatures are now centralized for the whole of France in the “SINBAD” data base.

Signature processing has enabled:

1. • distinguishing between mechanical phenomena and signature variation linked to unit operation: in particular, the impact on signature level of unit operating parameters such as initial fuel enrichment and burn-up rate was assessed;

2. • among the purely mechanical phenomena, pointing up slight changes in position of vessel internals and the first signs of structural wear: relaxation (in the hold-down spring and fuel rod assemblies) and wear on surfaces of contact between internals and reactor vessel were detected;

3. • lastly and most importantly, automatic recognition of the various types of vibratory behavior of internals.

It was consequently possible to draw up user requirement specifications for automated monitoring of internals, which should soon be integrated in PSAD, a system which groups several reactor monitoring functions.     

A model for predicting pellet-cladding interaction-induced fuel rod failure

A model for predicting pellet-cladding mechanical-interaction-induced fuel rod failure is presented. Cladding failure is predicted by explicitly modelling the formation and propagation of radial cladding cracks by the use of non-linear fracture mechanics concepts in a finite element computational framework. The failure model is intended for implementation in finite element fuel performance codes in which local pellet-clad interaction is modelled. Crack initiation is supposed to take place at pre-existing cladding flaws, the size of which is estimated by simple probabilistic concepts, and the subsequent crack propagation is assumed to be due to either iodine-induced stress corrosion cracking or ductile fracture. The novelty of the outlined approach is that the development of cladding cracks which may ultimately lead to fuel rod failure can be treated as a dynamic and time-dependent process. The influence of complex or cyclic loading, ramp rates and material creep on the failure mechanism can thereby be investigated. The presented failure model has been incorporated in the ABB Atom transient fuel performance code. Numerical results from some applications of the code are used to illustrate the usefulness of the model.     

A study of risk reduction for some mitigation features for PWRs

The risk reduction attainable with mitigation features in a large-dry pressurized water nuclear reactor (PWR) is evaluated. The calculations are made in a probabilistic risk analysis framework, and they are based on Zion Probabilistic Safety Study (ZPSS). Some of the modifications made to this study are also taken into account.The mitigation designs considered consist of features for simultaneously controlling late containment overpressure, containment basemat penetration, and hydrogen burning. The individual mitigation features include: a passive containment heat removal system (PCHRS), a filtered-vented containment system (FVCS), a core ladle, and controlled hydrogen burning. Emphasis is placed on comparison of PCHRS and FVCS design options. The results include calculations of the sensitivity to several failure mode probabilities and to the probability of core meltdowns with containment bypass.     

A nondestructive method for discriminating MOX fuel from LEU fuel for safeguards purposes

Plutonium-rich mixed oxide fuel (MOX) is increasingly used in thermal reactors. However, spent MOX fuel could be a potential source of nuclear weapons material and a safeguards issue is therefore to determine whether a spent nuclear fuel assembly is of MOX type or of LEU (Low Enriched Uranium) type.     

An automated system for predicting nuclear fuel operating capabilities

Translated from Atomnaya Énergiya, Vol. 65, No. 3, pp. 163–169, September, 1988.     

对于非能动压水堆核安全监管要求的变化

本文对非能动压水堆核安全监管要求的变化作了具体的叙述和分析.13项重要的改变涉及:非安全级系统的监管处理、安全停堆状态、全厂断电法则、未能自动停堆的预计瞬态法则、安全参数显示系统问题、事故后取样系统、蒸汽发生器多管破裂、氢的控制、重新定义运行基准地震、现实放射性源项、安全壳C型试验的最大时间隔、关于非能动流体系统的单一故障以及ITAAC问题.     

压水堆高放长寿命裂变产物堆内嬗变平衡循环设计

基于压水堆多燃料循环管理计算,进行长寿命裂变产物（LLFP）核素堆内嬗变分析。基于长寿命裂变产物核素在乏燃料中的比重及核素的放射毒性,¹²⁹I和⁹⁹Tc作为当前嬗变研究的主要裂变产物。为避免碘同位素分离,参照乏燃料中¹²⁷I和¹²⁹I的组分比例,设计当前的碘化物嬗变靶件。将嬗变核素均匀弥散在惰性慢化材料ZrH₂中,放置在控制棒导向管内进行嬗变分析计算。基于该嬗变组件设计方案,对不同的换料方案进行评价和比较,进而搜索嬗变平衡循环。计算显示,当前带有靶件组件的布料方案可达到平衡循环,并能实现LLFP的嬗变。进一步嬗变优化方案设计受限于当前嬗变组件设计。     

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。     

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.