Burning plutonium and minimizing radioactive waste in existing PWRs

Plutonium and other long-lived radioactive actinides are produced in light-water reactors (LWRs) using conventional fuel. “Innovative” fuel matrices may reduce the breeding of these nuclides. However, essential LWR safety features have to be preserved, which restricts the possibilities for new fuel-carrying matrices. Respective fuel assembly and LWR core safety studies indicate practicable new fuel options for the near future.     

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。     

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.     

Impact of partitioning and transmutation on the high level waste management

Nuclear energy generates 30% of the electricity in the EU. Still, different countries of EU27 have very different attitudes towards the future use of nuclear energy for electricity generation or other uses. However, independently of the political decision of continuation or phase out of the nuclear energy, all countries using nuclear energy to generate electricity are facing the question of the final management of its spent nuclear fuel and other high level radioactive wastes.Partition and Transmutation are emerging technologies that when integrated in advanced fuel cycles can strongly influence on the final wastes from the nuclear industry and its management. A review of the progress on the understanding of their real potentialities and main conclusions from a large number of international studies are presented in this paper. In particular, the conclusions from the main NEA working groups and the EURATOM RED-IMPACT project are jointly discussed.In this paper the emphasis is put on the effects of Partitioning and Transmutation on the inventory reduction, the heat source reduction and its implications to the repository capacity and on the performance assessment of the final repository.     

Safety assessment aspects of type B(U) packages containing wet intermediate level waste

Abstract

In Germany, the mechanical and thermal safety assessment of approved packages for the transport of RAM is carried out by BAM as the competent authority according to the International Atomic Energy Agency regulations. BAM was involved in several approval procedures with ductile cast iron containers containing wet intermediate level waste. These contents, which are not dried, only drained, consist of saturated ion exchange resin and a small amount of free water. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific points. The physical and chemical compatibility of the content itself and of the content with materials of the package must be shown. From the mechanical resistance point of view, the package has to withstand the forces resulting from the freezing liquid. The most interesting point, however, is the pressure build-up inside the package due to vapourisation. This could be caused by radiolysis of the liquid and must be taken into account for the storage period. The paper deals primarily with the pressure build-up inside the package caused by the regulatory thermal test (30 min at 800°C) as part of the cumulative test scenario under accident conditions of transport. To determine the pressure, the temperature distribution in the content must be calculated for the whole period from the beginning of the thermal test until cooling down. In this case, calculating the temperature distribution requires, besides the consideration of conduction and heat radiation, consideration of evaporation and condensation including the associated processes of transport.     

External corrosion of cladding in PWRs

In-reactor corrosion of Zircaloy is strongly influenced by the fast neutron flux and water chemistry of the primary coolant. Under typical PWR coolant conditions with low oxygen content the fast neutron flux increases the corrosion rate only slightly. On the other hand, under fast neutron irradiation at a high oxygen content in the primary coolant the corrosion is accelerated 5- to 10-fold. In addition localized oxide lenses (nodular corrosion) have been observed. However, hydrogen pick-up rates were generally low. The results are discussed in view of life-limiting aspects; under normal operating conditions of a PWR the external corrosion is not life limiting.     

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.     

Drying medium level liquid waste concentrates

Experiments on drying simulated medium level liquid waste concentrates on the laboratory and technical scales produced, under the selected conditions (40–80, about 20–200 mbar), a dry salt cake (residual moisture less than 0.6%). Splashing of the drying product during the final phase of drying has been largely avoided by addition of small amounts of sodium tetraborate. During the whole test (heating, drying, secondary drying, cooling) no exothermic reactions took place.     

Numerical prediction of stratified pipe flows in PWRs

The phenomenon of thermal stratification has been analysed on the l'EXPRESS experimental facility representing the pressurizer surge line of a Framatome PWR. This experimental approach has allowed to characterize flow regimes for different operating conditions. A numerical simulation approach has been performed by the TRIO code. The measured fluid temperatures have been compared to calculated values. A first validation of the numerical simulation was realized by comparing steady state results to experimental values, the second one by comparing transient conditions. Also the stratification onset has been estimated and compared to the experiment. The numerical simulation has allowed to obtain a good prediction of the quantities representative of the thermal loading.     

Heat transfer in high-level waste management

Heat transfer in the storage of high-level liquid wastes, calcining of radioactive wastes, and storage of solidified wastes are discussed. Processing and storage experience at the Idaho Chemical Processing Plant are summarized for defense high-level wastes; heat transfer in power reactor high-level waste processing and storage is also discussed.     

Cement mortar-degraded spinney waste composite as a matrix for immobilizing some low and intermediate level radioactive wastes: Consistency under frost attack

The increasing amounts of spinning waste fibers generated from cotton fabrication are problematic subject. Simultaneous shortage in the landfill disposal space is also the most problem associated with dumping of these wastes. Cement mortar composite was developed by hydrating mortar components using the waste slurry obtained from wet oxidative degradation of these spinney wastes. The consistency of obtained composite was determined under freeze–thaw events. Frost resistance was assessed for the mortar composite specimens by evaluating its compressive strength, apparent porosity and mass loss at the end of each period of freeze–thaw up to 45 cycles. Scanning electron microscopy, infrared spectroscopy and X-ray diffraction analyses were performed for samples subjected to frost attack aiming at evaluating the cement mortar in the presence of degraded spinney waste. The cement mortar composite exhibits acceptable resistance and durability against the freeze–thaw treatment that could be chosen in radioactive waste management as immobilizing agent for some low and intermediate level radioactive wastes.     

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.     

Anomalous peculiarity of coolant temperature feedback in PWRs

According to conventional thinking on power regulation in a reactor with a negative feedback, the power change has the same sign as the introduced reactivity causing it. In this study we consider the opposite situation, which is anomalous in the sense of the usual power behavior. This anomaly connects with spatial peculiarity of coolant temperature reactivity effect and predisposition to it is inherited for reactors, which are sufficiently large in the direction of the coolant flow. The analysis is performed using simplified one-dimensional reactor models with one feedback on the coolant temperature. In the perpendicular (radial) direction reactor is assumed to be uniform and infinite.     

Integrated waste management in a small programme

The Finnish nuclear waste management programme consists of handling, intermediate storage and final disposal of the spent fuel and operating waste as well as decommissioning of nuclear power plants and disposal of waste from dismantling. There are two final repositories for low- and intermediate-level operating waste, one of which is in operation at Olkiluoto and the other will be commissioned at Loviisa in 1997. A new company, Posiva, takes care of the necessary research, and later, will oversee construction and operation of the disposal facility for spent fuel. The next main target of the programme for spent fuel disposal is selection of a site in the year 2000. Construction of the final disposal facility should start during the 2010s and operation begin around 2020.     

Experimental determination of subcriticality at subcritical PWRs in Korea

A direct subcriticality measurement system (SMS) based on the Feynman-α method has recently been developed by KEPRI. It was applied to six commercial pressurized water reactors in Korea. However, the obtained Feynman curves failed to give proper multiplication factors. The objective of these tests was to investigate the performance of the Feynman method to predict directly the subcriticality of a given subcritical reactor by using the neutron pulse counts only without any reactor perturbation in the large commercial reactors. Recently, two methods developed by Hokkaido University and Westinghouse Electric Corporation. These methods have a defect due to being based on the modified neutron source multiplication method. To overcome this defect and derive operational benefits is necessary to estimate the subcriticality of a subcritical core directly from the neutron pulse counts only. The performance of the developed SMS was verified in the Kyoto University Critical Assembly and applied to eight 1000 MWe Optimized Pressurized Water Reactors (OPR1000) in Korea. The obtained results show that the SMS based on the Feynman method can be a useful tool to estimate the reactivity of a subcritical power reactor. Although the discrimination level of the signal-processing unit in OPR1000 suffers from noise and gamma ray effects, SMS can provide good Feynman curves and effective multiplication factors. However, the SMS has failed to give the reactivity for the entire measured data set. Improving the SMS and investigating the effects of different discriminator level settings at SPU in OPR1000 will be topics for further study.     

压水堆核电厂稳压器压力模糊控制器研究及仿真

为提高核电厂稳压器压力的控制性能,针对其升、降压不同的被控特性,结合可以实时调整控制器参数的模糊自适应比例、积分、微分(PID)控制,设计出基于2个模糊控制器的压力控制系统。通过MATLAB/simulink仿真表明,采用模糊控制器的稳压器压力控制系统使控制性能得到了明显的改善。     

压力容器气腔内放射性惰性气体含量估算

核电站运行过程中产生的放射性惰性气体对电站辐射安全有着重要的意义。本文依托某压水堆核电站大修期间一起放射性气体泄漏事件,介绍压水堆核电站大修期间,稳压器开口后,压力容器开大盖前,压力容器气腔内放射性惰性气体含量估算方法。     

高放废物地质处置性能评价

为建立我国高放废物地质处置性能评价方法而系统地介绍了性能评价的研究目的、研究内容、研究方法、国内外研究现状；以此为基础，提出了关于开展我国性能评价的若干建议。性能评价方法的建立将有利于我国高放废物地质处置事业的协调发展。     

高放废物地质处置场址安全要求

概述了国际原子能机构、美国、法国、瑞典、芬兰和日本的高放废物地质处置场址安全要求的内容、现状和发展趋势。场址安全要求主要从水文地质、地球化学、岩石特征、气候变化、人类干扰等方面来阐述场址的有利条件、不利条件以及潜在的有利与不利条件。以系统科学为指导,初步探讨了我国场址安全要求的研究框架和实施步骤。