Development of Multi-Functional Telerobotic Systems for Reactor Dismantlement

This report summarizes technological features of advanced telerobotic systems for reactor dismantling application developed at the Japan Atomic Energy Research Institute. Taking into consideration the special environmental conditions in reactor dismantling, major effort was made to develop multifunctional telerobotic system of high reliability which can be used to perform various complex tasks in an unstructured environment and operated in an easy and flexible manner.

The system development was carried out through constructing three systems in succession; a light-duty and a heavy-duty system as a prototype system for engineering test in cold environment, and a demonstration system for practical on-site application to dismantling highly radioactive reactor internals of an experimental boiling water reactor JPDR (Japan Power Demonstration Reactor). Each system was equipped with one or two amphibious manipulators which can be operated in either a push-button manual, a bilateral master-slave, a teach-and- playback or a programmed control mode. Different scheme was adopted in each system at designing the manipulator, transporter and man-machine interface so as to compare their advantages and disadvantages.

According to the JPDR decommissioning program, the demonstration system was successfully operated to dismantle a portion of the radioactive reactor internals of the JPDR, which used underwater plasma arc cutting method and proved the usefulness of the multi-functional telerobotic system for reducing the occupational hazards and enhancing the work efficiency in the course of dismantling highly radioactive reactor components.     

Developing Integrated Decommissioning Information Management System (IDIMS) of Nuclear Facilities

In Taiwan, the Taiwan Research Reactor (TRR) was shut down in January 1988, and a few nuclear facilities were accompanied to stop operation within Institute of Nuclear Energy Research (INER). For public health and safety reasons, INER dismantled step by step its expired nuclear facilities. Integrated Decommissioning Information Management System (IDIMS) was developed to ensure safety of dismantling and to record all activity data during the decommissioning project. These recorded activity data range from data of planning, licensing, post-operation to those of radioactive waste management and storage. In addition, IDIMS was expected to preserve decommissioning knowledge using information technology from practical data and problem solving. It also is anticipated that IDIMS will be an important knowledge repository and design base for decommissioning projects of nuclear power plants in Taiwan.     

重水研究堆退役废物再利用研究

实现废物再利用是废物最小化的重要措施之一,从废物流中将有潜在利用价值的物料分离出来实现再利用可大幅减少对环境的影响。本文以中国原子能科学研究院重水研究堆退役为实例研究了放射性废物再利用问题。通过全面分析和计算重水研究堆在退役期间产生的各类废物,得出具有一定数量的物料有潜在的利用价值,可直接或经适当处理后再利用在其他行业领域中。研究表明,通过采取废物最小化控制措施（如废物分类和废物流分离等）,采用适当的去污技术和执行清洁解控要求,至少可使重水研究堆退役过程中产生的几十吨钢铁、10 t铝材和5 t重水实现再利用。     

Verification of Source Term Estimation Method against Measured Data for Decommissioning Waste from a CANDU Reactor

The method for the establishment of an equilibrium core model proposed in the previous paper and the source term calculation method proposed in this paper for the characterization of decommissioning waste were verified by comparing the nuclide inventory estimated by MCNP/ORIGEN2 simulations with the measured nuclide inventory according to a chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. At first, the time-average pseudoequilibrium full-core model of Wolsong Unit 1 was developed on the basis of the previously proposed modeling method for the activation of in-core and ex-core structural components. Then, the application level of the neutron flux and cross section in the radionuclide buildup calculation were compromised. Fourteen major actinides and fission products were considered to represent the irradiated fuel condition, and a geometry simplification was also introduced in the burned full-core model for MCNP simulation. The assumption of a constant neutron flux and capture cross section as a function of the irradiation time was applied in the radionuclide buildup calculation in ORIGEN2. As a result, the values estimated from the analysis system agreed with the measured data within a difference range of 30%. Therefore, it was found that the MCNP/ORIGEN system and source term characterization method proposed can be viable to estimate the source terms of the decommissioning waste from a CANDU reactor.     

Application of Stable Adaptive Schemes to Nuclear Reactor Systems, (I)

Parameter identification and adaptive control schemes are presented for a point reactor with internal feedbacks which lead to the nonlinearity of the overall system. Both are shown stable with new representation of the system, which corresponds to the nonminimal system representation, in the vein of the Model Reference Adaptive System (MRAS) via the Lyapunov's method. For the sake of the parameter identification, model parameters can be adjusted adaptively as soon as measurements start, while plant parameters can also adaptively be compensated through control input to reduce the output error between the model and the plant for the case of the adaptive control.

In the case of the adaptive control, control schemes are presented for two cases, the case of the unknown decay constant of the delayed neutron and the case of the known constant. The adaptive control scheme for the latter case is shown extremely simpler than that for the former. Furthermore, when plant parameters vary slowly with time, computer simulations show that the proposed adaptive control scheme works satisfactorily enough to stabilize an unstable reactor and that it does even in the noise with small variance.     

Application of Stable Adaptive Schemes to Nuclear Reactor Systems, (II)

The parameter identification and adaptive control schemes applied in a previous study to a nonlinear point reactor are extended to the case of a loosely-coupled-core reactor with internal feedbacks, constituting a nonlinear overall system. Both schemes are shown to be stable, with the system newly represented on the pattern of the Model Reference Adaptive System (MRAS) with use made of the Lyapunov's method.

For either parameter identification or adaptive control of a loosely-coupled-core reactor, there exists no canonical form of multiple input-multiple output system which can be directly applied for deriving the MRAS with the matrix version of the Kalman-Yakubovich lemma as it was in the case of the point reactor. This difficulty is circumvented by the practical assumption that the neutron density can be directly measured on each core as reactivity change is applied as input into the coupled core as a whole. For parameter identification, the model parameters are adaptively adjusted to those of each core, while for the adaptive control, plant parameters of each core can be adaptively compensated, again through control inputs, to asymptotically reduce the output error between the model and the plant.

The point reactor is shown to correspond to a special case.     

Development of Chemical Decontamination Process with Sulfuric Acid-Cerium (IV) for Decommissioning

The dissolution of Cr-rich oxides in and the decontamination of contaminated specimens by sulfuric acid-Ce(IV) (SC) solution have been investigated. The results have been com-pared with those by the reported typical decontamination reagents. The SC solution can dissolve simultaneously not only Cr but also Fe and Ni in the Cr-rich oxides. The dissolution mainly depends on Ce(IV) concentration (1.3–10^?3 M) and temperature (40–90°C), but is independent of H₂SO₄-concentration (0.2–0.5 M at 5 x 10^?3M Ce(IV)). The DF is ca. 33 at 90°C for 24 h with 0.25 M H₂SO₄-5 x 10^?3M Ce(IV). Under electrolytic regeneration of Ce(IV) from Ce(III) during the decontamination, the DF is raised to 460 even at 80°C for 24 h with 0.25 M H₂SO₄-1.3 x 10^?3M Ce(IV). The SC solution can be applicable to the decontamination for decommissioning of both BWR and PWR. Fcrthermore the combination of the chemical cleaning with an additional ultrasonic cleaning provides a promissing cleaning procedure for post-disassembly components.     

Development of Decommissioning Engineering Support System (DEXUS of the Fugen Nuclear Power Station

The Fugen Nuclear Power Station (NPS) was shut down permanently in March 2003, and preparatory activities are underway to decommission the Fugen NPS.

An engineering system to support the decommissioning is being developed to create a dismantling plan using state-of-the-art software such as 3-dimensional computer aided design (3D-CAD) and virtual reality (VR). In particular, an exposure dose evaluation system using VR has been developed and tested.

The total system can be used to quantify radioactive waste, to visualize radioactive inventory, to simulate the dismantling plan, to evaluate workload in radiation environments and to optimize the decommissioning plan. The system will also be useful for educating and training workers and for gaining public acceptance.     

10 MW高温气冷实验堆燃料元件装卸系统研制

根据国际上类似系统的设计和运行经验，彻底改进了10MW高温气冷实验堆(HTR-10)燃料元件装卸系统的单列器、碎球分离器、提升器、控制系统，以使系统变得更为简单、可靠。改进后的设备均在全尺寸实验装置上进行过试验。系统的调试试验和初装料运行表明：该系统的性能满足HTR-10的要求。     

CARR工艺运输系统设计

介绍了中国先进研究堆(CARR)工艺运输系统的功能、工艺流程及设备分级,并对工艺运输系统的重要设备如装卸料机、双工位抓具、斜孔道和干法贮存室进行了结构描述。工艺运输系统的重要设备装卸料机已通过出厂验收,达到了功能要求,证明设计是合理的。     

钠冷快堆系统程序FR-Sdaso开发

为解决600 MW示范快堆（CFR600）事故分析和工况设计中的实际问题,自主开发了钠冷快堆系统程序FR-Sdaso,其建模范围包括堆芯、一回路、二回路、三回路、四回路和事故余热排出系统,主要物理模型包括点堆模型、单通道堆芯热工模型、多区钠池模型、四区蒸汽发生器模型等核岛设备或部件分析模型,汽轮机、凝汽器、给水加热器、除氧器等常规岛设备采用集总参数模型,泵、阀门、管道及控制体等采用通用模型。对程序进行了初步验证,结果表明,FR-Sdaso程序可用于分析全厂瞬态工况及超功率、失流、失热阱等典型事故过程。目前,FR-Sdaso程序已用于CFR600的设计和安全分析。     

Development of System Code FR-Sdaso for Sodium cooled Fast Reactor

To solve actual problems in the accident analysis and working condition design of the 600 MW demenstration fast reactor (CFR600), the sodium-cooled fast reactor (SFR) system code FR-Sdaso was developed, which could be used to model the reactor core, primary system, secondary system, tertiary system, quadruple system and the decay heat removal system of the SFR. The physical models can be divided into three categories: The models for nuclear island equipment including point reactor model, single-channel core thermal model, multi-zone sodium pool model and four-zone steam generator model, etc., the lump parameter models for conventional island equipment, including turbine, condenser, feed water heater, deaerator, etc., and the general models for pump, valve, pipe and control volume. Preliminary V&V work for FR-Sdaso was conducted, and the results show that FR-Sdaso can be used to analyze the transient conditions of the whole plant and typical SFR accidents such as overpower, loss of flow, and loss of heat sink. FR-Sdaso was used in the design and safety analysis of the CFR600.     

熔盐堆热管式非能动余热排出系统建模及程序开发

热管式非能动余热排出系统（HP-PRHRS）概念设计可有效提升熔盐堆非能动安全特性。基于HP-PRHRS结构和熔盐堆运行特点,建立了一套较为完整的数学物理模型,涵盖了熔盐堆堆芯物理热工耦合、高温热管和HP-PRHRS运行等。采用上述模型开发了HP-PRHRS分析程序PRAC,利用MSRE基准题和瞬态实验数据进行了对比验证。结果表明：PRAC程序计算值与基准题和实验结果吻合良好,证明了模型和程序的准确性。HP-PRHRS模型和PRAC程序能为后续开展HP-PRHRS深入设计提供模型和软件基础。     

Package Flow Models by Neural Network Representation for Understanding the Dynamic Behavior of Nuclear Reactor Systems

“Package Flow Model” (PFM) is a simple simulation model for intuitive understanding of various types of system dynamics. In the previous papers, the PFM was proposed and its application to the dynamic analysis of nuclear reactor systems was presented. In the present paper, the same model and same application are considered but a new representation method of the PFMs by a neural network is introduced, so that the dynamic simulation of the reactor subsystem can be performed through the calculation of corresponding neural network. Furthermore, the quasi optimum parameter values of each PFM are easily obtained by applying appropriate learning algorithm to get weight-values of the neural network.

Some case studies show that the learning process and the obtained optimum values can give us new useful information on approximate understanding of the dynamic behavior of actual processes in the system.     

池式快堆系统瞬态分析软件开发

为实现快堆系统分析软件国产化,在已开发的适用于稳态计算的池式快堆系统分析软件SAC-CFR的基础上,进一步开发了系统各部件的瞬态模型、控制系统和保护系统模型、瞬态工况热工水力学的求解逻辑,完成瞬态计算功能的开发。通过对日本文殊快堆45%功率汽机跳闸工况进行建模分析,验证了SAC-CFR用于系统瞬态分析的有效性,为进一步开发非能动余热排出系统分析模型打下了基础。     

Modeling and Code Development of HP-PRHRS for Molten Salt Reactor

The conceptual design of heat pipe cooled passive residual heat removal system (HP-PRHRS) was proposed to improve passive safety performance of molten salt reactor (MSR). Based on the structure of HP-PRHRS and the operation performance of MSR, a set of reasonable mathematical physical models were built, mainly including reactor core physical thermal model, high temperature heat pipe model and HP-PRHRS model. Analysis code PRAC for HP-PRHRS was developed for MSR adopting those models. The verification of the code was conducted using MSRE benchmark and the transient experimental data. The results show that the calculated value of PRAC code is in good agreement with the benchmark and experimental results, which proves the accuracy of the model and code. HP-PRHRS model and PRAC code can support and provide foundation for the future research on MSR.     

铅基快堆氧化物燃料性能分析程序FUTURE开发与初步验证

为了对铅基快堆氧化物燃料元件稳态工况下的服役性能和行为演化进行模拟计算,本文基于串行的半隐式耦合求解方法开发了铅基快堆氧化物燃料性能分析程序FUTURE。程序采用两步分析法实现了铅基快堆氧化物燃料棒全域热力分析与局部行为模型的多物理场耦合计算。通过各计算模块与模型算例、基准公式和现有程序的对比分析,对FUTURE程序进行了各分离效应的初步验证。结果表明,FUTURE程序能准确模拟铅基快堆稳态工况条件下氧化物燃料元件内部的温度演化、结构变形、应力分布和相互作用,并实现对燃料重构、氧和钚元素的迁移、裂变气体释放和服役期内液态铅铋腐蚀等内容的计算模拟。     

核动力堆热功率跟踪系统的自校正模型算法控制

针对核动力堆控制系统这一带不确定参数及干扰的复杂非线性系统 ,本工作用自校正模型算法控制原理提出了一种动力堆热功率跟踪控制的方法。该方法克服了一般多步模型算法控制律所存在的由于模型估计不准或模型参数大幅度变化引起系统动态特性和控制质量变坏的缺点。通过仿真 ,将本方法与一般多步模型算法控制相比较 ,结果证明了该方法可使系统具有更强的鲁棒性和更好的动态品质。     

200MW核供热堆功率调节系统设计原理

简要介绍了２００ＭＷ核供热堆的堆型结构和运动方式，着重介绍了２００ＭＷ核供热堆功率调节系统设计原理和调节控制棒与调节二回路流量协调控制方案，并介绍了调节二回路流理所采用调频电源作为二回路水泵调速的执行机构。