Improved analytical methodology for calculation assessment of material parameters in the nuclear installation decommissioning process

The process of nuclear installation decommissioning is, besides other features, characterized by production of large amount of various radioactive and non-radioactive materials or waste that have to be managed, taking into account its physical, chemical, toxic and radiological characteristics. Waste management is considered to be one of the key issues within the frame of the decommissioning process from the technological and also financial point of view. Because of that mentioned fact, the evaluation of costs and other parameters is necessary to be done as precise as possible in the decommissioning planning period. The calculation code OMEGA with its implemented module of integrated material flow, is suitable for the assessment and further optimization of the various decommissioning waste management scenarios considering the different input parameters.In the paper, the improved analytical methodology based on the identification of decommissioning materials, definition of detailed material streams, development of scenarios, calculation of output parameters and final optimization, is presented. The process of implementation of such methodology to the existing OMEGA material flow system, including the new or perspective technologies and methods for the waste managing, is also discussed more in details.Finally, the summarizing conclusions and recommendations resulting from the model calculation results, done for the verifying the suggested methodology and functionality of new improved material flow system of the OMEGA code, are presented.     

乏燃料后处理厂强放区域退役初始源项调查研究

随着退役治理专项工作的有序推进,我国早期乏燃料后处理等设施现已转入退役关键阶段,获得放射性特性数据等是退役前必须做的重要工作。本文首次依据我国遗留后处理厂退役初始源项调查科研任务,以工程现状、退役对源项的需求和测量技术基础作为出发点,确定了强放区域的调查原则、调查要求,通过系统设计,集成开发了以无损测量方法作为主要调查手段、面向在线工艺系统的放射性特性调查成套测量技术,为后处理厂强放区域退役奠定了源项基础。本文重点论述了总体设计中遇到的关键技术问题,以及如何运用这些技术解决问题。该方法的总体设计思路具有示范作用,可以作为设计复杂退役调查技术决策的重要依据。     

Systems approach as a tool for optimizing the technological disassembly processes in decommissioning nuclear power plants

VNIIAM Volgodonsk Affiliate of VNIIAM. Translated from Atomnaya énergiya, Vol. 77, No. 6, pp. 460–462, December, 1994.     

The calculation and analysis of the thermodynamic-cycle parameters for nuclear power stations

Methods are described for the calculation of the parameters and the absolute internal thermodynamic efficiency _i of a two-cycle nuclear power station, with a reactor using a gas for heat exchange, and two stages of steam pressure in the second cycle.The methods developed can be used for the complete calculation of a thermodynamic cycle on a digital computer.An analysis is carried out of the dependence of _i and other parameters on the quantities determining them.     

我国核设施退役治理标准化现状及建议

回顾了我国核设施退役治理标准化现状,总结了我国核设施退役治理标准存在的问题,对核设施退役治理标准规划和体系、核设施退役治理管理标准、技术标准以及标准的实施和监督等方面提出了标准化工作的建议。     

Structures and elements for the decommissioning cost estimations of nuclear research reactors

Decommissioning cost estimation is a very important technique when designing and planning a nuclear facilities’ decommissioning project. Decommissioning cost estimation should be made according to the phases of the decommissioning activities and the installed components of the nuclear facilities.     

Quantification model for estimating the radiological risks of decommissioning processes in nuclear facilities

This paper proposes a model for the quantification and estimating the radiological risks of decommissioning processes in nuclear facilities. Based on fuzzy linguistic variables, the membership function and inference rules were developed for quantifying the radiological risks of nuclear decommissioning processes. Also, the fuzzy inference system was developed and the proposed method was applied to the process of concrete decommissioning. The proposed model and system is flexible in that it allows a fast-computation of the subjective expert opinion when one or several input factors change. It is believed that the suggested model and system can be applied to evaluate the safety of complex systems by only changing the variable and inputs.     

核设施退役辐射检测及其仪器仪表应用

核设施退役辐射场检测技术在核设施退役的整个过程中起着非常重要的作用.较系统论述了核设施退役辐射检测的质量保证,最后从满足核设施退役工作的需要出发,评述了退役过程辐射检测仪器仪表的选择及其应用.指出国内辐射监测仪器仪表在满足核设施退役的需求上的较大差距,值得各有关方面关注.     

Failure experience as a tool for evaluation of nuclear structures

High reliability in nuclear structures is attainable only because previous failure experience with comparable installations can be factored into structural evaluations in such a way that unsatisfactory experience is forestalled. The use of corrective techniques to eliminate in-service faults in consumer products as a means of improving the characteristics of future products is a well-known quality control principle. This principle is only applicable to nuclear structures if the nature and causes of structural failure are fully understood.     

A study of the decommissioning procedure of an activated structure through an evaluation of the decommissioning cost for a research reactor

As decommissioning of a research reactor and a nuclear installation requires a long period of time from the decommissioning preparation work to the site remediation, the management of the data generated during the entire period of decommissioning is one of the most important tasks. In particular, the data obtained from research reactor decontamination and decommissioning activities can be important resources securing the safety and economic feasibility for other research reactor decommissioning. The owner of the research reactor and nuclear power plant need to submit decommissioning plan to the regulatory body at the starting stage of the research reactor and nuclear installation decommissioning project. The cost plan for decommissioning and the method for assessing the amount of exposure to protect workers must be stated in the decommissioning plan.This paper introduces the DES (Decommissioning Engineering System) that can be able to manage the data generated in the process of decommissioning of the TRIGA research reactor, to calculate an amount of waste, to evaluate decommissioning cost after deriving unit work productivity factor, and to predict the decommissioning process in advance. To verify the usability of this system and data integrity through connections among the unit systems, it describes the process to calculate the decommissioning cost using the data generated in dismantling an activated bio-shielding concrete in the TRIGA research reactor.As a result of the experiment to calculate the decommissioning cost with the TRIGA research reactor structure, it was found that the calculations were done precisely without flaw as the purpose of the experiment. Therefore, the DES can not only be used for other research reactors decommissioning, but also it is expected to be applied to other research reactors in the future. As a decommissioning cost between an activated concrete and a non-activated concrete according to the method of the dismantling procedure was significantly different, a study regarding the dismantling procedure needs more research.     

Concept of decommissioning nuclear plants in the USSR and the COMECON Member-Nations

Journal version of the report to the 5th All-Union Scientific Conference on Shielding Installations from Ionizing Radiation [in Russian], (Protvino, Institute of High-Energy Physics, September, 1989).     

Composition and structure of simulation models for evaluating decommissioning costs for nuclear power plant units

The methodological and practical approaches to realizing an iterative multiple-criterion analysis for evaluating the decommissioning costs of the power-generating units of nuclear power plants and the optimal structure of the analysis using information technologies are examined. The objective prerequisites which have been established and facilitate the development and use of decommissioning simulation models in practical work are analyzed.     

Identification and evaluation of facilitation techniques for decommissioning light water power reactors

This paper describes a study sponsored by the US Nuclear Regulatory Commission to identify practical techniques to facilitate the decommissioning of nuclear power generating facilities. The objectives of these “facilitation techniques” are to reduce public/occupational exposure and/or reduce volumes of radioactive waste generated during the decommissioning process.The paper presents the possible facilitation techniques identified during the study and discusses the corresponding facilitation of the decommissioning process. Techniques are categorized by their applicability of being implemented during the three stages of power reactor life: design/construction, operation, or decommissioning. Detailed cost-benefit analyses were performed for each technique to determine the anticipated exposure and/or radioactive waste reduction; the estimated cost for implementing each technique was then calculated. Finally, these techniques were ranked by their effectiveness to facilitate the decommissioning process.This study is a portion of the NRC's evaluation of decommissioning policy and supports the modification of regulations pertaining to the decommissioning process. The findings can be used by the utilities in the planning and establishment of the activities to ensure all objectives of decommissioning will be achieved.     

反应堆退役三维辐射场实时计算及可视化

本文基于DELMIA和VIRTOOLS平台开发的反应堆退役三维仿真原型系统,提出了仿真系统、数据库和计算内核既相互独立又集成统一的三维辐射场计算和可视化技术方案。利用点核积分算法建立了三维辐射场计算模型,得到了能量的对数与转换系数的多项式拟合公式,考虑了设备屏蔽和自吸收效应。采用VS语言和SQL server软件平台编制了三维辐射场计算程序,经验证,在关键点处的辐射水平计算值与测量值的比值小于10,并嵌入了仿真系统,实现了退役场景三维辐射场的实时计算和数据更新。提出了基于行走路径的人员受照剂量计算方法,并实现了可视化显示。     

Development of calculation methodology for estimation of radionuclide composition in wastes generated at Fukushima Daiichi nuclear power station

ABSTRACT

A calculation methodology for estimating the radionuclide composition in the wastes generated at the Fukushima Daiichi nuclear power station has been developed by constructing a skeleton overview of the distribution of radionuclides considering the material balance in the system and calculation flowcharts of the transportation of radionuclides into the wastes. The wastes have a distinctive feature that their level of contamination includes considerable uncertainties because the process behind the contamination with the radionuclides released from the damaged fuel during and after the accident is not yet fully understood. Here, the developed method can explicitly specify the intrinsic uncertainties as a band to be included in the estimated radionuclide composition in the wastes and can quantitatively describe the uncertainties by calibration using analytically measured data on actual waste samples collected at the site. Further studies to improve the quality of the calculation method, the introduction of a stochastic approach to describe uncertainties, and acquiring a quantitative understanding of the spatial distribution of radionuclides inside the reactor building are suggested as important steps toward reasonable and sustainable waste management as an integral part of the decommissioning of the Fukushima Daiichi nuclear power station.     

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct contact and thermal interaction of molten materials with coolant. The fragmented core materials form a sediment debris bed in the lower plenum. It is necessary to remove decay heat safely from this debris bed to achieve IVR. A simulation code to analyze the behavior of debris bed with decay heat was developed based on SIMMER-III code by implementing physical models, which simulate the interaction among solid particles in the bed. The code was validated by several experiments on the fluidization of particle bed by two-phase flow. These evaluation methodologies will serve as a basis for advanced safety assessment technology of SFRs in the future.     

Studies on the permeation of hydrogen and tritium in nuclear process heat installations

To realistically evaluate the important problems of hydrogen and tritium permeation in nuclear heated high temperature systems, estimates are made, among others, on the basis of the authors preliminary experimental data. Steam-methane reforming is used as the key process. The results show that oxide layers can decrease the hydrogen permeation rate by more than two orders of magnitude and that not only the oxidation potential and temperature but also the water partial pressure may be essential for the formation and possibly the structure of oxide layers, and consequently for the permeation rate. The consequences of the experimental data for the permeation of tritium are also discussed. The available empirical data and results of the measurements discussed here still contain large uncertainties. It will therefore be necessary to carry out, under as realistic conditions as possible, a broad parameter study of heat exchanger materials which are seriously considered for use in nuclear process heat installations.