核设施退役过程中的辐射场重构与拆除路径优化

辐射场分布和路径选择是核设施退役过程中影响工作人员吸收剂量的主要因素。针对核设施退役过程中多源项的三维辐射场,采用点核积分的方法进行重构,重构结果与蒙特卡罗程序计算结果吻合很好,验证了点核积分法的可行性。将核退役拆除路径问题抽象为一种类旅行商问题数学模型,构造不同拆除路径下所受外照射剂量对应的剂量矩阵,根据辐射防护ALARA(As Low As Reasonably Achievable)原则,利用遗传算法进行寻优。对于多源项退役拆除实例,计算给出了最优化拆除路径和三维可视化显示,并对优化效果进行了讨论。设计开发了面向核设施退役过程的辐射场重构与拆除路径优化功能软件。     

Risk assessment on hazards for decommissioning safety of a nuclear facility

A decommissioning plan should be followed by a qualitative and quantitative safety assessment of it. The safety assessment of a decommissioning plan is applied to identify the potential (radiological and non-radiological) hazards and risks. Radiological and non-radiological hazards arise during decommissioning activities. The non-radiological or industrial hazards to which workers are subjected during a decommissioning and dismantling process may be greater than those experienced during an operational lifetime of a facility. Workers need to be protected by eliminating or reducing the radiological and non-radiological hazards that may arise during routine decommissioning activities and as well as during accidents. The risk assessment method was developed by using risk matrix and fuzzy inference logic, on the basis of the radiological and non-radiological hazards for a decommissioning safety of a nuclear facility. Fuzzy inference of radiological and non-radiological hazards performs a mapping from radiological and non-radiological hazards to risk matrix. Defuzzification of radiological and non-radiological hazards is the conversion of risk matrix and priorities to the maximum criterion method and the mean criterion method. In the end, a composite risk assessment methodology, to rank the risk level on radiological and non-radiological hazards of the decommissioning tasks and to prioritize on the risk level of the decommissioning tasks, by simultaneously combining radiological and non-radiological hazards, was developed.     

A qualitative identification and analysis of hazards,risks and operating procedures for a decommissioning safety assessment of a nuclear research reactor

Korea Research Reactor-1(KRR-1, TRIGA Mark-II type reactor), the first nuclear research reactor in Korea, is being prepared for a decommissioning. The decommissioning methods and procedures of KRR-1 ought to be based on its structural conditions and radiological characteristics. Also, a systematic approach to the decommissioning tasks must be followed by reviews and assessments of the decommissioning workers’ safety.     

清洁解控和退役若干动向与新发展

对国际辐射防护协会第 1 0届大会 ( IRPA-1 0 )涉及的清洁解控和退役问题作了论述 ,包括排除、豁免、清洁解控和废物最少化 ;退役工程技术的发展 ,包括去污技术、切割解体技术、探测技术 ;介绍了一个研究堆退役例子和加速器退役 ;最后 ,还论及了退役中受关注的一些问题 ,如 :石墨废物、混凝土废物、重水堆退役的氚防护、退役时间和退役废物量等。     

Dismantling design for the loop rooms on the MR reactor

The recently completed international co-operation project was aimed at planning for decommissioning the MR reactor identified as a pilot plant for the decommissioning of the other shutdown reactors on the site. The MR reactor was a pool-type, materials testing reactor with the total thermal power of 50 MW which incorporated pressure tubes containing fuel under test. The MR facility includes the reactor with its nine loop rig rooms containing pumps, heat exchangers and experimental equipment as well as systems and equipment located in other buildings in the complex. The objective of the MR reactor decommissioning project was to identify dismantling equipment and the decommissioning methodology for the reactor, loop rooms and redundant services to permit the refit and re-use of the building for a different nuclear related purpose. The dismantling design comprises two separate, but combined, tasks, namely, the dismantling of reactor installation itself and dismantling of experimental loops. The techniques proposed to undertake the dismantling operations within the loop rooms are described. Two options have been developed for removing contaminated equipment from the high radiation field loop rooms and packaging the waste into approved waste containers. The benefits and detriments of both methods have been identified, which allows implementing the safe, timely and cost-effective decommissioning.     

反应堆压力容器退役包装屏蔽方案初探北大核心CSCD

为保证核电反应堆压力容器安全退役,本文以国内最早运行的秦山一期反应堆压力容器源项为参考,模拟设计保持压力容器完整和切割压力容器两种包装屏蔽方案,通过估算两种方案下废物体积、包装成本、运输及处置成本,对比分析发现切割压力容器方案更佳。研究成果可为核电站退役工作提供支持。     

Assessment of the radiation impact of steam generator dismantling on the workers,public and environment

Prediction of the exposure of workers and the impact on the public and environment is necessary for the planning of the decommissioning tasks. Planning and realisation of the dismantling process have to take into account many factors. This results in the creation of possible dismantling scenarios. Moreover, the input data such as nuclide composition and activity content often vary. In the case of a steam generator, the contamination level can differ even within the same nuclear power plant. The paper describes and applies the methodology used for complex analysis of the steam generator dismantling process in nuclear power plants using the VVER-440 reactor types.     

Development of a digital mock-up system for selecting a decommissioning scenario

The evaluation of decommissioning scenarios is critical to the successful development and execution of a decommissioning project. In the past, many experts have used a physical mock-up system to find the exact work processes and the working positions. Nowadays, these jobs are being done by a Digital Mock-Up (DMU) system. The DMU, which is a technology to realize an effective work process by using virtual environments through representing the physical and logical schema and the behavior of a real decommissioning work, can save on the cost and time, reduce the risk of making later changes, and develop various decommissioning scenarios. In this research, a decommissioning DMU system was developed for simulating the relevant dismantling processes. Decommissioning data-computing modules which can calculate a dismantling schedule, quantify a radioactive waste, visualize a radioactive inventory, estimate a decommissioning cost, and estimate a worker’s exposure were also developed to qualitatively assess the decommissioning information. And an analytic hierarchy process (AHP) model was developed to evaluate the decommissioning scenarios which reflected the quantitative and qualitative considerations. To establish the proper scenario for the thermal column in KRR-1, the developed decommissioning DMU system was applied to evaluate the two candidate scenarios of it.     

Aspects on decommissioning of the Greifswald nuclear power plant

The most important part of the nuclear facilities in the former German Democratic Republic is situated at Greifswald near the Baltic Sea. Shortly after reunification of the Germany states, a decision was taken to decommission all Russian pressurised water reactors. The dismantling of them will be the biggest decommissioning project of series reactors world-wide.The low level of radioactive contamination, especially in the primary circuit, makes recycling of much material after decontamination possible.     

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.     

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.     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

Safety requirements for the design of large containers for decommissioning waste

About 50% of the total volume of conditioned radioactive waste from nuclear power generation will finally result from the decommissioning of nuclear power plants (NPPs). Higher activated metallic waste from the core region of the reactors would, according to the International Transport Regulations (IAEA), require a type B container. This would, however, bring a significant increase in the costs of the management of such waste. A considerably cheaper solution of this problem can be achieved by separating the protection requirements for type B packaging (i.e. mechanical integrity and tightness). By using industrial packaging (IP) designed for a higher mechanical integrity, it is possible to cope with higher IAEA protection goals for the safe transport of dismantling waste without strictly following the extremely high and also expensive tightness requirements for type B packaging. This is because the radioactivity is bound in the metallic lattice of such activated decommissioning waste.The above mentioned strong IP for decommissioning waste can also be used for other highly activated waste from the core region (e.g. fuel element boxes, control rods or other activated equipment) during the operation of the NPPs.     

Experience with the dismantling of three secondary steam generators in unit A in Gundremmingen by the ‘ice-sawing’ technique

The unit A in Gundremmingen (KRB A), was the first commercial nuclear power plant in Germany. It had an electrical power of 250 MW_e and was in operation from 1966–1977. The plant was equipped with a dual cycle boiling water reactor of a former General Electric design and includes three recirculation lines each with a big recirculation pump and a steam generator comparable with those of pressurized water reactors. Therefore dismantling experience is gained for systems and components of boiling water reactors as well as pressurized water reactors.

In early 1980, it was decided to decommission the plant. Actual decommissioning work started in 1983 with the removal of the components and systems in the turbine house. Since 1990 the decommissioning activities have been expanded to all primary water systems inside the reactor building. In 1992 , KRB A obtained a licence for dismantling the remaining activated components like the reactor pressure vessel and the biological shield.

Meanwhile more than 5200 tons of contaminated components have been dismantled. Special cutting and handling tools were tested, developed and optimized for the purpose of working in radiation fields and under water. The dismantling work of the contaminated systems and components ends up in about 6000 tons of material with a rather low amount of waste, especially due to optimized decontamination techniques Eickelpasch et al. (1992). For the dismantling of the three secondary steam generators in the reactor building the ‘ice-sawing’ technique was developed and patented.     

Decommissioning of the TVR Research Heavy-Water Reactor at the Institute of Theoretical and Experimental Physics

The key work on decommissioning of the TVR research heavy-water reactor at the Institute of Theoretical and Experimental Physics is described. The processes involved in the preparation for hauling away the spent nuclear fuel, mothballing the high-activity heavy water, salvaging the contaminated water in the holding pond, and removal of radioactive contaminants from the holding pond are described. Special attention is devoted to selecting a variant of decommissioning of the TVR and solving the problems associated with the implementation of decommissioning. The processes of dismantling the connecting pipes of the horizontal experimental channels inside the reactor vessel and preparing the fittings and equipment for performing this work are described.It is reported that work has been performed on the dismantling of the heavy- and light-water and gas loops of the reactor and on the dismantling of the experimental setups operating on the neutron beams of the horizontal and vertical channels of the reactor.     

Calculation of the effective delayed neutron fraction and the prompt neutron lifetime with the indigenously modified CEBIS code: MCEBIS

The CEBIS code has been modified to enable the calcination of both the effective delayed neutron fraction and prompt neutron generation lifetime in any nuclear thermal reactor, especially reactor types such as TRIGA, SLOWPOKE, and MNSR. The new version, called MCEBIS, includes sonie special subroutines which will be called up as part of the input to calculate the above two dynamic parameters. In addition, some control flags have been added to recognize any important reactor components such as beryllium as a reflector or heavy water as moderator and pence calculate their photo-neutron fractions.

The MCEBIS code has been tested using two reactor models: TRIGA and MNSR. These models were developed mainly to verify the modified code. Each model represents a 1-D neutronics model of the reactor. Calculated results for the effective delayed neutron fraction and prompt neutron generation lifetime in both reactors have been compared with published data. Good agreement with published results has been established.     

Working plan for decommissioning of the Lovisa Atomic Power Plant

Conclusion The option with dismantling of radioactive parts immediately after completion of power operation was adopted for the decommissioning of the Lovisa Atomic Power Plant. If the engineering life of the power units is 30 yr, the decommissioning of the first power unit will begin in 2008 and that of the second power unit in 2012. The entire period of decommissioning of the atomic power plant from the time of shutdown of the first power unit until the burial facilities are sealed and license obligations are discharged will last more than 12 yr.Careful analyses showed that the dismantling of radioactive parts of the power units is possible through the use of methods that have already been developed.According to estimates and calculations, decommissioning operations for the power units of the atomic power plant will require approximately 3000 man-yr.Waste from dismantling can be buried safely in facilities built in the bedrock at the plant site. The irradiation doses to the public due to the burial of such waste remain low.The collective irradiation dose to personnel engaged in dismantling is estimated to be 23 man-Sv.The costs for decommissioning the power units of the atomic power plant, as calculated from the data compiled, are 800 million Finnish marks.Translated from Atomnaya Énergiya, Vol. 67, No. 2, pp. 83–88, August, 1989.     

放射性废石墨的焚烧处理

石墨用作燃料套管、慢化剂和反射层的反应堆退役后产生大量的放射性废石墨,面临处理。焚烧作为研究较为深入的处理技术之一,可实现其大幅度减容,且产物具有较高的安全性和稳定性。本文介绍了目前典型的焚烧技术有固定床焚烧法、流化床焚烧法、激光焚烧法等,其中流化床焚烧技术在燃烧效率和技术成熟度上具有优势。     

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.     

Dismantling of the graphite pile of Latina NPP: Characterization and handling/removal equipment for single brick or multi-bricks

This work describes the issues related the dismantling of graphite piles of the 1st generation gas cooled reactor of Latina NPP (Italy).The retrieval of the graphite is a strategic matter for the decommissioning of this type of plant: in this study were described and analysed the current approaches used to access the core and to perform the remote and dry extraction of graphite bricks from the top.Based on these data, the removal of the graphite of Latina NPP will be planned; the extraction of the graphite will be carried out layer by layer by means of a dedicated remote controlled handling systems. This equipment will be duly designed according to the nuclear, physical and mechanical constraints of the graphite piles in core. In doing that the issues regarding the irradiated graphite have been also analysed by FEM code, especially those related to the core geometry and to the proposed technique of hooking the graphite bricks by a ‘gripper’ tool inside the axial channel.Data on fresh nuclear grade and irradiated graphite, used for the numerical simulations, were obtained by means of experimental tests, which were carried out on samples extracted from the reactor, and from theoretical models.The results obtained could support the final design of proper lifting and gripper tools and handling equipment, for single brick or multi-bricks, and to implement waste management strategy for the graphite.