Channel reactors as an integral part of nuclear power in Russia

The place and role of channel reactors in nuclear power in our country and the main measures for upgrading and improving the power generating units of nuclear power plants with RBMK reactors are described. It is shown that the risk indicators for serious damage to the core of power generating units with RBMK reactors are lower after upgrading and the corresponding IAEA criterion established for operating nuclear power plants. Upgrading and implementation of a service life extension program has made it possible to obtain licenses for continuing operation of power generating units with first-generation RBMK reactors and predicting a service life increase to 45 years. The characteristics of nuclear power plants with channel reactors with more highly developed internal and natural safety properties are shown in evolutionary designs of the power generating units MKéR-860,-1000, and-1500, which have protective shells and which meet all requirements for power generating units built today. It is shown that innovative solutions for the channel reactor concept can be implemented on the basis of the designs of power generating units with nuclear superheating of steam or on the basis of designs for developing reactors with supercritical parameters. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 29–36, July, 2007.     

Development and adoption of measures for increasing the safety of power-generating units of nuclear power plants with RBMK reactors in 1986–2005

The basic stages and directions of upgrading and reconstruction of the power-generating units of nuclear power plants with RBMK reactors over a period of 20 yr since the Chernobyl accident are reflected in this paper. The greatest amount of work was done in 1986–1990 on all units with RBMK-1000 and-1500 reactors which were operating at that time in the USSR. The purpose of the upgrade was to improve the neutron-physical characteristics of the reactor cores, increase the response speed of the safety and control system and decrease the positive effect of water outflow from the cooling loop of this system, increase the flow capacity of the systems performing emergency discharge of the steam-gas mixture from the reactor, and improve the basic operating documentation. The subsequent stages of the upgrading and reconstruction concerned mainly the first-generation power-generating units (Nos. 1 and 2 units of the Leningrad and Kursk nuclear power plants). These works were performed to make the power-generating units conform to the requirements of the modern normative documents on safety and allow for the possibility of these units to remain in operation after the nominal 30-yr service life has been exhausted. __________ Translated from Atomnaya énergiya, Vol. 100, No. 4, pp. 312–320, April, 2006.     

Experience in operating and upgrading the No. 5 unit of the Novovoronezh nuclear power plant – practical base for developing a reliable source of nuclear energy

The No. 5 unit of the Novovoronezh nuclear power plant, starting commercial operations on September 26, 1980, is the first power-generating unit with a 1000 MW VVER in our country. The assimilation of its power gave invaluable experience to designers, builders, and equipment manufacturers; this experience was taken into account in the design solutions for next-generation power-generating units. A large volume of work on increasing the efficiency, reliability, and safety was performed over a 30-year service life. At present, the power-generating unit has been shut down for a major overhaul for upgrading according a program for extending the service life by 25–30 years.     

Probabilistic Service-Life Assessment for Safety Analysis for Preparing a Power-Generating Unit of a Nuclear Power Plant for Service-Life Extension

Information about the actual history of cyclic loading in the actual state of the metal in equipment components must be used for analyzing the safety of power-generating units in operating nuclear power plants, in preparation for service-life extension, and to determine the moment of onset of rapid aging of the metal and the end of the period of stable operation.The properties of the exponential distribution function and probability functions which are constructed for various loading scenarios using the hypothesis of probabilistic summation of fatigue damage for estimating the -percentage residual service life of equipment components are examined.Probabilistic estimates of the service life for operating nuclear power plants make it possible to control effectively the residual service life of the components of a nuclear power plant on the basis of information provided by the diagnostics systems and by the systems monitoring the state of the metal and the data on loading parameters from the control systems.     

Method of Integral Reliability Assessment of Equipment and Personnel of Long-Operating NPP Power Unit

Atomic Energy - A method of making an integral reliability assessment of the equipment and personnel for NPP with a long service life on the basis of an analysis of data from safety monitoring of...     

核电厂控制与保护系统动态仿真

分析了CHASHIMA核电站的测量系统、控制与保护模型、系统设备及设备失效模型、辅助系统管网模型。然后,基于C语言编制了控制与保护系统动态仿真程序模块PROSYS．并将其用于在工程模拟器,在模拟器上实现了CHASHIMA核电站控制与保护系统的动态仿真该工程模拟器已应用于核电站安全分析,以及为核电站先进主控室设计提供软件支持和验证服务：实际应用结果显示,该仿真软件能较好地模拟反应堆一、二回路的控制与保护功能。     

Thermoemission reactor-converters for nuclear power units in outer space

Conclusion Thermoemission nuclear power units with built-in generators in the nuclear reactor core can be regarded as a promising source of electric power for supplying the needs of space equipment for various purposes with a wide range of electric power demands over a long service life and with acceptable mass-limit characteristics.This article is a variant of a report presented at the Sixth Symposium on Nuclear Power in Outer Space (Albuquerque, January 1989).Translated from Atomnaya Énergiya, Vol. 66, No. 6, pp. 374–377, June, 1989.     

Thyristor crowbar system for the high current power supplies of ASDEX upgrade

The ohmic heating system and the poloidal field coils of ASDEX upgrade are supplied by 15 thyristor converter units with an installed apparent power of 600 MVA. To protect the thyristor converters against dc overvoltage arising from abnormal operations and resulting damages caused by the large energy stored in the AUG magnet coils an overvoltage protection system was required. The paper describes the motivation for—and the design and testing of the thyristor crowbar system representing the thyristor converter overvoltage protection system. It will present the layout, analyse the results of measurements obtained during commissioning, compare them to the calculated (design) values and report on the first experience of operation on the AUG coils improving the safety of the equipment.     

辐射场监测在合肥光源中的应用研究

合肥光源于近期完成了重大升级改造,对电子加速器和储存环的结构进行了重新调整,同时升级了辐射监测系统和束流损失监测系统。除了具备基本的机器保护与人员安全防护功能以外,通过加入时间信息、位置信息和加速器运行参数信息,新监测系统还具有一定的机器诊断能力,对加速器的调机和维护有辅助作用,同时也加深了对辐射源项的了解,对优化防护措施有指导意义。     

The swedish underground containment studies: State of art

Since 1973 studies of underground siting for nuclear power plants have been going on in Sweden. War protection, being the primary aim in accordance with the instructions, the first containment study has lead to siting in rock or in a pit. Rock siting gives better war protection than pit siting and also has less effect on the landscape, the cost being about equal. The second study was aimed at surveying the advantages and disadvantages of a rock sited 1000 MW BWA nuclear power plant from a reactor safety standpoint, compared to a plant above ground.Based on the instructions and considerations within the study group, the following criteria for the plant design have been established. (1) The plant should be designed to give protection against external acts of war with conventional weapons. (2) The plant should have a safety level equal to that of an above ground plant. It should fulfil the demands set by the authorities for above ground plants with respect to normal operation and accidents. No accidents that can be dealt with above ground may be permitted to result in more serious consequences, nor may they have a higher probability in a plant sited in rock. (3) The design of the plant should moreover utilize the possibilities of improving the safety afforded by rock siting. The criterion about war protection leads to siting in rock or pit, as shown in a previous CDL study. The study group has concentrated its work on rock siting.To clarify the two other criteria, the study group has outlined four alternative designs of a rock sited plant: (1) Reactor with complete pressure suppression (PS) containment placed together with central auxiliary equipment in a closed cavern with 48 m span that is placed about 50 m under the rock surface. (2) Reactor with complete PS containment placed together with auxiliary equipment in similar cavern as alternative (1) but open to the atmosphere. (3) Reactor placed ment. (4) Reactor placed in a containment directly surrounded by the rock. Auxiliary equipment placed in separate caverns.Standardization and quality improvement today are preferred to choosing new systems and more advanced technical solutions. Also, considering the desire from a safety standpoint, a rock sited plant should as much as possible exploit established technology. A consequence of the desire to use established technology is that the reactor cavern should be open to the atmosphere. If the cavern is closed, certain pipe rupture accidents may give overpressures that are difficult to master without large design alterations. The criterion about utilizing the possibilities for increasing the safety leads the interest to extreme improbable accidents, where certain advantages seem to be attainable with rock siting.A tight, strong cavern around the reactor would thereby be an ideal solution. This design appears, however, difficult to combine with the criterion about equal safety level as above ground. This criterion that controls the safety in the circumstances normally considered for nuclear power plants must be satisfied primarily, since extreme accidents have such very low probability. The tight cavern has therefore had to stand back for the open. The study shows, however, that an open reactor cavern can also be designed to significantly increase the protection of the surroundings in extreme improbable accidents compared to above ground plants.The chosen technical design of the reactor plant demands a cavern with a 45–50 m span. Caverns without strengthening efforts with such spans are used in mines, but have not previously been used for industrial plants. Studies of the stability of such caverns show that a safety level is attainable corresponding to the safety required for the other parts of the nuclear power plant. The conditions are that the rock is of high quality, that necessary strengthening measures are taken and that careful studies of the rock are made before and during the blasting, and also during operation of the plant.The third study was delivered to the government in 1977. One part in this study is going deeper in certain questions (safety, operation, maintenance, sabotage, war protection, cost and decommissioning). Another part aims to a broader view of risks and consequences in peace and war and also advantages and disadvantages of nuclear power plant for district heating.     

山西省部分放射治疗设备性能检测及防护设施评价

依据有关国家标准规定的检测项目和检测方法,山西省卫生厅卫生监督所对山西省39家医疗机构使用的28台60Co治疗机和21台医用电子加速器的防护性能进行了检测,并对其机房防护设施进行监测评价.检测结果显示,放射治疗设备防护性能总体较好,且医用加速器优于钴治疗机;机房防护设施符合国家标准.     

Prospects for using microelements in VVÉR reactors

The paper deals with prospects for using micropins in VVéR reactors in order to provide a radical improvement in station safety. The most efficient arrangement is direct micropin cooling by the water moderator in assemblies involving transverse flow around a micropin layer. That design provides neutron-physics and thermohydraulic characteristics similar to those of VVéR reactors containing rod elements. The micropins do not require any change in the reactor design, and pin assemblies containing them may be built to be identical with those in traditional VVéR reactors as regards dimensions, connecting units, and thermohydraulic and neutron-physics characteristics. The conditions occurring in major accidents show that such reactors containing micropins can provide radiation safety in any major accident, and it is particularly promising to use them to extend the working lives of first-generation power stations not fitted with pressurized vessels or sealed shells. Translated from Atomnaya énergiya, Vol. 86, No. 6, pp. 443–449, June, 1999.     

核电厂应急柴油发电机继电保护分析

文章主要介绍了核电厂应急柴油发电机功能与特点,在核事故情况下起到的重要性。详细介绍了继电保护的特点及保护配置,论述保护方案的合理性,能够完全满足应急柴油发电机的各种运行状态。对继电保护装置的原理、特性进行详细的论证分析,在应急柴油机出现不正常运行状态和严重故障情况下的保护动作,快速切除柴油机,防止柴油机的进一步损害;分析了应急柴油发电机在外部电源扰动情况下,继电保护装置如何快速响应,闭锁保护动作,避免应急柴油发电机不必要的情况下切除,造成应急电源失去,影响核电厂安全。同时,分析了继电保护装置保护定值的灵活应用,满足各种运行状况。阐述了核电厂由于其核安全的特殊性,导致继电保护配置的特殊性,分析了此种保护在其他设备上应用的可能性,并对设计提供的保护定值进行分析,提出自己的观点。     

浅谈民用核安全设备设计制造许可证的审查

《民用核安全设备监督管理条例》明确规定民用核安全设备设计制造等活动单位需要申请领取许可证或进行注册登记.国家核安全局修订颁布了相应的部门规章和配套文件.本文简要列出了申请许可证和注册登记单位的基本条件,讨论了许可证和注册登记审查过程中遇到的一些问题.     

Securing safe operation of VVÉR reactor systems in nuclear power plants

The safe operation of VVéR reactors has been discussed throughout the entire design process, taking account of the normative documentation, including the international requirements (IAEA, EUR). After the first domestic normative document “Basic principles for securing the safety of nuclear power plants” was approved in 1973, work began on the reconstruction of the first-generation VVéR-440 power-generating units. The measures taken to increase safety concerned all types of reactors VVéR-440 and-1000. Information on implementing these measures is presented. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 87–93, August, 2006.     

Present status of nuclear containments and ISI in Korea

Since the first nuclear power plant started in commercial service in 1978 in Korea, 20 units have been operated and maintained, and most recently several units were under construction and planned to be constructed in order to meet the demand of more electricity. The importance of nuclear containments always has been one of the hottest issues for the safety and protection of nuclear power plants. From 1970s to present year, various types of nuclear containments have been constructed until now. With the changes of times, nuclear containment systems have undergone a remarkable change, and finally a Korea standard nuclear power plant was defined. For those reasons, various regulatory issues, inspection technologies, technical requirements for periodic inspection have been applied differently depending on the specific nuclear containment types. In this study, overall status of nuclear power plants, development stages of nuclear containment systems, and inservice inspections in Korea were researched.     

三门核电AP1000机组辐射防护设计分析

三门核电AP1000机组为第三代核电机组,在辐射防护设计中采用了一回路加锌、较高pH值运行、停堆氧化操作、蒸汽发生器一回路水室电解抛光、优化设备维修、优化屏蔽设计、无线剂量监测等措施,以期降低机组辐射水平和职业照射剂量。本文介绍了三门核电AP1000机组在功率运行及大修期间的辐射水平和职业照射剂量数据,并与国内CPR1000机组的相关数据进行了对比,对AP1000机组的辐射防护设计进行分析,给出了三门核电AP1000机组在辐射防护运行管理及技术改进方面的建议。     

Optimal fuel utilization during decommissioning of nuclear power plants with RBMK reactors

When RBMK reactors are decommissioned successively in the same nuclear power plant, part of the fuel of the stopped reactors can be transferred to other units and additionally burned in continuing operations. The problem of minimizing the consumption of fresh fuel by optimal distribution of the additionally burned fuel over the reactors is examined. The limitations on the refueling rates, the holding time of fuel assemblies prior to transfer, the service life of fuel assemblies, and certain characteristics of reactors are taken into account. It is shown that the reuse of fuel in other units permits saving from one to almost two thousand fresh fuel assemblies and that the effect of optimizing the additional burn regime can reach several hundreds of saved fuel assemblies. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 284–290, May, 2007.     

星球表面用核反应堆临界安全分析

核反应堆电源具有寿命长、可全天候工作等特点,可作为星球表面及其他深空探测任务的电源。针对星球表面用核反应堆电源在发射过程中重返地面的临界安全问题,提出了星球表面用核反应堆的临界安全分析要求、分析假设与模型,并对反应堆临界安全特性及采取的临界安全措施进行了计算分析。计算结果表明,不同假设掉落环境下的星球表面用核反应堆的有效增殖因数均小于0.98,满足临界安全要求。反应堆通过采用Mo-14%Re合金结构材料、设置相对较厚的堆芯反射层以及在反射层包壳和堆芯外围涂覆Gd₂O₃涂层等措施有利于确保反应堆在事故时处于次临界状态。     

核电厂电气贯穿件设备延寿再鉴定方法研究

以秦山一期核电厂电气贯穿件(EPA)为实施对象,基于已有设备设计、制造、鉴定和实际运行数据,并结合国内外核电厂的设备鉴定、老化延寿管理等标准规范和最新研究成果,开展了核电厂设备延寿再鉴定方法研究,同时结合目前国内核电厂类似EPA等1E级核安全设备老化管理的现状,提出了设备老化管理建议。