CANDU6核电厂无过滤安全壳通风模式的研究

CANDU6核电厂早期设计未考虑严重事故对策,在严重事故下,CANDU6核电厂的安全壳容易失效。为了解决这一问题,本文研究了无过滤安全壳通风模式对CANDU6核电厂安全壳的影响。本文选取典型的全厂断电严重事故,利用重水蒸气回收系统作为无过滤安全壳通风的路径,初步研究了该通风模式下对安全壳完整性的保持和对裂变产物源项的滞留能力。研究表明:该通风模式可以有效保持安全壳的完整性,同时,对裂变产物源项也有一定的滞留能力。     

严重事故下开启双层安全壳环形空间通风过滤系统对缓解放射性向环境释放影响研究

核电站发生严重事故后,安全壳能包容从堆芯释放出的裂变产物,防止向环境的大量释放,但即使在安全壳完好的情况下,仍然会存在一定量泄漏。目前国际上的三代核电机型,大多采用双层安全壳的设计,对裂变产物具有一定的包容、滞留和过滤作用。本文基于我国自主设计的第三代核电机组,结合双层安全壳的设计特点和特定源项分析,对严重事故下双层安全壳之间的环形空间及其通风过滤系统对缓解裂变产物向环境释放的作用进行了定量分析,结果显示双层安全壳及环形空间通风过滤系统能够显著降低放射性气溶胶对环境的释放,对惰性气体也有一定的延缓排放作用。     

地下核电厂防止气载放射性核素扩散的研究

给出地下核电厂防止气载放射性核素扩散的总体原则,分析地下核电厂防止气载放射性核素扩散的工程措施,重点研究严重事故下的防护措施。提出一种严重事故下可实现安全壳及反应堆厂房洞室及时卸压的系统。该系统非能动响应,并根据严重事故的不同情况自动采取相应的应对措施,过滤排放安全壳及洞室内放射性气载物的同时为安全壳和洞室降压。通过这些工程措施可以更好地控制地下核电厂严重事故中产生的气载放射性核素,从设计上实现实际消除大量放射性核素释放的可能性。     

压水堆事故工况下放射性裂变产物向环境转移和释放的理论计算

本文建立了分析压水堆事故工况下惰性气体、元素碘、甲基碘和气溶胶粒子等气载裂变产物由安全壳向环境转移和释放的多仓室安全壳模型——FIPREA 模型。此模型考虑了单层、双层和半双层三种型式的安全壳中堆芯源项、自然沉积、过滤器捕集、喷淋液吸附及泄漏等因素对气载裂变产物浓度变化的影响。根据此模型编制了分析裂变产物去除及对环境释放情况的计算程序。本程序可用于核电站设计或安全评审时事故释放量的分析计算。     

百万千瓦级核电厂蒸汽发生器失去给水事故源项计算分析

对百万千瓦级压水堆核电厂的安全壳内进行隔间,应用IRSN和GRS等联合开发的ASTEC程序计算该类型核电厂在发生蒸汽发生器完全失去给水严重事故工况下放射性裂变产物在安全壳内释放迁移的情况,给出了主要隔间内的放射性活度.根据安全壳内喷淋系统能否正常启用对各个隔间内的放射性活度进行了比较.结算结果表明,喷淋能否启用,对Xe、Kr等惰性气体在各隔间内分布几乎无影响;但可以大大降低I、Br等易生成气溶胶、水溶性较好的裂变产物的浓度.对其他主要以气溶胶形态存在于安全壳气空间中的裂变产物也有很强的去除作用.喷淋的成功启用,将大部分放射性裂变产物冲刷入下部的地坑区,使得安全壳内上部空间的放射性活度有了明显的降低,但裂变产物聚积在地坑,使地坑的活度大大提高.     

小破口引发的严重事故工况及事故缓解的研究

利用MAAP4程序对方家山核电站进行建模,针对事故后果较为严重的小破口事件进行了计算分析,得到了假设事故下电厂系统的反应以及相应的严重事故现象.对事故中发生的DCH(安全壳直接加热)现象和安全壳失效以及裂变产物向环境的释放进行了分析.随后,本文根据相关的严重事故管理导则和该事故的特点,对缓解该事故的策略进行了研究和计算...     

基于MAAP5程序的秦山核电站严重事故分析

应用MAAP5程序建立了秦山核电站一、二回路,安全系统以及安全壳的模型,并以冷段双端断裂叠加高高、高、低压安注失效,安全壳喷淋系统失效为例,对该严重事故序列进行了模拟计算,给出了瞬态过程一些重要参数随时间的变化规律。结果表明：在72 h内无能动干预手段的条件下,安全壳的完整性可得到保证,相关数据可为秦山核电站严重事故预防和事故缓解措施的制定提供重要参考。     

严重事故下安全壳内裂变产物碘影响因素敏感性分析

反应堆发生事故最严重的后果是放射性裂变产物弥散到环境中,为了研究严重事故工况下放射性裂变产物碘在安全壳内的分布特点,本研究假设核电厂已经发生严重事故,一回路裂变产物碘释放到安全壳内。使用事故源项评估程序（ASTEC）构建核电厂安全壳结构模型,并设置边界条件,计算了裂变产物碘在不同pH值、有无金属银注入和气相辐照工况下的化学形态、化学特性、分布情况以及不同化合物的变化趋势。研究结果表明,碱性环境下可以降低安全壳内挥发性碘的生成;银的存在可以增加液相中碘的捕获和降低碘的挥发;气相辐照环境可以提高气相CH₃I 和IO_x的形成。本研究可以为严重事故工况下安全壳内放射性碘的去除提供支持。      

小破口冷却剂丧失始发安全壳失效事故源项行为分析

采用一体化严重事故仿真程序对600MW压水堆核电厂小破口冷却剂丧失(SB-LOCA)始发安全壳隔离失效、安全壳早期失效和晚期失效三类事故的源项行为进行分析。分析结果表明:(1)由于沉积作用或残留在熔融物中,挥发类和非挥发类裂变产物相对于惰性气体类,释入环境份额较小;(2)事故进程中安全壳与环境之间较小的压差和安全壳较晚的失效时间,分别使得在安全壳隔离失效和晚期失效事故中裂变产物较为缓慢地释入环境;(3)安全壳早期失效事故中,在安全壳直接加热(DCH)现象发生后熔融物颗粒与安全壳大气换热过程中,从熔融物释出的挥发性与非挥发性裂变产物在安全壳失效后快速地释入环境。上述结论可为严重事故源项缓解措施研究、厂外后果评价以及应急策略制定提供技术支持。     

采用ВВЭР-1000型反应堆装置的АЭС-91型核电站简介(二)

4.2　安全系统为防止或限制反应堆装置损坏以及包容核电站放射性裂变产物 ,考虑了下列安全系统 :　　 ● 　保护系统　　 ● 　包容系统　　 ● 　保障系统　　 ● 　控制系统安全系统的结构能动安全系统由 4个完全独立的通道组成。通道的容量、动作的速度和其它特性是根据当发生设计事件时能保障核和辐射安全性的条件而设计的。由于安全系统通道布置在单独房间和在安全壳中的单个地坑中 ,通道的实体隔离程度很高。安全系统通道包括从一个厂房到另一个厂房之间的连接通道 ,相互之间用防火实体屏障隔开。不允许不同安全通道之间有直接联系 ,…     

核供热堆安全壳设计方案的探讨

介绍了壳式核供热堆几种安全壳的设计特点。根据核供热堆的实践和该堆安全壳功能的分析比较,提出了取消“紧贴式”钢安全壳,采用大容积砼壳作为第三道安全屏障的可能性。     

Numerical method for the containment of fission products at a hypothetical accident in an underground nuclear power plant

As one means to expand the siting of nuclear power plants, construction of underground plants is now under study. An underground nuclear power plant has the feature that ground surrounding the underground cavity can contain the fission products of a hypothetical accident.If it is assumed that in a hypothetical reactor accident the cooling system loses its capacity wholly or partially, and gas containing fission products is emitted into the underground cavity. As a result, temperature, gas concentration and gas pressure in the cavity increase and it can be supposed that the gas leaks up to the surface through the ground, and that ground-water contains and carries fission products. The present paper numerically simulates a course of movement as mentioned above by the finite element method and gives the underground containment effect for fission products from a hypothetical accident.     

Safety consideration and economic advantage of a new underground nuclear power plant design

A conceptual design of an underground nuclear power plant is proposed to make undergrounding of nuclear reactors not only environmentally desirable but also economically feasible. Expedient to the underground environment, this design capitalizes on the pressure-containing and radiation filtering characteristics of the new underground boundary conditions. Design emphasis is on the containment of a catastrophic accident — that of a reactor vessel rupture caused by external means. The igh apacity apid nergy issipation nderground ontainment (HiC—REDUCE) system which efficiently contains loss-of-coolant accidents (LOCAs) and small break conditions is described. The end product is a radiation-release-proof plant which, in effect, divorces the safety of the public from the safety of the reactor.     

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.     

Different variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents probable variations of passive safety boiling water reactor (BWR). In order to improve safety and economy of passive safety BWR, the authors thought of use of a kind of improved Mark III type containment. The paper presents the basic configuration of the passive safety BWR that has an improved Mark III type containment. We tentatively call this passive safety BWR advanced safer BWR⁺ (ASBWR⁺) and the containment Mark X containment in the paper. One of the merits of the Mark X containment is double containment function against fission products (FP) release. Another merit is very low peak pressure at severe accidents without active cooling systems. The third merit is coolability by natural circulation of outside air. Therefore, the Mark X containment is very suitable for passive safety BWRs. It does not need a reactor building (R/B) as the secondary containment, because it is a double containment by itself. The Mark X containment is a general concept and also useful for half-passive safety BWRs that have both active and passive safety systems. In those examples, active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Effect analysis of the intentional depressurization on fission product behavior during TMLB' severe accident

It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH).Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs,but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.     

Leak-tightness characteristics concerning the containment structures of the HTTR

The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents, which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot-gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radioactivity and will maintain a correct pressure in the service area.The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept in a way that the design pressure becomes well below its allowable limitation by the emergency air purification system, which filters efficiency of particle removal and iodine removal well over the limited values.The obtained data demonstrate that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities.     

Evaluation of uncertainties in risk emanating from a number of severe accident issues

A study has been performed to estimate, for a particular pressurized water reactor, the uncertainty in risk associated with a number of key phenomenological issues. A second objective was to distinguish the individual importance of the various issues as contributors to the overall uncertainty in risk. The issues considered touched upon the areas of system behavior, containment loading, containment performance, and fission product source term behavior. It was found that the most important source of uncertainty for the plant in question (Surry) was direct containment heating (i.e., the transfer of heat from the core debris to the containment atmosphere when the debris is ejected at high pressure from the reactor vessel and dispersed throughout the atmosphere). Other significant issues included hydrogen burning, containment failure pressure, aerosol agglomeration uncertainties, the frequency of check valve failures leading to a loss-of-coolant accident (LOCA) outside containment, and the potential for having a LOCA induced by high temperatures in the reactor coolant system.     

Safety concept and operational criteria of a nuclear process heat plant

A plant design of a high-temperature nuclear reactor (HTR) of 500 MW(th) coupled with two coal gasification lines has been developed (PNP-500). The general objective of the safety concept is to maintain the same high standard as that developed for conventional nuclear power plants. As burnable gases are transported within the primary circuit of the reactor, the containment building and in close proximity outside of the containment, novel safety considerations are arising. The related safety concepts and the results of experiments with exploding gases are discussed. Also specific operating criteria are discussed which are due to the coupling between a nuclear reactor and the two gasification lines.