Inherently safe containments for nuclear power plants

Nuclear energy cannot be avoided in the near future. To regain public acceptance the safety of nuclear power plants has to be increased. Consequently, feasibility studies have been carried out for a containment proposal for future pressurized water reactors which will keep people unharmed even in the case of severe nuclear accidents under the assumption “all that can go wrong, will go wrong”. The main features of the design concept are a core melt cooling and retention device, a passively acting cooling system to remove the decay heat and a double-wall containment which is able to withstand high static and dynamic internal pressures due to hydrogen detonation. Internal structures are designed to resist extreme loadings resulting from various accident scenarios including in-vessel steam explosion and vessel failure under high system pressure.     

Gas cloud explosions and resulting blast effects

The design of nuclear power plant structures to resist blast effects due to chemical explosions requires the determination of load-time functions of possible blast waves. Whether an explosion of a hydrocarbon gas in the atmosphere will occur in the form of a deflagration or a detonation is largely dependent on the type of flame acceleration process which is closely related to the rate of energy release. Flame propagations at normal flame velocities in a free explosible gas cloud will certainly not lead to detonation. However, with sufficiently large clouds — particularly under adverse boundary conditions — the flame acceleration could become so high that an initial deflagration changes into a detonative process.Results of recent investigations, which will be discussed in detail, show that in a free cloud with deflagrative ignition (flame, heated wire, sparks) the occurrence of a gas detonation can practically be excluded. Apparently, free gas clouds can only be induced to detonate by a sufficiently strong detonative initiation. Independently of the initiating event in the practice of damage analysis, it has become customary to describe the consequences of an explosion by means of the so-called TNT equivalent. Therefore, it is attempted to specify this equivalent for hydrocarbons by means of energetic considerations including the propagation functions for the case of spherically symmetric detonations. Analogous to the safety distances required in the handling and storage of high explosives, a mass-distance relation of the form could be considered where L is the mass of spontaneously released hydrocarbon and k varies only with the structural shape of the blast loaded buildings.With the inclusion of an empirical relation which relates the quasi-static design pressure for a building with the normally reflected blast pressure of a blast wave, it is further attempted to establish a relation between the structural capacity of a building — i.e. the pressure resistance of a building structure for detonative dynamic loading and for quasi-static loading — and the unit-mass distance .     

Flame-quenching model of the quenching mesh for H2–air mixtures

A deflagration to detonation transition (DDT) occurrence is one of the most important issues concerning safety during severe accidents in nuclear power plants because it can damage the integrity of the containment. It is possible to arrest the acceleration of a flame which can cause DDT by installing quenching meshes between the compartments. To evaluate the applicability of a quenching mesh to nuclear power plants, it requires a means to evaluate a flame arrest of a quenching mesh under a given combustion condition. The flame-quenching models developed by previous researchers were derived to fit the experimental geometry and to consider various thermal boundary conditions from a flame to the mesh wall. Flame-quenching tests were carried out at the 10% hydrogen concentration in a dry air by changing atmospheric pressure to 2.2 bar as the initial pressure. The quenching criterion of a quenching mesh with a 0.3 mm gap distance for hydrogen–air mixtures is established by using the experimental data. The flame-quenching models are also evaluated by using the experimental data. A flame-quenching model that can be used to evaluate a flame arrest for various hydrogen–air mixtures in nuclear power plants is proposed.     

Estimation of ex-vessel steam explosion pressure loads

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment.In this article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel–coolant interactions. A parametric study was performed varying the location of the melt release (central, right and left side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to establish the influence of the varied parameters on the fuel–coolant interaction behaviour, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. For the most explosive central, right side and left side melt pour scenarios a detailed analysis of the explosion simulation results was performed. The study shows that for some ex-vessel steam explosion scenarios higher pressure loads are predicted than obtained in the OECD programme SERENA phase 1.     

Evaluation of hazards from industrial activities near nuclear power plants — deterministic and probabilistic studies

Among the potential hazards which could arise from industrial activity near nuclear power plants, fires and explosions of dangerous products are of particular concern. Indeed, thermal radiation from an adjacent fire could endanger the resistance of a plant's structures. Likewise, an accidental explosion would induce an overpressure wave which could affect buildings' integrity.

This paper presents the methodology developed by Electricité de France to evaluate the consequences of accidents affecting:

• - Industrial facilities: refineries, chemical and petrochemical plants, storage areas, pipelines of gaseous, liquid and liquefied materials.

• - Transportation routes (roads, railways, inland waterways) used to carry dangerous substances (solid explosives, liquid, gaseous or liquefied hydrocarbons).

Probabilistic methods have been developed by analysis of actual accident statistics (e.g. risks induced by transportation routes) and realistic and representative accident scenarios have been set up. Five sequences have been identified:

• - Formation of a fluid jet at a breach.

• - Evaporation and possible formation of a liquid layer.

• - Atmospheric dispersion and drift of a gaseous cloud.

• - Heat radiation from fire.

• - Unconfined explosion of a gaseous cloud.

This paper gives an overview of the methods and the main assumptions used to deal with each sequence. Those methods, presently applied by Electricité de France, provide a coherent and realistic approach for the evaluation of the risks at nuclear power plants induced by industrial activity.     

Hydrogen and steam distribution following a small-break LOCA in large dry containment

The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations. Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident （LOCA） without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump （RCP） compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.     

Relationship between the structure of vapor explosion and fragmentation mechanisms

A computer code PROVER-I for propagation phase of vapor explosion is developed to couple with MPS simulation. From the simulation results by MPS method, a new thermal fragmentation model is proposed with three types of time scale for modeling instant fragmentation, spontaneous nucleation fragmentation and normal boiling fragmentation. The structure of vapor explosion is investigated based on different fragmentation models. The vapor explosion with spontaneous nucleation fragmentation is characterized as a non-equilibrium overdriven detonation, and vapor explosion with hydrodynamic fragmentation is characterized as a non-equilibrium spontaneous burning. Sensitivity analyses are carried out for some parameters such as fragmentation time scale, delay time and vapor volume fraction. Spontaneous nucleation fragmentation with shorter delay time results in a vapor explosion structure situated between a non-equilibrium overdriven detonation and a non-equilibrium spontaneous burning, while a longer delay time gives a composite structure of vapor explosion which includes a shock and a non-equilibrium deflagration. The energy conversion ratio of a non-equilibrium overdriven detonation is much larger than that of a non-equilibrium spontaneous burning. The composite structure has the same destructive potential as non-equilibrium overdriven detonation. The variation of energy conversion ratio with the vapor volume fraction is consistent with the observations in fuel-coolant interaction experiments.     

Experimental study on hydrogen behavior at a subcompartment in the containment building

Hydrogen control in the case of severe accidents has been required by nuclear regulations to ensure the integrity of the nuclear containment building after Three Miles Island (TMI) accidents. Up to now, many experiments have been conducted to estimate the distribution of hydrogen during accidents in nuclear power plants. In this study, local hydrogen behavior has been experimentally investigated in a cylindrical multi-subcompartment mixing chamber of the SNU (Seoul National University) hydrogen mixing facility, measuring the local concentration in various conditions and mixture injection locations. Hydrogen is simulated by helium in the experiments. Results showed remarkably different local behavior of helium in experiments of several conditions, and the local analysis for hydrogen concentration rather than the lumped compartment analysis, used widely in most plants, would be important to ensure the equipment survivability or to determine the positions of ignitors.     

Penetration resistance of reinforced concrete containment structures

Containment structures not only provide a leak tight barrier, but also play a role in ensuring that the structures can withstand the impact load from projectile impacts or internal plant accidents. In assessing the containment structures of nuclear power plants, predicting the characteristics of impact resistance in relation to design and safety considerations is relevant. This investigation proposes a simple but effective method of performing numerical analysis on perforation resistance of reinforced concrete containment structures. In this work, normal and oblique impacting is considered to examine the residual velocity and impact phenomena of an ogive-nose steel projectile with various impact velocities against a reinforced concrete slab. Additionally, a phase diagram is devised to describe the ballistic terminal phenomena of projectile and target. This model could assess the resistance to penetration to results in the optimum design of the containment structures in nuclear power plants.     

基于火焰加速和燃爆转变准则的氢气点火安全性研究

采用点火器对可燃混合气体进行预先点火是严重事故下的1种可供选择的氢气缓解措施。基于σ准则和λ准则可以评估氢气燃烧时发生火焰加速(FA)和爆燃向爆炸的转变(DDT)的可能性。本文分析密闭房间中氢气早期和晚期点火的过程。分析结果表明，点火器在空间的合理布置和初次点火时间的控制，可有效移除事故前期的氢气。本方法能用于确定核电站干式安全壳内氢气点火器的数量、位置和点火时间。     

氢气催化复合器对核电厂严重事故的缓解效果

在严重事故条件下,安全壳内的氢气燃烧或爆炸威胁安全壳完整性,必须采取措施减小或消除安全壳的氢气风险。针对600MWe级核电厂的大型干式安全壳,以小破口失水诱发的严重事故序列为基准事故,计算分析了氢气催化复合器（PAR）消除安全壳内氢气的效果,及复合效应对安全壳压力温度的影响。研究表明：氢气催化复合器能够持续稳定地消除安全壳内氢气,但对于极其快速的氢气释放,它的消氢能力受到一定限制。     

Response of NPP structures to simultaneously acting air pressure loads and ground waves caused by a gas cloud explosion

Gas cloud explosions cause air pressure waves which propagate over the ground surface. The ground motion induced by these loads and their effect on structures are studied. The soil is modelled as a linear viscoelastic medium. A semianalytical method is used to compute the ground motion produced by a deflagration and by a detonation in a stiff and a soft layered soil. For a PWR reactor building subjected to the direct impact of an air pressure wave the additional effects of the ground waves on the motion of the building are studied. Whereas the vertical structural response is increased, the horizontal response decreases, when the effect of the ground waves is included. For the case studied the additional effect of the ground waves is small.     

Study on local hydrogen behaviors in a subcompartment of the NPP containment

Hydrogen control in the case of severe accidents has been required by nuclear regulations to ensure the integrity of containment after TMI accidents. Up to now, many experiments have been conducted to estimate the distribution of hydrogen during accidents in nuclear power plants. In this article, we proposed a computer code named HYCA3D developed to calculate the local hydrogen distribution with three-dimensional time-dependent governing equations, which can simulate the transport of multiple species. Also, local hydrogen behavior has been experimentally investigated in a cylindrical multi-subcompartment mixing chamber, measuring the local concentration in various conditions. Hydrogen is simulated by helium in the experiments. The proposed code was verified with these experimental results, followed by pre-tests with EPRI/HEDL standard problems. The calculation results show good agreement with the experimental data.     

核材料炸药化学爆炸条件下烟云源项的实验与仿真

为获取核材料化学爆炸事故烟云参数和气溶胶扩散规律,本文理论分析了核材料化学爆炸事故烟云扩散过程与机理,建立了基于CFD技术的烟云扩散数值仿真方法,并通过爆炸烟云外场扩散实验对该方法进行了验证。研究结果表明：CFD方法能实现烟云扩散的数值仿真;实验和仿真均显示小当量TNT爆炸高度远小于Church经验公式,宽度比例却增加;气溶胶颗粒在蘑菇云形态的双涡环烟云流场中分布不均匀,粒径大于50 μm的颗粒物大多位于烟柱中,而大部分可吸入颗粒物在烟云顶部聚集;气溶胶在烟云稳定前的驱散与沉降会改变其源项参数,以Operation Roller Coaster实验气溶胶积累质量分布为例的计算显示,空载释放率约为地面气化率的58%,可吸入比率由20%升高至31%,可吸入释放率约为18%。     

Transient cavitation effects in fluid piping systems

An accurate prediction of pressure transients and associated loadings in nuclear power plant piping systems requires a treatment of cavitation. A technique for calculating this effect in a general fluid-hammer analysis by the method of characteristics is developed. Cavitation is treated by a modified column separation model and is assumed to be a local phenomenon occurring whenever the pressure drops below the vapor pressure of the fluid. While the model is a simplification of the actual phenomena it reproduces the essential features of transient cavitation. Computational results obtained for a variety of piping arrangements demonstrate the versatility of the approach, and clearly illustrate the fact that neglecting cavitation leads to erroneous pressure-time loadings in the piping systems. Comparisons of calculated results with available experimental data, for a simple piping arrangement, show good agreement and provide validation for the computational cavitation model.     

核事故与国内现役核电机组核安全措施

根据核电厂核安全和辐射安全的设计防御准则,对核电史上三次重大事件进行分析,挖掘出核电事故主要因素：人因因素和超过设计值的自然灾害。同时结合国内核电厂的设计参数和运行参数,对发生类似事故进行研究比较,提出必要的预防方案。国内现役核电厂在运行安全技术上,已经可以充分预防人因事故的发生,对于超过设计值的自然灾害及外在因素引起的事故,还应重新考虑安全标准。核电厂在建设和运行过程中,需要充分考虑在极端环境下,如何将核辐射和泄漏的危害程度降低至政府以及公众能够接受的水平。     

百万千瓦级压水堆严重事故卸压阀高温瞬态分析

由于核电厂严重事故的恶劣工况,在卸压过程中严重事故卸压阀门可能会经历阀门无法承受的高温瞬态而导致不可用。本文在可能导致高压熔堆的事故序列中筛选出具有一定的包络性并包含各种典型严重事故现象的典型严重事故序列。针对该事故序列考虑严重事故管理中的开阀时间范围开展了高温瞬态计算,并针对重要的影响因素阀门开启时刻的稳压器水位开展分析。最终确定了百万千瓦级核电厂具备典型性及一定包络性的严重事故卸压阀工作条件,并得到了阀门开启前后阀门可能经历的最高流体温度及流体温度变化曲线,为严重事故卸压阀门的设备鉴定及功能应用提供了重要基础。      

核电厂蒸汽发生器一次侧向二次侧泄漏率取值研究北大核心CSCD

合理确定蒸汽发生器一次侧向二次侧泄漏率取值,并据此制定核电厂运行策略,对核电厂的安全及稳定运行意义重大。本文根据泄漏率数值使用目的,将泄漏率分为用于辐射防护设计的泄漏率取值、用于核电厂运行控制的泄漏率控制值、用于保证蒸汽发生器传热管完整性的泄漏率保护阈值三大类,并探讨了各类取值的确定依据。完成了对国内外核电厂蒸汽发生器一次侧向二次侧泄漏率取值情况的调研分析,结合研究情况,提出了我国核电厂蒸汽发生器一次侧向二次侧泄漏率取值及控制的建议。     

Subcritical power reactor with irradiation by a beam of accelerated protons

Conclusions The conversion of a nuclear power plant to operation with quite deeply subcritical reactors eliminates the primary reason for the appearance of reactivity accident situations associated with the probability of the reactor being transferred into a subcritical state and runaway of this state. There is no doubt that a nuclear power plant can in principle operate on the basis of a subcritical reactor and a high-power proton accelerator. To answer the question of whether or not it is desirable to equip nuclear power plants with accelerators, it must be kept in mind that besides achieving the main goal — complete elimination of the possibility of reactivity accidents and as a consequence of such accidents emission of solid radioactive products of uranium fission with enormous consequences for ecological and economic damage — such improvements have other important consequences. These include, for example, the possibility of constructing fuel cycles on the basis of the fuel depleted with respect to fissioning isotopes (^233,235U,²³⁹Pu), which will make it possible to decrease substantially the fuel component of the cost of a nuclear power plant; the possibility of more efficient utilization of nuclear fuel by increasing significantly the interval between loadings; and, control of the power and shielding of a reactor by changing the beam current of the accelerator. All this will make it possible, in principle, even at today's level of development of reactor and accelerator technology to build a subcritical power reactor with external irradiation with a high-energy particle beam. Institute of High Energy Physics. Translated from Atomnaya énergiya, Vol. 77, No. 4, pp. 300–308, October, 1994.     

Analysis on hydrogen risk mitigation in severe accidents for Pressurized Heavy Water Reactor

Hydrogen source term and hydrogen mitigation under severe accidents is evaluated for most nuclear power plants (NPPs) after Fukushima Daiichi accident. Two units of Pressurized Heavy Water Reactor (PHWR) are under operating in China, and hydrogen risk control should be evaluated in detail for the existing design. The distinguish feature of PHWR, compared with PWR, is the horizontal reactor core surrounded by moderator in calandria vessel (CV), which may influence the hydrogen source term. Based on integral system analysis code of PHWR, the plant model including primary heat transfer system (PHTS), calandria, end shield system, reactor cavity and containment has been developed. Two severe accident sequences have been selected to study hydrogen generation characteristic and the effectiveness of hydrogen mitigation with igniters. The one is Station Blackout (SBO) which represents high-pressure core melt accident, and the other is Large Break Loss of Coolant Accident (LLOCA) at reactor outlet header (ROH) which represents low-pressure core melt accident. Results show that under severe accident sequences, core oxidation of zirconium–steam reaction will produce hydrogen with deterioration of core cooling and the water in CV and reactor cavity can inhibits hydrogen generation for a relatively long time. However, as the water dries out, creep failure happens on CV. As a result, molten core falls into cavity and molten core concrete interaction (MCCI) occurs, releasing a large mass of hydrogen. When hydrogen igniters fail, volume fraction of hydrogen in the containment is more than 15% while equivalent amount of hydrogen generate from a 100% fuel clad-coolant reaction. As a result, hydrogen risk lies in the deflagration–detonation transition area. When igniters start at the beginning of large hydrogen generation, hydrogen mixtures ignite at low concentration in the compartments and the combustion mode locates at the edge of flammable area. However, the power supply to igniters should be ensured.