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.     

Analysis of hydrogen flame acceleration in APR1400 containment by coupling hydrogen distribution and combustion analysis codes

This study was conducted as part of the construction of an integrated system to mechanistically evaluate flame acceleration characteristics in a containment of a nuclear power plant during a severe accident. In the integrated analysis system, multi-dimensional hydrogen distribution and combustion analysis codes are used to consider three-dimensional effects of the hydrogen behaviors. GASFLOW is used for the analysis of a hydrogen distribution in the containment. For the analysis of a hydrogen combustion in the containment, an open-source CFD (computational fluid dynamics) code OpenFOAM is chosen. Data of the hydrogen and steam distributions obtained from a GASFLOW analysis are transferred to the OpenFOAM combustion solver by a conversion and interpolation process between the solvers. The combustion solver imports the transferred data and initializes the containment atmosphere as an initial condition of a hydrogen combustion analysis. The turbulent combustion model used in this study was validated by evaluating the F22 test of the FLAME experiment. The coupled analysis method was applied for the analysis of a hydrogen combustion during a station blackout accident in an APR1400. In addition, the characteristics of the flame acceleration depending on a hydrogen release location are comparatively evaluated.     

Numerical simulation of lateral phase distribution in turbulent upward bubbly two-phase flows

New constitutive models for the interfacial forces acting on bubbles were developed for accurately predicting the lateral phase distribution in turbulent bubbly two-phase flow in vertical channels. Several experimental measurements have revealed that the lateral void profile in bubbly two-phase flow varies from the void peaking near the wall to the almost flat distributions as the liquid velocity increases. However, within the authors' knowledge, the effect of liquid velocity on the void profile has not been successfully predicted by the existing models; this would indicate the strong limitation of the existing multidimensional two-phase flow models. In view of these, the validity of the present constitutive models was tested in varied conditions of the liquid velocity as well as the bubble size. Since several assumptions were required in the models mainly due to the insufficient knowledge of the bubble motion, further improvements should still be needed. Nevertheless, the predicted lateral phase distributions were found to be in reasonably good agreement with available experimental data. It is hence expected that the present constitutive models can effectively be used in the practical applications and also be the base of the more sophisticated ones.     

核电站安全壳内氢气扩散和燃烧的分析程序GASFLOW及其应用

介绍了由美国洛斯阿拉莫斯实验室(LANL)和德国卡尔斯鲁厄研究中心(FzK)共同开发的三维计算流体力学程序GASFLOW的基本数学物理模型和数值计算方法。该程序主要用于分析核电站严重事故下安全壳内氢气、水蒸气扩散分布和燃烧。列举了该程序在德国Konvio型压水堆氢气安全分析中的应用。     

Integral large scale experiments on hydrogen combustion for severe accident code validation-HYCOM

A joint research project was carried out in the EU Fifth Framework Programme, concerning hydrogen risk in a nuclear power plant. The goals were: Firstly, to create a new data base of results on hydrogen combustion experiments in the slow to turbulent combustion regimes. Secondly, to validate the partners CFD and lumped parameter codes on the experimental data, and to evaluate suitable parameter sets for application calculations. Thirdly, to conduct a benchmark exercise by applying the codes to the full scale analysis of a postulated hydrogen combustion scenario in a light water reactor containment after a core melt accident. The paper describes the work programme of the project and the partners activities. Significant progress has been made in the experimental area, where test series in medium and large scale facilities have been carried out with the focus on specific effects of scale, multi-compartent geometry, heat losses and venting. The data were used for the validation of the partners CFD and lumped parameter codes, which included blind predictive calculations and pre- and post-test intercomparison exercises. Finally, a benchmark exercise was conducted by applying the codes to the full scale analysis of a hydrogen combustion scenario. The comparison and assessment of the results of the validation phase and of the challenging containment calculation exercise allows a deep insight in the quality, capabilities and limits of the CFD and the lumped parameter tools which are currently in use at various research laboratories.     

Stabilization and termination of severe accidents in LWRs

The last 20 years of research on severe accident safety for light water reactors (LWRs) has resolved a number of issues. However, the issue of melt/debris coolability is still unresolved. At stake is the stabilization and termination of a severe accident, if ever it would occur. The stabilization and termination can be established only through the coolability of the melt or the particulate debris, which are found in-vessel, or ex-vessel, depending upon the extent of the progression of a postulated accident.This paper will review the state of the art of coolability during a severe accident for the current light water reactors (LWRs). It will also review whether the accident management actions will be effective in terminating a postulated severe accident. The attention paid to the stabilization and coolability in future LWRs will be discussed and the design solutions will be evaluated.     

Analysis of hexcan failures occurring during the simulation of severe accidents in liquid metal cooled reactors

Under the SIMBATH programme the physical phenomena of transient material movement and relocation during severe LMFBR accidents are investigated out-of-pile. In most of the SIMBATH bundle experiments a failure of the wrapper was observed. From the safety point of view this has implications on the issue of propagation. By openings into the inter-subassembly gaps pressure relief and material release are possible. From the development of failure, based on measurements made during the simulation tests, and from post-experiment investigations three types of failure mode have been identified:

• - Melt-through of the wrapper wall by a jet of hot material from a failing pin. This happened very early during the test. Sodium boiling in the annular bypass prior to failure has not been detected.

• - Melt-through in the simulated fuel region by severe ablation due to local crust instability combined with intense heat input from the flowing melt.

• - Melt-through in the simulated breeding regions close to blockages. This failure mode was always observed together with sodium gross boiling in the annular channel, i.e. reduced cooling of the wrapper wall.

No mechanical failure was detected as a result of the stress concentration in the corners of the hexcan walls. The influence of the internal overpressure is restricted mainly to final break-through after severe ablation and drives the material motions after wrapper failure; it does not control wrapper wall failure in these experiments.     

Hydrogen combustion in nuclear plant accidents and associated containment loads

The objective of this paper is to describe the reactor safety problems in severe accidents due to hydrogen combustion. The generation of hydrogen is discussed. The combustion of hydrogen is considered in terms of diffusion flames, deflagrations, flammability limits, incomplete combustion of very lean deflagrations, detonations and accelerated flames. An example of analyses of hydrogen combustion in a Mark III BWR is given. The example shows the strong dependence of the predicted peak pressure and number of burns on the assumptions of the analyses, the compartmentalization of the model and the inclusion of buoyancy-driven flows.     

Accommodation of potential hydrogen formation in LMFBR accidents

Results of design verification tests for the FFTF reactor cavity liner system are presented which suggest that steel liners would retain their integrity even under certain hypothetical accident conditions, thus, avoiding the formation of hydrogen. When liner failures are postulated in hypothetical reactor vessel melttrough accidents, hydrogen levels can be controlled by an air purging system. The design of a containment purging and effluent scrubbing system is discussed.     

Numerical simulation of plasma-assisted combustion of methane-air mixtures in combustion chamber

A two-dimensional mathematical model was developed to investigate the effects of dielectric barrier discharge (DBD) plasma on CH₄-air mixtures combustion at atmospheric pressure. Considering the physical and chemical processes of plasma-assisted combustion (PAC), plasma discharge, heat transfer and turbulent were simultaneously coupled into simulation of PAC. This coupling model consists of DBD kinetic model and methane combustion model. By comparing simulations and the original reference’s results, a high-accuracy of this model was validated. In addition, the effects of PAC actuation parameters on combustion characteristics were studied. Numerical simulations show that with an inlet airflow velocity of 10 ms ^-1, a CH₄-air mixtures’ equivalence ratio of 0.5, an applied voltage of 10 kV, a frequency of 1200 kHz, compared to conventional combustion (CC), the highest flame temperature rises by 32 K; outlet temperature distribution coefficient drops by 2.3%; the maximum net reaction rate of CH₄ and H₂O increase by 11.22% and 12.80% respectively; the maximum CO emission index decreases by 14.61%; the mixing region turbulence mixing time reduces by 89 ms.     

Pressure release of containments during severe accidents in Switzerland

As required by the Swiss Federal Nuclear Safety Inspectorate (HSK) all Switzerland's five nuclear power plants have to install a containment filtered venting system. The integrity of the containment (the last barrier for radioactive releases to the environment) can be threatened by overpressure due to inadequate heat removal. Design requirements have been provided for a specific class of severe accident scenarios. In general the capacity of the system is considered sufficient if it is able to vent the steam production corresponding to a decay heat level of 1% of the thermal reactor power. The mitigation capacity for the reduction of released radioactive material is specified by a retention factor of 1000 for aerosols to prevent or limit a long term ground contamination and a factor of 100 for elementary iodine for prevention or limiting of thyroid doses and to avoid short term evacuation. Besides existing requirements for design, maintenance and operation, additional claims such as passivity and operability at any pressure conditions inside the containment have to be met. Passivity implies that the system can be initiated after a severe accident without any operator action. The system also has to allow early manual venting. Various filtered venting systems are presently available. The nuclear power plants of Beznau, Gosgen, Leibstadt and Muhleberg have already selected such systems and already implemented them or are going to install them step by step. Beznau selected the Sulzer-EWI system which is using a water pool with nozzles-baffle plates and mixing elements to achieve the required filtration of the aerosols. In both Beznau units, the systems are installed and in standby mode. Gosgen, a pressurized water reactor as well as Beznau, is going to implement a filter system developed by Siemens-KWU, known as sliding pressure venting process, combining a venturi scrubber in a water pool and a mesh filter. The boiling water reactor of Leibstadt also selected the same system as Beznau while Müheberg choose the ABB system but not in the common design. The venturi pipes are thereby integrated in the water pool of the outer torus. The system in all five nuclear power plants is fully operable and in standby mode since December 1993.     

Experiments on air ingress during severe accidents in LWRS

Various kinds of experiments on the oxidation of Zircaloy-4 cladding material in different scales and under different conditions at temperatures 800–1300 °C (small scale) and up to 2000 °C (large scale) are presented. The focus of this work was on prototypic mixed air–steam atmospheres and sequential reaction in steam and air, where no data were available before. The separate-effects tests were performed to support the large scale bundle test QUENCH-10 and to deliver first data for model development.     

Progress in core and cermet fuel modelling to calculate severe accidents

The use of cermet type composite fuels leads to an optimised use of plutonium; a good thermomechanical behaviour due to a low operating temperature thanks to a high thermo-conductivity, that favours high burn-up due to the low fission gas release. However, the increase in the metallic mass, an alloy of zircaloy, in the core, as well as the composite nature of the fuel with two very different melting temperature ( 1873 K for the metal, and 2573 K for the ceramic) lead to a behaviour very different from that of the traditional ceramic fuel in the event of an accident.     

Effect of hydrogen combustion in the primary pump compartment

Hydrogen combustion in a nuclear power plant may threaten the integrity of some important systems and components.In this paper,the effect of hydrogen combustion in the primary pump compartment is analyzed by different initial hydrogen concentration and igniter locations using Computational Fluid Dynamics method.The results show that the combustion is confined to a limited area without pump damage at about 6.6%hydrogen volume fraction.Once igniting the hydrogen,the combustion affects the whole compartment at the 12%hydrogen volume fraction.The stress caused by the great temperature gradient or high temperature may damage the primary pump. Igniters at the lower location accelerate the combustion process and cause a threat to the pump integrity.     

Review of thermal-hydraulic researches in severe accidents in light water reactors

The Tohoku Region Pacific Coast Earthquake and subsequent severe accident (SA) in Fukushima Daiichi Nuclear Power Station caused unprecedented disaster in Japan. Before this accident, considerable researches on SAs had been carried out in Japan. However, unfortunately, such researches could not prevent the accident due to the unexpected huge Tsunami. However, the researches on SAs become more and more important in order to make clear the causes of the accident in Fukushima and improve the safety of nuclear power plants in Japan. In view of this, review on researches on thermal hydraulics in SAs in light water reactors was carried out. Important thermal-hydraulic phenomena in SAs were identified. Research activities on each phenomenon were surveyed mainly based on the articles published in Journal of Nuclear Science and Technology of Atomic Energy Society of Japan.     

Results of investigations of severe accidents involving water-cooled reactors

Russian Scientific Center “Kurchatovskii institut.” Translated from Atomnaya énergiya, Vol. 76, No. 4, pp. 282–302, April, 1994.     

Analysis of strategies for containment venting in case of severe accidents

During a core melt accident, a pressurization of the containment has to be expected, which could lead to a failure of the containment due to overpressurization. This failure mode is expected to be the most likely one for large dry containments under accident conditions. Also during a core melt accident, a large quantity of hydrogen may be generated, giving the potential of a loss of containment integrity due to violent hydrogen combustion. Timely venting of the containment atmosphere can prevent overpressurization and may perhaps make the hydrogen situation in the containment less severe. This paper discusses the thermodynamic consequences of different vent strategies for a large German PWR during core melt accidents.     

Thermodynamic consequences of leakages in a double containment during severe accidents

For a large nuclear power plant under normal operating conditions a leakage rate for the containment of 0.25 vol.%/day is admissible. During a successfully controlled LOCA leakages of the containment will be released through filters by the annulus^* air exhausting system into the environment. During a core melt accident a pressurization of the containment has to be expected, which could lead to a failure of the containment due to overpressurization. When openings in the containment steel shell occur before a catastrophic failure, a depressurization into the annulus takes place. The area of the openings determines the depressurization rate and the thermodynamic conditions in the annulus. Furthermore the behaviour of the components being necessary for accident mitigation is influenced too. This paper discusses the thermodynamic consequences of leaks in the containment shell of a German PWR during a core melt accident. The results of those calculations are the necessary boundary condition for the estimation of fission product retention in the annulus.     

Numerical simulation of turbulent flow in a 37-rod bundle

The unsteady Reynolds averaged Navier–Stokes equations, combined with a Reynolds stress model, were solved numerically to determine fully developed isothermal turbulent flow in a 60° sector of a 37-rod bundle. It was found that this flow contained large-scale coherent structures, which affected strongly the local velocity fluctuations, especially near the gaps between rods or between rods and the surrounding wall. The time-averaged mean velocity and Reynolds stresses were in good agreement with experimental results in a similar channel. Coherent velocity fluctuations at different locations throughout the entire rod bundle were strongly correlated with each other.