Analysis of hydrogen risk mitigation with passive autocatalytic recombiner system in CPR1000 NPP during a hypothetical station blackout

Hydrogen safety has attracted extensive concern in severe accident analysis especially after the Fukushima accident. In this study, a similar station blackout as happened in Fukushima accident is simulated for CPR1000 nuclear power plant (NPP) model, with the computational fluid dynamic code GASFLOW. The hydrogen risk is analyzed with the assessment of efficiency of passive autocatalytic recombiner (PAR) system. The numerical results show that the CPR1000 containment may be damaged by global flame acceleration (FA) and local detonation caused by hydrogen combustion if no hydrogen mitigation system (HMS) is applied. A new condensation model is developed and validated in this study for the consideration of natural circulation flow pattern and presence of non-condensable gases. The new condensation model is more conservative in hydrogen risk evaluation than the current model in some compartments, giving earlier starting time of deflagration to detonation transition (DDT). The results also indicate that the PAR system installed in CPR1000 could prevent the occurrence of the FA and DDT. Therefore, HMS such as PAR system is suggested to be applied in NPPs to avoid the radioactive leak caused by containment failure.     

Demonstrative testing of honeycomb passive autocatalytic recombiner for nuclear power plant

Passive autocatalytic recombiners (PAR) are widely being used as hydrogen control device in the current and advanced light water reactors (ALWRs). The PARs lend themselves to very effective means of circumventing buildup of combustible or detonable hydrogen gas mixtures in the reactor containment. Korea Nuclear Technology Inc. has recently developed a new PAR system with high porous catalyst material in the shape of honeycomb. The honeycomb PAR catalyst has a design characteristic of improved hydrogen removal performance by increasing the surface area and enhancing the flow rate through the catalyst at the same time, without increasing PAR size compared to the conventional PARs. The experimental study was focused on the development of the hydrogen depletion rate correlation of the honeycomb PAR. Two different sizes of PARs, KPAR-40 and KPAR-T2, have been employed in the tailor-made Integral Test Facility and Performance Test Facility. Multiple tests were conducted in various conditions of pressure, temperature, and hydrogen concentration. The hydrogen depletion rate correlation and the PAR performance constant were determined from the experimental results, which can be applied to the honeycomb PAR system. Also determined was the scale effect due to the PAR size, i.e., the number of catalysts in a PAR.     

A study on evaluating a passive autocatalytic recombiner PAR-system in the PWR large-dry containment

A systematic step-by-step framework for analyzing hydrogen behavior and implementing passive autocatalytic recombiners (PARs) to mitigate hydrogen deflagration or detonation risk in severe accidents (SAs) is presented. The procedure can be subdivided into five main steps: (1) modeling the containment based on the plant design characteristics, (2) selecting the typical severe accident sequences, (3) calculating the hydrogen generation including in- and ex-vessel period, (4) modeling the gas distribution in containment atmosphere and estimating the hydrogen combustion modes and (5) evaluating the efficiency of the PAR-system to mitigate the hydrogen risk with and without catalytic recombiners, according to the safety criterion. For the Chinese 600MWe pressurized water reactor (PWR) with a large-dry containment, large break loss-of-coolant accident (LB-LOCA) is screened out as the reference severe accident sequence, considering the nature of hydrogen generation and the probabilistic safety assessment (PSA) result on accident sequences. The results show that a certain number of recombiners could remove effectively hydrogen and oxygen, to protect the containment integrity against hydrogen deflagration or detonation.     

Fluid dynamic analysis of a catalytic recombiner to remove hydrogen

Catalytic reacting surfaces in recombiners are a reliable way to remove hydrogen as well as other burnable gases like CO in a passive way from the containment atmosphere of a nuclear power plant (NPP) during an accident. Industrial mature designs are ready to be installed in large dry containments to act as a mitigation measure preferably in the case of severe accidents. Experiments have been carried out to study manifold aspects of recombiners like the efficiency of hydrogen removal, start-up conditions, poisoning, oxygen starvation, steam and water impact and others. Mostly the global behaviour of a given device in a larger environment has been investigated in order to demonstrate the effectiveness and to facilitate the derivation of simplified models for long-term severe accident analyses. There are a number of reasons to look inside a recombiner to understand the interaction of chemistry and flow. This can help in understanding the dependencies of non-measurable variables (e.g. reaction rate), of local surface temperatures and more. It also offers possibilities to increase the chemical efficiency by optimising the geometry properly. Computational fluid dynamics (CFD) codes are available to be used as development tools to include the specifics of catalytic surface reactions. The present paper describes the use of the code system CFX (CFX 4.1 Flow Solver User Guide. 1995, Computational Fluid Dynamics Services, AEA Technology plc, Oxfordshire, UK) for creating a recombiner model. Finally its comparison with existing test data is discussed.     

Simulation of hydrogen mitigation in catalytic recombiner. Part-II: Formulation of a CFD model

This paper aims at formulation of a model compatible with CFD code to simulate hydrogen distribution and mitigation using a Passive Catalytic Recombiner in the Nuclear power plant containments. The catalytic recombiner is much smaller in size compared to the containment compartments. In order to fully resolve the recombination processes during the containment simulations, it requires the geometric details of the recombiner to be modelled and a very fine mesh size inside the recombiner channels. This component when integrated with containment mixing calculations would result in a large number of mesh elements which may take large computational times to solve the problem. This paper describes a method to resolve this simulation difficulty. In this exercise, the catalytic recombiner alone was first modelled in detail using the best suited option to describe the reaction rate ( [Prabhudharwadkar et al., 2005] and [Prabhudharwadkar et al., 2011]). A detailed parametric study was conducted, from which correlations for the heat of reaction (hence the rate of reaction) and the heat transfer coefficient were obtained. These correlations were then used to model the recombiner channels as single computational cells providing necessary volumetric sources/sinks to the energy and species transport equations. This avoids full resolution of these channels, thereby allowing larger mesh size in the recombiners. The above mentioned method was successfully validated using both steady state and transient test problems and the results indicate very satisfactory modelling of the component.     

Simulation of hydrogen mitigation in catalytic recombiner: Part-I: Surface chemistry modelling

This paper aims at accurate modelling of a Passive Catalytic Recombiner used for hydrogen mitigation in the nuclear power plant containments. In order to assess the performance of the recombiner through numerical simulations, it is required to accurately predict the catalytic reactions. There are various detailed reaction mechanisms available in the literature for prediction of hydrogen-oxygen reaction over a platinum surface. While a single step reaction rate expression is always sought in order to obtain numerical predictions economically, a detailed reaction mechanism that includes several elementary reactions and intermediate species is likely to produce more accurate predictions. The paper compares the solution from two of competing models, one a single step reaction and the other a multiple reaction model. A new single step rate expression is also derived from the detailed mechanism after simplifying it for the present problem. The paper also considers the diffusion controlled model that assumes rapid reaction rates for which the surface chemistry is not required at all. In order to find the best suited approach to model the surface chemistry, CFD simulations were performed with FLUENT code using available experimental data from the literature. The current study reports comparison up to 4% H₂ mole fraction in dry air with catalyst temperature varying from 300 K to 800 K. It is demonstrated that the new single step model is able to satisfactorily predict the data as well as the detailed chemistry model. The diffusion controlled model is shown to over-predict the data.     

Analysis of SBLOCA on CPR1000 with a passive system

Since the Fukushima accident in 2011,more and more attention has been paid to nuclear reactor safety.A number of evolutionary passive systems have been developed to enhance the inherent safety of reactors.This paper presents a passive safety system applied on CPR1000,which is a traditional generation Ⅱ+ reactor.The passive components selected are as follows:(1) the reactor makeup tanks (RMTs);(2) the advanced accumulators (A-ACCs);(3) the passive emergency feedwater system (PEFS);(4)the passive depressurization system (PDS);(5) the incontainment refueling water storage tank (IRWST).The model of the coolant system and the passive systems was established by utilizing a system code (RELAP5/MOD3.3).The SBLOCA (small-break loss of coolant) was analyzed to test the passive safety systems.When the SBLOCA occurred,the RMTs were initiated.The water in the RMTs was then injected into the pressure vessel.The RMTs' low water level triggered the PDS,which depressurized the coolant system drastically.As the pressure of the coolant system decreased,the A-ACCs and the IRWST were put to work to prevent the uncovering of the core.The results show that,after the small-break loss-of-coolant accident,the passive systems can prevent uncovering of the core and guarantee the safety of the plant.     

An electrochemical hydrogen meter for measuring hydrogen in sodium using a ternary electrolyte mixture

An electrochemical sensor for measuring hydrogen concentration in liquid sodium that is based on a ternary mixture of LiCl, CaCl₂ and CaHCl as the electrolyte has been developed. DSC experiments showed the eutectic temperature of this ternary system to be ∼725 K. Impedance spectroscopic analysis of the electrolyte indicated ionic conduction through a molten phase at ∼725 K. Two electrochemical hydrogen sensors were constructed using the ternary electrolyte of composition 70 mol% LiCl:16 mol% CaHCl:14 mol% CaCl₂ and tested at 723 K in a mini sodium loop and at hydrogen levels of 60-250 ppb in sodium. The sensors show linear response in this concentration range and are capable of detecting a change of 10 ppb hydrogen in sodium over a background level of 60 ppb. Identification of this electrolyte system and its use in a sensor for measuring hydrogen in sodium are described in this paper.     

Assessment of passive autocatalytic recombiners' capabilities in case of Zr oxidation during station blackout scenario for TVSA and TVSM fuel assemblies of WWER 1000

This paper presents the investigation of Passive Autocatalytic Recombiners' (PARs) capabilities for hydrogen recombination in case of Station blackout scenario. The assessment was performed for both types of WWER fuel assemblies – the old Modernistic type of fuel assemblies (TVSM) and recently installed new Alternative type of fuel assemblies (TVSA) in Kozloduy NPP. The main difference between both types of fuel assemblies is the different geometries, masses of internals materials as well as different burnable poisons. To investigate the PARs' capabilities it has been performed comparison of fuel behaviour of both types of fuel assemblies.To perform this analysis it has been used MELCOR “input model” for Kozloduy Nuclear Power Plant (KNPP) WWER-1000. The model was developed at the Institute for Nuclear Research and Nuclear Energy (INRNE) for investigation of severe accident scenarios. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides' release outside fuel cladding.To assess the PARs' capabilities it was used the acceptance criterion for containment integrity to be 8% hydrogen concentration. This criterion was based on the decision of RSK (Germany commission for reactor safety).Generally, based on the performed analyses it was made a conclusion that using both types of fuel assemblies it was not disturbance of PARs' capabilities and safety criterion of NPP.     

Evaluation of analytically scaled models of a pressurized water reactor using the RELAP5/MOD3 computer code

This paper performs analytical evaluations for the potential distortions caused by the scaled models using RELAP5/MOD3 computer codes. By use of scaling analysis, two scaled models with the same volumetric ratio are constructed for the Korean next generation reactor (KNGR), which is an advanced light water reactor. The scaling methodology adopted in this paper preserves the two-phase natural circulation similarities between prototype and scaled models. One scaled model is at full height with reduced flow area. The other model is at reduced height with reduced flow area. By using appropriate scale factors the RELAP5/MOD3 input models are developed. Then, the transient responses of the two ideal scaled models are simulated for small break loss of coolant accidents (SBLOCAs) by using the RELAP5/MOD3 computer code. The transient responses of the two scaled models are compared with those of the prototype. The results indicate that qualitative and quantitative similarities are well preserved for both models during SBLOCA with different break sizes.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. 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.     

Reliability assessment of passive containment isolation system using APSRA methodology

In this paper, a methodology known as APSRA (Assessment of Passive System ReliAbility) has been employed for evaluation of the reliability of passive systems. The methodology has been applied to the passive containment isolation system (PCIS) of the Indian advanced heavy water reactor (AHWR). In the APSRA methodology, the passive system reliability evaluation is based on the failure probability of the system to carryout the desired function. The methodology first determines the operational characteristics of the system and the failure conditions by assigning a predetermined failure criterion. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the PCIS performance. APSRA proposes to compare the code predictions with the test data to generate the uncertainties on the failure parameter prediction, which is later considered in the code for accurate prediction of failure surface of the system. Once the failure surface of the system is predicted, the cause of failure is examined through root diagnosis, which occurs mainly due to failure of mechanical components. The failure probability of these components is evaluated through a classical PSA treatment using the generic data. The reliability of the PCIS is evaluated from the probability of availability of the components for the success of the passive containment isolation system.     

Momentum and mass exchange during refill in scaled models of a PWR

Interfacial momentum and mass exchange between the liquid and gas phases in a PWR downcomer were investigated. A new momentum transfer correlation was developed from air-water experiments in - and models of a PWR with standard and distorted geometries. The correlation is based on the Kutateladze parameter and indicates that the overall momentum transfer between the phases does not depend on scale for geometrically similar models. Interphase mass exchange has been included by evaluating the effective gas flow for momentum transfer. Predictions of the modified correlation agree quite well with experimental results of steam-water flows in three different scaled models of PWR's.     

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.     

Failure probability evaluation of passive system using fuzzy Monte Carlo simulation

Passive systems have become an inherent feature of the advanced reactors. The main reason being the passive systems are, theoretically, more reliable than the active ones. Nevertheless the passive system may fail to fulfill its mission not only because of a consequence of classical mechanical failure of component (passive or active) of the passive system, but also due to the deviation from expected behavior due to physical phenomena mainly related to thermal hydraulic or due to different boundary or initial conditions. In this paper the methodology used for performing the passive system reliability analysis has been discussed. A case study on passive decay heat removal system (PDHRS) of large sized pressurized heavy water reactor (PHWR) has been discussed. Thermal hydraulic analysis have been carried out by using RELAP5 code to generate the response surface (from various ranges of identified key parameter values), keeping the criterion as clad surface temperature exceeding certain critical value. Some uncertainties, due to incomplete information, cannot be handled satisfactorily in the probability theory and the fuzzy set theory is more appropriate. In this study the random variables are considered as fuzzy numbers and the fuzzy set theory is employed. In addition, the Monte Carlo simulation technique is utilized to evaluate the probability of failure of system.     

Solution of the isotopic depletion equations using decomposition method and analytical solutions

In this paper an analytical calculation of the isotopic depletion equations is proposed, featuring a chain of major isotopes founding a typical PWR reactor. Part of this chain allows feedback reactions of (n, 2n) type. The method is based on decoupling the equations describing feedback from the rest of the chain by using the decomposition method, with analytical solutions for the other isotopes present in the chain. The method was implemented in a PWR reactor simulation code that makes use of the nodal expansion method (NEM) to solve the neutron diffusion equation, describing the spatial distribution of neutron flux inside the reactor core. Because isotopic depletion calculation module is the most computationally intensive process within simulation systems of nuclear reactor core, it is justified to look for a method that is both efficient and fast, with the objective of evaluating a larger number of core configurations in a short amount of time.     

Reliability analysis of a passive cooling system using a response surface with an application to the flexible conversion ratio reactor

A risk-informed methodology is applied to the selection of an ultimate heat sink for a Passive Secondary Auxiliary Cooling System. The reliability of the chosen design during the bounding transient, a station blackout, is calculated. The methodology considers both active component failures and the potential for inadequate cooling due to adverse thermal-hydraulic conditions. A response surface is developed as a surrogate for the thermal-hydraulic code and used for uncertainty propagation. The uncertainty introduced by the use of the response surface itself is explored. Two sensitivity studies are performed. The first study measures the sensitivity of peak clad temperature to initial ambient conditions and system degradation. The second study explores the sensitivity of system reliability to code error.     

Two types of a passive safety containment for a near future BWR with active and passive safety systems

The paper presents two types of a passive safety containment for a near future BWR. They are named Mark S and Mark X containment. One of their common merits is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. The PCV pressure can be moderated within the design pressure. Another merit is the capability to submerge the PCV and the RPV above the core level. The third merit is robustness against external events such as a large commercial airplane crash. Both the containments have a passive cooling core catcher that has radial cooling channels. The Mark S containment is made of reinforced concrete and applicable to a large power BWR up to 1830 MWe. The Mark X containment has the steel secondary containment and can be cooled by natural circulation of outside air. It can accommodate a medium power BWR up to 1380 MWe. In both cases the plants have active and passive safety systems constituting in-depth hybrid safety (IDHS). The IDHS provides not only hardware diversity between active and passive safety systems but also more importantly diversity of the ultimate heat sinks between the atmosphere and the sea water. Although the plant concept discussed in the paper uses well-established technology, plant performance including economy is innovatively and evolutionally improved. Nothing is new in the hardware but everything is new in the performance.     

非能动核电站安全壳外壁下降水膜的稳定性分析

阐述非能动核电站安全壳外壁冷却水膜稳定性的重要意义,探讨临界水膜厚度和临界空气流速的计算方法及其在非能动核电站中的适用性,并应用水膜脱离模型和表面波模型,探讨维持水膜稳定、避免水膜脱离和水膜破裂的条件,为各种非能动核电站安全壳外壁水膜稳定性的试验研究奠定基础。