Severe Accident Situations Research Articles

High-temperature Mechanical Analysis of Discharge Pipeline of Pressurizer

The severe accident is the beyond basis design accident due to several initial faults. The discharge pipeline of pressurizer is mainly used for pressure relief and evacuation of hydrogen in the primary circuit in severe accident. It can reduce the possibility of high-pressure core melt and the explosion of hydrogen. But in severe accident, the highest temperature of gas in the pipeline can reach 1200℃, which far exceeds the design temperature of pressurized water reactor. So it may cause the buckling or rupture of the pipeline. The pipeline may loss its discharge function. In this passage, we study the resistance of the discharge pipeline to high temperature in severe accident from the aspects of severe accident situation, high-temperature mechanical properties of pipeline material, the temperature field and stress of the pipeline, the evaluation of the stress and so on. In order to demonstrate the function of a certain nuclear power plant discharge pipeline, RCC-MR which is suitable for the high-temperature evaluation, is choose to evaluate the stress, the creep and the buckling of the pipeline. The results of the evaluation show that the discharge pipeline in this nuclear power plant can meet the functional requirements in the severe accidents.

Empirical modeling of ex-vessel debris bed formation in a pre-flooded reactor cavity

To secure the containment integrity of a nuclear power plant (NPP) in severe accident situation of ex-vessel corium release into the reactor cavity, a pre-flooding strategy of a reactor cavity has been adopted in various NPPs, including Korean PWRs. Molten corium released into the pre-flooded water pool is fragmentized and forms a particulate debris bed at the bottom of the cavity. To prevent the containment failure, it is important to properly remove the heat from the debris bed, which can be accomplished by water ingression into the bed. The water ingression characteristics are highly dependent on the debris bed shape. In this study, several main parameters affecting the debris bed shape were determined with respect to phenomena of free-fall & sedimentation, particle dispersion by steam spike and convection flow, and self-leveling. Then empirical model was developed and validated using the results of DEFCON (Debris bed Formation and COolability experimeNt) experiments. The model’s error was within ±10% when only free-fall & sedimentation were present. When steam-spike or convection was added to free-fall & sedimentation condition, the error increased to within ±20%. When all three were simultaneously present (free-fall & sedimentation, steam-spike, and convection), the model predicted most of the experimental results within ±20% error, except for several cases where convection had a significant impact. The effect of self-leveling itself was within ±10%. Considering all phenomena related to particulate debris bed formation, most cases had an error of ±20%. In addition, the debris bed shape formed by poly-disperse particles was predicted appropriately by the mode using the mass mean diameter of the particle size distribution, with an error range of ±20%. Consequently, it is expected that the proposed model of the debris bed shape can be utilized in an integrated severe accident analysis code for the evaluation on the ex-vessel debris bed coolability in the pre-flooded reactor cavity.

Development of a decision-making support system for the technical support center based on analysis of human performance under severe accident situation

After the Fukushima accident, hardship of group decision making under severe accident situation is one of human factor issues emerged by Fukushima accident. After core damage, the responsibility for plant control is shifted from main control room (MCR) operators to technical support center (TSC) members who need to follow severe accident management guideline (SAMG). Decision making is necessarily required from this time to establish a suitable strategy because SAMG is not a procedure but it is a broad guideline so it cannot give a precise correct answer to TSC members for mitigating accident sequence like EOP.The developed DM support system can shorten the group DM time through a semi automatic algorithm. In addition, it is possible to gather the opinions of all members of TSC by excluding unnecessary obstructive human factors under uncertain situations like lack of information.

Classification of ablation mode during impact of hot liquid jet on a solid

The ablation consecutive to hot jet impingement is a safety issue for future Sodium Fast Reactors (SFR) core-catcher that can occur during relocation of molten fuel (corium) under severe nuclear accident situation. Much is still unknown on the ablation phenomenon especially explanations on cavity shape are lacking. However, these data are required to specify the core-catchers which will be used in future nuclear reactors’ vessel to prevent the corium from drilling through the confinement vessel during severe accident. To tackle this subject, data from experiments are obtained and analyzed to identify first order physical mechanisms at stake, and their links to geometry of the cavity. Two ablation mechanisms are noticed, the film ablation regime, for which liquid exits the cavity as a liquid film followed by the pool effect for which the cavity is filled with liquid. The analysis of results shows that the cavity shape is fixed during the film ablation regime and translates as ablation proceeds. Modes of liquid exit from the cavity are analyzed as well as the shapes the cavity assumes. Liquid/air interface temperature is also determined, in the film ablation regime. An explanation of cavity shape is presented. Conditions based on dimensionless numbers are put forward to differentiate between different cavity shapes and liquid exit modes. A first model for transition between film and pool effect ablation regimes is presented. It is the first time to the best of our knowledge that such analyses are undertaken. These give new tools for ablation risk assessment.

Minimum diameter limit of particle size distribution and its effect on coolability of debris bed

The particle size distribution (PSD) of the debris bed is a major parameter influencing its coolability with respect to the two-phase pressure drop of the porous media. Most studies have focused on representative particle diameter as opposed to size distribution. Based on a severe accident situation (strong two-phase convection), we introduced a rearrangement of small particles and developed a corresponding size distribution formula by modifying the Rosin-Rammler distribution.A minimum particle diameter model was proposed as a function of dryout heat flux (DHF) based on a minimum fluidization velocity concept. In addition, a truncated Rosin-Rammler distribution was developed and a Sauter Mean Diameter (SMD) was formulated in the form of the gamma function with a new parameter: the minimum particle diameter (Dmin). Sensitivity analysis of DHF in a 1-D top flooded debris bed showed two notable results for the effect of Dmin on coolability. First, Dmin significantly reduced DHF uncertainty due to the distribution constant. Moreover, the calculated DHF increased by introducing the concept of a minimum particle diameter. In conclusion, treatment of small particles in the debris bed is essential for determining debris bed coolability. Additionally, the truncated Rosin-Rammler distribution could be utilized to provide the precise SMD in a porous media.

Adsorption behavior of methyl iodide on a silver ion-exchanged ZSM-5

Abstract An organic iodide, especially methyl iodide (CH3I), is one of the most safety concerns against a nuclear power plant severe accident because it is highly toxic and easily inhaled. Due to its high volatility, the dry filtration methods are generally employed to capture the gas phase CH3I. Although it is important to accurately assess the performance of dry filter in a postulated severe accident situation, no sufficiently accurate adsorption kinetic behaviors have been reported yet. In this study, we designed the highly accurate experimental set-up, thermogravimetric analysis-gas chromatography coupled system, which overcame the limitations of previous studies, and performed the CH3I adsorption tests on a silver ion-exchanged ZSM-5 (Ag-ZSM-5) zeolite at a wide temperature range (50–150 °C) under the postulated severe accident condition. The fresh and adsorbed Ag-ZSM-5 samples were characterized by powder XRD, elemental analysis, and SEM-EDS, and the further CH3I adsorption behaviors were discussed in depth. From the overall results, we obtained the accurate adsorption rates and activation energies, and an understanding of chemisorption and physisorption behaviors. Our findings provide the useful fundamentals for a model development of dry filter performance assessment.

원자력발전소의 중대사고 상황 관리를 지원하기 위한 생태학적 정보디스플레이 설계를 위한 정보요건의 도출

Objective: This study aims to identify and organize information requirements for designing ecological displays supporting responsive actions under severe accident situations of nuclear power plants (NPPs).BRBRBackground: One of the requisites for supporting human operators in handling severe accident situations in complex systems such NPPs is to design information displays that satisfy information needs of human operators. Ecological Interface Design (EID) has been regarded as a viable approach for developing user-centered information displays. This study follows the design process and principles of the EID framework.BRBRMethod: We firstly identified functional knowledge structure of a work domain related to severe accident situations of NPPs through work domain analysis (WDA) using abstraction hierarchy (AH). After validating the WDA model through expert review, we finally specified information requirements for ecological displays.BRBRResults: This study identified and organized information requirements for designing information displays that can be effectively used for supporting human operators in severe accident situations in NPPs.BRBRConclusion: A set of information requirements obtained from this study would be useful in designing information displays supporting human operators in coping with severe accident situations in NPPs.BRBRApplication: This study can contribute towards enhancing the safety of NPPs by providing a set of meaningful information requirements that should be offered to human operators in severe accident situations where their tasks could be very cognitively challenging.

Experimental Investigation of Decontamination Factor Dependence on Aerosol Concentration in Pool Scrubbing

Because a pool scrubbing is important for reducing radioactive aerosols to the environment for a nuclear reactor in a severe accident situation, many researches have been performed. However, decontamination factor (DF) dependence on aerosol concentration was seldom considered in an aerosol number concentration with limited aerosol coagulation. To investigate an existence of DF dependence on the concentration, DF in a pool scrubbing with 2.4 m water submergence was derived from aerosol measurements by light scattering aerosol spectrometers. It was observed that DF increased monotonically with decreasing particle number concentration in a constant thermohydraulic condition: a gradual increase from 10 to 32 in the range of 1.3×1011 - 8.0×1011/m3 at the inlet and a significant increase from 32 to 77 in the range of 3.6×1010 - 1.3×1011/m3. Two validation experiments were conducted in the range with the gradual DF increase to confirm whether the DF dependence is a real pool scrubbing phenomenon. In addition, characteristics of the DF dependence in different water submergences were investigated experimentally. It was found that the DF dependence became more significant in higher water submergence. Significant DF dependence was observed in the condition of the water submergence higher than 1.6 m and the inlet particle number concentration less than around 1×1011 /m3. It is recommended to perform further analysis for the DF dependence mainly in such condition since it could make a difference to both experiment and model of the pool scrubbing.

Analysis of composition and temperature of debris bed based on settling characteristics of relocated core debris in pool type SFR

The hypothetical core meltdown accident in sodium cooled fast reactors (SFR) is a severe accident situation during which the energetic dispersion of core materials and their relocation may pose a threat to the integrity of reactor containment. During core meltdown accident, part of the molten core materials relocated downwards would be fragmented in the lower plenum sodium pool and settle over the in-vessel core catcher plate and forms a particulate debris bed. The recriticality potential and coolability of the debris bed are strongly influenced by the composition of fragmented particles of different core materials and their size distribution. In this paper, simplified models have been developed to evaluate the incipient conditions i.e., composition profile and initial temperature of the particulate debris bed formed over an in-vessel core catcher plate following the core meltdown accident in pool type SFR. The particle mixture settling process subsequent to the fragmentation of molten core materials in sodium pool has been analyzed to evaluate the composition profile of the debris bed. A 1-D model for particle motion through the stagnant fluid under the influence of gravity has been considered to evaluate the settling characteristics of core materials debris (mainly UO2 and stainless steel) in liquid sodium for a wide range of possible sizes. By extending this model, the differential settling behaviour of mixture of UO2 and stainless steel particles in liquid sodium pool has been investigated. The composition profile of the debris bed along its thickness has been evaluated for the material relocation scenario considered with various mass ratios of stainless steel and UO2 mixture using the published experimental results based particle size distributions. Further, a 1-D transient conduction model is implemented to estimate the temperature of debris particles when they settle on the core catcher plate. Simulations revealed that the particles will be quenched to near the temperature of the liquid sodium in lower pool while reaching the core catcher. The results are useful in assessing the coolability characteristics of debris bed formed on the core catcher and its recriticality potential in the analysis of the Post Accident Heat Removal.

Neutron-gamma flux and dose calculations for feasibility study of DISCOMS instrumentation in case of severe accident in a GEN 3 reactor

The present paper presents the study carried out in the frame of the DISCOMS project, which stands for “DIstributed Sensing for COrium Monitoring and Safety”. This study concerns the calculation of the neutron and gamma radiations received by the considered instrumentation during the normal reactor operation as well as in case of a severe accident for the EPR reactor, outside the reactor pressure vessel and in the containment basemat. This paper summarizes the methods and hypotheses used for the particle transport simulation outside the vessel during normal reactor operation. The results of the simulations are then presented including the responses for distributed Optical Fiber Sensors (OFS), such as the gamma dose and the fast neutron fluence, and for Self Powered Neutron Detectors (SPNDs), namely the neutron and gamma spectra. Same responses are also evaluated for severe accident situations in order to design the SPNDs being sensitive to the both types of received neutron-gamma radiation. By contrast, fibers, involved as transducers in distributed OFS have to resist to the total radiation gamma dose and neutron fluence received during normal operation and the severe accident.

Out-containment mitigation of gaseous iodine by alkaline spray in severe accident situation

Radioactive iodine is one of the major constituents among fission products released into the atmosphere from damaged/leaked containment during severe core melt accident. This type of release can create significant impact on human health through thyroid cancer. To prevent the dispersion of such hazardous radionuclide, an emergency response spray system can be deployed outside the conventional physical barrier-the reactor containment. The purpose of this study is to examine the use of spray technology to mitigate the consequence of severe accidents outside the containment. The efficiency of spray method in the removal of the released fission products, in particular, gaseous iodine, was investigated under the hypothetical accident situation. Effects of different parameters such as spray flow rate, spray solution pH and gas release rate were investigated as part of the study. A laboratory-scale experimental system combining spray chamber with a nozzle and gaseous iodine release port was setup to examine the performance of the proposed system. The alkaline (aqueous NaOH.Na2S2O3) solutions of varying pH combined with water were used in the spray chamber to examine capture efficiency for gaseous iodine. The captured and un-captured fractions were determined by the UV–Visible spectroscopy and ion selective electrode method. Experimental results showed that use of proper spray solution pH and flow rate is important in determining iodine removal efficiency. The removal efficiency was also affected by the release rate of gaseous iodine. Temperature variation was also found to have effect on the volatility of the captured iodine in the sprayed solution.

Prediction of hydrogen concentration in containment during severe accidents using fuzzy neural network

Recently, severe accidents in nuclear power plants (NPPs) have become a global concern. The aim of this paper is to predict the hydrogen buildup within containment resulting from severe accidents. The prediction was based on NPPs of an optimized power reactor 1,000. The increase in the hydrogen concentration in severe accidents is one of the major factors that threaten the integrity of the containment. A method using a fuzzy neural network (FNN) was applied to predict the hydrogen concentration in the containment. The FNN model was developed and verified based on simulation data acquired by simulating MAAP4 code for optimized power reactor 1,000. The FNN model is expected to assist operators to prevent a hydrogen explosion in severe accident situations and manage the accident properly because they are able to predict the changes in the trend of hydrogen concentration at the beginning of real accidents by using the developed FNN model.

Transient refractory material dissolution by a volumetrically-heated melt

The present work addresses the question of corium–ceramic interaction in a core catcher during a core-melt accident in a nuclear power plant. It provides an original insight into transient aspects concerning dissolution of refractory material by a volumetrically heated pool. An experiment with simulant material (LIVECERAM) is presented.Test results clearly show that dissolution of solid refractory material can occur in a non-eutectic melt at a temperature which is lower than the melting temperature of the refractory material. During the dissolution transient, the interface temperature rises above the liquidus temperature, corresponding to the instantaneous average composition of the melt pool.With constant power dissipation in the melt and external cooling of the core-catcher, a final steady-state situation is reached. Dissolution stops when the heat flux (delivered by the melt to the refractory) can be removed by conduction through the residual thickness of the ceramic, with Tinterface=Tliquidus (calculated for the average composition of the final liquid pool). The final steady state corresponds to a uniform pool composition and uniform interface temperature distribution. Convection in the pool is governed by natural thermal convection and the heat flux distribution is therefore similar to what would be obtained for a single component pool.An interpretation of the experiment with two model-based approaches (0D and 1D) is presented. The mass transfer kinetics between the interface and the bulk is controlled by a diffusion sublayer within the boundary layer. During the dissolution transient, the liquid composition at the interface is concentrated in the refractory species. During the transient, the interface temperature is equal to the liquidus temperature corresponding to the local and instantaneous composition of the liquid at the interface.Regarding the design of a protective layer made of refractory materials, we can answer the question of how much ceramic can be dissolved and its impact on melt temperature evolution during the dissolution process. It also impacts on subsequent corium solidification since the additional mass of dissolved ceramic leads to increased volume of the molten material, significantly increasing the time required for complete solidification. For the long term, ceramic material does not offer better confinement than a crust made of solidified corium.This work served as support to a generalisation of the model of transient evolution of interface temperature in various severe accident situations (Seiler and Combeau, 2014).

Transient interface temperature on a vertical surface in multi-component solid–liquid systems with volume heating. Application to various severe accident situations

The question of the transient interface conditions in various severe accident situations, involving multi-component and multi-phase material mixtures has until now remained largely unresolved. The interface temperature controls melt temperature and transient heat-flux distribution. The present paper proposes a new approach to the transient interface temperature, drawing on the numerous past efforts.A single relation is proposed which applies to various severe accident situations: (i) thermal-hydraulic steady state for in-vessel retention (IVR); (ii) transient solidification; (iii) transient concrete ablation (MCCI); and (iv) refractory material (ceramic) ablation (core-catcher retention, liner design). This approach yields the following important conclusions:-When a thermal-hydraulic steady state can be reached (external cooling, in case of IVR and melt stabilization in an externally cooled core-catcher), the melt-solid interface temperature tends towards the liquidus temperature corresponding to the melt composition.-During the ablation transient:•the interface temperature is lower than the melt liquidus temperature if the wall-melting temperature is less than the melt liquidus temperature (MCCI),•the interface temperature is higher than the melt liquidus temperature if the melting temperature of the wall exceeds the melt liquidus (oxidic corium-refractory ceramic interaction).The new interface model also suggests that simultaneous ablation of two similar walls (for instance: two concrete walls) may be subject to instabilities which can potentially result in complete arrest of the ablation of one of the walls.In the case of simultaneous ablation of two walls of different nature, the wall with the lowest melting temperature is preferentially ablated during the transient phase.

PREDICTION OF THE REACTOR VESSEL WATER LEVEL USING FUZZY NEURAL NETWORKS IN SEVERE ACCIDENT CIRCUMSTANCES OF NPPS

Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

Monitoring Performance of the Cameras under the High Dose-Rate Gamma Ray Environments

CCD/CMOS cameras, loaded on a robot system, are generally used as the eye of the robot and monitoring unit. A major problem that arises when dealing with images provided by CCD/CMOS cameras under severe accident situations of a nuclear power plant is the presence of speckles owing to the high dose-rate gamma irradiation fields. To use a CCD/CMOS camera as a monitoring unit in a high radiation area, the legibility of the camera image in such intense gamma-radiation fields should therefore be defined. In this paper, the authors describe the monitoring index as a figure of merit of the camera's legibleness under a high dose-rate gamma ray irradiation environment. From a low dose-rate (10 Gy h) to a high dose-rate (200 Gy h) level, the legible performances of the cameras owing to the speckles are evaluated. The numbers of speckles generated by gamma ray irradiation in the camera image are calculated by an image processing technique. The legibility of the sensor indicator (thermo/hygrometer) owing to the number of speckles is also presented.

Smart Sensing of the RPV Water Level in NPP Severe Accidents Using a GMDH Algorithm

The reactor pressure vessel (RPV) water level is critical information for confirming the condition of core cooling in severe accident situations. However, the measured RPV water level signal cannot be trusted during severe accidents due to the unknown integrity of the sensor. In this study, the RPV water level was predicted under severe accident conditions using a group method of data handling (GMDH) algorithm. The prediction model was developed using data obtained from numerical simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code and validated using independent test data. The developed GMDH model performed very well. In addition, to investigate the effect of uncertainties in the input variables, the model was tested using input data with an artificially added random error. It was accurate enough to predict the RPV water level in severe accident situations when the RPV water level sensor cannot be trusted. Therefore, the developed GMDH model will be helpful for providing effective information for operators in severe accident situations.

福岛事故后对国内严重事故管理发展的思考

日本福岛核电厂事故之后，在世界范围内，包括核安全当局、核电业主、各核能研究单位以及普通民众，对核电厂严重事故的关注都提高到了一个前所未有的程度。本文对国内外严重事故管理情况以及严重事故管理发展的趋势进行了介绍，并对国内严重事故管理指南发展提出建议。 After the Fukushima accident, in the worldwide, the nuclear safety authorities, the owners of nuclear power, and nuclear research units, as well as ordinary people are concerned about the serious accident of the nuclear power plant. This paper firstly introduces the severe accident management situation at home and abroad as well as the development trend of severe accident management, and then make suggestions to the development of severe accident management guidelines in China.

A Performance Evaluation of a Notebook PC under a High Dose-Rate Gamma Ray Irradiation Test

We describe the performance of a notebook PC under a high dose-rate gamma ray irradiation test. A notebook PC, which is small and light weight, is generally used as the control unit of a robot system and loaded onto the robot body. Using TEPCO’s CAMS (containment atmospheric monitoring system) data, the gamma ray dose rate before and after a hydrogen explosion in reactor units 1–3 of the Fukushima nuclear power plant was more than 150 Gy/h. To use a notebook PC as the control unit of a robot system entering a reactor building to mitigate the severe accident situation of a nuclear power plant, the performance of the notebook PC under such intense gamma-irradiation fields should be evaluated. Under a similar dose-rate (150 Gy/h) gamma ray environment, the performances of different notebook PCs were evaluated. In addition, a simple method for a performance evaluation of a notebook PC under a high dose-rate gamma ray irradiation test is proposed. Three notebook PCs were tested to verify the method proposed in this paper.

SAMEX: A severe accident management support expert

A decision support system for use in a severe accident management following an incident at a nuclear power plant is being developed which is aided by a severe accident risk database module and a severe accident management simulation module. The severe accident management support expert (SAMEX) system can provide the various types of diagnostic and predictive assistance based on the real-time plant specific safety parameters. It consists of four major modules as sub-systems: (a) severe accident risk data base module (SARDB), (b) risk-informed severe accident risk data base management module (RI-SARD), (c) severe accident management simulation module (SAMS), and (d) on-line severe accident management guidance module (on-line SAMG). The modules are integrated into a code package that executes within a WINDOWS XP operating environment, using extensive user friendly graphics control. In Korea, the integrated approach of the decision support system is being carried out under the nuclear R&D program planned by the Korean Ministry of Education, Science and Technology (MEST). An objective of the project is to develop the support system which can show a theoretical possibility. If the system is feasible, the project team will recommend the radiation protection technical support center of a national regulatory body to implement a plant specific system, which is applicable to a real accident, for the purpose of immediate and various diagnosis based on the given plant status information and of prediction of an expected accident progression under a severe accident situation.