Conceptual Strategy for Mitigating the Risk of Hydrogen as an Internal Hazard in Case of Severe Accidents at Nuclear Power Plant Considering Existing Risks and Uncertainties Associated with the Use of Traditional Strategies

Hydrogen challenge mitigation stands as one of the main objectives in the management of severe accidents at Nuclear Power Plants (NPPs). Key strategies for hydrogen control include atmospheric inertization and hydrogen removal with Passive Autocatalytic Recombiners (PARs) being a commonly accepted approach. However, an examination of PAR operation specificity reveals potential inefficiencies and reliability issues in certain severe accident scenarios. Moreover, during the in-vessel stage of severe accident development, in some severe accident scenarios PARs can unexpectedly become a source of hydrogen detonation. The effectiveness of hydrogen removal systems depends on various factors, including the chosen strategies, severe accident scenarios, reactor building design, and other influencing factors. Consequently, a comprehensive hydrogen mitigation strategy must effectively incorporate a combination of strategies rather than be based on one strategy, taking into consideration the probabilistic risks and uncertainties associated with the implementation of PARs or other traditional methods. In response to these considerations, within the framework of this research it has been suggested a conceptual strategy to mitigate the hydrogen challenge during the in-vessel stage of severe accident development.

Influence of active and passive equipment for advanced pressurized water reactor on thermal hydraulic and source term behavior in severe accidents

Extensive studies have been carried out on the behavior of core degradation and fission products of common pressurized water reactors(PWRs).However,few of them have investigated the relationship between thermal hydraulic and fission product behavior in advanced passive PWRs.Due to the impact of thermal hydraulic be-haviors in different accident sequences on the release and transportation of fission products,an integrated severe accident analysis(ISAA)code with highly coupled thermal hydraulic and source term calculations is required to simultaneously analyze thermal hydraulic and source term behavior.For advanced passive PWRs,important safety systems that may affect the behavior of the core and fission products should be considered.It is therefore necessary to simulate the thermal hydraulic and fission product behavior of advanced passive PWRs.In this study,the ISAA code is adopted to simulate the occurrence of a hypothetical double ended cold leg LBLOCA of HPR1000 in three scenarios of equipment failure.The results show that the high-temperature fuel rods and cladding ma-terials exhibit delayed failure at the lower position of the active core,whereas earlier failure at higher position during the reflooding.Active and passive equipment affects fuel temperature,the oxidation conditions of the fuel,the interaction of fission products and structural materials,and the state of the fuel,thereby affecting the release of fission products in the fuel.HPR1000 only relies on passive equipment to relieve the core degradation in severe accidents,realize the in-vessel retention of melt,and eliminate the ex-vessel release possibility of fission product.It is hoped that the results can provide references for HPR1000 to formulate the severe accident management guidelines(SAMG).

Uncertainty and sensitivity analysis of in-vessel phenomena under severe accident mitigation strategy based on ISAA-SAUP program

The phenomenology involved in severe accidents in nuclear reactors is highly complex.Currently,integrated analysis programs used for severe accident analysis heavily rely on custom empirical parameters,which introduce considerable uncertainty.Therefore,in recent years,the field of severe accidents has shifted its focus toward applying uncertainty analysis methods to quantify uncertainty in safety assessment programs,known as“best estimate plus uncertainty(BEPU).”This approach aids in enhancing our comprehension of these programs and their further development and improvement.This study concentrates on a third-generation pressurized water reactor equipped with advanced active and passive mitigation strategies.Through an Integrated Severe Accident Analysis Program(ISAA),numerical modeling and uncertainty analysis were conducted on severe accidents resulting from large break loss of coolant accidents.Seventeen uncertainty parameters of the ISAA program were meticulously screened.Using Wilks'formula,the developed uncertainty program code,SAUP,was employed to carry out Latin hypercube sampling,while ISAA was employed to execute batch calculations.Statistical analysis was then conducted on two figures of merit,namely hydrogen generation and the release of fission products within the pressure vessel.Uncertainty calculations revealed that hydrogen production and the fraction of fission product released exhibited a normal distribution,ranging from 182.784 to 330.664 kg and from 15.6 to 84.3%,respectively.The ratio of hydrogen production to reactor thermal power fell within the range of 0.0578–0.105.A sensitivity analysis was performed for uncertain input parameters,revealing significant correlations between the failure temperature of the cladding oxide layer,maximum melt flow rate,size of the particulate debris,and porosity of the debris with both hydrogen generation and the release of fission products.

Experience gained in analyzing severe accidents for WWER RP using CC SOCRAT

The current Russian regulatory documents on the safety of nuclear power plant(NPP)specify the requirements regarding design basis accidents(DBAs)and beyond design basis accidents(BDBAs),including severe accidents(SAs)with core meltdown,in NPP design(NP-001-15,NP-082-07,and others).For a rigorous calculational justification of BDBAs and SAs,it is necessary to develop an integral CC that will be in line with the requirements of regulatory documents on verification and certification(RD-03-33-2008,RD-03-34-2000)and will allow for determining the amount of data required to provide information within the scope stipulated by the requirements for the structure of the safety analysis report(SAR)(NP-006-16).The system of codes for realistic analysis of severe accidents(SOCRAT)(formerly,thermohydraulics(RATEG)/coupled physical and chemical processes(SVECHA)/behavior of core materials relocated into the reactor lower plenum(HEFEST))was developed in Russia to analyze a wide range of SAs at NPP with water-cooled water-moderated power-generating reactor(WWER)at all stages of the accident.Enhancements to the code and broadening of its applicability are continually being pursued by the code developers(Nuclear Safety Institute of the Russian Academy of Sciences(IBRAE RAN))with OKB Gidropress JSC and other organizations.Currently,the SOCRAT/В1 code can be used as a base tool to obtain realistic estimates for all parameters important for computational justification of the reactor plant(RP)safety at the in-vessel stage of SAs with fuel melting.To perform analyses using CC SOCRAT/В1,the experience gained during execution of thermohydraulic codes is applied,which allows for minimizing the uncertainties in the results at the early stage of an accident scenario.This study presents the results of the work performed in 2010–2020 in OKB Gidropress JSC using the CC SOCRAT/В1.Approaches have been considered to develop calculational models and analyze SAs using CC SOCRAT.This process,which is clearly structured in OKB Gidropress JSC,provides a noticeable reduction in human involvement,and reduces the probability of erroneous results.This study represents the principal results of the work performed in 2010–2020 in OKB Gidropress JSC using the CC SOCRAT,as well as a list of the tasks planned for 2021–2023.CC SOCRAT/B1 is used as the base thermohydraulic SAs code.

Severe accident risks from external events

This paper reviews the early development of design requirements for seismic events in USA early developing nuclear electric generating fleet. Notable safety studies,including WASH-1400,Sandia Siting Study and the NUREG1150 probabilistic risk study,are briefly reviewed in terms of their relevance to extreme accidents arising from seismic and other severe accident initiators. Specific characteristic about the nature of severe accidents in nuclear power plant (NPP) are reviewed along with present day state-of-art analysis methodologies (methods for estimation of leakages and consequences of releases (MELCOR) and MELCOR accident consequence code system (MACCS)) that are used to evaluate severe accidents and to optimize mitigative and protective actions against such accidents. It is the aim of this paper to make nuclear operating nations aware of the risks that accompany a much needed energy resource and to identify some of the tools,techniques and landmark safety studies that serve to make the technology safer and to maintain vigilance and adequate safety culture for the responsible management of this valuable but unforgiving technology.

A novel approach for radionuclide diffusion in the enclosed environment of a marine nuclear reactor during a severe accident

A severe accident in a marine nuclear reactor leads to radionuclide leakage,which causes hidden dangers to workers and has adverse effects of environmental pollution.It is necessary to propose a novel approach to radionuclide diffusion in a confined environment after a severe accident in a marine nuclear reactor.Therefore,this study proposes a new method for the severe accident analysis program MELCOR coupled with computational fluid dynamics scSTREAM to study radioactive diffusion in severe accidents.The radionuclide release fraction and temperature calculated by MELCOR were combined with the scSTREAM calculations to study the radionuclide diffusion behavior and the phenomenon of radionuclide diffusion in different space environments of the reactor under the conditions of varying wind velocities of the ventilation system and diffusion speed.The results show that the wind velocity of the ventilation system is very small or zero,and the turbulent diffusion of radionuclides is not obvious and diffuses slowly in the form of condensation sedimentation and gravity settlement.When the wind speed of the ventilation system increases,the flow of radionuclides meets the wall and forms eddy currents,affecting the time variation of radionuclides diffusing into chamber 2.The wind velocity of the ventilation system and the diffusion speed has opposite effects on the time variation trend of radionuclide diffusion into the four chambers.

Status of Severe Accident Management Guidelines at Kozloduy Nuclear Power Plant

The objective of this paper is to present the current organization of the Emergency Procedures including Emergency Operating Procedures (EOP) and Severe Accident Management Guidelines (SAMG) in Kozloduy Nuclear Power Plant (KNPP) as a function of the severity of the accident conditions. Special attention is paid to SAMG. It is described when the SAMG are used and at which conditions in a transition between the EOPs and the SAMG should be made. The Critical Safety Function Restoration Guidelines and their connections with SAMGs and EOPs are also discussed. The arrangement of SAMG is described in detail, since in the KNPP exist 2 types of SAMGs for Main Control Room (MCR) and for the Accident Management Centre (AMC) and they contain the same strategies, but they are different in format. Both types are symptom oriented procedures, but those for MCR are in 2-column-format with interconnections, whereas those for the AMC are developed in a logical manner and simplified for people, who take decisions. In the paper, they are also discussed the adopted strategies in existing SAMG that should be followed to recover from a damaged core condition and to prevent or mitigate the release of fission products. In the paper, they are also described a number of technical measures for management and mitigation of severe accidents, which are implemented in KNPP before and after the Fukushima accident. Many of them are common for WWER-1000 type of reactors, but some of them are unique and plant specific. This information can be useful for operators of other WWER type reactors or even PWR reactors.

中法关于核能与环境的联合研究

本文聚焦核能的环境影响问题,将解决如下核能发电相关的主要环境问题:(1)控制正常运行条件下核设施的放射性排放,评估其非放射性环境影响(取水与非放射性水的排放);(2)乏燃料与放射性废物的长期管理,主要是通过地质处置库处理的乏燃料或放射性废物;(3)防止和缓解严重核事故以及核事故的放射性释放;(4)改善核安全水平,以限制核能的环境影响,提升公众的核能接受度。核能的温室气体排放水平非常低,可以根据需求供应大规模、可调度电力,在此方面核能具有独特的优势。在正常运行工况下,核电站每年释放到周围公众的有效辐射剂量可以忽略不计。国际社会已经开展了大量的努力,以确定可持续管理地质处置条件下高放长寿命放射性废物的方法。过去几次严重核事故中获得的经验为核能生产相关的安全问题提供了经验,也促成了重要的安全改进,其中包括反应堆的设计和运行管理方面的改善以及事故管理指导方针的制定等。事实证明,这些经验是非常宝贵的。严重事故的环境风险已经被大幅降低,相关的规约也已经建立起来,以最大限度地减少严重核事故条件下放射性物质的释放,并避免大规模的人员疏散。还需要继续采取措施,改善反应堆的安全性,提升核工业与核监管机构的透明度,以进一步降低核能的环境影响。

Fukushima accident study using MELCOR

The accidents at the Fukushima Daiichi nuclear power station stunned the world as the sequences played out over severals days and videos of hydrogen explosions were televised as they took place. The accidents all resulted in severe damage to the reactor cores and releases of radioactivity to the environment despite heroic measures had taken by the operating personnel. The following paper provides some background into the development of these accidents and their root causes,chief among them,the prolonged station blackout conditions that isolated the reactors from their ultimate heat sink - the ocean. The interpretations given in this paper are summarized from a recently completed report funded by the United States Department of Energy (USDOE).

Simulation of hydrogen distribution and effect of Engineering Safety Features (ESFs) on its mitigation in a WWER-1000 containment

In this study, thermal–hydraulic parameters inside the containment of aWWER-1000/v446 nuclear power plant are simulated in a double-ended cold leg accident for short and long times (by using CONTAIN 2.0 and MELCOR 1.8.6 codes), and the effect of the spray system as an engineering safety feature on parameters mitigation is analyzed with the former code. Along with the development of the accident from design basis accident to beyond design basis accident, the Zircaloy–steam reaction becomes the source of in-vessel hydrogen generation. Hydrogen distribution inside the containment is simulated for a long time (using CONTAIN and MELCOR), and the effect of recombiners on its mitigation is analyzed (using MELCOR). Thermal–hydraulic parameters and hydrogen distribution profiles are presented as the outcome of the investigation. By activating the spray system, the peak points of pressure and temperature occur in the short time and remain belowthe maximumdesign values along the accident time. It is also shown that recombiners have a reliable effect on reducing the hydrogen concentration below flame propagation limit in the accident localization area. The parameters predicted by CONTAIN and MELCOR are in good agreement with the final safety analysis report. The noted discrepancies are discussed and explained.

The General Design and Technology Innovations of CAP1400

A historical review of in-vessel melt retention （IVR） is given, which is a severe accident mitigation mea- sure extensively applied in Generation III pressurized water reactors （PWRs）. The idea of IVR actually originated from the back-fitting of the Generation 11 reactor Loviisa WER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse APIO00, the Korean APR1400 as well as Chinese advanced PWR designs HPRIO00 and CAP1400. The most influential phe- nomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel （RPV）. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contrib- ute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.

Moving particle semi-implicit simulation on the molten Wood's metal downward relocation process

In the case of a severe accident involving nuclear reactors,an important aspect that should be considered is the leakage of molten material from the inside of the reactor into the environment.These molten materials damage other reactor components,such as electrical tubes,grid plates and core catchers.In this study,the moving particle semi-implicit(MPS)method is adopted and improved to analyze the twodimensional downward relocation process of molten Wood’s metal as a representation of molten material in a nuclear reactor.The molten material impinges the Wood’s metal plate(WMP),which is mounted on a rigid dummy stainless steel in a cylindrical test vessel.The breaching process occurs because of heat transfer between the molten material and WMP.The formed breach areas were in good agreement with the experimental results,and they showed that the molten Wood’s metal spread above the WMP.The solid WMP fraction decreased with time until it reached the termination time of the simulation.The present results show that the MPS method can be applied to simulate and analyze the downward relocation process of molten material in the grid plate of a nuclear reactor.

Predicting the severity of traffic accidents on mountain freeways with dynamic traffic and weather data

Traffic accident severity prediction is essential for dynamic traffic safety management.To explore the factors influencing the severity of traffic accidents on mountain freeways and to predict the severity of traffic accidents,four models based on machine learning algorithms are constructed using support vector machine(SVM),decision tree classifier(DTC),Ada_SVM and Ada_DTC.In addition,random forest(RF)is used to calculate the importance degree of variables and the accident severity influences with high importance levels form the RF dataset.The results show that rainfall intensity,collision type,number of vehicles involved in the accident and toad section type are important variables influencing accident severity.The RF feature selection method improves the classification performance of four machine leaming algorithms,resulting in a 9.3%,5.5%,7.2% and 3.6% improvement in prediction accuracy for SVM,DTC,Ada_SVM and Ada_DTC,respectively.The combination of the Ada_SVM integrated algorithm and RF feature selection method has the best prediction performance,and it achieves 78.9% and 88.4% prediction precision and accuracy,respectively.

Nuclear power plant life extension:How aging affects performance of containments & other structures

This paper focuses on how aging can affect performance of safety-related structures in nuclear power plant (NPP). Knowledge and assessment of impacts of aging on structures are essential to plant life extension analysis,especially performance to severe loadings such as loss-of-coolant-accidents or major seismic events. Plant life extension issues are of keen interest in countries (like the United States) which have a large,aging fleet of NPPs. This paper addresses the overlap and relationship of structure aging to severe loading performance,with particular emphasis on containment structures.

Prediction of Accident Severity Using Artificial Neural Network: A Comparison of Analytical Capabilities between Python and R

Large amount of data has been generated by Organizations. Different Analytical Tools are being used to handle such kind of data by Data Scientists. There are many tools available for Data processing, Visualisations, Predictive Analytics and so on. It is important to select a suitable Analytic Tool or Programming Language to carry out the tasks. In this research, two of the most commonly used Programming Languages have been compared and contrasted which are Python and R. To carry out the experiment two data sets have been collected from Kaggle and combined into a single Dataset. This study visualizes the data to generate some useful insights and prepare data for training on Artificial Neural Network by using Python and R language. The scope of this paper is to compare the analytical capabilities of Python and R. An Artificial Neural Network with Multilayer Perceptron has been implemented to predict the severity of accidents. Furthermore, the results have been used to compare and tried to point out which programming language is better for data visualization, data processing, Predictive Analytics, etc.

Experimental Investigation on Fuel Coolant Interaction Using Simulant Ceramic Melts in Water: Insights and Conclusions

Steam explosion is one of the crucial and poorly understood phenomena which may occur during severe accident scenario and may lead to containment failure. In spite of several experimental and analytical studies, the root cause of steam explosion has not been understood. Recent claims in the literature suggest that the presence of fine fragmentation during steam explosion causes its occurrence. In order to investigate this and understand the root cause of steam explosion, series of experiments were performed with 50 g to 2500 g of CaO-B2O3, a corium simulant in 4.5 litre of water. It was observed that steam explosion may occur even in the absence of fine fragments, which is contrary to the claims in the literature. To investigate further, conversion efficiency analysis was performed. This suggested that the amount of thermal energy converted to mechanical energy is more important deciding factor in explaining the occurrence of steam explosion. The present study discusses the importance of conversion efficiency in deciding steam explosion and also gives a new perspective to look at steam explosion phenomenology.

Analysis of Pb-Bi Vessel Wall Ablation Experiment with High Temperature Liquid by MPS Method

In a severe accident of a light water reactor, ablation of the RPV （reactor pressure vessel） lower head by corium is a key phenomenon, which affects progression of the accident. The MPS （moving particle semi-implicit） method is one of particle methods that calculate behavior of incompressible fluid by semi-implicit method. In preceding studies, the MPS method has been extensively studied and developed for simulations of different phenomena involved in severe accident of nuclear reactors. This paper aims to investigate whether the MPS method is capable of analyzing the lower head ablation phenomenon. The small-scale experiment carried out at CRIEPI （Central Research Institute of Electric Power Industry） using Pb-Bi vessel and silicone oil was analyzed for the validation of the MPS method. The MPS analysis well reproduced the experimental phenomena qualitatively. However, with respect to some quantitative results, more investigation such as influence of the calculation particle size is necessary.

Analysis of QUENCH-ACM Experiments Using SCDAP/RELAP5

The QUENCH experimental programme at Karlsruhe under severe accident conditions, but while the geometry is still Institute of Technology investigates heat-up and reflooding of a core mainly rod-like. The recent QUENCH-ACM series of experiments, comprising QUENCH-12 （El 10 cladding alloy）, -14 （M5 alloy） and -15 （Zirlo^TM alloy）, together with QUENCH-06 （reference case, Zircaloy-4 alloy） addressed the effect of alternative cladding materials on oxidation and quenching under similar conditions. Superficial inspection of the experimental results reveals only minor differences in the thermal and oxidation response, except for the much larger hydrogen release during reflood in QUENCH-12. Post-test calculations were performed using a version of SCDAP/RELAP5/MOD3.2, modified to represent the QUENCH facility and to invoke alternative oxidation correlations. The calculations agreed rather well with experiments QUENCH-06, -14 and -15, but the significant hydrogen release during reflood in QUENCH-12 was not captured. Closer examination of the experimental results reveals further differences between QUENCH-12 which may be linked to breakaway oxidation of the E110 cladding. The analyses support the heuristic observation that there was no major difference between the influence of Zircaloy-4, M5 or ZirloTM, but the E-110 exhibited a contrasting behaviour with a consequent impact on the reflooding.

Quench Front Progression in a Superheated Porous Medium： Experimental Analysis and Model Development

In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a ＂debris bed＂. The typical particle size in a debris bed might reach few millimeters （characteristic length-scale： 1-5 mm）. The two-phase flow model for reflood of the degraded core is briefly introduced in this paper. It is implemented into the ICARE-CATHARE code, developed by IRSN （Institut de radioprotection et de surete nucleaire）, to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN sets up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, and validate safety models. The PRELUDE program studies the complex two phase flow （water/steam）, in a porous medium （diameter 180 mm, height 200 mm）, initially heated to a high temperature （400℃ or 700℃）. On the basis of the experimental results, thermal hydraulic features at the quench front have been analyzed. The two-phase flow model shows a good agreement with PRELUDE experimental results.

AREVA Control Room Accident Filtration System CRAFT

The AREVA CRAFT （control room accident filtration system） is a solution that maintains the proper air conditions in the main control room and emergency control facilities by filtering the air and removing noble gases in case of a severe accident in a nuclear power plant with increased activity concentration in the plant environment.