Analysis Of Accident Scenarios Research Articles

Fully ceramic microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Design basis accidents and fuel cycle cost

Fully ceramic microencapsulated (FCM) fuel may enhance the already strong inherent reactor safety characteristics of a conventional High-Temperature Gas-Cooled Reactor (HTGR). FCM fuel uses an SiC matrix that exhibits higher stability under irradiation with limited swelling relative to a conventional graphite matrix. Additionally, the SiC matrix is expected to exhibit improved mechanical performance relative to graphite. The feasibility of FCM-fueled HTGRs has been explored in a previous study (Lu et al., 2018), and three FCM fuel concepts were identified to be able to maintain the cycle length of the reference conventional General Atomics 350 MWt prismatic modular HTGR.In this work, the previous investigation of the reactor core safety characteristics of HTGRs fueled by these three FCM fuels was significantly extended to detailed thermal hydraulics and neutronics analysis of normal operation and design basis accident scenarios. We found that these FCM-fueled cores would have a maximum fuel temperature ∼20–40 K higher than that of the reference HTGR core at the beginning of cycle (BOC) and ∼30–60 K higher at the end of cycle (EOC) during normal operating conditions. Pressurized (P-) and depressurized (D-) loss of forced cooling (LOFC) accidents as well as the control rod withdrawal accident were explored. The FCM-fueled cores would at most have a maximum fuel temperature ∼30–50 K higher than that of the reference core in a P-LOFC accident, ∼10–30 K in a D-LOFC accident and ∼30–50 K in a rod withdrawal accident. The conclusion of the study is that FCM fuel has a small impact on the reactor performance and safety characteristics during normal operation and design basis accident conditions within the existing and generic 350 MWt prismatic modular HTGR design we considered.Additionally, a fuel cycle cost comparison was also performed in the current study. We found that the FCM fuels may be up to 40%–74% more expensive than the conventional fuel. Further analysis of the economics of FCM fuel are needed, because the uncertainties in the assumptions for this study are high.

A dynamic and quantitative risk assessment method with uncertainties for offshore managed pressure drilling phases

Drilling into offshore oil and gas fields often meets many challenges and uncertainties, such as a narrow window of drilling fluid density and shallow gas zones. Managed pressure drilling (MPD) techniques are increasingly used as alternatives to conventional drilling operations to manage such extreme conditions and reduce drilling costs and risks. Many safety and operational issues related to MPD process need to be investigated more thoroughly. Well kick is considered a typical hazardous event that may occur at different drilling phases, and such an event is prone to develop into a blowout. During offshore drilling phases, the risk of accidents may change with time, and such a dynamic characteristic should be recorded in risk assessment. This study presents a method for the application of dynamic Bayesian networks (DBNs) in conducting accident scenario analysis and dynamic quantitative risk assessment for MPD safety. This method can model the influence of uncertain risk factors, which have been ignored in existing research, by introducing an additional probability parameter. The effects of degradation are also taken into account. DBN inference is adopted to perform quantitative risk analysis and dynamic risk evolution. Then, the vulnerable root causes are identified by sensitivity analysis for accident prevention and mitigation measures. Well kick for four drilling cases is analyzed as a case study to demonstrate the feasibility of the proposed method. Three-step analysis partially validates the correctness and rationality of the proposed DBN model.

Adaptive simulation for failure identification in the Advanced Lead Fast Reactor European Demonstrator

The identification undesired or abnormal states of a nuclear power plant is of primary importance for defining accident prevention and mitigation actions. To this aim, computational models and simulators are frequently employed, as they allow to study the system response to different operational conditions. For complex systems like the nuclear power plants, this is in general challenging because the simulation tools are i) high-dimensional; ii) black-box; iii) dynamic and iv) computationally demanding.In this paper, an adaptive simulation framework recently proposed by some of the authors is tailored for the analysis of accident scenarios involving the control system of the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED).The results confirm that the adaptive simulation framework proposed is effective in identifying critical regions of operation with a limited number of calls to the computationally expensive model. The time of occurrence and magnitude of the failures of the components of the control system are identified as key factors to characterize the critical regions. In particular, it is shown that the order of occurrence of the components’ failures strongly affects the evolution of the accident scenarios.

Dynamic safety risk modeling of process systems using bayesian network

Process complex systems in particular oil and gas plants due to dealing with hazardous materials at severe process conditions are much prone to catastrophic accidents. In this context, safety risk analysis is a crucial tool to develop effective strategies to prevent accident and provide mitigative measures. Dynamic risk analysis (DRA) is one of the most practical approaches for risk analysis that helps provide safer operations of complex process systems. The present work is aimed at demonstrating the application of an integrated DRA approach to comprehensive quantitative modeling and analysis of the both aspects of risk, that is, probability and consequence assessments. In this approach, first, the worst case scenario is identified and then a robust tool is developed for dynamic accident scenario modeling and risk assessment by means of Bayesian Network. This approach is applied to risk analysis of a flammable liquid storage system at a gas refinery. The work provides valuable information on the identification and comprehensive analysis of worst case accident scenarios, their main consequences, critical basic events, and minimal cut sets which lead to accident scenarios and also for dynamic updating of probabilities and risk. The obtained results are more appropriate and rigorous to developing preventive and mitigative strategies for potential accident scenarios and thus increase the safety level in the complex process systems. © 2017 American Institute of Chemical Engineers Process Saf Prog 36: 399–407, 2017

Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria

The Nuclear Regulatory Commission (NRC) has considered revision of 10-CFR-50.46C rule (Borchard and Johnson, 2013, “10 CFR 50.46c Rulemaking: Request to Defer Draft Guidance and Extension Request for Final Rule and Final Guidance,” U.S. Nuclear Regulatory Commission, Washington, DC.) to account for burn-up rate effects in future analysis of reactor accident scenarios so that safety margins may evolve as dynamic limits with reactor operation and reloading. To find these limiting conditions, both cladding oxidation and maximum temperature must be cast as functions of fuel exposure. To run a plant model through a long operational transient to fuel reload is computationally intensive, and this must be repeated for each reload until the time of the accident scenario. Moreover for probabilistic risk assessment (PRA), this must be done for many different fuel reload patterns. To perform such new analyses in a reasonable amount of computational time with good accuracy, Idaho National Laboratory (INL) has developed new multiphysics tools by combining existing codes and adding new capabilities. The parallel highly innovative simulation INL code system (PHISICS) toolkit (Rabiti et al., 2016, “New Simulation Schemes and Capabilities for the PHISICS/RELAP5-3D Coupled Suite,” Nucl. Sci. Eng., 182(1), pp. 104–118; Alfonsi et al., 2012, “PHISICS Toolkit: Multi-Reactor Transmutation Analysis Utility—MRTAU,” PHYSOR 2012 Advances in Reactor Physics Linking Research, Industry, and Education, Knoxville, TN, Apr. 15–20.) for neutronic and reactor physics is coupled with the reactor excursion and leak analysis program—three-dimensional (RELAP5-3D) (The RELAP5-3D© Code Development Team, 2014, “RELAP5-3D© Code Manual Volume I: Code Structure, System Models, and Solution Methods,” Rev. 4.2, Idaho National Laboratory, Idaho Falls, ID, Technical Report No. INEEL-EXT-98-00834.) for the loss of coolant accident (LOCA) analysis and reactor analysis and virtual-control environment (RAVEN) (Alfonsi et al., 2013, “RAVEN as a Tool for Dynamic Probabilistic Risk Assessment: Software Overview,” 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Sun Valley, ID, May 5–9, pp. 1247–1261.) for the probabilistic risk assessment (PRA) and margin characterization analysis. For RELAP5-3D to process a single sequence of cores in a continuous run required a sequence of restarting input decks, each with different neutronics or thermal-hydraulic (TH) flow region and culminating in an accident scenario. A new multideck input processing capability was developed and verified for this analysis. The combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical pressurized water reactor (PWR).

Arctic shipping accident scenario analysis using Bayesian Network approach

This paper presents a methodology for the analysis of Arctic shipping accident scenarios using Bayesian Networks (BN). The proposed methodology is applied to a scenario involving a collision between a vessel and an iceberg. The study aims to identify the most significant causative factors to the potential accident scenarios. It is achieved by undertaking a sensitivity analysis study. The results inform the development of measures to avoid and control accidents during Arctic shipping.

Availability analysis of safety critical systems using advanced fault tree and stochastic Petri net formalisms

Failure scenarios analysis constitutes one of the cornerstones of risk assessment and availability analysis. After a detailed review of available methods, this paper identified two distinct formalisms to analyze failure scenarios and systems' availability: generalized stochastic Petri nets (GSPN) and Fault tree driven Markov processes (FTDMP). The FTDMP formalism is a combination of the Markov process and the fault tree. This aims to overcome fault tree limitations while maintaining the use of deductive logic. The GSPN is a Petri net with probabilistic analysis using Monte Carlo simulation. The effectiveness of both methods is studied through an emergency flare system including a knockout drum. It is observed that GSPN provides a robust and reliable mechanism for accident scenario analysis. It provides additional information such as events' frequencies at operating and failing modes and expected occurrence timing and durations resulting from different complex sequences. Even for multi-state variables which could be used to design a safety management system. Although FTDMP is a powerful formalism, it provides limited information.

Thermohydraulic Analysis of Accident Scenarios of a Fusion DEMO Reactor Based on Water-Cooled Ceramic Breeder Blanket: Analysis of LOCAs and LOVA

Thermohydraulic analysis of postulated accidents will identify system responses to accident scenarios and aid in developing design of safety systems and strategies to prevent or mitigate accident propagation. This paper reports analyses of four accident scenarios of a fusion DEMO reactor based on water-cooled ceramic-pebble breeder blanket. The accidents analyzed, which were selected based on the previous logical accident analysis, are ex-vessel loss of coolant of the primary cooling system, in-vessel loss of coolant of the first wall (FW) cooling pipes, loss of coolant in blanket modules, and loss of vacuum. The analyses have identified thermohydraulic responses of the DEMO systems to these accidents, pressure loads to confinement barriers for radioactive materials. Effectiveness of the safety systems and the integrity of the primary and final (secondary) confinement barriers against the accidents are discussed. As for the final confinement barrier, we show for the first time that implementation of a pressure suppression system (PSS) to the cooling system vault and a vacuum breaker to the tokamak pit is effective in significantly keeping the integrity of the final confinement barrier against the ex-vessel loss-of-coolant and loss-of-vacuum accidents, respectively. As for the primary confinement barrier, we show for the first time that limitation of the number of blankets from which a helium purge gas line collects the bred tritium will be a key technical issue to prevent propagation of loss of coolant in a blanket box through the purge gas line and then suppress the pressurization of the vacuum vessel (VV). For the in-vessel loss of coolant of the FW cooling pipes, further optimization of the PSS or design solutions regarding in-vessel components and plasma control will be necessary to decrease the pressure load to the VV and ensure the integrity of the primary confinement barrier.

Fire risk analysis based on one-dimensional model in nuclear power plant

Fire probabilistic safety assessment (fire PSA) is developed to give the insight of nuclear power plant risk induced by fire accident and the main contributors, and fire accident scenario analysis is one of the important parts in the work to get the key factors such as time to target damage and time to detection. The thermal–hydraulic model simulating fire accident should have enough accuracy and high speed to satisfy the request of fire PSA. Many researches indicate that in a large volume the hot fluid will concentrate in the upper part while the cold fluid will be in the lower part because of density difference under fire condition, that is, the gradients of temperature and of some other parameters in vertical direction are much greater than in horizontal direction. Based on such thermal stratification theory a one-dimensional model is developed, and the buoyant jet is used to simulate the process of heated air flowing up. In this paper a fire in a compartment is analyzed based on one-dimensional model and the temperature distribution is obtained, the results are compared with those of commercial software such as FDT, CFAST and FDS, then the time to target damage is evaluated based on results of different models and the fire non-suppression probabilities are evaluated. The results illustrate that one-dimensional model has better accuracy than FDT and CFAST since such a model can simulate the thermal stratification and natural circulation which exist in the volume simultaneously. Moreover when the fire power is low, the thermal stratification is apparent and air temperature in the hot upper layer is much lower than the critical value of target damage, the one-dimensional model has enough accuracy to be used directly in fire PSA. While the thermal stratification will be weakened when the fire power increases because of the effects of radiation heat transfer and the entrainment by the jet, so more detailed model such as FDS is needed for such situations, however the results of one-dimensional model can give the advice for the proper simulation time of FDS to improve the calculation efficiency when the upper part temperature is close to or higher than the critical value.

ICONE23-2096 COMPARISON BETWEEN PARCS/MCNP6 CODES AND SOME REFERENCE DATA ON VVER1000/V320 CORE

In response to the Fukushima accident the Research Centre Rez plans to simulate accident scenarios of NPP Temelin and NPP Dukovany analyzed in the stress tests ordered by the European Commission to all European NPPs. While in a number of safety analysis of various accident scenarios it is sufficient to use one point reactor kinetics there are selected types of accidents in which it is useful to model the space (3D) behaviour of neutron kinetics, in particular control rod ejections, boron dilution scenarios, including transitions from design basis to beyond design basis accidents. For example in order to analyse control rod ejections, boron dilution scenarios and local changes of moderator temperature one has to see the local changes (3D) of reactivity and power. In this paper the current model of VVER1000/V320 reactor created in PARCS 3.2 is benchmarked with MCNP6 code and some reference data from NPP Temelin. The geometry of fuel in the core used for the analyses comes from fuel assemblies of NPP Temelin (hexagonal shape). This benchmark may serve as one of validation tests of the PARCS code for VVER1000/V320.

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

Severe accident source terms for a sodium-cooled fast reactor

In order to support the demonstration of a risk-informed approach to the design optimization of a sodium-cooled fast reactor (SFR), it was necessary to make realistic estimates of the consequences of severe accident scenarios. This paper describes the database, models, and assumptions used to estimate the offsite consequences of characteristic severe accident scenarios. As required for comparison with the NRC’s technology neutral framework limit curve, the offsite dose at one mile from the plant boundary is calculated using conservative meteorology.The reference plant design is a 1000MWt pool-type design with metallic fuel. Because an integrated analysis tool comparable to MELCOR does not exist for SFR accident scenario analysis, it was necessary to write a computer code that would assess release of radionuclides from the fuel and transport within the reactor primary system and to link those analyses with results from existing computer codes that assess the dynamic response of the reactor, containment thermal–hydraulics, and radionuclide transport processes within the containment. The analyses indicate that the offsite source terms for SFR severe accident scenarios tend to be small because of the low melting temperature of the fuel, likelihood of significant retention of fission products within the sodium pool, augmentation of containment deposition processes by interaction with sodium oxide aerosols, and small driving force for release from the containment to the environment. A number of major sources of modeling uncertainty are identified as requiring further development effort. An integrated modeling capability, similar to the MELCOR code, is required to direct further research and ultimately to support the licensing of an SFR.

Safety analysis in process facilities: Comparison of fault tree and Bayesian network approaches

Safety analysis in gas process facilities is necessary to prevent unwanted events that may cause catastrophic accidents. Accident scenario analysis with probability updating is the key to dynamic safety analysis. Although conventional failure assessment techniques such as fault tree (FT) have been used effectively for this purpose, they suffer severe limitations of static structure and uncertainty handling, which are of great significance in process safety analysis. Bayesian network (BN) is an alternative technique with ample potential for application in safety analysis. BNs have a strong similarity to FTs in many respects; however, the distinct advantages making them more suitable than FTs are their ability in explicitly representing the dependencies of events, updating probabilities, and coping with uncertainties. The objective of this paper is to demonstrate the application of BNs in safety analysis of process systems. The first part of the paper shows those modeling aspects that are common between FT and BN, giving preference to BN due to its ability to update probabilities. The second part is devoted to various modeling features of BN, helping to incorporate multi-state variables, dependent failures, functional uncertainty, and expert opinion which are frequently encountered in safety analysis, but cannot be considered by FT. The paper concludes that BN is a superior technique in safety analysis because of its flexible structure, allowing it to fit a wide variety of accident scenarios.

Development of a IRSN code for dust mobilisation problems in ITER

The analysis of accident scenarios in the ITER Fusion facility such as loss of vacuum accidents (LOVA) or ingress of coolant accidents (ICE) is of primary importance in order to evaluate pressure loads generated by a potential dust explosion or mobilisation. Thus the DUST code for simulation of dust mobilisation has been developed in close collaboration between IRSN (France) and the Technical University of Cartagena (Spain), on the basis of the CAST3M code developed at the Commissariat à l’Energie Atomique (CEA, France). A high dilute model for the analysis of particle and gas mixtures has been implemented in the code. This type of mixtures is expected in the case of a LOVA inside ITER. The code will permit the analysis of dust spatial distribution, remobilisation and entrainment, and explosion. Special attention is paid to the study of the closure relationships which characterise the physical phenomena involved in this sort of problems and a model based on the force balance has been implemented as well. The implementation of the adhesion/re-suspension model is carefully described. To test the code qualitatively, two benchmarks have been studied taking into account the influence of different parameters. Some conclusions are finally drawn.

ＭＯＸ燃料加工施設における臨界事象発生頻度概略評価手法の開発

A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility.

Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

In 1975, as a result of a blockage of the coolant inlet flow, two plates of a fuel element of the BR2 reactor of the Belgian Nuclear Research Centre (SCK•CEN) were partially melted. The fuel element consisted of Al-clad plates with 90% 235U enriched UAl x fuel dispersed in an Al matrix. The element had accumulated a burn up of 21% 235U before it was removed from the reactor. Recently, the damaged fuel plates were sent to the hot laboratory for detailed PIE. Microstructural changes and associated temperature markers were used to identify several stages in the progression to fuel melting. It was found that the temperature in the center of the fuel plate had increased above 900–950 °C before the reactor was scrammed. In view of the limited availability of such datasets, the results of this microstructural analysis provide valuable input in the analysis of accident scenarios for research reactors.

Evaluation of remote pedestrian sensor system based on the analysis of car-pedestrian accident scenarios

There is a need to evaluate the effectiveness of a remote sensor system for pedestrian detection in a vehicle traffic environment. For this purpose, the car-pedestrian accident scenarios were analyzed using selected data from the Swedish Traffic Accident Data Acquisition (STRADA), a database developed by collecting accident data from the police and hospitals. The two most common scenarios were identified as cars entering and leaving intersections colliding with pedestrians crossing the road. The knowledge of these two scenarios was then developed in terms of the factors such as the pedestrian’s trajectory, the pedestrian’s speed, the car’s trajectory and the car’s velocity. Based on the developed knowledge, a mathematical model was presented and the remote pedestrian sensor system was evaluated by using this model. It was found that the sensor system can detect almost all the pedestrians in these two scenarios in time, in the case where the detective angle was larger than 60 degrees.

Comparison of techniques for accident scenario analysis in hazardous systems

In this paper, three accident scenario analysis techniques are presented and compared regarding their efficiency vs. the demanded resources. The complexity of modern industrial systems has prompted the development of accident analysis techniques that should thoroughly investigate accidents. The idea of criteria classification to fulfill this requirement has been proposed by other researchers and is examined here too. The comparison is done through the application of Event Tree analysis, Fault Tree analysis and Petri Nets technique—two relatively simple and a more demanding methodology—on the same hazardous chemical facility in view of analyzing an accident scenario of a hazardous transfer procedure. Accident scenario analysis techniques are essential not only in learning lessons from unfortunate events in the chemical industry but also in preventing the occurrence of such events in the future and in communicating risk more efficiently.

안전한 제품을 설계하기 위한 새로운 제품위험분석 방법

Today we are observing a lot of injuries. casualties. and property losses that are mainly caused by the defects of products. In order to derive safety designs. which minimize the possibility of such product liability-related accidents. we need to take intoaccount the user-product interaction as an important part of the danger factor analysis. Existing risk analysis techniques. however. have some limitations in detecting comprehensive danger factors that are peculiarly involved in human errors and the functional defects of products. Researches on danger factor analysis regarding the user-product interaction have been carried out actively in ergonomics.br/ In this paper. we suggest a novel product risk analysis technique. which is more objective and systematic compared to the previous ones. by combining a modified TAFEl (Task Analysis For Error Identification) technique with SASA (Systematic Approach to Accident Scenario Analysis) technique. By applying this technique to the product design practice in industry. corporations will be able to improve the product safety. consequently strengthening the competitiveness.br/

Analysis of Severe Accident Scenarios and Proposals for Safety Improvements for ADS Transmuters with Dedicated Fuel

So-called dedicated fuels will be utilized to obtain maximum transmutation and incineration rates of minor actinides (MAs) in accelerator-driven systems (ADSs). These fuels are characterized by a high-MA content and the lack of the classical fertile materials such as 238U or 232Th. Dedicated fuels still have to be developed; however, programs are under way for their fabrication, irradiation, and testing. In Europe, mainly the oxide route is investigated and developed. A dedicated core will contain multiple “critical” fuel masses, resulting in a certain recriticality potential under core degradation conditions. The use of dedicated fuels may also lead to strong deterioration of the safety parameters of the reactor core, such as, e.g., the void worth, Doppler or the kinetics quantities, neutron generation time, and βeff. Critical reactors with this kind of fuel might encounter safety problems, especially under severe accident conditions. For ADSs, it is assumed that because of the subcriticality of the system, the poor safety features of such fuels could be coped with. Analyses reveal some safety problems for ADSs with dedicated fuels. Additional inherent and passive safety measures are proposed to achieve the required safety level. A safety strategy along the lines of a defense approach is presented where these measures can be integrated. The ultimate goal of these measures is to eliminate any mechanistic severe accident scenario and the potential for energetics.