Accident Mitigation Research Articles

A human reliability analysis methodology based on an extended Phoenix method for severe accidents in nuclear power plants: Qualitative analysis framework

The mitigation of the Fukushima accident exposed the detrimental action of human errors once again. However, deficiencies in existing human reliability analysis (HRA) methods prevent the effective prediction and analysis of potential human errors during the mitigation of severe accidents following core damage. To address this problem, this study presents a qualitative analysis framework for emergency response personnel (ERP) reliability analysis under severe accident conditions. This framework is developed using the Phoenix HRA method in conjunction with analyzing the ERP response characteristics and human-related issues from previous accidents, literature surveys, drill observations and interviews, as well as an assessment of the applicability of existing HRA methods. The analysis framework includes the establishment of an ERP response tree and a human failure fault tree (FT), as well as the identification and compilation of ERP cognitive functions, human error modes (HEMs), performance-influencing factors (PIFs), and the construction of Bayesian belief networks, to provide methodological guidance for the prevention of human errors under severe accident conditions.

Development of geopolymer based sacrificial materials for GEN IV severe accident mitigation

• We have analysed the needs for ALLEGRO GEN IV reactor core catcher materials. • We have developed a new synthetic route for geopolymer materials. • The geopolymers has been designed as sacrificial materials in GEN IV core catcher. • The chemical and mechanical properties and irradiation stability has been examined. The development of Generation IV nuclear power plants includes the considerations about a hypothetical severe accident. Therefore, passive safety systems for severe accident mitigation can be engineered and designed as an integral part of the plant based on its needs and specifications. Presented work describes the development of geopolymer sacrificial materials for core catcher of a gas cooled fast reactor. Knowing the corium properties at the time of the reactor vessel failure we have designed a inovative sacrificial material using specific powder fillers powder (MgO, Fe 2 O 3 , Ca 3 (PO 4 ) 2 , Al 2 O 3 , Gd 2 O 3 ), which were immobilized by a silico-aluminate geopolymer matrix. The composite geopolymers were characterised by means of the solid-state analysis. Moreover, the gamma-radiation stability was examined together with the mechanical properties. We found that the geopolymer matrix can easily embed the powder fillers with high homogeneity. The mechanical properties are stable under gamma irradiation.

Use of Bayesian networks and improved SPAR-H for quantitative analysis of human reliability during severe accidents mitigation process in nuclear power plant

ABSTRACT Human reliability analysis (HRA) for severe accidents is an important component of a Level 2 probabilistic risk assessment (PRA) for nuclear power plants. In this study, a Bayesian network (BN) is used to construct a qualitative causal model of human error during severe accidents by identifying the main cognitive functions (MCFs), crew failure modes (CFMs), and performance influencing factors (PIFs) for emergency personnel implementing severe accident management guidelines (SAMGs) after core damage. This qualitative model is used to develop a quantitative HRA method for severe accidents in nuclear power plants by improving the definition of performance-shaping factors (PSFs), levels, and multiplier criteria of the standardized plant analysis risk human reliability analysis (SPAR-H) method and applying BNs to quantify the failure probability of human performance. The usability of the proposed method is verified using a case study of emergency personnel following SAMGs to depressurize a reactor cooling system.

Comparative Study of Physical Models for Particle Sedimentation Using SIMMER Code

Abstract In the context of improved safety requirements for Generation IV sodium-cooled fast reactors (SFR), an innovative severe accident mitigation scenario is being investigated. In the French frame of SFR research, the mitigation strategy consists of transfer tubes and a core catcher. The transfer tubes are dedicated to discharge molten fissile materials from the core center region and to guide them toward the core catcher where long-term cooling and subcritical state may be assured. The physical phenomena occurring during the discharge process are introduced in this paper. The current demonstration of the mitigation strategy uses best-estimate calculations with the reference computer code SIMMER. Previous analyses showed that the material discharge through the transfer tubes might be efficient; however, uncertainties of SIMMER approach are identified on the molten material mobility during the relocation process. It is related to a blockage formation due to particulate solid debris accumulation inside the transfer tube, in case of low energy accumulation in the degraded fuel, is believed to originate from the solid particle treatment in the code. As the performance of mitigation strategy strongly depends on the mobility of the relocating mixture, the most predictive behavior of particle flows is of great importance to SFR safety. Therefore, the SIMMER modeling of such flows is analyzed in this work. The first verification and validation test cases regarding the gravitational settling of particle clouds at varying volume fractions are presented. Recommendations for reactor calculations and first orientations for future research and development are highlighted.

Thermal-hydraulic analysis of FAIDUS design for severe accident mitigation strategy in a medium size sodium-cooled fast reactor

The impediment of core disruption is realized through the provision of diverse and redundant safety features in the design complemented by the use of inherent passive mechanisms that respond to the accident condition by ensuring convenient dispersal of molten core material from the core in the early stage of the accident. In light of this, a promising design strategy Fuel Assembly Inner DUct Structure (FAIDUS) core concept is proposed to enhance the axial discharge capability of molten fuel from core for the exclusion of core wide melt pool formation. The sequence of event during the course of relocation in the FAIDUS is analysed employing a transient integrated thermal-hydraulic model. The effectual performance of various FAIDUS design options in mitigating the accident is evaluated. Based on this, the preferred design choice conforming the laid down criteria is selected in the present study. The simulation results suggest that FAIDUS design concept delivers the best alternative way for extenuating accident progress through a comprehensive melt relocation strategy of greater reliability in an intrinsic way.

Study of the ablation consecutive to jet impingement on a meltable solid – Application to SFR core-catcher

In the scope of safety studies for future sodium cooled fast reactors (SFR), the use of discharge tubes and in-vessel core-catcher are foreseen for severe accident mitigation. During relocation of corium onto the core-catcher, complex thermal hydraulics phenomena can occur. This work focuses on the ablation of the core-catcher by the jet of corium. For that purpose, as experiments using corium and Na are difficult to achieve, we performed experiments using simulant materials: water jet on transparent ice. Transparent ice allows for ablation cavity real-time recording. Two main regimes were identified: a film flowing regime and a pool regime. We also studied the influence of temperature (30, 50, 70 °C) and velocity (1,2.5,5,7.5,10m/s) of the jet. Comparison with existing data have been done and finally, a scaling law in the first regime, for which the ablation rate is constant, has been found.

SBO analysis of a generic PWR-900 with ASTEC and MELCOR codes

After the Fukushima accident, the interest of the public to nuclear safety has growth and the international technical nuclear community has increased his attention in the investigation and the characterization of Severe Accident (SA) scenarios. In order to simulate the different, complex and multi-physical phenomena involved in a SA, computational tools, known as SA codes, have been developed in the last decades. In order to give some insights on the modelling capabilities of these tools and the differences in the calculation results, also related to the user-effect, an analysis of an unmitigated Station Black Out (SBO) occurring in a generic Western three-loops PWR 900 MWe has been carried out by the authors in the framework of the NUGENIA TA-2 ASCOM project. The simulation results of ASTEC code (study carried out with ASTEC V2, IRSN all rights reserved, [2019]), developed by IRSN, and MELCOR 2.2 code, developed by SANDIA for USNRC, have been compared and analyzed. The SBO scenario considered takes into account the intervention of the accumulators as only accident mitigation strategy. Several figures of merits related to the thermal-hydraulic (e.g. primary pressure, cladding temperature, etc.) and to the core degradation (e.g. hydrogen production, etc.) have been considered to describe the accident evolution until the vessel failure, for the two codes comparison.

Backtracking and prospect on LNG supply chain safety

The safety issues of Liquefied Natural Gas (LNG) in production, storage, loading/unloading, transportation/shipping, and re-gasification have became a major concern, since an accident in the LNG industry would be very costly. Understanding the threat of LNG not only contributes to the process safety and reliability in the research and development (R&D) system, but improves the efficiency of loss prevention, fire protection and emergency responses. As of April 2019, in order to obtain the present status and trend of LNG safety research, basing 1122 documents of the Web of Science database about safety research of LNG as a data source, CiteSpace and VOS viewer were used for network knowledge map analysis. A comprehensive knowledge map of LNG safety field was obtained from several research aspects including scientific research power, research hot spots and trends, research knowledge base and frontier. According to the study results, the development of LNG safety research was divided into four stages from 1970s to 2019, China and South Korea made a lot of contributions, and the United States is the most influential. Among them, the research from 2005 to 2019 was the most representative. Current research results indicate that a combination of Formal Safety Assessment (FSA) methodology and Dynamic Procedure for Atypical Scenarios Identification (DyPASI) will fully identify risks; The PHAST and TerEx programs quickly define safety zones. Computational Fluid Dynamics (CFD) software package can provide accurate quantitative data for the study of LNG safety. Research on quantitative risk assessment (QRA) and LNG evaporated gas (BOG) has been a hot topic and trend in this field. The application of expansion foam in LNG accident mitigation covers most of the research content in this field, and the optimization of LNG liquefaction process has a great influence on this industry. As the international demand for LNG energy output increases, floating liquefied natural gas (FLNG) will have considerable development, and increasingly researchers attach vital importance to the safety of LNG offshore production integrated unit.

Identification of Risk-Significant Components in Nuclear Power Plants to Reduce Cs-137 Radioactive Risk

The amount and frequency of Cs-137 release in a nuclear accident are regulated to enhance environmental radiation protection. While significant improvements have been accomplished, including the preparation of severe accident mitigation equipment and systems to satisfy Cs-137 regulations, no attempt has yet been made along these lines to manage the structures, systems and components (SSCs) in nuclear power plants. Therefore, this paper proposes a new approach to identify the significant SSCs by quantifying related risk importance measures using release frequency and Cs-137 radioactivity. The new approach includes a Level 1 and Level 2 probabilistic safety assessment (PSA) linked model to obtain the frequency and Cs-137 radioactivity in all accident scenarios. The developed approach was applied to an OPR-1000 full-power internal events PSA model, and as a result, the top 20 SSCs were identified based on Fussell–Vesely importance, risk reduction worth, and risk achievement worth measures. The new approach can support risk-informed management for not only Cs-137 regulations but also public risk by using early and latent cancer fatalities.

The holistic risk analysis of spent fuel pool considering the fuel route risk

The Fukushima nuclear disaster has raised the importance on the reliability and risk research of the spent fuel pool (SFP), including the risk of internal events, fire, external hazards and so on. From a safety point of view, the low decay heat of the spent fuel assemblies and large water inventory in the SFP has made the accident progress goes very slow, but a large number of fuel assemblies are stored inside the spent fuel pool and without containment above the SFP building, it still has an unignored risk to the safety of the nuclear power plant. In this paper, a standardized approach for performing a holistic and comprehensive evaluation approach of the SFP risk based on the probabilistic safety analysis (PSA) method has been developed, including the Level 1 SFP PSA and Level 2 SFP PSA and external hazard PSA. The research scope of SFP PSA covers internal events, internal flooding, internal fires, external hazards and new risk source-fuel route risk is also included. The research will provide the risk insight of Spent Fuel Pool operation, and can help to make recommendation for the prevention and mitigation of SFP accidents which will be applicable for the SFP configuration risk management.

Identification and evaluation of level crossing for train accident mitigation using geographic information system application

The condition of most of the railways in Indonesia, especially in southern West Java, intersects the highway. Increased train traveling and public transportation on these level crossings potentially risks accidents to occur between trains and public transportation. To prevent this accident and secure railroad travel, it is necessary to identify and evaluate the existence of the level crossings. This study aims to gather information, so that it becomes a database on the southern cross of West Java. The identity of the plot is in the form of: number of gatekeepers and KM, gatekeeper coordinates, intersection status, road status, alignment condition, road completeness (signs and markers), and gate type and road emplacement conditions. The method used in this study was a survey method. Data were analyzed using spatial analysis with the application of Geographic Information Systems. Based on the result, the number of level crossings in southern West Java consists of the Bandung – Tasikmalaya and Tasikmalaya – Maos tracks. The number of survey point locations is 563 points. With the availability of databases and track information on the southern cross of West Java, the data can be used for the purposes of railroad accident transportation mitigation in the form of a road map with a detailed and accurate identity.

Investigation of the pressure vessel lower head potential failure under IVR-ERVC condition during a severe accident scenario in APR1400 reactors

In the event of a core meltdown in a high-power reactor, the integrity of the reactor pressure vessel is presumably protected by severe accident mitigation systems such as in-vessel retention external reactor vessel cooling (IVR-ERVC). However, in the late phase of the accident, two possible locations on the RPV are prone to failure: the location of the focusing effect and location of in-core instrument penetration. These two potential points of damage in the RPV are investigated in this study. A numerical model for the prediction of the natural convection, melting, and solidification processes for IVR-ERVC is presented. The model is based on the enthalpy-porosity approach with an extension for continuous liquid fraction function. The model is implemented in open-source field operation and manipulation (OpenFOAM) computational fluid dynamic code to produce a new solver which is based on the combination of conjugate heat transfer solver and buoyant-driven natural convection solver and the new solver is validated against the melting Gallium experimental test, in-core instrumentation failure experimental test, and BALI experimental test. This numerical model is applied for the investigation of the RPV rupture at the location of the focusing effect and in-core instrumentation penetrations. Severe ablations of the cladding and the weld materials are observed at a heat load of about ~1800 K which is expected to lead to the ejection of the penetration tubes if the force holding the penetration tube in place is lower than the force exerted by the system pressure. Subsequently, a two-layer IVR configuration is assessed and the integrity of the RPV is found not to be compromised under external reactor vessel cooling. However, in the case of a boiling crisis, the temperature of the ex-vessel wall is expected to rise quickly and this is simulated by increasing the ex-vessel wall temperature. The RPV is found to fail near the beltline due to a phenomenon known as focusing effect when the ex-vessel wall temperature rises above 1200 K.

Safety cases for design-basis accidents in LWRs featuring passive systems

This paper presents results from a series of integral tests performed at Framatome’s INKA test facility in Karlstein (Germany) which simulates a KERENA boiling water reactor (BWR). The scope of the test series was on the behaviour of and interaction between the different passive systems and components under the conditions of extended loss of alternating power (ELAP). These SBO-like conditions were aggravated in three out of four tests by parallel LOCA (Loss of Coolant Accident). The scenarios of all four tests fully correspond to Design Basic Conditions (DBC). They were: main steam line break, feed water line break, reactor pressure vessel (RPV) bottom leak and station blackout (SBO, non-LOCA).In the tests, the passive systems integrated in KERENA and INKA, respectively, have fulfilled their design functions fully satisfactorily and as follows:The Passive Pressure Pulse Transmitter (PPPT) triggered the RPV depressurization without delay. The Emergency Condenser (EC) system removed decay heat along with stored energy from the RPV to the containment. The Containment Cooling Condenser (CCC) system forwarded said power to a heat sink outside of the containment. The passive containment pressure suppression system kept the containment pressure within the design range, partially displacing surplus thermal energy from the drywell to the wetwell, in particular in the early phases after occurrence of LOCA. The passive core flooding system replenished the coolant inventory of the RPV thereby ensuring water levels in the RPV which are fully sufficient for core cooling.Moreover, the systems have cooperated as anticipated by the designers, quietly and without perturbing each other.Hence the test results, which are reported and discussed more in detail within this paper, soundly confirm the underlying design and its passive features.Said tests were carried out as a part of the joint research project EASY (Evidence of Design Basis Accidents Mitigation solely with passive safety Systems), the overarching objective of which was the development and validation of the code system AC2 of GRS (Gesellschaft für Anlagen- und Reaktorsicherheit gGmbH).

Precipitation profile and dryout concentration of sea-water pool-boiling in 5 × 5 full-height BWR bundle

Sea water shall be injected into water-cooled nuclear reactors during severe accidents, which are located along coastal side to flood the nuclear fuel, which is heated by residual heat. Precipitation growth to narrower flow path area is a key to gain the confidence of accident mitigation procedure to cool down the reactor core during accidental conditions. A pool boiling experiment was conducted with a simulated 5 × 5 full-height BWR fuel-rod bundle with condensed (two and half times higher concentration) sea water. The temperature on the center rod surface in the top spacer rose rapidly, since the flow area inside the top spacer was filled with the precipitated salt. Dryout below the top spacer escalated temperatures of the heater surface. On the other hand, the heater above the top spacer was cooled stably by pool boiling. An example calculation estimates that the dryout due to salt precipitation may occur 19 h after sea water injection for an ABWR, which had operated at 3.926 GWt for 13 months on the basis of critical dryout concentration of 50 wt%.

Prediction of the internal states of a nuclear power plant containment in LOCAs using rule-dropout deep fuzzy neural networks

A serious threat to the integrity of the reactor core, reactor coolant system, or containment is incurred if proper and essential actions to mitigate accidents cannot be taken owing to insufficient information about the internal states of the nuclear power plant (NPP). Therefore, this study was carried out to develop a model capable of mitigating the risk of severe accidents by accurately predicting the internal states of an NPP containment. A deep fuzzy neural network (DFNN) is a method in which syllogistic fuzzy reasoning is relatively efficient and inference capability is enhanced. In this study, the internal states of an NPP containment, hydrogen concentration and pressure, are predicted using a rule-dropout DFNN, as little NPP information is available under the circumstances of severe accidents. In addition, the performance of the proposed rule-dropout DFNN model is compared with that of other fuzzy neural network variations to verify the enhancement in the accuracy of the DFNN. The developed rule-dropout DFNN model is expected to be capable of providing accident monitoring information in advance for accident mitigation, as its prediction error for the hydrogen concentration and pressure in the containment is low.

Thermophysical properties of nanofluids MWCNT (multi-walled carbon nanotubes) in water for emergency coolant from nuclear reactors

To evaluate the synthesis and characterization of MWCNT (Multi-walled Carbon Nanotubes) with different degrees of functionalization in distilled water. The thermophysical properties (thermal conductivity and viscosity) of these nanofluids were measured at a temperature range (20-60°C) and concentrations (0.005-0.05%) by volume. Increases in thermal conductivity and viscosity were found 9.3% and 4.7%, respectively, at a volumetric concentration of 0.01% at a temperature of 30°C. The study of new fluids that improve the rate of removal of heat is fundamental to obtain greater efficiency of energy systems. Among the several factors that compromise the efficiency of the energy systems, we can highlight the thermophysical limitations of the conventional fluids, inhibiting in a very significant way some industrial applications. In this work we intend to improve the heat transfer characteristics of fluids commonly used by the addition of nanoparticles, made up of carbon nanotubes, in water which is the most used fluid for the cooling of nuclear reactors in operation today. It is intended to improve the heat transfer characteristics of fluids commonly used by the addition of nanoparticles, made of carbon nanotubes, through the addition of nanoparticles, made up of carbon nanotubes, in water which is the most used fluid for refrigeration of nuclear reactors currently in operation. In order to assess its benefits for the applicability and nuclear systems, ie primary coolant, safety systems, major accident mitigation strategies.

Development of severe accident mitigation technology and analysis for SMART

Severe accident mitigation technology was developed and evaluated for a small integral reactor of SMART. The containment pressure and hydrogen behavior were analyzed using MELCOR computer code (Sandia National Laboratory, 2018) in the SBO (Station Black Out) sequence, which was selected from PSA (Probabilistic Safety Assessment) results for the SMART. The severe accident mitigation technology to improve the SMART safety include reactor vessel depressurization using the ADS (Automatic Depressurization System) to prevent DCH (Direct Containment Heating) in case of a reactor vessel failure, a reactor cavity flooding using the CFS (Cavity Flooding System) with the IRWST (In-containment Refueling Water Storage Tank) for the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) to prevent the reactor vessel failure, and a hydrogen control system of PARs (Passive Autocatalytic Recombiners) to remove hazards from hydrogen combustion considering the amount of hydrogen to be generated by 100% fuel cladding oxidation. The MELCOR results showed that the containment pressures during the SBO sequence was below the design pressure, which meant that the containment integrity is maintained during a severe accident in the SMART. The hydrogen mole fraction of the containment was much lower than the hydrogen safety criteria of 10 vol%, which meant that the possibility of the hydrogen burn in the containment was negligible. For this reason, it could be concluded that the hydrogen was controlled by the hydrogen moving path and the hydrogen control system of PARs in the SBO sequence of the SMART.

Maritime safety: Prevention versus mitigation?

Consideration of risk treatment measures is an essential step of a risk assessment. The type of treatment, preventive or mitigative, is decided on an ad hoc basis, and there seems to be a preferential bias towards mitigation measures in the maritime domain. The paper poses the question of whether this bias is warranted. To this end, we analysed seven studies where the formal safety assessment (FSA) was applied to a variety of cargo and passenger ships. The analysis presents a qualitative review of the recommended RCOs and studies any correlation between the presented RCOs and any bias towards mitigation or prevention. The analysis concluded that any assertion that improving accident mitigation is more practical, i.e. cost-effective, than incident prevention is unwarranted. However, we also showed that the credibility of the FSA results is weak, making this conclusion rather preliminary. The results also show that the reviewed risk analyses expose a potential bias towards mitigation, and the paper discusses possible reasons for that.

Pressure wave behavior and its effects on structure under In-box LOCA in a helium-cooled lead lithium blanket of hydrogen fusion reactors

The helium-cooled lead lithium (PbLi) blanket is considered as one of the candidate blanket concepts selected for the hydrogen fusion DEMO reactors and beyond, which has the advantages of simple structure, strong heat removal capacity and high tritium breeding ratio. However, due to the harsh environment such as high-energy neutron irradiation, high thermal load and great pressure gradient, there is a high possibility that one or some of the thousands of coolant channels will break in the breeding zone, which is so-called In-box Loss of Coolant Accident (In-box LOCA). When the accident occurs, the high pressure helium will rapidly inject into the lead lithium flow channel, generating a complex two-phase flow and great pressure shock effect, which may cause the peak pressure to exceed the design limit and threaten the integrity of the blanket structure. Therefore, it is of great significance to perform the transient analysis of in-box LOCA to improve the safety of the blanket and avoid the leakage of radioactive materials. In this paper, a two-way coupling model for fluid-solid interaction was established based on the ANSYS Workbench, and the model were validated through the experimental data obtained by injecting the high pressure helium gas into liquid lithium lead. Then the validated model was applied to the transient pressure wave propagation analysis and structural stress analysis of the Dual-Functional Lithium Lead (DFLL) blanket in order to explore the integrity of blanket structure under In-box LOCA. In addition, the effects of break location on pressure and structural stress was also investigated through six cases. The study found that the transient pressure in the DFLL blanket gone through three stages in any case: step rise, oscillate, and flatten out. Pressure peaks occurred during oscillations and their values were strongly dependent on the break location. The closer to the inlet/outlet, the higher the peak pressure was. The maximum pressure reached more than twice of the inlet pressure (up to ~16 MPa). As a result, the structural stress in some local areas has exceeded the allowable limits, and the corresponding suggestions for improvement have also been put forward. This study can provide guidance for safety design, operation and accident mitigation measures of helium-cooled lead lithium blankets.

A framework analyzing system status and human activities: Illustrated using 2011 Fukushima nuclear power plant accident scenarios

Systematically modeling human activities and their interactions with complex systems is important in decision making processes. In order to determine appropriate human activities bringing systems to some desired status, integrated analyses combining both components are necessary. However, this can be difficult due to the complexity of the problems and the lack of adequate data. In this work, a framework capable of analyzing system status and human activities as an integrated set is proposed based upon use of a time-dependent Bayesian Network. The 2011 Fukushima Daiichi nuclear power plant accident scenarios are utilized in this work in order to illustrate the framework. Scenarios presented in this work are developed based upon the authors’ interviews with Tokyo Electric Power Company (TEPCO) staff who were directly engaged in Fukushima accident mitigation. In this work, we focus upon problems that TEPCO staff found challenging during the Fukushima accident. Two types of tasks, mitigation activities during ongoing accidents and preparation tasks before potential accidents, are considered. The capability of the framework for providing system status, based upon rapidly changing human activities, can enable decision-makers to select appropriate mitigation strategies for use during emergencies. The capability of the framework in comparing multiple preparation efforts can enable decision-makers to identify appropriate strategies in order to prepare for similar future situations. Beyond the examples illustrated here, the proposed framework can be applied to broader fields and support corresponding decisions as long as corresponding human activity and system models are available.