Severe Nuclear Power Plant Research Articles

Methodology for analyzing dose consequence using atmospheric dispersion code A2CDOSE

Abstract The experience of Fukushima brings out the importance of the emergency response plan during or after a severe nuclear power plant (NPP) accident. To deal with the accident emergency response situation, the National Atomic Research Institute (NARI) in Taiwan developed an emergent radiological dose evaluation system with the Central Weather Administration (CWA) called “A2CDOSE”. A2CDOSE can evaluate the eight-day projected dose after a nuclear accident and provide the predicted dose consequences for decision-making. Nevertheless, a source term generation method was still needed during the accident. This study chose MELCOR as the comparison target for the source term simulation issue with RASCAL. A MELCOR calculation of Maanshan NPP was built and a 5 % containment leakage was set for both MELCOR and RASCAL source term calculations. It seems the calculations of noble gases were consistent between MELCOR and RASCAL. However, the Cs-137 and I-131 calculated by MELCOR were twice compared to the RASCAL results, which may have a chance to cause different calculations of public protective actions during an accident. After the source term analysis, several atmospheric dispersion cases were simulated by A2CDOSE and RASCAL for further comparison.

Improvement and validation of containment spray removal aerosol models based on single droplet collection particle mechanism

Nuclear safety is the lifeline for the development and application of nuclear energy. In severe nuclear power plant accidents, aerosols are the major carriers of fission products, which may leak into the environment and cause potential radioactive contamination. Containment spray is an important mitigation measure for severe accidents in nuclear power plants. It can effectively reduce containment atmospheric pressure and radioactive aerosol concentration. This paper improves the containment spray removal models so as to address the low accuracy of the original models of the ISAA code. Based on single droplet collection particle mechanism, new inertial impaction and interception models were used to accurately calculate the collection efficiency of large particles. Additionally, a new correlation was used to explain the Brownian diffusion collection mechanism of small particles. Moreover, thermophoresis and diffusiophoresis models were introduced to consider the contribution of steam condensation in the containment to the collection efficiency. A rear capture model was introduced to incorporate the influence of recirculation within the wake of large droplets on the collection efficiency. Furthermore, THAI, TOSQAN and CSE experiments were selected to validate and evaluate the improved ISAA code. According to the calculation results, the improved models can simulate the attenuation trend of suspended aerosols with higher accuracies and significantly improves the calculation accuracy of the spray removal constants. This work contributes to understand of the scavenging mechanism of aerosol particle by containment spray droplets, and provides an analysis method with higher simulation accuracy. At the same time, it points out shortcomings in the simulation of aerosol behavior in the current ISAA code and discusses future improvements to the code.

Influence of Jet on Aerosol Retention by Pool Scrubbing Under Multihole Injector

In severe nuclear power plant accidents, when the containment is in a serious challenging state, the gas mixture in the containment can be injected into the spent fuel pool through the multihole injector by the containment depressurization measure, to reduce the risk of containment overpressure failure and the release of radioactivity to the environment. The pool scrubbing efficiency of aerosol under the jet regime is studied on the small-scale aerosol pool scrubbing facility, focusing on the influence of mass flux, steam fraction, submergence, particle diameter, and pool initial temperature on the aerosol decontamination factor (DF). The results show that under the jet regime, the DF value is significantly greater than that in the bubble regime and the effect of jet flow on the mechanism of steam condensation and aerosol removal of the rising zone is weak under the conditions explored. DF increases with the increase of mass flux owing to the droplet interception and inertial collision aerosol removal mechanisms. Because the high pool temperature weakens the aerosol removal by steam condensation, DF decreases with the increase of initial pool temperature under the conditions explored. Based on the experimental data and the analysis of the removal mechanism under the jet regime, an empirical model of aerosol DF considering mass flux, steam fraction, pool temperature, submergence, and particle size is established and verified by the international experiments. The proposed model can be used to calculate DF in the process of containment overpressure discharge.

What’s better for our health? Conducting protective actions during a nuclear emergency or accepting a certain radiation dose?

The threat caused by ionising radiation has resulted in the establishment of strict radiation protection guidelines. This is especially true for severe nuclear power plant (NPP) accident scenarios, which may involve the release of significant amounts of ionising radiation. However, we believe that the fine balance between the benefit of a certain protective action (e.g. evacuation) and its risks is not always accounted for properly. Deaths and mental health problems have been associated with protective actions (e.g. evacuation) implemented in the response to the Fukushima Daiichi (NPP) accident in 2011. The protective actions were implemented consistent with international recommendations, to reduce radiation-induced health effects, even though the off-site effective doses were too low to indicate that there would be any discernible radiation-induced health effects. In this paper, we will provide a first step for the development of tools to evaluate the risk of protective actions versus the radiation-induced health risk. Over 50 papers were selected as useful from more than 600 reviewed papers to characterise the health impact of protective actions taken during different emergencies (including, technical and natural emergencies). An analysis was performed comparing the radiation-induced health effects averted by protective actions with the health effects associated with the protective actions. We concentrated our analysis on deaths and mental health problems associated with protective actions compared with the inferred radiation-induced deaths averted by the protective actions. Our analysis is stated in terms of absolute risk (cases per 1000) of health effects to allow for a direct comparison. It indicates that taking protective actions consistent with dose criteria typically used in many countries could result in more excess deaths than the inferred radiation-induced deaths prevented, as well as resulting in mental health problems. We identified that residents of facilities for long stays and the elderly are particularly vulnerable and a significant number of the deaths among the general public are associated with a lack of emergency preparedness provisions.

Tellurium transport in the RCS under conditions relevant for severe nuclear accidents

In the case of a severe nuclear power plant accident, tellurium is one of the more problematic and volatile fission products. If released it could become a health issue as it decays to iodine which accumulates in the thyroid gland. Research exists, that indicates tellurium likely interacts with caesium under severe accident conditions, thus it is important to further explore related phenomenon. In this work, tellurium was exposed to high temperature under oxidizing and inert conditions simulating severe accident conditions with and without airborne caesium iodide to determine the effect on the tellurium source term.The effect of caesium iodide was noticeable on tellurium transport behaviour in the gas phase under oxidizing and inert conditions. Under humid oxidizing conditions with caesium iodide, no significant impact on the total aerosol mass transport was noticed. However, less tellurium was transported through the model primary circuit and a potentially new compound was observed on the filter located after this model. Comparing inert dry to humid with caesium iodide showed an increase in the total aerosol mass transport whereas there was a decrease noticed of the tellurium reaching the filter after the model primary circuit. In the latter case, new unidentified compound(s) correlated to caesium, iodine and tellurium were observed on the filter located after the model.In this work, evidence was found that tellurium behaviour will be affected by caesium iodide under the investigated conditions. Moreover, it seems that under inert conditions the formed compounds may be stable at close to ambient temperatures. Unlike under oxidizing conditions, where dissociation likely occurred.

Improved prediction model for H2/CO combustion risk using a calculated non-adiabatic flame temperature model

During severe nuclear power plant (NPP) accidents, a H2/CO mixture can be generated in the reactor pressure vessel by core degradation and in the containment as well by molten corium-concrete interaction. In spite of its importance, a state-of-the-art methodology predicting H2/CO combustion risk relies predominantly on empirical correlations. It is therefore necessary to develop a proper methodology for flammability evaluation of H2/CO mixtures at ex-vessel phases characterized by three factors: CO concentration, high temperature, and diluents. The developed methodology adopted Le Chatelier’s law and a calculated non-adiabatic flame temperature model. The methodology allows the consideration of the individual effect of the heat transfer characteristics of hydrogen and carbon monoxide on low flammability limit prediction. The accuracy of the developed model was verified using experimental data relevant to ex-vessel phase conditions. With the developed model, the prediction accuracy was improved substantially such that the maximum relative prediction error was approximately 25% while the existing methodology showed a 76% error. The developed methodology is expected to be applicable for flammability evaluation in chemical as well as NPP industries.

Tellurium determination by three modes of instrumental neutron activation analysis in aerosol filters and trap solutions for the simulation of a severe nuclear accident

Tellurium belongs to the elements not frequently determined by neutron activation analysis (NAA) or other analytical methods. We present results of a new methodological study using three independent modes of instrumental NAA (INAA) using the 123mTe, 131Te and 131I radionuclides. We compare the results obtained in terms of accuracy, precision and limits of detection (LOD). We utilized the INAA procedures tested for the tellurium determination in aerosol filters and trap solutions in a model experiment aimed at reducing the knowledge gap concerning the behaviour of 132Te, a radiologically significant fission product, which constitutes a considerable health risk towards the public in case of its release in a severe nuclear power plant accident. We found that the nuclear reaction 130Te(n,γ)131Te and gamma-ray spectrometric measurement of 131I, a descendant of 131Te, is the most sensitive way of Te determination by INAA providing as low LOD values as 0.15 µg of Te in the Teflon aerosol filters and 0.22 µg mL-1 in the 0.1 M NaOH trap solutions. The three independent INAA modes allowed employing the self-verification principle of INAA for increasing the trustworthiness of our results. Finally, we also point to the indispensable role of the non-destructive feature of INAA for assay of samples, such as Teflon aerosol filters, that are difficult to be analysed by other analytical methods requiring complete sample destruction without analyte losses.

SOARCA uncertainty analysis of a short-term station blackout accident at the Sequoyah nuclear power plant

The U.S. Nuclear Regulatory Commission initiated the state-of-the-art reactor consequence analyses (SOARCA) project to develop realistic estimates of the offsite radiological health consequences for potential severe reactor accidents. The SOARCA analysis of an ice condenser containment plant was performed because its relatively low design pressure and its reliance on igniters make it potentially susceptible to early containment failure from hydrogen combustion during a severe accident. The focus was on station blackout accident scenarios where all alternating current power is lost. Accident progression calculations used the MELCOR computer code and offsite consequence analyses were performed with MACCS. The analysis included more than 500 MELCOR and MACCS simulations to account for uncertainty in important accident progression and offsite consequence input parameters.Consequences from severe nuclear power plant accidents modeled in SOARCA are smaller than previously calculated. The delayed releases calculated provide more time for emergency response actions. The results show that early containment failure is very unlikely, even without successful use of igniters. The modeled behavior of safety valves is very important to this conclusion, but there is sparse data and a lack of established expert consensus on the failure rates under severe accident conditions. Even for scenarios resulting in early containment failure, the calculated individual latent fatal cancer risks are very small. Early and latent-cancer fatality risks are one focus of this paper. Regression results showing the most influential parameters are also discussed.

Recovery countermeasure features that may preclude MCDA based decisions

Multi Criteria Decision Analysis (MCDA) with stakeholder participation has been suggested as a way to reach mutually acceptable and robust decisions regarding methods and strategies for recovery of inhabited land areas and food production systems severely contaminated by an airborne radioactive plume from a severe nuclear power plant accident. The paper describes a new approach to help ensuring that important issues are not overlooked when decisions are based on MCDA.

Two-phase flow pressure loss through packed particle bed for wide void fraction range

The understanding of the two-phase flow phenomena in porous media is a critical issue for debris bed coolability assessment in severe nuclear power plant accidents. Hence, two-phase-flow friction force modeling has been extensively attempted, based on pressure-loss measurement data. However, most previous two-phase flow pressure loss experiments were conducted at a restricted void-fraction range up to approximately 0.6, which is considerably lower than the dryout condition. Therefore, in this research, two-phase flow tests in packed particle beds are conducted to gather two-phase pressure loss data in the high-void-fraction region. The test beds are constructed by packing mm-scale particles, 3–7 mm in diameter. The experimental data are analyzed and compared with the predicted values of the previous two-phase friction models in term of not only the pressure gradient but also the interfacial friction force between the liquid and gas phases. As there was no valid data for the high-void-fraction region during the previous model development processes, an appropriate model for the interfacial friction force at void fractions greater than approximately 0.6 is unavailable. However, the interfacial friction force plays a significant role in the coolability assessment of the ex-vessel debris bed because it determines the water ingression rate under a cocurrent flow condition. The obtained experimental results in this research indicate the necessity to further improve the model in order to decrease the uncertainty in severe accident risk quantification.

Optimizing UAV-based radiation sensor systems for aerial surveys

In this study, we qualitatively and quantitatively analyzed unmanned aerial vehicle (UAV)-based radiation sensor systems suitable for specific survey missions. We examined a variety of UAVs, radiation sensors, and radiological survey missions; after reviewing previous studies that developed and conducted field tests, we categorized them by mission and suggested suitable types of UAVs and radiation sensors for each mission. To quantitatively analyze various system designs previously suggested, we proposed a new figure of merit (FOM) formula that can explain the mutual effects of parameters of both radiation sensors and UAVs on system performance. After targeting a radiological survey mission, we selected UAV and radiation sensor types qualitatively. Then, the quantitative FOM formula can be employed to efficiently assess whether the system achieves the required minimum detectable activity (MDA) without field test, while keeping the error of the MDA values in the order of 100. After defining the constraints including the proposed FOM formula, we replicated a severe nuclear power plant accident scenario. We found that one fixed-wing UAV and multiple rotary-wing UAVs are the minimum number of UAVs, and thus the optimal system for timely achievement of a satisfactory MDA and estimation of the three-dimensional radiation distribution contributed by both ground radiation and an atmospheric radioactive plume.

GIS-Based Integration of Social Vulnerability and Level 3 Probabilistic Risk Assessment to Advance Emergency Preparedness, Planning, and Response for Severe Nuclear Power Plant Accidents.

In the nuclear power industry, Level 3 probabilistic risk assessment (PRA) is used to estimate damage to public health and the environment if a severe accident leads to large radiological release. Current Level 3 PRA does not have an explicit inclusion of social factors and, therefore, it is not possible to perform importance ranking of social factors for risk-informing emergency preparedness, planning, and response (EPPR). This article offers a methodology for adapting the concept of social vulnerability, commonly used in natural hazard research, in the context of a severe nuclear power plant accident. The methodology has four steps: (1) calculating a hazard-independent social vulnerability index for the local population; (2) developing a location-specific representation of the maximum radiological hazard estimated from current Level 3 PRA, in a geographic information system (GIS) environment; (3) developing a GIS-based socio-technical risk map by combining the social vulnerability index and the location-specific radiological hazard; and (4) conducting a risk importance measure analysis to rank the criticality of social factors based on their contribution to the socio-technical risk. The methodology is applied using results from the 2012 Surry Power Station state-of-the-art reactor consequence analysis. A radiological hazard model is generated from MELCOR accident consequence code system, translated into a GIS environment, and combined with the Center for Disease Control social vulnerability index (SVI). This research creates an opportunity to explicitly consider and rank the criticality of location-specific SVI themes based on their influence on risk, providing input for EPPR.

Effect of honeycomb porous plate on critical heat flux in saturated pool boiling of artificial seawater

During a severe nuclear power plant accident, the integrity of the reactor pressure vessel must be assured. In response to a possible fuel meltdown, operators of the current generation of nuclear power plants are likely to inject water into the reactor pressure vessel cavity to cool down the reactor vessel wall, preserving its integrity and avoiding leakage of radioactive material. This study considers the use of seawater to flood a reactor pressure vessel cavity combined with the attachment of a honeycomb porous plate (HPP) on the vessel outer wall as a way to improve the safety margins for in-vessel retention of fuel. In long-duration experiments, saturated pool boiling of artificial seawater was performed with an upward-facing plain copper heated surface 30 mm in diameter. The resulting value for critical heat flux (CHF) was 1.6 MW/m2 at atmospheric pressure, a value significantly higher than the CHF obtained when the working fluid was distilled water (1.0 MW/m2). It was verified that sea-salt deposits could greatly improve surface wettability and capillarity, enhancing the CHF. The combination of artificial seawater and an HPP attached to the heated surface improved the boiling heat transfer coefficient and increased the CHF up to 110% (2.1 MW/m2) as compared to distilled water on a bare surface. After the artificial seawater experiments, most of the wall micropores of the HPP were clogged due to sea-salt aggregation on the HPP top and bottom surfaces. Thus, the CHF enhancement observed in this case was attributed mainly to the separation of liquid and vapor phases provided by the HPP channel structure and improvement of surface wettability and capillarity by sea-salt deposition.

Mass transfer experiments on the natural convection heat transfer of the oxide pool in a three-layer configuration

We investigated the heat load to the reactor vessel by the natural convection of the core melt in a hypothetical severe nuclear power plant accident condition. It is modeled as the natural convection heat transfer with the volumetric heat source in a chopped hemisphere. To achieve high Ra'H ranging from 1011 to 1013, mass transfer experiments were performed for three-layer configuration of the oxide pool with three different aspect ratios based on the analogy between heat and mass transfer. The local heat transfers to top, side, and bottom were measured, compared with the existing studies, and discussed phenomenologically. Major findings were that the side and bottom heat flux distributions are unaffected by the top plate cooling condition and that the angle dependent heat flux increases with the angle measured from the bottom. Three layer configuration is more conservative as its upward heat ratio was larger than that of two layer configuration. With the decrease of the aspect ratio, the upward heat ratios increased due to the reduced side wall cooling but with very small aspect ratio, multi-cell flow patterns were formed. The heat transfer around the edge of the bottom plate was impaired due to the formation of stagnant flows.

Intelligent Modeling for Nuclear Power Plant Accident Management

This paper explores the viability of using counterfactual reasoning for impact analyses when understanding and responding to “beyond-design-basis” nuclear power plant accidents. Currently, when a severe nuclear power plant accident occurs, plant operators rely on Severe Accident Management Guidelines. However, the current guidelines are limited in scope and depth: for certain types of accidents, plant operators would have to work to mitigate the damage with limited experience and guidance for the particular situation. We aim to fill the need for comprehensive accident support by using a dynamic Bayesian network to aid in the diagnosis of a nuclear reactor's state and to analyze the impact of possible response measures. The dynamic Bayesian network, DBN, offers an expressive representation of the components and relationships that make up a complex causal system. For this reason, and for its tractable reasoning, the DBN supports a functional model for the intricate operations of nuclear power plants. In this domain, it is also pertinent that a Bayesian network can be composed of both probabilistic and knowledge-based components. Though probabilities can be calculated from simulated models, the structure of the network, as well as the value of some parameters, must be assigned by human experts. Since dynamic Bayesian network-based systems are capable of running better-than-real-time situation analyses, they can support both current event and alternate scenario impact analyses.

A novel method to improve dose assessment due to severe NPP accidents based on field measurements and particle swarm optimization

Severe nuclear power plant (NPP) accidents are those which involve significant core degradation and lead the plant to conditions more severe than a design basis accident. Under such conditions the accident progression might become unpredictable and the source term estimation, imprecise by orders of magnitude. The consequence is a dose assessment very far from the reality and a deficient decision making support. This work presents a novel approach to improve accuracy of dose estimation, based on field measurements and particle swarm optimization (PSO) algorithm. The main idea is to determine a correction matrix, which once applied to the originally estimated (incorrect) dose distribution map, generates a corrected one, which better fits to the field measurements. The proposed correction matrix is the result of a concatenation of geometric transformations and an amplification/attenuation factor, aimed to fit the shape of the original map and radiation intensities in order to match the field measurements. Finding the optimum transformations (correction matrix) is, however a complex nonlinear optimization problem, which has been successfully solved by using a PSO algorithm. Results demonstrate that PSO was able to find good correction transformations, which can be used to better project future dose distributions and, consequently, improve decision making support.

Авария на ас Фукусима-дайити как стресс-тест для национальной системы защиты населения при тяжелой аварии на атомной станции

Авария на ас Фукусима-дайити как стресс-тест для национальной системы защиты населения при тяжелой аварии на атомной станции

Changes to soundscapes at outdoor gathering spaces in Fukushima, Japan, caused by the severe nuclear power plant accident

Five years have passed since the accident at the Fukushima Daiichi Nuclear Power Plant, which caused changes to daily life in the city of Fukushima, Japan. These daily life alterations caused changes to the soundscapes of Fukushima, which are still evolving now. In this study, the soundscape change at Fukushima’s outdoor gathering spaces is discussed using field recordings by the author. Shortly after the accident, few human voices could be heard at outdoor gathering spaces, but natural sounds such as birdsongs could be heard as usual per the time of year. Park decontamination started during summer 2011. In some parks where decontamination was successfully completed, people’s voices and sounds of children playing returned during spring 2012. However, in other parks where decontamination was done but ineffective, the lack of human voices and artificial sounds continued until radiation levels decreased sufficiently. In this way, soundscape change at outdoor gathering spaces represents people’s attitude toward the radioactive contamination; soundscape recording documents not only sonic environments but also people’s lives.

Development of an aerosol decontamination factor evaluation method using an aerosol spectrometer

During a severe nuclear power plant accident, the release of fission products into containment and an increase in containment pressure are assumed to be possible. When the containment is damaged by excess pressure or temperature, radioactive materials are released. Pressure suppression pools, containment spray systems and a filtered containment venting system (FCVS) reduce containment pressure and reduce the radioactive release into the environment. These devices remove radioactive materials via various mechanisms. Pressure suppression pools remove radioactive materials by pool scrubbing. Spray systems remove radioactive materials by droplet−aerosol interaction. FCVS, which is installed in the exhaust system, comprises multi-scrubbers (venturi-scrubber, pool scrubbing, static mixer, metal−fiber filter and molecular sieve). For the particulate radioactive materials, its size affects the removal performance and a number of studies have been performed on the removal effect of radioactive materials.This study has developed a new means of evaluating aerosol removal efficiency. The aerosol number density of each effective diameter (light scattering equivalent diameter) is measured using an optical method, while the decontamination factor (DF) of each effective diameter is evaluated by the inlet outlet number density ratio. While the applicable scope is limited to several conditions (geometry of test section: inner diameter 500mm×height 8.0m, nozzle shape and air-water ambient pressure conditions), this study has developed a numerical model which defines aerosol DF as a function of aerosol diameter (d) and submergences (x).

Influence of Concrete Properties on Molten Core-Concrete Interaction: A Simulation Study

In a severe nuclear power plant accident, the molten core can be released into the reactor pit and interact with sacrificial concrete. In this paper, a simulation study is presented that aims to address the influence of sacrificial concrete properties on molten core-concrete interaction (MCCI). In particular, based on the MELCOR Code, the ferrosiliceous concrete used in European Pressurized Water Reactor (EPR) is taken into account with respect to the different ablation enthalpy and Fe2O3 and H2O contents. Results indicate that the concrete ablation rate as well as the hydrogen generation rate depends much on the concrete ablation enthalpy and Fe2O3 and H2O contents. In practice, the ablation enthalpy of sacrificial concrete is the higher the better, while the Fe2O3 and H2O content of sacrificial concrete is the lower the better.