Severe Nuclear Accident Research Articles

Microstructural Analysis of CaO-Fe2O3 Heat Treated at Different Temperature

In present paper a study has been carried out to analyze the thermophysical properties and microstructural behavior of CaO.Fe2O3.A binary eutectic sample material is taken in the ratio of 22:78 by Wt. % in mixed powder form prepared manually. The sample powder is heat-treated at 950°C, 1050°C, and 1150°C for three hour holding time in programmable furnace. Further, the characterization is done with the help of different characterization technique like DSC, XRD, TGA to analyze its thermophysical properties and microstructural changes. This studies became necessary to analyze and interpret the data collected from prototype material found from nuclear severe accident occurred in the past. As heat transfer phenomena plays the crucial role in nuclear severe accident due to high temperature gradient developed inside the reactor vessel. Hence a non-prototype material is needed to study about heat transfer phenomena and in context of this study the material CaO.Fe2O3 is being considered. This study bears the first stage to validate the given material is simulant material. In the current study, heat absorbed, phase analysis, unit cell volume, melting point, crystallite size, dislocation density, and weight loss have been reported and analyzed.

Leaching behavior of prototypical corium samples: A step to understand the interactions between the fuel debris and water at the Fukushima Daiichi reactors

On 11 March 2011, the Great East Japan Earthquake occurred and caused a severe nuclear accident at the Fukushima Daiichi Nuclear Power Station (1F). A policy for corium debris retrieval will be decided in the near future. To this end, debris management after removal from reactors must be optimized. From the viewpoint of cost and waiting for definitive solutions, interim or long-term underwater storage is one of the considered options after debris removal. Leaching data on the long term behavior of debris in water are therefore needed before debris removal. In addition, it is important to understand potential differences in leaching behaviors between debris and Spent Fuel (SF) to get comparative elements.For these reasons, simulated in-vessel and ex-vessel corium debris, fabricated in the Colima experimental facility set up in the PLINIUS platform at CEA Cadarache, were selected and leaching experiments were carried out under oxidizing conditions. In parallel, geochemical modeling was performed to better understand the experimental concentrations, pH evolutions and secondary phase’s formation.Finally, the Fractional Release Rates of the (U, Zr)O2 matrix for the two kinds of samples (in-vessel and ex-vessel) were found to be close to or one order of magnitude lower than that of SF under oxidizing conditions (from 10−6 to 10−7 per day) but the release processes are different. Indeed, the Fraction of Inventory in Aqueous Phase of Mo, Ba, Sr, Rh, La, Nd, Ce and Y tended to be higher than that of uranium and zirconium in this study whatever the element considered. The release of simulating PFs, including low volatile element like trivalent rare earths, does not appear to be fully controlled by that of the (U, Zr)O2 matrix from a macroscopic and average point of view. The FPs could be partially located outside the (U, Zr)O2 solid solution as a function of the corium history (quenching rate and thermal history, chemical interactions…). The location of inventories and their chemical forms are not very well known moreover uncertainties related to heterogeneity are important and may also impact the calculation of the global released fractions over time. For the ex-vessel sample, releases into the solution of the main concrete elements (Si, Ca, Mg) seem well correlated and dissociated from the behavior of the (U, Zr)O2 phase. Further studies are needed especially on real or prototypical corium approaching the conditions of the Fukushima accident. The study of simpler and model systems could also provide information on the reactivity of the (U, Zr)O2 matrix.

On Line Measurement of Organic Iodine During a Severe Nuclear Accident

A severe nuclear accident may induce a dramatic dissemination of radioactive species into the environment. In that respect, improving the nuclear safety remains an important challenge to improve the society acceptability towards this energy. A solution may consist on implementing robust and reliable measurement systems operating near the Containment Venting Systems (CVS). These devices should be able to provide real time monitoring of the emitted fission products (FPs) in the course of a hypothetical accidental sequence. In the present study, a peculiar attention was devoted to iodine species (namely CH3I) measurement by complementary techniques (photoacoustic spectroscopy and gas chromatography). The most important results will be described here.

Effect of soaking time on properties of non-prototype material (CaO-Fe2O3; 21:79 by wt. %)

The various simulant materials (non-porotype material) have been used to study the severe accident phenomena in a nuclear reactor by melt coolability technique. The present material can be used as a simulant material to study the thermal phenomena and heat transfer of a nuclear reactor accident scenario. The study of thermophysical properties plays a crucial role in the prediction of actual phenomena that occurred in a nuclear severe accident. This work devoted to analysing the effect of soaking time on thermophysical properties of non-prototype material CaO-Fe2O3 (21C79F; C = CaO, F = Fe2O3; 21:79 by wt. %). In continuation, an effort has been made to validate the simulation properties of the material in the context of other existing materials that are used for predictive analysis during a nuclear severe accident. A binary mixture powder of non-eutectic material CaO-Fe2O3 (21:79 by wt. % confirmed by phase diagram) in the form of cylindrical pellets is heat-treated at 1100 °C for one, two, and three hours of soaking time and further ground to powder form and finally, the heat-treated powder were analysed through characterization techniques to evaluate the changes in thermal properties, crystalline structure, and morphology. The results were also compared with the initial powder sample accordingly. The properties of material revealed that the material is thermally stable and it can be used in molten form to investigate the severe accident phenomena. The properties of CaO-Fe2O3 (21C79F) is possibly near to the properties of corium (a liquid mixture of UO2 and steel) and also environment-friendly as well as for human beings.

Remote monitoring of Molten Core-Concrete Interaction experiment with Optical Fibre Sensors & perspectives to improve nuclear safety – DISCOMS project

The DISCOMS project (Distributed Sensing for Corium Monitoring and Safety) aimed at providing innovative solutions not requiring local electrical power supplies, for remote monitoring of a severe nuclear accident. The solutions are based on both long length SPNDs (Self Powered Neutron Detectors) and on distributed OFSs (Optical Fibre Sensors) capable to detect the onset of a severe accident, the corium pouring on the containment building concrete basemat, and its interaction with the concrete floor under the reactor vessel, until it spreads in the core catcher (EPR case). This paper mainly focuses on these last three detection targets achievable with distributed OFSs. It is based on the results of a Molten Core & Concrete Interaction (MCCI) experiment, namely VULCANO, held in June 2018 with a concrete crucible equipped with overall ~ 180 m long optical fibre sensing cables. This small scale experiment (50 kg of prototypical corium) has demonstrated the ability of distributed OFSs to remotely provide useful data during the MCCI run: i) temperature profiles images up to about 580°C (single wavelength Raman DTS reflectometer) until cooling down to room temperature, ii) high spatial-resolution frequency shifts profiles, due to combined (non-selective) strain and temperature influences (Rayleigh OFDR and Brillouin reflectometers), and iii) cables lengths ablated by the corium on sections weakened by the temperature (Raman DTS, Rayleigh OFDR, telecom and photon counting reflectometers).

A robust methodology for assessing thyroid absorbed doses based on individual monitoring data after a severe nuclear accident

The paper presents a “one-step” procedure for the direct dose assessment based on thyroid monitoring data obtained after a severe nuclear accident at a light water power reactor. The developed methodology is based on the “dose per content” concept that allows the direct conversion of the measured value to the dose estimate. The paper provides methods and data for assessments of the committed absorbed dose to the thyroid based on measurements of the activity of radioactive isotopes of iodine in the thyroid gland. Values of functions “absorbed dose to the thyroid integrated over 30 days per activity of I-131 measured in the thyroid” were calculated for adults and children. For in utero exposure, the absorbed dose to the thyroid of the offspring is provided per activity measured in the maternal thyroid. The contribution of short-lived radionuclides I-132, I-133, I-134, I-135, and Te-132, is considered by applying correction coefficients. Optionally, the methodology permits the use of measurements of the I-132 activity in the thyroid. Primary sources of uncertainty in the assessed thyroid dose are also discussed.

Limestone-siliceous and siliceous concretes thermal damaging at high temperature

Limestone-siliceous and siliceous concretes are used in reactor pits of French nuclear power plants. In case of severe nuclear reactor accident, failure of the reactor vessel would lead to interaction between molten corium (hot melt of nuclear fuel) and concrete. This paper focuses on the thermal degradation of both limestone-siliceous and siliceous concretes till 1000 °C. Thermo-Gravimetric Analysis (TGA) and Mercury Intrusion Porosimetry (MIP) are used to measure mass loss and porosity modification. As concretes are heterogeneous materials, sampling and representativeness have been addressed. TGA experiments show larger mass loss for limestone-siliceous concretes due to the decarbonation of calcium carbonate gravels when T > 800 °C. MIP experiments demonstrate a 100% increase of porosity for limestone-siliceous (resp. siliceous) concretes when T > 500 °C (resp. T > 800 °C). The consequences of these results are discussed in the frame of experimental tests on prototypical corium systems aimed at describing the key-phenomena involved in molten corium concrete interaction.

Nuclear safety in the unexpected second nuclear era

Nuclear energy development has entered an unexpected second nuclear era, which is mainly driven by developing countries. Despite major efforts to pursue a safe nuclear energy system in the first nuclear era, severe nuclear accidents occurred. A basic problem is that we do not have an adequate understanding of nuclear safety. From the viewpoints of risk and the close coupling of technical and social factors, this paper reexamines the nature of nuclear safety and reviews how previous experts understood it. We also highlight the new challenges that we are likely to confront in the unexpected second nuclear era and clarify some of the refinements that need to be made to the concept of nuclear safety from a sociotechnical perspective. These include the following: 1) Risk decisions should be made based on integrating social and technical elements (i.e., "social rationality"); 2) risk needs to be controlled based on the "Wuli-Shili-Renli" framework; 3) systems thinking should be substituted for reductionism in risk assessment, and social mechanisms need to be combined to address uncertainties; and 4) public-centered risk communication should be established. This contribution can provide a theoretical foundation for improving our understanding of the nature of nuclear safety and for transforming the concept of nuclear safety in the unexpected second nuclear era.

CFD-Assisted model development for estimation of hole-ablation diameter of a pressure vessel during severe accidents

To evaluate the released radioactive materials to an environment under a nuclear severe accident, it is necessary to analyze the ex-vessel phenomena. The releasing diameter is one of the most essential parameters to estimate the ex-vessel phenomena. The diameter is enlarged because the wall of a reactor vessel melts due to the high temperature of corium flow. In this study, a hole-ablation model was developed based on the one-dimensional heat balance equations for molten corium flow, crust layer, and molten layer of the reactor vessel, respectively. Moreover, the heat transfer coefficient for the situation of discharging flow was obtained using computational-fluid-dynamics. As a result, we proposed the enhancement factor, 1.57, to be multiplied by the heat transfer coefficient for an external turbulent flow. We found out that the strong effect of the critical thickness of the molten wall layer on the thicknesses of the crust layer and the molten wall layer. We suggested the relationship between the critical thickness of the molten wall layer and the ratio of the superficial velocity of the molten wall layer to that of the corium discharging flow. The proposed model was compared with the Pilch model: the mean averaged error (MAE) of the present model was 9.51%, whereas that of the Pilch model was 36.13%. In the case of the low superheat of the corium where the crust can be produced, the Pilch model without the crust formation model predicts the much larger final-hole diameter than the present model does.

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.

CFD analysis of the erosion of a light gas stratification by means of a hot air jet in the MiniPanda facility

Following the Fukushima-Daiichi accident, many countries decided to strengthen the safety systems of their nuclear power plants in order to increase the capabilities of managing severe nuclear accidents. To achieve this goal detailed analysis of postulated design and beyond-design-basis accidents is essential. The main tools available for the analysis of these complex scenarios are Advanced Lumped Parameters (LP) and Computational Fluid Dynamics (CFD) codes. Currently, the most limiting factor in the application of these tools is their validation. This activity is performed in support of the development and validation of a CFD model, built in the numerical environment of OpenFOAM, for hydrogen behavior in the containment of Light Water Reactors. For validation purpose, numerical simulations of the erosion of a light gas stratification by means of a hot air jet have been performed, and results of the simulations were analysed and compared against experimental results obtained by Ritterath (2012) in the MiniPanda facility. Two different scenarios have been considered, each characterized by a different value of the mass flow rate of the jet used to erode the stratification.

Modelling nuclear fuel behaviour with TAF-ID: Calculations on the VERDON-1 experiment, representative of a nuclear severe accident

The chemical behaviour of the main elements present in a PWR irradiated fuel sample (UO2 fuel, Zr from the Zircaloy-cladding, Pu, Np, and fission products such as Ba, Ce, Cs, Gd, I, La, Mo, Nd, Pd, Rh, Ru, Sr, Tc, Te, etc.) submitted to a nuclear severe accident type sequence (VERDON-1 experiment) has been investigated by thermodynamic calculations using the TAF-ID database. Particular emphasis has been placed on the chemical behaviour of the fission products Ba, Cs, Mo, and Zr, and the interactions between the Zircaloy-cladding and the UO2 matrix. Calculation results have been compared to experimental observations completed during the VERDON-1 experiment (instantaneous release of fission products and final chemical state of elements in the sample). Results presented in this manuscript do not account for non-equilibrium phenomena that have a major impact on fission product behaviour (e.g., chemical diffusion, mass transport induced by temperature gradients in the fuel pellet, etc.). None the less, these TAF-ID calculations have described accurately the melting of the sample following the formation of a mixed (U,Zr)O2-x phase because of the interaction between the fuel and the Zircaloy-cladding. Furthermore, calculations have assisted in the explanation of the instantaneous released fractions observed for Ba and Mo during the experiment (as a function of the experiment atmosphere and temperature).

The Risk of Extended Power Loss and the Probability of Emergency Restoration for Severe Events and Nuclear Accidents

This paper gives a complete account of recent work on the prediction of power restoration in severe accidents and events, and quantifies the benefits and impact of deploying back-up or emergency power systems in nuclear reactors. The overall outage data for all types of major events and disasters follow the same fundamental trends based on statistical learning theory, and are correlated by simple theoretically based exponential equations that include the degree of difficulty. This trend is shown to be completely independent of severe event type (e.g., hurricanes, ice storms, flooding, earthquakes, cyclones, and fires) but the rate systematically depends on the degree of difficulty. The paper emphasizes that the physics of learning, analytical methodology, technical statistical theory, and extensive event database already implicitly and fully include all human errors, actions, and decisions made during power restoration for the extremely adverse conditions prevalent in severe events and actual disasters. The theory shows flexible coping strategies/emergency power system FLEX/EPS reliability, deployment timescale, and severe event power restoration rate are intrinsically coupled together. The analytical results can be used to define the EPS design and reliability requirements and the potential risk benefit and deployment timescales in terms of the probability change in the risk of extended power loss.

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.

Hygroscopic growth of CsI and CsOH particles in context of nuclear reactor accident research

Studies of behavioral characteristics of fission product aerosols are extremely important in context of suspected environmental release during a severe nuclear reactor accident. Fission product aerosols generated in such a case are expected to travel from coolant systems to the containment and may get released to the environment in the most unlikely scenario of containment breach. Presence of steam also affects their dynamic behaviour and fate. Interaction of aerosol particles with water vapor in subsaturation domain modifies their physical and chemical characteristics. The objective of this study was to examine the hygroscopic nature of nuclear accident relevant cesium bound compounds viz. CsI, CsOH with the use of Hygroscopic Tandem Differential Mobility Analyzer (HTDMA). The growth factors for single component aerosols were obtained for different initial size of particles at different relative humidity levels. Growth factor curves as well as deliquescent parameters for CsI and CsOH particles have been obtained for the first time and have a crucial role towards environmental source term estimations in the event of nuclear accident. Experimental results were also compared with the theories available in literature.

Features of a control blade degradation observed in situ during severe accident conditions in boiling water reactors

ABSTRACT In the present paper, new results using in situ video are presented regarding boiling water reactor (BWR) control blade degradation up to 1750 K at the beginning of a nuclear severe accident. Energy-dispersive X-ray spectrometry (EDS) mapping indicated stratification of the absorber blade melt with formation of a chromium and boride-enriched layer. High-content-B- and C-containing material with increased melting temperature acted like a shielding and was found to prevent further relocation of control blade claddings. The interacted layers around the B4C-granules prevented direct steam attack of residual B4C. The results provide new insights for understanding of the absorber blade degradation mechanism under reducing conditions specific to Fukushima Dai-Ichi Unit 2 resulting from prolonged steam starvation.

Child and adult thyroid monitoring after a reactor accident (CAThyMARA): Technical recommendations and remaining gaps

Following a severe nuclear reactor accident large amounts of radioactive iodine are released in the atmosphere. Health consequences estimation for the affected population is based on the dose assessment which can be derived from in vivo measurements. Since the uptake of radioactive iodine increases the thyroid cancer risk of children it is of particular interest to dispose of technical recommendations on thyroid monitoring, particularly for children.This paper summarizes recommendations that have been issued by a group of European radiation protection specialists during the CAThyMARA project. It covers technical aspects such as the choice of instruments, the calibration process, the measurement preparedness, the dose assessment and communication issues. This paper also discusses remaining gaps and serves as an introduction to other papers of this special issue.

Characteristics of commercially available CdZnTe detector as gamma-ray spectrometer under severe nuclear accident

Characteristics of commercially available CdZnTe detector as gamma-ray spectrometer under severe nuclear accident

Transfer Effect Ratio of Loosely Coupled Coils for Wireless Power through CB Wall of NPPs

Unlike design basis accidents, some inherent uncertainties of the reliability of instrumentations are expected while subject to harsh environments (e.g., high temperature and pressure, high humidity, and radioactivity) occurring in severe nuclear accident conditions. By these conditions, instrumentations have had the bad situation as a station blackout (SBO) as the severe accident like Fukushima Daiichi nuclear disaster. In recent years, there has been an increasing interest in wireless power transfer technology, in particular, significant processing has been charted for inductively coupled systems. In this paper, we introduce some new method as transfer effect ratio of loosely coupled coils for wireless power through the container building (CB) wall as an alternative method under a station blackout of severe accident conditions in nuclear power plants. As a result, it can be shown an equivalent circuit model that can describe wireless energy transfer systems via coupled magnetic resonances for thickness wall of the container building using a simulation code.

Analysis of the radioactive atmospheric dispersion induced by ship nuclear power plant severe accident

For the nuclear powered ship, the radionuclides will be released near the sea surface after a severe nuclear accident. The influence of sea surface on the atmospheric diffusion of radionuclides cannot be ignored. Based on the Lagrangian particle tracking model, the simulation model for atmospheric diffusion of radionuclides above the sea was established. The influences of sea absorption on atmospheric diffusion of radionuclides emitted at different heights were analyzed. It is shown that if emission height is low, the time-integrated concentration of radionuclides in the air will be reduced significantly by the sea absorption. The lower the discharged height is, the more rapidly the concentration of radionuclides in the air near the source decreases. If the emission height is high, the time-integrated concentration of radionuclides in the air will increase firstly and then decrease with the increase of distance. The higher the emission height is, the lower the time-integrated concentration of radionuclides in the air near the sea surface is, and the smaller the amount of radionuclides absorbed by the sea is.