Severe Accident Conditions Research Articles

Advanced Analysis of Structural Performance in Novel Steel-Plate Concrete Containment Structures

This paper investigates the structural performance of novel steel-plate concrete containment structures, focusing on third-generation nuclear power plants. To address the challenges of increased complexities and costs associated with double-layer containment designs, this study explores the potential of steel-plate concrete structures to enhance safety, economic efficiency, and construction simplicity. The steel-plate concrete structure, characterized by its core concrete and dual steel plates, shows superior compressive strength, bending resistance, and elastoplasticity. Extensive numerical analyses, including finite element modeling and thermal-stress coupling, were conducted under various load conditions. Under structural integrity test conditions, the maximum radial displacement observed was 24.59 mm. Under design basis conditions, the maximum radial displacement was 47.61 mm; under severe accident conditions, it was 53.83 mm. The ultimate bearing capacity was 0.91 MPa, 2.17 times the design pressure. This study concludes that the steel-plate concrete containment structure maintains a high safety margin under all tested conditions, with stress and strain well within acceptable limits. It can effectively serve as a robust barrier against radioactive leakage and malicious impacts, providing a viable alternative to conventional containment designs.

Experimental study on heat transfer characteristics of steam condensation in the presence of air and CO2 outside a vertical tube

Under severe accident conditions, in addition to steam and air, there are also other non-condensable gases inside the containment, such as CO2 produced by the molten corium concrete interaction (MCCI). Due to the significantly higher molecular weight and density of CO2 compared to steam and air, CO2 will have a significantly different impact on the steam condensation process compared to air. However, few scholars have paid attention to the impact of CO2 on the condensation process, and there is no empirical correlation applicable to this condition. Therefore, this article investigates the effect of different CO2 concentrations on the steam condensation heat transfer coefficient (CHTC) in the range of system pressure from 0.2 to 1.3 MPa and subcooling from 30 to 130 °C through experimental methods. Meanwhile, based on relevant experimental data, a new correlation suitable for calculating the steam CHTC in the presence of CO2 and air was fitted. The results indicate that under low steam concentration conditions, the addition of CO2 will reduce CHTC, while under high steam concentration conditions, CO2 will increase CHTC. The increase in system pressure will strengthen the condensation promotion effect of CO2 and weaken its inhibitory effect on condensation. In addition, by comparing experimental data under different component conditions, it was found that the new empirical correlation has high computational accuracy, and the calculation results of this correlation can contain more than 90 % of the data within a 15 % error range.

Development of axisymmetric shell model for creep deformation analysis of reactor pressure vessel lower head

Under severe accident conditions of nuclear power plants, the lower head of a reactor pressure vessel can be at risk of creep deformation and potential failure. This study presents a computational model that can represent the creep deformation of the lower head through the application of shell theory. To account for the large deformation due to creep, kinematic equations are derived and implemented. Particularly, this study provides details of the mathematical formulation, which were lacking in previous studies. The analysis results using the developed model are compared with those of finite element analysis, considering deformation history, stress distribution, and deformed shape. A dimensionless time for the creep deformation is proposed based on the results, which can characterize the effective strain at failure. Finally, the developed model is applied to OECD lower head failure test for validation.

Exothermic characteristics evaluation of the Pt coated α-Al2O3 catalyst in bed for passive airborne radioactive material reduction system under forced flow conditions

After the severe accident at the Fukushima nuclear power plant, there has been heightened interest in the development of severe accident countermeasures. As part of these efforts, a passive filtration system capable of passive operation during power loss is under development.This paper focuses on evaluating the performance of a catalyst-filled bed designed to create a temperature gradient through exothermic reactions between combustible gases generated during accidents and a catalyst. The temperature gradient facilitates the formation of natural convection within the system, enabling passive operation. The catalytic bed is designed in a bed-type configuration filled with exothermic catalysts. To ensure physical and chemical stability under severe accident conditions, heat resistance tests were conducted using inorganic materials as a base to select an optimized exothermic catalyst. A test apparatus was designed for unit evaluation testing of the catalyst-filled bed equipped with the catalytic bed. Exothermic characteristics evaluation tests were conducted by varying the hydrogen volume fraction and face velocity, measuring heat generation in the catalytic bed under each condition. The correlation between these variables and the heat generation performance of the catalytic bed was determined. These findings will be utilized to inform the design parameters of the passive airborne radioactive material reduction system.

Concept design, application and optimization of operational parameters for new forced safety injection tank system

Background The engineered safety systems are designed to execute fundamental safety functions encompassing reactivity confinement, reactivity control and decay heat removal. Failure of any one of these functions can result in severe accident conditions. Passive systems have been implemented as a better option for plant safety. This research proposes a modification of existing safety injection tanks to the new forced safety injection tanks (FSITs) that utilize the backpressure from the pressurizer or steam generator to drive coolant into the reactor core under high pressure conditions. Methods FSITs aim to extend the coping-time during accidents like Station Blackout, providing an extended timing window for deployment of FLEX systems. The mathematical design is proposed and implemented into the thermal-hydraulic input of a nuclear power plant to demonstrate the system’s applicability that was demonstrated by leveraging the conventional PRA approach and risk quantification. Results The suggested system useful when used in the accident scenario of Station black out in-coincident with turbine-driven pumps fail to run. As the proposed design is a conceptual design, the optimization of associated operational parameters and set points is necessary. This optimization is performed using the risk analysis and virtual control environment-based Dynamic Probabilistic Risk Assessment framework. The method and findings of this study affirm that the coping-time for Station Blackout can be significantly extended, ensuring a substantial margin for the effective deployment of the FLEX. Conclusions A concept design of a passive forced safety injection system was suggested and demonstrated by integrating the mathematical model into a thermal-hydraulic model of a nuclear power plant. The parameters were optimized and the results demonstrated that the new system was effective in recovering the nuclear power plant from the accidents such as SBO with turbine-driven pumps fail to operate.

Three-dimensional CFD analysis of PHWR exposed core under postulated severe accident condition

A three-dimensional steady-state thermal–hydraulic analyses of a large sized Pressurized Heavy Water Reactors (PHWRs) with varying levels of exposure is carried out for degraded core heat transfer assessment and presented in this paper. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in PHWRs. Such exposure of fuel channels leads to their heat-up, governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer and thermo-chemical interactions among the core constituents. In order to understand such complex behavior and plan the severe accident management guidelines, numerical analyses of the exposed core are of necessary. This paper illustrates the influence of the temperature and velocity profiles vis-à-vis different regions of the exposed core on the fuel channel heat-up. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is reported. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, the advanced stages of fuel channel exposures are observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation from Zircaloy-made calandria tube surface and steam interaction. This leads to supplemental power addition for fuel channel heat-up. Moreover, it is interesting to note that the heat-up of fuel channels is restricted to central regions of the exposed core due to degraded steam cooling and higher-powered fuel channels. It is worth noting that fuel channel disassembly is observed in the advanced levels of fuel channel exposures. Moreover, fuel channel length equivalent to nearly 7–8 fuel bundle lengths undergo disassembly depending upon the fuel channel elevation and location in the core. The calandria vault water is observed to undergo film boiling at the onset of fuel channel disassembly.

Study on the lower head failure in severe accidents Part II: Thermal-mechanical coupling behavior analysis on CAP1400 reactor lower head

Under severe accident conditions with core meltdown, large amounts of molten material relocate to the lower head at high temperatures. Significant thermal and mechanical loads may lead to failure of the reactor pressure vessel (RPV) lower head, spreading radioactive fission products. Lower head failure time and location are crucial for accident mitigation and consequence assessment. This study developed a thermodynamic failure analysis model for the lower head by investigating material failure mechanisms, coupling this into an Integrated Severe Accident Analysis code (ISAA), and conducting experimental validation and applied research. Part I mainly introduces the thermodynamic analysis module development, validation against OLHF experiments, and preliminary failure criteria analysis. Results indicate the model accurately predicts lower head failure behavior, and Larson-Miller life fraction and Hedl-Dorn parameter criteria accurately assess lower head integrity. Part II presents applied research after validation and preliminary analysis. Using the improved ISAA, melt pool behavior and lower head thermal–mechanical response were analyzed for a CAP1400 small break loss-of-coolant accident (SBLOCA), assuming multiple safety system failures. Results indicate the bottom of the lower head melt pool didn’t form a hard shell due to high heat release rate, only an oxide shell layer around the second layer. Decay heat from melt pool components cannot be dissipated, leading to lower head failure. Further External Reactor Vessel Cooling (ERVC) heat transfer model analysis and research are needed in ISAA. The two developed mechanical analysis models predict material failure at high temperatures, and the timing and location of lower head failure are consistent, at approximately 16,000 s at 25.38°±5.47° on the lower head. The predicted failure mode is similar to Part I validation, indicating both Larson-Miller life fraction and Hedl-Dorn parameter criteria provide consistent lower head integrity judgments.

Protective layers of zirconium alloys used for claddings to improve the corrosion resistance

Abstract Zirconium alloys are used as a cladding material for fuel elements in nuclear reactors. In the case of severe accident conditions, the possible rapid oxidation of zirconium in steam or/and air may result in intense hydrogen generation and hydrogen–oxide mixture explosion. Advanced technologies for increasing the corrosion resistance of claddings are being investigated in two directions: (a) protective coatings on Zr alloys and (b) the use of new materials for claddings. Coatings with silicon may provide a more protective barrier than the ZrO2 films formed on an alloy cladding during nuclear plant operations. These coatings may also serve as a protective barrier during high-temperature accidents. The current work aimed at developing protective coatings with silicon on zirconium alloys. Multielemental Zr–Si–Cr coatings were formed on Zry-2 alloy using the physical vapor deposition (PVD) method. Long-term oxidation tests were carried out under the following conditions: 360°C/195 bar/63 days/water-simulating PWR water. Obtained results show the protective character of formed layers. The material in the form of silicon carbide grains covered with yttrium–aluminum garnet (SiC + YAG) was prepared using the sol–gel method. The formed powder is the main component for coating formation on Zr–1Nb alloy using the method of suspension plasma spraying (SPS).

Neutronic Analysis of Accident-Tolerant Cladding Materials in 3D Full Core BEAVRS PWR Benchmark Using OpenMC Code

After the Fukushima Daiichi nuclear accident in 2011, the performance of nuclear fuel during accidents became a matter of great concern. To address this, a new type of fuel technology called accident-tolerant fuel (ATF) has been developed with the goal of enhancing the ability of light water reactors (LWRs) to withstand severe accident conditions. Iron-based alloys have been suggested as potential candidates for fuel cladding due to their favourable thermomechanical properties, lower reactivity with steam, and lower hydrogen generation. This study evaluates the neutronic performance of C26M (a 2nd generation nuclear grade FeCrAl alloy), APMT™, 310SS, and 304SS cladding materials by comparing them with Zircaloy-4 cladding in a 3D PWR core at the beginning of the cycle (BOC) using OpenMC code. The results revealed that the neutronic penalty varied for different alternative cladding materials where C26M exhibited the lowest neutronic penalty value of -12551 pcm, while 310SS demonstrated the highest with a value of -17855 pcm. Additionally, important parameters in the reactor core such as neutron spectrum, reactivity coefficients, boron worth, control rod bank worth, power distribution, and radial thermal neutron flux distribution are evaluated and discussed. The analysis results showed that C26M provided a significantly higher level of neutronic performance compared to APMT™, 304SS, and 310SS. Although this study primarily focused on the neutronic performance of PWRs at BOC, future research should encompass fuel depletion analysis to delve deeper into the potential of alternative cladding materials.

Investigating the possibility of using a mixture of thorium with different fissile materials as a fuel in TRISO particles for the PBMR-400 reactor

The increasing demand for nuclear reactors as a source of clean energy forced nuclear researchers to search for ways to enhance the safety operation of nuclear reactors. This paper discusses the viability of using different thorium compounds ((233U, Th)O2, (235U, Th)O2, (RGPu, Th)O2) as an accident tolerant fuel/advanced technology fuel (ATF) for Pebble Bed Modular Reactor 400 (PBMR-400). A comprehensive analysis of thorium-based fuel performance was carried out and compared with UO2 to develop a novel nuclear fuel with the capability to tolerate severe accident conditions. The neutronic characteristics such as the infinity multiplication factor, initial heavy metal concentration, RGPu concentration and fission products concentration were analyzed. In addition to, the main safety parameters have been calculated to distinguish the effect of the proposed fuels on them. The thermal spectrum and thermal output power have been visualized to give a good insight view inside the core about the distribution of the heat. In general, the neutronic analysis demonstrated good behavior of the examined thorium-based fuel in the PBMR-400 and especially in the case of (RGPu, Th)O2. PBMR-400 showed excellent performance in incinerating a large amount of reactor-grade plutonium when it was used in a kernel fuel bed.

Effects of boric acid on volatile tellurium in severe accident conditions

Boric acid is used in light-water nuclear reactors to control the reactor and is expected to be present as part of the chemistry of a severe accident. Therefore, its influence on other prominent species expected in an accident must be investigated. One such species is tellurium. In the present study, tellurium is volatized, and boric acid is dissolved and injected into the system as a means of studying the interaction between it and tellurium. The experiments were evaluated with ICP-MS and XPS. Results suggest that while there is no direct interaction, boric acid still affects the tendency for tellurium to oxidize. In general, less oxidation was detected in the presence of boric acid than in its absence, especially at high temperatures. The species formed upon oxidation was determined to be TeO2. Since tellurium metal is more volatile than TeO2, this may have implication in a wider severe accident context.

Probabilistic Assessment of Ballooning of Pressure Tube under Severe Accident Conditions

For postulated accident scenario of loss of coolant accident coupled with the failure of emergency core cooling system, the pressure tube will deform due to presence of decay heat. The pressure tube deforms due to creep and this shall ultimately result in a contact with the calandria tube. In this work, the time for this contact to happen is calculated considering the uncertainties in mechanical and thermal material properties of the material. Finite element analysis of temperature distribution in the pressure tube has been carried out considering heat dissipation by conduction, convection and radiation. The source is heat generation due to decay heat of fuel bundles inside pressure tube. The ballooning of the pressure tube material is modelled using the Norton’s creep law. The finite element model is validated using experimental results from literature. As the finite element model is very complex and time intensive, it is difficult to employ the traditional uncertainty propagation methods like Monte Carlo simulations. The uncertainty analysis in this work is performed using Wilk’s method. The time duration when the contact of pressure tube and Calandria tube would happen and the temperature of the pressure tube at that instant is estimated. The results are presented in terms of the 95 percentile values with a confidence of 95 percent. The analysis gives an upper bound to the contact time between pressure tube and Calandria tube during loss of coolant accident condition subject to unavailability of emergency core cooling system.

CFD analysis of PHWR exposed core under postulated severe accident condition

A two-dimensional steady-state thermal-hydraulic analyses of the exposed core with varying levels of exposure is carried out for heat transfer assessment. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in Pressurised Heavy Water Reactors (PHWRs). Such exposure of fuel channels leads to their heat-up which is governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer. In order to understand such a complex behaviour as well as to plan the severe accident management guidelines, numerical analyses of the exposed core are required. In this paper, one-of-a-kind thermal-hydraulic analyses of the exposed reactor-scale core geometry are presented for row-wise exposures of fuel channels. These numerical analyses are terminated at the onset of fuel channel disassembly i.e., loss of geometrical intactness of the exposed core. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is brought out in this study. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, for the advanced stages of fuel channel exposures, the heat transfer is observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation arising from Zircaloy and steam interaction. Moreover, it is interesting to note that the disassembly of fuel channels is restricted to central regions of the exposed core.

Numerical analysis of sodium-argon two-phase flow in narrow channels under severe accident conditions in pool type SFR

A two-phase flow model for sodium-argon mixture flow through the reactor vessel head penetrations under the energetic core disruptive accident in a sodium fast reactor (SFR) is presented. The model evaluates the total sodium mass in sodium-argon mixture released to the reactor containment building (RCB) through penetrations. The driving pressure at the penetration entrance is given as input. The two-phase frictional and local pressure drops, inertial and piezo-metric heads are accounted for in the model. Depending on the argon volume fraction, either a homogenous or separated flow approach is adopted to evaluate the mixture flow. Reactor-scale analysis show that the presence of argon in vessel head penetrations can significantly mitigate the total sodium release mass during the slug impact phase. Also, the two-phase frictional pressure drop plays a mitigating role and needs to be included in the model formulation to avoid overestimation of the sodium release mass. Fluid-structure interaction effects such as vessel head bending and hold-down bolt elongation have negligible influence on sodium release mass as long as the deformations are within the elastic limit. The model is conceived to be part of an integrated code system to evaluate the sodium fire and the maximum RCB pressure under severe accident conditions in a pool-type SFR.

Gas phase interactions between tellurium and organic material in severe nuclear accident scenarios

One of the volatile fission products potentially released during a severe nuclear accident is tellurium. Recently, tellurium has been shown to form volatile organic tellurides from paint solvents in aqueous solutions which could further increase the release. However, not much is known about the possible interactions between tellurium and organic material in gas phase. In this study, tellurium was exposed to conditions simulating the containment atmosphere during an accident. Moreover, volatile organic compounds were introduced to the gas phase representing the presence of organic material in the containment during the accident. The results suggest that tellurium aerosols and organic material interact in the gas phase which was observed as an increase in the gaseous fraction and a change in the XPS spectra. Although no exact species were identified, the results raise questions about the behavior of tellurium in severe accident conditions, especially, regarding the reactions involving organic material.

The study on the main factors affecting the response analysis of nuclear fuel transport cask drop

The nuclear fuel transport cask is an important transport equipment in the nuclear power industry. Rational design requires the structure to maintain integrity and tightness after experiencing severe accident conditions such as a 9 m drop. Therefore, the design and analysis are the key links of container forensics. The calculation of drop analysis involves many factors such as nonlinear material constitutive, complex contact, high strain rate, and large deformation which brings great challenges to the reliability of simulation analysis. Summarizing and sorting out the key factors in the design and analysis of transport casks can solve the practical needs of engineering and guide the analysis and calculation of transport cask drop. This study considered the perspectives of impact limiter design, assembly clearance, analysis model, environmental factors, dropping angle, and the calculation and comparison of buffer materials' dispersion. A two-degree-of-freedom vibration system was used to simulate the impact of clearance collision, and a beam model was applied for the investigation of the response of assemblies. The considerations of the analysis model and the impact of the environment on buffer materials were expounded. Finally, the main conclusions are as follows: the dispersion of buffer materials should be considered in the design. The wood with low compressive strength will cause greater compression deformation, and the wood with high compressive strength will cause greater peak acceleration. The frequency, mass, and clearance of the internal components and the container all affect the acceleration response of the internal structure. The acceleration response is sensitive to the angle under the side and end drop conditions.

An analytical model to decompose mass transfer and chemical process contributions to molecular iodine release from aqueous phase under severe accident conditions

Radioactive iodine is a representative fission product to be quantified for the safety assessment of nuclear facilities. In integral severe accident analysis codes, the iodine behavior is usually described by a multi-physical model of iodine chemistry in aqueous phase under radiation field and mass transfer through gas-liquid interface. The focus of studies on iodine source term evaluations using the combination approach is usually put on the chemical aspect, but each contribution to the iodine amount released to the environment has not been decomposed so far. In this study, we attempted the decomposition by revising the two-film theory of molecular-iodine mass transfer. The model involves an effective overall mass transfer coefficient to consider the iodine chemistry. The decomposition was performed by regarding the coefficient as a product of two functions of pH and the overall mass transfer coefficient for molecular iodine. The procedure was applied to the EPICUR experiment and suppression chamber in BWR.

Dose Rate Assessment Around the PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko

Abstract Passive containment filtered vent (PCFV) was installed in Nuclear Power Plant Krsko (NEK) in 2013 as part of the safety upgrade program. It is intended for severe accident consequences prevention and mitigation by ensuring the containment integrity. When the pressure in the containment reaches limiting value, containment atmosphere is released into the environment through the PCFV system exhaust line. But, before release in the environment, the containment atmosphere passes through five aerosol filters in the containment and one iodine filter in the auxiliary building (AB) to reduce isotopic activity. In this paper, dose rates around the exhaust line of the PCFV system resulting from radioactivity release in case of a severe accident were determined in a four-step methodology. The assumed severe accident scenario is a beyond design basis accident station blackout (SBO) in NEK, which was simulated using the MELCOR code. Its results were input for the radionuclide transport and removal and dose estimation (RADTRAD) radiological calculations to obtain the activities released in the containment. These activities were then transformed into the gamma source intensity and spectrum using the ORIGEN-S libraries. This form of the source term is required for Monte Carlo calculations which were performed using the MCNP6.2. Two Monte Carlo calculations were performed. One for which the radiation source was modeled to emanate from the containment atmosphere and the other from the PCFV duct fluid. The main reason for the calculation was to assess limiting dose rates around PCFV duct (radiation monitor location) during actuation after severe accident. That is why the model is simple and conservative. The other task was to demonstrate that this location is not suitable for longer personnel presence in case of equipment failure during the PCFV actuation. Due to conservative assumptions, predicted dose rates are the highest expected at that location for any severe accident scenario.

Effect of molybdenum release on UO2/MOX fuel oxidation under severe light water reactor accident conditions

To investigate the Mo release behavior and its influence on the oxidation of UO2/mixed oxide (MOX) fuel, the oxidation and evaporation (O&E) behaviors of Mo powders and their influencing mechanisms with respect to oxygen partial pressure around powders (pO2s) were researched by using a thermogravimetric analysis technique under Ar-0.1 % O2, Ar-1.0 % O2, and Ar-1.5 % O2 atmospheres in the temperature range from 1473 to 1673 K. The study results reveal that Mo O&E could quickly consume oxygen, substantially decreasing pO2s. This decrease in pO2s facilitates the attainment of an oxygen mass balance by enhancing the diffusion rate of O2(g) towards the powder surfaces and decelerating the O&E rates of the powders. Subsequently, the oxygen consumption rate with respect to Mo release (FO-con) and the potential oxidation rate of UO2 nuclear fuels (FO-ads) under the typical accident conditions created in VERCORS were estimated using the latest analytical model of Mo and the existing fuel oxidation model, respectively. The results demonstrate that, as accident proceeds, FO-con has the potential to surpass FO-ads, thereby retarding the oxidation of nuclear fuel.

Two-Way Time-Dependent Prestress Losses of Prestressed Concrete Containment with Bonded Prestressing Strands

Prestressing plays a pivotal role in ensuring the tightness and integrity of prestressed concrete containment in nuclear power plants. The prestress loss reduces the compressive stress in concrete resulting from the prestressing strands and increases the risk of containment leakage under severe accident conditions. Therefore, the accurate prediction of prestress loss is essential for the design and in-service management of prestressed concrete containment. Unlike one-way beams or girders in building structures and bridges, two-way prestressing systems are used in prestressed concrete containment. In the current simplified method for evaluating time-dependent prestress loss, the interaction of concrete creep in two directions resulting from the two-way prestressing strands and the influence of the steel liner and mild steel rebars in two directions are neglected. In this study, based on the principle of creep superposition, the age-adjusted effective method for the creep estimation of concrete, and considering concrete shrinkage, concrete creep, and the relaxation of prestressing strands, as well as the influence of the steel liner and mild steel rebars in two directions, a sectional analysis is performed for prestressed concrete containment with bonded prestressing strands, and equations for calculating the two-way time-dependent prestress losses are derived. The results of the two-way time-dependent prestress losses predicted by the derived equations are compared with those of tests in the literature, and great agreement is achieved. Finally, a case study is given to show the application of the proposed method for the prediction of prestress loss in prestressed concrete containment in the nuclear power plant.