Reactor Building Research Articles

Seismic nonlinear interaction and uplifting effects between the nuclear island building and soil

Since the Fukushima nuclear accident, seismic design requirements have increased, necessitating a more refined evaluation of base uplift for weakly anchored nuclear island buildings. The elevated cooling water over 3000 t is a distinguishing feature of third-generation reactor buildings (RBs). To assess potential safety-adverse coupling effects, seismic wave propagation process through viscoelastic soils, discontinuous interfaces, and sloshing water was analysed. To facilitate nonlinear systematic simulation and parametric study, the traditional soil-structure interaction direct methods are improved and programmed on: thickness-independent viscous-spring elements for stable static-dynamic boundary transitions; seismic forces separated from the artificial boundary for flexible mesh planning; elevated water simulated with gravity stacking and hydrodynamic action. A nuclear island region encompassing water-cooled RBs and underlying soil cushion layer was investigated as a typical scenario to reveal rules using the proposed high-precision simulation platform. Non-negligible structural risks and correlations between multiple factors were quantitatively elucidated according to simulation findings. A time-lagged increase in the base overturning moment may occur because of water oscillations. As the water level increases to 85 %, the uplift height increases, adversely affecting the in-structure response spectra. Earthquake-induced intense uplift may lead to an increase in hydrodynamic pressure along the inner wall. The research results support the practical assessment of buildings with water cooling facilities located in soft composite sites with high seismic intensities.

Study on seismic performance and embedded effects on reactor buildings cluster considering nonlinear structure-soil-structure interaction

In the new nuclear power regulations, the effect of layered sites is essential for refined seismic analysis, especially since the Fukushima nuclear accident. There is a lack of systematic approaches for the coupling simulation of cluster-embedded nuclear structures in layered soils. To meet this need, three challenges arise: (a) refinement of modeling and mesh planning for the structure-soil-structure interaction (SSSI) system; (b) intensified nonlinearity of the surrounding soils due to relative motion in the cluster; (c) solving efficiency problems for large-scale SSSI analysis. A complicated-geometric reactor building-soil system was modeled using an improved splice mesh considering clusters and embeddings. An advanced global wave motion method mainly characterized by the surface elemental form of elastic support and radiation-damping of a semi-infinite site, enhanced with a GPU-accelerated dynamic solution and a viscous-elastic soil model. The proposed procedure was verified and benchmarked within engineering scenarios, leading to the following conclusions. In cluster simulations, embedding leads to some horizontal damping effect, but the vertical response spectrum can increase substantially on higher floors. Ignoring the cluster effect can lead to a conservative seismic design of the reactor core. Batch parametric discussions provide empirical curves aiding qualitative estimations in nuclear engineering. A systematic perspective for high-accuracy seismic evaluation of structural clusters is supported.

Development and Experimental Validation of a Heat Transfer Model for Spilled Molten Salt Pools

A spill of radionuclide-bearing molten salt is one of the major postulated events that needs to be analyzed for liquid fluorine salt–cooled high-temperature reactor (FHR) or molten salt reactor licensing purposes. In this postulated event, radioactive source term materials (RSTMs) in the molten salt are discharged from the reactor vessel to the reactor building. The release of RSTMs from the spilled salt pool to the gas space in the reactor building is expected to be controlled by the cooling behavior of the spilled salt, including the growth and shrinkage of the solid crust on the surface of the spilled salt pool. This paper presents a simulation model for spilled salt pool heat transfer and validation efforts. The validation data come from two molten salt spill tests that were performed recently: the PELE2 test by the Rapid Experimental Laboratory of Kairos Power LLC (KP) and the Argonne salt cooling test conducted by Argonne National Laboratory. The former was a large-scale test involving kilograms of molten spilled FLiNaK salt, and the latter was a relatively smaller-scale test targeting various processes associated with a salt spill event. Both tests generated valuable data sets that can be used to assess salt cooling and validate evaluation models. This paper provides a new one-dimensional model that can simulate the cooling process of a spilled salt pool as well as the thermal responses of heat structures, such as the stainless steel liner and the concrete below the salt. The model has been implemented as part of KP-SAM code, which is a branch of the systems code SAM specific to KP FHR. The simulation results of the model are compared with the data of the PELE2 and Argonne tests, and reasonable agreements are observed between the model and test data.

VIKTORIA experiments in the frame of R&D project on sump filtration during a loss of coolant accident

During a Loss of Coolant Accident (LOCA), in PWR’s, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability and by the Containment Spray System (CSS) to remove residual heat and to maintain containment integrity. After the drainage of the Refueling Water Storage Tank (RWST), water is taken from sumps in the lower part of the reactor building. A filtering system is implemented to collect debris produced by the pipe break as well as other latent materials, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the ECCS and CSS systems. IRSN has launched an experimental R&D project investigating the clogging of sump filters by integral tests performed in the VIKTORIA loop. The main objectives are to investigate the head loss of the filter (physical and chemical clogging) for prototypic upstream debris source term in relevant thermal hydraulic conditions (water temperature and flow velocity on the filter surface) in compliance with the temperature profile and the chemistry of the water in sumps during a LOCA transient. For that, the VIKTORIA loop was equipped successively with two types of 2 m2 filters used in 900MWe NPP’s. The tests highlight the settling of the largest particles (concrete, painted chips) and part of the fibers; the transport of debris (roughly 55 to 65 % of the injected debris source term) leads to the physical clogging of the filter. The debris carried to the filter generate at 80 °C (with chemistry) a very quick increase of the pressure drop across the filter (≈7 kPa) that could be due to rapid chemical effects further to fibers corrosion. At the end of the 80 °C plateau, a pressure drop increase was observed due to the temperature decrease in agreement with the water viscosity evolution. The two types of filters (rectangular pockets or planar types) behave very differently with rather low head losses for the second type; ≈1.5 kPa. Nevertheless, downstream debris source term appears to be more significant with the planar filter (compared to the filter with pockets) during the first hour of injection before the fibrous debris bed formation. We have also observed a more stabilized “cake”, during the (7 days) medium term evolution compared to that for rectangular pockets filter due to motion of debris inside pockets. The recent experiments performed with less amount of fibers (by replacement of part of fibrous materials by RMI metallic insulation) led to significantly reduce the head loss without any consequences on the downstream behavior.

Comprehensive integral simulation of first 3 weeks of severe accident at Fukushima Daiichi Unit 1. Part II – Simulation of the in-vessel and ex-vessel phases

This paper is the second part of the summary presenting the results of IBRAE work over last 12 years on severe accident investigation at Unit 1 of Fukushima Daiichi with the SOCRAT code. The paper supplements the part I that was published previously, and is dedicated to the description of one-through integral simulation results of the 3 weeks of accident unfolding starting from the tsunami arrival. The integral simulation encompasses the different processes at in-vessel and ex-vessel phases of the accident including the processes in reactor building. The important issues discussed are the hypothesis on the induced leak from a steam line, the operation of isolation condenser, the possible early combustion of hydrogen in the reactor building room that houses the drywell head, as well as the time, conditions and regime of hydrogen combustion in the reactor hall, the flowrate of alternative water injection, and the depth of concrete ablation. The main criterion for checking the adequacy of assumptions and the simulation results is the agreement with factual information available including the measurement data and the results of inverse calculations of the source term. The shortcomings of current simulation with the SOCRAT code are briefly considered as a perspective for improving the realism of results in the next studies.

Analysis of the Long-Term Interaction Between Molten Core and Dry Concrete at Fukushima Daiichi Unit 1

The latest investigations of Fukushima Daiichi Unit 1 have demonstrated that corium attack to the pedestal walls and pedestal floor has occurred in Fukushima Daiichi Unit 1 to a certain extent. The results of past analytical benchmarks, such as the Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF project), have agreed with this finding. However, the latest investigation does not show evidence of unlimited molten core–concrete interaction (MCCI), which is one of the main discrepancies from the BSAF project. More recently a MCCI benchmark has been launched in the context of the OECD/NEA project ARC-F (Analysis of Information from Reactor Building and Containment Vessels of Fukushima Daiichi Nuclear Power Station). In the benchmark, common geometry, boundary, and initial conditions have been selected among all the participants. The results show an improved agreement among different codes for what concerns overall erosion, corium temperature, and hydrogen generation, confirming that to some extent, the earlier scatter found in these variables came from differences in the MCCI scenario modeled by each partner. However, common unlimited erosion, not observed by onsite visual inspections, is still predicted. Understanding the origin of this deviation might provide insights into boundary conditions, model drawbacks, or ill-posed assumptions that might need to be revisited (e.g. interfacial temperature, effective heat transfer coefficients, concrete heat transfer). In this paper, a summary of the overall results and a discussion of modeling and boundary conditions is presented to disclose the results of the activity and the future steps to be taken in the OECD/NEA project FACE (Fukushima Daiichi Nuclear Power Station Accident Information Collection and Evaluation).

Conversion augmentation of an industrial NH3 oxidation reactor by geometry modification to improve the flow and temperature pattern uniformity using CFD modeling

BackgroundThe optimization of flow and temperature patterns in industrial reactors is crucial for achieving efficient and uniform chemical reactions. This study's major goals are to pinpoint possible areas for improvement in the NH3 oxidation reactor's performance and to deal with the problem of uneven flow distribution inside the reactor. MethodsIn this study, the reactor building design has been changed by extending the feed pipeline vertically and increasing the number of incoming feed streams in order to achieve uniformity in the property distribution on the catalyst surface of an industrial NH3 Oxidation reactor. Thus, using the CFD approach and the finite volume method, a three-dimensional model has been suggested. The results are contrasted with the actual geometrical configuration. The property alteration along the catalyst surface and the reactor length have been assessed. Significant findingsBy expanding the feed pipeline, the flow pattern at the reactor entry is fully developed and becomes uniform. As a result, NO2 production could go up by as much as 11 %. The rates of NH3 conversion, NO yield, and HNO3 generation consequently increased by 12.5 %, 3.1 %, and 8.0 %, respectively. Additionally, this alteration results in a uniform distribution of temperature and pressure across the catalytic surface, prolonging the lifetime of the catalyst. The pressure and temperature difference over the surface of the catalyst with the original reactor configuration was also found to be approximately 250 Pa and 423.15 K, according to the data. Pressure and temperature difference were reduced to 15 Pa and 273.15 K, respectively, as the feed line's length was increased at the same time.

Analysis of particles containing alpha emitters in stagnant water in Fukushima Daiichi Nuclear Power Station’s Unit 3 reactor building

Particles containing alpha (α) nuclides were identified from sediment in stagnant water in the Unit 3 reactor building of the Fukushima Daiichi Nuclear Power Station (FDiNPS). We analyzed different concentrations of α-nuclide samples collected at two sampling sites, the torus room and the main steam isolation valve (MSIV) room. The solids in the stagnant water samples were classified, and the uranium (U) and total alpha concentrations of each fraction were measured by dissolution followed by inductively coupled plasma mass spectrometry and α-spectrometry. Most of the α-nuclides in the stagnant water samples from the torus and MSIV rooms were in particle fractions larger than 10 μm. We detected uranium-bearing particles ranging from sub-µm to 10 µm in size by scanning electron microscopy–energy-dispersive X-ray (SEM–EDX) observations. The chemical forms of U particles were determined in U–Zr oxides, oxidized UO2, and U3O8 with micro-Raman spectroscopy. Other short-lived α-nuclides (plutonium [Pu], americium [Am], and curium [Cm]) were detected by alpha track detection, and the particles with α-nuclides was characterized by SEM–EDX analysis. α-nuclide-containing particles with several tens to several 100 µm in size mainly comprised iron (Fe) oxyhydroxides. In addition, we detected adsorbed U onto Fe oxyhydroxide particles in the MSIV room sample, which indicated nuclear fuel dissolution and secondary U accumulation. This study clarifies the major characteristics of U and other α-nuclides in sediment in stagnant water in the FDiNPS Unit 3 reactor building, which significantly contribute to the consideration of removal methods for particles containing α-nuclides in the stagnant water.

Design standardisation and seismic protection of SMRs through modular metafoundations

This paper investigates the seismic protection of the Nuward™ small modular reactor (SMR) building, focusing on design loading and beyond design basis earthquake (bDBE) conditions. The study aims to achieve two primary objectives: i) to enhance seismic mitigation of a SMR building under bDBE conditions, through the use of the innovative modular single-layer (SLM) and multi-layer (MLM) metafoundations (MFs); ii) to effectively standardise and harmonise SMR building designs in locations prone to beyond design basis conditions. To accomplish these goals and demonstrate the protective capabilities of the MFs, the study employs non-linear time-history analyses (NLTHAs) for both DBE and bDBE conditions. Along these lines, a reduced-order model was developed from a refined finite element (FE) model of the SMR building using the Craig-Bampton mode synthesis technique. Then, finite locally resonant modular MFs were designed and analysed using NLTHAs. Specifically, physics-based ground motion models (GMMs) were used to generate and select seismic triplets that mimicked DBE and bDBE scenarios for NLTHAs. Successively to achieve improved seismic performance, the optimization of the MFs was pursued by targeting the optimal number of columns, resonator parameters, and unit cell dimensions. Additionally, the deployment of inerters was considered, to significantly reduce the size of the MFs and enable their application in multiple layers for ultra-low frequency attenuation. The overall findings suggest that modular MFs meet seismic protection requirements, and positively contribute to the standardization process of SMR buildings, even in areas characterized by beyond-design seismic conditions.

Seismic mitigation analysis of three-dimensional base-isolated nuclear structures with soil-dependent isolation system under extreme earthquakes

The Fukushima nuclear accident in March 2011 has raised more stringent requirements for the ability of nuclear power plants (NPPs) to withstand extreme disasters such as extreme earthquakes. To address this challenge, this study proposed a new hybrid multi-dimensional passive control system that combines three-dimensional base isolation and geotechnical seismic isolation. In this system, high-damper rubber bearings were employed to isolate horizontal seismic motions, whereas disc springs, accounting for frictional forces, were used to mitigate vertical seismic motions. Additionally, a geotechnical seismic isolation system arranged under the base mat was leveraged to further reduce the seismic response of the superstructure. Taking a large-scale reactor building in complex layered sites as an example, the effectiveness of the hybrid multi-dimensional isolation system was studied using a wave motion method. The research results indicate that when subjected to extreme earthquakes, hybrid multi-dimensional isolation exhibits superior isolation performance compared with three-dimensional base isolation, which can significantly reduce the three-dimensional dynamic response of nuclear structures without increasing the displacement of the isolation bearings. The proposed hybrid multi-dimensional isolation system offers an effective solution to the challenges faced by three-dimensional isolation systems, thereby enhancing the seismic resistance of nuclear structures against extreme earthquakes.

Study on spent fuel heatup during spent fuel pool complete loss of coolant accident

Abstract After years of operation, nuclear power plants accumulate a substantial amount of spent fuel stored in spent fuel pools. Prior to the completion of dry storage facilities, this spent fuel remains in the spent fuel pools for several decades. The progression of accidents in spent fuel pools leading to fuel damage typically spans several days or even weeks, which may lead people to overlook the importance of spent fuel pool safety. However, accidents involving spent fuel pools have a characteristic of low probability but high consequences. Their significance has been reemphasized, especially after the Fukushima accident. This study conducts an analysis of the scenario where a spent fuel pool experiences a complete loss of water. A plant equipped with a BWR-4 reactor and a MARK-I containment structure is used as the reference plant. The analysis utilizes the MAAP5 code to assess the impact of different fuel cooling times and ventilation flow rate of secondary containment building on the fuel temperature variation, considering only air circulation for cooling. Based on the analysis results, the plant utilizes the normal ventilation system of the reactor building, which can maintain the fuel temperature below 565 °C with a cooling time of 1 year. Additionally, it was found in the study that adequate ventilation flow rates and sufficient cooling time can ensure that the fuel temperature remains below 565 °C, thereby preserving fuel integrity.

Experimental investigation on effective aerosol scavenging using different spray configurations with pre-injection of water mist for Fukushima Daiichi decommissioning

During the decommissioning of the Fukushima Daiichi nuclear power plant, it is important to consider the retrieval of resolidified debris both in air and underwater configurations. For the subsequent retrieval of debris from the reactor building, the resolidified debris must be cut into smaller pieces using various cutting methods. During the cutting process, aerosol particles are expected to be generated at the submicron scale. It has been noted that such aerosols sizing within the Greenfield gap (0.1–1 μm) are difficult to remove effectively using traditional spraying methods. Therefore, to improve the aerosol removal efficiency of the spray system, a new aerosol agglomeration method was recently proposed, which involves injecting water mist to enlarge the sizes of the aerosol particles before removing them using water sprays. In this study, a series of experiments were performed to clarify the proper spray configurations for effective aerosol scavenging and to improve the performance of the water mist. The experimental results showed that the spray flow rate and droplet characteristics are important factors for the aerosol-scavenging efficiency and performance of the water mist. The results obtained from this study will be helpful for the optimization of the spray system design for effective aerosol scavenging during the decommissioning of the Fukushima Daiichi plant.

The Application of Laser-Scanning 3D Model Reconstruction Technology for Visualizing a Decommissioning Model of the Heavy Water Research Reactor

Currently, over 100 nuclear power units globally have been in operation for more than 40 years. Hindered by the limitations of computer technology at the time, these nuclear facilities lack detailed electronic drawings. Activities such as equipment replacement and process circuit system modifications during operation result in discrepancies between paper drawings and actual conditions. Given the complexity and irreversibility of nuclear facility decommissioning activities, virtual simulation technology is often employed before the decommissioning process begins to assist in designing and validating decommissioning plans. Consequently, the creation of high-precision 3D models is crucial for subsequent decommissioning designs. Through innovatively utilizing laser-scanning 3D model reconstruction technology in the reconstruction of the model of China’s first heavy water research reactor undergoing decommissioning, this paper provides an overview of the process of laser-scanning 3D model reconstruction and its application in reconstructing the heavy water research reactor model. Using a 3D laser scanner, four decommissioning areas of the heavy water research reactor, including the reactor building, secondary water pump room, ventilation center, and low-level radioactive wastewater storage tank area, were subjected to 3D laser scanning. The acquired point cloud data from 572 scanning stations were processed using point cloud processing software for denoising, stitching, and triangulation. The triangulated model was then imported into modeling software for 3D reconstruction, ultimately establishing a digitalized model of the heavy water research reactor suitable for subsequent decommissioning simulation and design.

Detailed visualization of radioactive hotspots inside the unit 1 reactor building of the Fukushima Daiichi Nuclear Power Station using an integrated Radiation Imaging System mounted on a Mecanum wheel robot

ABSTRACT In the decommissioning of the Fukushima Daiichi Nuclear Power Station, understanding the distribution of radioactive substances and dose-equivalent rates is crucial to develop detailed decontamination plans and minimize worker exposure. In this study, we remotely visualized radioactive hotspots and dose-equivalent rate distribution in Unit 1 reactor building of the station using a Mecanum wheel robot equipped with a Compton camera, simultaneous localization and mapping device, and survey meter. We successfully visualized high-concentration radioactive hotspots on the U-shaped piping of the drywell humidity control system and the atmospheric control piping in the ceiling in front of the transverse in-core probe room. Furthermore, the hotspot location was identified in three dimensions using the Compton camera used to analyze the atmospheric control piping. By simultaneously analyzing the dose-equivalent rate data acquired by the survey meter and the hotspot locations visualized by the Compton camera, it was confirmed that the hotspots caused elevated dose-equivalent rates in the surrounding area. In the future, if this robotic system is used in unexplored areas, such as the upper floors of reactor buildings, it can provide information about the locations of radioactive hotspots and the distribution of dose-equivalent rates.

A Study on the Effects of Photogrammetry by the Camera Angle of View Using Computer Simulation

During the decommissioning activities, a movie was shot inside the reactor building during the investigation of the primary containment vessel by applying photogrammetry, which is one of the methods for three-dimensional (3D) reconstruction from images, to the images from this movie, it is feasible to perform 3D reconstruction of the environment around the primary containment vessel. However, the images from this movie may not be suitable for 3D reconstruction because they were shot remotely by robots owing to limited illumination, high-dose environments, etc. Moreover, photogrammetry has the disadvantage of easily changing 3D reconstruction results by simply changing the shooting conditions. Therefore, this study investigated the accuracy of the 3D reconstruction results obtained by photogrammetry with changes in the camera angle of view under shooting conditions. In particular, we adopted 3D computer graphics software to simulate shooting target objects for 3D reconstruction in a dark environment while illuminating them with light for application in decommissioning activities. The experimental results obtained by applying artificial images generated by simulation to the photogrammetry method showed that more accurate 3D reconstruction results can be obtained when the camera angle of view is neither too wide nor too narrow when the target objects are shot and surrounded. However, the results showed that the accuracy of the obtained results is low during linear trajectory shooting when the camera angle of view is wide.

Selection method for observation points using Bayesian LASSO at estimating radiation source distribution from air dose rates

When we decommission a reactor building, it is essential to identify the radiation source distribution for safety. It has been reported that the source distribution can be predicted from the measured air dose rates at appropriate observation points by minimizing an evaluation function using the Least Absolute Shrinkage and Selection Operator (LASSO). However, it is challenging to decide on suitable points in advance. Therefore, we estimate the posterior distribution from the prior distribution of the source amounts, calculated by the standard LASSO, using the Bayesian LASSO. We then assess the predictive distribution of the air dose rates at the candidate observation points from the posterior distribution. We select the additional observation points based on the variances of the predictive distributions. We confirmed that the method can estimate the source distribution with fewer additional observation points than when adding them randomly in most cases.

Evaluation of radiation dose caused by bremsstrahlung photons generated by high-energy beta rays using the PHITS and GEANT4 simulation codes

ABSTRACT High-energy beta-ray sources generate bremsstrahlung photons in the building materials, generated by the radioactive nuclides inside the nuclear reactor buildings emitted by the Fukushima Daiichi nuclear reactor accident. Therefore, evaluating the bremsstrahlung dose subjected to the workers in the building is crucial for radiation protection. The precision for assessing the bremsstrahlung dose was investigated by comparing the Particle and Heavy Ion Transport code System (PHITS) and the GEometry ANd Tracking (GEANT4) simulation code results. Behind various shielding plates (lead, copper, aluminum, glass, and polyethylene with multiple thicknesses), the water cylinder was set to obtain the absorbed dose and the deposited energy spectrum by the bremsstrahlung photons. In comparing the deposited energy spectrum, the spectral shapes have consistent trends. Below several tens of keV, a peak is observed in the PHITS spectrum for lead. While comparing the absorbed dose, most of the results from both codes correlate within an ~10% difference for 2.280 MeV beta-ray sources and an ~20% difference for 0.5459 MeV beta-ray sources, except for ~30% for 20 mm thick lead. Although some differences occurred, the evaluation results of both codes correlated well with the above precision.

Numerical Simulation of the Effect of Fusion Welding Processes on Residual Stress in Stainless Steel Pipe Weld Joints Used in Fast Reactors

The main structural material used in the building of Fast Breeder Reactors (FBR) is stainless steel of type 316L(N). Finite Element Model (FEM)-based simulations were carried out to understand the effect of various welding processes on the residual stresses induced in 316L(N) stainless steel primary sodium-carrying pipes with an inner diameter of 208 mm and a wall thickness of 5.6 mm. The welding processes considered include multi-pass Tungsten Inert Gas (TIG) welding, A-TIG, laser, hybrid laser-TIG, and Hybrid laser-Metal Inert Gas (MIG). First, single or hybrid heat sources are chosen based on the welding process, and then weld bead profiles obtained during simulation are made to match with that of the actual weld bead profiles obtained during experiments. Then, the optimised heat source is employed for the thermal analysis. The thermal analysis output is then successively linked and provided as input to the mechanical analysis, which uses an isotropic hardening model to predict hoop and axial residual stresses in pipe weld joints. The results show that the A-TIG weld joint exhibited lower hoop stress (165 MPa at inner diameter and -2 MPa at outer diameter) in comparison with that of the other weld joints. The lower residual stress is attributed due to the straight-sided weld bead, lower peak temperature, and slower cooling rate of the A-TIG welding. Therefore, A-TIG welding is recommended among the chosen fusion welding processes for welding of 316L(N) stainless steel primary sodium carrying pipes to minimise the residual stresses and mitigate the premature failure of the pipe weld joints.

Strategic analysis on sizing of flooding valve for successful accident management of small modular reactor

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

Assessment of radiological dose and emergency planning zones of TRIGA Mark-II research reactor of Bangladesh due to a postulated severe accidental condition

Radiological dose and emergency planning zones assessment has been conducted for dry and rainy seasons in Bangladesh for the TRIGA Mark-II research reactor. A postulated severe accident of a plane crash was considered in this analysis. Due to the full collapse of the reactor building and core damage, core inventory radionuclides were released at their highest release fraction. In this instance, doses due to the release of radionuclides such as Xe, Kr, I, and Cs were estimated using two different atmospheric dispersion models utilizing meteorological data collected from the Bangladesh Meteorological Department. The source term analysis has been performed using the ORIGEN 2.1 code. The HotSpot 3.1.2 and RASCAL 4.3 codes were used for the assessment of radiological doses. Based on the results of both the RASCAL 4.3 and HotSpot 3.1.2 codes, the TEDE and thyroid CDE were found to be substantially higher than the allowable limit in a large area. The dose values were compared to the dose standards of the IAEA General Safety Requirements (GSR) Part 7 and the United States Environmental Protection Agency (U.S. EPA) to identify the evacuation, sheltering and iodine prophylactic administration distances for both dry and rainy seasons. The evacuation and sheltering distances were obtained at approximately 2.25–5 km, 4.5–15 km, and 3–6.6 km, 3.7–8.5 km according to U.S. EPA dose criteria in the dry and rainy seasons, respectively. On the other hand, according to IAEA GSR part 7 dose criteria, the evacuation and sheltering distances are obtained at 1.5–3.5 km and 2.5–5.25 km in the dry and rainy seasons, respectively. Thus required emergency protective actions must be undertaken by the local authority to lessen the health impact on the general public.