Accepted Manuscripts

Laurent Cantrel, Thierry Albiol, Loic Bosland, Juliette Colombani, Frédéric Cousin, Anne-Cécile Grégoire, Olivia Leroy, Sandrine Morin, Christian Mun, Ohnet Marie Noëlle, Sidi Souvi, Céline Monsanglant-Louvet, Florent Louis, Bruno Azambre and Christophe Volkringer
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038223
This paper deals with near past, ongoing and planned R&D works on fission products (FPs) behavior in Reactor Cooling System (RCS), containment building and in Filtered Containment Venting Systems (FCVS) for severe accident (SA) conditions. The FCVS topic, in link with the post-Fukushima accident management, is again on the agenda [1]. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in simulation softwares. After being initiated in the International Source Term Program (ISTP), researches on iodine transport through the RCS are still ongoing and for containment, the last advances are linked to the Source Term Evaluation and Mitigation (STEM) OECD/NEA project. The objective is to improve the evaluation of Source Term for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major FPs: iodine and ruthenium. A follow-up, called STEM2 has just started to reduce some remaining issues and be closer to reactor conditions. For FCVS works, the efficiencies for trapping iodine covering scrubbers and dry filters are examined to get a clear view of their abilities in SA conditions. Another part is focused on specific porous materials able to trap volatile iodine. Influence of zeolites materials parameters (nature of the counter-ions, structure, Si/Al ratio...) will be tested. New kind of porous materials constituted by Metal organic Frameworks will also be looked at because they can constitute a promising way of trapping.
TOPICS: Accidents, Containment, Porous materials, Simulation, Nuclear fission, Ions, Metals, Chemistry, Filters, Nuclear reactor cooling, Nuclear power stations, Ruthenium, Containment buildings, Uncertainty, Accident management, Fukushima nuclear disaster, Japan, 2011
Yuzhou Chen, Minfu Zhao, Keming BI, Bin Yang, Dongxu Zhang and Kaiwen Du
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038215
Critical heat flux experiment was performed in a tube of 8.2 mm in inner diameter with 2.4 m in heated length. The pressure covered the ranges from 8.6 to 21.0 MPa, mass flux 1157 to 3776 kg/m2s, inlet quality -2.79 to -0.08 (subcooling 19 to 337 oC) and local quality -0.97 to 0.53. For the pressure close to the near-critical point the critical heat flux decreased with the pressure increasing. For the subcooling larger than a certain value the critical heat flux was related to the local conditions, and the experimental results were in agreement with the previous correlation. But for low subcooling and saturated flow the critical condition was related to the total power. Based on the present experimental results an empirical correlation of the critical heat flux was presented.
TOPICS: Water, Critical heat flux, Pressure, Subcooling, Flow (Dynamics)
Attila Kiss, Andrey Churkin, D.S. Pilkhwal, Abhijeet Mohan Vaidya, Walter Ambrosini, Andrea Pucciarelli, Krishna Podila, Yanfei Rao, Laurence K.H. Leung, Yuzhou Chen, Mark Anderson and Meng Zhao
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038162
Two Computational Fluid Dynamic (CFD) benchmarks have been performed to assess the prediction accuracy of CFD codes for heat transfer in different geometries. The first benchmark focused on heat transfer to water in a tube (1st benchmark), while the second benchmark covered heat transfer to water in two different channel geometries (2nd benchmark) at supercritical pressures. In the first round with the experimental data unknown to the participants (i.e., blind calculations), CFD calculations were conducted with initial boundary conditions and simpler CFD models. After assessment against measurements, the calculations were repeated with the refined boundary conditions and material properties in the follow-up calculation phase. Overall, the CFD codes seem to be able to capture the general trend of heat transfer in the tube and the annular channel but further improvements are required in order to enhance the prediction accuracy. Finally, sensitivity analyses on the numerical mesh and the boundary conditions were performed. It was found that the prediction accuracy has not been improved with the introduction of finer meshes and the effect of mass flux on the result is the strongest among various investigated boundary conditions.
TOPICS: Computational fluid dynamics, Boundary-value problems, Heat transfer, Water, Materials properties, Sensitivity analysis
Robin McDougall, Scott B. Nokleby and Ed Waller
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038185
This paper presents a novel methodology for generating radiation intensity maps using a mobile robotic platform and an integrated radiation model. The radiation intensity mapping approach consists of two stages. First, radiation intensity samples are collected using a radiation sensor mounted on a mobile robotic platform, reducing the risk of exposure to humans from an unknown radiation field. Next, these samples, which need only to be taken from a subsection of the entire area being mapped, are then used to calibrate a radiation model of the area. This model is then used to predict the radiation intensity field throughout the rest of the area that could not be directly measured. The performance of the approach is evaluated through experiments. The results show that the developed system is effective at achieving the goal of generating radiation maps using sparse data.
TOPICS: Radiation (Physics), Robotics, Risk, Sensors
Muhsin Mohd Amin, Yu Duan and Shuisheng He
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038161
It is now well known that heat transfer to fluid at supercritical pressure shows complex behaviours. This is due to the strong variations of the thermal physical properties resulting from the changes of pressure and temperature. To improve the reliability and efficiency of the supercritical water-cooled reactors to be designed, the understanding of supercritical fluid flow in the fuel assemblies is very important. The study reported here reconsiders a simplified geometry, including a trapezoid channel enclosing an inner rod to simulate the triangular arrangement of a fuel assembly. Large eddy simulation (LES) with the WALE model is used to simulate the forced convection flow in the channel. Supercritical water at 25MPa is used as the working fluid. The Reynolds number based on the hydraulic diameter and the bulk velocity was 10540, while the heat flux from the inner rod wall was varied from 10kW/m^2 to 75kW/m^2. Due to the non-uniformity of the cross-section of the flow channel, large unsteady flow structures are observed. The characteristics of the flow structures and their effect on the local heat transfer are analysed using the instantaneous velocities, spectrum analysis and correlation analysis. The swinging flow structures in the wide gap are much weaker than those in the narrow gap. The behaviours of such large flow structures are influenced by the strong spatial and temporal variations of the properties. When the temperature distribution follows T_b
TOPICS: Pressure, Forced convection, Annulus, Water, Flow (Dynamics), Heat transfer, Fluids, Fuels, Manufacturing, Reynolds number, Reliability, Supercritical fluids, Emission spectroscopy, Spectroscopy, Temperature, Large eddy simulation, Heat flux, Geometry, Temperature distribution, Unsteady flow
Expert View  
Samuel Miranda
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038160
Nuclear safety criteria are based upon the concept that plant situations that are expected to have a high frequency of occurrence must not pose a danger to the public, and that plant situations that pose the greatest danger to the public must be limited to situations that have a very low expected frequency of occurrence. This concept is implemented by grouping postulated plant situations (or events) into categories that are defined according to their expected frequencies of occurrence. Events in each category must be shown to yield consequences that remain within the limits that are defined for that category. Events must not be allowed to develop into the more serious events that belong in other, higher-consequence categories. In other words, nuclear plant designs must not allow high-frequency, low-consequence events to degrade into high-frequency, high-consequence events. The development of this system of frequency-based categorization is discussed, followed by an evaluation of various methods that have been proposed and applied to show how minor events could be prevented from becoming major events.
TOPICS: Safety, Accidents, Nuclear power stations
Rafael Bocanegra, Valentino Di Marcello, Victor H. Sánchez-Espinoza and Gonzalo Jiménez
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038062
A VTT Fukushima Daiichi Unit 3 MELCOR model [1] was modified to simulate the Fukushima Daiichi Unit 2 accident. Five simulations were performed using different modeling approaches. The model 1F2 v1 includes only the basic modifications to reproduce the 1F2 accident. The model 1F2 v2 includes the same modifications used in 1F2 v1 plus the WW improvement. In the 1F2 v3 model, the RCIC logic was modified to avoid the use of tabular functions for the mass flow inlet and outlet. As a result of this analysis it is concluded that there is a strong dependency on parameters which still have many uncertainties, such as the RCIC two-phase flow operation, the alternative water injection, the suppression pool behavior, the rupture disk behavior and the containment failure modes which affect the final state of the reactor core.
TOPICS: Simulation, Accidents, Fukushima nuclear disaster, Japan, 2011, Flow (Dynamics), Failure mechanisms, Modeling, Two-phase flow, Underground injection, Disks, Rupture, Containment, Uncertainty
Bruno Gonfiotti and Sandro Paci
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038059
The estimation of Fission Products (FPs) release from the containment system of a nuclear plant to the external environment during a Severe Accident (SA) is a quite complex task. In the last 30-40 years several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments, and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state-of-the-art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus FPT-1 test employing the ASTEC (F) and MELCOR (USA) codes. The analysis focuses on the stand-alone containment aspects of the test, and three different modellisations of the containment vessel have been developed showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to correctly predict thermal-hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results, and presents different sensitivity analyses carried out on the aerosols and FPs behavior.
TOPICS: Containment, Aerosols, Accidents, Nuclear power stations, Sensitivity analysis, Thermal hydraulics, Containment systems, Containment vessels, Coolants, Nuclear fission
Thomas Schulenberg and Hongbo Li
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038061
While supercritical water is a perfect coolant with excellent heat transfer, a temporary decrease of the system pressure to sub-critical conditions, either during intended transients or by accident, can easily cause a boiling crisis with significantly higher cladding temperatures of the fuel assemblies. These conditions have been tested in an out-of-pile experiment with a bundle of 4 heated rods in the SWAMUP facility co-constructed by CGNPC and SJTU in China. Some of the transient tests have been simulated with a one-dimensional MATLAB code, assuming quasi-steady state flow conditions, but time dependent temperatures in the fuel rods. Heat transfer at supercritical and at near-critical conditions was modelled with a recent look-up table of Zahlan (2015), and sub-critical film boiling was modelled with the look-up table of Groenveld et al. (2003). Moreover, a conduction controlled rewetting process was included in the analyses, which is based on an analytical solution of Schulenberg and Raqué (2014). The method could well reproduce the boiling crisis during depressurization from supercritical to subcritical pressure, including rewetting of the hot zone within some minutes, but the peak temperature was somewhat under-predicted. Tests with a lower heat flux, which did not cause such phenomena, could be predicted as well. In another test with increasing pressure, however, a boiling crisis was also observed at a heat flux, which was significantly lower than the critical heat flux predicted by the CHF look-up table of Groeneveld et al. (2006).
TOPICS: Pressure, Fuels, Manufacturing, Transient heat transfer, Supercritical water reactors, Boiling, Temperature, Heat transfer, Transients (Dynamics), Heat flux, Critical heat flux, Fuel rods, Cladding systems (Building), Accidents, Water, China, Film boiling, Matlab, Rods, Flow (Dynamics), Heat conduction, Coolants
Changbing Tang, Shuo Xing, Hua Pang, Ping Chen and Yi Zhou
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038060
Because of the high temperature and high pressure characteristics of SCWR, the thermal hydraulic performance of SCWR is greatly different from PWR, which makes the current PWR fuel rod performance analysis codes are no longer applicable to SCWR. In this research, the irradiation swelling, irradiation densification, thermal expansion, thermal creep, irradiation creep and irradiation hardening of UO2 pellet were considered, the irradiation swelling, thermal expansion, thermal creep and plastic deformation of stainless steel cladding were considered, the gas conductance and radiant conductance of gap heat transfer were considered, the forced convective heat transfer on the outer surface of cladding was considered. Meanwhile, the irradiation effects and the thermal effects on the materials parameters such as thermal conductivity, specific heat and young's modulus were also considered in this research. With the help of ABAQUS software, the related user-defined subroutines were developed, and the irradiation effects and thermal effects of SCWR fuel were introduced into the numerical simulation, and then completed the analysis of SCWR fuel rods' performance under the steady power conditions. Some reference suggestions for the design and development of SCWR fuel were provided by the establishment of this numerical simulation method.
TOPICS: Computer simulation, Supercritical water reactors, Fuel rods, Irradiation (Radiation exposure), Creep, Thermal expansion, Fuels, Temperature effects, Cladding systems (Building), Electrical conductance, Pressurized water reactors, Young's modulus, Convection, Design, Computer software, Stainless steel, High temperature, Thermal conductivity, Specific heat, Deformation, Heat transfer, Hardening, High pressure (Physics)
Amr Abdelhady
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4038058
This paper studies the radiological consequences resulting from withdrawal of nuclear fuel element from a core of open pool type reactor during normal operation. The reactor is cooled by water in a vertical upward direction which acts on the fuel element by an upward force. In case of accidental failure of fuel element clamp, it will be released and ejected vertically from the core and assumed to reach the pool surface. A negative reactivity insertion in the core after fuel element withdrawal and the reactor will be shutdown. MCNP code was used in this study to calculate the radiation dose rate levels in the reactor hall and inside the control room during fuel element withdrawal.
TOPICS: Fuels, Control rooms, Clamps (Tools), Failure, Water, Nuclear fuels, Radiation (Physics)
Sami Penttila, Pekka Moilanen, Wade Karlsen and Aki Toivonen
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037897
The presented work in this paper consists of a test set-up study of a new pneumatic loading device based on double bellows system and with miniature autoclaves enabling applications at temperature and pressure up to 650°C and 35 MPa, respectively. It has been demonstrated that it is technically feasible to carry out well defined and controlled material testing in the supercritical water (SCW) environment using this testing system. By using this type of system it makes possible to investigate the intrinsic role of the applied stress on the deformation behavior of material in light water reactor (LWR) conditions and also in other harsh environments like SCW conditions. In addition to this the compactness and versatility of the set-up makes this system particularly attractive for deployment in a hot-cell for testing of irradiated materials.
TOPICS: Materials testing, Bellows (Equipment), Water, Light water reactors, Testing, Stress, Pressure, Deformation, Temperature
Zhou Yuan, Wang Yangle, Chen Jingtan, Xia Zhaoyang and Junfeng Wang
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037896
The Supercritical Carbon Dioxide (S-C ) Brayton gas turbine cycle has been studied as an effcient and cost effective option for advanced power systems. One major safety issue for any power cycle is a pipe break and the associated discharge of the working fluid and subsequent decrease in system pressure. In this paper, a S-C critical flow in the nozzle tube is analyzed numerically by FLUENT 15.0. The Redlich-Kwong real gas equation is selected to calculate carbon dioxide density and the standard K-epsilon turbulence model is selected. Experimental data (Guillaume et al., 2014) are used as a benchmark to examine the capability of the current approach. Compared with experimental data, the simulation results overestimate the critical mass flux; the error range is between 15% and 25%. The simulation results show that as L/D ratio increases, critical mass flow decreases. As stagnation temperature increases, critical mass flow decreases. The complex thermal hydraulic behavior in the nozzle tubes is analyzed. Three flow patterns in the nozzle tube during transient critical flow are obtained and discussed. From inlet to outlet of the tube, C may undergo the following phases in turn:1)supercritical phase;2)supercritical phase - gas phase;3)supercritical phase - gas phase - liquid phase. The simulation results are also helpful for further experimental and theoretical research.
TOPICS: Flow (Dynamics), Computer simulation, Nozzles, Supercritical carbon dioxide, Simulation results, Thermodynamic power cycles, Pipes, Carbon dioxide, Cycles, Errors, Transients (Dynamics), Gas turbines, Density, Pressure, Temperature, Fluids, Power systems (Machinery), Turbulence, Safety
Ahmad Moghrabi and David Novog
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037895
The Canadian Pressure-Tube Super Critical Water-cooled Reactor (PT-SCWR) is an advanced Generation IV (GEN-IV) reactor concept which is considered as an evolution of the conventional CANada Deuterium Uranium (CANDU) reactor that includes both pressure tubes and a low temperature and pressure heavy water moderator. The Canadian PT-SCWR fuel assembly utilizes a Plutonium and Thorium fuel mixture with SuperCritical light Water (SCW) coolant flowing through the High-Efficiency Re-entrance Channel (HERC). In this work, the impact of fuel depletion on the evolution of lattice physics phenomena were investigated starting from fresh fuel to burnup conditions (25 MW·d·kg-1 [HM]) through sensitivity and uncertainty analyses using the lattice physics modules in SCALE (Standardized Computer Analysis for Licensing Evaluation). Given the evolution of key phenomena such as void reactivity and fuel temperature coefficient of reactivity in traditional CANDU reactors with burnup, this study focuses on the impact of fission products, 233U breeding, and actinides on fuel performance. The work shows that the most significant change in fuel properties with burnup is the depletion of fission isotopes of Pu and the buildup of high-neutron cross section fission products, resulting in a decrease in cell k8 with burnup as expected. Other impacts such as the presence of Protactinium and Uranium-233 are also discussed.
TOPICS: Physics, Pressure, Fuels, Water, Nuclear fission, Supercritical water reactors, Uranium, Licensing, Uncertainty analysis, Neutrons, Isotopes, Temperature, Manufacturing, Coolants, Low temperature, Computers
Stephen A Hambric, Samir Ziada and Richard Morante
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037898
The United States Nuclear Regulatory Commission (USNRC) has approved several Extended Power Uprates (EPU) for Boiling Water Reactors (BWRs). In some of the BWRs, operating at the higher EPU power levels and flow rates led to fatigue damage of Steam Dryers, including the generation of loose parts. Since those failures occured all BWR owners proposing EPUs have been required by the USNRC to ensure that the steam dryers would not experience fatigue cracking. This paper provides an overview of BWR steam dryer design; the fatigue failures that occurred at the Quad Cities (QC) nuclear power plants and their root causes; a brief history of BWR EPUs; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluation methods (static and alternating stress).
TOPICS: Stress, Boiling water reactors, Steam, Power uprate, Dryers, Fatigue failure, Evaluation methods, Failure, Fatigue cracks, Fatigue damage, Nuclear power stations, Design, Flow (Dynamics), Quality control
Nima Fathi, Patrick McDaniel, Charles Forsberg and Cassiano de Oliveira
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037806
The intermittency of renewable power generation systems on the low carbon electric grid can be alleviated by using nuclear systems as quasi-storage systems. Nuclear Air-Brayton systems can produce and store hydrogen when electric generation is abundant and then burn the hydrogen by Co-Firing when generation is limited. The rated output of a nuclear plant can be significantly augmented by Co-Firing. The incremental hydrogen to electricity efficiency can far exceed that of hydrogen in a stand-alone gas turbine. Herein we simulate and evaluate this idea on a 50 MW small modular liquid metal/molten salt reactor. Considerable power increases are predicted for Nuclear Air-Brayton systems by Co-Firing with hydrogen before the power turbine.
TOPICS: Carbon, Energy storage, Thermodynamic power cycles, Hydrogen, Co-firing, Liquid metals, Nuclear power stations, Renewable energy, Storage, Molten salt reactors, Gas turbines, Turbines, Electric power generation
Metin Yetisir, Holly Hamilton, Rui Xu, Michel Gaudet, David Rhodes, Mitch King, Andrew Kittmer and Ben Benson
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037818
The Canadian SCWR is a 2540 MWt channel-type supercritical water-cooled nuclear reactor (SCWR) concept that employs 336 fuel channels in the reactor core. Each fuel channel includes a pressure tube that is submerged in a heavy water moderator and contains a removable fuel assembly. To achieve high (>45%) power conversion efficiency, the Canadian SCWR operates at a supercritical water pressure (25 MPa) and high temperatures (350 ºC at the inlet, 625 ºC at the outlet). These conditions lead to fuel cladding temperatures close to 800 ºC. Because of reduced material strength at this temperature and higher fission gas production of the fuel, collapsible fuel cladding is selected over internally-pressurized cladding. To increase heat transfer and to reduce cladding temperatures, turbulence-inducing wire-wraps are employed on fuel elements. Numerical models have been developed to analyze the thermal-structural behavior of Canadian SCWR fuel at normal and accident conditions. It was found that the maximum fuel cladding temperature for the worst-case accident scenario is below the melting point by a small margin. This paper presents a description of the Canadian SCWR fuel assembly concept, its unique features, the rationale used in the concept development and the results of various numerical analyses demonstrating the performance and characteristics of the Canadian SCWR fuel channel.
TOPICS: Manufacturing, Fuels, Water, Supercritical water reactors, Cladding systems (Building), Temperature, Accidents, Numerical analysis, Nuclear reactors, Melting point, High temperature, Turbulence, Computer simulation, Heat transfer, Nuclear fission, Water pressure, Wire, Strength (Materials), Energy conversion, Pressure
Laurence K.H. Leung and Armando Nava Dominguez
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037807
The thermal-hydraulics program in support of the development of the Canadian Super-Critical Water-cooled Reactor (SCWR) concept have undergone several phases. It focused on key parameters such as heat transfer, critical flow and stability of fluids at supercritical pressures. Heat-transfer experiments were performed with tubes, annuli and bundles in water, carbon dioxide or refrigerant flows. Data from these experiments have led to enhancement in understanding of the phenomena, improved prediction methods and verified analytical tools. In addition, these experiments facilitated the investigation of separate effects on heat transfer (such as geometry, diameter, spacing device, transient, etc.). Chocking flow characteristics were studied experimentally with sharp-edged nozzles of two different sizes of opening. Experimental data have been applied in improving the critical-flow correlation in support of accident analyses. A one-dimensional analytical model for instability phenomena has been developed and assessed against the latest experimental data for quantifying the prediction capability and applicability.
TOPICS: Thermal hydraulics, Supercritical water reactors, Flow (Dynamics), Heat transfer, Water, Stability, Fluids, Transients (Dynamics), Accidents, Nozzles, Annulus, Carbon dioxide, Geometry, Refrigerants
Krishna Podila and Yanfei Rao
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037747
Within the Generation-IV International Forum, Canadian Nuclear Laboratories (CNL) led the conceptual fuel bundle design effort for the Canadian supercritical water cooled reactor (SCWR). The proposed fuel rod assembly for the Canadian SCWR design comprised of 64-elements with spacing between elements maintained using the wire-wrap spacers. Experimental data and correlations are not available for the fuel-assembly concept of the Canadian SCWR. To analyze the thermalhydraulic performance of the new bundle design, CNL is using computational fluid dynamics (CFD) as well as the subchannel approach. Simulations of wire-wrapped bundles can benefit from the increased fidelity and resolution of a CFD approach due to its ability to resolve the boundary layer phenomena. Prior to the application, the CFD tool has been assessed against experimental heat transfer data obtained with bundle subassemblies to identify the appropriate turbulence model to use in the analyses. In the present paper, assessment of CFD predictions was made with the wire-wrapped bundle experiments performed at Xi'an Jiaotong University in China. A 3D CFD study of the fluid flow and heat transfer at supercritical pressures for the rod bundle geometries was performed with the key parameter being the fuel rod wall temperature. It was found that the CFD simulation tends to overpredict the fuel wall temperature, and the predicted location of peak temperature differs from the measurement by up to 65 degrees.
TOPICS: Pressure, Heat transfer, Wire, Simulation, Computational fluid dynamics, Water, Fuel rods, Supercritical water reactors, Design, Fuels, Manufacturing, Wall temperature, China, Turbulence, Fluid dynamics, Temperature, Resolution (Optics), Boundary layers
Tamas Janos Katona and Andras Vilimi
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4037715
Paks Nuclear Power Plant identified the post-Fukushima actions for mitigation and management of severe accidents caused by external events that include updating of some hazard assessments, evaluation of capacity / margins of existing severe accident management facilities, and construction of some new systems and facilities. In all cases, the basic question was, what level of margin has to be ensured above design basis external hazard effects, and what level of or hazard has to be taken for the design. Paks Nuclear Power Plant developed certain an applicable in the practice concept for the qualification of already implemented measures and design the new post-Fukushima measures that is outlined in the paper. The concept is based on the generalization of the procedure and assumptions used in the definition of acceptable margins for seismic loads, analysis of the steepness of the hazard curves and features of the hazards. Justification of the definition of exceedance probability of the design basis effects for the design of severe accident management systems is given on the basis of first order reliability theory. The application of the concept is presented on several practical examples.
TOPICS: Design, Nuclear power stations, Hazards, Accident management, Fukushima nuclear disaster, Japan, 2011, Hazard analysis, Construction, Stress, Accidents, Probability, Reliability theory

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