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[en] Highlights: • A containment design is proposed for the I2S-LWR, including passive safety systems. • Assessment of the safety systems demonstrates indefinite passive cooling. • The impact of the ADS valves and accumulator on plant response is investigated. - Abstract: The Integral Inherently Safe Light Water Reactor (I2S-LWR) is a novel reactor design concept which aims at delivering an electric power output level comparable to that of large LWRs (approximately 1000 MWe), while at the same time achieving an overall level of safety that is enhanced with respect to large Generation III+ LWRs. One of the main safety goals is to achieve indefinite cooling following design basis accidents (DBAs) using atmosphere air as ultimate heat sink. In order to accommodate these goals, the I2S-LWR incorporates several innovative safety features, including an integral Reactor Pressure Vessel (RPV), enhanced passive Decay Heat Removal (DHR) systems and several containment passive cooling systems. The present work is focused on a passive and reliable containment design, which plays a significant role in LOCA scenario and is the last boundary to prevent the release of radioactivity to the environment. In this paper, several innovative passive systems located in the I2S-LWR containment are proposed, including Core Make-up Tank (CMT), Accumulator (ACC), Passive Suppression System (PSS), Automatic Depressurization System (ADS), Passive Containment Cooling System (PCCS) and Passive Reactor Cavity Cooling System (PRCCS). The best-estimate thermal-hydraulic code RELAP5 has been used to model the I2S-LWR RPV and containment passive systems. The inadvertent opening of PORV (Power Operated Relief Valve) accident scenario has been simulated in order to study the containment response, including the coupling between RPV and containment. The results show that, through the activation of the ADS, pressure equilibrium between containment and reactor pressure vessel is achieved, while maintaining the reactor core covered at all times. The containment passive cooling systems ensure that the containment pressure remains at acceptable levels throughout the transient. A sensitivity study on the number of operating ADS valves, and on the availability of accumulators is also performed.
[en] Highlights: ► A MHD stability analysis on an electromagnetic pump was carried out. ► Small perturbations for MHD fields were considered in sinusoidal form. ► Critical Reynolds number depends on the Hartman number and perturbed wave number. ► A magnetic field has a significant stabilizing effect on liquid sodium flow. - Abstract: A stability analysis of a viscous, incompressible, and electrically conducting liquid sodium flow in an annular linear induction electromagnetic pump for sodium coolant circulation of a Sodium Fast Reactor (SFR) was carried out when transverse magnetic fields permeate the sodium fluid across the narrow annular gap. Due to a negligible skin effect and the presence of a magnetic core outside the gap, radial magnetic field is assumed to be constant over the narrow channel gap, and the steady state solution of the axial velocity is obtained as a function of radius. Small perturbations for MHD fields were considered in sinusoidal form as a function of the angular frequency and wave number, and the resulting equations were linearized. The solutions of the perturbed equations were sought in the form of a linear combination of independent orthogonal functions in a non-dimensional radial interval (0, 1), and each orthogonal function was chosen to satisfy the boundary conditions of adhesion to the solid walls of the channel. Under the assumption that solutions of the equations were not oscillated rapidly according to the radial coordinate, finite numbers of orthogonal polynomials were considered. As a result, simultaneous equations with coefficients of steady-state solutions were arranged, and dispersion relations between angular frequency and wave number of perturbed state were sought. The imaginary part of the angular frequency was taken into consideration from the condition of existence of a nontrivial solution of the system, which leads to the relation between critical Reynolds number (Recr) and Hartmann number (Ha). In the present study, critical Reynolds number and wave numbers were plotted against the Hartmann number for a long wave perturbation; thus, it was shown that a magnetic field has a significant stabilizing effect on liquid sodium flow.
[en] Highlights: ► Numerical fluctuation calculations for a commercial PWR. ► Fluctuations in neutron flux, fuel and moderator temperature and coolant velocity. ► Coupled models for neutronics and thermal–hydraulics. ► Validated against FLUENT and RELAP5/PARCS. - Abstract: This paper describes a tool for estimating fluctuations in neutron flux, fuel temperature, moderator density and flow velocity in Pressurized Water Reactors by coupling a dynamic thermal–hydraulic module and a dynamic neutron kinetic module. The code calculates the static solution first, giving the profile of the static fuel temperature, moderator density, velocity and neutron flux. The fluctuations (called noise in this work) are the differences between the actual time-dependent values and the corresponding mean values. The fluctuations are in general induced by perturbations in the thermal–hydraulic parameters, e.g. moderator temperature or density, at the inlet of the core. There is also a possibility to directly define the perturbations in the macroscopic cross-sections and to supply them to the neutronic part of the model. The model was validated against two separate calculations using two different commercial tools.
[en] Highlights: ► An automatic core reload design tool was developed for a pressurized water reactor. ► Three different algorithms, i.e., the rank-based ant system, max–min ant system, and Ant-Q are adopted. ► Safety requirements are formulated as penalty terms of the quality function. ► Firstly, fuel assemblies are permutated to some degree and then fuel assemblies are rotated. - Abstract: An automatic core reload design tool was developed for a pressurized water reactor (PWR). A loading pattern (LP) was searched for using three different algorithms: the rank-based ant system (RAS), max–min ant system (MMAS), and Ant-Q which are variants of the ant colony algorithm. The fuel assemblies (FAs) were permuted in a one eighth core position and then the LP was copied to the other one eighth core with mirror symmetry, to form a quarter core LP, which was extended to a full core LP with rotational symmetry. Heuristic information was implemented to reduce search space and thus computation time. Safety requirements, such as the hot channel factor FΔH and moderator temperature coefficient (MTC), which must be satisfied, were formulated as penalty terms of the quality function. The search procedure contained two steps. The first step was to place the FA so that FΔH and MTC might be slightly violated, and the second step was to rotate the FA, which would improve the FΔH and MTC and the fuel cycle length. When the LP was designed, the SIMULATE-3 code calculated the FΔH, MTC, and cycle length, which were used to update the pheromone. The results demonstrated that the developed tool can obtain a LP which possesses the desired cycle length and also satisfies safety requirements.
[en] Highlights: ► It describes the Korean radiation dose assessment systems. ► The characteristic of the developed systems are introduced. ► The validation of the systems are also described. - Abstract: This paper describes radiological dose assessment systems which have been developed to support a decision making process against a nuclear emergency. Local scale system FADAS has been used in a national radiological emergency preparedness system. For a validation of FADAS, field tracer experiments were conducted over nuclear sites. Long range model named LADAS has been developed to estimate the effects of radiological emergencies in neighboring countries. METRO-K has been developed for a simulation of the radiological effects due to radioactive materials which have been deposited on various surfaces of an urban area. These systems are useful to provide the radiological information which is essential to choose the suitable countermeasures in case of a nuclear emergency.
[en] Highlights: ► A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. ► The code is applicable to the extent of subcooled boiling of coolant. ► Safety analysis of IAEA benchmark reactor core is carried out. ► Results corroborate well with the results available in the literature. - Abstract: A point reactor kinetics code, coupled with thermal hydraulics of plate type fuel, is developed in order to carry out the reactivity initiated transient analysis of a reactor. Point kinetics equations are solved by matrix algebra, which involves piecewise constant approximation (PCA) i.e. the time dependent functions-reactivity and neutron source terms are assumed to be constant within an interval of time. This code is able to simulate various kinds of reactivity insertion like step, ramp, sinusoidal etc. for transient analysis. Effect of reactivity feedback is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where thermal energy conservation equations are solved by finite difference method along with Crank–Nicolson technique to find out the fuel, clad and coolant temperatures during transients. Results of the code corroborate well with the benchmark results.
[en] Highlights: ► Up to now the spectral parameters of thermal neutrons are measured with activation foils that are not always reliable in low flux systems. ► We applied a solid state nuclear track detector to measure the absolute neutron flux in the light water sub-critical reactor (LWSCR). ► Experiments concerning fission track detecting were performed and were investigated using the Monte Carlo code MCNP. ► The neutron fluxes obtained in experiment are in fairly good agreement with the results obtained by MCNP. - Abstract: In the present paper, a solid state nuclear track detector is applied to measure the absolute neutron flux in the light water sub-critical reactor (LWSCR) in Nuclear Science and Technology Research Institute (NSTRI). Up to now, the spectral parameters of thermal neutrons have been measured with activation foils that are not always reliable in low flux systems. The method investigated here is the irradiation method. Experiments concerning fission track detecting were performed. The experiment including neutron flux calculation method has also been investigated using the Monte Carlo code MCNP. The analysis shows that the values of neutron flux obtained by experiment are in fairly good agreement with the results obtained by MCNP. Thus, this method may be able to predict the absolute value of neutron flux at LWSCR and other similar reactors.
[en] Highlights: ► Evaluate performance of SCALE depletion capabilities and ENDF/B-VII libraries. ► Validate using measured isotopic assay data for ∼100 spent fuel samples. ► Compare results obtained with ENDF/B-V and ENDF/B-VII cross-sections. ► Fission product prediction with ENDF/B-VII yields remarkable improvements. ► SCALE results for isotopic compositions are comparable to that of other codes. - Abstract: New isotopic depletion capabilities and ENDF/B-VII data libraries have been implemented in the recent release 6.1 of SCALE, a comprehensive modeling and simulation suite for nuclear safety analysis and design developed and maintained by Oak Ridge National Laboratory. An assessment of the effect of the new developments on the code performance is the subject of this paper. The analysis is focused on evaluating the code performance in predicting isotopic compositions in spent nuclear fuel by using an extensive, measured isotopic assay database. The analysis results obtained using the latest ENDF/B-VII cross-section data and different resonance processing methods in SCALE are compared to the results of previous validation studies that used ENDF/B-V data. The performance of SCALE depletion capabilities with respect to other computational systems is assessed based on recent published results that were obtained using ENDF/B-VII libraries.
[en] Highlights: ► The performance of GA, HNN and combination of them in BPP optimization in PWR core are adequate. ► It seems HNN + GA arrives to better final parameter value in comparison with the two other methods. ► The computation time for HNN + GA is higher than GA and HNN. Thus a trade-off is necessary. - Abstract: In the last decades genetic algorithm (GA) and Hopfield Neural Network (HNN) have attracted considerable attention for the solution of optimization problems. In this paper, a hybrid optimization method based on the combination of the GA and HNN is introduced and applied to the burnable poison placement (BPP) problem to increase the quality of the results. BPP in a nuclear reactor core is a combinatorial and complicated problem. Arrangement and the worth of the burnable poisons (BPs) has an impressive effect on the main control parameters of a nuclear reactor. Improper design and arrangement of the BPs can be dangerous with respect to the nuclear reactor safety. In this paper, increasing BP worth along with minimizing the radial power peaking are considered as objective functions. Three optimization algorithms, genetic algorithm, Hopfield neural network optimization and a hybrid optimization method, are applied to the BPP problem and their efficiencies are compared. The hybrid optimization method gives better result in finding a better BP arrangement.
[en] Highlights: ► Water is used as the coolant in the PBWR. ► The “near-miss” model is applied. ► The minimum temperature and maximum velocity are located at the neutral position of the four narrow gap regions. ► The high-temperature region and pressure drop of the coolant decreased with the increase of diameter. - Abstract: The thermal hydraulic characteristics of the pebble bed water cooled reactor (PBWR) have been studied by computational fluid dynamics (CFD). The “near-miss” model is applied considering the problem on the mesh quality. The standard κ–ε model is used as the turbulence model. The velocity and temperature fields of the coolant are obtained and analyzed in different fuel diameters (such as 3 and 6 mm). The effects of the diameters on the thermal hydraulic characteristics are also discussed. The distributions of the temperature, velocity, pressure and Nusselt number (Nu) of the coolant on the surface of the pellet are obtained and analyzed. This study can conduct the experimental and mechanism research of PBWR.