• Nuclear Engineering and Technology
• /
• v.51 no.7
• /
• pp.1738-1748
• /
• 2019

Previous research has shown that the inclined condition has an impact on the natural circulation (natural circulation) mode operation of Floating Nuclear Power Plant (FNPP) mounted on the movable marine platform. Due to its compact structure, small volume, strong maneuverability, the Integral Pressurized Water Reactor (IPWR) is adopted as marine reactor in general. The OTSGs of IPWR are symmetrically arranged in the annular region between the reactor vessel and core support barrel in this paper. Therefore, many parallel natural circulation loops are built between the core and the OTSGs primary side when the main pump is stopped. and the inclined condition would lead to discrepancies of the natural circulation drive head among the OTSGs in different locations. In addition, the flow rate and temperature nonuniform distribution of the core caused by inclined condition are coupled with the thermal hydraulics parameters maldistribution caused by OTSG group operating mode on low power operation. By means of the RELAP5 codes were modified by adding module calculating the effect of inclined, heaving and rolling condition, the simulation model of IPWR in inclined condition was built. Using the models developed, the influences on natural circulation operation by inclined angle and OTSG position, the transitions between forced circulation (forced circulation) and natural circulation and the effect on natural circulation operation by different OTSG grouping situations in inclined condition were analyzed. It was observed that a larger inclined angle results the temperature of the core outlet is too high and the OTSG superheat steam is insufficient in natural circulation mode operation. In general, the inclined angle is smaller unless the hull is destroyed seriously or the platform overturn in the ocean. In consequence, the results indicated that the IPWR in the movable marine platform in natural circulation mode operation is safety. Selecting an appropriate average temperature setting value or operating the uplifted OTSG group individually is able to reduce the influence on natural circulation flow of IPWR by inclined condition.

• The Bulletin of The Korean Astronomical Society
• /
• v.46 no.2
• /
• pp.37.2-37.2
• /
• 2021

Turbulence produces the density and velocity fluctuations in molecular clouds, and dense regions within the density fluctuation are the birthplace of stars. Also, turbulence can produce non-thermal pressure against gravity. Thus, turbulence plays a crucial roles in controlling star formation. However, despite many years of study, the detailed relation between turbulence and star formation remain poorly understood. As part of the Taeduk Radio Astronomy Observatory (TRAO) Key Science Program (KSP), "mapping Turbulent properties In star-forming MolEcular clouds down to the Sonic scale (TIMES; PI: Jeong-Eun Lee)", we mapped two star-forming molecular clouds, the Orion A and the ρ Ophiuchus molecular clouds, in six molecular lines (¹³CO 1-0/C¹⁸O 1-0, HCN 1-0/HCO⁺ 1-0, and CS 2-1/N₂H⁺ 1-0) using the TRAO 14-m telescope. We applied the Principal Component Analysis (PCA) to the observed data in two different ways. The first method is analyzing the variation of line intensities in velocity space to evaluate the velocity power spectrum of underlying turbulence. We investigated the relation between the star formation activities and properties of turbulence. The other method is analyzing the variation of the integrated intensities between the molecular lines to find the characteristic correlation between them. We found that the HCN, HCO⁺, and CS lines well correlate with each other in the integral shaped filament in the Orion A cloud, while the HCO⁺ line is anti-correlate with the HCN and CS lines in L1688 of the Ophiuchus cloud.

• Smart Structures and Systems
• /
• v.9 no.3
• /
• pp.231-251
• /
• 2012

The stochastic optimal control for a piezoelectric spherically symmetric shell subjected to stochastic boundary perturbations is constructed, analyzed and evaluated. The stochastic optimal control problem on the boundary stress output reduction of the piezoelectric shell subjected to stochastic boundary displacement perturbations is presented. The electric potential integral as a function of displacement is obtained to convert the differential equations for the piezoelectric shell with electrical and mechanical coupling into the equation only for displacement. The displacement transformation is constructed to convert the stochastic boundary conditions into homogeneous ones, and the transformed displacement is expanded in space to convert further the partial differential equation for displacement into ordinary differential equations by using the Galerkin method. Then the stochastic optimal control problem of the piezoelectric shell in partial differential equations is transformed into that of the multi-degree-of-freedom system. The optimal control law for electric potential is determined according to the stochastic dynamical programming principle. The frequency-response function matrix, power spectral density matrix and correlation function matrix of the controlled system response are derived based on the theory of random vibration. The expressions of mean-square stress, displacement and electric potential of the controlled piezoelectric shell are finally obtained to evaluate the control effectiveness. Numerical results are given to illustrate the high relative reduction in the root-mean-square boundary stress of the piezoelectric shell subjected to stochastic boundary displacement perturbations by the optimal electric potential control.

• Nuclear Engineering and Technology
• /
• v.54 no.4
• /
• pp.1253-1260
• /
• 2022

This paper reports the control method for the core power of the China initiative Accelerator Driven System (CiADS) facility. In the CiADS facility, an intense external neutron source provided by a proton accelerator coupled to a spallation target is used to drive a sub-critical reactor. Without any control rod inside the sub-critical reactor, the core power is controlled by adjusting the proton beam intensity. In order to continuously change the beam intensity, an adjustable aperture is considered to be used at the Low Energy Beam Transport (LEBT) line of the accelerator. The aperture size is adjusted based on the Proportional Integral Derivative (PID) controllers, by comparing either the setting beam intensity or the setting core power with the measured value. To evaluate the proposed control method, a CiADS core model is built based on the point reactor kinetics model with six delayed neutron groups. The simulations based on the CiADS core model have indicated that the core power can be controlled stably by adjusting the aperture size. The response time in the adjustment of the core power depends mainly on the adjustment time of the beam intensity.

• Advances in concrete construction
• /
• v.2 no.1
• /
• pp.13-27
• /
• 2014

The principal objective of this study is to deepen the characterization studies already led on sand concretes in previous works. Indeed, it consists in studying the effect of the sand type on the main properties of sand concrete: fracture and mechanical properties. We particularly insist on the determination of the fracture characteristics of this material which apparently have not been studied. To carry out this study, four different types of sand have been used: dune sand (DS), river sand (RS), crushed sand (CS) and river-dune sand (RDS). These sands differ in mineralogical nature, grain shape, angularity, particle size, proportion of fine elements, etc. The obtained results show that the particle size distribution of sand has marked its influence in all the studied properties of sand concrete since the sand having the highest diameter and the best particle size distribution has given the best fracture and mechanical properties. The grain shape, the angularity and the nature of sand have also marked their influence: thanks to its angularity and its limestone nature, crushed sand yielded good results compared to river and dune sands which are characterized by rounded shape and siliceous nature. Finally, it should further be noted that the sand concrete presents values of fracture and mechanical properties slightly lower than those of ordinary concrete. Compared to mortar, although the mechanical strength is lower, the fracture parameters are almost comparable. In all cases, the sand grains are debonded from the paste cement during the fracture which means that the crack goes through the paste-aggregate interface.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1083-1095
• /
• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2023-2030
• /
• 2022

A novel pulse fitting analysis (PFA) method is presented for the acquisition of nuclear spectra. The charging process of the feedback capacitor in the resistive feedback charge-sensitive preamplifier is equivalent to the impulsive pulse, and its impulse response function (IRF) can be obtained by non-linear fitting of the falling edge of the nuclear pulse. The integral of the IRF excluding the baseline represents the energy deposition of the particles in the detector. In addition, since the non-linear fitting process in PFA method is difficult to achieve in the conventional architecture of spectroscopy system, a new multi-channel analyzer (MCA) based on Zynq SoC is proposed, which transmits all the data of nuclear pulses from the programmable logic (PL) to the processing system (PS) by high-speed AXI-Stream in order to implement PFA method with precision. The linearity of new MCA has been tested. The spectrum of ¹³⁷Cs was obtained using LaBr₃(Ce) scintillator detector, and was compared with commercial MCA by ORTEC. The results of tests indicate that the MCA based on PFA method has the same performance as the commercial MCA based on pulse height analysis (PHA) method and excellent linearity for γ-rays with different energies, which infers that PFA method is an effective and promising method for the acquisition of spectra. Furthermore, it provides a new solution for nuclear pulse processing algorithms involving regression and iterative processes.

• Nuclear Engineering and Technology
• /
• v.54 no.2
• /
• pp.523-531
• /
• 2022

In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U₂₃₅(n, f), $ U_{235}(n\bar{\nu})$ and H₁(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.

• Nuclear Engineering and Technology
• /
• v.53 no.6
• /
• pp.1769-1785
• /
• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.

• Nuclear Engineering and Technology
• /
• v.54 no.7
• /
• pp.2367-2375
• /
• 2022

In this work, we propose a new mechanistic model for the treatment of helium behaviour at the grain boundaries in oxide nuclear fuel. The model provides a rate-theory description of helium inter-granular behaviour, considering diffusion towards grain edges, trapping in lenticular bubbles, and thermal resolution. It is paired with a rate-theory description of helium intra-granular behaviour that includes diffusion towards grain boundaries, trapping in spherical bubbles, and thermal re-solution. The proposed model has been implemented in the meso-scale software designed for coupling with fuel performance codes SCIANTIX. It is validated against thermal desorption experiments performed on doped UO₂ samples annealed at different temperatures. The overall agreement of the new model with the experimental data is improved, both in terms of integral helium release and of the helium release rate. By considering the contribution of helium at the grain boundaries in the new model, it is possible to represent the kinetics of helium release rate at high temperature. Given the uncertainties involved in the initial conditions for the inter-granular part of the model and the uncertainties associated to some model parameters for which limited lower-length scale information is available, such as the helium diffusivity at the grain boundaries, the results are complemented by a dedicated uncertainty analysis. This assessment demonstrates that the initial conditions, chosen in a reasonable range, have limited impact on the results, and confirms that it is possible to achieve satisfying results using sound values for the uncertain physical parameters.