Published by De Gruyter

Volume 82 Issue 2

April 2017

Issue of Kerntechnik

Contents
Journal Overview

Publicly Available February 28, 2022

Frontmatter

Page range: 141-143

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February 28, 2022

Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

T. Kaliatka, A. Kaliatka, E. Uspuras, M. Vaisnoras, H. Mochizuki, W. F. G. van Rooijen

Page range: 148-160

Abstract

Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 × 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

February 28, 2022

Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

E. Zarifi, G. Jahanfarnia, K. Sepanloo

Page range: 161-175

Abstract

This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al 2 O 3 nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

February 28, 2022

PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

K. Hadad, M. Esmaeili-Sanjavanmareh

Page range: 176-183

Abstract

PCTRAN capability to simulate a large break loss of coolant accident concurrent with the loss of offsite power in Bushehr Nuclear Power Plant is enhanced and investigated. Following the correction of the accident scenario for Bushehr nuclear power plant in PCTRAN, simulation results are compared with the final safety assessment report of that plant. As a result, the primary loop thermal hydraulics parameters including pressure, total flow rates, leakage flow rates and reactor power are in a good agreement with the reference data. Hot and cold leg temperature variations have the same trends as reference data but have a maximum of 80 8C disagreement at the transient initiation. The reason for this disagreement is explained and its adjustment is discussed. Improvements of PCTRAN simulator are mainly due to enhancing user control for atmospheric steam dump valve, containment pressure and emergency core cooling systems which are thoroughly described in this paper.

February 28, 2022

Experimental study of natural circulation flow instability in rectangular channels

T. Zhou, S. Qi, M. Song, Z. Xiao

Page range: 184-189

Abstract

Experiments of natural circulation flow instability were conducted in rectangular channels with 5 mm and 10 mm wide gaps. Results for different heating powers were obtained. The results showed that the flow will tend to be instable with the growing of heating power. The oscillation period of pressure D-value and volume flow are the same, but their phase positions are opposite. They both can be described by trigonometric functions. The existence of edge position and secondary flow will strengthen the disturbance of fluid flow in rectangle channels, which contributes to heat transfer. The disturbance of bubble and fluid will be strengthened, especially in the saturated boiling section, which make it possible for the mixing flow. The results also showed that the resistance in 5 mm channel is bigger than that in 10 mm channel, it is less likely to form stable natural circulation in the subcooled region.

February 28, 2022

Exerimental method and preliminary studies of the passive containment water film evaporation mass transfer

C. Li, L. Yang, W. Zhao, S. Zhou, W. Du, Z. Gao, H. Li

Page range: 190-195

Abstract

For larger containments and higher operation parameters, characteristics of the outside cooling of the PCCS are very important for the analysis on the containment integrity. A preliminary analysis was made and a four-step experimental method was used to numerically analyze the falling water film evaporation for the advanced passive containment. Then, the water flow stability along the outside wall of the containment was studied. The results fit well with those correlations without airflow when the air velocity is less than 5.0 m/s. However, when the air velocity is larger than 5.0 m/s, the influence of the air velocity on the water film will appear and the mean water film thickness will be thicker. Based on the prototype operation parameters, experimental studies were carried and the results were compared with the Dittus-Boelter correlation within the operation ranges. A modification factor was proposed for the conservative application of this correlation for nuclear safety analysis.

February 28, 2022

Automated generation of burnup chain for reactor analysis applications

V.-P. Tran, H.-N. Tran, A. Yamamoto, T. Endo

Page range: 196-205

Abstract

This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO 2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.

February 28, 2022

Fission source sampling in coupled Monte Carlo simulations

B. Olsen, J. Dufek

Page range: 206-209

Abstract

We study fission source sampling methods suitable for the iterative way of solving coupled Monte Carlo neutronics problems. Specifically, we address the question as to how the initial Monte Carlo fission source should be optimally sampled at the beginning of each iteration step. We compare numerically two approaches of sampling the initial fission source; the tested techniques are derived from well-known methods for iterating the neutron flux in coupled simulations. The first technique samples the initial fission source using the source from the previous iteration step, while the other technique uses a combination of all previous steps for this purpose. We observe that the previous-step approach performs the best.

February 28, 2022

Hysteresis phenomenon in nuclear reactor dynamics

B. Pirayesh, A. Pazirandeh, M. Akbari

Page range: 210-216

Abstract

This paper applies a nonlinear analysis method to show that hysteresis phenomenon, due to the Saddle-node bifurcation, may occur in the nuclear reactor. This phenomenon may have significant effects on nuclear reactor dynamics and can even be the beginning of a nuclear reactor accident. A system of four dimensional nonlinear ordinary differential equations was considered to study the hysteresis phenomenon in a typical nuclear reactor. It should be noted that the reactivity was considered as a nonlinear function of state variables. The condition for emerging hysteresis was investigated using Routh-Hurwitz criterion and Sotomayor’s theorem for saddle node bifurcation. A numerical analysis is also provided to illustrate the analytical results.

February 28, 2022

Investigation of neutronic and safety parameters variation in 5 MW research reactor due to U₃O₈Al fuel conversion to ThO₂ + U₃O₈Al

Z. Gholamzadeh, S. A. H. Feghhi, Z. Alipoor, M. Vahedi, S. M. Mirvakili, H. Bagheri, C. Tenreiro

Page range: 217-224

Abstract

Thorium-based fuels could comprise several advantages and are being investigated as a potentially competitive option with uranium-based fuels for research or power reactors. The present study investigates computationally the application of two different thorium-based fuels in a research reactor. Void and temperature reactivity coefficients, safety factor, power peaking factor, neutron generation time, effective delayed neutron fraction and 135 Xe worth parameter were investigated for the fuel conversions. The results showed both the investigated fuels would not significantly disturb neutronic and safety parameters of the modeled core in comparison with its routine fuel loading. However, 235-enriched thorium based fuel concluded in noticeably reduction of High Level Waste (HLW) but 233-enriched type could be taken in attention because of its longer fuel cycle (̴15%) and integrated neutron flux (̴23%).

February 28, 2022

Implementation of meso-scale radioactive dispersion model for GPU

Sunarko, Z. Suud

Page range: 225-231

Abstract

Lagrangian Particle Dispersion Method (LPDM) is applied to model atmospheric dispersion of radioactive material in a meso-scale of a few tens of kilometers for site study purpose. Empirical relationships are used to determine the dispersion coefficient for various atmospheric stabilities. Diagnostic 3-D wind-field is solved based on data from one meteorological station using mass-conservation principle. Particles representing radioactive pollutant are dispersed in the wind-field as a point source. Time-integrated air concentration is calculated using kernel density estimator (KDE) in the lowest layer of the atmosphere. Parallel code is developed for GTX-660Ti GPU with a total of 1 344 scalar processors using CUDA. A test of 1-hour release discovers that linear speedup is achieved starting at 28 800 particles-per-hour (pph) up to about 20 × at 14 4000 pph. Another test simulating 6-hour release with 36 000 pph resulted in a speedup of about 60 ×. Statistical analysis reveals that resulting grid doses are nearly identical in both CPU and GPU versions of the code.

February 28, 2022

Solution of the multilayer multigroup neutron diffusion equation in cartesian geometry by fictitious borders power method

R. Zanette, C. Z. Petersen, M. Schramm, J. R. Zabadal

Page range: 232-238

Abstract

In this paper a solution for the one-dimensional steady state Multilayer Multigroup Neutron Diffusion Equation in cartesian geometry by Fictitious Borders Power Method and a perturbative analysis of this solution is presented. For each new iteration of the power method, the neutron flux is reconstructed by polynomial interpolation, so that it always remains in a standard form. However when the domain is long, an almost singular matrix arises in the interpolation process. To eliminate this singularity the domain segmented in R regions, called fictitious regions. The last step is to solve the neutron diffusion equation for each fictitious region in analytical form locally. The results are compared with results present in the literature. In order to analyze the sensitivity of the solution, a perturbation in the nuclear parameters is inserted to determine how a perturbation interferes in numerical results of the solution.

February 28, 2022

Half-space albedo problem with modified FN method for linear and quadratic anisotropic scattering

R. G. Türeci, D. Türeci

Page range: 239-245

Abstract

One speed, time independent and homogeneous medium neutron transport equation can be solved with the anisotropic scattering which includes both the linear anisotropic and the quadratic anisotropic scattering properties. Having solved Case’s eigenfunctions and the orthogonality relations among these eigenfunctions, some neutron transport problems such as albedo problem can be calculated as numerically by using numerical or semi-analytic methods. In this study the half-space albedo problem is investigated by using the modified F N method.

About this journal

Kerntechnik is an independent journal for nuclear engineering (including design, operation, safety and economics of nuclear power stations, research reactors and simulators), energy systems, radiation (ionizing radiation in industry, medicine and research) and radiological protection (biological effects of ionizing radiation, the system of protection for occupational, medical and public exposures, the assessment of doses, operational protection and safety programs, management of radioactive wastes, decommissioning and regulatory requirements). For more than 75 years Kerntechnik offers original scientific and technical contributions, review papers and conference reports.
All articles are subject to thorough, independent peer review.

Volume 82 Issue 2

Issue of Kerntechnik

Frontmatter

Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

Abstract

Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

Abstract

PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

Abstract

Experimental study of natural circulation flow instability in rectangular channels

Abstract

Exerimental method and preliminary studies of the passive containment water film evaporation mass transfer

Abstract

Automated generation of burnup chain for reactor analysis applications

Abstract

Fission source sampling in coupled Monte Carlo simulations

Abstract

Hysteresis phenomenon in nuclear reactor dynamics

Abstract

Investigation of neutronic and safety parameters variation in 5 MW research reactor due to U3O8Al fuel conversion to ThO2 + U3O8Al

Abstract

Implementation of meso-scale radioactive dispersion model for GPU

Abstract

Solution of the multilayer multigroup neutron diffusion equation in cartesian geometry by fictitious borders power method

Abstract

Half-space albedo problem with modified FN method for linear and quadratic anisotropic scattering

Abstract

Investigation of neutronic and safety parameters variation in 5 MW research reactor due to U₃O₈Al fuel conversion to ThO₂ + U₃O₈Al