scholarly journals SMALL MODULAR PWR DESIGN FOR TRU RECYCLING WITH McCARD-MASTER TWO-STEP PROCEDURE

2021 ◽  
Vol 247 ◽  
pp. 01003
Author(s):  
Dae Hee Hwang ◽  
Ser Gi Hong
Keyword(s):  
Mass Flow ◽  
Cross Sections ◽  
Treatment Method ◽  
Monte Carlo Code ◽  
Core Analysis ◽  
Code System ◽  
Lattice Calculation ◽  
Group Constant ◽  
Step Procedure ◽  
Neutronic Characteristics

In our previous study, a small modular PWR core was designed for TRU (Transuranics) recycling with multi-recycling scheme with a typical two-step procedure using DeCART2D/MASTER code system in which the lattice analysis for producing homogenized group constant was performed by DeCART2D while whole core analysis was conducted by MASTER code. However, the neutron spectrum hardening of the LWR core loaded with TRU requires validating the multi-group cross section library and resonance self-shielding treatment method in lattice calculation. In this study, a new procedure using McCARD/MASTER was used to analyze the SMR core, in which the lattice calculation was performed by a Monte Carlo code called McCARD with a continuous energy library to generate homogenized two-group assembly cross sections. The SMR core analysis was performed to show neutronic characteristics and TRU mass flow in the SMR core with TRU multi-recycling. The result shows that the analyses on the neutronic characteristics and TRU mass flow using the McCARD/MASTER code system showed good agreement with the previous ones using the DeCART2D/MASTER code system. The neutronic characteristics of each cycle of the core satisfied the typical limit of a commercial PWR core and the SMR core consumes effectively TRU with net TRU consumption rates of 8.46~14.33 %.

Development of REFT Code System for the Analysis of Highly Heterogeneous Research Reactors

2014 ◽  
Author(s):  
Tengfei Zhang ◽  
Hongchun Wu ◽  
Youqi Zheng ◽  
Liangzhi Cao ◽  
Yunzhao Li
Keyword(s):  
Least Square Method ◽  
Least Square ◽  
Core Analysis ◽  
Code System ◽  
Lattice Calculation ◽  
Control Rod ◽  
Research Reactors ◽  
The Core ◽  
User Friendly ◽  
Control Rods

As an effort to enhance the accuracy in simulating the operations of research reactors, a fuel management code system REFT was developed. Because of the possible complex assembly geometry and the core configuration of research reactors, the code system employed HELIOS in the lattice calculation to describe arbitrary 2D geometry, and used the 3D triangular nodal SN method transport solver, DNTR, to model unstructured geometry in the core analysis. Flux reconstruction with the least square method and micro depletion model for specific isotopes were incorporated in the code. At the same time, to make it more user friendly, a graphical user interface was also developed for REFT. In the analysis of the research reactors, the calculations involving the control rod movement are encountered frequently. The modeling of the control rods differential worth behavior is important in that the movement of the control rod may introduce variations on the reactivity. To handle the problem two effective ways of alleviating the control rod cusping effect are recently proposed, based on the established code system. The methodologies along with their application and validation will be discussed.

scholarly journals Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

2015 ◽  
Vol 2015 ◽  
pp. 1-11 ◽  
Author(s):  
Wonkyeong Kim ◽  
Jinsu Park ◽  
Tomasz Kozlowski ◽  
Hyun Chul Lee ◽  
Deokjung Lee
Keyword(s):  
Comparative Analysis ◽  
Cross Sections ◽  
Power Distribution ◽  
Reactor Core ◽  
Computer Code ◽  
Anisotropic Scattering ◽  
Monte Carlo Code ◽  
Core Analysis ◽  
Modeling Approach ◽  
Significant Difference

A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

The Preparation of Submicron Precision Cross Sections by Dimpling With A ‘Flatting Tool’

1997 ◽  
Vol 480 ◽  
Author(s):  
Helen L. Humiston
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Microstructural Analysis ◽  
Complex Materials ◽  
Ion Milling ◽  
Step Procedure ◽  
Simple Step ◽  
Primary Advantage ◽  
Preparation Techniques ◽  
Support Grid

AbstractThe complex materials systems in VLSI devices require specialized preparation techniques for TEM microstructural analysis. For this purpose, it is desirable to obtain electron transparency in all material layers from the oxides used in dielectrics to refractory metals such as tungsten. The primary advantage of dimpling these materials is that ideal specimens are obtained for low angle ion milling. By dimpling both sides of the cross section with a padded flatting tool, a thicker specimen of 130μm at the outer rim of the 3mm disc is produced that narrows to the 125nm thickness fringes in the center. These samples do not require a copper support grid, thereby allowing for a lower milling angle of 2.5 degrees on both sides of the specimen. This technique provides a cross section that is electron transparent in all layers without the loss of oxides due to differential thinning rates of various materials at higher milling angles.It is generally thought that precision thinning through a submicron feature is not possible on the dimpler. However, a simple step-by-step procedure for this technique will be demonstrated and discussed.

Validating the Serpent-Ants Calculation Chain Using BEAVRS Fresh Core HZP Data

2021 ◽  
Author(s):  
Ville Valtavirta ◽  
Antti Rintala ◽  
Unna Lauranto
Keyword(s):  
Monte Carlo ◽  
Best Practices ◽  
Power Distribution ◽  
Measured Data ◽  
Monte Carlo Code ◽  
Generation Process ◽  
Isothermal Temperature ◽  
Control Rod ◽  
Temperature Coefficients ◽  
Group Constant

Abstract The Serpent Monte Carlo code and the Serpent-Ants two step calculation chain are used to model the hot zero power physics tests described in the BEAVRS benchmark. The predicted critical boron concentrations, control rod group worths and isothermal temperature coefficients are compared between Serpent and Serpent-Ants as well as against the experimental measurements. Furthermore, radial power distributions in the unrodded and rodded core configurations are compared between Serpent and Serpent-Ants. In addition to providing results using a best practices calculation chain, the effects of several simplifications or omissions in the group constant generation process on the results are estimated. Both the direct and two-step neutronics solutions provide results close to the measured values. Comparison between the measured data and the direct Serpent Monte Carlo solution yields RMS differences of 12.1 mg/kg, 25.1 × 10-5 and 0.67 × 10-5 K-1 for boron, control rod worths and temperature coefficients respectively. The two-step Serpent-Ants solution reaches a similar level of accuracy with RMS differences of 17.4 mg/kg, 23.6 × 10-5 and 0.29 × 10-5 K-1. The match in the radial power distribution between Serpent and Serpent-Ants was very good with the RMS and maximum for pin power errors being 1.31 % and 4.99 % respectively in the unrodded core and 1.67 %(RMS) and 8.39 % (MAX) in the rodded core.

Characterization of Radioactive Waste From Side Structural Components of a CANDU Reactor for Decommissioning Applications in Korea

2010 ◽  
Author(s):  
Rizwan Ahmed ◽  
Gyunyoung Heo ◽  
Dong-Keun Cho ◽  
Jongwon Choi
Keyword(s):  
Radioactive Waste ◽  
Cross Sections ◽  
Reactor Core ◽  
International Standards ◽  
Monte Carlo Code ◽  
Capture Cross ◽  
Structural Components ◽  
The Core ◽  
Waste Classification ◽  
Capture Cross Sections

Reactor core components and structural materials of nuclear power plants to be decommissioned have been irradiated by neutrons of various intensities and spectrum. This long term irradiation results in the production of large number of radioactive isotopes that serve as a source of radioactivity for thousands of years for future. Decommissioning of a nuclear reactor is a costly program comprising of dismantling, demolishing of structures and waste classification for disposal applications. The estimate of radio-nuclides and radiation levels forms the essential part of the whole decommissioning program. It can help establishing guidelines for the waste classification, dismantling and demolishing activities. ORIGEN2 code has long been in use for computing radionuclide concentrations in reactor cores and near core materials for various burn-up-decay cycles, using one-group collapsed cross sections. Since ORIGEN2 assumes a constant flux and nuclide capture cross-sections in all regions of the core, uncertainty in its results could increase as region of interest goes away from the core. This uncertainty can be removed by using a Monte Carlo Code, like MCNP, for the correct calculations of flux and capture cross-sections inside the reactor core and in far core regions. MCNP has greater capability to model the reactor problems in much realistic way that is to incorporate geometrical, compositional and spectrum information. In this paper the classification of radioactive waste from the side structural components of a CANDU reactor is presented. MCNP model of full core was established because of asymmetric structure of the reactor. Side structural components of total length 240 cm and radius 16.122 cm were modeled as twelve (12) homogenized cells of 20 cm length each along the axial direction. The neutron flux and one-group collapsed cross-sections were calculated by MCNP simulation for each cell, and then those results were applied to ORIGEN2 simulation to estimate nuclide inventory in the wastes. After retrieving the radiation level of side structural components of in- and ex-core, the radioactive wastes were classified according to the international standards of waste classification. The wastes from first and second cell of the side structural components were found to exhibit characteristics of class C and Class B wastes respectively. However, the rest of the waste was found to have activity levels as that of Class A radio-active waste. The waste is therefore suitable for land disposal in accordance with the international standards of waste classification and disposal.

scholarly journals Modeling of the ORNL PCA Benchmark Using SCALE6.0 Hybrid Deterministic-Stochastic Methodology

2013 ◽  
Vol 2013 ◽  
pp. 1-9 ◽  
Author(s):  
Mario Matijević ◽  
Dubravko Pevec ◽  
Krešimir Trontl
Keyword(s):  
Cross Sections ◽  
Neutron Transport ◽  
National Laboratory ◽  
General Purpose ◽  
Monte Carlo Code ◽  
Oak Ridge ◽  
Benchmark Data ◽  
Selection Of ◽  
Good Agreement ◽  
Reactor Dosimetry

Revised guidelines with the support of computational benchmarks are needed for the regulation of the allowed neutron irradiation to reactor structures during power plant lifetime. Currently, US NRC Regulatory Guide 1.190 is the effective guideline for reactor dosimetry calculations. A well known international shielding database SINBAD contains large selection of models for benchmarking neutron transport methods. In this paper a PCA benchmark has been chosen from SINBAD for qualification of our methodology for pressure vessel neutron fluence calculations, as required by the Regulatory Guide 1.190. The SCALE6.0 code package, developed at Oak Ridge National Laboratory, was used for modeling of the PCA benchmark. The CSAS6 criticality sequence of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, as well as MAVRIC/Monaco hybrid shielding sequence, was utilized for calculation of equivalent fission fluxes. The shielding analysis was performed using multigroup shielding library v7_200n47g derived from general purpose ENDF/B-VII.0 library. As a source of response functions for reaction rate calculations with MAVRIC we used international reactor dosimetry libraries (IRDF-2002 and IRDF-90.v2) and appropriate cross-sections from transport library v7_200n47g. The comparison of calculational results and benchmark data showed a good agreement of the calculated and measured equivalent fission fluxes.

The Analysis of Influence of Fuel Pellet and Coolant Temperature Distributions on the VVER-1000 Active Core Characteristics

2013 ◽  
Author(s):  
A. A. Mishin ◽  
V. V. Galchenko
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Fuel Pellet ◽  
Coolant Temperature ◽  
Radial Temperature ◽  
Group Constant ◽  
Temperature Distributions ◽  
Active Core ◽  
Core Characteristics

The accuracy and quality of neutron-physical calculations of the active core characteristics depend heavily on the few-group constant preparation procedure. The method, based on using average in the fuel assembly fuel and coolant parameters is currently used for preparing macroscopic cross-sections. The question is what impact would considering the uneven distribution of those parameters, made on the few-group constant preparation stage exert on further analysis of the reactor facility behavior during steady-state and transients operation. The study carries out comparative analysis of the neutron-physical characteristics of the VVER-1000 core using the standard approach and using distributed in the fuel assembly fuel and coolant parameters while preparing few-group constants. It’s revealed that the fuel pellet and coolant radial temperature distributions affect the multiplication factor and temperature reactivity effect values.

Methods for the Gas Dynamic Calculation of the Cooled Multistage Turbine Performance

2014 ◽  
Author(s):  
B. I. Mamaev ◽  
M. M. Petukhovsky ◽  
A. V. Pozdnyakov ◽  
E. K. Ryabov
Keyword(s):  
Mass Flow ◽  
Cross Sections ◽  
Rotation Velocity ◽  
Test Results ◽  
Good Correspondence ◽  
Airfoil Surface ◽  
Flow Parameters ◽  
Operation Modes ◽  
Cooling Air ◽  
The Impact

Method of 1D calculation was developed. It is applicable to various ideal gases and their mixtures with regard to variability of their physical properties within wide temperature range. Four checking cross-sections were set for each vane/blade row: at the row inlet, in the row throat, at the row outlet, and downstream the row at the distance of the inter-row gap. Relations for the calculation of the flow outlet angle and maximum outlet velocity when the velocity in the throat reaches sonic values were a basis for a mathematical model. These relations considered the row geometry, tapering of a meridional streamline, a compressibility of gas, and losses variable along a channel. They made it possible to determine a gas mass flow at various flow outlet velocities. The row loss coefficient was determined from numerous test results. Losses in uncooled row were composed of a friction, edge, wave (due to incidence), and tip losses. A cascade density, the row height, the edge thickness, the airfoil surface roughness, inlet and outlet angles, Re, and the flow turbulence were also considered. Semi-empirical relations, describing an impact of a cooling air blow-off through a perforation in the airfoil and tip shrouds, trailing edge slot, at the shroud mating area, and in the gap downstream the row, were applied for the cooled turbine calculation. The influence of the blade cooling air pre-swirl on its temperature and turbine power was taken into account as well. Axial flows and the impact of a circumferential flow were defined when the radial gap flow for shroudless blades was calculated. For shrouded blades flows through linear and step labyrinths calculated. Influence of the rotation direction on the flow and gas heating in the labyrinth were also taken into account. Disk-gas losses are also calculated. Inputs were geometry of each row, a rotation velocity, a pressure, a temperature, a flow inlet angle, cooling air mass flow, temperature and pressure. Outputs could be any operation parameter of a turbine or a stage and flow parameters at checking cross-sections. Calculation covered whole range of turbine operation modes, including stage operation modes with energy consumption. A high-performance code was written on a basis of this method. It was verified against test results of some model turbines. A comparison showed quite good correspondence to tests both for turbine integral parameters and flow parameters at the checking cross-sections.

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