scholarly journals Cubic Phases in the Gd2O3-ZrO2 and Dy2O3-TiO2 Systems for Nuclear Industry Applications

2015 ◽  
Vol 2015 ◽  
pp. 1-7 ◽  
Author(s):  
Maria Teresa Malachevsky ◽  
Diego Rodríguez Salvador ◽  
Sergio Leiva ◽  
Claudio Alberto D'Ovidio
Keyword(s):  
Cross Section ◽  
Reactor Core ◽  
Neutron Cross Section ◽  
Nuclear Industry ◽  
Cubic Phase ◽  
Oxide Ceramic ◽  
Inert Matrix ◽  
Industry Applications ◽  
Preliminary Study ◽  
High Neutron

Neutron absorbers are elements with a high neutron capture cross section that are employed at nuclear reactors to control excess fuel reactivity. If these absorbers are converted into materials of relatively low absorption cross section as the result of neutron absorption, they consume during the reactor core life and so are called burnable. These elements can be distributed inside an oxide ceramic that is stable under irradiation and thus called inert. Cubic zirconium oxide is one of the preferred materials to be used as inert matrix. It is stable under irradiation, experiments very low swelling, and is isomorphic to uranium oxide. The cubic phase is stabilized by adding small amounts of dopants like Dy2O3 and Gd2O3. As both dysprosium and gadolinium have a high neutron cross section, they are good candidates to prepare burnable neutron absorbers. Pyrochlores, like Gd2Zr2O7 and Dy2Ti2O7, allow the solid solution of a large quantity of elements besides being stable under irradiation. These characteristics make them also useful for safe storage of nuclear wastes. We present a preliminary study of the thermal analysis of different compositions in the systems Gd2O3-ZrO2 and Dy2O3-TiO2, investigating the feasibility to obtain oxide ceramics useful for the nuclear industry.

scholarly journals Adaptation of HELIOS Models for WWER-440 Fuel Assemblies to Application by the TRACE/PARCS Program

2019 ◽  
pp. 46-51
Author(s):  
I. Ovdiienko ◽  
O. Kuchyn ◽  
M. Ieremenko ◽  
P. Vlasenko
Keyword(s):  
Fuel Assembly ◽  
Cross Section ◽  
Neutron Transport ◽  
Reactor Core ◽  
Neutron Cross Section ◽  
Safety Analysis ◽  
Neutron Transport Equation ◽  
Fuel Assemblies ◽  
Adaptation Procedure

The preparation of a few-group neutron cross-section library is an important step in implementation of the computer packages that are based on solution of the neutron transport equation in the few-group diffusion approximation into the safety analysis practices. The accuracy of modelling the physical neutron kinetic processes in the reactor core directly depends on the quality of few-group cross-section library. It is important to note that such cross-section library should be prepared in the format applied in the computer package and with use of a spectral code that models the fuel assembly quite adequately. The best option for preparing the few-group neutron crosssection library for the PARCS few-group diffusion code, which is being introduced into SSTC NRS safety analysis practices as a part of the TRACE/PARCS coupled neutron kinetic/thermal hydraulic package, is to adapt the previously developed and validated models of fuel assemblies for the HELIOS spectral program. The adaptation procedure for HELIOS models for WWER-440 including the fuel follower and transition part forming the input file structure required for correct work of the GenPMAXS program is presented. The approaches to the choice of reference states and branch parameters in the PARCS code format are presented. The results from correctness analysis of the adaptation of the HELIOS WWER-440 fuel assembly computer models are presented. The results are based on a comparative analysis of the fuel assembly multiplication properties obtained by the HELIOS model that was developed for preparation of the cross-section libraries for the DYN3D program (validated and widely used at SSTC NRS at present), and by the HELIOS model that was adapted for the GENPMAX program.

The Testing and Application of Continuous-Energy Neutron Cross Section Library HENDL-ADS/MC

2013 ◽  
Author(s):  
Chong Chen ◽  
Jun Zou ◽  
Dezheng Xu ◽  
Qin Zeng ◽  
Minghuang Wang
Keyword(s):  
Cross Section ◽  
Neutron Energy ◽  
Fission Products ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Neutron Shielding ◽  
Section Data ◽  
Nuclear Data Library ◽  
Data Library ◽  
High Neutron

A point-wise cross-section data library HENDL-ADS/MC (Hybrid Evaluated Nuclear Data Library) has been produced by FDS team to do the nuclear analysis for the ADS system. The HENDL-ADS/MC library contained 408 nuclide cross-section files including actinides, fission products and structural materials for neutron energy up to 150 MeV. The nuclear library also contained several sub-libraries with different temperatures. A series of neutron integral experiments and critical safety benchmarks have been performed to test the availability and reliability of the HENDL-ADS/MC data library. To validate and qualify the reliability of the high neutron energy cross section for HENDL-ADS/MC library further, a series of high neutron shielding experiments have been performed using MCNP. The testing results indicated the accuracy and reliability of HENDL-ADS/MC library.

Point Defect Absorption and Dislocation Loop Formation at Grain Boundaries in Hvem-Irradiated Zr and Its Alloys:1. Influence of Solute Addition

1985 ◽  
Vol 43 ◽  
pp. 350-351
Author(s):  
R.A. Herring ◽  
M. Griffiths ◽  
M.H Loretto ◽  
R.E. Smallman
Keyword(s):  
Grain Boundary ◽  
Grain Boundaries ◽  
Point Defect ◽  
Reactor Core ◽  
Solute Concentration ◽  
Nuclear Industry ◽  
Dislocation Loops ◽  
Loop Formation ◽  
Component Material ◽  
Impurity Interaction

Because Zr is used in the nuclear industry to sheath fuel and as structural component material within the reactor core, it is important to understand Zr's point defect properties. In the present work point defect-impurity interaction has been assessed by measuring the influence of grain boundaries on the width of the zone denuded of dislocation loops in a series of irradiated Zr alloys. Electropolished Zr and its alloys have been irradiated using an AEI EM7 HVEM at 1 MeV, ∼675 K and ∼10-6 torr vacuum pressure. During some HVEM irradiations it has been seen that there is a difference in the loop nucleation and growth behaviour adjacent to the grain boundary as compared with the mid-grain region. The width of the region influenced by the presence of the grain boundary should be a function of the irradiation temperature, dose rate, solute concentration and crystallographic orientation.

Investigation of the total neutron cross section for natural hafnium on the filtred neutron beam with energy 59 keV

2011 ◽  
Vol 30 (0) ◽  
pp. 142-147
Author(s):  
О. О. Грицай ◽  
А. К. Гримало ◽  
В. В. Колотий ◽  
В. М. Венедиктов ◽  
С. П. Волковецький ◽  
...  
Keyword(s):  
Cross Section ◽  
Neutron Beam ◽  
Neutron Cross Section ◽  
Total Neutron ◽  
Total Neutron Cross Section

Measurement of the Thermal Neutron Cross Section of the90Sr(n, γ)91Sr Reaction

1994 ◽  
Vol 31 (3) ◽  
pp. 173-179 ◽  
Author(s):  
Hideo HARADA ◽  
Toshiaki SEKINE ◽  
Yuichi HATSUKAWA ◽  
Noriko SHIGETA ◽  
Katsutoshi KOBAYASHI ◽  
...  
Keyword(s):  
Thermal Neutron ◽  
Cross Section ◽  
Neutron Cross Section ◽  
Thermal Neutron Cross Section

A Bent Crystal Neutron Spectrometer and Its Application to Neutron Cross-Section Measurements

1947 ◽  
Vol 72 (2) ◽  
pp. 109-115 ◽  
Author(s):  
R. B. Sawyer ◽  
E. O. Wollan ◽  
S. Bernstein ◽  
K. C. Peterson
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Neutron Spectrometer ◽  
Bent Crystal

Total Neutron Cross Section of Deuterium below 1000 keV

1972 ◽  
Vol 29 (26) ◽  
pp. 1745-1748 ◽  
Author(s):  
P. Stoler ◽  
N. N. Kaushal ◽  
F. Green ◽  
Edward Harms ◽  
Luciano Laroze
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Total Neutron ◽  
Total Neutron Cross Section

Cross Section and Fission Yields Induced Uncertainty in the Vver-440 Burnup Calculation

2021 ◽  
Author(s):  
Branislav Vrban ◽  
Stefan Cerba ◽  
Jakub Luley ◽  
Filip Osuský ◽  
Vladimir Necas
Keyword(s):  
Cross Section ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Reactor Operation ◽  
Precise Estimation ◽  
Stochastic Techniques ◽  
Core Characteristics ◽  
Fission Yields ◽  
Operational Data ◽  
Burnup Calculation

Abstract The properties of nuclear fuel depend on the actual isotopic composition which develops during a reactor operation. In practice, the prediction accuracy of burnup calculations serves as the basis for the future precise estimation of a core lifetime and other safety-based core characteristics. The present study quantifies nuclear data induced uncertainties of nuclide concentrations and multiplication factors in VVER-440 fuel depletion analysis. The well-known SCALE system and the TRITON sequence are used with the NEWT deterministic solver in the SAMPLER module that implements stochastic techniques to assess the uncertainty in computed results. The propagation of uncertainties in neutron cross section and fission yields is studied through the depletion calculation of 2D heterogeneous VVER-440 fuel assembly with an average enrichment of 4.87 wt % of 235U and six gadolinium rods with 3.35 % of Gd2O3. In the paper, fixed nominal depletion conditions are based on the real operational data of the Slovak NPP Bohunice unit 4 during cycle 30. In total 250 cases with uncertain parameters are computed and the results are evaluated by an auxiliary tool.

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