scholarly journals Activities in Divertor Reflector and Linear Plates Using WCLL and HCPB Breeding Blanket Concepts

Energies ◽  
2021 ◽  
Vol 14 (24) ◽  
pp. 8305
Author(s):  
Simona Breidokaite ◽  
Gediminas Stankunas
Keyword(s):  
Cross Sections ◽  
Thermal Effects ◽  
Nuclear Reactions ◽  
Nuclear Data ◽  
Alpha Particles ◽  
Vacuum Vessel ◽  
Plasma Facing Components ◽  
Nuclear Data Library ◽  
Safe Performance ◽  
Data Library

In fusion devices, such as European Demonstration Fusion Power Reactor (EU DEMO), primary neutrons can cause material activation due to the interaction between the source particles and the targeting material. Subsequently, the reactor’s inner components become activated. For safety and safe performance purposes, it is necessary to evaluate neutron-induced activities. Activities results from divertor reflector and liner plates are presented in this work. The purpose of liner shielding plates is to protect the vacuum vessel and magnet coils from neutrons. As for reflector plates, the function is to shield the cooling components under plasma-facing components from alpha particles, thermal effects, and impurities. Plates are made of Eurofer with a 3 mm layer of tungsten, while the water is used for cooling purposes. The calculations were performed using two EU DEMO MCNP (Monte Carlo N-Particles) models with different breeding blanket configurations: helium-cooled pebble bed (HCPB) and water-cooled lithium lead (WCLL). The TENDL–2017 nuclear data library has been used for activation reactions cross-sections and nuclear reactions. Activation calculations were performed using the FISPACT-II code at the end of irradiation for cooling times of 0 s–1000 years. Radionuclide analysis of divertor liner and reflector plates is also presented in this paper. The main radionuclides, with at least 1% contribution to the total value of activation characteristics, were identified for the previously mentioned cooling times.

scholarly journals Calculated Radioactivity Yields of Gallium-67 using Matlab Codes

2020 ◽  
Vol 21 (1) ◽  
pp. 21
Author(s):  
Imam Kambali
Keyword(s):  
Cross Sections ◽  
Nuclear Reactions ◽  
Nuclear Data ◽  
Proton Bombardment ◽  
Matlab Code ◽  
Nuclear Data Library ◽  
Gallium 67 ◽  
Data Library ◽  
Mev Protons ◽  
Evaluated Nuclear Data

In nuclear medicine, gallium-67 (67Ga) is potentially applied for imaging a certain type of tissue. In this investigation, 67Ga is theoretically studied in terms of its potential radioactivity yields at the end of various energetic proton bombardments.  Nuclear cross-sections derived from the Talys Evaluated Nuclear Data Library (TENDL) 2017 were used as the input files, while a Matlab code was developed to perform the yield calculations of 67Zn(p,n)67Ga and 68Zn(p,2n)67Ga nuclear reactions to produce 67Ga. Two different targets – enriched 67Zn and natZn targets – were simulated in the calculations. The calculated yields suggested that a maximum of 27.37 MBq/µAh could be achieved when enriched 67Zn target was irradiated with 15-MeV protons, whereas 46.99 MBq/µAh could be generated following a 30 MeV proton bombardment of enriched 68Zn target. Various radioactive gallium impurities, i.e. 63,64,65,66,68,70Ga and stable 69Ga isotope were also expected to be generated mostly via (p,n) and (p,2n) reactions when natZn target was used in the 67Ga production. In contrast, radioactive 66Ga and 68Ga impurities were mainly produced following bombardment of enriched 67Zn and 68Zn targets. This study can be used as a reference for future 67Ga radionuclide production.

scholarly journals CENDL-3.2: The new version of Chinese general purpose evaluated nuclear data library

2020 ◽  
Vol 239 ◽  
pp. 09001
Author(s):  
Zhigang Ge ◽  
Ruirui Xu ◽  
Haicheng Wu ◽  
Yue Zhang ◽  
Guochang Chen ◽  
...  
Keyword(s):  
Cross Sections ◽  
Nuclear Data ◽  
General Purpose ◽  
Reaction Cross ◽  
Main Reaction ◽  
Science And Engineering ◽  
Nuclear Data Library ◽  
Evaluation Techniques ◽  
Data Library ◽  
Evaluated Nuclear Data

A new version of Chinese Evaluated Nuclear Data Library, namely CENDL-3.2, has been completed under the joint efforts of CENDL working group. This library is constructed with the general purpose to provide high-quality nuclear data for the modern nuclear science and engineering. 272 nuclides from light to heavy are covered in CENDL-3.2 in total and the data for 134 nuclides are new or updated evaluations in energy region of 10-5 eV-20 MeV. The data of most of the key nuclides in nuclear application like U, Pu, Th, Fe et al. have been revised and improved, and various evaluation techniques have been developed to produce the nuclear data with good quality. Moreover, model dependent covariances data for main reaction cross sections are added for 70 fission product nuclides. To assess the accuracy of CENDL-3.2 in application, the data have been tested with the criticality and shielding benchmarks collected in ENDITS-1.0.

scholarly journals BETA EFFECTIVE SENSITIVITY TO NUCLEAR DATA WITHIN THE APOLLO3® NEUTRONIC PLATFORM

2021 ◽  
Vol 247 ◽  
pp. 15003
Author(s):  
G. Valocchi ◽  
P. Archier ◽  
J. Tommasi
Keyword(s):  
Sensitivity Analysis ◽  
Cross Sections ◽  
Uncertainty Propagation ◽  
Direct Calculation ◽  
Nuclear Data ◽  
The Core ◽  
Nuclear Data Library ◽  
Lattice Core ◽  
Effective Sensitivity ◽  
Data Library

In this paper, we present a sensitivity analysis of the beta effective to nuclear data for the UM17x17 experiment that has been performed in the EOLE reactor. This work is carried out using the APOLLO3® platform. Regarding the flux calculation, the standard two-step approach (lattice/core) is used. For what concerns the delayed nuclear data, they are processed to be directly used in the core calculation without going through the lattice one. We use the JEFF-3.1.1 nuclear data library for cross-sections and delayed data. The calculation of k-effective and beta effective is validated against a TRIPOLI4® one while the main sensitivities are validated against direct calculation. Finally, uncertainty propagation is performed using the COMAC-V2.0 covariance library.

Temperature Dependence of the Westcott g-factor for Neutron Reactions in Activation Analysis

1999 ◽  
Vol 71 (12) ◽  
pp. 2309-2315 ◽  
Author(s):  
N. E. Holden
Keyword(s):  
Cross Section ◽  
Neutron Capture ◽  
Nuclear Reactions ◽  
Nuclear Data ◽  
Reaction Rates ◽  
Nuclear Data Library ◽  
Neutron Reactions ◽  
Using Data ◽  
Data Library ◽  
G Factors

The Westcott g-factors, which allow the user to determine reaction rates for nuclear reactions taking place at various temperatures, have been calculated using data from the Evaluated Neutron Nuclear Data Library, ENDF/B-VI. Nuclides chosen have g-factors which are significantly different from unity and result in different reaction rates compared to nuclides whose neutron capture cross section varies as the reciprocal of the neutron velocity. Values are presented as a function of temperature up to 673.16 K (400 °C).

scholarly journals Uncertainty quantification and sensitivity studies on Thorium-fueled reactors

2020 ◽  
Vol 239 ◽  
pp. 22008
Author(s):  
Eliot Party ◽  
Xavier Doligez ◽  
Philippe Dessagne ◽  
Maëlle Kerveno ◽  
Greg Henning
Keyword(s):  
Cross Sections ◽  
Nuclear Data ◽  
Pressurized Water ◽  
Correlation Matrices ◽  
Target Uncertainty ◽  
Nuclear Data Library ◽  
Sensitivity Studies ◽  
Water Reactors ◽  
Data Library ◽  
Good Agreement

This paper shows how Total Monte Carlo (TMC) method and Perturbation Theory (PT) can be applied to quantify uncertainty due to nuclear data on reactor static calculations of integral parameters such as keff and βeff. This work focuses on thorium fueled reactors and it aims to rank different cross sections uncertainty regarding criticality calculations. The consistency of the two methods are first studied. The cross sections set used for the TMC method is computed to build adequate correlation matrices. Those matrices are then multiplied by the sensitivity coefficients obtained thanks to the PT to obtain global uncertainties that are compared to the ones calculated by the TMC method. Results in good agreement allow us to use correlation matrix from the state of the art nuclear data library (JEFF 3-3) that provide insight of uncertainty on keff and βeff for thorium fueled Pressurized Water Reactors. Finally, maximum uncertainties on cross sections are estimated to reach a target uncertainty on integral parameters. It is shown that a strong reduction of the current uncertainty is needed and consequently, new measurements and evaluations have to be performed.

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