Investigation of (n,γ) reactions in fissionable fluids in a hybrid reactor system
-
M. Günay
Abstract
In this study, the effect of (n,γ) reactions in spent fuel-grade (SFG)-PuO2, UO2, NpO2 and UCO contents was investigated in a designed hybrid reactor system. In this system, the molten salt-heavy metal mixtures 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % UO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % NpO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % UCO were used as fluids. Beryllium (Be) is used as neutron multiplier by (n,2n) reactions. Thence, a Be zone with a width of 3 cm was used in order to contribute on fissile fuel breeding between the liquid first wall and blanket. (n,γ) reactions were calculated in the liquid first wall, blanket and shield zones, which SFG-PuO2, UO2, NpO2 and UCO contents. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code and nuclear data library ENDF/B-VII.0.
Kurzfassung
In dieser Studie wurden in einem konzipierten Reaktorsystem (n,γ) Reaktionen bei Spent-Fuel-grade Komponenten (SFG)-PuO2, UO2, NpO2 und UCO untersucht. In diesem System wurden als Fluide gelöste Salz-Schwermetal-Mischungen aus 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 und 2 – 10 % UO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 und 2 – 10 % NpO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 und 2 – 10 % UCO verwendet. Beryllium wird über die (n, 2n) Reaktion als Neutronenmultiplikator verwendet. Deshalb wurde zwischen der ersten Flüssigwand und dem Blankett ein 3 cm dicker Be-Bereich eingesetzt, um so die Herstellung von spaltbarem Brennstoff zu fördern. Die (n,γ) Reaktion wurde an der aus SFG-PuO2, UO2, NpO2 und UCO, bestehenden ersten Flüssigwand, am Blankett- und am Shield-Bereich berechnet. Als Baumaterial wurde ein vier Zentimeter dicker ferritischer Stahl – 9Cr2WVTa – verwendet. Dreidimensionale nukleonische Berechnungen wurden mit Hilfe der neuesten Version des Monte Carlo Codes MCNPX-2.7.0 und unter Verwendung der nuklearen Datenbibliothek ENDF/B-VII.0 durchgeführt.
References
1 Şahin, S.; Übeyli, M.: Radiation Damage Studies on The First Wall of a HYLIFE-II Type Fusion Breeder. Energy Conversion and Management46 (2005) 318510.1016/j.enconman.2005.03.007Suche in Google Scholar
2 Şahin, S.; et al.: Effects of spectral shifting in an inertial confinement fusion system. Kerntechnik70 (2005) 23310.3139/124.100251Suche in Google Scholar
3 Şarer, B.; Günay, M.; et al.: Three-Dimensional Neutronic Calculations For The Fusion Breeder Apex Reactor. Fusion Science and Technology52 (2007) 107Suche in Google Scholar
4 Günay, M.; et al.: Three-Dimensional Neutronic Calculations for a Fusion Breeder Apex Reactor Using Some Libraries. Annals of Nuclear Energy38 (12) (2011) 275710.1016/j.anucene.2011.08.007Suche in Google Scholar
5 Günay, M.; et al.: Three-dimensional Monte Carlo Calculation of Gas Production in Structural Material of APEX Reactor for Some Evaluated Data Files. Annals of Nuclear Energy55 (2013) 29210.1016/j.anucene.2013.01.001Suche in Google Scholar
6 Günay, M.; et al.: The Effect on Radiation Damage of Structural Material in a Hybrid System By Using a Monte Carlo Radiation Transport Code. Annals of Nuclear Energy63 (2013) 15710.1016/j.anucene.2013.07.038Suche in Google Scholar
7 IAEA: International Atomic Energy Agency IAEA-TECDOC-1349, 2003Suche in Google Scholar
8 Christofilos, N. C.: Design for A High Power-Density Astron Reactor. Journal of Fusion Energy8 (1989) 9710.1007/BF01050784Suche in Google Scholar
9 Moir, R. W.: Liquid First Walls For Magnetic Fusion Energy Configurations. Nuclear Fusion37 (1997) 55710.1088/0029-5515/37/4/I13Suche in Google Scholar
10 Abdou, M. A.; et al.: Chapter 1: Overview. APEX Interim Report, 1999Suche in Google Scholar
11 Abdou, M. A.; et al.: On The Exploration of Innovative Concepts for Fusion Chamber Technology. Fusion Engineering and Design54 (2001) 18110.1016/S0920-3796(00)00433-6Suche in Google Scholar
12 Abdou, M. A.: Preface. Fusion Engineering and Design72 (2004) 110.1016/j.fusengdes.2004.10.001Suche in Google Scholar
13 Abdou, M. A.; et al.: Overview of Fusion Blanket R&D in The US Over The Last Decade. Nuclear Engineering and Technology37 (5) (2005) 401Suche in Google Scholar
14 Piera, M.; et al.: Hybrid Reactors: Nuclear Breeding or Energy Production?Energy Conversion and Management51 (2010) 175810.1016/j.enconman.2010.01.025Suche in Google Scholar
15 Ying, A.; et al.: Chapter 5: Thick Liquid Blanket Concept, APEX Interim Report, 1999Suche in Google Scholar
16 Şarer, B.; et al.: Calculations of Neutron-Induced Production Cross-Sections of 180,182,183,184,186W up to 20 MeV. Annals of Nuclear Energy36 (2009) 41710.1016/j.anucene.2008.11.025Suche in Google Scholar
17 Günay, M.: Investigation of radiation damage in structural material of APEX reactor by using Monte Carlo method. Annals of Nuclear Energy53 (2013) 5910.1016/j.anucene.2012.06.038Suche in Google Scholar
18 Günay, M.; Kasap, H.: Neutronic Investigation of The Application of Certain lutonium-Mixed Fluids in a Fusion-Fission Hybrid Reactor. Annals of Nuclear Energy63 (2014) 43210.1016/j.anucene.2013.08.024Suche in Google Scholar
19 Chadwick, M. B.; et al.: ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology. Nuclear Data Sheets, 107 (2006) 293110.1016/j.nds.2006.11.001Suche in Google Scholar
20 Pelowitz, D. B.: MCNPX User’s Manual, Version 2.7.0, LA-CP-11-00438, 2011Suche in Google Scholar
21 Günay, M.: Investigation of Neutronic Effects in Structural Material of a Hybrid Reactor by using The MCNPX Monte Carlo Transport Code. Kerntechnik 78:3 (2013) 198Suche in Google Scholar
22 Şarer, B.; et al.: Comparisons of The Calculations Using Different Codes Implemented in MCNPX Monte Carlo Transport Code for Accelerator Driven System Target. Fusion Science Technology61 (2012) 302Suche in Google Scholar
© 2014, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Simulation of thermal fluid dynamics in parabolic trough receiver tubes with direct steam generation using the computer code ATHLET
- Measurement of velocity profiles of nanofluids in laminar channel flow using Particle Image Velocimetry
- Prediction of correlation between two-phase natural circulation flows in heated and unheated channels of a parallel channel system – based on electrical analogy
- Investigation of (n,γ) reactions in fissionable fluids in a hybrid reactor system
- In-pile modelling of nuclear fuel element for the MTR type reactors – Part 2
- One-step synthesis of Pt-reduced graphene oxide composites based on high-energy radiation technique
- Hamming generalized corrector for reactivity calculation
- Experimental study of flow inversion in MTR upward flow research reactors
- Technical Note
- The effect of burn up on the kinetic parameters for a pressurized water reactor fueled by MOX using MCNPX code
- Diffusion length calculations for one-speed neutrons in a slab with backward, forward and linear anisotropic scattering
- Improvement of passive shielding to reduce background components to determinate radioactivity at low energy gamma rays
- A study of the energy enhancement of electron in radio frequency (RF) linear accelerator of iris loaded waveguards
- Age-dependent effective doses for radionuclides uniformly distributed in air
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Simulation of thermal fluid dynamics in parabolic trough receiver tubes with direct steam generation using the computer code ATHLET
- Measurement of velocity profiles of nanofluids in laminar channel flow using Particle Image Velocimetry
- Prediction of correlation between two-phase natural circulation flows in heated and unheated channels of a parallel channel system – based on electrical analogy
- Investigation of (n,γ) reactions in fissionable fluids in a hybrid reactor system
- In-pile modelling of nuclear fuel element for the MTR type reactors – Part 2
- One-step synthesis of Pt-reduced graphene oxide composites based on high-energy radiation technique
- Hamming generalized corrector for reactivity calculation
- Experimental study of flow inversion in MTR upward flow research reactors
- Technical Note
- The effect of burn up on the kinetic parameters for a pressurized water reactor fueled by MOX using MCNPX code
- Diffusion length calculations for one-speed neutrons in a slab with backward, forward and linear anisotropic scattering
- Improvement of passive shielding to reduce background components to determinate radioactivity at low energy gamma rays
- A study of the energy enhancement of electron in radio frequency (RF) linear accelerator of iris loaded waveguards
- Age-dependent effective doses for radionuclides uniformly distributed in air
