Study of the Space-Time Neutron Multiplication Formula

2010 ◽  
Author(s):  
Jianli Hao ◽  
Wenzhen Chen ◽  
Shaoming Wang ◽  
De Zhang
Keyword(s):  
Neutron Flux ◽  
Discrete Time ◽  
Neutron Transport ◽  
Space Time ◽  
Neutron Density ◽  
Time Process ◽  
Delayed Neutrons ◽  
Multiplication Formula ◽  
Neutron Flux Density ◽  
Neutron Multiplication

The process of neutron multiplication is a discrete-time process, but the neutron transport theory takes neutron multiplication as a continuous neutron source, which ignores the discrete-time process of neutron multiplication, which would take in errors, so it is necessary for describing the process of neutron multiplication as a discrete-time process. “The neutron doubling formula including delayed neutrons” has been established which describes the process of neutron multiplication as a discrete-time process, but it has nothing to do with space. “The neutron doubling formula including delayed neutrons” could not be used to describe the variety of distributing of neutron density in transient process; it also could not be used to deal with the problem of three-dimensional space. In order to solve the problems mentioned above, the space-time neutron multiplication formula is established. Based on the theory of neutron multiplication, the concept of space is introduced to the neutron multiplication formula and the space-time neutron multiplication formula is established by taking into account of neutron transport. The formula can describe the inherent physical process of neutron multiplication in fission chain reaction system. The test of space-time neutron multiplication formula is done, which proves the formula is right. Given the initial neutron density as well as the multiplication factor, the formula can strictly describe the variety of neutron density (neutron flux density) with time. It could be used for setting a standard for estimating error for the measurement of neutron flux density as well as numerical calculation; the space-time neutron multiplication has larger applicability compared with the “neutron doubling formula including delayed neutrons”.

Simulation on Neutron Space-Time Kinetics for Accelerator Driven Sub-Critical System

2017 ◽  
Author(s):  
Yingming Song ◽  
Qingyu Gao ◽  
Ke Wang ◽  
Yaping Guo ◽  
Lu Zhang ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Transport Theory ◽  
Neutron Transport ◽  
Space Time ◽  
Outlet Temperature ◽  
Critical System ◽  
Static Approximation ◽  
Energy Group ◽  
Mc Method

Monte Carlo transport theory was applied to the variables space and time separated framework of neutron space-time kinetics calculation for Accelerator driven sub-critical reactor. The improved quasi-static approximation was combined with Monte Carlo neutron transport code (IQS/MC) for neutron space-time kinetic process of ADS sub-critical system. Besides, the IQS/MC simulation calculation program with visualization operation platform for ADS sub-critical system was developed simultaneously. The beam transient was analysed simulatedly based on the physical model of CIADS. Three-dimensional grid distributions of relative neutron flux of energy group were separated along time can be obtained by computing energy group separated of neutron flux, meanwhile the totally relative power, fuel temperature and outlet temperature of coolant at the core varied as the time were also obtained. The calculated results of IQS/MC method and point reactor method were compared, which agreed well with the relevant physics laws and verify that the IQS/MC method is applicable to the simulation of ADS neutron space-time kinetics and ADS neutronics transient security analysis.

scholarly journals Experimental methods for determining the effective neutron multiplication factor of the “Yalina-Thermal” subcritical assembly

2020 ◽  
Vol 65 (2) ◽  
pp. 235-242 ◽  
Author(s):  
I. A. Edchik ◽  
T. N. Korbut ◽  
A. V. Kuzmin ◽  
S. E. Mazanik ◽  
V. P. Togushov ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Probabilistic Method ◽  
Nuclear Research ◽  
Fuel Load ◽  
Neutron Density ◽  
Density Level ◽  
Multiplication Factor ◽  
Nuclear Facilities ◽  
Neutron Multiplication ◽  
Effective Neutron

To study the kinetics of subcritical systems and determine the optimal conditions for the transmutation of longlived radioactive waste in the neutron spectrum of ADS-systems the “Yalina” research nuclear facility was created at Joint Institute for Power and Nuclear Research – Sosny (Minsk, Belarus). The main safety indicator of a subcritical system (active zone reactivity) was measured for a “Yalina-Thermal” assembly via three independent methods: inverse multiplication, probabilistic and impulse ones. For the inverse multiplication method, the neutron flux density was monitored during assembly loading. For a fuel load of 285 EK-10 rods the neutron multiplication was M = 22.3±0.6, and the effective neutron multiplication coefficient was keff = 0.9551± 0.0016. The probabilistic method (Feynman-alpha method), based on measuring fluctuations in the neutron density level within a system with a fission chain reaction, gave the ratio of the variance to the average counting rate value D/n = 1.779±0.005, which corresponds to keff = 0.9597 ±0.0003. The pulse method is aimed at studying the neutron flux behavior of after the neutron pulse injection into the breeding system. Measurements were held with the same setup, used in the Feynman-alpha method. The measured decay constant of instantaneous neutrons is α = –670±0.7 1/s, which corresponds to keff = 0.9560±0.0001. The effective multiplication factor keff of the subcritical assembly “Yalina-Thermal”, obtained via three different independent methods, is around average value of keff = 0.9569 ± 0.0018. The methods considered can be used for subcritical level monitoring for ADS-systems and research nuclear facilities.

scholarly journals Solution of neutron-transport multigroup equations system in subcritical systems

2018 ◽  
Vol 4 (1) ◽  
pp. 79-85 ◽  
Author(s):  
Igor V. Shamanin ◽  
Sergey V. Bedenko ◽  
Vladimir N. Nesterov ◽  
Igor O. Lutsik ◽  
Anatoly A. Prets
Keyword(s):  
Neutron Flux ◽  
Transport Equation ◽  
Flux Density ◽  
Diffusion Approximation ◽  
Neutron Transport ◽  
Nuclear Data ◽  
Neutron Transport Equation ◽  
Neutron Flux Density ◽  
Data Libraries ◽  
Nuclear Data Libraries

An iteration method has been implemented to solve a neutron transport equation in a multigroup diffusion approximation. A thermoelectric generator containing plutonium dioxide, used as a source of thermal and electric power in spacecraft, was studied. Neutron yield and multigroup diffusion approximation data was used to obtain a continuous and group distribution of neutron flux density spectra in a subcritical multiplying system. Numerical multigroup approaches were employed using BNAB-78, a system of group constants, and other available evaluated nuclear data libraries (ROSFOND, BROND, BNAB, EXFOR and ENDSF). The functions of neutron distribution in the zero iteration for the system of multigroup equations were obtained by approximating an extensive list of calculated and experimental data offered by the EXFOR and ENDSF nuclear data libraries. The required neutronic functionals were obtained by solving a neutron transport equation in a 28-group diffusion approximation. The calculated data was verified. The approach used is more efficient in terms of computational efforts (the values of the neutron flux density fractions converge in the third iteration). The implemented technique can be used in nuclear and radiation safety problems.

Mathematical Modeling of Silicon Doping by Neutron Transmutation Doping Method for High Efficient Solar Cells

MRS Advances ◽  
2017 ◽  
Vol 2 (53) ◽  
pp. 3135-3140
Author(s):  
Valery L. Dshkhunyan ◽  
Alexander A. Dyakov ◽  
Sergey M. Karabanov ◽  
Andrey V. Kozlov ◽  
Dmitry V. Markov ◽  
...  
Keyword(s):  
Mathematical Modeling ◽  
Solar Cells ◽  
Neutron Flux ◽  
Flux Density ◽  
Neutron Transport ◽  
Neutron Transmutation Doping ◽  
Si Solar Cells ◽  
Neutron Flux Density ◽  
Transmutation Doping ◽  
Neutron Transmutation

ABSTRACTIt is known that n-Si solar cells have higher efficiency than p-Si solar cells. One of the problems connected with n-Si application for solar cell production is the difficulty of using Czochralski method for growing n-Si ingots, uniform in structure. The present paper examines the possibility of production of n-Si ingots, uniform in resistance, by neutron transmutation doping (NTD) for photovoltaics using the mathematical modeling method. The provided calculation data are obtained by MCU-RFFI/A accounting code with DLC/MCUDAT-1.0 constant library developed by «Kurchatov Institute» Russian Research Center. The MCU accounting code is used for solution of the neutron-transport equation by Monte-Carlo procedure on the basis of estimated nuclear data for arbitrary three-dimensional geometry systems.The present paper provides the estimation of uniformity of neutron-flux density along the ingot length and radius; dependence of silicon resistance on duration of irradiation. These studies established the neutron flux density distribution along the ingot length and radius; regularities of silicon resistance changes on duration and intensity of irradiation.

Comparison of Neutron Flux Density in Carbon Prism filled with NaCl and Air

2020 ◽  
Author(s):  
Kamil Stevanka ◽  
Dusan Kral ◽  
Ondrej Stastny ◽  
Robert Holomb ◽  
Karel Katovsky ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Flux Density ◽  
Neutron Flux Density

KNVK vacuum fission chambers for measuring neutron flux density

1987 ◽  
Vol 62 (3) ◽  
pp. 232-237
Author(s):  
E. K. Malyshev ◽  
S. V. Chuklyaev ◽  
O. I. Shchetinin
Keyword(s):  
Neutron Flux ◽  
Flux Density ◽  
Neutron Flux Density ◽  
Fission Chambers

Nuclear Instrumentation Systems in Prototype Fast Breeder Reactor

2004 ◽  
Author(s):  
P. M. Vijayakumaran ◽  
C. P. Nagaraj ◽  
C. Paramasivan Pillai ◽  
R. Ramakrishnan ◽  
M. Sivaramakrishna
Keyword(s):  
Neutron Flux ◽  
Early Warning System ◽  
Warning System ◽  
Radiation Monitoring ◽  
Fast Breeder Reactor ◽  
Delayed Neutrons ◽  
The Core ◽  
Flux Monitoring ◽  
Nuclear Instrumentation ◽  
Fission Chambers

The nuclear instrumentation systems of the Prototype Fast Breeder Reactor (PFBR) primarily comprise of global Neutron Flux Monitoring, Failed Fuel Detection & Location, Radiation Monitoring and Post-Accident Monitoring. High temperature fission chambers are provided at in-vessel locations for monitoring neutron flux. Failed fuel detection and location is by monitoring the cover gas for fission gases and primary sodium for delayed neutrons. Signals of the core monitoring detectors are used to initiate SCRAM to protect the reactor from various postulated initiating events. Radiation levels in all potentially radioactive areas are monitored to act as an early warning system to keep the release of radioactivity to the environment and exposure to personnel well below the permissible limits. Fission Chambers and Gamma Ionisation Chambers are located in the reactor vault concrete for monitoring the neutron flux and gamma radiation levels during and after an accident.

scholarly journals Uncertainty Analysis of WWER-1000 Core Macroscopic Cross Sections due to Spectral Effects

2020 ◽  
pp. 39-46
Author(s):  
О. Kukhotska ◽  
I. Ovdiienko ◽  
M. Ieremenko
Keyword(s):  
Energy Spectrum ◽  
Isotopic Composition ◽  
Uncertainty Analysis ◽  
Neutron Flux ◽  
Flux Density ◽  
Cross Section ◽  
Cross Sections ◽  
Fuel Burnup ◽  
Thermophysical Parameters ◽  
Neutron Flux Density

The paper presents the results of uncertainty analysis of WWER‑1000 core macroscopic cross sections due to spectral effects during WWER‑1000 fuel burnup and the analysis of cross section sensitivity from thermophysical parameters of the calculated cell, which affect energy spectrum of neutron flux density. The calculation of changes in the isotopic composition during burnup and the preparation of macroscopic cross sections used the developed HELIOS computer model [1] for TVSA, which is currently operated at most Ukrainian WWER‑1000 units. The GRS approach applying Software for Uncertainty and Sensitivity Analyses (SUSA) [2] was chosen to assess the uncertainty of the macroscopic cross sections due to spectral effects and analysis of cross section sensitivity from thermophysical parameters. The spectral effect on macroscopic cross sections was taken into account by calculating the fuel burnup for variational sets of thermophysical parameters (fuel temperature, coolant temperature and density, boric acid concentration) prepared in advance by the SUSA program, as a result of which fuel isotopic composition vectors were obtained. After that, neutronic constants for the reference state were developed for each of the sets of isotopic composition, which corresponded to a certain set of thermophysical parameters. At the next stage, the uncertainty of macroscopic cross sections of the interaction due to the spectral effects on the isotopic composition of the fuel was analyzed using SUSA 4, followed by the analysis of cross section sensitivity from thermophysical parameters of the calculated cell affecting energy spectrum of neutron flux density. In the future, the uncertainty of two-group macroscopic diffusion constants can be used to estimate the overall uncertainty of neutronic characteristics in large-grid core calculations, in particular, in the safety analysis.

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