Neutron Flux

Page 1. 1484 IEEE TRANSACTIONS ON NUCLEAR SCIENCE, VOL. 40, NO. 6, DECEMBER 1993 Altitude and Latitude Variations in Avionics SEU and Atmospheric Neutron Flux E. Normand and T. J. Baker Boeing Defense & Space Group, Seattle, WA 98124-2499 Abstract ...

A moderator of paraffin wax assembly has been demonstrated where its thickness can be optimized to thermalize fast neutrons. The assembly is used for measuring fast neutron flux of a neutron probe at different neutron energies, using BF 3 (U1 00 and 2 00) and 3 He(U0.5 00) neutron detectors. The paraffin wax thickness was optimized at 6 cm for the neutron probe which contains an Am–Be neutron source. The experimental data are compared with Monte Carlo simulation results using MCNP5 version 1.4. Neutron flux comparison and neutron activation techniques are used for measuring neutron flux of the neutron probe to validate the optimum paraffin moderator thickness in the assembly. The neutron fluxes are measured at (1.17 ± 0.09) 9 10 5 and (1.19 ± 0.1) 9 10 5 n/s, being in agreement with the simulated values. The moderator assembly can easily be utilized for essential requirements of neutron flux measurements. Keywords Am–Be and 252 Cf neutron sources Á BF 3 & 3 He detectors Á Paraffin wax Á Neutron flux Á Monte Carlo simulation

A one dimensional steady state thermal analytical model has
been developed to study the heat transfer and temperature distributions
in a quartz ampoule filled with MoO3 powder. The
source heat generation inside the ampoule is released from the
high neutron flux (1.4 · 1014 neutrons/cm2s) interaction with
MoO3. Natural and forced convections heat transfer boundary
conditions are adopted during the irradiation process. The
peak temperatures in MoO3 powder and quartz are calculated
and compared with their melting temperatures to ensure the irradiation
safety criteria.

A B ST R A C T If a nuclear reactor happened to be prompt critical - even very slightly - the number of neutrons would increase exponentially at a high rate, and very quickly the reactor would become uncontrollable by means of cybernetics. The prompt neutron flux spectrum of the compact core of the Ghana’s miniature neutron source reactor (MNSR) was understudy using the Monte Carlo method. 20484 energy groups combined for all three categories of the energy distribution, thermal, slowing down and fast regions were modeled to create small energy bins. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The prompt thermal neutrons recorded it highest flux in the inner irradiation channel with a peak flux of (1.2091 ± 0.0008) × 10 12 n/cm 2 ·s, followed by the outer irradiation channel with a peak flux of (7.9393 0.0056) × 10 11 n/cm 2 ·s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3328 0.0004) × 10 11 n/cm 2 ·s. The peak values of the thermal energy range occurred in the energy range 1.8939× 10 -08 MeV – 3.7880× 10 -08 MeV. The inner irradiation channel again recorded the highest flux of (1.8361 0.0301) × 10 09

An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting t...

In this article, two-dimensional partial differential equations with time representation of nuclear power reac-tor kinetics are considered for spatial reactor dynamics and thermo hydraulic behavior analysis of a large thermal advanced gas cooled reactor (AGR) type used for nuclear power generation. The equations include the neutron flux equation and delayed neutron precursor concentration, together with taking into account the equations to represent the thermo hydraulic behavior of the fuel, coolant and moderator temperatures. These equations are solved numerically using the finite difference method. For time propagation, an implicit method is applied. The desired initial condition for the reactor to stay at stable critical condition is estab-lished by finding the correct value of reactivity. The reactivity disturbance effect in the reactor is studied for different cases and presented for high reactivity values. The model was developed for the analysis of a large AGR with 2000 MWe f...