Measurement of neutron energy spectrum at the radial channel No. 4 of the Dalat reactor
نویسندگان
چکیده
INTRODUCTION Several compositions of neutron filters have been installed at the channel No. 4 of the Dalat research reactor to produce quasi-monoenergetic neutron beams. However, this neutron facility has been proposed to enhance the quality of the experimental instruments, and to characterize the neutron spectrum parameters for new filtered neutron beams of 2 keV, 24 keV, 59 keV and 133 keV. CASE DESCRIPTION In order to meet the demand of neutron spectrum information for calculation and design of filtered neutron facilities at the Dalat nuclear research reactor (DNRR), the experimental determinations of neutron flux and energy spectrum, up to 8 MeV, has been performed at the inner entrance of the horizontal channel No. 4 from the core of DNRR. The Westcott neutron fluxes as well as the α-parameter that represents the deviation of epithermal neutron distribution from the 1/E law were measured by applying the cadmium ratio and the multi-foils activation methods. The fast neutron spectrum was measured based on the iterative adjustment procedure with threshold reactions. DISCUSSION AND EVALUATION A set of pure metal thin foils with the diameter of 1.27 cm and thickness of 0.125 mm were used as threshold detectors to measure the integrated fluxes, and a calculation procedure on iterative adjustment was implemented to derive the differential neutron energy spectrum from the integrated values. CONCLUSIONS The neutron fluxes and spectrum parameters were characterized with the measured values of 4.80 × 10(9), 1.98 × 10(7), 5.06 × 10(8) cm(-2) s(-1) and 0.0448 for the thermal, epithermal, fast neutron fluxes and the α-shape factor, respectively. The present result has been significantly applied to the input data for the Monte Carlo simulations in the developments of filtered mono-energetic neutron beam facility at the institute.
منابع مشابه
Estimation of neutron and gamma dose in the MNSR research reactor
In this study, the neutron and gamma doses in the dry channel and in the internal irradiation site of the Miniature Neutron Source research reactor (MNSR) has been calculated and measured. The MNSR reactor is a light water reactor with a maximum power of 30 kW and equipped with various irradiation facilities, including five irradiated sites, five irradiation sites and a dry channel. The interna...
متن کاملبررسی امکان استفاده از چشمه های نوترونی رادیوایزوتوپی در نوترون درمانی با بور
Background : Performing successful BNCT experiments needs a suitable neutron source. Important factors of the neutron beam are flux and energy that are very important in the selection of neutron source. In most centers that use this method for treatment, reactor is a neutron source, which according to characteristics of the reactor appropriated neutrons are very high. High cost of constructin...
متن کاملModeling the measurement of VVER-1000 reactor power by neutron and gamma radiation with MCNP code
The present study deals with a new method for measuring the power of a reactor. This method uses gamma and neutron radiation resulted from the entire reactor structure, without changing its structure (online). In terms of functionality, this method can measure the reactor power in real-time and report it instantly. In order to obtain the relationship between reactor power and gamma and neutron ...
متن کاملDetermination of neutron temperature in irradiation channels of reactor
The neutron temperature is a characteristic parameter in irradiation channels of reactor. For nuclides which have resonances in the thermal neutron energy range, their Westcott g-factors are different from unity. The values of g-factors and cross-sections of (n, γ) reaction of these nuclides are temperature dependence. The standard energy for tabulation of thermal neutron cross-section (σ0) is ...
متن کاملمحاسبات دز نوترون در حادثه بحرانیت JCO در Tokaimura با کد MCNPX
Recognizing the accident and estimating absorbed doses at the incident time, is one of the requirements for radiation safety. The aim of this paper is designing a model for assessment of nuclear criticality effectiveness in non-reactor units and evaluation of the effect of variation of distances on dose rate and neutron energy spectrum. In this study neutron dose-rate was simulated between 0.5m...
متن کامل