Characterization of an ²⁴¹Am-Be neutron irradiation facility at Institute for Nuclear Science and Technology

An automated panoramic irradiator with a 241Am-Be neutron source of 5 Ci is installed in a bunker-type medium room at the Institute for Nuclear Science and Technology (INST) for calibration of neutron devices. Bonner Sphere Spectrometer (BSS) formed by 6 spheres plus bare detector, with cylindrical, almost point like, 6LiI(Eu) scintillator and 2 different spectral unfolding FRUIT and BUNKIUT codes are used to characterize the neutron field in different measurement points along the irradiation bench. The neutron field is also simulated by MCNP5 software and compared with measurements performed by the BSS. The paper shows the main results obtained in terms of neutron spectra at fixed distances from the source as well as their neutron fluence rate (totaland direct) and ambient dose equivalent rate. These values measured by the BSS with two unfolding FRUIT and BUNKIUT codes are in good agreement with that of simulated by MCNP5 within 10%.

Research on Thermal Neutron Shielding Performance of TiB2-Al Composite Materials

<p>Although the various excellent properties and preparation methods of TiB₂-based composites have been extensively studied, their neutron shielding properties have not received as much attention. In this article, the neutron shielding performance of the previously prepared TiB₂-Al composite will be studied. The photo neutron source device was used to carry out neutron irradiation tests on test samples with a thickness of 10 mm. The average thermal neutron shielding rate of TiB₂-based boron-containing composites is 17.55%, and the shielding rate increases with the increase of BN content. The macroscopic cross-section of thermal neutrons of the composites generally shows a stable trend, and when the BN content is 10%, the thermal neutrons macroscopic cross section reaches the maximum value of 7.58cm^-1. With the increase of the BN content, the thermal neutron fluence rate shows a gradually decreasing trend.</p>

Research on Thermal Neutron Shielding Performance of TiB2-Al Composite Materials

<p>Although the various excellent properties and preparation methods of TiB₂-based composites have been extensively studied, their neutron shielding properties have not received as much attention. In this article, the neutron shielding performance of the previously prepared TiB₂-Al composite will be studied. The photo neutron source device was used to carry out neutron irradiation tests on test samples with a thickness of 10 mm. The average thermal neutron shielding rate of TiB₂-based boron-containing composites is 17.55%, and the shielding rate increases with the increase of BN content. The macroscopic cross-section of thermal neutrons of the composites generally shows a stable trend, and when the BN content is 10%, the thermal neutrons macroscopic cross section reaches the maximum value of 7.58cm^-1. With the increase of the BN content, the thermal neutron fluence rate shows a gradually decreasing trend.</p>

ABSOLUTE MEASUREMENT OF 14.57 MeV NEUTRON FLUENCE RATE USING PROTON RECOIL NEUTRON TELESCOPE

A single stage vacuum-type proton recoil neutron telescope (PRT) was used for accurate measurement of 14.57 MeV neutron fluence rate from an indigenously developed D-T neutron generator at Purnima, BARC. The telescope consists of a polyethylene radiator having 4 cm diameter and CsI (Tl) scintillation crystal having thickness 1.5 mm and 4 cm diameter separated by 20.5 cm kept in a vacuum chamber. The neutron detection efficiency of the telescope for 14.57 MeV neutrons was calculated analytically using n-p scattering cross section data from Evaluated Nuclear Data File VII and also evaluated using fluka simulation. The relativistic transformation of n-p differential scattering cross section from centre-of-mass to laboratory system was used for calculating the efficiency of PRT. The 14.57 MeV neutron fluence rate was also measured using copper foils. The comparison of fluence rate measured using PRT and copper foil activation techniques is presented in this paper. The total uncertainty in measurement using PRT and copper foil activation technique is found to be 3.93 and 6.97%, respectively.

The Preliminary Shielding Analysis of CAP1400 Based on JMCT-S Code

JMCT-S[1] 3-D high resolution Monte Carlo neutron and photon transport code is proposed for shielding design due to its significant advantages such as complex geometry modeling, accurate calculation method, high efficient parallel computing and visible mass data analysis. To demonstrate its feasibility and benefits in engineering, this paper presents the verification of JMCT-S code based on primary shielding analysis for CAP1400 nuclear power station. Firstly, by employing JMCT-S code this paper completes the modeling and simulations of CAP1400 nuclear power station with full-core pin-by-pin and reactor pressure vessel internals inside of primary shield wall[2]. Secondly, neutron fluence rates are calculated based on the equilibrium cycle information. Particularly, the full-core neutron source is calculated from 3-D power distribution, fissionable nuclide fractions, fission spectrum, particle numbers and energy released per fission. Finally, JMCT-S code is verified by comparing results with those of the MCNP[3] code. According to the calculation results, JMCT-S code has similar accuracy to the MCNP code. The relative error of neutron fluence rate in active core is within ±1.5%, and deviation of neutron fluence rate for fast neutron fluence rate on inner surface of reactor pressure vessel is less than ±10%. Consequently, JMCT-S code is reliable in shielding design of reactor plant which lays the foundation for usage of JMCT-S code.