Benzoxazine resin as an interesting building block for advanced neutrons shields

The design of lightweight neutrons shields has been restricted for quite some time to the use of the epoxy thermosets as the main building blocks. Meanwhile, the recent developments in the field of polymers suggest otherwise. Indeed, benzoxazine resins have taken the lead over the traditional thermosets in many exigent applications. Therefore, in a vision to introduce newer matrices with better performances and to further expand the applications of the benzoxazine resins into the nuclear field, the neutron shielding efficiency along with the thermal and thermomechanical performances of the neat benzoxazine polymer and its subsequent B4C-reinforced composites were investigated. The neutron shielding measurements were performed using an optimized experimental setup at NUR research reactor, Algiers. The neat benzoxazine polymer displayed almost similar thermal neutrons screening performances than the epoxy with a macroscopic cross-section (Σ) of a 0.724 cm− 1 equivalent to a mean free path (λ) of 0.957 cm. The effect of the particle amount was also studied to maximize the shielding ability of the developed materials. For instance, the benzoxazine composite containing 20 wt.% of B4C displayed the outstanding screening ratio of about 96% for a sample thickness of 13 mm. Finally, the remarkable findings were put into context by providing multifaceted comparisons with the available shielding materials.

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>

VERIFICATION OF THE OpenMC HOMOGENIZED MYRRHA-1.6 CORE MODEL

The OpenMC code is being employed both as a multi-group nodal macroscopic cross-section generator and a reference multi-group Monte Carlo (MGMC) solution. The aim is to do a neutronic benchmark verification study versus a deterministic model (based on the MYRRHA-1.6 core) performed by the PHISICS simulator. MYRRHA, a novel research accelerator driven system concept that is also foreseen to work as a critical configuration, offers a rich opportunity of testing state-of-the art methods for reactor physics analysis due to its strong heterogeneous configuration utilized for both thermal and fast spectra irradiation purposes. The original core configuration representing MYRRHA-1.6 and formed by 169 assemblies, was launched in OpenMC for producing a homogenous nodal model that, when executed in its multi-group Monte Carlo mode, it produced a keff that differs in almost 500 pcm from the original case. This means that in the future, such approximation should correct the nodal cross-sections to preserve the reaction rates in order to match those ones from the heterogeneous model. Nevertheless, such MGMC mode of operation offered by OpenMC could be exploited in order to verify deterministic core simulators. By inputting the same nodal multi-group cross-section model into the transport solver of the PHISICS toolkit, the neutronic benchmark showed a difference of 171 pcm in eigenvalue while comparing it to its OpenMC MGMC counterpart. Also, local multi-group and energy-integrated nodal profiles of the neutron flux showed a maximum relative difference between methodologies of 15% and 1%, respectively. This means that the MGMC capabilities offered by OpenMC can be employed to verify other deterministic methodologies.

ADVANCED HOMOGENIZATION METHODS FOR PRESSURIZED WATER REACTORS

Macroscopic cross section generation is key part of core calculation. Commonly, the data are prepared independently without a knowledge of fuel loading pattern. The fuel assemblies are simulated in infinite lattice (with mirror boundary conditions). Rehomogenization method is based on combination of actual neutron flux in fuel assembly with macroscopic data from infinite lattice. Rehomogenization method was implemented into the macrocode Andrea and tested on a reference cases. Cases consist of fuel cases, cases with strong absorber, cases with absorption rods, or cases with reflector assemblies. Testing method is based on a comparisons of homogenized and rehomogenized macroscopic cross sections and later on a comparisons of relative power of each fuel assembly. Above that there is comparison of eigenvalue.

Effect of Boron Carbide Content on Properties of Thermoplastic Natural Rubber Composite as Thermal Neutron Shielding

In this work, thermoplastic natural rubber (TPNR) composites were produced through melt blending method. Boron carbide (B4C) as filler was added into the polymer blend (TPNR) with different weight percent from 0% to 30% and the effect of different B4C contents on mechanical and thermal neutron attenuation properties of TPNR composites has been studied. The phase formation in composites was analyzed using XRD technique. From the results, it showed that the incorporation of B4C fillers into TPNR matrix has enhanced the macroscopic cross section of the composites, however it lessens the tensile strength. Macroscopic cross section of the composites were increased from 3.34 cm-1 to 14.8 cm-1, while the tensile strength of the composites decreased from 3.79 MPa to 1.06 MPa with increasing B4C from 0 wt% to 30 wt%. B4C diffraction peaks were also increased in intensity with increasing B4C content.