International Journal of Advanced Engineering Application

B₄C-Polyethylene Composite Shielding Material Design, MCNP6 Monte Carlo Simulation and Reactor Safety Transient Analysis

Author(s):Sanjiv Kumar, MeeraTripathi, Rahul Misra

Affiliation: Department of Nuclear Engineering and Technology, Banaras Hindu University, Varanasi, Uttar Pradesh, India

Page No: 91-94

Volume issue & Publishing Year: Volume 3, Issue 4, 2026/04/16

Journal: International Journal of Advanced Engineering Application (IJAEA)

ISSN NO: 3048-6807

DOI: https://doi.org/10.5281/zenodo.19703475

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Abstract:
India's nuclear power programme, comprising 22 operating reactors (totalling 6,780 MWe) with 8 additional reactors under construction including two 700 MWe PHWRs at Kakrapar and Gorakhpur, faces evolving radiation shielding requirements driven by life extension of existing 220 MWe PHWRs beyond their original 25-year design life, the commissioning of higher-power 700 MWe units with increased neutron fluence environments, and the Atomic Energy Regulatory Board's (AERB) updated radiation protection standards aligned with ICRP Publication 103 recommendations. The occupational dose limit reduction from 30 mSv/year to 20 mSv/year under the revised AERB Safety Code requires shielding upgrades at existing units and optimised shielding design for new construction. This paper presents the design, characterisation, and Monte Carlo N-Particle (MCNP6) simulation validation of a B₄C-polyethylene nanocomposite shielding material incorporating 20 wt% B₄C microparticles (10 µm mean diameter) and 5 wt% high-density polyethylene (HDPE) matrix with nanoclay compatibiliser, targeting simultaneous attenuation of both gamma radiation (using polyethylene's hydrogen content for neutron thermalisation and B-10 neutron capture) and fast neutrons. Mass attenuation coefficients are measured by gamma transmission experiments and compared with NIST XCOM database values. Neutron moderation and capture cross-sections are evaluated by activation analysis. MCNP6 simulations of a representative PHWR bioshield geometry validate the composite material's performance against heavy concrete and standard polyethylene benchmarks. A reactor safety transient analysis — control rod ejection with and without SCRAM activation — is performed using the RELAP5-3D thermal-hydraulic code coupled to MCNP6 neutronics to establish the design basis accident response of the proposed shielding configuration.

Keywords: radiation shielding, B₄C, HDPE, MCNP6, neutron, gamma, PHWR, nuclear safety, RELAP5, AERB, Monte Carlo, dose rate, bioshield, India

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