Full core analysis of IRIS reactor by using MCNPX.

Area of Science:

• Nuclear Engineering
• Computational Physics

Background:

• Accurate neutronic analysis is crucial for nuclear reactor safety and efficiency.
• The International Reactor Innovative and Secure (IRIS) reactor design requires detailed neutronic assessment.
• Verification of computational methods against established codes is essential for reliable simulations.

Purpose of the Study:

• To conduct a comprehensive neutronic analysis of a fresh-fueled IRIS reactor using the MCNPX code.
• To validate MCNPX results by comparing criticality and power distribution calculations with established codes.
• To investigate the impact of Boron-10 enrichment on the effective multiplication factor (k-eff) using WIMS-D5.

Main Methods:

• Neutronic analysis utilizing the MCNPX Monte Carlo code.
• Criticality calculations, including effective multiplication factor (k-eff) determination.
• Assessment of radial and axial power distributions, nuclear peaking factor, and axial offset.
• Comparison of MCNPX results with HELIOS/NESTLE, CASMO/SIMULATE, CORD-2, and SAS2H/KENO-V.
• Simulation of Boron-10 enrichment effects using WIMS-D5 code.

Main Results:

• The MCNPX code yielded an effective multiplication factor (k-eff) value closely matching that from modified CORD-2 calculations.
• Radial and axial power distributions calculated by MCNPX showed good agreement with published SAS2H/KENO-V results.
• WIMS-D5 analysis indicated the effect of Boron-10 enrichment on the fuel pin's k-eff at various burnup stages, consistent with HELIOS code findings.

Conclusions:

• MCNPX is a reliable tool for neutronic analysis of the IRIS reactor, with results validated against multiple established codes.
• The study provides a benchmark for future neutronic assessments of the IRIS reactor design.
• Understanding the impact of neutron absorbers like Boron-10 is vital for reactor core management and safety.