MCNPX vs. DORT for SNS shielding design studies.

Area of Science:

• Nuclear Engineering
• Radiation Physics
• Accelerator Technology

Background:

• Radiation transport is a critical consideration in accelerator facility design.
• Shielding effectiveness must be validated to ensure operational safety and minimize environmental impact.
• The Spallation Neutron Source (SNS) requires robust shielding solutions for its access ways and tunnels.

Purpose of the Study:

• To analyze radiation transport through an 18m access way and a 2.2m thick shielding door at the SNS.
• To evaluate the performance of typical accelerator shielding materials like concrete and steel.
• To compare the accuracy and agreement of Monte Carlo (MCNPX) and discrete ordinates (DORT) methods for radiation shielding calculations.

Main Methods:

• Utilized MCNPX code for Monte Carlo simulations of radiation transport.
• Employed DORT code for discrete ordinates calculations.
• Input radiation sources were derived from beam losses during accelerator operation.

Main Results:

• Radiation penetration was analyzed through various shielding materials including concrete and steel.
• Simulations covered an 18m access way and a 2.2m massive shielding door.
• Results from MCNPX and DORT methods demonstrated good agreement.

Conclusions:

• Both Monte Carlo and discrete ordinates methods provide reliable predictions for radiation transport in accelerator shielding.
• The study validates the use of these computational methods for designing effective shielding for facilities like the SNS.
• Findings support the use of conventional materials like concrete and steel for robust accelerator shielding.