Radiation Evaluation Methodology for Concrete Structures (NUREG/CR-7281 ORNL/SPR-2020/1572)

On this page:

• Publication Information
• Abstract

Download complete document

• NUREG/CR-7281 (PDF – 78.64 MB)

Publication Information

Manuscript Completed: December 2020
Date Published: July 2021

Prepared by:
J. Risner
A. Alpan
J. Yang (formerly with ORNL)

Oak Ridge National Laboratory
Oak Ridge, TN 37831-6283

Madhumita Sircar, NRC Project Manager

Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington DC 20555-0001

Availability Notice

Abstract

The ability to accurately calculate neutron fluence levels, radiation heating rates (neutron and gamma), and gamma dose rates in the concrete biological shields (bioshields) of light water reactors has become increasingly important as plant life extensions of 60 years have become common and extensions to 80 years of operation are proceeding This study evaluates the impact of concrete composition, the size and location of reinforcing steel, the presence of a bioshield liner or reflective thermal insulation, and the size of the reactor cavity gap on neutron flux, total heating rate (the rate of radiation energy deposition), and gamma dose rate using parametric studies with a representative three-loop pressurized water reactor (PWR) model. The analyses are performed using state-of-the-art hybrid radiation transport calculations, which provide the ability to explicitly model complex geometries and eliminate the discretization effects in space, energy, and angle that are present in the  commonly used discrete ordinates transport methodology. The results of these analyses provide insights into the effect of material and geometrical variations in the representative PWR model on the magnitude of the neutron flux, heating rate, and gamma dose rate incident to the bioshield as well as their attenuation through the reinforced concrete structure.