title: MCNPX vs. DORT for SNS shielding design studies.
abstract: Radiation transport occurs through the 18 m long access way adjacent to the Spallation Neutron Source accelerator tunnel and the 2.2 m thick massive shielding door which closes the access way. A variety of typical materials for accelerator shielding, such as concrete and steel, were used for construction of the door to study radiation penetration. A comparison was carried out using both Monte Carlo (code MCNPX) and discrete ordinates (code DORT) methods. The beam losses during the accelerator operation are the sources for the radiation calculations. Analyses show that the results from the two methods are in good agreement.
authors.affiliation: Oak Ridge National Laboratory, Spallation Neutron Source 701 Scarboro Road, Oak Ridge, TN 37830, USA. popovai@ornl.gov
authors.forename: Irina I
authors.initials: II
authors.lastname: Popova
mesh_terms: D003198:Computer Simulation; D017076:Computer-Aided Design; D005160:Facility Design and Construction / Q000379:methods*; D015233:Models, Statistical; D009010:Monte Carlo Method; D009502:Neutrons; D010315:Particle Accelerators / Q000295:instrumentation*; D011829:Radiation Dosage; D011834:Radiation Monitoring / Q000295:instrumentation / Q000379:methods*; D011835:Radiation Protection / Q000295:instrumentation* / Q000379:methods; D018570:Risk Assessment / Q000379:methods; D012307:Risk Factors; D012984:Software; D012986:Software Validation; D013714:Tennessee