Potential MCNP enhancements for NCT
MCNP a Monte Carlo radiation transport code, is currently widely used in the medical community for a variety of purposes including treatment planning, diagnostics, beam design, tomographic studies, and radiation protection. This is particularly true in the Neutron Capture Therapy (NCT) community. The current widespread medical use of MCNP after its general public distribution in about 1980 attests to the code's general versatility and usefulness, particularly since its development to date has not been influenced by medical applications. This paper discusses enhancements to MCNP that could be implemented at Los Alamos for the benefit of the NCT community. These enhancements generally fall into two categories, namely those that have already been developed to some extent but are not yet publicly available, and those that seem both needed based on our current understanding of NCT goals, and achievable based on our working knowledge of the MCNP code. MCNP is a general, coupled neutron/photon/electron Monte Carlo code developed and maintained by the Radiation Transport Group at Los Alamos. It has been used extensively for radiation shielding studies, reactor analysis, detector design, physics experiment interpretation, oil and gas well logging, radiation protection studies, accelerator design, etc. over the years. MCNP is a three-dimensional geometry, continuous energy physics code capable of modeling complex geometries, specifying material regions such as organs by the intersections of analytical surfaces.
- Research Organization:
- Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
- Sponsoring Organization:
- USDOE; USDOE, Washington, DC (United States)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 6899826
- Report Number(s):
- LA-UR-92-3587; CONF-9209280-1; ON: DE93003807
- Resource Relation:
- Conference: Neutron capture therapy for cancer, Columbus, OH (United States), 13-17 Sep 1992
- Country of Publication:
- United States
- Language:
- English
Similar Records
MCNP (Monte Carlo Neutron Photon) capabilities for nuclear well logging calculations
Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code
Related Subjects
NEUTRON CAPTURE THERAPY
DOSIMETRY
MONTE CARLO METHOD
OPTIMIZATION
PHANTOMS
PLANNING
CALCULATION METHODS
MEDICINE
MOCKUP
NEUTRON THERAPY
NUCLEAR MEDICINE
RADIOLOGY
RADIOTHERAPY
STRUCTURAL MODELS
THERAPY
560101* - Biomedical Sciences
Applied Studies- Radiation Effects- Dosimetry & Monitoring- (1992-)