Papers using OpenMC

Announcements
124

Preliminary Neutronics Analysis for Generation III+ VVER Design - VVER-1200

1 Like
125

Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code

1 Like
126

The Random Ray Method Versus Multigroup Monte Carlo: The Method of Characteristics in OpenMC and SCONE

1 Like
127

Reactivity control optimization of boron-free small modular pressurized water reactor with helical-cruciform metallic fuel

1 Like
128

Cross-Evaluation of the PU-MET-FAST criticality benchmark experiments in ICSBEP handbook (by the end of 2022)

1 Like
129

Fundamental Design Considerations for Nuclear Space Power Systems

1 Like
130

An alternative approach for excess reactivity control of TRISO particles

1 Like
131

Comparison and evaluation of resolved resonance region covariance representations

132

Assessing shielding material Performance: Benchmarking of Monte Carlo codes for oil shale and Basalt-Boron fiber concretes

133

Neutronic analysis of the BYU molten salt micro reactor

1 Like
134

OpenMC-based high-fidelity model for the Moroccan TRIGA Mark II research reactor

135

Prospects for a high-field, compact break-even axisymmetric mirror (BEAM) and applications

136

Some recent cool work by @nuclearbae and friends:
3D high-fidelity automated neutronics guided optimization of fusion blanket designs

1 Like
137

Improvement of LLFPs transmutation efficiency by moderator loading in lead-based fast neutron reactor

138

Non-uniform fuel distribution and thermo-mechanical analysis of a 1 MW thermal power micronuclear heat pipe reactor

1 Like
139

Irregular, Backward-Compatible PWR Assembles with More Pins to Facilitate Uprates

1 Like
140

Modeling of the Advanced Test Reactor Using OpenMC,
Cubit, and Griffin

1 Like
141

Optimizing the Fixed Number Detector Placement for the Nuclear Reactor Core Using Reinforcement Learning

Core design and reactivity control optimization of small lead-cooled reactor LFR-180

Towards inferring reactor operations from high-level waste

ERSN-OpenMC-Py: A python-based open-source software for OpenMC Monte Carlo code

Multiphysics analysis of a metal hydride moderated megawatt heat pipe reactor with burnable poisons

1 Like
142

Pre-Experimental Assessment of Uranyl Nitrate Solution Irradiation in Kartini Reactor Facilities

143

Angular distribution uncertainty influence in a large sodium-cooled fast reactor with mixed-oxide fuel