Results 31 to 40 of about 10,040 (154)
Frontiers in Energy Research, 2022 A conceptual design for a 100 MW(th) organic cooled reactor is proposed with the application of organic fluid HB-40 as a coolant. In order to obtain the axial and radial power distribution, a physical model for the proposed core is developed using the ...
Feng Wang +7 more
doaj +1 more source
Photodissociation of p-process nuclei studied by bremsstrahlung induced activation
, 2005 A research program has been started to study experimentally the near-threshold photodissociation of nuclides in the chain of cosmic heavy element production with bremsstrahlung from the ELBE accelerator. An important prerequisite for such studies is good
A. R. Junghans +22 more
core +1 more source
, 2016 The MCNP5 code was employed to simulate the {\gamma}ray shielding capacity of tungstate composites. The experimental results were applied to verify the applicability of the Monte Carlo program.
Li, Yingjun +5 more
core +1 more source
Current status of MCNP6 as a simulation tool useful for space and accelerator applications
, 2012 For the past several years, a major effort has been undertaken at Los Alamos National Laboratory (LANL) to develop the transport code MCNP6, the latest LANL Monte-Carlo transport code representing a merger and improvement of MCNP5 and MCNPX. We emphasize
Bull, J. S. +4 more
core +1 more source
Nuclear Engineering and Technology, 2020 The control rod drive mechanism needs to be debugged after reactor fresh fuel loading. It is of great importance to monitor the subcriticality of this process accurately.
Wencong Wang, Caixue Liu, Liyuan Huang
doaj +1 more source
Nuclear Engineering and Technology, 2021 A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations.
Tanja Goričanec +4 more
doaj +1 more source
Benchmark study of TRIPOLI-4 through experiment and MCNP codes [PDF]
2011 2nd International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and their Applications, 2011 Reliability on simulation results is essential in nuclear physics. Although MCNP5 and MCNPX are the world widely used 3D Monte Carlo radiation transport codes, alternative Monte Carlo simulation tools exist to simulate neutral and charged particles' interactions with matter. Therefore, benchmark are required in order to validate these simulation codes.
Michel, Maugan +4 more
openaire +2 more sources
Correlated Prompt Fission Data in Transport Simulations
, 2018 Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and $\gamma$-ray~observables. Beyond simple average quantities, the study of distributions and correlations
Andrews, M. T. +17 more
core +1 more source
Vitess 3.8: a modernized framework for Monte Carlo neutron tracing simulations
Journal of Applied Crystallography, Volume 59, Issue 2, Page 678-686, April 2026.VITESS 3.8 introduces modernized neutron‐source modeling with new artificial‐intelligence‐based and KDSource modules, plus expanded component support including a prism module and NCrystal‐integrated sample module. It also delivers major upgrades to existing modules and multiple new features, advancing the simulation framework.VITESS is a modular Monte ...
José Ignacio Robledo +5 more
wiley +1 more source
Extending OpenMC validation to spent fuel canisters: A criticality benchmark against MCNP
Nuclear Engineering and TechnologyOpenMC is an open-source Monte Carlo code with increasing relevance in criticality safety and reactor physics applications. While its validation has covered a broad range of systems, its performance in spent nuclear fuel storage scenarios remains limited
J. Ruiz-Pineda +4 more
doaj +1 more source