2012
DOI: 10.13182/nt12-a15345
|View full text |Cite
|
Sign up to set email alerts
|

Foreword: Special Issue on the Initial Release of MCNP6

Help me understand this report

Search citation statements

Order By: Relevance

Paper Sections

Select...
2
2
1

Citation Types

1
62
0

Year Published

2014
2014
2023
2023

Publication Types

Select...
7
1

Relationship

0
8

Authors

Journals

citations
Cited by 45 publications
(63 citation statements)
references
References 0 publications
1
62
0
Order By: Relevance
“…Errors are propagated from imperfect knowledge of the reactor thermal power, and from the extrapolation of this power into the associated anti-neutrino spectrum. Reactor operators are able to provide precision measurements of the thermal output power, as well as estimates (based on simulation with the code MCNP [37]) of the isotopic fuel composition and fission fractions f i /F , where f i is the absolute fission rate of species i and F ≡ f i . Uncertainty estimates on the order of 2-3% are typical, although it may be possible to reduce this to around a half of a percent (cf.…”
Section: -100 Evmentioning
confidence: 99%
“…Errors are propagated from imperfect knowledge of the reactor thermal power, and from the extrapolation of this power into the associated anti-neutrino spectrum. Reactor operators are able to provide precision measurements of the thermal output power, as well as estimates (based on simulation with the code MCNP [37]) of the isotopic fuel composition and fission fractions f i /F , where f i is the absolute fission rate of species i and F ≡ f i . Uncertainty estimates on the order of 2-3% are typical, although it may be possible to reduce this to around a half of a percent (cf.…”
Section: -100 Evmentioning
confidence: 99%
“…The multi-particle Monte Carlo transport code MCNP6 [5] was used to simulate the proton-induced nuclear interactions and the subsequent transport of secondary particles. The major advantage of using MCNP6 in this study is the accuracy of the underlying cross sections for proton and neutron transport.…”
Section: Calculation Methodsmentioning
confidence: 99%
“…2, we considered two different geometries to generate shielding parameters, without and with the inclusion of the partition wall in the model. A coupling technique called the surface source write and read (SSW/SSR) in MCNP6 was implemented on the exterior surfaces of the BSA to reduce the computing burden of source generation in repeated Monte Carlo shielding simulations [5]. Note that the shielding parameters generated in this way were obviously the BSA dependent and intended only for use in shielding analyses of this facility.…”
Section: Calculation Methodsmentioning
confidence: 99%
“…MCNP6 [1] is used in various applications involving fission reactions at low, but also at intermediate and high energies. It is critical that it describes such reactions as well as possible, therefore it is often validated and verified against available experimental data and calculations by other models (see, e.g., [1,2] and refrences therein).…”
Section: Introductionmentioning
confidence: 99%