MCNP

The moderator material and optimum sizes, the extraction channel configuration, as well as the converter and reflector material and sizes, were determined using the MCNP-4B code and the steepest descent method to attain a maximum flux density of thenual neutrons at the position of an object to be studied. The calculated data were experimentally verified, which showed good agreement. 

J. F. Biresmeister, MCNP - A General Monte-Carlo N-Particle Transport Code, Version 4B, LA-12625-M, 1997. 

A radioisotope battery is one of the choice for energy source of meteorological obseiwation and development of undersea and space[l]. We have considered a strontium-90 (half-life 28.8y) heat-source model of a radioisotope battery and improved it in two aspects—radiation dose reduction and improvement of thermal conductivity—adding graded structure to the model[2]. The present study reports the dose reduction of bremsstrahlung photons from -ray of - "Sr and its daughter nuclide yttrium-90. The calculation was carried out by a continuous energy Monte Carlo code, MCNP 4A[3]. 

We performed two kinds of calculation using MCNP 4A code. One was simplified calculation. Its source was set to 2.245 MeV as monoenergy and cutoff energy for electron set to 2.24 MeV. The calciilation could save time and search for the tendency of the dose variation to the inner and outer radiuses of a mixed layer. 

The possibility of absorbed dose reduction for the heat source model of a Sr radioisotope battery was studied by adding graded structure to the model using MCNP 4A code. The results showed 17, 19 and 22 % reduction to the dose of the base structure for three, four and five layers, respectively. 

B riesmeister, J. F. 1990. MCNP-A general Monte Carlo code for neutron and photon transport. Version 4.2. Technical Report LA-7396-M, Los Alamos National Laboratory. 

To examine its viability, a variety of parametric scans were performed using MCNP-5 [LANL, 2003] on a model of the proposed core. The core was designed with sufficient reactivity to ensure 10 years of full-power operation even when including such negative feedback coefficients as temperature, temperature based expansion, and bumup. The neutronics runs done include ... 

Much attention is being paid to the generation of nuclear data. For the PANTHER-THERMIX system, neutron cross sections are generated by WIMS-E and SCALE-4 codes. The Monte-Carlo code MCNP is used for tiiis as well, and to check certain reactor calculations. 

The validation effort was pursued in dose cooperation with CEA Cadarache. Its aims were twofold comparison between our deterministic route (MICROX-2/TWODANT) and the Monte Carlo (MCNP-4A) one, on the one side, and support for the validation effort on the European code system ECCO/ERANOS, on the other side. The work concentrated on three numerical benchmarks derived from the ZONA-2 series of the CIRANO experimental program (performed in the zero power facility MASURCA at Cadarache). For detailed results, see references [5, 6, 7j here only a brief summary of the main findings is given. 

The influence of the basic data was studied with the stochastic route (MCNP-4A employed continues energy data based on ENDF/B-V, ENDF/B-VI and JEF-2.2). The deterministic results were obtained for JEF-2.2 data. 

The Monte Carlo N-Particle (MCNP) (version 4C2) code was used for most of the neutronics analyses by both organizations. 

Calculations of core cell bum-up in the N2 PWR were performed by the spectral code CETERA [16]. The fission product and actinide activities were estimated using the RECOL [17] library data base, which was generated on the bases of the latest versions of the evaluated nuclear data files, ENDF/B-V, with corrections based on the results of critical experiments [18]. The criticality problem was solved for a realistic 3-D geometry model of a TFC by Monte-Carlo with RECOL and checked with MCNP [19] for fresh fuel load. One-group cross-sections were prepared for bum-up calculation of critical loads of both fresh and spent fuel and input to ORIGEN-2 [20] for detailed radionuclide content calculations. 

In general the three-dimensional Monte Carlo transport codes MCNP-4A, -4B, -4C are used to predict the measured spectra Point-wise cross sections from FENDL/E-1 are used mainly in... 

In large samples, the y-ray count rate of a PGNAA system is a multivariable function of the elemental dry composition, density, water, contents, and thickness of the material. The experimental calibration curves require tremendous laboratory work using a large number of standards with well-known compositions. The Monte Carlo simulation code MCNP helps to reduce the experimental standards as described by Oliveira et al. (1997) in their attempt to optimize the PGNAA instrument design for cement raw materials. 

Thermal Reactor Benchmark Calculations with the MCNP Monte Carlo Code, Richard E. Prael(LASL)... 

A number of thermal reactor benchmark calculations have been performed using the continuous-energy Monte Carlo code MCNP with ENDF/B-IV data. The principal motivations for these calculations were ... 

LASL Group TD-6, MCNP-A General Monte Carlo Code for Neutron and Photon Transport, LA-7396-M, Los Alamos Scientific Lab. (July 1978). 

Homogeneous Thermal Benchmark Calculations with ENDF/B-V Data Using MCNP,... 

X-5 Monte Carlo Team. 2003. MCNP—A General Monte Carlo N-particle Transport Code, Version 5 Volume IT. User s guide. Los Alamos, NM LA-GP-03-0245 Los Alamos National Laboratory (revised February 1,2008). 

The idea of reducing the detector to a point is the basis of the virtual point-detector concept. The importance of this concept is that it allows approximations to be made that simplify, what would otherwise be, complicated mathematical calculations. Mahling et al (2006) took the dimensions of 49 actual detectors and, using the Monte Carlo program MCNP, were able to derive an empirical equation for calculating d based upon the radius and height of the detector and two energy dependent parameters, which have been quantified. The concept has been explored for planar and semi-planar detectors (Alfassi et al, 2006). The rather curious conclusion was that the concept is valid, but the virtual point for small detectors can be outside the physical dimensions of the detector. 

A common program used for the calculation of efficiencies in volume sources is MCNP-Monte Carlo N-Particle... 

Physical dimensions of the detector, including such things as detector-to-cap distance. (These are needed when assessing summing, variation of count rate with sample-to-detector distance and for setting up MCNP and similar catibration models.)... 

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