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This paper describes the use of the Monte Carlo code MCNPX-PoliMi for nuclear-nonproliferation applications, with particular emphasis on the simulation of spontaneous and neutron-induced nuclear fission. New models for the outgoing neutrons and gamma rays emitted in spontaneous and induced fission are described. For spontaneous fission, the models include prompt neutron energy distributions that depend on the number of neutrons emitted in the individual fission events. For neutron-induced fission, due to lack of data, the prompt neutron energy distributions are independent of the number of neutrons emitted in the individual fission events. Gamma rays are sampled independently of the neutrons. Code validation is performed on well-characterized mixed-oxide fuel and plutonium-oxide samples.
In the past few years, efforts to develop new measurement systems to support nuclear nonproliferation and homeland security have increased substantially. Monte Carlo radiation transport is one of the simulation methods of choice for the analysis of data from existing systems and for the design of new measurement systems; it allows for accurate description of geometries, detailed modeling of particle–nucleus interactions, and event-by-event detection analysis.
This paper describes the use of the Monte Carlo code MCNPX-PoliMi for nuclear-nonproliferation applications, with particular emphasis on the simulation of spontaneous and neutron-induced nuclear fission. In fact, of all possible neutron–nucleus interactions, neutron-induced fission is the most defining characteristic of special nuclear materials (such as 235U and 239Pu), which is the material of interest in nuclear-nonproliferation applications. The MCNP-PoliMi code was originally released from the Radiation Safety Shielding Center (RSSIC) at Oak Ridge National Laboratory in 2003 [1]. Original motivations for the code development include the ability to perform simulations of pulse height tallies of neutron interactions in existing and new organic scintillators, time-dependent cross-correlation measurements, and neutron and gamma ray multiplicity measurements.
The MCNPX-PoliMi code contains many enhancements and is based on MCNPX ver. 2.7.0. The new code includes the ability to model the neutrons and gamma rays emitted from photonuclear interactions, contains new built-in fission sources, and new models for spontaneous and induced fission emissions. This paper focuses on the latter capability and describes code validation with comparisons with experiments performed on 252Cf and bulk fissile material using liquid scintillators and helium-3 systems. MCNPX-PoliMi ver. 2.0 was released through RSICC in 2012 as a patch to MCNPX ver. 2.7.0 and as an executable [2].
The Monte Carlo modeling of the neutron and gamma-ray emissions from spontaneous- and induced-fission events introduces many approximations to the complex physical processes that occur during nuclear fission. The modeling requires the knowledge of the number, spectral, directional, and temporal distributions of neutrons and gamma rays of the various nuclides being modeled. Early models of fission did not include the full description of the multiplicity (i.e., the number) distributions of
The validation of the microscopic models of spontaneous and induced fission is a complex procedure that requires extensive comparisons of simulations to carefully planned and executed experiments [12], [13]. Many of the quantities mentioned in Section 2 are not directly observable in typical experimental procedures.
252Cf is frequently used for code validation because this type of source is readily available and can be used in a university laboratory setting [14]. However, most radioisotope
MCNPX-PoliMi can be used to model neutron and gamma emissions of special nuclear materials and various isotopic sources. We have implemented in MCNPX-PoliMi ver. 2.0 new models of neutron and gamma-ray emissions from spontaneous and induced fission for several isotopes of interest. These models include the multiplicity, energy spectrum, and angular distributions of neutrons emitted from spontaneous and induced fission. We have validated these models with experiments using (a) liquid
This research was funded in part by the National Science Foundation and the Domestic Nuclear Detection Office of the Department of Homeland Security through the Academic Research Initiative Award # CMMI 0938909.
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Because TNI can theoretically estimate an unknown object’s geometry and total uranium mass, corrections to the point kinetics model could be implemented using TNI measurement information to correct for this assay bias [1,10].Monte Carlo simulations using MCNPX-PoliMi v2.0.0 [11] were used to simulate neutron fission chain propagation in uranium metal under TNI.MCNPX-PoliMi was used due to its improved modeling of correlations between neutron energy and fission multiplicity and convenient output files that track collisions and fission multiplicities.
The shielding modifies the intensity and the energy spectrum of the neutrons that impinge on the radiation detectors.The detectors are simulated using the Monte Carlo code MCNPX-PoliMi [31,32] and post-processing code MPPost [33] to evaluate our ability to simulate these scenarios.Code validation is performed by comparing experimental and simulation results for each of the scenarios.
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