This open access book is a guide to using the Monte Carlo N-Particle (MCNP) code for nuclear safeguarding and non-proliferation applications.

Neutron coincidence and multiplicity counting is a standard technique used to measure uranium and plutonium masses in unknown samples for nuclear safeguards purposes, but background sources of radiation can obscure the results. In particular, high energy cosmic rays can produce large coincidence count contributions. Since some of the events occur in the sample itself, it is impossible to measure the background separately. This effect greatly increases the limit of detection of some low level neutron coincidence counting applications.The cosmic ray capability of MCNP6 was used to calculate the expected coincidence rates from cosmic rays for different sample configurations and experimental measurements were conducted for comparison. Uranium enriched to 66%, lead bricks, and an empty detector were measured in the mini Epithermal Neutron Multiplicity Counter, and MCNP6 simulations were made of the same measurements. The results show that the capability is adequate for predicting the expected background rates.Additional verification of MCNP6 was given by comparison of particle production rates to other publications, increasing confidence in MCNP6's use as a tool to lower the limit of detection. MCNP6 was then used to find particle and source information that would be difficult to detect experimentally. The coincidence count contribution was broken down by particle type for singles, doubles, and triples rates. The coincidence count contribution was broken down by source, from(a, n), spontaneous fission, and cosmic rays, for each multiplicity. (C) 2016 Elsevier B.V. All rights reserved.

The globally-averaged concentrations of radioelements K and Th within Vesta's regolith are determined from gamma ray spectra acquired by Dawn's Gamma Ray and Neutron Detector (GRaND). Spectra measured by GRaND's bismuth germanate (BGO) scintillator, while in close proximity to Vesta, are analyzed. Improvements in data reduction and analysis methods enable detection and quantification of K and Th. Ample precision is achieved using the entire data set acquired by Dawn during 5 months of low-altitude operations. A simple, analytic model, which can be applied to measurements of Vesta and Ceres, is used to determine radioelement concentrations from measured counting rates. Systematic errors in the analysis are evaluated using simulated gamma ray spectra for representative vestan meteorite compositions. Concentrations of K and Th within Vesta's global regolith, measured by GRaND, are consistent with eucrite-rich howardite, and are distinct from most achondrites, all chondrites, and Mars meteorites. The K/Th ratio of Vesta (900 +/- 400) is similar to the average ratio for howardite (approximately 1200). These radioelement data, along with major element ratios determined by nuclear spectroscopy, strongly support the hypothesis that Vesta is the parent body of the HEDs. The depletion of moderately-volatile elements implied by the measured K/Th ratio is consistent with early accretion of Vesta from a hot, incompletely condensed solar nebula and/or, less likely, subsequent removal of volatiles by energetic collisions or degassing of magmas. (C) 2015 The Authors. Published by Elsevier Inc.

MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of these two computer codes. MCNP6 is the result of six years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory’s X Computational Physics Division, Monte Carlo Codes Group (XCP-3) and Nuclear Engineering and Nonproliferation Division, Radiation Transport Modeling Team (NEN-5) respectively, have combined their code development efforts to produce the next evolution of MCNP. While maintenance and major bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. In fact, the initial release of MCNP6 contains numerous new features not previously found in either code. These new features are summarized in this document. Packaged with MCNP6 is also the new production release of the ENDF/B-VII.1 nuclear data files usable by MCNP. The high quality of the overall merged code, usefulness of these new features, along with the desire in the user community to start using the merged code, have led us to make the first MCNP6 production release: MCNP6 version 1. High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, our automated nightly software debugger tests, the underlying high quality nuclear and atomic databases, and significant testing by many beta testers.

Transport codes like MCNPX (Monte Carlo N-Particle eXtended) which are based primarily on experimentally derived data libraries at energies below approximately 100 MeV are reliable and predictive in transporting and producing neutrons and gammas. However, they are deficient in the microscopic sense because of the inability to produce correlated secondary particles. Neutrons will produce the average number of gammas at each collision regardless of the specific neutron reaction. Consequently, capture gammas may be produced by elastic collisions and inelastic gammas may be produced by capture. To remedy this deficiency the Cascading Gamma-ray and Multiplicity (CGM) V3.4 code has been integrated into MCNPX V2.7.0. The CGM code is based on Hauser-Feshbach and optical models. This paper details an overview of the CGM code integration into MCNPX followed by a test case highlighting the gamma spectrum and multiplicities from the CGM/MCNPX code. The test case provides gamma spectrum and multiplicities for the (n,gamma) reaction on Fe-56 using thermal neutrons. A second test case highlights gamma correlation for the (n,gamma) reaction on Fe-56 using thermal neutrons using three coincidence detectors. Published by Elsevier Ltd.

The NASA Dawn Mission s Gamma Ray and Neutron Detector (GRaND) [1] acquired mapping data during 5 months in a polar, low altitude mapping orbit (LAMO) with approx.460-km mean radius around main-belt asteroid Vesta (264-km mean radius) [2]. Neutrons and gamma rays are produced by galactic cosmic ray interactions and by the decay of natural radioelements (K, Th, U), providing information about the elemental composition of Vesta s regolith to depths of a few decimeters beneath the surface. From the data acquired in LAMO, maps of vestan neutron and gamma ray signatures were determined with a spatial resolution of approx.300 km full-width-at-half-maximum (FWHM), comparable in scale to the Rheasilvia impact basin (approx.500 km diameter). The data from Vesta encounter are available from the NASA Planetary Data System. Based on an analysis of gamma-ray spectra, Vesta s global-average regolith composition was found to be consistent with the Howardite, Eucrite, and Diogenite (HED) meteorites, reinforcing the HED-Vesta connection [2-7]. Further, an analysis of epithermal neutrons revealed variations in the abundance of hydrogen on Vesta s surface, reaching values up to 400 micro-g/g [2]. The association of high concentrations of hydrogen with equatorial, low-albedo surface regions indicated exogenic delivery of hydrogen by the infall of carbonaceous chondrite (CC) materials. This finding was buttressed by the presence of minimally-altered CC clasts in howardites, with inferred bulk hydrogen abundances similar to that found by GRaND, and by studies using data from Dawn s Framing Camera (FC) and VIR instruments [8-10]. In addition, from an analysis of neutron absorption, spatial-variations in the abundance of elements other than hydrogen were detected [2].

Monte Carlo simulations are increasingly used to reconstruct dose distributions in radiotherapy research studies. Many studies have used the MCNPX Monte Carlo code with a mesh tally for dose reconstructions. However, when the number of voxels in the simulated patient anatomy is large, the computation time for a mesh tally can become prohibitively long. The purpose of this work was to test the feasibility of using lattice tally instead of mesh tally for whole-body dose reconstructions. We did this by comparing the dosimetric accuracy and computation time of lattice tallies with those of mesh tallies for craniospinal proton irradiation. The two tally methods generated nearly identical dosimetric results, within 1% in dose and within 1 mm distance-to-agreement for 99% of the voxels. For a typical craniospinal proton treatment field, simulation speed was 4 to 17 times faster using the lattice tally than using the mesh tally, depending on the numbers of proton histories and voxels. We conclude that the lattice tally is an acceptable substitute for the mesh tally in dose reconstruction, making it a suitable potential candidate for clinical treatment planning.

With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

Ever since there has been spent fuel (SF), researchers have made nondestructive assay (NDA) measurements of that fuel to learn about its content. In general these measurements have focused on the simplest signatures (passive photon and total neutron emission) and the analysis has often focused on diversion detection and on determining properties such as burnup (BU) and cooling time (CT). Because of shortcomings in current analysis methods, inspectorates and policy makers are interested in improving the state-of-the-art in SF NDA. For this reason the U.S. Department of Energy, through the Next Generation Safeguards Initiative (NGSI), targeted the determination of elemental Pu mass in SF as a technical goal. As part of this research effort, 14 nondestructive assay techniques were studied . This wide range of techniques was selected to allow flexibility for the various needs of the safeguards inspectorates and to prepare for the likely integration of one or more techniques having complementary features. In the course of researching this broad range of NDA techniques, several cross-cutting issues were. This paper will describe some common issues and insights. In particular we will describe the following: (1) the role of neutron absorbers with emphasis on how these absorbers vary in SFmore » as a function of initial enrichment, BU and CT; (2) the need to partition the measured signal among different isotopic sources; and (3) the importance of the “first generation” concept which indicates the spatial location from which the signal originates as well as the isotopic origins.« less

MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, but it is much more than the sum of those two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Decision Applications Division, Radiation Transport and Applications Team (D-5), respectively, have combined their code development efforts to produce the next evolution of MCNP. While maintenance and bug fixes will continue for MCNP5 1.60 and MCNPX 2.7.0 for upcoming years, new code development capabilities only will be developed and released in MCNP6. In fact, the initial release of MCNP6 contains 16 new features not previously found in either code. These new features include the abilities to import unstructured mesh geometries from the finite element code Abaqus, to transport photons down to 1.0 eV, to transport electrons down to 10.0 eV, to model complete atomic relaxation emissions, and to generate or read mesh geometries for use with the LANL discrete ordinates code Partisn. The first release of MCNP6, MCNP6 Beta 2, is now available through the Radiation Safety Information Computational Center, and the first production release is expected in calendar year 2012. High confidence in the MCNP6 code is based on its performance with the verification and validation test suites, comparisons to its predecessor codes, the regression test suite, its code development process, and the underlying high-quality nuclear and atomic databases.

The first MCNP based in-line Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerfulmore » capability as well as dictate a path forward for future development to improve the usefulness of the technology. (authors)« less

The existing Monte Carlo N-Particle (MCNPX) particle tracking (PTRAC) coincidence capture file allows a full list of neutron capture events to be recorded in any simulated detection medium. The originating event history number (e.g. spontaneous fission events), capture time, location and source particle number are tracked and output to file for post-processing. We have developed a new MCNPX PTRAC coincidence capture file capability to aid detector design studies. New features include the ability to track the nuclides that emitted the detected neutrons as well as induced fission chains in mixed samples before detection (both generation number and nuclide that underwent induced fission). Here, the power of this tool is demonstrated using a detector design developed for the non-destructive assay (NDA) of spent nuclear fuel. Individual capture time distributions have been generated for neutrons originating from Curium-244 source spontaneous fission events and induced fission events in fissile nuclides of interest: namely Plutonium-239, Plutonium-241, and Uranium-235. Through this capability, a full picture for the attribution of neutron capture events in the detector can be simulated.

The angular distribution of scattered photons is incorrect in MCNPX and MCNP5 because the incoherent and coherent form factors are obsolete. The obsolete data affect all photon transport problems with E > 74 keV. Elastic backscatter for E > 105 keV is completely missing. Consequently, a new ACE-format photoatomic data library, tentatively named MCPLIB05 and referred to herein as MCPLIB05T, has been developed for MCNP/X. Data in MCPLIB05T other than form factors are identical to that in its predecessor photoatomic library, MCPLIB04. The new form factor data in MCPLIB05T come directly from ENDF/B-VII (rev. 0) and are in a format incompatible with older versions of MCNP/X. Consequently, a new version of MCNP/X has been developed to identify and use the new MCPLIB05T data and yet retain backward compatibility, including tracking, when MCPLIB04 is used. The NJOY nuclear data processing system is undergoing development to enable future generations of photoatomic data libraries with modern form factor data in the new format.

This paper describes the techniques that are available to calculate the performance of 3He alternative detectors using MCNPX. Calculations of the performance of safeguards detectors that use 3He have been successfully carried out for many years. In the case of coincidence or multiplicity counting, specific tallies have been implemented to calculate the Singles, Doubles and Triples counting rates. The implementation of the method was done in such a way that it equates every capture in some nuclide in the detection zone with the production of an electronic pulse from the detector. This is a very good approximation for 3He detectors and BF3 detectors. However it is not appropriate for detectors such as boron-lined proportional counters, in which the fraction of capture events leading to an electronic pulse above threshold is very dependent on the geometric arrangement, in particular the thickness and composition of the boron-containing layer. This paper gives calculations of the ideal pulse height distributions to be expected from different detector types and gives values for the probability, as a function of detector energy threshold, that a neutron capture reaction will cause an electronic pulse from the detector. This is termed the electronic efficiency. This electronic detection efficiency depends very little on the energy of the captured neutron, which in most practical cases are heavily weighted towards thermal energies. It does not depend on the position of the source neutron or moderation in the sample. For cases of interest to nuclear safeguards, measurement of uranium and plutonium in specially designed detectors, the spectrum of detected neutrons is fairly constant and thus the electronic detection efficiency becomes a detector constant. The paper discusses how the electronic detection efficiency needs to be included in the calculation of Singles, Doubles and Triples, and describes proposals to improve the tallying capability of MCNPX for such cases.

high mobility and stability. These models were implemented in the MARMOT phase field code, which is used to calculate effective Xe diffusivities for various irradiation conditions. The effective diffusivities were used in BISON to calculate fission gas release for a number of test cases. The results are assessed against experimental data and future directions for research are outlined based on the conclusions.