Michael L Fensin

Initial MCNP6 Release Overview

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 MCNPX Monte Carlo Radiation Transport Code

MCNPX (Monte Carlo N‐Particle eXtended) is a general‐purpose Monte Carlo radiation transport code with three‐dimensional geometry and continuous‐energy transport of 34 particles and light ions. It contains flexible source and tally options, interactive graphics, and support for both sequential and multi‐processing computer platforms. MCNPX is based on MCNP4c and has been upgraded to most MCNP5 capabilities. MCNP is a highly stable code tracking neutrons, photons and electrons, and using evaluated nuclear data libraries for low‐energy interaction probabilities. MCNPX has extended this base to a comprehensive set of particles and light ions, with heavy ion transport in development. Models have been included to calculate interaction probabilities when libraries are not available. Recent additions focus on the time evolution of residual nuclei decay, allowing calculation of transmutation and delayed particle emission. MCNPX is now a code of great dynamic range, and the excellent neutronics capabilities allow new opportunities to simulate devices of interest to experimental particle physics, particularly calorimetry. This paper describes the capabilities of the current MCNPX version 2.6.C, and also discusses ongoing code development.

THE ENHANCEMENTS AND TESTING FOR THE MCNPX 2.6.0 DEPLETION CAPABILITY

Monte Carlo–linked depletion methods have gained recent interest due to the ability to model complex three-dimensional geometries using continuous-energy cross sections. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a completely self-contained Monte Carlo-linked depletion capability in a single Monte Carlo code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross-section data permit. The objective of this work is (a) describe the MCNPX depletion methodology dating from the original linking of MONTEBURNS and MCNP to the first public release of the integrated capability (MCNPX 2.6.B, June 2006) that has been reported previously, (b) further detail the many new depletion capability enhancements since then leading to the present Radiation Safety Information Computational Center (RSICC) release, MCNPX 2.6.0, (c) report calculation results for the H. B. Robinson benchmark, and (d) detail new features available in MCNPX 2.7.A. Each version of MCNPX depletion starting from MCNPX 2.6.A leading to the official RSICC release of MCNPX 2.6.0 and the new beta release MCNPX 2.7.A included significant upgrades that addressed key issues from earlier versions. This paper details these key issues and the approach utilized to address the issues as enhancements for MCNPX 2.6.0. The MCNPX 2.6.0 depletion capability enhancements include (a) allowing the modeling of as large a system as computer memory capacity permits; (b) tracking every fission product available in ENDF/B VII.0; (c) enabling depletion in repeated structures geometries such as repeated arrays of fuel pins; (d) including metastable isotopes in burnup; and (e) manually changing the concentrations of any isotope during any time step by specified atom fraction, weight fraction, atom density, or gram density. These enhancements allow better detail to model the true system physics as well as to improve the robustness of the capability. H. B. Robinson benchmark calculations were completed to assess the validity of nuclide predictability of MCNPX 2.6.0. The results show comparisons of key actinide and fission products as compared to experiment and the SCALE-4 SAS2H sequence 27-group cross-section library (27BURNUPLIB) results. MCNPX 2.6.0 depletion results are within 4% of the experimental results for most major actinides. Two major depletion enhancements are available in the MCNPX 2.7.A beta release: improved 63-group flux querying and parallelization of the burnup interface routines in multiprocessor mode. Fixing the energy group querying routine does correctly tally the energy flux for use with isotopes not containing transport cross sections; however, results show <1% change in nuclide prediction for the benchmark test case. MCNPX 2.7.A parallelizes the depletion interface routines and running of CINDER90 so that different burnable regions of a given depletion system can be preprocessed, burned, and postprocessed on separate slave processors. The parallelization involves minimal communication between processors and therefore leads to significant computational performance enhancement. The combination of new enhancements and testing of the MCNPX 2.6.0 depletion computational system make this capability a valuable Monte Carlo-linked depletion tool. Additional testing and feature enhancements are under development to further improve the usefulness of the computational tool.

Improved Reaction Rate Tracking and Fission Product Yield Determinations for the Monte Carlo-Linked Depletion Capability in MCNPX

Abstract As advanced reactor concepts challenge the accuracy of current modeling technologies, a higher-fidelity depletion calculation is necessary to model time-dependent core reactivity properly for accurate cycle length and safety margin determinations. The recent integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a completely self-contained Monte Carlo-linked depletion capability. Two advances have been made in the latest MCNPX capability based on problems observed in prereleased versions: continuous-energy collision density tracking and adequate fission yield selection. Prereleased versions of the MCNPX depletion code calculated the reaction rates for (n,2n), (n,3n), (n,p), and (n,α) by matching the MCNPX steady-state 63-group flux with 63-group cross sections inherent in the CINDER90 library and then collapsing to one-group collision densities for the depletion calculation. The accuracy of this procedure is therefore dictated by the adequacy of the 63-group energy structure of the cross-section set to accurately capture the spectrum of a specific model. Different types of models would therefore require different types of cross-section energy group structure. MCNPX 2.6.A eliminates this dependency by using the continuous-energy reaction rates determined during the MCNPX steady-state calculation to calculate energy-integrated collision rates to be used by CINDER90. MCNPX 2.6.A now also determines the proper fission yield to be used by the CINDER90 code for the depletion calculation. The CINDER90 code offers a thermal, fast, and high-energy fission yield for each fissile isotope contained in the CINDER90 data file. MCNPX 2.6.A determines which fission yield to use for a specified problem by calculating the integral fission rate for the defined energy boundaries (thermal, fast, and high energy), determining which energy range contains the majority of fissions, and then selecting the appropriate fission yield for the energy range containing the majority of fissions. The MCNPX depletion capability enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. This study focuses on the methodology development of the two improvements described here. Further improvements are under development to enhance the usefulness of this new capability.

Testing the Delayed Gamma Capability in MCNP6

Abstract The mission of the Domestic Nuclear Detection Office is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism. A delayed particle capability was initially added to MCNPX 2.6.A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. To meet the objectives of detection scenario modeling, the capability was designed to sample a particular time for emitting particular multiplicity of a particular energy. Because the sampling process of selecting both time and energy is interdependent, to linearize the time and emission sampling, atom densities are computed at several discrete time steps, and the time-integrated production is computed by multiplying the atom density by the decay constant and time step size to produce a cumulative distribution function for sampling the emission time, energy, and multiplicity. The delayed particle capability was initially given a time-bin structure to help reasonably reproduce, from a qualitative sense, a fission benchmark by Beddingfield, which examined the delayed gamma emission. This original benchmark was only qualitative and did not contain the magnitudes of the actual measured data but did contain relative graphical representation of the spectra. A better benchmark with measured data was later provided by Hunt, Mozin, Reedy, Selpel, and Tobin at the Idaho Accelerator Center; however, because of the complexity of the benchmark setup, sizable systematic errors were expected in the modeling, and initial results compared to MCNPX 2.7.0 showed errors outside of statistical fluctuation. Presented here is a more simplified approach to benchmarking, utilizing closed form analytic solutions to the granddaughter equations for particular sets of decay systems. We examine five different decay chains (two-stage decay to stable) and show the predictability of the MCNP6 delayed gamma feature. Results do show that while the default delayed gamma calculations available in the MCNP6 1.0 release can give accurate results for some isotopes (e.g., 137Ba), the percent differences between the closed form analytic solutions and the MCNP6 calculations were often >40% (28Mg, 28Al, 42K, 47Ca, 47Sc, 60Co). With the MCNP6 1.1 Beta release, the tenth entry on the DBCN card allows improved calculation within <5% as compared to the closed form analytic solutions for immediate parent emissions and transient equilibrium systems. While the tenth entry on the DBCN card for MCNP6 1.1 gives much better results for transient equilibrium systems and parent emissions in general, it does little to improve daughter emissions of secular equilibrium systems. Hypotheses were presented as to why daughter emissions of secular equilibrium systems might be mispredicted in some cases and not in others.

MCNPX Improvements for Threat Reduction Applications

Enhancements contained in the current MCNPX 2.6.0 Radiation Safety Information Computational Center (RSICC) release will be presented, including stopped‐muon physics, delayed neutron and photon generation, and automatic generation of source photons. Preliminary benchmarking comparisons with data taken with a muon beam at the Paul Scherrer Institute Spallation Neutron Source accelerator will be discussed. We will also describe current improvements now underway, including Nuclear Resonance Fluorescence (NRF), pulsed sources, and others. We will also describe very new work begun on a threat‐reduction (TR) user interface, designed to simplify the setup of TR‐related calculations, and introduce standards into geometry, sources and backgrounds.