In order to model the activated isotopes and resulting dose from a nuclear detonation in an urban environment, the Activation and Transmutation of Isotopes in an Unstructured Mesh (ACTIUM) Python toolkit has been developed to combine the unstructured mesh-based particle transport capability of MCNP6.2 with the CINDER2008 transmutation code to produce quantities of interest for the post-detonation nuclear forensics and weapons effects communities. The ACTIUM toolkit has been implemented and validated with a number of test cases from a simple analytic model to a case study of the urban detonation in Nagasaki, Japan. The ACTIUM approach is the first of its kind to couple the latest release of CINDER2008 as a part of the Activation in Accelerator Radiation Environments (AARE) package with MCNP6.2 and produce transmuted quantities per time step on an unstructured mesh for the nuclear forensics and weapon effects communities. ACTIUM uses the latest ENDF/B-VIII.0, TENDL2017, and JENDL4 cross-section libraries for the transmutation calculations and includes methods for producing material cards for the initial MCNP6.2 unstructured mesh calculation based on highly detailed materials often found in urban environments on a city-specific basis.

The Monte Carlo Application ToolKit (MCATK) is a component-based software library designed to build specialized applications and to provide new functionality for existing general purpose Monte Carlo radiation transport codes. We will describe MCATK and its capabilities along with presenting some verification and validations results.

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.

This special issue of Nuclear Technology features four papers on topics relating to the newly released Monte Carlo N-Particle version 6 ~MCNP6! radiation transport code, the result of six years of merging MCNPX into MCNP5. MCNP6 Beta 2 was released in February 2012, and MCNP6 Beta 3 was frozen in June 2012 for subsequent release. The plan is that the Beta 3 release contains all the capabilities found in the production release. The first paper in this special issue, “Initial MCNP6 Release Overview,” presents an overview of MCNP6: its development and release history; new geometry, physics, tally, and variance reduction capabilities; the underlying physics models and nuclear and atomic data; and discussions on the confidence in the code’s accuracy. The second and third papers in this issue focus on two of these new features. “MCNP6 Unstructured Mesh Initial Validation and Performance Results” discusses the new ability for MCNP6 to read unstructured mesh geometry and associated materials from the commercial finite element code Abaqus. While many MCNP users will be interested in the multiphysics analysis now possible ~e.g., radiation-induced melting and thermomechanical deformation!, many more will simply be interested in the ability to import mesh geometries created by computer-aided design0 computer-aided engineering tools into MCNP. The next paper, “The MCNP6 Delayed-Particle Feature,” discusses MCNP6 delayed neutron and gamma capabilities and validation efforts. The fourth paper, “MCNP Variance Reduction Developments in the 21st Century,” is an overview of methods implemented since 2000: pulse-height tallies, nested dxtran spheres, precollision and next-event estimators. In my mind, these papers are proof that MCNP maintains an ongoing and vibrant development. Our endeavors develop new, robust, and valuable methods and features for our diverse user community. Our active user base continues to grow and to use our code for the betterment of society. I also believe that MCNP has a bright future. We strive to extend physics, parallel capability, variance reduction methods, and more. I am particularly fond of three new capabilities: the aforementioned mesh capability, an On-The-Fly Doppler cross-section broadening for temperature effects, and a continuous-energy cross-section sensitivity keff analysis tool. The latter two are described on our mcnp.lanl.gov website. What will MCNP look like in its second 35 years? I cannot say. But, I do know that it takes a creative, dedicated staff to achieve what we have so far, and they should be applauded for it. I especially wish to thank the longtime staff members who have recently retired: Tom Booth, Art Forster, John Hendricks, Denise Pelowitz, Dick Prael, and Laurie Waters. Together they have devoted more than 150 person-years of effort to MCNP. Finally, we appreciate and acknowledge the funding from the U.S. Departments of Energy ~principally, Advanced Scientific Computing Research!, Defense, and Homeland Security, and ultimately, the U.S. taxpayer.

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 interaction of radiation with matter can cause activation or fission reactions producing unstable residuals that decay with the emission of delayed-neutron and/or delayed-gamma radiation. This delayed radiation can be exploited for a variety of purposes, including homeland security, health physics, instrumentation and equipment design, and nuclear forensics. Here we report on capability that has been developed to provide automated simulations of delayed-neutron and/or delayed-gamma radiation using MCNP6. We present new high-fidelity delayed-gamma simulation results for models based on the neutron-fission experiments conducted by Beddingfield and Cecil to illustrate and validate this powerful feature.

The increasing uses of voxelied human body geometry have created challenges in many aspects of Monte Carlo simulations . This paper presents an investigation on the maximum number of voxels the MCNP code can handle by using an extremely detailed VIP-Man model . The VIP-Man tomographic model is a radiation dosimetry model developed from the segmented Visible Human color cross sectional images . Several variables that MCNP uses to store the information related to the MCNP "universe" have been changed in this study to maximize the efficient usage of the system memory . Two modifications have been made in the MCNP source code for the variable, laf, which greatly reduce the amount of system memory required for these kinds of MCNP geometries. Another variable, mazu, which is not necessarily used when running the voxel models, is only allocated if it is used. The VIP-Man model having about 25 million voxels caused MCNP5 (MCNPS_RSICC 1 .20) to crash. However, a simple voxel model was tested and results showed that the modified version of MCNP5 can handle lattice geometries of 400 million voxels in size. This modified version was used to calculate the transmission radiograph of a portion of the VIP-Man with 100 million voxels . From this study, we conclude that, although there have been many improvements in computer speed and tallies in the MCNP, both the current and modified versions of the MCNP remains unable to handle the whole VIPMan model at the original voxel size of 0 .33mm x 0 .33 mm x 1 .0 mm, with 3 .7 billion voxels, which would exceed the operating system limits of 32 bit MS Windows PCs if every voxel's material was stored in memory. However, a VIP-Man with larger resolution, such as 1 mm cubic voxels, would potentially be able to run on the modified version of MCNPS . Key Word.s: VIP-Man, MCNP, voxel, do.rimetry

In July of 2004, an updated version of MCNP5{trademark} (MCNP5-RSICC-1.30) was released to the Radiation Shielding Information Computational Center. This updated version has three new features, thirteen bug fixes and several minor coding improvements. The new features are: support for 8 byte integers, specialized tally treatment of large lattices, and mesh tally enhancements. Of the thirteen bug fixes, only four resulted in incorrect answers in specific circumstances. In addition to the standard RSICC distribution of the MCNP5 source, executables and patches, the patch file (only) is available on the MCNP website: http://www-xdiv.lanl.gov/x5/MCNP/theresources.html. The three new MCNP5 features are discussed. Several new improvements have also been made to the manual and development environment. All of the features, bug fixes, coding improvement issues and related documentation are now maintained in Sourceforge. Fortran and C source code and regression test problems are now under version control with CVS.

With the release of MCNP 5, much more emphasis has been placed on improving its functionality on PCs running Microsoft Windows (9X/NT/2000/XP). Enhanced installation and build systems, support for more Fortran compilers, integration with X11 graphics build files, and MPI and PVM parallel capabilities have been implemented in MCNP 5 for Windows PCs. MCNP 5 can be installed with an InstallShield setup programs, similar to other Windows programs, for users who only need to install executables and data libraries. For those users who need to compile the source, the GNU make utility can used in conjunction with three supported Fortran compilers: Compaq Visual Fortran (CVF), Lahey Fortran 95 (LF95) or Absoft Fortran 95 (AF95). Alternatively, Compaq Developer Studio can be used to compile MCNP 5. The X-Windows plotting capabilities have been improved, and all the appropriate open source X11R6 files for compiling MCNP 5 are bundled with the MCNP 5 source code. X client software is still needed, however, to display geometry, cross-section or tally plots. Parallel capabilities which exist on other platforms have been extended to Windows PCs, allowing users to utilize dual CPU PCs, clusters of homogeneous Windows PCs (preferably with MPI), or heterogeneous clusters (preferably with PVM). Wall-clock runtimes show that MCNP 5 compiled with CVF runs 1.4 times faster than when it is compiled with LF95 or AF95. Wall-clock runtimes also show MCNP 5 with MPI more effectively utilizes a dual-processor Windows 2000 PC than MCNP 5 with PVM.

Dried peas, soybeans, and other legumes have been investigated in relation to their generation of ascorbic ácid, thiamine, riboflavin, and nicotinic acid following germination. It has been determined that, generally speaking, such sprouted legumes are an excellent source of ascorbic acid, fair as a source of riboflavin and nicotinic acid, and poor for thiamine. The various varieties of peas were more productive of the vitamins concerned than were soybeans, which, however, were of considerable merit and markedly superior to the other types of beans investigated. A temperature effect was determined, indicating that both germination and the generation of vitamins proceeded at a faster rate at the higher temperature range. The applicability tcmilitary conditions of this method of producing an accessory supply of water soluble vitamins is indicated.