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Both high-fidelity & toy simulations are being used to understand measured signals and improve the Area 11 NDSE diagnostic. We continue to gain more and more confidence in the ability for MCNP to simulate neutron and photon transport from source to radiation detector.
This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/k$. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$_2$, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the control rod models in MAMMOTH and adding the BISON thermo-elastic models and thermal-fluids heat transfer.
In order to fully verify temperature measurements made on the target window using infrared (IR) optical non-contact methods, actual comparative measurements are made with a real beam distribution as the heat source using Argonne National Laboratory’s (ANL) 35 MeV electron accelerator. Using Monte Carlo N-Particle (MCNP) simulations and thermal Finite Element Analysis (FEA), a cooled mock target window with thermocouple implants is designed to be used in such a test to achieve window temperatures up to 700°C. An uncoated and blackcoated mock window is designed to enhance the IR temperature measurements and verify optical transmitted radiation (OTR) imagery. This allows us to fully verify and characterize our temperature accuracy with our current IR camera method and any future method we may wish to explore using actual production conditions. This test also provides us with valuable conclusions/concerns regarding the calibration method we developed using our IR test stand at TA-53 in MPF-14.
When neutrons are scattered by target nuclei at elevated temperatures, it is entirely possible that the neutron will actually gain energy (i.e., up-scatter) from the interaction. This phenomenon is in addition to the more usual case of the neutron losing energy (i.e., down-scatter). Furthermore, the motion of the target nuclei can also cause extended neutron down-scattering, i.e., the neutrons can and do scatter to energies lower than predicted by the simple asymptotic models. In recent years, more attention has been given to temperature-dependent scattering cross sections for materials in neutron multiplying systems. This has led to the inclusion of neutron up-scatter in deterministic codes like Partisn and to free gas scattering models for material temperature effects in Monte Carlo codes like MCNP and cross section processing codes like NJOY. The free gas scattering models have the effect of Doppler Broadening the scattering cross section output spectra in energy and angle. The current state of Doppler-Broadening numerical techniques used at Los Alamos for scattering resonances will be reviewed, and suggestions will be made for further developments. The focus will be on the free gas scattering models currently in use and the development of new models to include high-Z resonance scattering effects. These models change the neutron up-scattering behavior.
The U.S. Stockpile Stewardship Program [1] is designed to sustain and evaluate the nuclear weapons stockpile while foregoing underground nuclear tests. The maintenance of a smaller, aging U.S. nuclear weapons stockpile without underground testing requires complex computer calculations [14]. These calculations in turn need to be verified and benchmarked [14]. A wide range of research facilities have been used to test and evaluate nuclear weapons while respecting the Comprehensive Nuclear Test-Ban Treaty (CTBT) [2]. Some of these facilities include the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory, the Z machine at Sandia National Laboratories, and the Dual Axis Radiographic Hydrodynamic Test (DARHT) facility at Los Alamos National Laboratory. This research will focus largely on DARHT (although some information from Cygnus and the Los Alamos Microtron may be used in this research) by modeling it and comparing to experimental data. DARHT is an electron accelerator that employs high-energy flash x-ray sources for imaging hydro-tests. This research proposes to address some of the issues crucial to understanding DARHT Axis II and the analysis of the radiographic images produced. Primarily, the nature of scatter at DARHT will be modeled and verified with experimental data. It will then be shown that certain design decisions can be made to optimize the scatter field for hydrotest experiments. Spectral effects will be briefly explored to determine if there is any considerable effect on the density reconstruction caused by changes in the energy spectrum caused by target changes. Finally, a generalized scatter model will be made using results from MCNP that can be convolved with the direct transmission of an object to simulate the scatter of that object at the detector plane. The region in which with this scatter model is appropriate will be explored.
FY17 work focused on automation, both of the measurement analysis and comparison of simulations. The experimental apparatus was relocated and weeks of continuous measurements of the spontaneous fission source 252 Cf was performed. Programs were developed to automate the conversion of measurements into ROOT data framework files with a simple terminal input. The complete analysis of the measurement (which includes energy calibration and the identification of correlated counts) can now be completed with a documented process which involves one simple execution line as well. Finally, the hurdles of slow MCNP simulations resulting in low simulation statistics have been overcome with the generation of multi-run suites which make use of the highperformance computing resources at LANL. Preliminary comparisons of measurements and simulations have been performed and will be the focus of FY18 work.
Verification and validation of our solutions for calculating the neutron reactivity for nuclear materials is a key issue to address for many applications, including criticality safety, research reactors, power reactors, and nuclear security. Neutronics codes solve variations of the Boltzmann transport equation. The two main variants are Monte Carlo versus deterministic solutions, e.g. the MCNP [1] versus PARTISN [2] codes, respectively. There have been many studies over the decades that examined the accuracy of such solvers and the general conclusion is that when the problems are well-posed, either solver can produce accurate results. However, the devil is always in the details. The current study examines the issue of self-shielding and the stress it puts on deterministic solvers. Most Monte Carlo neutronics codes use continuous-energy descriptions of the neutron interaction data that are not subject to this effect. The issue of self-shielding occurs because of the discretisation of data used by the deterministic solutions. Multigroup data used in these solvers are the average cross section and scattering parameters over an energy range. Resonances in cross sections can occur that change the likelihood of interaction by one to three orders of magnitude over a small energy range. Self-shielding is the numerical effect that the average cross section in groups with strong resonances can be strongly affected as neutrons within that material are preferentially absorbed or scattered out of the resonance energies. This affects both the average cross section and the scattering matrix.
The 235 U mass assay of bulk uranium items, such as oxide canisters, fuel pellets, and fuel assemblies, is not achievable by traditional gamma-ray assay techniques because of the limited penetration of the item by the characteristic 235 U gamma rays. Instead, fast neutron interrogation methods such as active neutron coincidence counting must be used. For international safeguards applications, the most commonly used active neutron systems, the Active Well Coincidence Counter (AWCC), Uranium Neutron Coincidence Collar (UNCL), and 252 Cf Shuffler, rely on fast neutron interrogation using an isotopic neutron source [i.e., 252 Cf or Am(Li)] to achieve better measurement accuracies than are possible using gamma-ray techniques for high-mass, high-density items. However, the Am(Li) sources required for the AWCC and UNCL systems are no longer manufactured, and newly produced systems rely on limited supplies of sources salvaged from disused instruments. The 252 Cf shuffler systems rely on the use of high-output 252 Cf sources, which while still available have become extremely costly for use in routine operations and require replacement every 5–7 years. Lack of a suitable alternative neutron interrogation source would leave a potentially significant gap in the safeguarding of uranium processing facilities. Previously we examined the performance of Oak Ridge National Laboratory’s Large Volume Active Well Coincidence Counter modified to accept a commercially available deuterium-deuterium (DD) neutron generator to examine the potential of the DD neutron generator as an alternative to the isotopic sources [1]. That work indicated that the DD generator was a viable replacement for the AWCC applications, but because of the large dimensions of that generator, the DD/AWCC would have to be run as a delayed neutron counting system to provide equivalent measurement precision. In this work we examine the performance of a standard UNCL modified to accept the DD neutron generator for the assay of 235 U based on the results of Monte Carlo N-Particle (MCNP) simulations and measurements of depleted uranium and highly enriched uranium items. The projected performance of the modified UNCL for fresh pressurized water reactor fuel assemblies operated in both traditional coincidence counting and delayed neutron counting modes is presented.
The subcritical reactivity, ρ, of a neutron multiplying system is of interest to many factions of the nuclear community. The subcritical reactivity in a nuclear system can be inferred from measurements of the prompt neutron decay constant, α. Near critical, the reactivity is linearly proportional with α, but the linearity between these quantities does not hold in systems with low multiplication or significantly above prompt critical. This work examines a large range of subcritical prompt neutron decay constant measurements on a thermal HEU metal system. This system is designed to mimic a solution system and to have the optimal H/ 235 U ratio to minimize the critical mass. The reactivity of these measurements varies between 0$\$$ and 70$\$$ below critical. In addition to experimental measurements on the thermal system, simulations are performed on both the thermal HEU system and a fast HEU metal system. Both the fast and thermal results are compared to baseline simulations performed using criticality eigenvalue (KCODE) mode in MCNP®6.2. These simulations are used for comparisons on the linearity between α and ρ. Prompt neutron decay constant measurements in this work are performed using the Rossi-α method. The measurements are taken using a 3 He neutron detection system and a Los Alamos National Laboratory designed list-mode data acquisition system. A specially designed computer program determines α from list-mode data. The α-eigenvalue, or α at delayed critical, for the thermal HEU system is -199:4±4:4 s -1 as measured during this work. The α-eigenvalue for the fast HEU system was previously measured to be -8:5 x 10 -5 s -1 . The results from the thermal HEU system study suggest that the relationship between ρ and α is linear when k eff > 0:80 based on the simulation results. The experimental results also agree with this result, but have less fidelity. No experimental results are available for the fast HEU system.
The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very high void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.
MCNP6.2 was released in 2018 and is the latest production release of MCNP ® available from RSICC. Previous releases were MCNP6.1 [3] in 2013 and MCNP6.1.1 in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods using Whisper for NCS validation
This report summarizes the efforts to investigate the gamma-ray radiation signatures present in High Active Liquid (HAL) to determine the feasibility of developing advanced plutonium monitoring technology for plutonium to improve safeguards verification measurements and nuclear material accountancy at the Tokai Reprocessing Plant. This includes the development of MCNP models of a waste tank and ion chamber and using experimental results to benchmark the model. With this model the expected dose rate and gamma radiation signatures could be determined for various volumes of liquid waste in the tank.
This paper is a user’s guide to the current version of the MARS15 Monte Carlo code. MARS allows inclusive and, in many cases, exclusive simulations of three-dimensional hadronic and electromagnetic cascades and modeling of heavy ion, muon and low energy neutron-photon transport in accelerator, detector, spacecraft and shielding components. Particles can have energies from a fraction of an electronvolt up to about 100 TeV. The code has undergone substantial improvements since the last documented version MARS13(95) and since previous releases of MARS15; all these as well as other specific features of the MARS code system are explained in detail. Descriptions of general input and output with commentary and recommendations are given. Examples are given for running the program with distributed sources, complex compounds, arbitrary geometries, and magnetic fields. A built-in graphical-user interface allows the visualization and detailed analysis of the geometry description. Enhanced options are described for the MARS code being coupled with: the DPMJET hadron event generator, the MCNP code for low-energy neutron transport, the ANSYS system for thermal and stress analysis, physics analysis and graphics packages, the MAD accelerator lattice description via a universal MAD-MARS beam line builder, the STRUCT program for tracking particles in accelerator lattices with beam loss recording, the DETRA code for nuclide decay and transmutation calculations, the EGS5 code for precise modeling of low-energy electromagnetic showers, and the ROOT system for description and visualization of arbitrary geometries. Use of the code in a multistage mode, coupled with event generators (DPMJET), with the STRUCT program for tracking particles in accelerator lattices with beam loss recording, and with physics analysis and graphics packages is demonstrated with typical input and output examples.
Neutron transport methods used to establish subcriticality require validation by comparison to critical experiments considered to be benchmarks. Whisper is a sensitivity/uncertainty analysis tool developed to assist with the task of validation in nuclear criticality safety. Details on the Whisper methodology can be found in References 1-3 on the MCNP® reference collection website at https://mcnp.lanl.gov. During the process of validation there can be cases where a benchmark experiment may be found to be a statistical outlier, in which the calculated k-effective value and the experiment k-effective value differ by an amount atypical for similar experiments. A methodology optionally employed by Whisper is the exclusion of statistical outliers based upon the iterative diagonal chi-squared statistical rejection technique. Alternatively, there is an option to include all benchmarks in the Whisper library collection, even those benchmarks found to be statistical outliers, when computing the bias, bias uncertainty and margin of subcriticality (MOS) leading to establishment of the baseline upper subcritical limit (USL). A comparison study has been done to compute USLs with and without statistical outliers in the Whisper benchmark collection to determine what effect rejection of statistical outliers has on the recommended USL. The results show little overall difference in the recommended baseline USLs developed by Whisper when excluding statistical outliers. Additionally, there does not appear to be a clear trend in predicting whether the baseline USL will be higher or lower when rejecting statistical outliers from the benchmark critical experiment collection used for validation.
A library is introduced that enables users to conduct Monte Carlo tallies on heterogeneous computing systems. The library can be built with or linked to a Monte Carlo transport code to provide tally capabilities for the project. The library provides a set of tools and an interface for constructing tallies that can be computed on CPUs or GPUs regardless of where the sample data was generated. The library can be incorporated into existing Monte Carlo transport codes at both the research and production level with minimal intrusion into the transport code. The tally library is capable of computing a variety of tallies including track-length histogram (mesh) tallies, collisionbased kernel density estimators, functional expansion tallies, eigenvalue sensitivity coefficients via iterated fission probability, and sensitivities of reaction rates or adjoint-weighted quantities via differential operator sampling. The parallelization methods are discussed and speedups are shown for 2-D and 3-D problems with comparisons to both the tally library CPU implementations as well as the tally capabilities in the host transport code. Performance results and examples are shown for MCNP, MCATK, and OpenMC.
The Nuclear Inspection Node Event SIMulator (NINESIM) is a tool that is capable of modeling a node that monitors passing traffic for radiological and nuclear materials. This task requires the consideration of a number of factors. One such factor is clutter, or the level of traffic passing through a node. Clutter might affect the performance of radiation detectors due to attenuation or scattering that particles undergo as they move through the material. This paper discusses the results from MCNP simulations that were performed to determine in what scenarios clutter had a significant impact on detector performance, where the receiver operating characteristic (ROC) curve was used as the metric to evaluate detector performance. Clutter was found to have an insignificant impact on neutron detection, therefore, clutter may be ignored when modeling neutron detectors. However, there were a number of scenarios where clutter was found to affect the performance of photon detectors; the amount of traffic was found to be directly proportional to the effectiveness of photon detectors. The results of this work were then used to develop a photon clutter scaling algorithm that will be implemented into NINESIM. The Domestic Nuclear Detection Office, and possibly other agencies such as the Defense Threat Reduction Agency, will benefit from this work by eventually obtaining a tool that can enhance their ability in countering the illicit movement of radiological and nuclear materials.
A comparison of the neutronic properties between the legacy high specific activity (HSA) cobalt capsule design and the new HSA cobalt capsule design was conducted utilizing the full advanced test reactor core MCNP model. The neutron flux and effective neutron absorption cross section for the desired (n,gamma) reaction as well as the undesired reactions (n,2n), (n, alpha), and (n,p) for the cobalt target material for each design was calculated and compared. It was found that the effective neutron absorption cross section for the desired reaction (n,gamma) on the target material that directly produces Co-60 is slightly higher in the new design than the legacy design, and there is little statistical difference between the two target designs for the undesired reactions (n,2n), (n, ), and (n,p). Based on these results, neutronically, there does not appear to be a significant difference between the two capsule designs.
A critical experiment was conducted at NCERC on 6 Jun 2016. In this experiment, two sets of activation foils (Cu, Ni, Sc and Ti) were placed within the core. After the irradiation, the foils were withdrawn, packaged and set to the LANL C-NR Countroom and the foils were counted over a period of a few weeks. The results were tabulated and a simple MCNP model was created to simulate a neutron spectrum consistent with the foil results.