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DOE OSTI2011

3-D Deep Penetration Neutron Imaging of Thick Absorgin and Diffusive Objects Using Transport Theory

A current area of research interest in national security is to effectively and efficiently determine the contents of the many shipping containers that enter ports in the United States. This interest comes as a result of the 9/11 Commission Act passed by Congress in 2007 that requires 100% of inbound cargo to be scanned by 2012. It appears that this requirement will be achieved by 2012, but as of February of 2009 eighty percent of the 11.5 million inbound cargo containers were being scanned. The systems used today in all major U.S. ports to determine the presence of radioactive material within cargo containers are Radiation Portal Monitors (RPM). These devices generally exist in the form of a gate or series of gates that the containers can be driven through and scanned. The monitors are effective for determining the presence of radiation, but offer little more information about the particular source. This simple pass-fail system leads to many false alarms as many everyday items emit radiation including smoke detectors due to the Americium-241 source contained inside, bananas, milk, cocoa powder and lean beef due to the trace amounts of Potassium-40, and fire brick and kitty litter due to their high clay content which often contains traces of uranium and thorium. In addition, if an illuminating source is imposed on the boundary of the container, the contents of the container may become activated. These materials include steel, aluminum and many agricultural products. Current portal monitors also have not proven to be that effective at identifying natural or highly enriched uranium (HEU). In fact, the best available Advanced Spectroscopic Portal Monitors (ASP) are only capable of identifying bare HEU 70-88% of the time and masked HEU and depleted uranium (DU) only 53 percent of the time. Therefore, a better algorithm that uses more information collected from better detectors about the specific material distribution within the container is desired. The work reported here explores the inverse problem of optical tomography applied to heterogeneous domains. The neutral particle transport equation was used as the forward model for how neutral particles stream through and interact within these heterogeneous domains. A constrained optimization technique that uses Newtons method served as the basis of the inverse problem. Optical tomography aims at reconstructing the material properties using (a) illuminating sources and (b) detector readings. However, accurate simulations for radiation transport require that the particle (gamma and/or neutron) energy be appropriate discretize in the multigroup approximation. This, in turns, yields optical tomography problems where the number of unknowns grows (1) about quadratically with respect to the number of energy groups, G, (notably to reconstruct the scattering matrix) and (2) linearly with respect to the number of unknown material regions. As pointed out, a promising approach could rely on algorithms to appropriately select a material type per material zone rather than G2 values. This approach, though promising, still requires further investigation: (a) when switching from cross-section values unknowns to material type indices (discrete integer unknowns), integer programming techniques are needed since derivative information is no longer available; and (b) the issue of selecting the initial material zoning remains. The work reported here proposes an approach to solve the latter item, whereby a material zoning is proposed using one-group or few-groups transport approximations. The capabilities and limitations of the presented method were explored; they are briefly summarized next and later described in fuller details in the Appendices. The major factors that influenced the ability of the optimization method to reconstruct the cross sections of these domains included the locations of the sources used to illuminate the domains, the number of separate experiments used in the reconstruction, the locations where measurements were collected, the optical thickness of the domain, the amount of signal noise and signal bias applied to the measurements and the initial guess for the cross section distribution. All of these factors were explored for problems with and without scattering. Increasing the number of source and measurement locations and experiments generally was more successful at reconstructing optically thicker domains while producing less error in the image. The maximum optical thickness that could be reconstructed with this method was ten mean free paths for pure absorber and two mean free paths for scattering problems. Applying signal noise and signal bias to the measured fluxes produced more error in the produced image. Generally, Newtons method was more successful at reconstructing domains from an initial guess for the cross sections that was greater in magnitude than their true values than from an initial guess that was lower in magnitude.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

A Brief User's Guide for PARTISN. Revision 1

This document is intended as a very basic primer on the use of the PARTISN multigroup discrete-ordinates transport code for neutron criticality and time-independent neutron leakage problems. One-dimensional (r only) spherical and two-dimensional (r-z) cylindrical geometries and only the most basic options are discussed. It is assumed that the Nuclear Data Interface (NDI) is the source of nuclear data, that volume sources are isotropic, and that external boundaries are vacuums. The location and meaning of useful information in the PARTISN output is discussed. A sample input deck for a cylindrical problem that stacks a leakage calculation and a k eff calculation is listed. The method for inputting cross sections by hand (e.g., for analytic benchmark comparisons) is presented.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2013

Criticality Safety Validation of SCALE 6.1 (Revised)

The computational bias of criticality safety computer codes must be established through the validation of the codes to critical experiments. A large collection of suitable experiments has been vetted by the International Criticality Safety Benchmark Experiment Program (ICSBEP) and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). A total of more than 350 cases from this reference have been prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID) maintained by the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The performance of the KENO V.a and KENO-VI Monte Carlo codes within the SCALE 6.1 code system with ENDF/B-VII.0 cross-section data in 238-group and continuous energy is assessed using the VALID models of benchmark experiments. The TSUNAMI tools for sensitivity and uncertainty analysis are utilized to examine some systems further in an attempt to identify potential causes of unexpected results. The critical experiments available for validation of the KENO V.a code cover eight different broad categories of systems. These systems use a range of fissile materials including a range of uranium enrichments, various plutonium isotopic vectors, and mixed uranium/plutonium oxides. The physical form of the fissile material also varies and is represented as metal, solutions, or arrays of rods or plates in a water moderator. The neutron energy spectra of the systems also vary and cover both fast and thermal spectra. Over 300 of the total cases used utilize the KENO V.a code. The critical experiments available for the validation of the KENO-VI code cover three broad categories of systems. The fissile materials in the systems vary and include high- and intermediate-enrichment uranium and mixed uranium/plutonium oxides. The physical form of the fissile material is either metal or rod arrays in water. As with KENO V.a, both fast and thermal neutron energy spectra are represented in the systems considered. The results indicate generally good performance of both the KENO V.a and KENO-VI codes across the range of systems analyzed. The bias of calculated k eff from expected values is less than 0.9% Δk in all cases. All eight categories of experiments show biases of less than 0.5% Δk in KENO V.a with the exception of intermediate enrichment metal systems using the 238-group library. The continuous energy library generally manifests lower biases than the multigroup data. The KENO-VI results show slightly larger biases, though this may primarily be the result of modeling systems with more geometric complexity, which are more difficult to describe accurately, even with a generalized geometry code like KENO-VI. Several additional conclusions can be drawn from the results of this validation effort. These conclusions include that the TSUNAMI tools can be used successfully to explain the cause of aberrant results, that some evaluations in the IHECSBE should be updated to provide more rigorous expected k eff values and uncertainties, and that potential cross-section errors can be identified by detailed review of the results of this validation. It also appears that the overall cross-section uncertainty as quantified through the SCALE covariance library is overestimated. Overall, the KENO V.a and KENO-VI codes are shown to provide consistent, low-bias results for a wide range of physical systems of potential interest in criticality safety applications.

Marshall, William (BJ) J.↗

DOE OSTI2013

Advances in Sensitivity Analysis with MCNP [Slides]

Monte Carlo codes such as MCNP are used extensively in preventing inadvertent nuclear criticalities in processes involving fissionable material. Their ability to do this relies on the accuracy of the simulation tools, the underlying nuclear data libraries, and their validation of relevant experiments. Recent development of an inline method for importance or adjoint weighting with continuous-energy Monte Carlo in MCNP has allowed for the calculation of sensitivity coefficients without the need of meshes or multigroup collapse, unlike previously available tools. These capabilities allow for quantitative comparisons of processes to benchmark experiments and uncertainty quantification for the calculation of biases and safety margins. The current state-of-the-art is discussed, as well as future research avenues.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2013

Wedgehog Manual and Interface Guide, Jayenne-7_0_0

Wedgehog is an IMC package developed by the CCS–2’s Monte Carlo Team1 . Wedgehog is designed to interface with a time-split radiation-hydrodynamics (rad-hydro) code and supports one-dimensional R, two-dimensional RZ or three dimensional XY Z geometry. The geometry is confined to logically rectangular Adaptive Mesh Refinement (AMR) meshes with unit aspect ratios and 2:1 refinement between adjacent cells. Wedgehog supports both two and three temperature simulations. Available scattering models for photon-electron interaction include Compton and isotropic scattering. It maintains a census of particles in between time steps to preserve the contents of the radiation field. Wedgehog can accommodate frequency dependence by using a “Multigroup” (MG) approximation that essentially averages the frequency-dependent opacity information into discrete bins. Additionally, Wedge hog includes a capability to use Opacity Distribution Factors (ODFs) that may be generated with the XCP-5 code TOPS. A previous version of Wedgehog that utilizes IBM’s Cell chip is available for the Roadrunner petascale computer at Los Alamos National Laboratory (circa 2006-2013, estimated). However, as Roadrunner is going out of service in the near future, we have dropped support for the interfaces that were deprecated in Jayenne 6 4 0 and removed in Jayenne-7 0 0. The Roadrunner version had the same interface as the mainline code and is available in RZ and XYZ geometry with all the same features except for flux tallies across a surface (i.e., tally sphere and RZ line-segment).

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2016

New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple_ace.pl and simple_ace_mg.pl.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

The MCNP6 Analytic Criticality Benchmark Suite

Analytical benchmarks provide an invaluable tool for verifying computer codes used to simulate neutron transport. Several collections of analytical benchmark problems [1-4] are used routinely in the verification of production Monte Carlo codes such as MCNP® [5,6]. Verification of a computer code is a necessary prerequisite to the more complex validation process. The verification process confirms that a code performs its intended functions correctly. The validation process involves determining the absolute accuracy of code results vs. nature. In typical validations, results are computed for a set of benchmark experiments using a particular methodology (code, cross-section data with uncertainties, and modeling) and compared to the measured results from the set of benchmark experiments. The validation process determines bias, bias uncertainty, and possibly additional margins. Verification is generally performed by the code developers, while validation is generally performed by code users for a particular application space. The VERIFICATION_KEFF suite of criticality problems [1,2] was originally a set of 75 criticality problems found in the literature for which exact analytical solutions are available. Even though the spatial and energy detail is necessarily limited in analytical benchmarks, typically to a few regions or energy groups, the exact solutions obtained can be used to verify that the basic algorithms, mathematics, and methods used in complex production codes perform correctly. The present work has focused on revisiting this benchmark suite. A thorough review of the problems resulted in discarding some of them as not suitable for MCNP benchmarking. For the remaining problems, many of them were reformulated to permit execution in either multigroup mode or in the normal continuous-energy mode for MCNP. Execution of the benchmarks in continuous-energy mode provides a significant advance to MCNP verification methods.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Performance Improvements for Coarse Mesh Finite Difference Acceleration L3:RTM.PRT.P13.02

The development of VERA-CS in recent years has focused on developing the capability to simulate multiple cycles of operating commercial nuclear power plants. Now that these capabilities have advanced to the point where it is being deployed to users, the focus is on improving the computational performance of various components in VERA-CS. In this work, the focus is on the Coarse Mesh Finite Difference (CMFD) solution in MPACT. CMFD serves multiple purposes in the 2D/1D solution methodology. First, it is a natural mechanism to tie together the radial MOC transport and the axial SP3 solution. Because the CMFD system solves the multigroup three-dimensional core in one system, it pulls together the global response of the system. In addition, the CMFD solution provides a framework to accelerate the convergence of the eigenvalue problem.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

MPACT Theory Manual, Version 2.2.0

This theory manual describes the three-dimensional (3-D) whole-core, pin-resolved transport calculation methodology employed in the MPACT code. To provide sub-pin level power distributions with sufficient accuracy, MPACT employs the method of characteristics (MOC) solutions in the framework of a 3-D coarse mesh finite difference (CMFD) formulation. MPACT provides a 3D MOC solution, but also a 2D/1D solution in which the 2D planar solution is provided by MOC and the axial coupling is resolved by one-dimensional (1-D) lower order (diffusion or P3) solutions. In Chapter 2 of the manual, the MOC methodology is described for calculating the regional angular and scalar fluxes from the Boltzmann transport equation. In Chapter 3, the 2D/1D methodology is described, together with the description of the CMFD iteration process involving dynamic homogenization and solution of the multigroup CMFD linear system. A description of the MPACT depletion algorithm is given in Chapter 4, followed by a discussion of the subgroup and ESSM resonance processing methods in Chapter 5. The final Chapter 6 describes a simplified thermal hydraulics model in MPACT.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

VERA-CS Verification & Validation Plan

This report summarizes the current status of VERA-CS Verification and Validation for PWR Core Follow operation and proposes a multi-phase plan for continuing VERA-CS V&V in FY17 and FY18. The proposed plan recognizes the hierarchical nature of a multi-physics code system such as VERA-CS and the importance of first achieving an acceptable level of V&V on each of the single physics codes before focusing on the V&V of the coupled physics solution. The report summarizes the V&V of each of the single physics codes systems currently used for core follow analysis (ie MPACT, CTF, Multigroup Cross Section Generation, and BISON / Fuel Temperature Tables) and proposes specific actions to achieve a uniformly acceptable level of V&V in FY17. The report also recognizes the ongoing development of other codes important for PWR Core Follow (e.g. TIAMAT, MAMBA3D) and proposes Phase II (FY18) VERA-CS V&V activities in which those codes will also reach an acceptable level of V&V. The report then summarizes the current status of VERA-CS multi-physics V&V for PWR Core Follow and the ongoing PWR Core Follow V&V activities for FY17. An automated procedure and output data format is proposed for standardizing the output for core follow calculations and automatically generating tables and figures for the VERA-CS Latex file. A set of acceptance metrics is also proposed for the evaluation and assessment of core follow results that would be used within the script to automatically flag any results which require further analysis or more detailed explanation prior to being added to the VERA-CS validation base. After the Automation Scripts have been completed and tested using BEAVRS, the VERA-CS plan proposes the Watts Bar cycle depletion cases should be performed with the new cross section library and be included in the first draft of the new VERA-CS manual for release at the end of PoR15. Also, within the constraints imposed by the proprietary nature of plant data, as many as possible of the FY17 AMA Plant Core Follow cases should also be included in the VERA-CS manual at the end of PoR15. After completion of the ongoing development of TIAMAT for fully coupled, full core calculations with VERA-CS / BISON 1.5D, and after the completion of the refactoring of MAMBA3D for CIPS analysis in FY17, selected cases from the VERA-CS validation based should be performed, beginning with the legacy cases of Watts Bar and BEAVRS in PoR16. Finally, as potential Phase III future work some additional considerations are identified for extending the VERA-CS V&V to other reactor types such as the BWR.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

SCALE Code System 6.2.2

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including 3 deterministic and 3 Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2011

Production and Testing of the VITAMIN-B7 Fine-Group and BUGLE-B7 Broad-Group Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data

New coupled neutron-gamma cross-section libraries have been developed for use in light water reactor (LWR) shielding applications, including pressure vessel dosimetry calculations. The libraries, which were generated using Evaluated Nuclear Data File/B Version VII Release 0 (ENDF/B-VII.0), use the same fine-group and broad-group energy structures as the VITAMIN-B6 and BUGLE-96 libraries. The processing methodology used to generate both libraries is based on the methods used to develop VITAMIN-B6 and BUGLE-96 and is consistent with ANSI/ANS 6.1.2. The ENDF data were first processed into the fine-group pseudo-problem-independent VITAMIN-B7 library and then collapsed into the broad-group BUGLE-B7 library. The VITAMIN-B7 library contains data for 391 nuclides. This represents a significant increase compared to the VITAMIN-B6 library, which contained data for 120 nuclides. The BUGLE-B7 library contains data for the same nuclides as BUGLE-96, and maintains the same numeric IDs for those nuclides. The broad-group data includes nuclides which are infinitely dilute and group collapsed using a concrete weighting spectrum, as well as nuclides which are self-shielded and group collapsed using weighting spectra representative of important regions of LWRs. The verification and validation of the new libraries includes a set of critical benchmark experiments, a set of regression tests that are used to evaluate multigroup crosssection libraries in the SCALE code system, and three pressure vessel dosimetry benchmarks. Results of these tests confirm that the new libraries are appropriate for use in LWR shielding analyses and meet the requirements of Regulatory Guide 1.190.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

(U) Second-Order Sensitivity Analysis of Uncollided Particle Contributions to Radiation Detector Responses Using Ray-Tracing

The Second-Level Adjoint Sensitivity System (2nd-LASS) that yields the second-order sensitivities of a response of uncollided particles with respect to isotope densities, cross sections, and source emission rates is derived in Refs. 1 and 2. In Ref. 2, we solved problems for the uncollided leakage from a homogeneous sphere and a multiregion cylinder using the PARTISN multigroup discrete-ordinates code. In this memo, we derive solutions of the 2nd-LASS for the particular case when the response is a flux or partial current density computed at a single point on the boundary, and the inner products are computed using ray-tracing. Both the PARTISN approach and the ray-tracing approach are implemented in a computer code, SENSPG. The next section of this report presents the equations of the 1st- and 2nd-LASS for uncollided particles and the first- and second-order sensitivities that use the solutions of the 1st- and 2nd-LASS. Section III presents solutions of the 1st- and 2nd-LASS equations for the case of ray-tracing from a detector point. Section IV presents specific solutions of the 2nd-LASS and derives the ray-trace form of the inner products needed for second-order sensitivities. Numerical results for the total leakage from a homogeneous sphere are presented in Sec. V and for the leakage from one side of a two-region slab in Sec. VI. Section VII is a summary and conclusions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2017

A Comparison of Monte Carlo and Deterministic Solvers for keff and Sensitivity Calculations

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.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

FY17 Status Report on NEAMS Neutronics Activities

Under the U.S. DOE NEAMS program, the high-fidelity neutronics code system has been developed to support the multiphysics modeling and simulation capability named SHARP. The neutronics code system includes the high-fidelity neutronics code PROTEUS, the cross section library and preprocessing tools, the multigroup cross section generation code MC2-3, the in-house meshing generation tool, the perturbation and sensitivity analysis code PERSENT, and post-processing tools. The main objectives of the NEAMS neutronics activities in FY17 are to continue development of an advanced nodal solver in PROTEUS for use in nuclear reactor design and analysis projects, implement a simplified sub-channel based thermal-hydraulic (T/H) capability into PROTEUS to efficiently compute the thermal feedback, improve the performance of PROTEUS-MOCEX using numerical acceleration and code optimization, improve the cross section generation tools including MC2-3, and continue to perform verification and validation tests for PROTEUS.

47 OTHER INSTRUMENTATION↗

DOE OSTI2018

Studying effects of non-equilibrium radiative transfer via HPC

This report presents slides on Ph.D. Research Goals; Local Thermodynamic Equilibrium (LTE) Implications; Calculating an Opacity; Opacity: Pictographic Representation; Opacity: Pictographic Representation; Opacity: Pictographic Representation; Collisional Radiative Modeling; Radiative and Collisional Excitation; Photo and Electron Impact Ionization; Autoionization; The Rate Matrix; Example: Total Photoionization rate; The Rate Coefficients; inlinlte version 1.1; inlinlte: Verification; New capabilities: Rate Matrix – Flexibility; Memory Option Comparison; Improvements over previous DCA solver; Inter- and intra-node load balancing; Load Balance – Full Picture; Load Balance – Full Picture; Load Balance – Internode; Load Balance – Scaling; Description; Performance; xRAGE Simulation; Post-process @ 2hr; Post-process @ 4hr; Post-process @ 8hr; Takeaways; Performance for 1 realization; Motivation for QOI; Multigroup Er; Transport and NLTE large effects (1mm, 1keV); Transport large effect, NLTE lesser (1mm, 750eV); Blastwave Diagnostici – Description & Performance; Temperature Comparison; NLTE has effect on dynamics at wall; NLTE has lesser effect in the foam; Global Takeaways; The end.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2016

SCALE Code System

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE depletion with TRITON (T5-DEPL/T6-DEPL),• CE sensitivity/uncertainty analysis with TSUNAMI-3D,• Simplified and efficient LWR lattice physics with Polaris,• Large scale detailed spent fuel characterization with ORIGAMI and ORIGAMI Automator,• Advanced fission source convergence acceleration capabilities with Sourcerer,• Nuclear data library generation with AMPX, and• Integrated user interface with Fulcrum.Enhanced capabilities include:• Accurate and efficient CE Monte Carlo methods for eigenvalue and fixed source calculations,• Improved MG resonance self-shielding methodologies and data,• Resonance self-shielding with modernized and efficient XSProc integrated into most sequences,• Accelerated calculations with TRITON/NEWT (generally 4x faster than SCALE 6.1),• Spent fuel characterization with 1470 new reactor-specific libraries for ORIGEN,• Modernization of ORIGEN (Chebyshev Rational Approximation Method [CRAM] solver, API for high-performance depletion, new keyword input format)• Extension of the maximum mixture number to values well beyond the previous limit of 2147 to ~2 billion,• Nuclear data formats enabling the use of more than 999 energy groups,• Updated standard composition library to provide more accurate use of natural abundances, andvi• Numerous other enhancements for improved usability and stability.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2018

SCALE Code System

97 MATHEMATICS AND COMPUTING↗