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DOE OSTI2017

Cloud Computing for Complex Performance Codes.

This report describes the use of cloud computing services for running complex public domain performance assessment problems. The work consisted of two phases: Phase 1 was to demonstrate complex codes, on several differently configured servers, could run and compute trivial small scale problems in a commercial cloud infrastructure. Phase 2 focused on proving non-trivial large scale problems could be computed in the commercial cloud environment. The cloud computing effort was successfully applied using codes of interest to the geohydrology and nuclear waste disposal modeling community.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

A User’s Guide to the PLTEMP/ANL Code

PLTEMP/ANL V4.2 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2017

ALE3D: An Arbitrary Lagrangian-Eulerian Multi-Physics Code

ALE3D is a multi-physics numerical simulation software tool utilizing arbitrary-Lagrangian- Eulerian (ALE) techniques. The code is written to address both two-dimensional (2D plane and axisymmetric) and three-dimensional (3D) physics and engineering problems using a hybrid finite element and finite volume formulation to model fluid and elastic-plastic response of materials on an unstructured grid. As shown in Figure 1, ALE3D is a single code that integrates many physical phenomena.

42 ENGINEERING↗

DOE OSTI2017

SIERRA Code Coupling Module: Arpeggio User Manual Version 4.44

The SNL Sierra Mechanics code suite is designed to enable simulation of complex multiphysics scenarios. The code suite is composed of several specialized applications which can operate either in standalone mode or coupled with each other. Arpeggio is a supported utility that enables loose coupling of the various Sierra Mechanics applications by providing access to Framework services that facilitate the coupling. More importantly Arpeggio orchestrates the execution of applications that participate in the coupling. This document describes the various components of Arpeggio and their operability. The intent of the document is to provide a fast path for analysts interested in coupled applications via simple examples of its usage.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

DOE OSTI2016

Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Advanced Technology and Mitigation (ATDM) SPARC Re-Entry Code Fiscal Year 2017 Progress and Accomplishments for ECP.

The SPARC (Sandia Parallel Aerodynamics and Reentry Code) will provide nuclear weapon qualification evidence for the random vibration and thermal environments created by re-entry of a warhead into the earth’s atmosphere. SPARC incorporates the innovative approaches of ATDM projects on several fronts including: effective harnessing of heterogeneous compute nodes using Kokkos, exascale-ready parallel scalability through asynchronous multi-tasking, uncertainty quantification through Sacado integration, implementation of state-of-the-art reentry physics and multiscale models, use of advanced verification and validation methods, and enabling of improved workflows for users. SPARC is being developed primarily for the Department of Energy nuclear weapon program, with additional development and use of the code is being supported by the Department of Defense for conventional weapons programs.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2017

Energy Referencing in LANL HE-EOS Codes

Here, We briefly describe the choice of energy referencing in LANL's HE-EOS codes, HEOS and MAGPIE. Understanding this is essential to comparing energies produced by different EOS codes, as well as to the correct calculation of shock Hugoniots of HEs and other materials. In all equations after (3) throughout this report, all energies, enthalpies and volumes are assumed to be molar quantities.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2017

R-Matrix Codes for Charged-Particle Induced Reactions in the Resolved Resonance Region (3), Summary Report of an IAEA Consultants' Meeting IAEA Headquarters, Vienna, Austria, 28-30 June 2017

A Consultants Meeting was held at the IAEA Headquarters, from 28 to 30 June 2017, to discuss the results of a test exercise that had been defined and assigned to all participants of the previous meeting held in December 2016. Five codes were used in this exercise: AMUR, AZURE2, RAC, SFRESCO and SAMMY. The results obtained from these codes were compared and further actions were proposed. Participants’ presentations and technical discussions, as well as proposed additional actions have been summarized in this report.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2017

Hardening of DOE UQ Codes for Ultra-Deep Proactive Geosteering in Oilfield Service

The primary objective of this project was to harden the DOE uncertainty quantification (UQ) code to solve the inverse problem for the geosteering service in the oilfield drilling and exploration. The primary focus is to verify the feasibility of using QUESO to solve geosteering inverse problems through statistical inversion. QUESO is a software library developed by QUEST, a DOE's Scientific Discovery through Advanced Computing (SciDAC) institute. In the process of drilling a borehole, geosteering is the act of adjusting the well trajectory on the fly to reach the geological targets or to keep the wellbore within desired formation. Geosteering inversion is to infer underground formation structure using measurements such as azimuthal resistivity logging data. The geosteering process relies on the real-time reconstruction of the subsurface earth model, and a robust and efficient inversion algorithm is always necessary to satisfy the job requirement. The new generation of azimuthal resistivity geosteering tool is capable of detecting multiple formation boundaries 20-100 ft away, and the corresponding geosteering inversion problem can be very challenging as many parameters need to be inverted within such a large depth of investigation. The success of the project will support US energy security by helping to advance subsurface modeling and characterization technologies in oilfield applications. An effective subsurface inversion algorithm and the implementation as a software package will promote the usage of ultra-deep azimuthal resistivity measurements, which can help well placement in complicated underground environments, reduce costs and risks of drilling, and increase oil and gas production. It will help medium and small size independent service companies and tool manufactures to enter the more advanced and lucrative directional drilling area. In addition, by hardening the QUESO code and building a transparent interface for easy manipulation of the QUESO functions, we provide a user-friendly tool for a larger scientific and industry communities that rely heavily on modeling and simulation tools for forecasting, parameter studies, design, and decision making. These are all activities that can strongly benefit from meaningful representations of modeling uncertainty supported by our product.

58 GEOSCIENCES↗

DOE OSTI2017

FY2017 Updates to the SAS4A/SASSYS-1 Safety Analysis Code

The SAS4A/SASSYS-1 safety analysis software is used to perform deterministic analysis of anticipated events as well as design-basis and beyond-design-basis accidents for advanced fast reactors. It plays a central role in the analysis of U.S. DOE conceptual designs, proposed test and demonstration reactors, and in domestic and international collaborations. This report summarizes the code development activities that have taken place during FY2017. Extensions to the void and cladding reactivity feedback models have been implemented, and Control System capabilities have been improved through a new virtual data acquisition system for plant state variables and an additional Block Signal for a variable lag compensator to represent reactivity feedback for novel shutdown devices. Current code development and maintenance needs are also summarized in three key areas: software quality assurance, modeling improvements, and maintenance of related tools. With ongoing support, SAS4A/SASSYS-1 can continue to fulfill its growing role in fast reactor safety analysis and help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2015

MELCOR Computer Code Manuals Volume 1: Primer and Users' Guide

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2015

MELCOR Computer Code Manuals

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2018

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Final Technical Report (FTR): Testing the Relationship of Comprehensive Training to Increased Energy Code Compliance

This report outlines the State of Maryland's contribution to a large national study of residential energy efficiency attainment. The study documents baseline practices in new single-family construction, target areas for improvement relative to the state energy code, and quantify related savings potential. The project team collectively implemented a three-phased approach. Data gathered through phases I and III has been submitted to Pacific Northwest National Laboratory (PNNL) for analysis. Findings available to date are published as PNNL technical reports on the DOE Building Energy Codes Program website with final analysis pending publication later in 2018.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2018

User's Guide for RESRAD-ONSITE Code

The RESRAD-ONSITE code is equipped with a user-friendly interface that has many features to facilitate its use and comprehension. The design of the interface provides various navigation options, including entering data, performing calculations, and displaying the results of the calculations. General and context-specific help are available, offering information on editing and viewing the radionuclide database, the definitions of input parameters and their use in the calculations, and the selection and output of the results of the calculations into other applications. Two types of sensitivity analysis are supported by the code, deterministic and probabilistic, which can be used to study the influence of input parameters on the calculated results.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2017

FUEL PERFORMANCE MODELING OF REACTIVITY-INITIATED ACCIDENTS USING THE BISON CODE

The Fukushima Daiichi nuclear accidents in 2011 sparked considerable interest in the U.S. to develop new nuclear fuel with enhanced accident tolerance. Throughout the development of these new fuel concepts they will be extensively modeled using specialized computer codes and experimentally tested for a variety of different postulated accident scenarios. One accident scenario of interest, reactivity-initiated accident, is a nuclear reactor event involving a sudden increase in fission rate that causes a rapid increase in reactor power and temperature of the fuel which can lead to the failure of the fuel rods and a release of radioactive material. The focus of this work will be on the fuel performance modeling of reactivity-initiated accidents using the BISON code being developed at Idaho National Laboratory. The overall goal of this work is to provide the best possible modeling predictions for future experimental tests. Accurate predictive capability modeling using BISON is important for safe operation of these tests and provides a cheaper alternative to the expensive experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Solid Mechanics Code Verification: Contrasting Classical and Manufactured Reference Solutions

Verification results for Sierra/SM using inexact reference solutions have often exhibited unsatisfactory convergence behavior. With an understanding of the convergence behavior for these types of tests, one can avoid falsely attributing pathologies of the test with incorrectness of the code. Simple theoretical results highlight that for an inexact reference solution two conditions must be met to observe asymptotic convergence. These conditions, and the resulting types of convergence behaviors, are further illustrated with graphical examples depicting the exact, inexact reference, and sequence of numerical solutions as vectors (in a function space). A stress concentration problem is adopted to contrast convergence behaviors when using inexact (classical linear elastic) and exact (manufactured) reference solutions. Convergence is not initially attained with the classical solution. Convergence with the manufactured solution indicates the convergence failure with the classical reference did not result from code error and provides insight on how for this problem asymptotic convergence could be attained with the classical reference solution by modifying the computational models.

42 ENGINEERING↗

DOE OSTI2018

DOSECALCX, A Bremsstrahlung Radiation Dose Code for Electron Beam Targets

The Dual-Axis Radiographic Hydrodynamic Test (DARHT) Facility is a two-axis, linear induction electron accelerator designed to produce Bremsstrahlung radiation for imaging hydrodynamic physics within rapidly compressed targets. The radiation dose code DOSECALC, used to obtain approximate results for experiments, was modified to include new materials and renamed DOSECALCX. Calculations were compared with results from the MCNP code. Photon spectra was found to agree visibly and quantitatively while angular dose data was only comparable within an order of magnitude.

43 PARTICLE ACCELERATORS↗