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DOE OSTI2020

Combined Environments: Driving multiple loads on Z

Pulsed power drivers such as the Z generator of Sandia National Laboratories typically deliver high current (>20MA) to single experiments. This project is intended to develop and assess ways to simultaneously drive multiple targets on a single pulsed power driver (specifically a neutron and an x-ray producing target driven in a single experiment). The combined x-ray/neutron environment produced will then be used to investigate potential synergistic effects in integrated circuits. A pre-requisite for being able to design and study multiple targets on Z is first adapting simulation tools to be able to model them effectively. This will enable us to assess the tradeoffs between the different ways multiple targets can be combined, and to better understand how existing and future pulsed power machines can be used to generate combined testing environments. This report is limited to documenting the initial development of a parallel load modeling capability that is presently being applied to design experiments to produce combined neutron/x-ray environments on Z.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Proceedings of RIKEN BNL Research Center Workshop: Jet Observables at the Electron-Ion Collider (Volume 136) [Slides]

As the realization of an Electron Ion Collider (EIC) moves forward, efforts from the nuclear physics community continue to grow. In addition to the ongoing detector R&D efforts, plans for novel analysis topics must be demonstrated to aid in the detector designs, so that we can maximize the physics output of the EIC. This relies on input from both the experimental and theory communities. The aim of this workshop is to gather experts as well as those with a developing interest in the EIC so that theorists and experimentalists with experience measuring jets in a variety of hadronic collision systems can discuss the possible advantages and challenges of making measurements of jets in e+p and for the first time ever jets in e+A collisions at the EIC. In the time since the EIC white paper was written in 2011, there has been a growing recognition that jet observables could be a powerful probe of many of the physics topics which will be addressed by the EIC. Recent years have seen a multitude of both theoretical and experimental papers exploring the utility of jets for topics as diverse as determining the hadronic structure of photons, studying the 3D structure of the nucleon, and characterizing the properties of the matter found in nuclei. The aim of this workshop is to provide a forum for both theorists and experimentalists to discuss the possible advantages and challenges of jet measurements in e+p and e+A collisions at the EIC, learn about the status of necessary simulation tools, consider requirements on detector performance, and propose new ideas for jet observables and measurements. This proceedings report is a compilation of slides from the presentations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

Three-Phase Modeling in Sockeye

The heat pipe simulation tool Sockeye is given the capability to allow the presence of the solid phase of the heat pipe working fluid, which is necessary for modeling transients to and from room temperature, since microreactor designs feature high-temperature heat pipes employing metal working fluids. The liquid phase equations are converted to equations for the mixture of liquid and solid phases, with a simple latent heat relation determining the fractions of these phases. This capability is introduced while preserving equivalent solutions for the liquid-vapor regime. Theory is presented, along with numerical examples to illustrate the propagation of a melt front and startup from the solid state.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2020

FY20 SAM Code Developments and Validations for Transient Safety Analysis of Advanced non-LWRs

The System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. This report summarizes major progress in SAM code development, capability enhancements, demonstration, and validation to support transient safety analysis of advanced non-LWRs. Rapid developments continued in fiscal year 2020 (FY20) to support various needs of the advanced reactor community, especially the NRC and industry on the licensing safety analysis of advanced reactor designs. Significant code changes were made to provide various capability enhancements, bug fixes, and user friendliness improvements. Major code updates are summarized in Section 1, while four important enhancements are detailed in Sections 2-5, including a multi-dimension flow model; reactivity feedback and decay heat models; control and trip system modeling, and additional fluid and solid thermophysical property models. Code validation activities in FY20 include using test data from the Fast Flux Test Facility (FFTF), the High Temperature Test Facility (HTTF), and several separate effects test facilities for pebble-bed modeling.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Xyce XDM Netlist Translator User Guide (Version 2.2)

This manual describes the installation and use of the Xyce™ XDM Netlist Translator. XDM simplifies the translation of netlists generated by commercial circuit simulator tools into Xyce-compatible netlists. XDM currently supports translation from PSpice and HSPICE netlists into Xyce™ netlists.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2020

NEAMS Workbench Status and Capabilities

The mission of the US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is to develop, apply, deploy, and support state-of-the-art predictive modeling and simulation tools for the design and analysis of current and future nuclear energy systems. This is accomplished by using computing architectures that range from laptops to leadership-class facilities. The NEAMS Workbench is a new initiative that will facilitate the transition from conventional tools to highfidelity tools by providing a common user interface for model creation, review, execution, output review, and visualization for integrated codes. The Workbench can use common user input, including engineering-scale specifications that are expanded into application-specific input requirements through the use of customizable templates. The templating process can enable multifidelity analysis of a system from a common set of input data. Additionally, the common user input processor can provide an enhanced alternative application input that provides additional conveniences compared with native input, especially for legacy codes. Expansion of the integrated codes and application templates available in the Workbench will broaden the NEAMS user community and will facilitate system analysis and design. Current and planned capabilities of the NEAMS Workbench are detailed herein.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

State-of-the-Art Reactor Consequence Analysis Project: Sensitivity Analysis Using Dakota

The State-of-the-Art Reactor Consequence Analyses (SOARCA) project has focused on best estimate analyses and uncertainty analysis for postulated accidents at specific nuclear power plants. The consequences of these accidents are estimated using the simulation tools MELCOR and MACCS. To understand which uncertain input variables are important to determining these consequences, analysts have performed sensitivity analyses. The tool used to perform these sensitivity analyses in previous SOARCA work, CompModSA, is no longer supported. Therefore, the current work focuses on migrating these analyses to another tool and evaluating its performance. Dakota, which is a tool developed at Sandia National Laboratories, is used in this work. Sensitivity results are created for three analyses from the SOARCA Surry UA. Though CompModSA and Dakota vary slightly in their algorithms and implementation, their sensitivity results generally agree, which gives confidence in the Dakota approach and increases confidence in the original analyses. It is likely that this methodology is extendable to the rest of SOARCA analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

FEST: Facility Energy Saving and Securing Technology Using Multi-Source Data: (Milestone 2 Report)

This project aims to demonstrate three technologies developed in-house at LLNL and University of Michigan-Dearborn (UMD), at a military site, including i) Grid Data Crossing (called GriD-Xing) for increasing smart meter data usability, ii) Facility Energy Optimization (called Facility E-GO) for improving facility energy efficiency in both operation and planning perspectives, and iii) Co-simulation tool (called Co-Sim) for enhancing smart meter and energy facility networks resilience and security. In this report, Millstone 2 - Integration A: Integrate GriD-Xing and facility optimization tools is documented.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2021

Increasing the Usability of Sockeye: Implementing Customer-Driven Modeling Improvements, Features, and Documentation of Sockeye

A number of tasks for the heat pipe simulation tool Sockeye were identified based on customer needs and included both user documentation and code development: (a) Implement saturation-dependence in the capillary pressure model; (b) Provide an analytic utility to analyze the frozen startup limit; (c) Demonstrate custom friction closures; (d) Develop examples for different types of heat transfer sources and boundary conditions; (e) Demonstrate the capillary and sonic limits; and (f) Provide discretization details in the theory manual. These tasks were all completed, and results are given where applicable.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2021

Improving Fission Products at CARIBU: Near Field Detection (Q2/FY21 Quarterly Progress Report)

The 95 Zr manuscript describing our experimental approach to measure branching ratios of long-lived fission products has been completed and successfully published in NIM A. This article is an important milestone in this effort and will serve as a great reference in any future reports describing results of other measurements we plan to perform. We have also expended our team, and Daniel Hoff joined us as a postdoc. We have been training him on how to use our analysis and simulation tools, as well as bringing him up to speed on the technical details of our experiment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2021

SAS4A/SASSYS-1 Lead Verification Test Suite

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over forty-five years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. Although SAS4A/SASSYS-1 was developed to support the analysis of any liquid-metal-cooled nuclear reactor, it has primarily been utilized to design and analyze Sodium Fast Reactors (SFRs). As a result, most of the verification basis for SAS4A/SASSYS-1 utilizes sodium as a coolant and geometry descriptions that are typical of a SFR facility. In this work, new verification problems are defined to extend the applicability of the SAS4A/SASSYS-1 verification basis into Lead Fast Reactor (LFR) design space. A review of the existing SAS4A/SASSYS-1 SFR verification test suite is performed to determine the number of test cases that are SFR specific. SFR specific test cases are replicated using lead as a coolant and system layouts that are representative of LFRs. This report provides definitions and reference solutions for seven new test cases that scope lead-specific features and capabilities of the software. For a verification test to be considered acceptable, comparisons of analytical solutions to SAS4A/SASSYS-1 predictions must produce negligible or justifiable errors. All LFR specific test cases are found to be acceptable.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Pilot Testing of a Modular Oxygen Production System Using Oxygen Binding Adsorbents (Final Report)

In this project, RTI and Air Liquide have focused on the development of innovative oxygen separation materials and technologies based on reversibly binding of oxygen to enable smaller and cheaper air separations. In alignment with that object the team has had multiple achievements to advance the technology including: Synthesized and characterized novel M-CoSalen (RTIO2Sorb-1) material which showed a dynamic oxygen adsorption capacity in excess of 1wt%. Developed synthesis routes for RTIO2Sorb-1 with commercially relevant techniques and produced batches of 0.25kg and over 4 kg of total synthesis. Synthesized and characterized extrudate forms of the RTIO2Sorb-1 with relevant crush strength and maintained dynamic oxygen adsorption capacity in near 1wt%. Developed techniques for forming structured beds in fiber shapes with conventional materials. Developed a model for O 2 sorption processes to analyze various bed configurations and cycle parameters. Studied classic O 2 solid sorbents reported in the literature to learn oxygen binding mechanisms. Design a 10 kg/day O 2 VPSA pilot system with 2 or 4-bed operation. Completed a techno-economic analysis based on the experimental results and modeling. The overall objective of the project was to design, fabricate, and test a modular O 2 production system and perform a techno-economic analysis (TEA) after testing to determine the cost-benefit of the advanced modular air separation system. The goals of this technology development project were to achieve a bed-size factor (BSF) of less than 600 lb-adsorbent/TPD O 2 (ton/day O 2 ) (as compared with the state-of-the-art BSF of 850), and O 2 -purity greater than 95% at a cost that is projected to be equivalent or lower than the current state of the art (SOTA), commercially available large-scale cryogenic air separation systems. To achieve these goals and objective, the project team executed on (1) oxygen binding adsorbent optimization and scale up, (2) adsorbent material formation process studies to form the adsorbent material into structured beds for rapid pressure swing adsorption (PSA) cycles with low pressure drop, fast mass transfer, and low attrition, (3) cycle development studies to optimize the PSA process, and (4) develop simulation tools for rapid cycle modeling and numerical evaluation/optimization.

20 FOSSIL-FUELED POWER PLANTS↗

DOE OSTI2021

Qualification of System-Level Advanced Reactor Safety Analysis Software for Lead Systems: Final CRADA Report

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over forty-five years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. Although SAS4A/SASSYS-1 was developed to support the analysis of any liquid-metal-cooled nuclear reactor, it has primarily been utilized to design and analyze Sodium Fast Reactors (SFRs). As a result, most of the qualification basis for SAS4A/SASSYS-1 has utilized sodium as a coolant and geometry descriptions that are prototypic of SFR configurations. In this project, which partnered with Westinghouse Electric Company, LLC, the initial foundation for a qualification basis centered on prototypic pool-type lead-cooled systems has been established, where the end goal is to extend support for utilization of SAS4A/SASSYS-1 in Lead Fast Reactor (LFR) licensing or authorization. This project included three fundamental technical tasks: LFR V&V test suite development (Task 1); qualification support for LFRs (Task 2); and LFR modeling capabilities evaluation and improvement (Task 3)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Status of EM Pump Modeling Capability for VTR

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. Most recently, SAS4A/SASSYS-1 has been selected as the safety analysis software for the Versatile Test Reactor (VTR), a new materials testing Sodium Fast Reactor (SFR) to be built by the Department of Energy. In order to support analysis of the VTR, which will utilize ElectroMagnetic (EM) pumps as the primary coolant pumps, the development of new EM pump modeling capabilities has been initiated for SAS4/SASSYS-1. The new physics-based EM pump model captures systems-level transient performance based on equivalent circuit theory and is highly flexible to model a wide range of electromagnetic pumps of the three-phase linear induction type subject to changes in power supply and plant conditions. Within the equivalent circuit framework, two sub-models are developed to provide options for analysts both with and without a detailed geometric characterization of their pump such that minimal information is required to achieve a realistic working model, but specific details of a particular pump design may be captured if sufficient information is provided. In addition, heat generated from pump operation and its transport through the plant system can be modeled to better capture the behavior of certain transients. This paper will outline the model and derive the governing equations of the new EM pump model. An initial comparison between the simple EM pump model and experimental data is made. This comparison shows that the equivalent circuit model can capture pump performance, with greater accuracy near the rated conditions. In order to demonstrate the transient behavior of the new model, a simple transient with an EM pump replacing the centrifugal pumps in the Advanced Burner Test Reactor is included.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Present Early Test Results at SECARB Industry Briefing [Slides]

Contents include the following presentations: 1) SECARB Phase III Early Test Cranfield, MS; 2) SECARB Phase III Early Test at Cranfield, March 9, 2010; 3) Update on Results of SECARB "Early" Test of Monitoring Large Volume Injection at Cranfield; 4) Progress Update SECARB "Early" Test Monitoring 3.5 Million Tons at Cranfield, March 7, 2012; 5) Cranfield Geophysical Monitoring Status and Results, March 8, 2012; 6) Large Volume CO 2 Injection at Cranfield, Early Field Test of the SECARB Phase III–Near-Surface Monitoring, March 7, 2012; 7) Shallow Groundwater Study at the Cranfield Site, March 12, 2013; 8) Subsurface Monitoring Studies–Results and Integration, March 12, 2013; 9) Review of the Petrophysics for the Cranfield Phase III Project, March 12, 2013; 10) Geomechanics at Cranfield, March 5, 2014; 11) Gulf Coast Carbon Center Near-Surface Update, March 4-5, 2014; 12) Monitoring Design Best Practices, March 5, 2014; 13) Assessing Impacts of CO 2 Leakage on Groundwater Quality and Monitoring Network Efficiency–Case Study at a CO 2 -EOR site, March 11, 2013; 14) Near-surface Stratigraphy and Geophysics, March 11, 2015; 15) Update on Reservoir Modeling and Final Testing and P&A Underway, March 11, 2013; 16) SECARB Early Test Retrospective; 17) Fluid Source Attribution at CCS and CO 2 -EOR Sites, March 8, 2016; 18) SECARB "Early Test" at Cranfield, March 8-9, 2017; 19) Analysis of Seismic Time-Lapse Observations at Cranfield; 20) SECARB Phase III Cranfield Project (Early Test), March 8, 2018; 21) Enhanced Analytical Simulation Tool (EASiTool) Dynamic Capacity Estimation; 22) Net Carbon Negative Oil Mass Accounting–Cranfield Case Study; 23) Knowledge Sharing; 24) Advancing Toward Commercialization Final Stages of the Early Test; 25) CO 2 EOR Feasibility in SECARB Region Limited by CH 4 ; 26) Understanding of CO2 Migration and Entrapment–Cranfield Site

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2021

Multiscale Modeling of Cr 2 O 3 -doped UO 2 Creep and Fracture

The pursuit of UO 2 fuel with various dopants (e.g., Cr 2 O 3 , Al 2 O 3 ) for improved accident tolerance due to larger grain sizes and suspected reductions in fission gas release and creep requires advanced modeling and simulation tools. These tools enable rapid multiscale development and understanding of material behavior in regimes where experimental measurements may be lacking. This work builds upon previous multiscale modeling efforts for fission gas behavior of Cr 2 O 3 -doped UO 2 to begin looking at the effects of dopants on mechanical properties with a particular focus on creep and fracture. The results indicate that dislocation climb is the most dominate creep mechanism at almost all temperatures with sensitivity analyses further providing evidence in support of that claim. Limited experimental data from the Advanced Fuels Campaign (AFC) program in the U. S. Department of Energy on tensile strength of Cr 2 O 3 -doped UO 2 indicates that fracture may be less severe in doped specimens. The results on creep and fracture are preliminary given the large uncertainty associated with the models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Xyce™ XDM Netlist Translator User Guide (Version 2.4)

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2021

U.S. Domestic Pebble Bed Reactor: Security-by-Design

U.S. nuclear power facilities face increasing challenges in meeting dynamic security requirements caused by evolving and expanding threats while keeping cost reasonable to make nuclear energy competitive. The past approach has often included implementing security features after a facility has been designed and without attention to optimization, which can lead to cost overruns. Incorporating security in the design process can provide robust, cost effective, and sufficient physical protection systems. The purpose of this work is both to develop a framework for the integration of security into the design phase of High Temperature Gas Reactors (HTGRs) that utilize pebble-based fuels and increase the use of modeling and simulation tools to optimize the design of physical protection systems. Specifically, this effort focuses on integrating security into the design phase of a model HTGR that meets current Nuclear Regulatory Commission (NRC) physical protection requirements and providing advanced solutions to improve physical protection and decrease costs. A suite of tools, including SCRIBE3D©, PATHTRACE© and Blender© were used to model a hypothetical, generic domestic HTGR facility. Physical protection elements such as sensors, cameras, barriers, and guard forces were added to the model based on best practices for physical protection systems. Multiple outsider sabotage scenarios were examined with four-to eight adversaries to determine security metrics. The results of this work will influence physical protection system designs and facility designs for U.S. domestic HTGRs. This work will also demonstrate how a series of experimental and modeling capabilities across the Department of Energy (DOE) Complex can impact the design of and complete Safeguards and Security by Design (SSBD) for SMRs. The conclusions and recommendations in this document may be applicable to all SMR designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗