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At least 973 records · Page 54

Fundamental Power Coupler Technical Development - Final CRADA Report

The SLAC ‘Linac Coherent Light Source -II’ project (LCLS II) is a hard X-ray Free Electron Laser now under construction in support of photon-based material science. LCLS II uses a superconducting linac design based on the twenty year-long TESLA- Technology development effort. The TESLA proposal, the European XFEL (EU-XFEL) project, and the ILC design employ a pulsed superconducting linac with roughly 0.5% duty factor. This aspect was changed for LCLS II so that the FEL could be operated continuous-wave (CW) namely with a 100% duty factor. Several critical changes to the linac design were required including RF power distribution, cryogenic heat-load management and cavity preparation. This note lists the changes required to the fundamental power coupler (FPC). All of these changes were made by the SLAC physics and engineering teams and reviewed by the cryomodule design lead (Fermilab) and international partners. Testing was done by the LCLS II cryomodule teams (Jefferson Lab and Fermilab). The SLAC LCLS II project team managed the design and procurement process. The LCLS II fundamental power coupler design is based upon the DESY TESLA Test Facility (TTF-3) design. DESY shared their drawings with LCLS II and SLAC copied and modified them to meet the LCLS II CW operation power and heat-load requirements. LCLS II had an external reviewer evaluate and compare SLAC drawings to EU-XFEL drawings. SLAC used the same industrial partners that EU-XFEL used.

47 OTHER INSTRUMENTATION↗

Development of a LOCA Experimental Benchmark for BISON - Milestone L2:FMC.P15.08

Since 2013 CASL has been planning and working towards a PWR LOCA Challenge Problem. Since BISON is the CASL fuel rod simulation platform, it plays a central role in that activity. The goal of the FY-17 CASL milestone reported here was to develop an Experimental Benchmark for LOCA analysis using BISON. Activities encompassed both code development and validation. This report provides a current snapshot of BISON’s capability for LOCA behavior including: 1) a summary of code extensions to facilitate accident analysis, 2) a series of separate effects tests, and 3) initial code validation to complex integral rod LOCA behavior. Conclusions from each of these main activities are summarized below. The key material and behavior models required to address transient high-temperature phenomena occurring during LOCAs in a standard PWR have now been implemented in BISON. These apply specifically to UO2 fuel, Zircaloy cladding and water coolant. During FY-17 important new capability to address axial UO2 fuel relocation (for 1.5D geometry) and account for oxidation energy deposition in cladding, were included. Planned future development efforts include extending the axial fuel relocation model from 1.5D to 2D/3D and improving BISON’s fission gas release model to include transient gas release associated with the high burnup structure (HBS) in high burnup fuel. Additionally, investigation of potential anisotropic creep behavior in Zircaloy cladding is planned, assuming sufficient experimental data are available to support this effort. A substantial number of separate effects validation cases (42 tests from 3 experimental series) have been completed to compare BISON predictions to measured ballooning and burst behavior for Zircaloy cladding. Such experiments include a wide variety of pressures, temperatures and loading rates. In general, BISON predictions of burst temperature, pressure and burst time are very reasonable. For one experimental series, however, involving both very high temperatures and strain rates, BISON systematically over-predicts the cladding hoop strain. Investigation of this discrepancy is an important and planned activity. With another experiment an effort was made to investigate 3D cladding response due to an azimuthal temperature variation. Results indicated 3D effects are potentially important in fuel rod analysis during LOCAs and will be further investigated in the future. BISON validation to a series of integral fuel rod experiments has also been completed. These experiments involve all fuel and cladding phenomena relevant to LOCA conditions, and can include complexities associated with irradiated fuel relative to fresh fuel. Such experiments also generally include complex thermal-hydraulic boundary conditions. Four experiments (6 rods) have been considered to date including simulated fuel (ZrO2) and both fresh and high-burnup UO2. Test rods ranged from rodlets to full length commercial PWR fuel rods. As with the separate effects experiments, BISON predictions of burst temperature, pressure and burst time are generally very reasonable. Comparisons to cladding peak strain and rod outer diameter axial profiles are less satisfactory, and identify material models and possibly modeling approximations (e.g., 2D-RZ vs 3D geometries) requiring additional investigation. Validation of BISON for integral rod LOCA behavior is by no means complete, with additional cases planned.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

ATDM ASD report to ECP FY18Q4

Kokkos and DARMA teams worked together on several significant proposals for the C++ standards body, including the MDSpan and Exectutors proposals. This work has potential for long-term impact into the language standard reflecting core abstractions needed for our HPC programming and development efforts. Kokkos team collaborating with AMD via the Path Forward effort. This is advancing Kokkos support for [AMD] backends, necessary for future exascale/HPC architectures.

97 MATHEMATICS AND COMPUTING↗

Accelerator Physics Center at Fermilab : History and Accomplishments (2007-2018)

Fermilab’s Accelerator Physics Center (APC) was created in June 2007 with mission to carry out research and development to keep the US leading high-energy physics laboratory at the forefront of accelerator science, technology and facility operation. In support of the FNAL high-energy physics research mission, APC scientists and enginners conducted accelerator R&D aimed at next-generation and beyond accelerator facilities; provided accelerator physics support for existing operational programs and the evolution thereof; trained accelerator scientists and engineer and established experimental programs for a broad range of accelerator R&D that can be accessed by both Fermilab staff and the US and world HEP community. APC was a center-place for in-depth design, research and development efforts which allowed the Laboratory to make intelligent decisions on the ILC in the US, on the Muon Collider, as well as originate projects like PIP-II (through the Proton Driver/Project-X work), LHC-AUP (via LARP) and the IOTA/FAST R&D facility. APC was also the birthplace and host of many national and international collaborations and several educational/training programs in beam physics resulted in 27 PhD theses. In the Fall 2018, following an important milestone of the first beam circulating in the IOTA ring, the APC was re-organized into several departments in Accelerator Division, charged to carry out the accelerator physics research with IOTA/FAST beams and making it relevant for future upgrades of the Fermilab accelerator complex.

43 PARTICLE ACCELERATORS↗

Integrated FHR technology development: Tritium management, materials testing, salt chemistry control, thermal hydraulics and neutronics, associated benchmarking and commercial basis

This report summarizes the results of a 3-year Integrated Research Project (IRP) sponsored by the U.S. Department of Energy to address major technical challenges in the development of the Fluoride-salt-cooled High-temperature Reactor (FHR). The IRP universities included the Massachusetts Institute of Technology (MIT), the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The FHR is a new reactor concept less than 20 years old that combines (1) a clean fluoride salt coolant, (2) the graphite-matrix coated-particle fuel originally developed for High- Temperature Gas-Cooled Reactors and (3) passive decay heat removal systems from sodium fast reactors. The base-line design is a pebble-bed FHR that delivers heat between 600 to 700°C with a 7 Li 2 BeF 4 (flibe) salt coolant to the power cycle or industrial users. The Executive Summary describes what was accomplished and summarizes major conclusions. The main report contains stand-alone sections in each major area of work. Highlights are summarized below. Tritium. The coolant salt produces tritium necessitating understanding and control of tritium. Experimental work included measuring hydrogen uptake on carbon, tritium uptake during irradiation of graphite in 700°C salt and the permeability of tritium through metals with and without coatings. An FHR systems model was developed to predict tritium behavior and enable design of tritium control systems. The leading candidate for tritium removal from the flibe salt is carbon bed outside the reactor core. Thermal hydraulics and Neutronics. A 10-meter high FHR integral effects test facility, using Dowtherm A®, has been built and used to simulate the integral response of a representative FHR to improve understanding of FHR thermal hydraulics and validate thermal hydraulics codes used for the FHR. Neutronic analysis included code-to-code benchmarking, multiphysics modeling, uncertainty analysis for nuclear data uncertainties and shielding analysis. Twisted-tube heat exchangers were examined that may provide significant cost reductions because of their improved heat-transfer performance with liquid salts. The impacts on safety and design caused by uncertainties in physical properties of the salt were assessed with recommendations on where added physical property measurements are needed. Work began on understanding radiative heat transfer that becomes important in salt systems at higher temperatures. Materials. Static corrosion tests were conducted on graphite and materials in 700°C salt -- both in the UW laboratory and in the MIT reactor. In addition to static corrosion testing, a circulating salt corrosion loop was built and is operating. Methods to measure salt properties including redox were developed. The presence of carbon increases corrosion. With good redox control, the evidence indicates stainless steel can be used as a material of construction. Benchmarking. An additional series of workshops where held to enable benchmarking and comparison of results between different investigators. Commercialization. The FHR delivers higher average-temperature (~650°C) heat to the power cycle or industry than other reactors. This enables power cycles with lower-cost heat storage for variable electricity to the grid while the reactor operates at base-load -- providing a major competitive advantage by maximizing revenue from electricity sales. In October, 2016, a new startup company named Kairos Power was formed, and opened its first office in Jack London Square, Oakland, California, in January 2017. Kairos Power exists largely due to the work that was accomplished by this IRP and the previous DOE-funded FHR IRP that performed research on FHR technology from 2012 to 2015. In September, 2018, as this IRP is concluding, Kairos Power was staffed at 60 full time employees, working to develop technology derived from this IRP. Many of these employees worked with the IRP before joining Kairos Power 1 (https://kairospower.com). Other students and researchers from this IRP have joined other startup companies, including the Terrapower Molten Chloride Fast Reactor and the Terrestrial Energy development efforts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Preparation of Four Large-format Hot Isostatically Pressed I-AgZ Waste Form Samples for Performance Testing

Hot isostatic pressing (HIP) is being considered as a process to directly convert 129 I-bearing materials to a durable radiological waste form. Zeolites, specifically silver-exchanged mordenite (AgZ), have been studied extensively as potential radioactive iodine sorbents and will contain 129 I as chemisorbed AgI, a relatively insoluble compound with a high melting point. Oak Ridge National Laboratory has conducted several recent studies on the HIP of both I–AgZ and other iodine-bearing zeolites, resulting in an inventory of samples prepared with a suite of zeolites with varied iodine incorporation methods, pressing conditions, and sizes. These samples have been made available to researchers at other national laboratories in support of test method development to assess the stability of heterogenous iodine-bearing waste forms. The work described here generated four large-format HIPed I–AgZ samples. These samples are 1.5 in. diameter and 2 in. length and contain 49 g of I–AgZ. Two samples were prepared at an iodine loading of 64 mg I/g I– AgZ, and two samples were prepared at an iodine loading of 135 mg I/g I–AgZ. These two sets of duplicate samples will be made available upon request in support of iodine waste form durability method development efforts.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program (Rev. 2)

The beyond design basis accident at Fukushima Daiichi in March 2011 and the Three Mile Island accident in 1979 imply that severe nuclear accidents may occur at somewhat higher frequencies than previously predicted, and that the financial liabilities of such accidents can cripple a utility. While current fuel is more than adequate for design basis accidents, the potential for harm from beyond design basis accidents (BDBA) has been a driver for Westinghouse development efforts since 2003. After Fukushima, other fuel vendors also began pursuing new fuel materials that can provide significant increases in the time, enabling reactor operators to respond to unforeseen events before significant releases of fuel materials and fission products occur. In order to be commercially successful, accident tolerant fuel (ATF) products must also provide significant fuel cost savings, operating cost improvements as well as safety improvements. This report documents the engineering, design and testing that occurred from October 1, 2016, through September 30, 2018, by the Westinghouse team. In addition, it includes: 1. The results of modeling and simulation work completed to date that supports technical feasibility of the ATF concept in meeting the goals of the enhanced ATF program including a core level analysis predicting how an existing LWR will perform when loaded with ATF. 2. Regulatory Analysis showing path to LTA/LTR acceptance. 3. Business Analysis showing commercial viability of the ATF concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Enhancements to the Subplane Capability in MPACT

In recent years, much of the development effort for the Virtual Environment for Reactor Applications (VERA) has been devoted to improving performance and targeting industry-class computers. This includes reducing both runtime (improving efficiency) and core-count requirements to enable simulation on smaller machines. In MPACT, in particular, many aspects of the neutronic iteration strategy have been evaluated and improved, such as the subgroup and method of characteristics (MOC) runtimes, and many efforts have focused on improving the coarse mesh finite difference (CMFD) runtime, which has become a dominant component of the runtime as other aspects have been improved.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Effluent Management Facility Evaporator Bottoms: Waste Streams Formulation and Waste Form Qualification Testing

This report describes the results from grout formulation and cementitious waste form testing performed by Pacific Northwest National Laboratory for Washington River Protection Solutions, LLC (WRPS). These results are part of a screening test that investigates grout formulations proposed for encapsulating a wide range of compositions predicted to be present in evaporator bottoms wastes from the Hanford Effluent Management Facility (EMF). This work supports the technical development need for alternative treatment and disposition paths for the EMF evaporator bottoms waste and future direct-feed low-activity waste (DFLAW) operations at the Hanford Site. High-priority activities included simulant production, grout formulation, and cementitious waste form testing. The work contained within this report relates to waste form development and testing, but does not directly support the 2017 Integrated Disposal Facility (IDF) performance assessment (PA). However, this work contains information useful for future PA updates [beyond fiscal year (FY) 2017] and future waste form development efforts. These analytical results can be used by (i) cementitious waste form scientists to further the understanding of cementitious leach behavior of contaminants of concern (COCs), (ii) decision makers interested in off-site waste form disposal, and (iii) the U.S. Department of Energy, their Hanford Site contractors, and stakeholders as they continue to assess the IDF PA program at the Hanford Site. The reported results help fill existing data gaps, support final selection of a cementitious waste form for the EMF evaporator bottoms waste, and improve the technical defensibility of long-term waste form risk estimates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Hierarchical Task-Data Parallelism using Kokkos and Qthreads

This report describes a new capability for hierarchical task-data parallelism using Sandia's Kokkos and Qthreads, and evaluation of this capability with sparse matrix Cholesky factorization and social network triangle enumeration mini-applications. Hierarchical task-data parallelism consists of a collection of tasks with executes-after dependences where each task contains data parallel operations performed on a team of hardware threads. The collection of tasks and dependences form a directed acyclic graph of tasks - a task DAG. Major challenges of this research and development effort include: portability and performance across multicore CPU; manycore Intel Xeon Phi, and NVIDIA GPU architectures; scalability with respect to hardware concurrency and size of the task DAG; and usability of the application programmer interface (API).

97 MATHEMATICS AND COMPUTING↗

Multiple Degradation Mechanisms in Reinforced Concrete Structures, Modeling and Risk Analysis

The overarching goal of this project is to complement ongoing Department of Energy (DOE) Light Water Reactor Sustainability (LWRS)-funded Grizzly concrete modeling development efforts by providing improved multi-physics models for incorporation into the Grizzly and BlackBear codes. Objectives of this work identified at the outset of this project include: 1. Coupling between mechanical damage and transport processes. The mechanical damage will be characterized by nonlinear mechanical constitutive models appropriate for concrete used in nuclear power plants (NPPs). 2. Improved representation of the coupling among the transport process, building on the current moisture and thermal transport models in Grizzly. The parameters for coupling among heat and mass transport in the multi-physics model will be experimentally determined. 3. Coupling among different scale levels of concrete constituents, fine and coarse aggregates, cement paste, hydration products, and pore structure. 4. Probabilistic analysis of the random nature of the heterogeneous concrete structure and the environmental factors (ambient temperature and humidity) and their multiple effects on concrete deterioration. 5. Benchmark verification, validation and uncertainty quantification of the thermo-hygro- chemo-mechanical (THCM) concrete formulation. The end goal of this work was to provide a comprehensive and robust simulation capability that can be applied at the engineering scale for analysis of realistic deterioration scenarios in NPP concrete structures.

36 MATERIALS SCIENCE↗

Baseline Characterization Database Verification Report – IG-110 Billet 10X69

The purpose of this engineering calculations and analysis report (ECAR) is to present the data being collected in the Baseline Graphite Characterization program. This program is directly tasked with supporting Idaho National Laboratory’s (INL’s) research and development efforts in the Advanced Reactor Technologies (ART) program. This program populates a comprehensive database that will reflect the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical- and mechanical-property information being collected will be transferred to the Nuclear Data Management and Analysis System (NDMAS), and that database will help populate the handbook of property data available to member nations of the Generation-IV International Forum. The transfer of these data from the applicable technical lead to the dissemination databases available to other end users requires a full review of test procedures and data-collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents that analysis for IG-110, Billet 10X69 and facilitates the release of the associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Baseline Characterization Database Verification Report – 2114 Billet A20570

The purpose of this engineering calculations and analysis report (ECAR) is to present data collected in the Baseline Graphite Characterization Program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Advanced Reactor Technologies (ART) Program. This program populates a comprehensive database that reflects the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical- and mechanical-property information being collected will be transferred to the Nuclear Data Management and Analysis System (NDMAS), and that database will help populate the handbook of property data available to member nations of the Generation-IV International Forum. Transfer of these data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data-collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents the analysis for 2114 Billet A20570 and facilitates release of associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

IAEA ITC28 Project Report. 28th International Training Course on the Physical Protection of Nuclear Materials and Nuclear Facilities

Sandia National Laboratories (SNL) has hosted the International Training Course on the Physical Protection of Nuclear Materials and Nuclear Facilities since 1978. This course is the flagship training course of the International Atomic Energy Agency (IAEA). On behalf of the National Nuclear Security Administration (NNSA), SNL manages, develops, and coordinates all course materials, and works closely with the IAEA to arrange all logistical details for the course. ITC-28 incorporated several new approaches based on feedback and experience with ITC-27 and earlier versions of the course. For ITC-28, an addition to the Integrated Security Facility (ISF) at SNL was a mock reactor hall with a mock reactor pool. Other facilities at the ISF include a mock processing facility, material receiving area, and central alarm station. The physical protection system at the ISF—an area that formerly housed Category I nuclear material—provides many opportunities for hands-on, real world training in the design and evaluation of a physical protection system (PPS). This document provides a brief description of ITC-28, including a summary of lessons learned and key recommendations for future development efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Baseline Characterization Database Verification Report – PCEA Billet 02S8 7

The purpose of this Engineering Calculations and Analysis Report is to present the data being collected in the Baseline Graphite Characterization program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Advanced Reactor Technologies (ART) Program. This program is populating a comprehensive database that will reflect the baseline properties of nuclear grade graphite with regard to individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the Nuclear Data Management and Analysis System (NDMAS), and that database will help populate the handbook of property data available to member nations of the Generation IV International Forum. Transfer of these data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents the analysis for PCEA billet 02S8 7 and facilitates release of associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Baseline Characterization Database Verification Report – PCEA Billet XPC01D3-36

The purpose of this Engineering Calculations and Analysis Report is to present the data being collected in the Baseline Graphite Characterization program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Advanced Reactor Technologies (ART) program. This program is populating a comprehensive database that will reflect the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the NGNP Data Management and Analysis System (NDMAS), and from that database will help populate the handbook of property data available to member nations of the Generation IV International Forum (GIF). The transfer of this data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents that analysis for PCEA billet XPC01D3-36 and facilitates the release of the associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Baseline Characterization Database Verification Report – IG-110 Billet 08-9-052-7

The purpose of this Engineering Calculations and Analysis Report is to present the data being collected in the Baseline Graphite Characterization program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Advanced Reactor Technologies (ART) program. This program is populating a comprehensive database that will reflect the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical and mechanical property information being collected will be transferred to the NGNP Nuclear Data Management and Analysis System (NDMAS), and from that database will help populate the handbook of property data available to member nations of the Generation IV International Forum (GIF). The transfer of this data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents that analysis for IG-110 billet 08-9-052-7 and facilitates the release of the associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Wet Waste Hydrothermal Liquefaction and Biocrude Upgrading to Hydrocarbon Fuels (2019 State of Technology)

Each year, the DOE Bioenergy Technologies Office (BETO) assesses progress in their research and development efforts toward sustainable production of renewable fuels. Technical and cost targets were previously established for the wet waste hydrothermal liquefaction and biocrude upgrading pathway and summarized in a design report. The present report summarizes the research and associated techno-economic analysis (TEA) in support of the 2019 state of technology (SOT) assessment for this pathway. Data from Pacific Northwest National Laboratory’s Conversion hydrothermal liquefaction (HTL) program for wet waste was used to update the pathway techno-economic analysis (TEA) for the fiscal year 2019 State of Technology (2019 SOT). An overview of the current process model, experimental data and plant economics for the SOT is presented.

09 BIOMASS FUELS↗