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DOE OSTI2018

UW Translational Imaging Biomarker and Therapeutic Development

The development and validation of radiotracers is a critical step in the clinical translation of molecular imaging and theranostics to support personalized medicine. The proposed program is to support such efforts by an interdisciplinary team of investigators at the University of Washington (UW) and Fred Hutchinson Cancer Research Center (FHCRC) and to provide opportunities for trainees to learn the spectrum of translational research in imaging biomarkers and therapeutic development. The proposed translational research projects include "Preclinical Evaluation of PET Imaging for Use in the Translation of Targeted Alpha Therapy of Blood-Related Cancers to Clinical Studies" (Project 1-Wilbur, leader); "Hypoxia Imaging: Radiopharmaceuticals, Preclinical Studies, Mechanisms and Clinical trials" (Project 2-Krohn, leader); and "Diagnostic Applications of Radiolabeled Biologics in Solid Tumors" (Project 3-Minoshima, leader). These projects use radiotracers with long half-lives, such as F-18, Zr-89, I-124, and At-211, that can potentially be handled commercially if translation research is successful. The proposed research method is to use an interdisciplinary 'team' of expert investigators and trainees working on a specific project for 1 or 2 years along the timeline and milestones set by the Program Directors. Each team includes 1) a clinician expert; 2) a radiochemistry expert; 3) an imaging or radiation therapy expert; 4) a physicist; 5) trainees; and 6) a UW technology transfer officer. Two groups of trainees will be involved: clinical and basic science (radiochemistry) trainees. Clinical trainees will be recruited from 3 training programs: Nuclear Medicine Resident Program; Radiology Residency Program; and Residency Program in Radiation Oncology. Basic science trainees will be graduate students from degree-awarding departments (chemistry, material sciences, bioengineering) or post-doctoral students (radiology, radiation oncology, or collaborating departments). A goal of training is to develop future investigators who will be able to conduct independent translational research with knowledge and expertise earned in this interdisciplinary team environment. To address clinically relevant questions, the program will organize regular brainstorming meetings participated by clinicians from various medical disciplines; program investigators; trainees; technology transfer officers to discuss medical questions, relevant targets, and technical feasibility. UW provides outstanding research and clinical facilities and resources that can be accessed by investigators. The Departments of Radiology and Radiation Oncology at UW have been at the forefront of translational imaging and therapeutics research efforts in cancer, neurological disorders, and cardiovascular diseases. This funding opportunity will solidify ongoing efforts of interdisciplinary collaborations at UW into a cohesive program that will train a new generation of investigators. This will, in turn, accelerate the translational potential of new radiotracers by rapid developments of proof-of-concept and first-in-man studies along with in vivo validations in a clinically relevant context.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2018

Developing a Human Factors Engineering Program Plan and End State Vision to Support Full Nuclear Power Plant Modernization

The Light Water Reactor Sustainability (LWRS) Program, which is sponsored by the United States (U.S.) Department of Energy (DOE), conducts research and development (R&D) to establish the technical bases to sustain U.S. nuclear assets. One area in the LWRS program is the Plant Modernization pathway, which includes human factors R&D, human factors engineering (HFE), and ergonomics. LWRS program researchers in this pathway conduct targeted R&D to address aging and reliability concerns with the legacy instrumentation and control (I&C) and related information systems in commercial nuclear power plants (NPPs). One key aspect of this pathway is to develop a strategy for full nuclear plant modernization, which previous LWRS plant modernization milestone reports have begun to describe. To further support this comprehensive plan and vision for full plant modernization, additional details need to be further developed. In particular, a core aspect of a strategy for full nuclear plant modernization is an HFE program plan and an end state vision for plant modernization that is well-integrated with both 1) a technically defensible approach to migrating the existing, mostly analog I&C infrastructure to a digital I&C infrastructure, and 2) a valid business case methodology to cost-justify the modernization activity. Thus, the purpose of this LWRS R&D activity and milestone report is threefold. First, the report provides guidance for operating commercial NPPs to support their development and evaluation of an HFE program management plan (HFE PMP) and an end state vision for plant modernization. The report presents best practices and lessons learned related to HFE PMPs based on MCR modifications in which analog systems were replaced by digital I&C systems. Second, the report summarizes the rationale for full digital I&C integration, highlights a number of factors that are critical to consider, and describes how HFE should merge with full digital I&C integration. Finally, the report summarizes and integrates the economic factors that must be considered to cost-justify HFE involvement in plant modernization.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2018

Integration of Solar Training into Allied Professional Development Platforms - Final Technical Report

The Interstate Renewable Energy Council (IREC) designed the “Integration of Solar Training into Allied Professional Development Platforms” project to surmount barriers to the expansion of solar technology caused by gaps in knowledge impacting on-the-job performance of allied professionals. At project start, questionnaires deployed nationally indicated that only 9% of firefighters felt certain they could shut down a PV system to the extent possible, and only a third of code officials thought they could even discern the difference between a PV and solar thermal system. Therefore, the project objective was to measurably and significantly increase the number of code officials and fire service professionals with solar knowledge directly applicable to their job function. IREC identified specific gaps in current knowledge, and integrated research-proven adult learning principles and technology solutions to enable a cost-effect but impactful reach. The project design ensured that solar training is a continuing component of the targeted sector professional development toolkit. To this end, the key partners recruited and cultivated were existing, trusted training partners in the targeted sectors: International Code Council Solar Rating and Certification Corporation (ICC-SRCC), International Association of Electrical Inspectors (IAEI), International Association of Fire Fighters (IAFF) and National Association of State Fire Marshals (NASFM). During the project, IREC held 79 courses in 25 states that trained 4,156 people, offered a PV in the 2017 National Electric Code webinar attended by 787 people, developed on-line training for code officials with 2,298 on-line learners, and developed an on-line training for firefighters with 4,930 on-line learners. In addition, the project participated in 28 conferences and reached over 537,000 professionals with solar information.

14 SOLAR ENERGY↗
DOE OSTI2018

Abstract - Cooperative Research and Development Agreement between Ramaco Carbon and National Energy Technology Laboratory

NETL and Ramaco Carbon have developed an umbrella Cooperative Research and Development Agreement (CRADA). This umbrella CRADA will be used to initiate detailed technical discussions for assessing projects in 3 general areas: 1) The conversion of coal to value-added solid carbon products such as carbon fiber and carbon nanomaterials, 2) The use of computational modeling for developing coal beneficiation processes, and 3) The development of technologies for extracting rare earth elements from coal, beneficiated coal, coal-byproducts and mining byproducts.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2018

Developing a Carbon Model for the U.S. Economy

In this project, we have assessed the feasibility of developing a stock and flow model of carbon for the United States (U.S.) economy and applying it to assess opportunities for substitution of carbon dioxide (CO 2 ) as a feedstock for industrial processes. We have assessed the feasibility of developing a model that could answering these questions by integrating the results of the following three tasks: (1) A survey of data in existing government databases of commodity stocks and flows with the aim of identifying relevant data sources, data gaps, and how they could be filled. (2) Developing a quantitative “topdown” description of carbon stocks in and flows through the U.S. economy, in which we present the results in a Sankey diagram for carbon that will compliment those prepared by LLNL for the U.S. energy system. (3) Conducting a quantitative case study of a selected material production pathway with the aim of understanding the potential to assemble a national picture of carbon flows from the “bottom-up”. We integrate the results of these tasks and identify the way in which we could develop a more refined national model of carbon stocks and flows to assess and intercompare carbon dioxide utilization (CO 2 U) pathways.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

ICME Guided Development of Advanced Cast Aluminum Alloys for Automotive Engine Applications (ADAPA)

Project Objectives: To develop a new class of advanced, cost competitive aluminum casting alloys providing a 25% improvement in component strength relative to components made with A319 or A356 alloys using sand and semi-permanent casting processes for high-performance engine applications. To demonstrate the power of Integrated Computational Materials Engineering (ICME) tools for accelerating the development of new materials and processing techniques, as well as to identify the gaps in ICME capabilities. To develop comprehensive cost models to ensure that components manufactured with these new alloys do not exceed 110% of the cost using incumbents alloys A319 or A356. To develop a technology transfer and commercialization plan for deployment of these new alloys in automotive engine applications. The key targets and baseline are shown in the table below.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Fast Reactor Methods Development Plan

Fast reactor design and safety analysis methods (computer codes) have been under development for approximately sixty years. During the first forty years, code development was driven by the immediate needs of operational facilities such as Experimental Breeder Reactors I and II and the Fast Flux Test Facility. Development was further driven by programmatic needs to support advanced reactor programs such as the Clinch River Breeder Reactor Project and the Advanced Liquid Metal Reactor Program. By the early 1990s, design and safety analysis codes had been well established to addressed the unique characteristics of fast reactors. During the subsequent twenty years, however, development was nearly stagnant as a result of U.S. Government policy changes that abruptly canceled nearly all of the fast reactor research in the United States. As a result, the state of fast reactor design and safety analysis methods has deteriorated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Pretreatment of Crystalline Silicotitanate (CST) and development of a digestion standard to support Tank Closure Cesium Removal (TCCR)

In support of Savannah River Remediation’s (SRR) deployment of an ion exchange process to remove radioactive cesium from waste supernate, referred to as the Tank Closure Cesium Removal (TCCR) process, SRNL was tasked with validating the field protocol for pretreating the crystalline silicotitanate (CST) material to be used in the columns, preparing CST for use in in-tank batch contact testing, as well as developing a digestion standard to verify complete dissolution of the CST prior to characterization after each in-tank batch contact test. A procedure (SRNL L29 Manual, ITS-0229) was developed to document the protocol for pretreating CST utilizing conditions similar to what will be performed in the field. Analysis of material pretreated following the procedure confirmed similar conversion to the Na+ form of the resin as had been obtained previously. Portions of the batch of material pretreated according to the developed protocol were also used for development of a digestion standard. Results from digestion and analysis of this standard were used to establish a set of control limits for the measured Ti, Zr, and Nb concentrations in digested CST material. An aliquot of this standard CST material will be digested and analyzed alongside each in-tank batch contact sample, and the measured concentrations of Ti, Zr, and Nb will be compared against the established control limits to confirm complete digestion has occurred.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of Critical Experiments to Provide Validation Data for Multiphysics Coupling Methods (Final Report)

Computational simulation has become one of the most important tools in the modern toolbox for nuclear reactor design. Reactor simulations, like the reactors themselves, involve multiple physical processes including neutronics, fluid flow and heat transfer (thermal-hydraulics), thermo-mechanics, fuel behavior, chemistry, and balance of plant. Each process can affect the behavior of the others, and advanced computational tools have been developed that couple separate physics models together to simulate the feedback effects. The objective of this work is to generate experimental data to support the validation of multi-physics reactor codes. It specifically targets the SHARP toolset from the NEAMS Reactor Product Line, but it is intended to be useful in any set of coded that couples thermal-hydraulics with neutronics simulations. The project takes advantage of the inherent flexibility of the RPI Reactor Critical Facility (RCF), a low power open-pool research reactor. The standard configuration of the RCF is used to explore the effects of moderator temperature on neutronic behavior and reactivity. Because the low power of the reactor generates no appreciable temperature change in the moderator, the moderator temperature is artificially changed with two 18 kW electric immersion heaters. The temperature feedback effects can be observed by recording the change in the reactor power as the reactor goes from a supercritical state, to critical, to a subcritical state. To expand the temperature range and temperature rate of change that can be investigated, an experimental apparatus consisting of a heated water loop was designed and installed in the RCF. In this setup, water from a heated reservoir tank is pumped through a pipe placed through the center of the reactor core. The water in the reservoir can be heated to around 70 °C and pumped through the system. The pump has a variable flow rate and electronic control, so a number of scenarios can be developed. A slug of water can be pushed through the system to generate a reasonably fast change in reactivity. The pump speed can be increased or decreased over the course of a measurement, or even oscillated at different periods, to observe the response of the reactor. To complement the experimental data generated by the RCF, a set of computational models have been developed. They are intended to illustrate the usefulness of the experimental data generated in validating coupled multi-physics simulations. A model previously constructed in MCNP has been used in conjunction with safety analysis at the RCF for many years and was modified to support the design of the heated water loop experiment. This model was translated into Serpent 2, a similar Monte Carlo code, which provided the added advantage of being able to produce collapsed cross-section data files for use in the deterministic neutron transport codes that are part of the SHARP toolset. A complete reconstruction of the RCF model, both with and without the heated fluid loop experiment, were necessary for use with the SHARP codes, PROTEUS (neutronics) and Nek5000 (thermal hydraulics), as these require an unstructured mesh format to operate. The computational requirements for fully coupled three dimensional simulations are significant, and challenges to full implementation and comparison with RCF experiments remain. A wide variety of experiments have been conducted for eventual benchmarking and validation of multi-physics reactor codes. A complete database of detailed time- and temperature-dependent measurements have been generated and stored. Several examples of the experimental results are presented here, with refinement and analysis of experiments ongoing. While these experiments still only address a finite range of operating conditions, in particular being limited to water in the sub-boiling regime, they represent a unique (to date) look at the feedback between temperature and neutronic behavior suited for validation of the coupling routines in multi-physics codes. It is hoped that these will be the foundation of future benchmarks that will provide confidence in the tools that will be used to develop the next generation of nuclear reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of a Novel Accelerator for Neutron Transport Solution Using the Galerkin Spectral Element Methods (Final Report)

This report summarizes the results of a three-year research project sponsored by the U.S. Department of Energy (DOE) Nuclear Energy University Program (NEUP) to develop and implement advanced acceleration schemes for the DOE NEAMS neutronics code PROTEUSSN. The project team included the University of Massachusetts Lowell, The Ohio State University, University of Michigan, and Argonne National Laboratory. The PROTEUS code is a high-fidelity capable deterministic neutron transport code based on unstructured finite element meshes, which solves the steady-state and transient neutron transport problem using the 2nd-order discrete ordinate method (SN2ND), the method of characteristics (MOC), and the advanced nodal transport method (NODAL). The DSA scheme has been implemented in PROTEUS to speed up SN transport calculations. The existing DSA scheme employs a consistent finite element formulation on the same fine mesh structure as the SN solution While the use of consistent discretizations makes the DSA method effective, it does not necessarily make it efficient because the numerical solution of a discretized elliptic diffusion problem itself can be costly, particularly when the problem size becomes large. We developed a new discontinuous Galerkin (DG) discretization of the diffusion equation, called DG-DSA, which can effectively and efficiently accelerate the SN transport iterations. As compared with the previous work, the novelty of our method is that the diffusion equation is solved on a coarse-mesh grid using the DG methods, and the DG diffusion discretization incorporates local hp adaptation, i.e., local adaptation of mesh size and/or polynomial degree, based on local total cross section (or optical thickness). Therefore, the resulting number of degrees of freedom (DOF) of the DG discretization is much less than the conventional consistent DSA discretizations, and thus DG-DSA can achieve significant improvement in computational efficiency. We implemented this scheme in PROTEUS-SN. In addition, we developed a new stabilization scheme named linear prolongation CMFD (lpCMFD). A novel feature of this scheme is that the conventional flat flux ratio–based scaling approach is replaced with a linear interpolation of the scalar flux differences at the coarse-mesh cell edges between the neutron transport and CMFD calculations. Fourier convergence analysis and numerical results show that lpCMFD is unconditionally stable and more effective than other CMFD based on acceleration schemes such as pCMFD, odCMFD. We also developed a new nonlinear diffusion acceleration scheme for solving neutron transport equations. This scheme, called LR-NDA, employs a local refinement approach on the framework of CMFD by solving a local boundary value problem of the scalar flux on the coarse-mesh structure to replace the piecewise constant scalar flux obtained by CMFD. The refined flux is then used to update the scalar flux in the neutron transport source iteration. It has been demonstrated that that LR-NDA is much more effective and stable than CMFD for a wide range of optical thicknesses. LR-NDA is a local adaptive method, which means LR-NDA does not necessarily require local refinement for all the coarse-mesh cells on the problem domain, i.e., it can be used only for relatively optically thick regions where the standard CMFD scheme would encounter the convergence problem.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Development of the Ultimate One-Dimensional Time to Explosion (UODTX) Apparatus

With recent advancements in experimental capabilities and worldwide improvement of digital data acquisition and control, there is sufficient justification to develop a new version of the One Dimensional Time to eXplosion (ODTX) system to allow for additional diagnostics and improved control to increase sample throughput and increase the adaptability of future experiments. The legacy ODTX system was designed to measure time to explosion (TTX) for a given temperature, which it does well, but it was not equipped to handle the recent experimental advances of Pressure-ODTX (PODTX) and Composition-ODTX (CODTX). PODTX involves fielding a pressure transducer as an additional diagnostic and applying arbitrary heating-profiles to the system anvils. CODTX furthers this concept by including in-situ gas-composition analysis via infrared (IR) absorption techniques with plans to extend CODTX to solid composition detection and in-situ mass spectrometry (MS). The legacy ODTX system was not built with these experiments in mind and performing them involves substantial modifications and downtime to the existing machine. Developing a standalone system that can perform all C/P/ODTX experiments and is adaptable to unconceived experiments would allow for the continued use of ODTX as a work-horse for Small Scale Safety Testing (SSST), but also provide a testbed for new explosive safety research. This system would ultimately be housed in the neighboring ODTX cell and would live at the High Explosive Application Facility (HEAF) until the end of its service life. HEAF, Chemistry, and Engineering would be the primary stakeholders for the system’s design as they all have experiments that they want to perform. Engineering would be responsible for the system development and would work closely with HEAF to ensure that the system is built with high explosive (HE) safety in mind. By the end of development and after product realization, HEAF/Chemistry would be the system owner, operator, and primary user. The proposed system design is highly influenced by the existing ODTX design. The system will have load frame to support the experiment, anvils to confine the explosive, and some means of applying confining pressure. The samples will be able to be heated and cooled through the confining the anvils. The biggest change from the legacy design is the ability to add diagnostics to the experiment as needed: thermocouples (TCs), resistance temperature detectors (RTDs), pressure transducers (PTs), infrared transmittance/absorbance (IR), mass spectrometry (MS), strain gages (SGs), photonic doppler velocimeter (PDV), as well as arbitrary 4-20 mV signals and +/- 10 VDC signals. The current design does not allow for flexibility in that respect and is hardwired with a few diagnostic channels. The onus will fall on designing a load frame that can support system modifications as well as a data acquisition system that can handle the multitude of signal types.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Prototype Ceramic HEPA Filter Testing and Filter Media Research and Development for Nuclear Facility Ventilation

This is a brief summary of recent and past Ceramic HEPA Filter research and development activities under an NNSA NSR&D funded project and describes on-going plans for FY19 work as presently planned. Prior work has evaluated concepts for a ceramic HEPA filter including ceramic support materials, filter media including electrospun filter media, and metal housing designs to accommodate the conservative temperature performance target of up to 500°C. This has been supported by modelling efforts discussed with nuclear facility system engineers and their management. The temperature and time-attemperature are based on conservative, bounding temperatures from fire analysis presented as an EFCOG best practice paper for nuclear facility safety analysis. The overall goal is to improve safety of DOE nuclear facilities during fire scenarios by developing improved filtration technology including increasing temperature performance while maintaining acceptable filter pressure drop. Ceramic filtration technology can provide robust, passive mitigation against radiological releases. The goal of the ceramic HEPA filter is to provide a HEPA filter that provides increased temperature performance in an equivalent form factor (size and shape) that is similar to existing conventional HEPA filters used in facilities. Once development is complete and qualified for use in facilities, ceramic HEPA filters could simplify and reduce safety and support system costs, as well as lower life-cycle costs. Installation, monitoring, and maintenance of active fire suppression systems, as well as other related systems, are a considerable on-going cost to nuclear facilities. Reducing the level of dependence on Safety Class and Safety Significant fire alarm and fire suppression systems could allow significant cost savings. At present we are continuing research and development efforts towards the goal of a ceramic filter design, including improved ceramic filter media, that could reduce total operation cost for facilities utilizing HEPA filters.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Development of a Liquid-Fueled Molten Salt Reactor Safeguards Model.

This work describes the ongoing work to develop a molten salt reactor (MSR) model and associated tools for safeguards analysis. A new flowsheet was developed in collaboration with Oak Ridge National Laboratory (ORNL) for the Molten Salt Demonstration Reactor (MSDR). This design was chosen by ORNL as a generic baseline design that could be used for safeguards research. The model has simple chemical processing that is less extensive than the two-fluid flowsheet developed in the last year. A detailed TRITON reactor physics model, provided by ORNL, was implemented into the process model. The process model now includes reactor parameters such as K-eff and decay heat, which could be used as part of an advanced safeguards approach. Finally, a set of generic safeguards tools based on current safeguards approaches were developed. These tools are flexible and can be used with most MSR flowsheets. ACKNOWLEDGEMENTS This work was funded by the Materials Protection Accounting and Control Technologies (MPACT) working group as part of the Fuel Cycle Technologies Program under the U.S. Department of Energy, Office of Nuclear Energy. The authors would also like to acknowledge Ben Betz ler for his work on the reactor physics models that were incorporated into the work and the continued collaboration with ORNL staff.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2019

Prototype Ceramic HEPA Filter Testing and Filter Media Research and Development for Nuclear Facility Ventilation

This is a brief summary of recent and past Ceramic HEPA Filter research and development activities under an NNSA NSR&D funded project and describes on-going plans for FY19 work as presently planned. Prior work has evaluated concepts for a ceramic HEPA filter including ceramic support materials, filter media including electrospun filter media, and metal housing designs to accommodate the conservative temperature performance target of up to 500°C. This has been supported by modelling efforts discussed with nuclear facility system engineers and their management. The temperature and time-at-temperature are based on conservative, bounding temperatures from fire analysis presented as an EFCOG best practice paper for nuclear facility safety analysis. The overall goal is to improve safety of DOE nuclear facilities during fire scenarios by developing improved filtration technology including increasing temperature performance while maintaining acceptable filter pressure drop. Ceramic filtration technology can provide robust, passive mitigation against radiological releases. The goal of the ceramic HEPA filter is to provide a HEPA filter that provides increased temperature performance in an equivalent form factor (size and shape) that is similar to existing conventional HEPA filters used in facilities. Once development is complete and qualified for use in facilities, ceramic HEPA filters could simplify and reduce safety and support system costs, as well as lower life-cycle costs. Installation, monitoring, and maintenance of active fire suppression systems, as well as other related systems, are a considerable on-going cost to nuclear facilities. Reducing the level of dependence on Safety Class and Safety Significant fire alarm and fire suppression systems could allow significant cost savings, while maintaining safety. At present we are continuing research and development efforts towards the goal of a ceramic filter design, including improved ceramic filter media, that could reduce total operation cost for facilities utilizing HEPA filters.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Development of a Corrosion Monitoring System for Advanced Molten Salt Reactors

Molten salt reactors (MSRs) have been identified as one of several possible next-generation advanced reactor designs that could replace the aging fleet. There are many inherent advantages to MSRs, including higher power density, lower pressure, low stored energy, and prompt negative temperature reactivity coefficient Ezell et al. [2018]. However, there are challenges related to the operating temperatures, as well as, the corrosive salts attacking the structural materials. During the Molten Salt Reactor Experiment (MSRE), which was operated at Oak Ridge National Laboratory (ORNL) (Ignatiev et al. [2014], Forsberg et al. [2007], Forsberg et al. [2008], Delpech et al. [2009], Serp et al. [2014], Ignatiev and Surenkov [2017]) corrosion was monitored by sampling the salt for chromium. However, this type of monitoring does not identify the location of corrosion. In FY18, ORNL developed a novel in-situ corrosion sensor that could be placed on the outside of the structure to monitor changes in magnetic susceptibility of salt-wetted alloys as corrosion occurs Holcomb et al. [2019]. Initial testing with this novel sensor demonstrated a correlation between the change in signal output from the sensor to the various levels of pre-corroded test specimens. However, all initial testing was performed at room temperatures, which is not realistic for the harsh MSR environment. Based on this research, FY19 funding allowed for development of a high-temperature version of the sensor to be installed at various locations on the reactor infrastructure. The high-temperature sensor can operate at up to 750°C and can monitor development of corrosion inside the pipe. This sensor measures the increase in magnetic susceptibility that occurs as chromium dissolves from the inner surface of the pipe, when paramagnetic structural alloys become ferromagnetic with the loss of chromium. This report will discuss the development of corrosion monitoring sensor and the testing with results of corroded specimens at increasing temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

The Development of the Generalized Spallation Model [Thesis]

The generalized spallation model (GSM) is a general spallation physics event generator with a physics-rich history dating back to the early 1970s during the development of the CEM. GSM estimates the outcome of high-energy hadron-nucleus and nucleus-nucleus collisions, where collision energies often exceed hundreds of MeV. Such a collision may include a neutron incident on carbon, for example. The thesis presented here outlines the continued development and modernization of GSM, whereby it was migrated from a legacy Fortran code to utilize an advanced and modern Fortran architecture. Verification and validation of GSM also took place throughout the continued development of GSM. The results of this work have transformed GSM into a highly modern and powerful event generator, where numerous benefits have been fostered and precipitated, including but not limited to portability, modularization, flexibility, robustness, and scalability. GSM is a robust and well developed software library that clients and end-users may confidently use, within the scope outlined here, as a result of this work.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Progress on Grizzly Development for Reactor Pressure Vessels and Reinforced Concrete Structures

The Grizzly code has been under development for the U.S. Department of Energy’s Light Water Reactor Sustainability program to provide predictive tools for the evolution of material degradation in critical light water reactor structural components due to long-term exposure to the environmental conditions of normal reactor operation, and for the effects of this degradation on the safety of these components. This development has primarily targeted reactor pressure vessels and reinforced concrete structures. Work performed during Fiscal Year 2019 added important features for both of these types of structures. For probabilistic fracture mechanics analysis of reactor pressure vessels, capabilities were added to account for the effects of warm prestressing and residual stresses, as well as to model crack initiation, growth, and arrest. For modeling microstructure evolution in reactor pressure vessels, a cluster dynamics model for evolution of Mn-Ni-Si phases has been implemented. For modeling degraded reinforced concrete structures, capabilities to consider the combined effects of damage and creep in the concrete constitutive model and model nonlinear behavior of reinforcing bars has been added. These newly-developed features fill important gaps in Grizzly’s feature set for engineering-scale analysis of degraded reactor pressure vessels and reinforced concrete structures. Grizzly development in both of these areas has reached a point where it is ready for extensive testing on relevant engineering problems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

ROCM+Intel-PathForward+RemoteSpaces Development

This report documents the completion of milestone STPRO4-25 Harden and optimize the ROCm based AMD GPU backend, develop a prototype backend for the Intel ECP Path Forward architecture, and improve the existing prototype Remote Memory Space capabilities. The ROCM code was hardened up to the point of passing all Kokkos unit tests - then AMD deprecated the programming model, forcing us to start over in FY20 with HIP. The Intel ECP Path Forward architecture prototype was developed with some initial capabilities on simulators - but plans changed, so that work will not continue. Instead SYCL will be developed as a backend for Aurora. Remote Spaces was improved. Development is ongoing part of a collaboration with NVIDIA.

97 MATHEMATICS AND COMPUTING↗