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DOE OSTI2015

IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V&V) of available analytical capabilities for HTGR simulation for design and safety evaluations. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Final Report (2010-2015) for the Topical Collaboration on Quantitative Jet and Electromagnetic Tomography (JET) of Extreme Phases of Matter in Heavy-ion Collisions

During the 5-year funding period (2010-2015), the JET Collaboration carried out a comprehensive research program with coordinated efforts involving all PI members and external associated members according to the plan and milestones outlined in the approved JET proposal. We identified important issues in the study of parton energy loss and made significant progress toward NLO calculations; advanced event-by-event hydrodynamic simulations of bulk matter evolution; developed Monte Carlo tools that combine different parton energy loss approaches, hydrodynamic models and parton recombination model for jet hadronization; and carried out the first comprehensive phenomenological study to extract the jet transport parameter.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2022

Simulation of CEFR neutronic start-up tests with FENNECS

This paper presents simulation results of selected Neutronic Start-up Tests of the China Experimental Fast Reactor (CEFR) obtained by the neutronics code FENNECS that have been performed within the frame of the IAEA Coordinated Research Program I31032. The Finite Element Neutronics code FENNECS is developed at GRS and solves the few-group steady-state and transient diffusion equation using a Galerkin-based finite element approach. Its main purpose is the safety assessment of Small Modular Reactors and Micro Reactors with complex geometry (e.g., rotating control drums) which gain increased interest internationally. Serpent has been applied to create reference models of the CEFR and cross-section libraries in 10 energy groups for FENNECS. Using these libraries, the following experiments have been simulated with FENNECS: net criticality, control rod integral and differential worth, void reactivity effect, subassembly exchange reactivity effects and reaction rate distribution. The obtained satisfactory agreements with the measurements represent a valuable contribution to the validation of FENNECS. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

GEN-IV Benchmarking of Triso Fuel Performance Models under accident conditions modeling input data

This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: • The modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release. • The modeling of the AGR-1 and HFR-EU1bis safety testing experiments. • The comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from “Case 5” of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. “Case 5” of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to “effects of the numerical calculation method rather than the physical model” [IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary. 09/2016: Tables 6 and 8 updated. AGR-2 input data added

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-TEMPERATURE GAS-COOLED TEST REACTOR

Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Generation IV Benchmarking of TRISO Fuel Performance Models Under Accident Conditions Final Report

The Generation IV International Forum (GIF) is a co-operative international endeavor of fourteen members organized to carry out the research and development needed to establish the feasibility and performance capabilities of the next generation nuclear energy systems. GIF selected six reactor technologies, amongst which is the Very High Temperature Reactor (VHTR) that is primarily dedicated to the cogeneration of electricity and hydrogen. The technical basis for VHTR is the tristructural isotropic (TRISO)-coated particle fuel, the graphite as the core structure, helium coolant, as well as the dedicated core layout and lower power density to removal decay heat in a natural way. At the heart of safety features of the VHTR concept lie the TRISO fuel particles that are designed to keep their structural integrity and retain fission products at temperatures up to 1600°C. As part as the design and future operation of VHTRs, a key aspect is the accurate prediction of fuel performance under irradiation and accident conditions. Modeling and simulation allow prediction of TRISO fuel behavior when subject to neutron flux and in high temperature accident scenarios. The refinement of the fuel performance models and codes is performed by comparison to in-pile and out-of-pile experimental data that reproduce the expected irradiation conditions in high temperature gas-cooled reactors (HTGRs). Historically, the International Atomic Energy Agency (IAEA) developed a benchmark dedicated to the validation of predictive methods for fuel and fission product behavior through the Coordinated Research Program CRP-2 (IAEA, 1997). CRP-2 was later updated to cover fuel fabrication, quality assurance, irradiation performance, safety testing, and spent fuel. The scope of the resulting CRP-6 benchmarks focused on HTGR fuel performance and fission product release (IAEA, 2012). Taking advantage of additional TRISO fuel fabrication, irradiation, and safety testing campaigns, GIF launched a Generation IV Benchmarking of TRISO Fuel Performance Models under Accident Conditions in late 2015. This GIF benchmark is a three-year program steered by Idaho National Laboratory (INL, USA). The other participants include the Japan Atomic Energy Agency (JAEA, Japan) and the Korea Atomic Energy Research Institute (KAERI, Korea). The objectives of the benchmark are to: follow on the IAEA CRP benchmarks, (2) model fission product release under accident conditions, (3) compare results obtained by the fuel performance modeling codes of the benchmark participants, and (4) compare these code predictions to experimental data. Safety tests chosen for modeling include the first and second experiments of the Advanced Gas Reactor program (AGR-1 and AGR-2) and the High Flux Reactor (HFR) EU1bis experiment. This report presents the results obtained by the three research institutions using their respective fuel performance modeling codes. Comparisons of the corresponding fission product release predictions are made with experimental data from AGR-1, AGR-2, and HFR-EU1bis. The benchmark results show good agreements between all participants but also show a general trend of over-prediction of the experimental release data, which is mainly attributed to the use of over-estimated diffusion coefficients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Bison Verification and Validation Activities for TRISO

Numerical modeling and simulation (M&S) tools play a key role in the research, development, and overall safety assessments of next-generation nuclear energy systems. One such tool, Bison, is a nuclear fuel performance code that is applicable to many fuel forms (e.g., light-water reactor fuel, oxide and metallic fuel for fast reactors, tri-structural isotropic (TRISO) fuel, and plate fuel), and it uses the finite element method to model the thermo- mechanical response of nuclear fuels. One fuel form widely utilized in Generation-IV high-temperature gas-cooled and fluoride- salt-cooled nuclear reactor concepts is TRISO fuel. Recently, Bison’s capabilities were significantly expanded to enable it to model the performance of TRISO particles and compacts. It is important that Bison’s computational results be reliable and predictive, since this code is used to inform high-consequence decisions. The various processes developed to address this issue generally entail two fundamental steps: verification and validation (V&V). Verification ensures that the code functions correctly and is reliable. Code/solution verification, code benchmark, and software quality assurance exercises are examples of verification activities. On the other hand, validation is the process of assessing a code’s capability to accurately model physical problems. Comparisons between code results and experiments quantify the validation level. Application of V&V procedures is crucial to the development of computational tools that are free of coding mistakes and can accurately represent reality. The current study presents an overview of Bison V&V activities relevant to the TRISO fuel concept, which include code/solution verification exercises, CRP-6 Benchmark—a Coordinated Research Program through the International Atomic Energy Agency (IAEA)—exercises, and validation exercises with the Advanced Gas Reactor (AGR)- 1/2/3/4 experiment series.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Reflecting Sunlight: Recommendations for Solar Geoengineering Research and Research Governance

The National Academies of Sciences, Engineering, and Medicine (NASEM) report "Reflecting Sunlight: Recommendations for Solar Geoengineering Research and Research Governance" focused on three atmospheric-based Solar Geoengineering strategies: Stratospheric Aerosol Injection, Marine Cloud Brightening, Cirrus Cloud Thinning. The study recommends that a national (and internationally-coordinated) Solar Geoengineering research program be established and coordinated by the U.S. Global Change Research Program—with a focus on developing information needed by decision makers, not advancing a path to deployment. The study recommends a detailed framework for governance of this research, at both national and international levels. The study also recommends a detailed research agenda for this program, encompassing over a dozen transdisciplinary research topics that can be loosely grouped into the categories of (i) technical dimensions and impacts, (ii) social dimensions, and (iii) context and goals for solar geoengineering research itself.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2016

New High-Throughput Measurement Systems for Radioactive Wastes Segregation and Free Release - 16166

The European Metrology Research Program (EMRP) is a coordinated research-and-development program that pursues the closer integration of national research programs. The EMRP Metrology for Decommissioning Nuclear Facilities 'MetroDecom' Joint Research Project ENV54 (JRP ENV54), a 36-month project started in September 2014, delivers research products addressing all aspects of the decommissioning process. This includes the characterization of waste materials present on the decommissioning site, pre-selection prior to free release or repository acceptance measurement, free-release measurement, the measurement of waste packages thermal power prior to the repository storage, and the monitoring of stored wastes and repositories in the very long term. This paper describes the track record of two JRP ENV54 working packages, which address the measurement facilities for pre-selection of waste materials prior to measurement for repository acceptance or possible free release (segregation measurement system); and free release (free release measurement system), based on a single standardized concept characterized by unique, patented lead-free shielding. The key objective is to improve the throughput, accuracy, reliability, modularity and mobility of segregation and free-release measurement. This will result in more reliable decision-making with regard to the safe release and disposal of radioactive wastes into the environment and, resulting in positive economic outcomes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Can Radiation Therapy Quality Assurance Improve Nasopharyngeal Cancer Outcomes in Low- and Middle-Income Countries: Reporting the First Phase of a Prospective International Atomic Energy Agency Study

Most new nasopharyngeal cancer cases occur in low-income and middle-income countries, and these patients experience poorer overall survival than that of new nasopharyngeal cancer cases in high-income countries. The goal of this research project is to determine whether the introduction of a radiation therapy quality assurance program can ultimately improve outcomes for nasopharyngeal cancer patients in lower-income and middle-income countries. This study reports the results of the first phase of the International Atomic Energy Agency Coordinated Research Project (325-E3-TM-47712).

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2015

On-Going International Research Program on Irradiated Concrete Conducted by DOE, EPRI and Japan Research Institutions. Roadmap, Achievements and Path Forward

The Joint Department of Energy (DOE)-Electric Power Research Institute (EPRI) Program (Light Water Reactor Sustainability (LWRS) Program–Material Pathway–Concrete and Long-Term Operation (LTO) Program) and US Nuclear Regulatory Commission (NRC) research studies aim at understanding the most prominent degradation modes and their effects on the long-term operation of concrete structures to nuclear power generation. Based on the results of the Expanded Materials Degradation Analysis (EMDA), (NUREG/CR-7153, ORNL/TM-2011/545), irradiated concrete and alkali-silica reaction (ASR)-affected concrete structures are the two prioritized topics of on-going research. This report focuses specifically on the topic of irradiated concrete and summarizes the main accomplishments obtained by this joint program, but also provides an overview of current relevant activities domestically and internationally. Possible paths forward are also suggested to help near-future orientation of this program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Germany-US Nuclear Theory Exchange Program for QCD Studies of Hadrons & Nuclei 'GAUSTEQ'

GAUSTEQ was a Germany-U.S. exchange program in nuclear theory whose purpose was to focus research efforts on QCD studies of hadrons and nuclei, centered around the current and future research programs of Jefferson Lab and the Gesellschaft fur Schwerionenforschung (GSI) in Germany. GAUSTEQ provided travel support for theoretical physicists at US institutions conducting collaborative research with physicists in Germany. GSI (with its Darmstadt and Helmholtz Institute Mainz braches) served as the German “hub” for visits of U.S. physicists, while Jefferson Lab served as the corresponding “hub” for visits of German physicists visiting U.S. institutions through the reciprocal GUSTEHP (German-US Theory Exchange in Hadron Physics) program. GAUSTEQ was funded by the Office of Nuclear Physics of the U.S. Department of Energy, under Contract No.DE-SC0006758 and officially managed through Old Dominion University in Norfolk, Virginia. The program ran between 2011 and 2015.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2016

ATF Neutron Irradiation Program Technical Plan

The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Reference Cross Sections for Charged-particle Monitor Reactions

Evaluated cross sections of beam-monitor reactions are expected to become the de-facto standard for cross-section measurements that are performed over a very broad energy range in accelerators in order to produce particular radionuclides for industrial and medical applications. The requirements for such data need to be addressed in a timely manner, and therefore an IAEA coordinated research project was launched in December 2012 to establish or improve the nuclear data required to characterise charged-particle monitor reactions. An international team was assembled to recommend more accurate cross-section data over a wide range of targets and projectiles, undertaken in conjunction with a limited number of measurements and more extensive evaluations of the decay data of specific radionuclides. Least-square evaluations of monitor-reaction cross sections including uncertainty quantification have been undertaken for charged-particle beams of protons, deuterons, 3 He- and 4 He-particles. In conclusion, recommended beam monitor reaction data with their uncertainties are available at the IAEA-NDS medical portal www-nds.iaea.org/medical/monitor_reactions.html.

43 PARTICLE ACCELERATORS↗

DOE OSTI2015

Uranium Biomineralization by Natural Microbial Phosphatase Activities in the Subsurface

In this project, inter-disciplinary research activities were conducted in collaboration among investigators at The University of Alabama (UA), Georgia Institute of Technology (GT), Lawrence Berkeley National Laboratory (LBNL), Brookhaven National Laboratory (BNL), the DOE Joint Genome Institute (JGI), and the Stanford Synchrotron Radiation Light source (SSRL) to: (i) confirm that phosphatase activities of subsurface bacteria in Area 2 and 3 from the Oak Ridge Field Research Center result in solid U-phosphate precipitation in aerobic and anaerobic conditions; (ii) investigate the eventual competition between uranium biomineralization via U-phosphate precipitation and uranium bioreduction; (iii) determine subsurface microbial community structure changes of Area 2 soils following organophosphate amendments; (iv) obtain the complete genome sequences of the Rahnella sp. Y9-602 and the type-strain Rahnella aquatilis ATCC 33071 isolated from these soils; (v) determine if polyphosphate accumulation and phytate hydrolysis can be used to promote U(VI) biomineralization in subsurface sediments; (vi) characterize the effect of uranium on phytate hydrolysis by a new microorganism isolated from uranium-contaminated sediments; (vii) utilize positron-emission tomography to label and track metabolically-active bacteria in soil columns, and (viii) study the stability of the uranium phosphate mineral product. Microarray analyses and mineral precipitation characterizations were conducted in collaboration with DOE SBR-funded investigators at LBNL. Thus, microbial phosphorus metabolism has been shown to have a contributing role to uranium immobilization in the subsurface.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2015

FY-2015 FES (Fusion and Energy Sciences) Joint Research Target: Final Report for the Period October 1st, 2014, through September 30th, 2015

Annual JRT-15 Target: Conduct experiments and analysis to quantify the impact of broadened current and pressure profiles on tokamak plasma confinement and stability. Broadened pressure profiles generally improve global stability but can also affect transport and confinement, while broadened current profiles can have both beneficial and adverse impacts on confinement and stability. This research will examine a variety of heating and current drive techniques in order to validate theoretical models of both the actuator performance and the transport and global stability response to varied heating and current drive deposition.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

FY16 Summary Report: Participation in the KOSINA Project

Salt formations represent a promising host for disposal of nuclear waste in the United States and Germany. Together, these countries provided fully developed safety cases for bedded salt and domal salt, respectively. Today, Germany and the United States find themselves in similar positions with respect to salt formations serving as repositories for heat-generating nuclear waste. German research centers are evaluating bedded and pillow salt formations to contrast with their previous safety case made for the Gorleben dome. Sandia National Laboratories is collaborating on this effort as an Associate Partner, and this report summarizes that teamwork. Sandia and German research groups have a long-standing cooperative approach to repository science, engineering, operations, safety assessment, testing, modeling and other elements comprising the basis for salt disposal. Germany and the United States hold annual bilateral workshops, which cover a spectrum of issues surrounding the viability of salt formations. Notably, recent efforts include development of a database for features, events, and processes applying broadly and generically to bedded and domal salt. Another international teaming activity evaluates salt constitutive models, including hundreds of new experiments conducted on bedded salt from the Waste Isolation Pilot Plant. These extensive collaborations continue to build the scientific basis for salt disposal. Repository deliberations in the United States are revisiting bedded and domal salt for housing a nuclear waste repository. By agreeing to collaborate with German peers, our nation stands to benefit by assurance of scientific position, exchange of operational concepts, and approach to elements of the safety case, all reflecting cost and time efficiency.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

IAEA Coordinated Research Project on HTGR Physics, Thermal-Hydraulics, and Depletion Uncertainty Analysis Prismatic HTGR Benchmark Specification: Phase II

In February 2009, the Technical Working Group on Gas Cooled Reactors of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Project (CRP) on the High-Temperature Gas Cooled Reactor (HTGR) Uncertainty Analysis in Modeling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organisation for Economic Cooperation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on verification and validation of available analytical capabilities for HTGR simulation for design and safety evaluations. Within the framework of these activities different numerical and experimental benchmark problems were performed, and insight was gained about specific physics phenomena and the adequacy of analysis methods. Because the prismatic design specification included in this report is based directly on the OECD/NEA Modular HTGR (MHTGR) 350 MW benchmark, participants in both activities can leverage their core models developed for the OECD/NEA benchmark for this CRP benchmark with only minor changes. The CRP on the HTGR UAM Phase II specifications is also directly linked to the Phase I specification document, since the core models required for Phase II exercises are constructed using the lattice models and cross-section libraries developed for Phase 1. It was decided at the Third Research Coordination Meeting (RCM) in May 2017 that the initial scope of CRP activities will be reduced to match the formally approved remaining contract period (June 2019), as well as limited funding and resource availability. A summary of the revised Phases I–IV exercise definitions is provided here.

42 ENGINEERING↗