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At least 487 records · Page 27

All sources NASA NTRS DOE OSTI

DOE OSTI2018

FY18Q1 Quarterly Report: Radiation Enhanced Diffusion of Ag, Ag-Pd, Eu, and Sr in Neutron Irradiated PyC/SiC Diffusion Couples

Obtaining accurate diffusion kinetics in materials representative of those found in tristructural-isotopic (TRISO) coated particle fuel is needed to predict the diffusive release of the same fission products in reactor. Planar diffusion couples with representative pyrocarbon (PyC) and silicon carbide (SiC) layers are being produced using the same fluidized-bed chemical vapor deposition (FB-CVD) technology used to produce TRISO particles from the first irradiation experiment of the Advanced Gas Reactor Fuel Qualification and Development Program (AGR). The layer properties of the planar diffusion couples are tailored to meet the specified PyC density and microstructure of the SiC layer as defined by the AGR program. The influence of these variables on diffusion is also being explored by producing PyC and SiC variants. The pathway to producing the diffusion couples is discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Updating and Refining BISON TRISO Assessment

Several advanced reactor concepts consider tristructural isotropic (TRISO) fuel in their designs. To support the advanced reactor industry, the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program within the U.S. Department of Energy (DOE) is developing modeling and simulation capabilities within the BISON fuel performance code. Assessment of these capabilities against available experi- mental data, such as the Advanced Gas Reactor (AGR) program, is essential in instilling confidence in the code predictions. Associated uncertainties in model inadequacy, experimental measurements, and manufacturing are also of importance. This report discusses the updating and standardization of the existing assessment suite in BISON for TRISO fuel performance analysis. The cases considered the AGR-1, AGR-2, and AGR-3/4 experimental programs. The report also includes a preliminary investigation into the sources of model inadequacy associated with silver (Ag) release fractions using the AGR-2 Compact 2-1-2 as a case study. The results indicate that uncertainties in the operating temperatures supplied to the model and the form of those temperature conditions (spatially averaged versus spatially resolved) both affect Ag release predictions. This suggests that previously observed model inadequacy can be attributed to both physical models and the operating conditions supplied to them. These findings will be leveraged to guide development of assessment cases for additional experiments and refinement of physical models for behaviors that potentially influence Ag release, such as palladium penetration.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1973

Cavity temperature and flow characteristics in a gas-core test reactor

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Putre, H. A.↗

DOE OSTI2014

AGR-2 Data Qualification Report for ATR Cycle 154B

This report provides the data qualification status of Advanced Gas Reactor-2 (AGR-2) fuel irradiation experimental data from Advanced Test Reactor (ATR) Cycle 154B as recorded in the Nuclear Data Management and Analysis System (NDMAS). This is the last cycle of AGR-2 irradiation, as the test train was pulled from the ATR core during the outage portion of ATR Cycle 155A. The AGR-2 data streams addressed in this report include thermocouple (TC) temperatures, sweep gas data (flow rates including new Fission Product Monitoring (FPM) downstream flows from Fission Product Monitoring System (FPMS) detectors, pressure, and moisture content), and FPMS data (release rates and release-to-birth rate ratios [R/Bs]) for each of the six capsules in the AGR-2 experiment. The final data qualification status for these data streams is determined by a Data Review Committee (DRC) comprised of AGR technical leads, Sitewide Quality Assurance (QA), and NDMAS analysts. The Data Review Committee reviewed the data acquisition process, considered whether the data met the requirements for data collection as specified in QA-approved Very High Temperature Reactor (VHTR) data collection plans, examined the results of NDMAS data testing and statistical analyses, and confirmed the qualification status of the data as given in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Multiscale, mechanistic modeling of irradiation-enhanced silver diffusion in TRISO particles

Tristructural isotropic (TRISO) particles are under consideration for use in several proposed advanced nuclear reactor concepts. The silicon carbide (SiC) layer in TRISO acts as a barrier to prevent the release of the fission products. However, despite remarkable retention, silver (Ag) release has been observed from intact particles, which requires investigation since the Ag isotope ( 110m Ag) has a long half-life. Previous work focused on developing a multiscale, mechanistic model for Ag diffusion accounting for temperature and microstructure effect and has been successfully validated. In this work, we expand the previous model to account for irradiation-enhanced Ag diffusivity in SiC and improve its accuracy over a wider grain size and temperature ranges relevant for advanced reactor conditions. A temperature, grain size, and flux dependent diffusivity is therefore derived using the mesoscale code MARMOT and implemented in the fuel performance code BISON. The irradiation-enhanced Ag diffusivity in SiC is compared against experimental data and validated using BISON against Ag release measurements from the Advanced Gas Reactor Fuel Development and Qualification Program (AGR-1 and AGR-2). Herein, we quantify the impact of SiC grain size, irradiation, and temperature on Ag release. In agreement with previous studies, we find accounting for SiC grain size improves agreement between BISON predictions and experimental observations for most cases. In conclusion, we also find that accounting for irradiation improves agreement for cases where Ag release was underestimated, but the impact was less significant than accounting for microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Test Plan for Reirradiation and 1400°C Heating Test of AGR-3/4 Compact 10-1

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 are in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015b). This test plan describes the reirradiation and subsequent heating test of AGR-3/4 Compact 10-1. One of the main goals of this test is to assess the retention of short-lived fission products I-131 (t1/2 = 8.02 d) and Xe-133 (t1/2 = 5.2 d) in fuel kernels. To generate short-lived fission products, the compact will first be reirradiated in the Neutron Radiography Reactor (NRAD). Following reirradiation, the compact will be removed from NRAD, gamma scanned on the Precision Gamma Scanner (PGS), and then loaded into the Fuel Accident Condition Simulator (FACS) furnace for heating at 1400°C. Fission gases and condensable fission products released from the compact will be collected in the Fission Gas Monitoring System (FGMS) and the FACS condensation plates, respectively.

42 ENGINEERING↗

DOE OSTI2019

Test Plan for Reirradiation and 1600°C Heating Test of AGR-3/4 Compact 3-1

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 are in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015b). This test plan describes the reirradiation and subsequent heating test of AGR-3/4 Compact 3-1. One of the main goals of this test is to assess the retention of short-lived fission products I-131 (t1/2 = 8.02 d) and Xe-133 (t1/2 = 5.2 d) in fuel kernels. To generate short-lived fission products, the compact will first be reirradiated in the Neutron Radiography Reactor (NRAD). Following reirradiation, the compact will be removed from NRAD, gamma scanned on the Precision Gamma Scanner (PGS), and then loaded into the Fuel Accident Condition Simulator (FACS) furnace for heating at 1600°C. Fission gases and condensable fission products released from the compact will be collected in the Fission Gas Monitoring System (FGMS) and the FACS condensation plates, respectively. The planned uses of AGR-3/4 compacts (including heating testing and destructive analysis) are discussed in the AGR-3/4 PIE plan (Demkowicz 2017).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

AGR-3/4 PIE at Oak Ridge National Laboratory

The Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the third and fourth test irradiations, Advanced Gas Reactor-3/4 (AGR-3/4), began at INL in April 2015. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing technical input, preparations for PIE testing and analysis, and contributing general expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

AGR-3/4 PIE at Oak Ridge National Laboratory

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Design characteristics of the Idaho National Laboratory high-temperature gas-cooled test reactor - 18703

A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG-TR) has been developed by Idaho National Laboratory as part of a U.S. Department of Energy initiative to explore and potentially expand existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200-MW HTG-TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325/650 deg. C and utilizes tri-structural isotropic particle fuel with an enrichment of 15.5-wt% {sup 235}U. The primary mission of the HTG-TR is material irradiation, and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast-neutron fluxes. The highest thermal-neutron flux (3.90 E+14 n/cm{sup 2}/s) occurs in the outer reflector, and the maximum fast-flux levels (1.17 E+14 n/cm{sup 2}s) are produced in the central reflector column, where most of the graphite has been removed. The core features a large number of irradiation positions with large test volumes and long test lengths, providing an ideal environment for thermal-neutron irradiation of large test articles. The total available test volume is more than 1,100 L. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, and light water) from the helium primary coolant system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

High-temperature irradiation-resistant thermocouple instability model for in-pile reactor use

This article presents an instability model for the high-temperature irradiation-resistant thermocouple (HTIR-TC). Here the term instability defines the superposition of both drift and inhomogeneity of TC thermoelements occurring simultaneously. The HTIR-TC is an advanced thermocouple (TC) that uses the refractory metals niobium and molybdenum as sensing thermoelements for generating electromotive force (EMF) in a field of neutrons and at temperatures upward of 1,600°C. In the Advanced Gas Reactor (AGR) 5/6/7 tests conducted at Idaho National Laboratory’s Advanced Test Reactor (ATR), the HTIR-TCs showed low to moderate instability throughout the life of the test. The instability model reveals that HTIR-TCs can, when the operating temperature of the reactor fuel is normal, maintain performance throughout an 18-month refueling cycle typical of nuclear power plants, reflecting an instability of less than ±1%. The HTIR-TC is also qualified for incorporation into a test fixture during the testing of new fuels.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

Unpacking model inadequacy: The quantification of silver release from TRISO fuel by considering empirical and mechanistic approaches

Increasing adoption of the proposed tristructural isotropic (TRISO) particle fuel for both advanced and existing reactors makes it critical to assess and address any uncertainties and inadequacies of TRISO fission product release models. Model inadequacy stems from simplifications made to the computational model when compared to the experiments. The modeling and simulation efforts conducted using the BISON fuel performance code, along with the experimental campaigns carried out under the Advanced Gas Reactor Fuel Development and Qualification Program, afford a unique opportunity to conduct a rigorous modeling inadequacy assessment within the Bayesian uncertainty quantification (UQ) framework. Here, this study compares the standard Bayesian framework against the Kennedy-O'Hagan (KOH) framework, which explicitly represents modeling inadequacy, in regard to UQ for TRISO silver release models. For this purpose, both the traditional Arrhenius equation fitted to experimental data and the more advanced lower-length-scale (LLS)-informed model, which considers microstructure information, are independently considered. Applying the inverse UQ process on the AGR-2 and -3/4 datasets revealed modeling inadequacy to be the most dominant source of uncertainty. Experimental noise uncertainty is also significant; however, model parameter uncertainty can be considered negligible. Interestingly, both the Arrhenius equation and the LLS-informed model demonstrated similar levels of modeling inadequacy. For the forward predictive UQ, the KOH framework improved both the accuracy and quality of quantified uncertainties in comparison to the standard Bayesian framework. This is true for both the Arrhenius equation and the LLS-informed model. In comparing these modeling approaches, both demonstrated similar performance at the engineering scale, while the LLS-informed model expectedly outperformed the Arrhenius equation at the mesoscale. These conclusions highlight the importance of explicitly accounting for modeling inadequacy in the UQ process, and reinforce the need for continuous refinement of physics-based models in order to address the modeling inadequacy.

Advanced reactors↗

DOE OSTI2018

AGR-3/4 PIE at Oak Ridge National Laboratory

The Advanced Reactor Technologies (ART) Technology Development Office (TDO) is currently supporting a tristructural isotropic (TRISO) fuel development and qualification program, which includes fuel fabrication, test irradiations, and post-irradiation examination (PIE) and safety testing to assess fuel performance during normal irradiation and under accident conditions. PIE work on fuel from the third and fourth test irradiations, Advanced Gas Reactor-3/4 (AGR-3/4), will begin at Idaho National Laboratory (INL) in approximately April 2015. The work scope in this memorandum purchase order (MPO) includes Oak Ridge National Laboratory (ORNL) providing technical input, developing equipment and techniques for small-area deconsolidation of compacts, and initiating a system design for hot cell operations (expected to begin in about April 2015), and contributing expertise to this effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

PARFUME Theory and Model Basis Report

The success of gas reactors depends upon the safety and quality of the coated particle fuel. The fuel performance modeling code PARFUME simulates the mechanical, thermal and physico-chemical behavior of fuel particles during irradiation. This report documents the theory and material properties behind vari¬ous capabilities of the code, which include: 1) various options for calculating CO production and fission product gas release, 2) an analytical solution for stresses in the coating layers that accounts for irradiation-induced creep and swelling of the pyrocarbon layers, 3) a thermal model that calculates a time-dependent temperature profile through a pebble bed sphere or a prismatic block core, as well as through the layers of each analyzed particle, 4) simulation of multi-dimensional particle behavior associated with cracking in the IPyC layer, partial debonding of the IPyC from the SiC, particle asphericity, and kernel migration (or amoeba effect), 5) two independent methods for determining particle failure probabilities, 6) a model for calculating release-to-birth (R/B) ratios of gaseous fission products that accounts for particle failures and uranium contamination in the fuel matrix, and 7) the evaluation of an accident condition, where a particle experiences a sudden change in temperature following a period of normal irradiation. The accident condi¬tion entails diffusion of fission products through the particle coating layers and through the fuel matrix to the coolant boundary. This document represents the initial version of the PARFUME Theory and Model Basis Report. More detailed descriptions will be provided in future revisions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Concept of Operations for Microreactor Transportation

Microreactors are very small nuclear reactors with a power output of 20 MWe or less and are designed to be factory-built, modular in nature, and highly portable. The Idaho National Laboratory (INL) has been identified as a potential location to demonstrate microreactors and two microreactor concepts currently under consideration for demonstration at the INL include a tristructural isotropic (TRISO) fueled high-temperature gas reactor and a sodium cooled reactor using a SNAP-10A style fuel pin design. The objective of this report is to identify a concept of operations (CONOPS) for shipping these microreactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

A Research Program for Fission Product/Dust Transport in HTGR’s

High and Very High Temperatures Gas Reactors (HTGRs/VHTRs) have five barriers to fission product (FP) release: the TRISO fuel coating, the fuel elements, the core graphite, the primary coolant system, and the reactor building. This project focused on measurements and computations of FP diffusion in graphite, FP adsorption on graphite and FP interactions with dust particles of arbitrary shape. Diffusion Coefficients of Cs and Iodine in two nuclear graphite were obtained by the release method and use of Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Instrumented Neutron Activation Analysis (INAA). A new mathematical model for fission gas release from nuclear fuel was also developed. Several techniques were explored to measure adsorption isotherms, notably a Knudsen Effusion Mass Spectrometer (KEMS) and Instrumented Neutron Activation Analysis (INAA). Some of these measurements are still in progress. The results will be reported in a supplemental report later. Studies of FP interactions with dust and shape factors for both chain-like particles and agglomerates over a wide size range were obtained through solutions of the diffusion and transport equations. The Green's Function Method for diffusion and Monte Carlo technique for transport were used, and it was found that the shape factors are sensitive to the particle arrangements, and that diffusion and transport of FPs can be hindered. Several journal articles relating to the above work have been published, and more are in submission and preparation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1973

Space shuttle aps propellant thermal conditioner study

An analytical and experimental effort was completed to evaluate a baffle type thermal conditioner for superheating O2 and H2 at supercritical pressures. The thermal conditioner consisted of a heat exchanger and an integral reactor (gas generator) operating on O2/H2 propellants. Primary emphasis was placed on the hydrogen conditioner with some effort on the oxygen conditioner and a study completed of alternate concepts for use in conditioning oxygen. A hydrogen conditioner was hot fire tested under a range of conditions to establish ignition, heat exchange and response parameters. A parallel technology task was completed to further evaluate the integral reactor and heat exchanger with the side mounted electrical spark igniter.

Fulton, D. L.↗