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DOE OSTI2015

PIE on Safety-Tested AGR-1 Compact 5-1-1

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

PIE on safety-tested AGR-1 compact 4-2-2

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program [Petti et al. 2010]; this work continues to be funded by the Department of Energy's Office of Nuclear Energy in support of Advanced Reactor Technologies (ART). AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for 3 years in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE recently completed at ORNL on irradiated AGR-1 Compact 4-2-2.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Development of a Radial Deconsolidation Method

A series of experiments have been initiated to determine the retention or mobility of fission products* in AGR fuel compacts [Petti, et al. 2010]. This information is needed to refine fission product transport models. The AGR-3/4 irradiation test involved half-inch-long compacts that each contained twenty designed-to-fail (DTF) particles, with 20-μm thick carbon-coated kernels whose coatings were deliberately fabricated such that they would crack under irradiation, providing a known source of post-irradiation isotopes. The DTF particles in these compacts were axially distributed along the compact centerline so that the diffusion of fission products released from the DTF kernels would be radially symmetric [Hunn, et al. 2012; Hunn et al. 2011; Kercher, et al. 2011; Hunn, et al. 2007]. Compacts containing DTF particles were irradiated at Idaho National Laboratory (INL) at the Advanced Test Reactor (ATR) [Collin, 2015]. Analysis of the diffusion of these various post-irradiation isotopes through the compact requires a method to radially deconsolidate the compacts so that nested-annular volumes may be analyzed for post-irradiation isotope inventory in the compact matrix, TRISO outer pyrolytic carbon (OPyC), and DTF kernels. An effective radial deconsolidation method and apparatus appropriate to this application has been developed and parametrically characterized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Silicon Carbide Temperature Monitor Processing Improvements. Status Report

Silicon carbide (SiC) temperature monitors are used as temperature sensors in Advanced Test Reactor (ATR) irradiations at the Idaho National Laboratory (INL). Although thermocouples are typically used to provide real-time temperature indication in instrumented lead tests, other indicators, such as melt wires, are also often included in such tests as an independent technique of detecting peak temperatures incurred during irradiation. In addition, less expensive static capsule tests, which have no leads attached for real-time data transmission, often rely on melt wires as a post-irradiation technique for peak temperature indication. Melt wires are limited in that they can only detect whether a single temperature is or is not exceeded. SiC monitors are advantageous because a single monitor can be used to detect for a range of temperatures that occurred during irradiation. As part of the process initiated to make SiC temperature monitors available at the ATR, post-irradiation evaluations of these monitors have been previously completed at the High Temperature Test Laboratory (HTTL). INL selected the resistance measurement approach for determining irradiation temperature from SiC temperature monitors because it is considered to be the most accurate measurement. The current process involves the repeated annealing of the SiC monitors at incrementally increasing temperature, with resistivity measurements made between annealing steps. The process is time consuming and requires the nearly constant attention of a trained staff member. In addition to the expensive and lengthy post analysis required, the current process adds many potential sources of error in the measurement, as the sensor must be repeatedly moved from furnace to test fixture. This time-consuming post irradiation analysis is a significant portion of the total cost of using these otherwise inexpensive sensors. An additional consideration of this research is that, if the SiC post processing can be automated, it could be performed in an MFC hot cell, further reducing the time and expense of lengthy decontaminations prior to processing. Sections of this report provide a general description of resistivity techniques currently used to infer peak irradiation temperature from silicon carbide temperature monitors along with some representative data, the proposed concepts to improve the process of analyzing irradiated SiC temperature monitors, the completed efforts to prove the proposed concepts, and future activities. The efforts detailed here succeeded in designing and developing a real-time automated SiC resistivity measurement system, and performed two initial test runs. Activities carried out include the assembly and integration of the system hardware; the design and development of a preliminary monitor fixture; the design of a technique to automate the data analysis and processing; the development of the communication, coordination, and user software; and the execution and troubleshooting of test run experiments using the box furnace. Although the automation system performed as required, the designed fixture did not succeed in establishing the needed electrical contacts with the SiC monitor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Status Report on Irradiation Capsules Designed to Evaluate FeCrAl-UO 2 Interactions

This status report provides the background and current status of a series of irradiation capsules that were designed and are being built to test the interactions between candidate FeCrAl cladding for enhanced accident tolerant applications and prototypical enriched commercial UO 2 fuel in a neutron radiation environment. These capsules will test the degree, if any, of fuel cladding chemical interactions (FCCI) between FeCrAl and UO 2 . The capsules are to be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory to burn-ups of 10, 30, and 50 GWd/MT with a nominal target temperature at the interfaces between the pellets and clad of 350°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Acceptance Test Data for BWXT Coated Particle Batch 93164A Defective IPyC Fraction and Pyrocarbon Anisotropy

Coated particle fuel batch J52O-16-93164 was produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used as demonstration production-scale coated particle fuel for other experiments. The tristructural-isotropic (TRISO) coatings were deposited in a 150-mm-diameter production-scale fluidizedbed chemical vapor deposition (CVD) furnace onto 425-μm-nominal-diameter spherical kernels from BWXT lot J52L-16-69316. Each kernel contained a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO) and was coated with four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (i.e., 93164A).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Acceptance Test Data for Candidate AGR-5/6/7 TRISO Particle Batches BWXT Coater Batches 93165 93172 Defective IPyC Fraction and Pyrocarbon Anisotropy

Coated particle fuel batches J52O-16-93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). Some of these batches may alternately be used as demonstration coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μmnominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Improvement of INVS Measurement Uncertainty for Pu and U-Pu Nitrate Solution

In the Tokai Reprocessing Plant (TRP) and the Plutonium Conversion Development Facility (PCDF), a large amount of plutonium nitrate solution which is recovered from light water reactor (LWR) and advanced thermal reactor (ATR), FUGEN are being stored. Since the solution is designated as a direct use material, the periodical inventory verification and flow verification are being conducted by Japan Safeguard Government Office (JSGO) and International Atomic Agency (IAEA).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

SSHAC Level 1 Probabilistic Seismic Hazard Analysis for the Idaho National Laboratory

A Probabilistic Seismic Hazard Analysis (PSHA) was completed for the Materials and Fuels Complex (MFC), Naval Reactors Facility (NRF), and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) (Figure 1-1). The PSHA followed the approaches and procedures appropriate for a Study Level 1 provided in the guidance advanced by the Senior Seismic Hazard Analysis Committee (SSHAC) in U.S. Nuclear Regulatory Commission (NRC) NUREG/CR-6372 and NUREG-2117 (NRC, 1997; 2012a). The SSHAC Level 1 PSHAs for MFC and ATR were conducted as part of the Seismic Risk Assessment (SRA) project (INL Project number 31287) to develop and apply a new-risk informed methodology, respectively. The SSHAC Level 1 PSHA was conducted for NRF to provide guidance on the potential use of a design margin above rock hazard levels. The SRA project is developing a new risk-informed methodology that will provide a systematic approach for evaluating the need for an update of an existing PSHA. The new methodology proposes criteria to be employed at specific analysis, decision, or comparison points in its evaluation process. The first four of seven criteria address changes in inputs and results of the PSHA and are given in U.S. Department of Energy (DOE) Standard, DOE-STD-1020-2012 (DOE, 2012a) and American National Standards Institute/American Nuclear Society (ANSI/ANS) 2.29 (ANS, 2008a). The last three criteria address evaluation of quantitative hazard and risk-focused information of an existing nuclear facility. The seven criteria and decision points are applied to Seismic Design Category (SDC) 3, 4, and 5, which are defined in American Society of Civil Engineers/Structural Engineers Institute (ASCE/SEI) 43-05 (ASCE, 2005). The application of the criteria and decision points could lead to an update or could determine that such update is not necessary.

58 GEOSCIENCES↗
DOE OSTI2017

X-ray Analysis of Defects and Anomalies in AGR-5/6/7 TRISO Particles

Coated particle fuel batches J52O-16-93164, 93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used for other tests. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.4%-enriched uranium carbide and uranium oxide (UCO), with the exception of Batch 93164, which used similar kernels from BWXT lot J52L-16-69316. The TRISO-coatings consisted of a ~50% dense carbon buffer layer with 100-μmnominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. Each coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (e.g., 93164A). Secondary upgrading by sieving was performed on the upgraded batches to remove specific anomalies identified during analysis for Defective IPyC, and the upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B). Following this secondary upgrading, coated particle composite J52R-16-98005 was produced by BWXT as fuel for the AGR Program’s AGR-5/6/7 irradiation test in the INL ATR. This composite was comprised of coated particle fuel batches J52O-16-93165B, 93168B, 93169B, and 93170B.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Heating Tests of AGR-2 Compact 6-4-1 Loose Kernels and AGR-3/4 Compacts 3-2 and 10-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching goal of the program is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety/heating testing, fuel performance modeling, and fission product transport (INL 2016). Several fuel irradiation experiments have been performed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), and additional irradiations are planned. These experiments are intended to provide data on fuel performance under irradiation, support fuel fabrication process development, qualify fuel for operating and accident conditions, provide irradiated fuel for accident testing, and support development of fuel performance and fission product transport models.

42 ENGINEERING↗
DOE OSTI2012

AGR-1 Irradiation Test Final As-Run Report

This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 10 25 n/m 2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10 -7 with only one capsule significantly exceeding this value. A maximum R/B of around 2 x 10 -7 was reached at the end of the irradiation in Capsule 5. Several shakedown issues were encountered and resolved during the first three cycles. These include the repair of minor gas line leaks; repair of faulty gas line valves; the need to position moisture monitors in regions of low radiation fields for proper functioning; the enforcement of proper on-line data storage and backup, the need to monitor thermocouple performance, correcting for detector spectral gain shift, and a change in the mass flow rate range of the neon flow controllers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

ATF material model development and validation for priority fuel concepts

Development, implementation, and validation of material and behavior models for accident tolerant fuel (ATF) concepts in the Bison fuel performance code began in 2014 in response to the events that occurred at the Fukushima Daichii nuclear power plant in March 2011. Early on the focus was on U 3 Si 2 fuel and FeCrAl cladding as part of a high impact problem through the Nuclear Energy and Advanced Modeling Simulation (NEAMS) program. Then, developments for Cr 2 O 3 -doped UO 2 fuel, and SiC-SiC and Cr-coated zirconium-based claddings began based upon industry interests. In late fiscal year 2018 the Consortium for Advanced Simulation of Light Water Reactors (CASL) took over further ATF work in Bison in support of the Nuclear Regulatory Commission (NRC) engagement. Discussions with the NRC identified their list of priority fuel and cladding concepts, which included Cr 2 O 3 -doped UO 2 and U 3 Si 2 fuels, and Cr-coated zirconium-based and FeCrAl claddings. In particular, the NRC suggested that reports similar in form to NUREG/CR-7024 [1] that was developed for traditional LWR materials UO 2 and zirconium-based claddings (i.e., Zircaloy-4, M5®, ZIRLO) be created for the priority ATF concepts. The approach to ATF capability development in Bison since the beginning has been two-fold: (1) empirical correlations and (2) multiscale model development. Both approaches have uncertainty inherent to them. Uncertainty in empirical correlations is bounded by the experimental data upon which with the correlation was developed. Models developed through a multiscale approach have uncertainty associated with the lower length scale calculations and input parameters that must be propagated to the engineering scale model in Bison. In this report, the recommended models, their range of applicability (e.g., temperature, burnup), and associated uncertainty for the NRC priority fuel concepts Cr 2 O 3 -doped UO 2 and U 3 Si 2 are presented in a manner similar to the aforementioned NUREG. In addition, the Cr 2 O 3 -doped UO 2 models are validated to the Halden Reactor tess IFA-677.1 rods 1 and 5 and IFA-716.1 rod 1. For U 3 Si 2 the models are validated to the recent post-irradiation examination (PIE) data ATF-13 and ATF-15 rods that were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory. Finally, an uncertainty quantification (UQ) and sensitivity analysis (SA) is completed on a single rod for each fuel concept that takes into account the defined uncertainty in select models. The results of the UQ and SA analyses indicate that given the large uncertainty in some of the input ATF models that the uncertainty on the Bison predictions of fuel performance metrics of interest bound the available experimental data. Correlation coefficients are also reported that identified the uncertain inputs with the strongest correlation (positive or negative) with the outputs of interest. Fuel performance metrics investigated include fuel elongation, rod internal pressure, fuel centerline temperature, and fission gas release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Preliminary Evaluation of Loading DOE Standardized Canisters in the CPP-603 Irradiated Fuel Storage Facility

This evaluation looks at the equipment and operations necessary to load United States Department of Energy (DOE) -owned Spent Nuclear Fuel (SNF) into DOE SNF Standardized Canisters in the CPP-603 Irradiated Fuel Storage Facility in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL). Two types of fuels are looked at in this evaluation: Advanced Test Reactor (ATR) fuel (uranium-aluminide fuel with aluminum cladding) and Peach Bottom fuel (thorium-uranium carbide fuel in a graphite matrix). The ATR fuel ready for loading would come from fuel storage canisters in the CPP-603 facility, while the Peach Bottom fuel ready for loading could come from either fuel storage canisters in the CPP-603 facility and or containers in the CPP-749 Underground Fuel Storage Facility – Generation I Vaults.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development of a Fast Spectrum Test Loop at the Advanced Test Reactor

The United States currently has no infrastructure to provide a fast neutron flux for research purposes. With the increased interest in Sodium Fast Reactor (SFR) technology, this type of infrastructure is necessary for further development of key components such as Versatile Test Reactor. The purpose of this research is to characterize the thermal hydraulics of a test loop capable of providing SFR like operating conditions. This includes showcasing feasibility in mechanical design, reaching a high fast to thermal neutron flux ratio ideally near 15, an inlet temperature of 600 K, a temperature increase of 50 K from inlet to outlet, and test chamber radial temperature profiles similar to SFR fuel rods. Existing infrastructure must be used to provide the neutron flux for the test loop, and it was decided that the Advanced Test Reactor (ATR) was the optimal choice due to its availability and previous work conducted similar to this project. To achieve the goals of this project, several mechanical designs were put forward with the most feasible design being chosen. The neutronics and thermal hydraulics analysis were conducted in tandem, with an iterative process used to maximize efficiency of both systems. Analysis of the neutronics system was done using Attila and MCNP while the thermal hydraulic analysis was done using RELAP5. This document outlines the thermal hydraulic analysis, though it includes the results from the mechanical design and neutronics analysis. The loop design includes a primary system with molten sodium, a set of “booster fuel plates” to increase the fast flux within the test section, a water system to cool the booster fuel plates, a helium system to separate the sodium and water systems, and a thermal neutron shield to decrease the thermal flux. Each of these systems is modeled in the neutronics and thermal hydraulic analyses. The RELAP5 model showcased the feasibility of the design proposed by this report. For all designs tested, the desired 50 K temperature increase from inlet to outlet was reached. The fuel rod radial temperature profile was lower than was expected, likely due to a lower rod thermal power than expected. A parametric study was also conducted to characterize the system with three fuel pins. This study showed the effect of thermal power, fluid pressure, fluid velocity, and inlet temperature on the system. ©Copyright by Anton Higgins December 11, 2019 All Rights Reserved

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

AGC-3 Specimen Load Calculations by Stack

This Engineering Calculations and aAnalysis Report (ECAR) documents the results of the threshold averaging on the load cell data taken during the third advanced graphite creep (AGC) test, AGC 3. Specimens were irradiated in the Advance Test Reactor (ATR) during Cycles 152B, 154B, 155A, and 155B (the AGC-3 capsule was not in ATR for Cycles 153A/B and 154A). During each cycle, the specimens located in the upper portions of the stacks were subjected to a compressive load. The applied load of each stack was monitored and load data were recorded in 1 minute intervals. Collecting data detailing the loads applied over the life of the experiment is necessary for use in future analyses. This load summary data will be used in creep rate estimations and post irradiation examination of material properties.

42 ENGINEERING↗
DOE OSTI2017

Technical Evaluation Report for the NDMAS AGR-5/6/7 TRISO Irradiation Monitoring System

The Nuclear Data Management and Analysis System (NDMAS) manages irradiation experiment data collected under the Advanced Gas Reactor (AGR) Fuel Development and Qualification program within the Advanced Reactor Technologies Technology Development Office. The AGR-5/6/7 irradiation experiment combines the fifth, sixth, and seventh planned irradiation campaigns of the U.S. Department of Energy AGR Fuel Development and Qualification Program. These combined irradiation tests will consist of fuel qualification tests (AGR-5/6) and a margin test (AGR-7) for the industrially produced uranium oxycarbide (UCO) tristructural isotropic (TRISO) coated particle fuel developed by the AGR program. The objective of the AGR-5/6/7 irradiation experiment is to establish the performance of UCO TRISO fuel fabricated using engineering-scale equipment in a production environment under and beyond normal operating conditions (PNL-5245). Experimental irradiation measurements for this experiment (i.e., thermocouple temperature, gas flow rate, fission product release rate, gross gamma count, and Advanced Test Reactor (ATR) operational parameters) are captured automatically to the NDMAS database. Data derived from numerical simulation or other mathematical analyses (i.e., fast fluence, burnup, temperature, and fission product release to birth ratio) are captured to the database when provided by the researchers. The NDMAS database is a relational database accessed via structured query language and managed with Microsoft SQL Server. While select users can directly access the database, that is not the most convenient way to interact with the data. The NDMAS team constructs numerous tables that combine related datasets, and delivers those data to researchers via graphs and spreadsheets on a Microsoft SharePoint-based website. For the AGR-5/6/7 experimental data, a new subsite has been developed, along with a set of web pages. These pages will be used by members of fuel irradiation team for monitoring irradiation conditions according to specification (SPC-1749), by the physics modelers to access necessary model inputs, and by the nuclear engineering community to view and access data such as fuel temperatures, fission product release from capsules, and ATR operational parameters. The deliverable represented in this document is a SharePoint website containing tables of the AGR-5/6/7 measured and calculated data, graphs displaying the data, and links to source documents and references.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Additional Confirmatory LBL Analysis of AGR-5/6/7 Compacts and Over-Coated Particles

Fuel compacts for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) were fabricated by BWX Technologies Nuclear Operations Group (BWXT NOG) located in Lynchburg, Virginia. Two compact packing fractions (PFs) were produced - nominally, 40% PF and 25% PF - in which the tristructural-isotropic (TRISO)-coated particle volume was targeted to be approximately 40% and 25% of the total compact volume, respectively. The TRISO coatings were deposited using a 150 mm diameter fluidized-bed chemical vapor deposition (CVD) furnace on spherical kernels that were nominally 425 μm in diameter. The kernels were from kernel composite Lot J52R-16-69317, which contained low-enriched uranium (15.5% 235 U) in a mixture of uranium carbide and uranium oxide (UCO). Kernels were coated with four concentric CVD layers: a porous carbon buffer layer that was nominally 100 μm thick, an inner pyrolytic carbon (IPyC) layer that was nominally 40 μm thick, a silicon carbide (SiC) layer that was nominally 35 μm thick, and an outer pyrolytic carbon (OPyC) layer that was nominally 40-μm-thick. Coated particle composite J52R-16-98005 was overcoated (OC) with a graphite/resin blend, and these OC TRISO particles were pressed to form cylindrical compacts that were nominally a half inch in diameter and one inch long.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗