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DOE OSTI2013

Summary of the Preliminary Optical ICHMI Design Study: A Preliminary Engineering Design Study for a Standpipe Viewport

This summary report examines an in-vessel optical access concept intended to support standoff optical instrumentation, control and human-machine interface (ICHMI) systems for future advanced small modular reactor (AdvSMR) applications. Optical-based measurement and sensing systems for AdvSMR applications have several key benefits over traditional instrumentation and control systems used to monitor reactor process parameters, such as temperature, flow rate, pressure, and coolant chemistry (Anheier et al. 2013). Direct and continuous visualization of the in-vessel components can be maintained using external cameras. Many optical sensing techniques can be performed remotely using open optical beam path configurations. Not only are in-vessel cables eliminated by these configurations, but also sensitive optical monitoring components (e.g., electronics, lasers, detectors, and cameras) can be placed outside the reactor vessel in the instrument vault, containment building, or other locations where temperatures and radiation levels are much lower. However, the extreme AdvSMR environment present challenges for optical access designs and optical materials. Optical access is not provided in any commercial nuclear power plant or featured in any reactor design, although successful implementation of optical access has been demonstrated in test reactors (Arkani and Gharib 2009). This report outlines the key engineering considerations for an AdvSMR optical access concept. Strict American Society of Mechanical Engineers (ASME) construction codes must be followed for any U.S. nuclear facility component (ASME 2013); however, the scope of this study is to evaluate the preliminary engineering issues for this concept, rather than developing a nuclear-qualified design. In addition, this study does not consider accident design requirements. In-vessel optical access using a standpipe viewport concept serves as a test case to explore the engineering challenges and performance requirements for sodium fast reactor (SFR) and high-temperature gas reactor (HTGR) AdvSMR applications. The expected environmental conditions for deployment are reviewed for both AdvSMR designs. Optical and mechanical materials that maximize component lifetime are evaluated for the standpipe viewport design under these conditions. Optical components and opto-mechanical designs that provide robust optical-to-metal seals and stress-free optical component mounting are identified, and then key performance specifications are developed for a sapphire optical viewport concept. Design strategies are examined that protect the internal optical surfaces from liquid-coolant condensation and impurity deposits. Finally, a conceptual standpipe viewport design that is suggestive of how this concept could be assembled using standard nuclear-qualified pipe components, is presented.

42 ENGINEERING↗

DOE OSTI2023

Safeguards Facility Model for Lead-cooled and Gas-cooled Fast Reactors

Efforts have been underway for many years on the development of safeguards for fast neutron reactors. The efforts have been primarily for sodium-cooled fast reactors (SFRs) in cooperation with Japan, to properly implement these safeguards as the reactors are conceptualized and ultimately commercialized. The primary challenge to development of adequate safeguard approaches for fast reactors has been in understanding how to apply different aspects of safeguards when facilities widely vary from one type of fast reactor to another and is largely dependent on several key aspects of the specific reactor type. For this report, the following key areas were identified for further analysis to determine where Lead-cooled Fast Reactors (LFRs) and Gas-Cooled Fast Reactors (GCFR) fuel material flow and key measurement points would differ from previously studied SFR safeguards.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Irradiated AGR 2 Compact 4-3-1 Examination Plan

This plan describes post-irradiation examination activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory on irradiated Compact 4-3-1 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with the Statement of Work for PIE and Safety Testing of Pebble Bed Modular Reactor Limited Fuel at ORNL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Thermal Model Heat Rate Predictions of the AGR-5/6/7 Experiment

Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train. AGR-5/6/7 will test low-enriched uranium oxycarbide (UCO – a heterogeneous mixture of uranium oxide and uranium carbide) TRISO fuel. The AGR-5/6 portion of the experiment will provide data to support qualification of the selected reference fuel design, while the AGR-7 portion will serve as a margin test, irradiating the fuel beyond normal operating conditions. During irradiation, the temperature in each capsule is controlled by varying the helium / neon gas mixture flowing through the capsule until the desired thermocouple temperature is reached. Further adjustments can be made by changing the lobe power and the neutron filter. Three neutron filters are available to use as the experiment burns up the fissionable fuel. A thermal finite element model has been created for the five capsules comprising the AGR-5/6/7 experiment. Heat rates from a physics analysis are imported into the model along with fast neutron fluence. Graphite shrinkage due to the fast neutron fluence is incorporated into the model. Gas gaps change as a function of fast neutron fluence. Radiation heat transfer is a major contributor in this model. This is a large model with more than 1 million finite element brick elements. More than 150 different parts are modeled in the finite element model and communicate with each other from a heat transfer sense. More than 50 thermocouples are used in the experiment and are compared to actual measurements. The experiment is composed of five separate stainless steel capsules all welded together. This paper presents a method used to predict the gas mixture in the event that all of the thermocouples fail in a particular capsule. Curve fitting of heat rates from previous cycles and projecting them into the future is the basis of the method.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Thermal Model of the AGR-5/6/7 Experiment

Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train, AGR-5/6/7 will test low-enriched uranium oxycarbide (UCO – a heterogeneous mixture of uranium oxide and uranium carbide) TRISO fuel. The AGR-5/6 portion of the experiment will provide data to support qualification of the selected reference fuel design, while the AGR-7 portion will serve as a margin test, irradiating the fuel beyond normal operating conditions. A thermal finite element model has been created for the five capsules comprising the AGR-5/6/7 experiment. Heat rates from a physics analysis are imported into the model along with fast neutron fluence. Graphite shrinkage due to the fast neutron fluence is incorporated into the model. Gas gaps change as a function of fast neutron fluence. Radiation heat transfer is a major contributor in this model. This is a large model with more than 1 million finite element brick elements. More than 150 different parts are modeled in the finite element model and communicate with each other from a heat transfer sense. More than 50 thermocouples are used in the experiment and are compared to actual measurements. The experiment is composed of five separate stainless steel capsules all welded together. Each capsule has its own Helium/.Neon gas mixture to help control the temperature of the capsule.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

An Experimental and Theoretical Examination of Air Ingress Rates During Small- and Medium-Break Air Ingress Accidents

Here, air ingress measurements using the 1/20th-scaled Helium Air Ingress gas Reactor Experiment (HAIRE) facility show key geometric variables of interest and their effect on air ingress in small- and medium-sized breaks in High Temperature Gas cooled Reactors. These variables include but are not limited to break diameter, break angle, and break wall thickness. Differing wall thicknesses for the same break diameter can have order-of-magnitude changes to the air ingress rate, which is a key figure of merit in the air ingress accident scenario. Additionally, different break sizes can change the importance of the angle in the break scenario. With smaller breaks, the flow will not transition from intermittent flow, to countercurrent flow, to diffusive flow as the break rotates from vertically upward toward vertically downward. This would lead to less variability with smaller breaks, which in turn would make the accident scenario more predictable for smaller-sized breaks.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2019

Completion of Level 2 Milestone – M2AT 19IN030401044, Complete construction of the gas and moisture supply systems and receive the custom-designed tube furnace from the vendor

The Advanced Gas Reactor (AGR) program has designed and is now constructing the Air/Moisture Ingress Experiment (AMIX) facility to test irradiated tristructural isotropic (TRISO) fuels under air or moisture-ingress accidents postulated for high-temperature gas-cooled reactors. AMIX will ultimately be installed in the Fuel Conditioning Facility at the Materials and Fuels Complex (MFC) at Idaho National Laboratory. AMIX consists of a gas supply (capable of supplying mixtures of air/helium and steam/helium), a tube furnace capable of accepting samples up to 6 cm in diameter and heating them to 1650°C, gamma detectors for measuring the release of condensable, gamma-emitting fission products, and cold traps with gamma detectors for measuring the release of fission gases. Data obtained from AMIX experiments will be used to qualify this new generation of TRISO fuels with the U.S. Nuclear Regulatory Commission, and inform models and simulations of TRISO fuel accident performance and source term.

42 ENGINEERING↗

DOE OSTI2022

AGR-5/6/7 Final Release-to-Birth Ratio Data Analysis

AGR-5/6/7 is the last of a series of Advanced Gas Reactor (AGR) experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low-enriched fuel for use in the high-temperature gas cooled reactor (HTGR). AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through the use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO-coated particle fuel. The AGR tests provide valuable irradiation-performance data to support fuel process development, qualify fuel for normal operation and accident conditions, and support development and validation of fuel performance and fission-product (FP) transport models and codes. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. Release-to-birth ratios (R/B) for fission-gas isotopes released from each capsule are calculated from release rates, measured by germanium detectors in the Fission Product Monitoring System (FPMS) installed downstream from each capsule, and birth rates calculated using numerical models of FG generation. The R/Bs are a critical measure of the ability of the fuel kernel, the particle coating layers, and the compact matrix to retain fission-gas atoms, preventing their release into the sweep-gas flow, and the impact of initially defective particles and/or particle-coating failures that occur during irradiation. For fission-gas isotopes, particle failure is defined as failure of all coating layers, allowing gaseous fission atoms to escape from a particle. During the first five cycles (162B ? 165A), R/Bs were stable in the 10-8?10-6 range, and no in-pile particle failures were observed, based on the gross gamma counts. The maximum R/B value of around 2 ? 10-6 for Kr-85m resulted from the presence of as-fabricated exposed kernels (based on the high exposed kernel fraction), the dispersed uranium, and high fuel particle temperatures in Capsule 1. Comparison of capsule-measured R/Bs from these early cycles to predictions using the previously developed AGR R/B model demonstrated FG release from the AGR-5/6/7 TRISO fuel was comparable to that of previous experiments. In addition, the Kr-85m R/B per-exposed-kernel values are comparable to R/B values obtained in AGR-3/4 irradiation experiment and four irradiation experiments performed during 1980s: (1) HRB-17/18, (2) COMEDIE-BD1, (3) HFR B1, and (4) HRB-21. In contrast, all measured R/B values are lower than predictions by the commonly used Richards and German models, which are intentionally conservative. A large number of in-pile particle failures occurred in Capsule 1 by the end of Cycle 166A. During the final four cycles (166A ? 168A), apparent damage to the Capsule 1 gas line appeared to cause FG leakage from that capsule into the other four capsules, resulting in an increase in fission gas (FG) detected in the effluent for all capsules. Isolation of the Capsule 1 gas line during the last three cycles also prevented measurement of its FG release. Thus, R/Bs in all capsules after Cycle 166A are highly uncertain because of undefined amount of leakage from Capsule 1, especially for long-lived isotopes. A few hundred in-pile particle failures were estimated for Capsule 1 before the end of Cycle 166A, but the total number of failures is unknown due to the lack of FG release data in the later cycles. Based primarily on evidence from the gross gamma counts during Cycle 168A, approximately 15 particles failed in Capsule 3 and four particles failed in Capsule 2. In-pile failures in Capsule 3 were anticipated because this capsule was designed to operate beyond the HTGR normal operating temperature range. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR model, developed in INL/EXT-14-32970, for R/B of the short-lived isotopes (Kr-89 and Xe-137) with minimal leakage from Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1969

Comments on the feasibility of developing gas core nuclear reactors

Recent developments in the fields of gas core hydrodynamics, heat transfer, and neutronics indicate that gas core nuclear rockets may be feasible from the point of view of basic principles. Based on performance predictions using these results, mission analyses indicate that gas core nuclear rockets may have the potential for reducing the initial weight in orbit of manned interplanetary vehicles by a factor of 5 when compared to the best chemical rocket systems. In addition, there is a potential for reducing total trip times from 450 to 500 days for chemical systems to 250 to 300 days for gas core systems. The possibility of demonstrating the feasibility of gas core nuclear rocket engines by means of a logical series of experiments of increasing difficulty that ends with ground tests of full scale gas core reactors is considered.

Rom, F. E.↗

DOE OSTI2015

Fission Products Distribution in TRISO Coated Fuel Particles Irradiated to 3.22 x 1021 n/cm2 Fast Fluence at 1092°C

Mechanisms by which fission products (especially Ag) migrate across the intact coating layers of tristructural isotropic (TRISO) coated fuel particles designed for next generation nuclear reactors have been the subject of numerous studies but remain poorly understood. This paper presents results obtained from the electron microscopic examination of selected irradiated tristructural isotropic (TRISO) coated particles (CPs) taken from fuel Compact 3-1-1 irradiated in Advanced Gas Reactor (AGR) experiment AGR-1. Studying this particles are of specific interest as it was fabricated using a different carrier gas composition ratio for the SiC layer deposition compared with the baseline fuel coated particles reported previously. These fabrication parameters are expected to reduce defects in the SiC layers resulting from uranium dispersion and potentially reducing the SiC layer permeability for metallic fission products. Basic scanning electron microscopy (SEM) and SEM montage investigations of the particles indicate that the number density of fission products in the inner pyrolytic carbon (IPyC) layer is far higher than that in the SiC layer, and that the number density of fission products in the SiC layer is far lower than the corresponding value for compacts with baseline fuel. Preliminary comparative fission product distribution montages suggest that the SiC layers of particles in Compact 1-3-1 are less permeable to fission products than those in compacts with baseline fuel. Transmission electron microscopy (TEM) samples were sectioned by focused ion beam (FIB) technique from the IPyC layer, the SiC layer and the IPyC-SiC interlayer of the particle. Detailed TEM and scanning transmission electron microscopy (STEM) coupled with energy dispersive X-ray spectroscopy (EDS) or electron energy loss spectroscopy (EELS) were performed to identify fission products and characterize their distribution across the IPyC and the SiC layers. Based on the distribution of the fission products, transport mechanisms are discussed and proposed, with emphasis placed on Ag. Our study contributes to enhancing the understanding of fission product transport mechanisms in TRISO coated fuel particles with SiC layer possessing reduced concentration of processing defects.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

AGR-5/6/7 Irradiation Disassembly and Metrology First Look

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 high-temperature gas-cooled reactors (HTGRs), which are graphite-moderated nuclear reactors cooled with helium. This work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing, deployment, and operation of HTGRs in the United States. To achieve these goals, the program includes fuel fabrication, irradiations of TRISO fuels and high-temperature materials (e.g., graphite), safety testing and post-irradiation examination (PIE), fuel performance modeling, and fission product transport and source term determination. The ART AGR program has conducted four distinct fuel irradiation experiments in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The first of these irradiation tests, designated AGR-1, began in ATR in December of 2006 and ended in November 2009. This experiment was primarily to act as a shakedown test of the multi-capsule test train design and to provide early data on fuel performance that would be used in fuel fabrication process development. AGR-1 fuel kernels were produced on an engineering scale, but the TRISO coatings and cylindrical fuel compacts were fabricated on a laboratory scale. The AGR-1 PIE was completed and the final report was published in 2015. The second irradiation test, AGR-2, started in ATR in June 2010 and ended in October 2013. The AGR-2 irradiation test was designed to provide fuel performance data for coated particles fabricated on an engineering-scale pilot line using a coater with an internal chamber diameter of 150 mm (6 in.). The final PIE report was published in 2021. AGR-3/4, a single irradiation that combined what were originally conceived as the third and fourth tests, was to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport. PIE of the AGR-3/4 experiment is still in progress as of this writing. The subject of this report is AGR-5/6/7, the final qualification test of AGR TRISO fuel made entirely at the engineering scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Status Report on FY2022 Model Development within the Integrated Energy Systems HYBRID Repository

This publication details newly created energy storage models developed within the HYBRID Modelica repository as part of the Department of Energy Office of Nuclear Energy (DOE-NE) Integrated Energy Systems (IES) program, led by Idaho National Laboratory (INL). Model development to-date includes creation of dynamic systems-level models of a pebble bed high temperature gas reactor (HTGR), liquid air energy storage (LAES), and compressed air energy storage (CAES) in the IES-based HYBRID repository. Models are developed using the latest publicly available data and incorporate the possibility of control strategy inclusion for use with the existing IES modeling, analysis, and optimization toolset. Simulations showcase the abilities of each technology to flexibly operate in ways consistent with IES operation expectations. When these models are available, they can be utilized within different integrated energy park concepts to understand optimal system operation, control, and dispatching. Moreover, given the generic nature of the models, industrial partner technologies can be quickly added to the repository using the existing models as a basis. Additional dynamic models for thermal energy storage concepts can be developed and added to the HYBRID repository as needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Confirmatory Defect Analyses of AGR-5/6/7 Fuel

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program has executed a series of experiments to test and qualify tristructural isotropic (TRISO) particle fuels for use in high-temperature gas-cooled reactors. The most recent irradiation experiments, AGR-5/6/7, are testing fuel that was fabricated in near production-scale equipment. There are three important defect fractions that must be characterized, using a deconsolidation-leach-burn-leach (DLBL) method, to better predict and understand fuel performance during irradiation. These are the dispersed uranium fraction (DUF), exposed kernel fraction (EKF), and silicon carbide defect fraction (SDF). The DLBL characterization data from analyses of the four compact batches, fabricated and characterized by the BWX Technologies Nuclear Operations Group (BWXT-NOG), showed unexpectedly higher values for all three defect fractions relative to the parent TRISO particle defect fractions. The DUF and SDF should not have changed significantly during compact fabrication and the apparent increase in the EKF, a metric for particle damage, was higher than anticipated for the compacts with 40% volumetric packing fraction (PF). Analytical results showed no net increase in the EKF for the 25% PF compacts. Compacts from one compact batch of each packing fraction were sent to Oak Ridge National Laboratory (ORNL), along with uncompacted tristructural isotropic (TRISO) particles, overcoated to a thickness targeting a 40% PF, for confirmatory analyses of the defect fractions and determination of the source(s) of damage. The ORNL results did not confirm the BWXT-NOG results for either the DUF or the SDF; the ORNL results showed no significant change in these defect fractions relative to the parent TRISO particle lot. The ORNL results did, however, confirm that a large increase in the exposed kernel fraction was found in the 40% PF compacts and that the 25% PF compacts sustained little if any damage. Data suggest that a large majority of the mechanical damage to the TRISO particles occurred in the overcoater when targeting the 40% PF. The analytical laboratories at both BWXT-NOG and ORNL had inconsistent outcomes on analyses of individual batches. Although the ORNL data is more consistent with the parent TRISO lot and expectations, the data set is too limited in size to justifiably reject suspect data from either laboratory at this time. It is recommended that an additional analysis be done to provide sufficient data for estimating the true defect fractions for interpreting the radionuclide release data during the AGR-5/6/7 irradiation experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Post-irradiation 1600°C Heating Test of AGR-1 UCO Fuel Kernels

The Advanced Gas Reactor Fuel Development and Qualification (AGR) Program was established to perform research and development on tristructural isotropic (TRISO) coated-particle fuel to support deployment of high temperature gas-cooled reactors (HTGRs) . This talk will discuss the results from a post-irradiation heating test of five individual fuel kernels extracted from four TRISO fuel particles that were deconsolidated from an irradiated AGR-1 fuel compact.

heating test↗

DOE OSTI2014

Comparison of Fission Product Release Predictions Using PARFUME with Results from the AGR-1 Irradiation Experiment

This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive layers of the TRISO-coated particle and potentially through the compact matrix. The release of these fission products was measured in PIE and modeled with PARFUME.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

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

42 ENGINEERING↗

DOE OSTI2020

AGR-2 Compact 6-4-1 Post-Irradiation Examination Results

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite 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 program goal 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 testing, fuel performance modeling, and fission product transport (INL 2015). A series of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗