Binh T. Pham

Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-2 Experiment in the Advanced Test Reactor

A temperature sensitivity evaluation has been performed for an individual test capsule in the AGR-2 TRISO particle fuel experiment. The AGR-2 experiment is the second in a series of fueled test experiments for TRISO coated fuel particles run in the Advanced Test Reactor at the Idaho National Laboratory. A series of cases were compared to a base case by varying different input parameters in an ABAQUS finite element thermal model. Most input parameters were varied by ±10%, with one parameter ±20%, to show the temperature sensitivity to each parameter. The most sensitive parameters were the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. The thermal conductivity of the fuel compacts and thermal conductivity of the graphite holder were of moderate sensitivity. The least sensitive parameters were the emissivities of the stainless steel and graphite, along with gamma heat rate in the non-fueled components. Sensitivity calculations were also performed for the fast neutron fluence, which showed a general, but minimal, temperature rise with increasing fluence.Copyright

Thermal Predictions of the AGR-2 Experiment with Variable Gas Gaps

Abstract A new daily as-run thermal analysis was performed at the Idaho National Laboratory for the advanced gas cooled reactor (AGR) test experiment number two (AGR-2) in the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment due to graphite shrinkage resulting from neutron damage. The purpose of this analysis was to calculate the daily average temperatures of each TRISO (tristructural isotropic)–particle fuel compact. A steady-state thermal analysis was performed daily for each capsule with the commercial finite element heat transfer code ABAQUS. These new thermal predictions show the compact fuel temperature dependence on the variable gas gap method. Comparison between measured and calculated temperatures is discussed.

Diagnostic and Prognostic Models for Generator Step-Up Transformers

................................................................................................................................................. v SUMMARY ................................................................................................................................................ vii ACKNOWLEDGMENTS ........................................................................................................................... ix ACRONYMS ............................................................................................................................................. xiii

UNCERTAINTY QUANTIFICATION OF CALCULATED TEMPERATURES FOR ADVANCED GAS REACTOR FUEL IRRADIATION EXPERIMENTS

Abstract This paper presents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) fuel irradiation experiments conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technologies Fuel Development and Qualification Program. The predicted temperatures with associated uncertainty for AGR tests using the ABAQUS finite element heat transfer code are used to validate the fission product transport and fuel performance simulation models. To quantify the uncertainty of calculated temperatures, this study identifies and analyzes model parameters of potential importance to the predicted fuel temperatures. The selection of input parameters for uncertainty quantification is based on the ranking of their influence on the variation of temperature predictions. Thus, selected input parameters include those with high sensitivity and those with large uncertainty. The propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of the calculated temperatures. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using an experimental design, an analysis of pairwise interactions of model parameters was performed to establish the sufficiency of the first-order (linear) expansion terms in constructing the response surface. To achieve completeness, the uncertainty propagation made use of pairwise noise correlations of model parameters. The AGR-2 overall fuel temperature uncertainties reported here are less than 5% (or 60°C).

Sensitivity Evaluation of the AGR 3-4 Experiment Thermal Model Irradiated in the Advanced Test Reactor

A temperature sensitivity evaluation has been performed on a thermal model for the AGR-3/4 fuel experiment on an individual capsule. The experiment was irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Four TRISO fuel irradiation experiments are planned for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program which supports the development of the Very High Temperature Gas-cooled Reactor under the Next-Generation Nuclear Plant project.AGR-3/4 is the third TRISO-particle fuel test of the four planned and is intended to test tri-structural-isotropic (TRISO)-coated, low-enriched uranium oxy-carbide fuel. The AGR-3/4 test was specifically designed to assess fission product transport through various graphite materials. The AGR-3/4 irradiation test in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) TRISO-particle fueled compacts were inserted into 12 separate capsules for the experiment (four compacts per capsule).The purpose of this analysis was to assess the sensitivity of input variables for the capsule thermal model. A series of cases were compared to a base case by varying different input parameters into the ABAQUS finite element thermal model. These input parameters were varied by ±10% to show the temperature sensitivity to each parameter. The most sensitive parameter was the compact heat rates, followed by the outer control gap distance and neon gas fraction. Thermal conductivity of the compacts and thermal conductivity of the various graphite layers vary with fast neutron fluence and exhibited moderate sensitivity. The least sensitive parameters were the emissivities of the stainless steel and graphite, along with gamma heat rate in the non-fueled components. Separate sensitivity calculations were performed varying with fast neutron fluence, showing a general temperature rise with an increase in fast neutron fluence. This is a result of the control gas gap becoming larger due to the graphite shrinkage with neutron damage. A smaller sensitivity is due to the thermal conductivity of the fuel compacts with fast neutron fluence.Heat rates and fast neutron fluence were input from a detailed physics analysis using the Monte Carlo N-Particle (MCNP) code. Individual heat rates for each non-fuel component were input as well. A steady-state thermal analysis was performed for each sensitivity calculation. ATR outer shim control cylinders and neck shim rods along with driver fuel power and fuel depletion were incorporated into the physics heat rate calculations. Surface-to-surface radiation heat transfer along with conduction heat transfer through the gas mixture of helium-neon (used for temperature control) was used in the sensitivity calculations.Copyright

Thermal Predictions of the AGR-3/4 Experiment Using PIE-Measured Time Varying Gas Gaps

A thermal analysis was performed for the Advanced Gas Reactor test experiment (AGR-3/4) with Post Irradiation Examination (PIE) measured time varying gas gaps. The experiment was irradiated at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Several fuel irradiation experiments are planned for the AGR Fuel Development and Qualification Program which supports the development of the Very-High-Temperature gas-cooled Reactor (VHTR) under the Next-Generation Nuclear Plant (NGNP) project.AGR-3/4 combines two tests in a series of planned AGR experiments to test tri-structural-isotropic (TRISO)-coated, low-enriched uranium oxy-carbide fuel. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. The AGR-3/4 test irradiation in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) TRISO fueled compacts were inserted into twelve separate capsules for the experiment (four compacts per capsule).The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures using PIE-measured time (fast neutron fluence) varying gas gaps and to compare with experimentally measured thermocouple data. PIE-measured experimental data was used for the graphite shrinkage versus fast neutron fluence. Heat rates were input from a detailed physics analysis using the Monte Carlo N-Particle (MCNP) code for each day during the experiment. Individual heat rates for each non-fuel component were input as well. A steady-state thermal analysis was performed for each daily calculation. A finite element model was created for each capsule using the commercial finite element heat transfer and stress analysis package ABAQUS. The fission and neutron gamma heat rates were calculated with the nuclear physics code MCNP. ATR outer shim control cylinders and neck shim rods along with driver fuel power and fuel depletion were incorporated into the daily physics heat rate calculations. Compact and graphite thermal conductivity were input as a function of temperature and fast neutron fluence with the field variable option in ABAQUS. Surface-to-surface radiation heat transfer along with conduction heat transfer through the gas mixture of helium-neon (used for temperature control) was used in these models. Model results are compared to thermocouple data taken during the experiment.Copyright

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.

Online Monitoring Technical Basis and Analysis Framework for Emergency Diesel Generators - Interim Report for FY 2013

................................................................................................................................................ iii EXECUTIVE SUMMARY .......................................................................................................................... v ACRONYMS ............................................................................................................................................... ix

Data Collection Plan to Populate the Light Water Reactor Sustainability Failure Mode Degradation Library

Interest in implementing advanced Prognostic Health Management (PHM) systems in commercial nuclear power plants (NPPs) has increased rapidly in recent years, with an overarching goal of implementing of improving the safety, reliability, and economics/profitability of the aging nuclear fleet and extending their service life in the most cost-effective manner. The PHM system utilizes prognostic tools to estimate the remaining useful life (RUL) of a component or system of components based on current and predicted operating conditions. An effective implementation of the PHM system will anticipate and identify unique age-dependent degradation modes to provide early warning of emerging problems. Selection of the components and structures to be monitored is a crucial step for successful PHM implementation in NPPs. A selection framework is recommended for risk significant components (both safety-related and non-safety related) based on the Fussell-Vesely (F-V) Importance Measure and the Risk Achievement Worth (RAW) measure. For the selected components, a failure mode degradation library will be developed consisting of data corresponding to different failure/degradation modes. In lieu of constructing an expensive scaled test facility, several data sources are identified for populating the failure mode degradation library, including various national laboratories, universities, agencies, and industries.

Communication Pathways in the Light Water Reactor Sustainability Online Monitoring Project

Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Large amounts of operational data, including failure data, are necessary for the development and calibration of diagnostic and prognostic algorithms. The ability to use data from all available resources will provide the most expeditious avenue to implementation of online monitoring in existing NPPs; however, operational plant data are often considered proprietary. Secure methods for transferring and storing data are discussed, along with a potential technology for implementation of online monitoring.