scholarly journals Thermal Predictions of the AGR-3/4 Experiment With Time-Varying Gas Gaps

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki
Keyword(s):  
National Laboratory ◽  
Transfer Model ◽  
Time Varying ◽  
Commercial Code ◽  
Reactor Test ◽  
Enriched Uranium ◽  
Test Reactor ◽  
Abaqus Model ◽  
Very High ◽  
Carbide Fuel

A thermal analysis was performed for the Advanced Gas Reactor test experiment (AGR-3/4) with 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 that 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 tristructural-isotropic (TRISO)-coated, low-enriched uranium oxy-carbide fuel. Forty-eight TRISO-fueled compacts were inserted into 12 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 time (fast neutron fluence)-varying gas gaps and to compare with experimentally measured thermocouple data. A finite-element heat transfer model was created for each capsule using the commercial code ABAQUS. Model results are compared to thermocouple data taken during the experiment.

scholarly journals Coupled Thermal-Hydraulic Analysis and Species Mass Transport in a Versatile Experimental Salt Irradiation Loop (VESIL) Using CTF

2021 ◽  
Vol 2 (3) ◽  
pp. 309-317
Author(s):  
Samuel A. Walker ◽  
Abdalla Abou-Jaoude ◽  
Zack Taylor ◽  
Robert K. Salko ◽  
Wei Ji
Keyword(s):  
Mass Transfer ◽  
Fission Product ◽  
Noble Metals ◽  
National Laboratory ◽  
Species Interaction ◽  
Transfer Model ◽  
Two Fluids ◽  
Gas System ◽  
Fuel Salt ◽  
Test Reactor

With the resurgence of interest in molten salt reactors, there is a need for new experiments and modeling capabilities to characterize the unique phenomena present in this fluid fuel system. A Versatile Experimental Salt Irradiation Loop (VESIL) is currently under investigation at Idaho National Laboratory to be placed in the Advanced Test Reactor (ATR). One of the key phenomena this proposed experiment plans to elucidate is fission product speciation in the fuel-salt and the subsequent effects this has on the fuel-salt properties, source term generation, and corrosion control. Specifically, noble gases (Xe & Kr) will bubble out to a plenum or off-gas system, and noble metals (Mo, Tc, Te, etc.) will precipitate and deposit in specific zones in the loop. This work extends the mass transfer and species interaction models in CTF (Coolant-Boiling in Rod Arrays—Two Fluids) and applies these models to give a preliminary estimation of fission product behavior in the proposed VESIL design. A noble metal–helium bubble mass transfer model is coupled with the thermal-hydraulic results from CTF to determine the effectiveness of this insoluble fission product (IFP) extraction method for VESIL. Amounts of IFP species extracted to the off-gas system and species distributions in VESIL after a 60-day ATR cycle are reported.

Thermal Analysis of the FSP-1 Experiment in the Advanced Test Reactor

2016 ◽  
Author(s):  
Grant L. Hawkes ◽  
Nicolas E. Woolstenhulme
Keyword(s):  
Thermal Analysis ◽  
Flow Instability ◽  
National Laboratory ◽  
Safety Evaluation ◽  
Thermal Safety ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Safety Margins ◽  
Enriched Uranium ◽  
Test Reactor

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Three different types of fuel plates with matching pairs for a total of six plates were analyzed. These three types of plates are: full burn, intermediate power, and thick meat. A thermal analysis has been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A thermal safety evaluation was performed to demonstrate that the FSP-1 irradiation experiment complies with the thermal-hydraulic safety requirements of the ATR Safety Analysis Report (SAR). The ATR SAR requires that minimum safety margins to critical heat flux and flow instability be met in the case of a loss of commercial power with primary coolant pump coast-down to emergency flow. The thermal safety evaluation was performed at 26 MW NE lobe power to encompass the expected range of operating power during a standard cycle. Additional safety evaluations of reactivity insertion events, loss of coolant event, and free convection cooling in the reactor and in the canal are used to determine the response of the experiment to these events and conditions. This paper reports and shows that each safety evaluation complies with each safety requirement of the ATR SAR.

scholarly journals Thermal Predictions of the AGR-3/4 Experiment With Time Varying Gas Gaps

2014 ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki
Keyword(s):  
Heat Transfer ◽  
Thermal Analysis ◽  
Finite Element ◽  
Fast Neutron ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Surface Radiation ◽  
Time Varying ◽  
Fuel Component ◽  
Enriched Uranium

A thermal analysis was performed for the Advanced Gas Reactor test experiment (AGR-3/4) with 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 time (fast neutron fluence) varying gas gaps and to compare with experimentally measured thermocouple data. Previous 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 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.

scholarly journals Flow Testing and Analysis of the FSP-1 Experiment

2017 ◽  
Author(s):  
Grant L. Hawkes ◽  
Warren F. Jones ◽  
Wade Marcum ◽  
Aaron Weiss ◽  
Trevor Howard
Keyword(s):  
Flow Rate ◽  
High Performance ◽  
National Laboratory ◽  
Water Channel ◽  
Flow Characteristics ◽  
Separate Branch ◽  
Flow Test ◽  
Enriched Uranium ◽  
Test Reactor ◽  
Fuel Meat

The U.S. High Performance Research Reactor conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Size Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water channel velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A pressure differential versus flow rate curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported.

Progressive Evolution of the Core of the Fast Breeder Test Reactor

2010 ◽  
Author(s):  
S. Varatharajan ◽  
K. V. Sureshkumar ◽  
K. V. Kasiviswanathan ◽  
G. Srinivasan
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Power Reactor ◽  
Test Bed ◽  
Fast Breeder Test Reactor ◽  
The Core ◽  
Linear Heat ◽  
Mox Fuel ◽  
Test Reactor ◽  
Carbide Fuel

The second stage of Indian nuclear programme envisages the deployment of fast reactors on a large scale for the effective use of India’s limited uranium reserves. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is a loop type, sodium cooled fast reactor, meant as a test bed for the fuels and structural materials for the Indian fast reactor programme. The reactor was made critical with a unique high plutonium MK-I carbide fuel (70% PuC+30%UC). Being a unique untested fuel of its kind, it was decided to test it as a driver fuel, with conservative limits on Linear Heat Rating and burn-up, based on out-of-pile studies. FBTR went critical in Oct 1985 with a small core of 23 MK-I fuel subassemblies. The Linear Heat Rating and burn-up limits for the fuel were conservatively set at 250 W/cm & 25 GWd/t respectively. Based on out-of-pile simulation in 1994, it was possible to raise the LHR to 320 W/cm. It was decided that when the fuel reaches the target burn-up of 25 GWd/t, the MK-I core would be progressively replaced with a larger core of MK-II carbide fuel (55% PuC+45%UC). Induction of MK-II subassemblies was started in 1996. However, based on the Post-Irradiation Examination (PIE) of the MK-I fuel at 25, 50 & 100 GWd/t, it became possible to enhance the burn-up of the MK-I fuel to 155 GWd/t. More than 900 fuel pins of MK-I composition have reached 155 GWd/t without even a single failure and have been discharged. One subassembly (61 pins) was taken to 165 GWd/t on trial basis, without any clad failure. The core has been progressively enlarged, adding MK-I subassemblies to compensate for the burn-up loss of reactivity and replacement of discharged subassemblies. The induction of MK-II fuel was stopped in 2003. One test subassembly simulating the composition of the MOX fuel (29% PuO2) to be used in the 500 MWe Prototype Fast Breeder Reactor was loaded in 2003. It is undergoing irradiation at 450 W/cm, and has successfully seen a burn-up of 92.5 GWd/t. In 2006, it was proposed to test high Pu MOX fuel (44% PuO2), in order to validate the fabrication and fuel cycle processes developed for the power reactor MOX fuel. Eight MOX subassemblies were loaded in FBTR core in 2007. The current core has 27 MK-I, 13 MK-II, eight high Pu MOX and one power reactor MOX fuel subassemblies. The reactor power has been progressively increased from 10.5 MWt to 18.6 MWt, due to the progressive enlargement of the core. This paper presents the evolution of the core based on the progressive enhancement of the burn-up limit of the unique high Pu carbide fuel.

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