Thermal Analysis Safety Margins Using ABAQUS for the MP-2 Experiment in the Advanced Test Reactor

Author(s):  
Grant L. Hawkes ◽  
Douglas S. Crawford ◽  
Gregory K. Housley

The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR). MP-2 is considered a non-instrumented drop-in test where small aluminum-clad fuel plate samples are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates were analyzed. A thermal analysis has been performed on the MP-2 experiment to be irradiated in the ATR at Idaho National Laboratory (INL). A new technique for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) using the commercial finite element and heat transfer code ABAQUS is demonstrated. This new technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node. Pressure drop data is fed into the calculations in order to geometrically calculate the water saturation temperature. Results from the DNBR and FIR calculations are displayed with the ABAQUS post processor named Viewer. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.

Author(s):  
Grant L. Hawkes

Abstract The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10-day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculations that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated. Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.


Author(s):  
Grant L. Hawkes ◽  
Nicolas E. Woolstenhulme

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.


Author(s):  
Hilda B. Klasky ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
Sarma B. Gorti ◽  
Randy K. Nanstad ◽  
...  

The Oak Ridge National Laboratory (ORNL) performed a detailed technical review of the 2015 Electrabel (EBL) Safety Cases prepared for the Belgium reactor pressure vessels (RPVs) at Doel 3 and Tihange 2 (D3/T2). The Federal Agency for Nuclear Control (FANC) in Belgium commissioned ORNL to provide a thorough assessment of the existing safety margins against cracking of the RPVs due to the presence of almost laminar flaws found in each RPV. Initial efforts focused on surveying relevant literature that provided necessary background knowledge on the issues related to the quasi-laminar flaws observed in D3/T2 reactors. Next, ORNL proceeded to develop an independent quantitative assessment of the entire flaw population in the two Belgian reactors according to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Appendix G, “Fracture Toughness Criteria for Protection Against Failure,” New York (both 1992 and 2004 versions). That screening assessment of the EBL-characterized flaws in D3/T2 used ORNL tools, methodologies, and the ASME Code Case N-848, “Alternative Characterization Rules for Quasi-Laminar Flaws”. Results and conclusions derived from comparisons of the ORNL flaw acceptance assessments of D3/T2 with those from the 2015 EBL Safety Cases are presented in the paper. The ORNL screening analyses identified fewer flaws than EBL that were not compliant with the ASME Section XI (1992) criterion; the EBL criterion imposed additional conservatisms not included in ASME Section XI. Furthermore, ORNL’s application of the updated ASME Section XI (2004) criterion produced only four non-compliant flaws, all due to design-basis loss-of-coolant loading transients. Among the latter, only one flaw remained non-compliant when analyzed using the warm-prestress (WPS) cleavage fracture model typically applied in USA flaw assessments. ORNL’s independent refined analysis of that flaw (#1660, which was also non-compliant in the EBL screening assessments) rendered it compliant when modeled as a more realistic individual quasi-laminar flaw using a 3-dimensional XFEM (eXtended Finite Element Method) approach available in the ABAQUS© finite element code. Taken as a whole, the ORNL-specific results and conclusions confirmed the structural integrity of Doel 3 and Tihange 2 under all design transients with ample margin in the presence of the 16,196 detected flaws.


Author(s):  
Edward Shitsi ◽  
Prince Amoah ◽  
Emmanuel Ampomah-Amoako ◽  
Henry Cecil Odoi

Abstract Research reactors all over the world are expected to operate within certain safety margins just like pressurized water reactors and boiling water reactors. These safety margins mainly include onset of nucleate boiling ratio (ONBR), departure from nucleate boiling ratio (DNBR), and flow instability ratio (FIR) in addition to the maximum clad or fuel temperature and saturation temperature or boing point of the coolant inside the core of the reactor. This study carried out steady-state safety analysis of the Ghana Research Reactor-1 (GHARR-1) with low enriched uranium (LEU) core. Monte Carlo N-particle (MCNP) code was used to obtain radial and axial power peaking factors used as inputs in the preparation of the input file of plate temperature code of Argonne National Laboratory (PLTEMP/ANL code), which was then used to obtain the mentioned safety parameters of GHARR-1 with LEU core in this study. The data obtained on the ONBR were used to obtain the initiation of nucleate boiling boundary data with respect to the active length of the reactor core for various reactor powers. The obtained results for LEU core were also compared with that of the high enriched uranium (HEU) core. The results obtained show that the 34 kW GHARR-1 with LEU core is safe to operate just as the previous 30 kW HEU core was safe to operate.


2014 ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki

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.


Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki

A thermal analysis was performed for the Advanced Gas Reactor test experiment number three/four (AGR-3/4) 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 tristructural-isotropic (TRISO)-coated, low-enriched uranium oxycarbide fuel. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. The AGR-3/4 test was inserted in the ATR beginning in 2011 and is currently still in the reactor. Forty-eight (48) TRISO fueled compacts were inserted into twelve separate capsules for the experiment. The purpose of this analysis was to calculate the temperatures of each compact and graphite layer to obtain daily average temperatures and to compare with experimentally measured thermocouple data. Heat rates were input from a detailed physics analysis using the 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 this analysis 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. The kinetic theory of gases was used to correlate the thermal conductivity of the gas mixture. Model results are compared to thermocouple data taken during the experiment. Future thermal analysis models will consider control temperature gas gaps and fuel compact–graphite holder gas gaps varying from the original fabrication dimensions as a function of fast neutron fluence.


Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki ◽  
Binh T. Pham

A thermal analysis was performed for the advanced gas reactor test experiment (AGR-3/4) with post irradiation examination (PIE) measured time (fast neutron fluence) 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 under the advanced reactor technologies project. The AGR-3/4 test was designed primarily to assess fission product transport through various graphite materials. Irradiation in the ATR started in December 2011 and finished in April 2014. Forty-eight (48) tristructural-isotropic-fueled compacts were inserted into 12 separate capsules for the experiment. 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 compare with experimentally measured thermocouple (TC) data. PIE-measured experimental data were used for the graphite shrinkage versus fast neutron fluence. PIE dimensional measurements were taken on all the fuel compacts, graphite holders, and all of the graphite rings used. Heat rates were input from a detailed physics analysis for each day during the experiment. Individual heat rates for each nonfuel 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.


2016 ◽  
Author(s):  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
John T. Maki ◽  
Binh T. Pham

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.


1980 ◽  
Vol 53 (3) ◽  
pp. 437-511 ◽  
Author(s):  
D. W. Brazier

Abstract An attempt has been made to review the development of thermoanalytical procedures as they have been applied to elastomers and elastomer systems over the past 10 years. For all rubber industry products, temperature and its effects, either alone or in conjunction with the chemical environment, play an important role from the production stage through to the final failure of the product in the field. It is thus not surprising that thermal analysis, in which temperature is the prime variable, has found such diverse applications in elastomer studies. The identification and quantitative analysis of rubber formulations have received most attention. Such formulations produce characteristic “fingerprints” when studied in DTA, DSC, TG, or TMA. In DSC, the determination of the glass transition characteristics, the observation and determination of crystallinity, the detection of cyclization reactions, and the monitoring of thermal and oxidative degradation characteristics can all be observed in a single experiment covering the temperature range from −150 to +600°C. At normal heating rates, e.g., 20°C/min, such information is available in 40 min. TG/DTG analysis can yield the elastomer or elastomers content, oil and plasticizer, carbon black (level and often type), and inorganic ash in less than 60 min. Processing and curing can also be studied. Blend compatibility can be assessed on the basis of both Tg and crystallinity measurements and the data used to determine optimum mixing times. Sulfur vulcanization and peroxide curing of elastomers is readily monitored by DSC and can be used for confirmation analysis of the presence of curatives. Limitations in such analysis exist, but as understanding and ability to interpret cure exotherms increase, valuable information about the mechanism and the nature of the cured network will be obtained. The testing of rubber compounds involves many hours of labor by current procedures. The rapidity of thermal analysis promises to offer some relief. In addition to DSC and TG, TMA, a relatively new technique, offers a rapid approach to low-temperature testing. Dynamic mechanical analysis (DMA) offers a rapid route to determining dynamic properties, but as yet, relatively little has been published on the application of this new technique to elastomers. As environmental concern increases, techniques such as evolved gas analysis (EGA) and combined techniques such as TG/gas chromatography are predicted to play an important role. As for the future, it is readily apparent that the principles of the methods have been established and, in several cases, it now remains to reduce them to a practical level. In some areas, such as vulcanization studies, much remains to be undertaken to improve our interpretive skills. Although there is some indication that certain industries have produced “in-house” standards for the analysis of rubber compounds by DSC and TG/DTG, it will only be when national and international standards organizations study and produce standard procedures, that the techniques will be generally adopted. Maurer's prediction in 1969 of increased applications of DTA and TG in elastomer studies has undoubtedly proved correct, and with the proliferation of reliable commercial instrumentation, significant developments can be anticipated in the next decade.


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