As-Built Simulation of the High Flux Isotope Reactor

The Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) is an 85 MWt flux trap-type research reactor that supports key research missions, including isotope production, materials irradiation, and neutron scattering. The core consists of an inner and an outer fuel element containing 171 and 369 involute-shaped plates, respectively. The thin fuel plates consist of a U3O8-Al dispersion fuel (highly enriched), an aluminum-based filler, and aluminum cladding. The fuel meat thickness is varied across the width of the involute plate to reduce thermal flux peaks at the radial edges of the fuel elements. Some deviation from the designed fuel meat shaping is allowed during manufacturing. A homogeneity scan of each fuel plate checks for potential anomalies in the fuel distribution by scanning the surface of the plate and comparing the attenuation of the beam to calibration standards. While typical HFIR simulations use homogenized fuel regions, explicit models of the plates were developed under the Low-Enriched Uranium Conversion Program. These explicit models typically include one inner and one outer fuel plate with nominal fuel distributions, and then the plates are duplicated to fill the space of the corresponding fuel element. Therefore, data extracted from these simulations are limited to azimuthally averaged quantities. To determine the reactivity and physics impacts of an as-built outer fuel element and generate azimuthally dependent data in the element, 369 unique fuel plate models were generated and positioned. This model generates the three-dimensional (i.e., radial–axial–azimuthal) plate power profile, where the azimuthal profile is impacted by features within the adjacent control element region and beryllium reflector. For an as-built model of the outer fuel element, plate-specific homogeneity data, 235U loading, enrichment, and channel thickness measurements were translated into the model, yielding a much more varied azimuthal power profile encompassed by uncertainty factors in analyses. These models were run with the ORNL-TN and Shift Monte Carlo tools, and they contained upwards of 500,000 cells and 100,000 unique tallies.

TRANSIENT ANALYSIS OF SIMULTANEOUS LOFA AND RIA IN RSG-GAS REACTOR AFTER 32 YEARS OPERATION

During the operation of the research reactor RSG-GAS, there are many design parameters should be verified based on postulated accidents. Several design basis accidents (DBA) such as loss of flow accident (LOFA) and reactivity-initiated accident (RIA) also have been conducted separately. This paper discusses about possibility of simultaneous accidents of LOFA and RIA. The accident analyses carry out calculation for transient condition during RIA, LOFA, and postulated accident of simultaneous LOFA-RIA. This study aims to conduct a safety analysis on simultaneous LOFA and RIA, and investigate the impact on safety margins. The calculations are conducted by using the PARET code. The maximum temperature of the center fuel meat at nominal power of 30 MW and steady state conditions is 126.10°C and MDNBR of 2.94. At transients condition, the maximum center fuel meat temperature for LOFA, RIA and simultaneous LOFA-RIA are consecutively 132.99°C, 135.67°C and 138.21°C, and the time of reactor trip are 3.2593s, 3.6494s and 2.7118s, respectively. While the MDNBR for LOFA, RIA and simultaneous LOFA-RIA are respectively at transient condition are 2.88, 2.58 and 2.63, respectively. It is shown that, simultaneous LOFA-RIA has the fastest trip time. In this case, the low flow trip occurs first in advance to over power trip. From these results, it can be concluded that the RSG-GAS has adequate safety margin against transient of simultaneous LOFA-RIA.Keywords: RSG-GAS, Simultaneous, LOFA, RIA, PARET

SENSITIVITY ANALYSIS ON THERMOHYDRAULIC CODE FOR MODIFIED PLATE-FUELED 2 MW TRIGA

SENSITIVITY ANALYSIS OF THERMOHYDRAULIC CODE FOR MODIFIED PLATE-FUELED 2 MW TRIGA. The plan to modify TRIGA 2000 Bandung from using regular TRIGA fuel to plate-type fuel should be supported by the use of appropriate computer codes. This research proposes three codes to design reactor thermohydraulics at transient condition. Analysis has been performed to identify code sensitivity using the same input and correlation. The codes used were COOLOD-N2, Heathyd, and PARET-ANL. The input was obtained from preliminary analysis of a flow rate calculation of 70 kg/s and a nominal power of 2 MW. The comparison of these three codes did not consider uncertainty factor for neutronic and technical aspects. The sensitivity analysis on thermohydraulic codes used to calculate heat transfer in the fuel plate of TRIGA reactor at steady state condition indicates similar temperature trend lines for the coolant, plate, and fuel meat. Temperature calculation results obtained from COOLOD-N2, Heathyd and PARET ANL give consistent sensitivity with the differences of coolant temperature from 2.83% to 12.5%; cladding temperature from 2.14% to 31.30%; and fuel meat temperature from 6.63% to 18.64%. The margins of flow instability were 5.03; 5.68 and 4.21, respectively for COOLOD-N2, Heathyd, and PARET-ANL. These values show that flow instability has not yet occurred. The results of the analysis show that the use of those three codes for steady state condition using the same input, in which uncertainty factor is neglected, give similar trend for coolant, cladding, and fuel meat temperature. As the modelling in each code is different, the values obtained are not exactly the same.Keywords: sensitivity analysis, TRIGA Plate, COOLOD-N2, Heathyd, PARET-ANL

Thermo-mechanical behavior simulation coupled with the hydrostatic-pressure-dependent grain-scale fission gas swelling calculation for a monolithic UMo fuel plate under heterogeneous neutron irradiation

Monolithic UMo fuels have a higher uranium density than previously developed fuels. They have become the most promising fuels to be used in high-flux research and test reactors after the US Office of Material Management and Minimization Reactor Conversion Program (formerly Reduced Enrichment Research and Test Reactor (RERTR) program). In this study, a computational method is established to couple the macro-scale irradiation-induced thermo-mechanical behavior simulation with the hydrostatic-pressure-dependent fission gas swelling calculation in the UMo grain. The stress update algorithms and consistent stiffness moduli are respectively presented for UMo fuel, in which both the hydrostatic-pressure-dependent irradiation swelling and deviatoric-stress-directed irradiation creep are taken into account. Accordingly, the user subroutines to define the thermo-mechanical non-homogeneous constitutive relations for the UMo fuel meat and Al cladding are developed and validated. The in-pile behavior in a monolithic UMo fuel plate under a location-dependent irradiation condition is calculated and discussed.

Effect of Fuel Meat Thickness on the Non-Uniform Irradiation-Induced Thermo-Mechanical Behavior in Monolithic UMo/Al Fuel Plates

Monolithic UMo/Al fuel plates have a promising prospect in the advanced research and test reactors because of their high equivalent uranium density and stable irradiation performance. They will undergo complicated in-pile thermo-mechanical behavior, which may affect their lifetime and the safety of nuclear reactors. It is necessary to capture the effect of fuel meat thickness on in-pile thermo-mechanical behavior evolution in the fuel plates in order to realize their optimized design and control their service safety. In this study, considering a non-uniform irradiation condition, several 3D finite element models are built to simulate the in-pile behavior in different-thickness UMo/Al plates. The user subroutines are programmed based on the thermo-mechanical constitutive relations and stress update algorithms of the constituent materials. The influences of fuel meat thickness on the temperature field, the main deformations and the interfacial normal stresses are numerically investigated based on the obtained results.

Flow Testing and Analysis of the FSP-1 Experiment

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.

INTERAKSI BAHAN BAKAR U3Si2-Al DENGAN KELONGSONG AlMg2 PADA ELEMEN BAKAR SILISIDA TMU 2,96 gU/cm3 PASCA IRADIASI

INTERAKSI BAHAN BAKAR U3Si2-Al DENGAN KELONGSONG AlMg2 PADA ELEMEN BAKAR SILISIDA TMU 2,96 gU/cm3 PASCA IRADIASI. Telah dilakukan analisis interaksi bahan bakar U3Si2-Al dengan kelongsong AlMg2 pada pelat elemen bakar (PEB) U3Si2-Al tingkat muat uranium (TMU) 2,96 gU/cm3 pasca iradiasi. Penelitian ini bertujuan untuk mengetahui pengaruh radiasi terhadap perubahan mikrostruktur PEB selama di reaktor. Untuk mengetahui pengaruh radiasi terhadap mikrostruktur PEB U3Si2-Al perlu dipahami interaksi kelongsong AlMg2 dengan inti elemen bakar U3Si2-Al pra maupun pasca iradiasi. Pengujian pra iradiasi dilakukan pemanasan PEB U3Si2-Al TMU 2,96 gU/cm3 dengan ukuran 10x10 mm di dalam tungku DTA (Differential Thermal Analysis) dengan variasi temperatur 450, 550, 650, 900 dan 1350oC. PEB U3Si2-Al TMU 2,96 gU/cm3 pasca iradiasi dilakukan pemotongan di dalam hotcell dengan ukuran 2x10 mm sebanyak 3 (tiga) sampel bagian bottom, middle dan top PEB. Potongan PEB U3Si2-Al TMU 2,96 gU/cm3 pra maupun pasca iradiasi dikenakan preparasi metalografi meliputi mounting, grinda, poles, dan etsa. Pengamatan mikrostruktur interaksi bahan bakar U3Si2 dengan kelongsong AlMg2 dalam PEB U3Si2-Al pra iradiasi dilakukan menggunakan Scanning Electron Microscope (SEM-EDS), sedangkan pengamatan mikrostruktur PEB U3Si2-Al pasca iradiasi dilakukan menggunakan mikroskop optik di dalam hotcell. Hasil interaksi U3Si2dengan matrik Al maupun kelongsong AlMg2 pada PEB U3Si2-Al pra iradiasi terjadi aglomerat dengan pembentukan senyawa baru U(Al,Si)x dan UAlx. Pembentukan aglomerat semakin besar dengan meningkatnya temperatur pemanasan. Interaksi U3Si2 dengan matrik Al maupun kelongsong AlMg2 pada PEB U3Si2-Al pasca iradiasi diperoleh hasil bahwa pada kelongsong bagian atas dan bawah terjadi lapisan oksida dan pada bagian tengah PEB terbentuk layer senyawa U(Al,Si)x berwarna abu-abu terang dengan ketebalan sekitar 1-3 mikron. Dari hasil analisis ini diperoleh bahwa PEB U3Si2-Al pra maupun pasca iradiasi ke duanya menghasilkan senyawa intermetalik U(Al,Si)xINTERACTION OF U3Si2-Al FUEL ELEMENT WITH AlMg2 CLADDING ON POST IRRADIATION WITH LOADING OF URANIUM 2.96 gU/cm3. Interaction of U3Si2-Al fuel element with AlMg2 cladding on post irradiation of 2.96 gU/cm3 loading of uranium (TMU) of U3Si2-Al fuel elements plate (PEB) has been analyzed. The purpose of this research is to study the changes of microstructure of nuclear fuel elements during iradiation in reactor core. Understanding on interaction of U3Si2-Al fuel meat with AlMg2 cladding onpre and post irradiation needed to study the influence of radiation on fuel elements plate. PEB U3Si2-Al with 2.96 gU/cm3 by size 10 × 10 mm were heated in DTA (Differential Thermal Analysis) furnace with temperature variation at 450, 550, 650, 900 and 1350oC to perform pre irradiation test.Post irradiation samples were cut by size 2 × 10 mm as many as three samples taken from bottom, middle, and top of PEB in hotcell.The metallography preparation for each pieces of pre and post irradiation samples of U3Si2-Al fuel elements platewith 2.96 gU/cm3 weredone through steps mounting, grinding, polishing, and etching.Scanning Electron Microscope (SEM-EDS) were used to observe the pre irradiation microstructure of fuel elements U3Si2-Al with AlMg2 cladding interaction, while the post irradiationmicrostructure were observed by optical microscope in hot cell. The result show the interaction of U3Si2 with Al matrix or AlMg2 cladding in pre irradiation PEB U3Si2-Aloccurred agglomeration formed new compouds of U(Al,Si)x and UAlx formation. Agglomeration formation on heated pre irradiation samples were bigger while heating temperature increased. The post irradiation sampels shoed the oxide layer were formed outside the AlMg2 cladding and the inner side of caldding that contact to the fuel meat formedlight-grey U(Al,Si)xlayer at 1-3 micron of thickness.

Effects of Thermal Treatment on the Co-rolled U-Mo Fuel Foils

ABSTRACTA monolithic fuel design based on U–Mo alloy has been selected as the fuel type for conversion of United States’ high-performance research reactors (USHPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U–Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U–Mo foil during fabrication alters the microstructure of both the U–10Mo fuel meat and the U–Mo/Zr interface. This work studied the effects of post-rolling annealing treatment on the microstructure of the co-rolled U–Mo fuel meat and the U–Mo/Zr interaction layer. The U–Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184kJ/mole, consistent with a previous diffusion-couple study. The phases in the U–Mo/Zr interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion-couple study.

Conceptual Nuclear Design of a 20 MW Multipurpose Research Reactor

This paper presents some of studied results of a pre-feasibility project on a new research reactor for Vietnam. In this work, two conceptual nuclear designs of 20 MW multi-purpose research reactor have been done. The reference reactor is the light water cooled and heavy water reflected open-tank-in-pool type reactor. The reactor model is based on the experiences from the operation and utilization of the HANARO. Two fuel types, rod and flat plate, with dispersed U3Si2-Al fuel meat are used in this study for comparison purpose. Analyses for the nuclear design parameters such as the neutron flux, power distribution, reactivity coefficients, control rod worth, etc. have been done and the equilibrium cores have been established to meet the requirements of nuclear safety and performance.