Plate-type Fuels Research Articles

Conceptual Design On N16 Decay Chamber For Modified TRIGA-2000 With Plate-Type Fuel

Conceptual Design On N16 Decay Chamber For Modified TRIGA-2000 With Plate-Type Fuel. the TRIGA-2000 is a research reactor in bandung that will be modified using plate-type fuel. The reactor core cooling system is changed from the natural convection cooling mode to the forced convection mode. The purpose of the study is to assess the conceptual design for the decay chamber of N16 nuclide in the primary cooling system of the reactor. In this design, the hold-up system decays the nuclide of N16 resulted from neutron activation product. In the period of 50 seconds, the activity of N16 (T1/2= 7.13 seconds) decays 7 time from half life to low level. The cube shape of decay chamber is provided a plate with 4 hollows and facility to flush the cavitation bubbles. The decay chamber, which is submerged into the bulk shielding as located outside of the reactor pool. The conceptual design uses the Fluent software compared with the analytical estimation for flow velocity in the decay chamber. The result shows a good agreement range with the analytical estimations. The uniform flow profile can be obtained at the velocity of about 0.4 m/s. Water flow life time of 50 seconds in the decay chamber with the capacity of 3.5 m3 is able to decay the N16 nuclide to low level. This decay chamber is expected to contribute in completing the design of reactor primary coolant system using the forced convection mode.

Investigation of flapper valve malfunction in Tehran Research Reactor with porous media model

In the majority of research reactors, the core cooling is a combination of forced and Natural Circulation (NC). Typically, in normal operation the generated heat is removed by forced convection whilst the NC prepares residual decay heat removal. In this study, the focus is on evaluation of Tehran Research Reactor (TRR) core in failure of the natural cooling system. In present work, the integrity of hot fuel plate in the NC cooling process of the TRR, with flapper valve (FV) malfunction has been analyzed. In this way, for evaluation of the incore Thermal-Hydraulics (TH) the COBRA-EN code is enhanced, numerically. A finite element solver for plate type fuels is included and the appropriate heat transfer correlations in laminar and turbulent flow are added. For CFD study of the pool, regarding the fluid-solid domain in the TRR, the porous media model has been applied and the resistance coefficients for porous media model are extracted. The CFD streamlines have been used in COBRA-EN for TH study of hot and normal plates in inversions. In this analysis, the flow patterns in reactor core and pool for normal operation of the FV and its stuck in opening during NC are evaluated. The analysis shows that clad temporarily exceeds from allowable temperature for 74% opening of the FV and remains for seconds. In following, for 56% of opening the void generation is seemed in addition to permanent clad temperature exceeding. Therefore, the 56% of the FV opening due to failure is introduced as criteria for core and fuel plate damage.

Preliminary study for alternative conceptual core design of the MTR research reactor

The utilization of the research reactor is increasingly widespread, especially for radioisotope production and testing of advanced materials and preference to use a compact core. The reactor core design has been determined on the maximum thermal flux in the middle of the core per MW. BATAN has designed several alternative research reactor cores. The purpose of this research reactor is to obtain the optimum reactor core configurations with the criteria to have a thermal neutron flux in the centre of the core with minimum of 1.0×1015 n/cm2 s. Power level of the research reactor is 60 MWthwith U9Mo/Al fuel 85 cm of height. Design of plate-type fuels with a higher core results in the heat transfer to the coolant optimal. All 16 fuel assemblies and the 4 control rodsare inserted into the core for this reactor. The core design calculations were carried out with the WIMSD-5B and BATAN-FUEL codes. Conceptual design calculation results show that the core configuration with 5 × 5 grids, all the fresh fuel, fuel loading of 470 g, a D2O reflector, a maximum thermal neutron flux in the central core is 1.09 x 1015 n/cm2s and the cycle length is 33 days. The reactor core design is the most optimal for MTR type. For the equilibrium core, a fuel loading of 600 g results in the maximum thermal flux of 1.07x1015 n/cm2s and the two safety rods should be used in the core.

Natural convection cooling characteristics in a plate type fuel assembly of Kyoto University Research Reactor during loss of coolant accident

The transient behavior of nuclear fuel during loss of coolant accident (LOCA) must be evaluated not only for power reactors but also for research reactors. Kyoto University Research Reactor (KUR) is a light-water moderated tank-type reactor operated at the rated thermal power of 5MW and has been widely utilized for a lot of scientific researches. The safety of the plate-type fuels in the KUR has been studied by a reactor coolant system analysis code. However, for the accurate prediction of the natural convection phenomena of air during LOCA, the empirical correlation for natural convection should be selected properly and an experimental validation should be performed. In this study, the natural convection heat transfer characteristics were experimentally investigated using a simulated fuel assembly which has the same dimension with the KUR fuels. Furthermore, a simplified unsteady heat conduction simulation was performed by taking the axial heat density distribution into account, and the cooling feature of the fuel plate during LOCA was discussed by estimating the time to the meltdown.

Effects of the Foil Flatness on Irradiation Performance of U10Mo Monolithic Mini-Plates

Monolithic plate-type fuels comprise a high-density, low-enrichment, U10Mo fuel foil encapsulated in a cladding material. This concept generates several fabrication challenges, including flatness, centering, or thickness variation. There are concerns whether these parameters have implications on overall performance. To investigate these inquiries, the effects of the foil flatness were studied. For this, a representative plate was simulated for an ideal case. The simulations were repeated for additional cases with various foil curvatures to evaluate the effects on the irradiation performance. The results revealed that the stresses and strains induced by fabrication process are not affected by the flatness of the foil. Furthermore, fabrication stresses in the foil are relieved relatively fast in the reactor. The effects of the foil flatness on peak irradiation stress-strains are minimal. There is a slight increase in temperature for the case with maximum curvature. The major impact is on the displacement characteristics. While the case with a flat foil produces a symmetrical swelling, if the foil is curved, more swelling occurs on the thin-cladding side and the plate bows during irradiation.

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

The Advanced Test Reactor (ATR) is conducting scoping studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low-enriched uranium (LEU) composition, through the Reduced Enrichment for Research and Test Reactors Program, within the Global Threat Reduction Initiative. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary reactor physics scoping and feasibility analysis of TRIGA fuel within the current ATR fuel element envelope and compares it to the functional requirements delineated by the Naval Reactors Program, which includes >4.8×1014 fissions/s·g−1 of 235U in test positions, a fast–to–thermal neutron flux ratio that has a <5% deviation from its current value, a desired steady cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other design parameters outside those put forth by the Naval Reactors Program that are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time. The result of this study demonstrates potential promise for implementation of TRIGA fuel in the ATR from a reactor physics perspective; discussion of observations and limitations are provided herein.

Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

A novel research reactor concept based on coated particle fuel

This work presents a preliminary study of a novel plate-type fuel concept for a high-performance and ultra-safe research reactor. This new fuel type consists of coated particle fuel (CPF) randomly dispersed in an aluminum matrix with a certain packing fraction that can be adjusted depending on the reactor design requirements. The CPF can also be varied in the fuel kernel material between UC and UO2. For the purpose of this study, UO2 was used as the reference fuel type. Using this novel fuel type, a 20MWth pool-type research reactor was investigated to determine the preliminary performance and safety characteristics of the new fuel. The core thermal analysis was done using the MATRA-P code. The neutronics analysis was done using the Monte Carlo Serpent code for an equilibrium cycle resulting from a multi-batch fuel management. In this analysis, it was found that the Doppler effect is significantly enhanced through the implementation of CPF, in turn improving the inherent safety of the reactor. In addition to the notable improvement in safety, the new fuel type also promises to be able to achieve a high thermal neutron flux, improving the performance and utility of the reactor. It is concluded that the CPF-based fuel concept presented in this paper can enable new high-flux reactor designs using simple plate-type fuels with improved safety.

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U–Mo Alloy Plate-Type Fuel

The uranium–molybdenum (U–Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium–molybdenum (U–Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium–molybdenum (DU–Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U–Mo alloys as a function of temperature.

Transient analysis of break below the grid in Tehran research reactor using the newly enhanced COBRA-EN code

Recent availability of high performance computers and computational methods along with continuing increase in operational experience necessitate revising some operational constrains and conservative safety margins. Implementation of advanced methods constitute a real necessity in the safety and design analysis and allow obtaining more realistic simulation of the processes taking place during the normal operations and transients.In this paper, COBRA-EN thermal–hydraulic code is extended and upgraded to include analysis of plate type fuels in research reactors. Furthermore, the convection heat transfer coefficient in the code is modified and a simple model of natural convection is added and a procedure to take hot channel data or other required information in each time step is included for post-processing of the results. To evaluate the capability of the upgraded code three cases of break below the grid accident in Tehran research reactor (TRR) core which contains plate type fuels are simulated. The safety analysis report of TRR confirms the results, convergence rate and reliability of developed code.

Assessment of residual stresses on U10Mo alloy based monolithic mini-plates during Hot Isostatic Pressing

This article presents an assessment of the residual stresses in U–10 wt.% Mo (U10Mo) alloy based monolithic fuel plates and the elasto-plastic response to thermo-mechanical processing. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Understanding of the three-dimensional residual stress field is important for understanding the in-reactor performance of these plate-type fuels. To define fuel-cladding stress–strain characteristics, a thermo-mechanical finite element model was developed. During fuel plate fabrication, the hot pressing temperature approaches the melting temperature of the cladding, so that temperature dependent material properties were incorporated to improve the accuracy of the model. By using elasto-thermo-plastic material models, it was determined that the cladding material (Al6061-O) is subjected to tensile stresses that exceed its proportional limits. The fuel foil is subject to compressive stresses and remains below yield. The residual stresses in the plates are significant, and therefore, should not be neglected. In particular, the simulations indicate the presence of high stress gradients at the fuel/cladding interface, thus emphasizing the need for a high quality bond.

Effects of Applied Load on 6061-T6 Aluminum Joined Employing a Novel Friction Bonding Process

Friction bonding is under consideration for use in mass production of plate-type nuclear fuels for research reactors. This article discusses the effects of applied load (the most important process parameter for fabrication of these fuels) on temperature distribution, microstructure, and mechanical properties. Friction bonding experiments showed that tool geometry caused temperature gradients across the tool surface. Temperatures at the joint interface suggested the advancing side of the tool produced a majority of the frictional heat, while the retreating side of the tool mainly forged the plasticized material while bonding increased with applied load. The microstructure across the tool surface was also altered and, as a function of applied load, affected the mechanical properties. The 6061 aluminum alloy had mechanical properties close to a T4 temper after processing. Results documented in this article will aid in continual enhancement of friction bonding for nuclear fuel plate fabrication, and will hopefully contribute to continued advancement of friction stir welding (FSW) state of the art.

Hydrogen Absorption-Desorption and Gamma-UMo Nuclear Fuel Powder Production

Gamma uranium-molybdenum alloys has been considered as the fuel phase in plate type fuel elements for MTR reactors due to its performance under irradiation and metallurgical processing. To its usage as dispersion phase in aluminum matrix, a necessary step is the conversion of the as cast structure into powder, and the technique considered at IPEN / CNEN - Brazil was HDH (hydration-dehydration). This work has the aim to study the hydrogen incorporation by gamma-UMo alloys with 8% weight molybdenum. The samples were thermally treated under constant flow of hydrogen, for temperatures varying from 500oC up to 600oC and times of 1 to 4 hours. Some of the curves relating mass incorporation and time for the above temperatures were obtained, and the results related to its microstructures and ease of fragmentation.

Flow regime transition criteria for upward two-phase flow in vertical narrow rectangular channels

In relation to the cooling system of high performance microelectronics, a high power research reactor with plate type fuels and plasma facing components of a fusion reactor, study of two-phase flow in a narrow rectangular channel has been paid considerable attention, recently. For the two-fluid model, direct geometrical parameters such as the void fraction should be used in flow-regime criteria. From this point of view, flow-regime transition criteria for vertical upward flows in narrow rectangular channels have been developed considering the mechanisms of flow-regime transitions. The basic concept of the present modeling followed the Mishima and Ishii model for vertical upward two-phase flows in round tubes. Newly developed criteria have been compared with the existing experimental data for air–water flows in narrow rectangular channels with the gaps of 0.3–17 mm. The present criteria showed satisfactory agreements with those data. Further comparisons with data for steam–water in a rectangular channel at relatively high system pressures have been made. The results confirmed that the present flow-regime transition criteria could be applied over wide ranges of parameters as well as to boiling flow.

Improvement of Critical Heat Flux Correlation for Research Reactors using Plate-Type Fuel

In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as “Loss of the primary coolant flow”. Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal.The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF.A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and non-uniform heat flux conditions. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed. The new correlation could be adopted under the conditions of the atmospheric pressure, the inlet subcooling less than 78K, the channel gap size between 2.25 to 5.0mm, the axial peaking factor between 1.0 to 1.6 and L/De between 71 to 174 which were the ranges investigated in this study.

Improvement of Critical Heat Flux Correlation for Research Reactors using Plate-Type Fuel.

In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as “Loss of the primary coolant flow”. Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal.The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF.A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and non-uniform heat flux conditions. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed. The new correlation could be adopted under the conditions of the atmospheric pressure, the inlet subcooling less than 78K, the channel gap size between 2.25 to 5.0mm, the axial peaking factor between 1.0 to 1.6 and L/De between 71 to 174 which were the ranges investigated in this study.

Transient Behavior of Low Enriched Uranium Silicide Plate-Type Fuel for Research Reactors during Reactivity Initiated Accident Conditions

Abstract This paper describes the results of transient experiments using a low enriched uranium silicide miniplate fuel for research reactor. The pulse irradiation was performed in the Nuclear Safety Research Reactor (NSRR) at the Japan Atomic Energy Research Institute (JAERI). The results obtained in this study are summarized as follows : (1) The tested fuel plates were damaged with energy depositions above 94 cal/g·fuel, but remained intact below 82cal/g·fuel. A failure threshold should therefore exist between these two values. (2) Four of the fuel plates that showed peak cladding surface temperatures <330°C were damaged by the thermal stress during quenching. These damaged fuel plates revealed small intergranular cracks that propagated perpendicularly to the axial direction of the plate, from the Al cladding surface to the fuel core, without significant dimensional changes. On the other hand, when peak cladding surface temperatures were >400°C, the test fuel plates were damaged mainly by melting of the Al cladding, accompanying significant dimensional changes. (3) The thermal stress of the damaged fuel plates calculated on the basis of the maximum transient temperature drop during quenching was greater than the tensile stress that occurred during fabrication.

Transient Behavior of Low Enriched Uranium Silicide Plate-Type Fuel for Research Reactors during Reactivity Initiated Accident Conditions.

In many research reactors and test reactors, a low enriched ( 94 cal/g • fuel plate. A failure threshold must be existed between these two values.

Identification of Flow Patterns by Neutron Noise Analysis during Actual Coolant Boiling in Thin Rectangular Channels

The primary objective of this paper is to introduce results of coolant boiling experiments in a simulated materials test reactor-type fuel assembly with plate fuel in an actual reactor environment. The experiments have been performed in the Hoger Onderwijs Reactor (HOR) research reactor at the Interfaculty Reactor Institute, Delft, The Netherlands. In the analysis, noise signals of self-powered neutron detectors located in the neighborhood of the boiling region and thermocouple in the channel wall and in the coolant are used. Flow patterns in the boiling coolant have been identified by means of analysis of probability density functions and power spectral densities of neutron noise. It is shown that boiling has an oscillating character due to partial channel blockage caused by steam slugs generated periodically between the plates. The observed phenomenon can serve as a basis for a boiling detection method in reactors with plate-type fuels.

Dimensional Stability of Low Enriched Uranium Silicide Plate-Type Fuel for Research Reactors at Transient Conditions.

This paper describes the result of transient experiments using low enriched uranium silicide plate-type fuel for research reactors. The pulse irradiation was carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute. The results obtained were: 1. At fuel plate temperature of below 400°C, a good dimensional stability of the tested fuel was kept. No fuel failure occurred.2. At a plate temperature of about 540°C, a local crack was initiated on the AI-3%Mg alloy cladding. Once the cladding temperature exceeded the melting point of 640°C, the fuel plate was degraded much by increased bowing and cracking of the denuded fuel meat occurred after relocation of molten Al cladding. Despite of these degradation, neither fragmentation of the fuel plate nor mechanical energy generation occurred up to the cladding temperature of 971 °C.3. At the temperatures of around 925°C, the reaction of silicide particles with molten Al in the matrix and that of cladding occurred, forming Al riched U (Al, Si) compounds and Si riched (U, Si) compounds at the outermost surface of the silicide particles.