Plate-type Fuels Research Articles

High fidelity core flow measurement experiment for an advanced research reactor using a real scale mockup

Owing to spatial effects and vortex flow, flow in research reactors that use plate-type fuels can be maldistributed to the parallel channels of the core, which significantly impacts the reactor safety. In this study, the core flow of an advanced research reactor was measured in a real-scale facility under various hydraulic conditions. For flow measurement, integrated pressure lines were embedded in the mockups of 22 fuel assemblies and six fission molybdenum assemblies. Each assembly mockup was individually calibrated to obtain the relationship between the pressure drop and flow rate. Real-scale facility, which implements the characteristics of the hydraulic conditions in research reactors, was then used to evaluate the assembly-to-assembly flow distribution under normal operating condition, a partially withdrawn condition for the follower fuel assemblies, no flow for the pool water management system, and 1:1.5 asymmetric inlet flow condition. As a parallel channel system, core flow distribution was analyzed with conventional header design approach. Taking into account the measuring uncertainty, the core flow was uniformly distributed within 5 % under all conditions. This was mainly because the core flow resistance was sufficiently high and the vortex flow was minimized by the perforated plate.

Neutronic and thermal-hydraulic assessment of the TRR with new core designed based on tubular fuels

Abstract Herein, the feasibility study of the Tehran Research Reactor (TRR) with a new core designed based on tubular fuels from the neutronic, thermal-hydraulic, safety, and operational points of are investigated using MCNPX, WIMS, CITATION, Computational Fluid Dynamics (CFD), and RELAP codes. According to the results, the total neutron flux in the new core with tubular fuels is increased by more than 14.3 % compared with the current core of the TRR with plate-type fuels. Moreover, due to the higher fuel amount in the tubular compared with plate-type fuels (about 17 % in similar conditions), its effective multiplication factor is much higher than the TRR with plate-type fuels. Moreover, the results show that the maximum cladding temperature is sufficiently lower than 105 °C and the produced heat in the tubular fuel are removed without changing the current flow rate of the core. Furthermore, the maximum fuel temperature in tubular fuel is about 10 °C lower than the maximum fuel temperature in the current standard fuel element.

Fluid structural interaction simulation of the MP-1 plate performance irradiated in the advanced test reactor

Due to the high neutron fluxes they generate and increased heat transfer performance, plate type fuels are used in the U.S. high-performance research reactors. During irradiation, a significant amount of fission energy (i.e., ∼200 MeV per atom) is released by the U-235 chain reaction that is carried away by the coolant. Through thermal hydraulic analysis, the coolant’s heat transfer capability is investigated to ensure that the desired fuel temperature can be maintained. At high temperatures, the plate undergoes elastic/plastic deformation, creep, and swelling as a result of both the temperature gradients and the fission gas production within the fuel. These effects are studied via fuel performance analysis. When the plate deformation is small enough that the coolant flow remains relatively unchanged, conducting these two types of analyses independently will suffice. But at high fuel burnups, the swollen plates may encroach into the coolant channels that separate the fuel plates from each other and cause these channels to narrow. Large reductions in channel gap size imperil cooling performance, causing fuel temperatures to rise. If the plate deformation is asymmetric, the fuel centerline will shift toward one side of the channel, causing an uneven reduction in coolability. In addition, a boehmite (oxide) layer will, over time, grow on the plate surface, further obstructing the heat transfer process. To precisely predict fuel temperatures/deformation, a complete coupled analysis that considers coolant flow, heat transfer, oxide growth, elastic/plastic deformation, creep, and swelling is needed; however, this type of analysis method is not available in the literature. To fill the gap, this research developed a fluid structure interaction (FSI) approach to the fuel plate analysis, then successfully applied it to the Mini-Plate (MP)-1 experiment, which was irradiated in the Advanced Test Reactor (ATR) for both one and two cycles. The complete analysis coupled STAR-CCM+, a computational fluid dynamic (CFD) software for calculating flow, with Abaqus, a finite element analysis code for calculating plate deformation. Improvements in the results were found when comparing the fully coupled analysis to the independently conducted analyses but they were not significant due to the miniature size of the plates and the relatively short irradiation time. In the future, the fully coupled approach presented herein will be applied to full-size fuel plates with longer irradiation cycles once additional experiments become available.

Development and validation of a general two-fluid thermal-hydraulic analysis code for reactors with plate-type fuel assemblies

A general two-fluid thermal-hydraulic analysis code, TTHAC-PFA (Two-fluid Thermal-Hydraulic Analysis Code for reactors using Plate-type Fuel Assemblies), has been developed. The code was developed on basis of the fundamental governing equations. Reasonable mathematic and physical models for parallel rectangular channels and fuel plates have been built up to meet the command of the safety analysis for plate-type fuel reactors. The coolant distribution is conducted on basis of the pressure drop equation theory. Normal subchannel approach is adopted in the code. Fortran language is adopted. The RA-6 experiment is used to evaluate the heat transfer models. The max deviation of the simulations from the measured value is about 5.7 °C. Then, experiments on two-phase flow pressure drop are adopted to validate the code. Approximately 97% of the data points are within ±20% of the deviation range. In addition, CARR (China Advanced Research Reactor) is adopted as the benchmark for normal condition validation. Flow distribution validation and temperature profile validation are conducted. The flow validation presents negligible mean discrepancy and the temperature differences are within the scope of 6 °C. All validations are in good agreement, which indicates the developed code meet the requirements of analyzing thermal-hydraulic characteristics for parallel rectangular channels in reactors. Furthermore, TTHAC-PFA is applied to the investigate channel blockage accident occurred in the CARR. The mass flow is redistributed among channels due to the obstruction. In the obstructed channel and two nearby channels, the coolant temperatures increase a lot. The coolant temperature in the obstructed channel rises by 60 °C. In the rest of the channels, the coolant temperatures decrease with the increase of coolant flow. Besides, the two fuel plates, which form the blocked channel, are asymmetrically cooled due to flow maldistribution. The temperature difference between the cladding on both sides of the fuel plate can reach 40 °C. As a result of partial blockage, the cladding temperature on the side of the blocked channel rises by 63 °C. During the blockage accident, the plate temperature is under the melting temperature. The results indicate the significance of studying the effects of the adjacent channels near the obstructed channel for reactors with plate-type fuel elements.

A new coupling method based on MOC and the SN nodal method for the neutron transport calculations

The ray spacing must be fine enough to traverse every region of the reactor to get accurate results when using the method of characteristics. However, when calculating some research reactors, the active zone is relatively small compared to the entire problem. It would be time-consuming and memory costing to apply MOC for the entire geometry. A new approach that couples MOC and the SN nodal method is proposed and implemented to solve those issues. The calculation region is divided into the MOC domain, which includes active zone or complex structures, and the SN domain, such as the moderator and reflector region. The MOC and SN nodal methods are coupled by the interface angular flux. The 2D C5G7 benchmark and a 2D JRR-3 plate-type fuels reactor problem are tested to demonstrate the accuracy and efficiency of the coupling method, and the numerical results show that the coupling method achieves superior efficiency.

Micro-tensile testing of the bond line in hot isostatic pressed aluminum

Considerable effort is being devoted to development and regulatory qualification of low enriched fuels for research and test reactors by many agencies worldwide. One promising fuel configuration being examined for United States higher power research and test reactors (USHPRRs) are plate-type fuels composed of a metallic uranium-molybdenum foil clad in an aluminum alloy. The two pieces of aluminum alloy cladding are bonded using a hot isostatic pressing method. The mechanical properties of the resulting bond line in the aluminum alloy cladding will vary by the HIP'ing parameters, requiring a need to characterize the bond line. Small scale mechanical testing can provide a path for evaluating the mechanical properties and deformation behavior of the bond line both prior to and following irradiation. In this research, room temperature micro-tensile specimens of non-irradiated and irradiated samples containing an Al alloy (AA 6061) bond line were tested to evaluate its strength and deformation behavior. Observations indicated that the strain rate did not affect the deformation behavior or strength and most of the micro-tensile specimens failed in a ductile mode in grains around the bond line. There was no indication that the microstructural features from the bond line affected the mechanical properties of the micro-tensile specimens. An initial examination was performed on irradiated material but further systematic studies of the effects of irradiation can be performed in the future.

Multi-Objective Optimization Design of Plate-type Fuel via Co-simulation Method

Plate-type fuel element has been implemented and studied widely because of its good heat transfer capability. For the actual design of fuel assembly, the design objectives usually to be in competition, and the constraints also need to be considered. Therefore, this is a challenging constrained multi-objective optimization problem. A co-simulation method has been developed in this study to optimize the design of fuel plate for the thermal–hydraulic performance. Moreover, due to the versatility of this co-simulation method, extra objectives and constraints could be added as needed. First, a two-dimensional thermal-fluid–solid multiphysics model of plate-type fuel assembly was established and solved by the finite element method. After that, the values of relevant physical variables were verified with the three-dimensional model. Finally, the numerical simulation of multiphysics model was incorporated with the multi-objective evolutionary algorithm (NSGA-II, adaptive NSGA-III and CMOEA-MS) which could handle constraints to optimize the thickness parameters of fuel plate. Pareto optimal solutions were obtained separately after the co-simulation with three algorithms, and the qualitative and quantitative analyses of the Pareto fronts showed that CMOEA-MS and NSGA-II had better performance in this study. This co-simulation could provide comprehensive and reliable design information according to the design objectives and requirements.

Power transient critical heat flux in a narrow rectangular channel under downward flow

Power transient critical heat flux (CHF) experiments were conducted under downward flow conditions in a narrow rectangular channel simulating a sub-channel consisting of plate-type fuels for a research reactor. The test section is designed as a sub-channel with a wetted width of 66.6 mm, heated width of 62 mm, gap size of 2.35 mm, and length of 640 mm. In the experiment, a mass flux of 3750 kg/m2s with a downward flow, an inlet temperature of 37 °C, and outlet pressure of 170 kPa were maintained constantly at a given heat flux transient condition. The heat flux was exponentially increased with the power increase rate ranging from 0.03 to 1 s−1 for the simulation of power transients in the reactivity-initiated accidents of the research reactor. It was found that the CHF increased significantly as the power increase rate increased at a given flow condition. The power transient CHF was analyzed using a dimensionless velocity consisting of parameters representing the liquid-sublayer thinning mechanism that is expected in the vicinity of the heated wall. Finally, we developed a power transient CHF correlation based on the experimental data set consisting of the present data, data of Kataoka et al., and the R-12 data of Celata et al. The developed correlations showed an average error of 0.24% and root mean square error of 3.09% against the experimental data set for power increase rates of 0.03–1 s−1.

Numerical Investigation of Conjugated Heat Transfer of the Plate-Type Fuel Assembly in the Research Reactor

In nuclear reactors, the research of conjugated heat transfer between the fuel and coolant in the fuel assembly is fundamental for improving the safety, reliability and economy. The numerical approach based on Computational Fluid Dynamics (CFD) can be used to realize the rapid analysis of the conjugated heat transfer. Besides, the numerical simulation can provide detailed physical fields that are useful for the designing and optimizing of the fuel assembly. The plate-type fuels are generally used to enhance heat transfer in research reactors with high power density. In this study, a standard plate-type fuel assembly in the research reactor was taken into consideration. The solid-fluid conjugated heat transfer of the fuel assembly and coolant was numerically investigated. In the fluid region, the subcooled flow boiling simulation model was established by implementing the Rensselaer Polytechnic Institute model into the Euler multi-phase flow method. The results show that the conjugated heat transfer of the fuel assembly and coolant can be simulated using the model established in this work. The influence of fluid velocity, power density and the width of the flow channel on the temperature distribution and the conjugated heat transfer was investigated and discussed.

RESONANCE SELF-SHIELDING CALCULATION FOR PLATE-TYPE FUEL ASSEMBLIES

Owing to its ability to handle arbitrary geometry and high computation efficiency, subgroup method is a widely used resonance self-shielding method. Ordinarily, the subgroup parameters are generated from homogenous resonance integral tables, and then can be used to calculate heterogeneous problems via the equivalence theory. However, it cannot provide accurate results especially for the plate-type assemblies with strong heterogeneity. What’s more, the fuel enrichment in plate-type assemblies is relatively much higher than the common rod-type ones. As a result, the resonance interference effect in plate-type fuels is particularly intense. To solve this problem and to provide accurate effective self-shielded cross-sections for plate-type fuel assemblies, 1-D plate problems are used to generate subgroup parameters. In order to deal with resonance interference, the resonance interference factors are generated by equivalent homogenous problems solved by 0-D hyperfine group solver. To avoid the computational burden caused by too many hyperfine group calculations, the importance of resonance isotopes is calculated in advance to select important resonance isotopes. Important and non-important ones are handled by different ways respectively. JRR-3M plate-type assemblies are used to test the newly method. Numerical results show that the relative errors of effective self-shielded cross-sections are generally less than 1.5% compared with the reference.

Prediction of critical heat flux for narrow rectangular channels in a steady state condition using machine learning

The subchannel of a research reactor used to generate high power density is designed to be narrow and rectangular and comprises plate-type fuels operating under downward flow conditions. Critical heat flux (CHF) is a crucial parameter for estimating the safety of a nuclear fuel; hence, this parameter should be accurately predicted. Here, machine learning is applied for the prediction of CHF in a narrow rectangular channel. Although machine learning can effectively analyze large amounts of complex data, its application to CHF, particularly for narrow rectangular channels, remains challenging because of the limited flow conditions available in existing experimental databases. To resolve this problem, we used four CHF correlations to generate pseudo-data for training an artificial neural network. We also propose a network architecture that includes pre-training and prediction stages to predict and analyze the CHF. The trained neural network predicted the CHF with an average error of 3.65% and a root-mean-square error of 17.17% for the test pseudo-data; the respective errors of 0.9% and 26.4% for the experimental data were not considered during training. Finally, machine learning was applied to quantitatively investigate the parametric effect on the CHF in narrow rectangular channels under downward flow conditions.

ANALYSIS OF UNCONTROLLED REACTIVITY INSERTION TRANSIENT OF TRIGA MARK 2000 BANDUNG USING MTR PLATE TYPE FUEL ELEMENT

ANALYSIS OF UNCONTROLLED REACTIVITY INSERTION TRANSIENT OF TRIGA MARK 2000 BANDUNG USING MTR PLATE TYPE FUEL ELEMENT. Analysis of uncontrolled reactivity insertion is very important for the safety of reactor operations. Determination of melting point limit, critical heat fluxes and melting temperatures of cladding are the main objectives for most of these studies to determine whether fuel temperature can withstand the transient insertion of reactivity. In this study, uncontrolled reactivity insertion transient was carried out due to the withdrawal of control rods in nominal power of 1 MW and 2 MW. Analysis of reactivity transient was carried out using the WIMSD/5B and MTRDYN codes. The WIMSD/5B code is used to generate cross sections and the MTRDYN program is used for analysis under transient conditions. Based on calculations on the initial power of 1 MW and 2 MW with an insertion of reactivity of greater than 0.5 $/s the reactor operation is not safe because the fuel temperature exceeds the design limit. For reactivity insertion 0.5 $/s allows increased power can be stabilized by feedback reactivity. For 1 MW of nominal power, the maximum coolant temperature, cladding and fuel are 86.39 oC, 164.86 oC and 165.33 oC, respectively. For 2 MW of nominal power, the maximum coolant temperature, cladding and fuel are 89.09 oC, 176.96 oC and 177.602 oC, respectively. Based on calculation, It is concluded that the feedback mechanism can protect the fuel cladding from a local meltdown if reactivity insertion 0.5 $/s and the reactor is in nominal power of 1 MW and 2 MW.

A study on the numerical analysis for OFI in a narrow rectangular flow channel

The prediction of the OFI is important for open pool type research reactors since the flow instability may lead to a premature CHF and cause a fuel failure. Hence, we implemented two-phase models on the commercial CFD code using user-defined functions to simulate a subcooled boiling and predict OFI flow at low pressure conditions. The capability of this CFD code with the two-phase models was assessed by the W&F experiment, which is for OFI in a rectangular flow channel with the same heat fluxes on both sides. Numerical assessments for OFI in a narrow rectangular channel with different heat fluxes on both sides was also simulated considering the practical operating conditions of typical open pool type research reactors with plate type fuels. Simulation results show that a commercial CFD code with the models implemented in this study properly predict the OFI at low pressure in a narrow rectangular channel.

Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core

The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

Establishment of the design stress intensity value for the plate-type fuel assembly using a tensile test

In this paper, the design stress intensity values for the plate-type fuel assembly for research reactor are presented. Through a tensile test, the material properties of the cladding (aluminum alloy 6061) and structural material (aluminum alloy 6061-T6), in this case the yield and ultimate tensile strengths, Young’s modulus and the elongation, are measured with the temperatures. The empirical equations of the material properties with respect to the temperature are presented. The cladding undergoes several heat treatments and hardening processes during the fabrication process. Cladding strengths are reduced compared to those of the raw material during annealing. Up to a temperature of 150 °C, the strengths of the cladding do not significantly decrease due to the dislocations generated from the cold work. However, over 150 °C, the mechanical strengths begin to decrease, mainly due to recrystallization, dislocation recovery and precipitate growth. Taking into account the uncertainty of the 95% probability and 95% confidence level, the design stress intensities of the cladding and structural materials are established. The presented design stress intensity values become the basis of the stress design criteria for a safety analysis of plate-type fuels.

Instability measurement for plate-type fuels subjected to narrow-parallel flow

Plate-type fuel assemblies are utilized in many research reactors. Some are open-loop, pool-type reactors, but others are pressurized reactors with high coolant flow rates. Parallel flows between the fuel plates can make these plates unstable. Several studies in the early 60s reported that the plates could lose their stability through a static type of instability. In accordance with studies in the 70s, however, the type of instability, whether static or dynamic, may depend on the boundary conditions. In this study, we present experimental observations of the instability of a fuel plate subjected to a narrow-axial flow between several different fuel plates. We sought to measure and compare the critical flow velocity with predictions according to Miller’s theory, which is generally used during the design of a plate-type fuel. For this experiment, we installed strain gages on a target plate (instrumented plate) positioned at the center of five simulated plates to reflect the fluid interaction in the narrow gaps between the plates. Pressure sensors were also installed to measure the static pressure differences between two channels adjacent to the target plate in an effort to identify the moment of instability. We observed that the stability of the target plate is lost due to a static type of instability at a flow velocity of 16.4 m/s, slightly lower than the flow velocity predicted using Miller’s equation for a fixed–fixed boundary condition.

Development of phenomena identification and ranking tables (PIRTs) to implement research reactor-specific capability in SPACE code

The benefits of research reactor technology are always essentially accompanied by nuclear safety, which is ensured by a series of supporting analysis such as safety analysis. In order to extend the applicability of the Safety and Performance Analyzing Code (SPACE code) to safety analysis of typical open-tank-in-pool type research reactors using plate-type fuels, a Phenomena Identification and Ranking process was performed for most of the relevant internal event scenarios of a reference research reactor, i.e., the KiJang Research Reactor, designed and under development by the Korea Atomic Energy Research Institute (KAERI). The Structures, Systems, and Components (SSCs) important to nuclear safety of the reactor were included in the PIR processes. A summary of important results from the PIRT study on the research reactors is presented with a proposal on a preliminary set of data bases for a validation and verification matrix in addition to research issues.

Evaluation of Fuel Burn-up and Radioactivity Inventory in the 2 MW TRIGA-Plate Bandung Research Reactor

Currently plate-type fuels are planned to use in the research reactor TRIGA-2000 Bandung, Indonesia (further called as TRIGA-Plate reactor). Plate type fuel consists of enriched uranium sandwiched between metal cladding. Plate-type fuel is used in some research reactors to obtain high neutron flux and is used to study material irradiation or isotope production. For the purpose of Safety Analysis Report (SAR) on the fuel-type modification plan of TRIGA-2000 Bandung, it is necessary to calculate radioactivity inventory from the reactor core under operating conditions. The radioactivity inventory of materials irradiated in a reactor depends on the initial material composition at the BOC, burn-up history from BOC to EOC and the power peaking factor. The purpose of the present work is to evaluate the fuel burn-up and radioactivity inventory in fuel materials of TRIGA-Plate research reactor. The mass of U-235 consumed and the mass of Pu-239 and Th-232 produced are also evaluated. The calculation results obtained using ORIGEN2 code are: the mass of U-235 consumed is 1.40E+02 gram; while the mass of Pu-239 produced is 1.36E+00 gram and Th-232 produced is 1.03E-06 gram. The largest radioactivity inventories of the reactor at the EOC sequentially are: Kr group is about 2.29E+02 Ci; for group I is 7.77E+03 Ci and for groups Cs is 1.92E+02Ci.

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and multipurpose reactors. To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions. We consider the chain or cascade method in two steps to facilitate the analysis of whole core. In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly. In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel. The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination. The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling. To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor. The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%. The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor. The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.

Investigation of the CHF correlation for a narrow rectangular channel under a downward flow condition

In present study, CHF experiments were conducted to evaluate and improve existing CHF correlations for narrow rectangular channels under a downward flow condition. A test section heated on both sides simulates a subchannel of plate-type fuels in a research reactor. The experiment covers a wide range of flow conditions: 0–2739 kg/m2 s for mass flux, 37–99 °C for inlet temperatures and 120–224 kPa for outlet pressures. The experimental data obtained in the present study are compared with prediction results of existing CHF correlations. From the evaluation, the correlation proposed by Kaminaga et al., (hereafter referred to as Kaminaga correlation) shows the best prediction performance for the experimental data among them. However, it was also found that the Kaminaga correlation still contains a large degree of uncertainty relative to the experimental data. In this study, we attempted to improve the Kaminaga correlation to minimize its prediction uncertainty. For this purpose, we investigated the gap size, mass flux and subcooling effects on the CHF in a narrow rectangular channel based on both the present and Mirshak CHF data. Finally, the Kaminaga correlation was modified to consider the effects of the gap size, mass flux, and subcooling on the CHF.