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

2020 ◽  
Vol 23 (2) ◽  
pp. 69
Author(s):  
Surian Pinem ◽  
Tukiran Surbakti ◽  
Imam Kuntoro
Keyword(s):  
Fuel Element ◽  
Cross Sections ◽  
Feedback Mechanism ◽  
Heat Fluxes ◽  
Coolant Temperature ◽  
Fuel Temperature ◽  
Element Analysis ◽  
Reactivity Insertion ◽  
Type Fuel ◽  
Plate Type

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.

scholarly journals ANALISIS KONVEKSI ALAM TERAS REAKTOR TRIGA BERBAHAN BAKAR TIPE PELAT MENGGUNAKAN COOLOD-N2

2015 ◽  
Vol 17 (2) ◽  
pp. 67 ◽  
Author(s):  
Sudjatmi K A ◽  
Endiah Puji Hastuti ◽  
Surip Widodo ◽  
Reinaldy Nazar
Keyword(s):  
Natural Convection ◽  
Fuel Element ◽  
Flow Instability ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Coolant Temperature ◽  
Triga Reactor ◽  
Fuel Elements ◽  
Type Fuel ◽  
Plate Type

ABSTRAK Analisis Konveksi Alam Teras Reaktor Triga Berbahan Bakar Tipe Pelat MENGGUNAKAN COOLOD-N2. Rencana penghentian produksi elemen bakar jenis TRIGA oleh produsen elemen bakar reaktor TRIGA, sudah seharusnya diantisipasi oleh badan pengoperasi reaktor TRIGA untuk menggantikan elemen bakar tipe silinder tersebut dengan tipe pelat yang tersedia di pasaran. Pada penelitian ini dilakukan perhitungan untuk model teras reaktor dengan spesifikasi utama menggunakan bahan bakar U3Si2Al dengan pengayaan uranium  sebesar 19,75% dan tingkat muat 2,96 gU/cm3. Analisis dilakukan menggunakan program COOLOD-N2 yang tervalidasi pada konfigurasi teras TRIGA konversi berbahan bakar tipe pelat, yang tersusun atas 16 elemen bakar, 4 elemen kendali dan 1 fasilitas iradiasi yang terletak tepat di tengah teras. Hasil analisis menunjukkan bahwa dengan temperatur pendingin masuk ke teras sebesar 37oC, dan rasio faktor puncak daya radial ≤ 1,92 maka daya maksimum yang dapat dioperasikan pada moda operasi konveksi bebas adalah 600 kW. Karakteristik termohidrolika yang diperoleh antara lain adalah temperatur pendingin di sisi outlet, kelongsong dan meat masing-masing sebesar 82,39oC, 108,88oC, dan 109,02oC, pada ΔTONB (Temperature Onset of Nucleate Boiling) =7,18oC dan nilai OFIR (Onset of flow instability ratio) =1,03 Hasil yang diperoleh dari perhitungan ini diharapkan dapat dijadikan acuan untuk menentukan tingkat daya reaktor TRIGA berbahan bakar pelat. Kata kunci: TRIGA Konversi, COOLOD-N2, karakteristik termohidrolika, konveksi alam, elemen bakar tipe pelat.  ABSTRACT ANALYSIS OF NATURAL CONVECTION IN TRIGA REACTOR CORE PLATE TYPES FUELED USING COOLOD-N2. Any pretensions to stop the production of TRIGA fuel elements by TRIGA reactor fuel elements manufacturer should be anticipated by the operating agency of TRIGA reactor to replace the cylinder type fuel element with plate type fuel element that available on the market. In this study, the calculation of U3Si2Al fuel with uranium enrichment of 19.75 % and a load level of 2.96 gU/cm3 was performed. Analyses were performed using the validated COOLOD - N2 program. TRIGA conversion core configurations of fuel plate type are composed of 16 fuel elements, 4 control elements and 1 irradiation facilities which are located in the middle of core. The calculation results showed that if the cooling temperature was 37°C, and the ratio of radial power peaking factor ≤ 1.92, then the maximum power that can be operated on free convection mode of operation was 600 kW. The thermalhydraulic characteristic obtained such as coolant temperature at the outlet side, cladding and meat were 82.39°C, 108.88°C and 109.02°C respectively, while the ΔTONB (Temperature Onset of Nucleate Boiling) was 7.18°C and OFIR (Onset of flow instability ratio) value was 1.03. The results are expected to be used as a reference for determining the power level of the TRIGA reactor core plate types fueled. Keywords: TRIGA Convertion, COOLOD-N2, Thermalhydraulics characteristic, natural convection, plate type fuel element.

scholarly journals THE THERMOHYDRAULIC ANALYSIS OF THE BANDUNG RESEARCH REACTOR CORE WITH PLATE TYPE FUEL ELEMENTS USING THE CFD CODE

2018 ◽  
Vol 20 (3) ◽  
pp. 123
Author(s):  
Reinaldy Nazar ◽  
Sudjatmi KA ◽  
Ketut Kamajaya
Keyword(s):  
Fuel Element ◽  
Surface Temperature ◽  
Forced Convection ◽  
Flow Rate ◽  
Research Reactor ◽  
Saturation Temperature ◽  
Cooling Flow ◽  
Fuel Elements ◽  
Type Fuel ◽  
Plate Type

Due to TRIGA fuel elements are no longer produced by General Atomic, it is necessary to find a solution so that the Bandung TRIGA 2000 reactor can still be operated. One solution is to replace the type of fuel elements. Study on using the MTR plate type fuel elements as used in RSG-GAS Serpong has been done for the Bandung TRIGA 2000. Based on the results of the study using CFD computer program, it is found that Bandung TRIGA 2000 with plate type fuel elements cannot be operated up to 2000 kW power by natural convection cooling mode. Therefore, the reactor must be cooled by forced convection. The analysis using forced convection showed that for cooling flow rate of 50 kg/s and various temperatures of 35oC, 35.5 oC and 36 oC, the surface temperature of the fuel element is between 110.37 oC and 111.27 oC. Meanwhile, the cooling water temperature in the corresponding position is between 61.03 oC and 61.95 oC. In this operation condition, the surface temperatures of fuel elements can approach the saturation temperature and nucleat boiling started to occur. Hence, the use of cooling flow rate entering core less than 50 kg/s should be avoided. The surface temperature of fuel elements decreased under saturation temperature if cooling flow rate is greater than 65 kg/s. The surface temperature of fuel elements is achieved at 96.65 oC and coolant temperature in the corresponding position was 54.38 oC. Keywords: Bandung research reactor, plate type fuel element, thermohydraulic, CFD code ANALISIS TERMOHIDROLIK TERAS REAKTOR RISET BANDUNG BERELEMEN BAKAR TIPE PELAT MENGGUNAKAN PROGRAM CFD. Mengingat tidak diproduksinya lagi elemen bakar TRIGA oleh General Atomic, maka perlu diusahakan suatu solusi agar reaktor TRIGA 2000 Bandung dapat tetap beroperasi. Salah satu solusi adalah dengan melakukan penggantian tipe elemen bakar. Pada studi ini telah dianalisis penggunaan elemen bakar tipe pelat yang sejenis dengan yang digunakan di RSG-GAS Serpong, untuk digunakankan pada teras reaktor TRIGA 2000 Bandung. Berdasarkan hasil penelitian yang telah dilakukan dengan menggunakan program komputer CFD, diketahui bahwa reaktor TRIGA berelemen bakar tipe pelat tidak dapat dioperasikan pada daya 2000 kW dengan menggunakan moda pendinginan konveksi alamiah seperti yang digunakan saat ini. Untuk kondisi ini, pendinginan dilakukan dengan moda pendinginan konveksi paksa. Hasil analisis konveksi paksa menunjukkan bahwa dengan menggunakan laju alir pendingin pompa 50 kg/s dan variasi temperatur pada 35 oC, 35,5 oC dan 36 oC, diperoleh temperatur permukaan pelat elemen bakar antara 110,37 oC – 111,27 oC dan temperatur pendinginnya pada posisi terkait antara 61,03 oC – 61,95 oC. Temperatur permukaan pelat elemen bakar ini mendekati temperatur saturasi dan tentunya telah mulai terjadi pendidihan inti, sehingga penggunaan laju alir pendingin masuk teras reaktor kurang dari 50 kg/s perlu dihindari. Temperatur permukaan pelat elemen bakar mulai menurun menjauhi temperatur saturasi jika digunakan laju alir pendingin lebih besar dari 65 kg/s, dengan temperatur permukaan pelat elemen bakar 96,65 oC dan temperatur pendinginnya pada posisi terkait 54,38 oC.Kata kunci: Reaktor riset Bandung, elemen bakar tipe pelat, termohidrolik, program CFD

Coupled Calculation on Fluid Structure Interaction in Plate-Type Fuel Element

2018 ◽  
Author(s):  
Yiqi Yu ◽  
Elia Merzari ◽  
Jerome Solberg
Keyword(s):  
Fuel Element ◽  
Fluid Structure Interaction ◽  
Initial Step ◽  
University Of Missouri ◽  
Fluid Structure ◽  
Structure Interaction ◽  
Coupled Calculation ◽  
Enriched Uranium ◽  
Type Fuel ◽  
Plate Type

In nuclear reactors that use plate-type fuel, the fuel plates are thermally managed with coolant flowing through channels between the plates. Depending on the flow rates and sizes of the fluid channels, the hydraulic forces exerted on a plate can be quite large. Currently, there is a worldwide effort to convert research reactors that use highly enriched uranium (HEU) fuel, some of which are plate-type, to low-enriched uranium (LEU). Because of the proposed changes to the fuel structure and thickness, a need exists to characterize the potential for flow-induced deflection of the LEU fuel plates. In this study, as an initial step, calculations of Fluid-Structure Interaction (FSI) for a flat aluminum plate separating two parallel rectangular channels are performed using the commercial code STAR-CCM+ and the integrated multi-physics code SHARP, developed under the Nuclear Energy Advanced Modeling and Simulation program. SHARP contains the high-fidelity single physics packages Diablo and Nek5000, both highly scalable and extensively validated. In this work, verification studies are performed to assess the results from both STAR-CCM+ and SHARP. The predicted deflections of the plate agree well with each other as well as exhibiting good agreement with simulations performed by the University of Missouri utilizing STAR-CCM+ coupled with the commercial structural mechanics code ABAQUS. The study provides a solid basis for FSI modeling capability for plate-type fuel element with SHARP.

Electron Beam Weldability of AA6061-T6 Aluminum Straps for U-Mo Follower Fuel Assembly Manufacturing

2016 ◽  
Author(s):  
Soo-sung Kim ◽  
Yong-jin Jeong ◽  
Jong-man Park ◽  
Yoon-sang Lee ◽  
Chong-tak Lee
Keyword(s):  
Electron Beam ◽  
Fuel Assembly ◽  
Cross Sections ◽  
Electron Beam Welding ◽  
High Vacuum ◽  
Welding Parameters ◽  
Depth Of Penetration ◽  
Shearing Strength ◽  
Type Fuel ◽  
Plate Type

A procedure for Electron Beam Welding (EBW) was developed for the manufacturing of a follower fuel assembly made of an AA 6061-T6 aluminum straps for a U-Mo plate-type fuel proposed to be used in the future in Korea’s Kijang Research Reactor (KJRR) project. The initial welding trials of the weld samples were carried out with a high vacuum chamber using the EBW process. After investigating the welds, EB welding parameters for the full-sized samples were optimized for the required depth of penetration and weld quality. Subsequently, the weld samples made by the filler specimens showed higher shearing strengths than those of the non-filler specimens. This procedure made by EBW process was also confirmed based on the results of the shearing strength test, an examination of the macro-cross sections, and the fracture surfaces of the welded specimens.

scholarly journals Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Javier González Mantecón ◽  
Miguel Mattar Neto
Keyword(s):  
Fuel Element ◽  
Turbulence Models ◽  
Nuclear Research ◽  
Index Method ◽  
Ansys Cfx ◽  
Commercial Code ◽  
Grid Convergence ◽  
Grid Convergence Index ◽  
Type Fuel ◽  
Plate Type

The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydroelastic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-e model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions.

scholarly journals Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

2008 ◽  
Vol 23 (1) ◽  
pp. 19-30 ◽  
Author(s):  
Ahmed Khedr
Keyword(s):  
Steady State ◽  
Flow Instability ◽  
Continuous Process ◽  
Nucleate Boiling ◽  
Fortran Program ◽  
Coolant Temperature ◽  
Regulatory Body ◽  
Research Reactors ◽  
Type Fuel ◽  
Plate Type

The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

scholarly journals RESONANCE SELF-SHIELDING CALCULATION FOR PLATE-TYPE FUEL ASSEMBLIES

2021 ◽  
Vol 247 ◽  
pp. 03007
Author(s):  
Fan Xia ◽  
Hongchun Wu ◽  
Yunzhao Li ◽  
Jiewei Yang
Keyword(s):  
Cross Sections ◽  
Arbitrary Geometry ◽  
Computational Burden ◽  
Computation Efficiency ◽  
Fuel Assemblies ◽  
Equivalence Theory ◽  
The Common ◽  
Relative Errors ◽  
Type Fuel ◽  
Plate Type

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.

scholarly journals Heat transfer analysis of plate type fuel element of reactor core

2018 ◽  
Author(s):  
Reinaldy Nazar ◽  
Jupiter Sitorus Pane ◽  
Ketut Kamajaya
Keyword(s):  
Heat Transfer ◽  
Fuel Element ◽  
Reactor Core ◽  
Heat Transfer Analysis ◽  
Type Fuel ◽  
Plate Type

Axial Power and Coolant-Temperature Profiles for a Non-Re-Entrant Pressure-Tube Supercritical Water-Cooled Reactor Fuel Channel

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Vitali Kovaltchouk ◽  
Eleodor Nichita ◽  
Eugene Saltanov
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Pressure Tube ◽  
Coolant Temperature ◽  
Outlet Temperature ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Fuel Channel ◽  
Technical Specifications ◽  
Institute Of Technology

The axial power and coolant-temperature distributions in a fuel channel of the Generation IV pressure-tube super-critical water-cooled reactor (PT-SCWR) are found using coupled neutronics-thermal-hydraulics calculations. The simulations are performed for a channel loaded with a fresh, 78-element Th-Pu fuel assembly. Neutronics calculations are performed using the DONJON diffusion code using two-group homogenized cross sections produced using the lattice code DRAGON. The axial coolant temperature profile corresponding to a certain axial linear heat generation rate is found using a code developed in-house at University of Ontario Institute of Technology (UOIT). The effect of coolant density, coolant temperature, and fuel temperature variation along the channel is accounted for by generating macroscopic cross sections at several axial positions. Fixed-point iterations are performed between neutronics and thermal-hydraulics calculations. Neutronics calculations include the generation of two-group macroscopic cross sections at several axial positions, taking into account local parameters such as coolant temperature and density and average fuel temperature. The coolant flow rate is adjusted so that the outlet temperature of the coolant corresponds to the SCWR technical specifications. The converged axial power distribution is found to be asymmetric, resembling a cosine shape skewed toward the inlet (reactor top).

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