Design of U3Si2-Al Plate-Type Fuel Element for China Advanced Research Reactor

2010 ◽  
Author(s):  
Aimin Zhang ◽  
Yalun Kang
Keyword(s):  
Fuel Element ◽  
Neutron Flux ◽  
Research Reactor ◽  
High Heat ◽  
Reactor Power ◽  
High Heat Flux ◽  
Water Tank ◽  
Enriched Uranium ◽  
Fuel Elements ◽  
Plate Type

China Advanced Research Reactor (CARR), which will be critical in China Institute of Atomic Energy (CIAE) in 2010, is a multipurpose, high neutron flux and tank-type (inverse neutron trap) reactor with compact core. Its nominal reactor power is 60MW and the maximum thermal neutron flux is about 8.0×1014n/cm2·s in heavy water tank. It has a cylindrical core having a diameter of about 450mm and a height of 850mm. The CARR’s core consists of seventeen plate-type standard fuel elements and four follower fuel elements, initially loaded with 10.97 kg of 235U. The fuel element has been designed with U3S2-Al dispersion containing 235U of (19.75±0.20)wt.% low enriched uranium (LEU) and having a density of 4.3gU/cm3. The aluminum alloy is used as the cladding. There are twenty-one and seventeen fuel plates in the standard and follower fuel element, respectively. There are specific requirements for design of the fuel element and strict limitation for the operation parameters due to the high heat flux and high velocity of coolant in CARR. Irradiation test of fuel element had been carried out at fuel element power of 3.1±20%MW at Russia MIR reactor. Average burnup of fuel element is up to 40%. This paper deals with the detailed design of fuel element for CARR, out-pile and in-pile test projects, including selection of fuel and structure material, description of element structure, miniplates and fuel element irradiation experiment, measurement of properties of fuel plate, fabrication of fuel element and test results.

scholarly journals Changes of the Neutron Flux of the Nuclear Reactor Triga Mark III Since the Conversion from High to Low 235U Enrichment

2021 ◽  
Vol 8 (2) ◽  
pp. 149-153
Author(s):  
C. Vázquez-López ◽  
O. Del Ángel-Gómez ◽  
R. Raya-Arredondo ◽  
S. S. Cruz-Galindo ◽  
J. I. Golzarri-Moreno ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Nuclear Reactor ◽  
Research Reactor ◽  
Nuclear Research ◽  
Reactor Power ◽  
Uranium Fuel ◽  
Acrylic Sheet ◽  
Enriched Uranium ◽  
Reactor Operating ◽  
Fuel Elements

The neutron flux of the Triga Mark III research reactor was studied using nuclear track detectors. The facility of the National Institute for Nuclear Research (ININ), operates with a new core load of 85 LEU 30/20 (Low Enriched Uranium) fuel elements. The reactor provides a neutron flux around 2 × 1012 n cm-2s-1 at the irradiation channel. In this channel, CR-39 (allyl diglycol policarbonate) Landauer® detectors were exposed to neutrons; the detectors were covered with a 3 mm acrylic sheet for (n, p) reaction. Results show a linear response between the reactor power in the range 0.1 - 7 kW, and the average nuclear track density with data reproducibility and relatively low uncertainty (±5%). The method is a simple technique, fast and reliable procedure to monitor the research reactor operating power levels.

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.

scholarly journals Effects of the Foil Flatness on the Stress-Strain Characteristics of U10Mo Alloy Based Monolithic Mini-Plates

2014 ◽  
Author(s):  
Hakan Ozaltun ◽  
Pavel Medvedev
Keyword(s):  
Mechanical Performance ◽  
Peak Stress ◽  
Stress Strain ◽  
Strain Behavior ◽  
Uranium Fuel ◽  
Considerable Impact ◽  
Enriched Uranium ◽  
Fuel Elements ◽  
Cladding Material ◽  
Plate Type

The effects of the foil flatness on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation were studied. Monolithic plate-type fuels are a new fuel form being developed for research and test reactors to achieve higher uranium densities. This concept facilitates the use of low-enriched uranium fuel in the reactor. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. To evaluate the effects of the foil flatness on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate from RERTR-12 experiments (Plate L1P756) was considered. Both fabrication and irradiation processes of the plate were simulated by using actual irradiation parameters. The simulations were repeated for various foil curvatures to observe the effects of the foil flatness on the peak stress and strain magnitudes of the fuel elements. Results of fabrication simulations revealed that the flatness of the foil does not have a considerable impact on the post fabrication stress-strain fields. Furthermore, the irradiation simulations indicated that any post-fabrication stresses in the foil would be relieved relatively fast in the reactor. While, the perfectly flat foil provided the slightly better mechanical performance, overall difference between the flat-foil case and curved-foil case was not significant. Even though the peak stresses are less affected, the foil curvature has several implications on the strain magnitudes in the cladding. It was observed that with an increasing foil curvature, there is a slight increase in the cladding strains.

scholarly journals Characterization of neutron spectrum parameters at irradiation channels for neutron activation analysis after full conversion of the Dalat nuclear research reactor to low enriched uranium fuel

2014 ◽  
Vol 4 (1) ◽  
pp. 70-75
Author(s):  
D. Vu C. ◽  
Q. Thien T. ◽  
V. Doanh H. ◽  
D. Quyet P. ◽  
T. Anh T.T. ◽  
...  
Keyword(s):  
Neutron Activation Analysis ◽  
Neutron Flux ◽  
Neutron Spectrum ◽  
Thermal Neutron Flux ◽  
Research Reactor ◽  
Nuclear Research ◽  
Fast Neutron Flux ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Full Core

In the framework of the program on Russian Research Reactor Fuel Return (RRRFR) and the program on Reduced Enrichment for Research and Test Reactor (RERTR), the full core conversion of the Dalat Nuclear Research Reactor (DNRR) to low enriched uranium (LEU, 19.75% 235U) fuel was performed from November 24, 2011 to January 13, 2012. The reactor is now operated with a working core consisting of 92 WWR-M2 LEU. After the full core conversion, the neutron spectrum parameters which are used in k0-NAA such as thermal neutron flux (fth), fast neutron flux (ffast), f factor, alpha factor (a), and neutron temperature (Tn) have been re-characterized at four different irradiated channels in the core. Based on the experimental results, it can be seen that the thermal neutron flux decreases by 6÷9% whereas fast neutron flux increases by 2÷6%. The neutron spectrum becomes‘harder’ at most of irradiated positions. The obtained neutron spectrum parameters from this research are used to re-establish the procedures for Neutron Activation Analysis (NAA) according to ISO/IEC 17025:2005 standard at NuclearResearch Institute.

scholarly journals RELAP5 SIMULATION FOR SEVERE ACCIDENT ANALYSIS OF RSG-GAS REACTOR

2018 ◽  
Vol 20 (1) ◽  
pp. 23 ◽  
Author(s):  
Andi Sofrany Ekariansyah ◽  
Endiah Puji Hastuti ◽  
Sudarmono Sudarmono
Keyword(s):  
Fuel Element ◽  
Secondary Flow ◽  
Research Reactor ◽  
Natural Circulation ◽  
Safety Analysis ◽  
Reactor Power ◽  
Severe Accident ◽  
Accident Analysis ◽  
Plate Temperature ◽  
The One

The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational chararacteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element.Keywords: loss of flow, blockage, fuel plate, RSG-GAS, RELAP5 SIMULASI RELAP5 UNTUK ANALISIS KECELAKAAN PARAH PADA REAKTOR RSG-GAS. Reaktor riset di dunia diketahui lebih aman dari pada reaktor daya karena desainnya yang lebih sederhana pada teras dan karakteristika operasinya. Namun demikian, potensi bahaya reaktor riset terhadap publik dan lingkungan tidak bisa diabaikan karena beberapa fitur tertentu. Oleh karena itu, level keselamatan reaktor riset harus jelas ditunjukkan dalam Laporan Analisis Keselamatan (LAK) dalam bentuk analisis keselamatan yang dilakukan dengan berbagai macam pendekatan dan metode dan didukung dengan alat komputasi. Tujuan penelitian ini adalah untuk mensimulasikan beberapa kecelakaan parah pada reaktor RSG-GAS yang dapat menyebabkan kerusakan bahan bakar untuk memperkuat hasil analisis kecelakaan parah yang sudah ada dalam LAK. Simulation dilakukan dengan program perhitungan RELAP5/SCDAP/Mod3.4 yang memiliki kemampuan untuk memodelkan elemen bahan bakar tipe pelat di RSG-GAS. Tiga kejadian telah disimulasikan yaitu hilangnya aliran primer dan sekunder dengan kegagalan reaktor untuk dipadamkan, tersumbatnya beberapa kanal pendingin bahan bakar pada daya penuh, dan hilangnya aliran primer dan sekunder yang diikuti dengan tersumbatnya beberapa kanal pendingin bahan bakar setelah reaktor padam. Kejadian pertama akan membahayakan pelat bahan bakar dengan naiknya temperatur kelongsong hingga titik lelehnya yaitu 590 °C. Tersumbatnya satu atau beberapa kanal pada satu elemen bahan bakar menyebabkan konsekuensi yang berbeda pada pelat bahan bakar, dimana paling sedikit tersumbatnya 2 kanal akan merusak satu pelat bahan bakar, apalagi tersumbatnya satu elemen bahan bakar. Kombinasi antara hilangnya aliran pendingin primer dan sekunder yang diikuti dengan tersumbatnya satu kanal bahan bakar setelah reaktor dipadamkan menyebabkan naiknya temperatur kelongsong di bawah titik lelehnya yang berarti sirkulasi alam yang terbentuk dan daya yang terus turun cukup untuk mendinginkan elemen bahan bakar.Kata kunci: kehilangan aliran, penyumbatan, pelat bahan bakar, RSG-GAS, RELAP5

Research Reactor VR-1 and Determination of Neutron Spatial Distribution in Its Radial Experimental Channel

2009 ◽  
Author(s):  
Marija Mileticˇ
Keyword(s):  
Spatial Distribution ◽  
Neutron Flux ◽  
Flux Density ◽  
Research Reactor ◽  
Reactor Power ◽  
Activation Detector ◽  
Experimental Channel ◽  
Neutron Flux Density ◽  
Radial Channel

The research reactor VR-1 is operated by Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University (CTU) in Prague. It is a pool-type, light-water reactor, with low enriched uranium. Maximum thermal power is 1kW (equal to 1·108 impulses/second when compared with reactors with higher power). Research on VR-1 reactor is mainly used for the education of university students, preparation and testing of new educational methodologies, investigation of reactor lattice parameters, reactor dynamics study, research in the control equipment field, neutron detector calibration, etc. One of the applications performed by students is the determination of the absolute value of the neutron flux density (also known as Neutron Spatial Distribution) in the radial experimental channel in reactor VR-1. The method used for this measurement is Neutron Activation Analysis. The principle of this method consists in neutron capture in a nucleus of the material forming the activation detector which is irradiated in the experimental channel. The activity of the produced radioactive products (radioisotopes) is then measured by means of appropriate counter system (in our case, High Purity Germanium detector). For this measurement totally 34 gold foils were irradiated at different reactor power levels and various positions in radial channel in aim to determine the neutron spatial distribution in radial channel. Interesting results about symmetry, value and dependence on reactor power level of neutron flux density were obtained.

scholarly journals CALCULATION OF NEUTRON FLUX DISTRIBUTION AT PIERCING BEAM PORTS OF PLATE TYPE RESEARCH REACTOR BANDUNG

2020 ◽  
Vol 21 (1) ◽  
pp. 25
Author(s):  
Epung Saepul Bahrum ◽  
Prasetyo Basuki ◽  
Alan Maulana ◽  
Jupiter Sitorus Pane
Keyword(s):  
Neutron Flux ◽  
Flux Distribution ◽  
Research Reactor ◽  
Neutron Flux Distribution ◽  
Plate Type
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