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

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.

Download Full-text

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

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA ◽

10.17146/tdm.2015.17.2.2317 ◽

2015 ◽

Vol 17 (2) ◽

pp. 67 ◽

Cited By ~ 5

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.

Download Full-text

Development of thermal hydraulic and margin analysis code for steady state forced and natural convective cooling of plate type fuel research reactors

Progress in Nuclear Energy ◽

10.1016/j.pnucene.2013.11.006 ◽

2014 ◽

Vol 71 ◽

pp. 39-51 ◽

Cited By ~ 17

Author(s):

Daeseong Jo ◽

Jonghark Park ◽

Heetaek Chae

Keyword(s):

Steady State ◽

Convective Cooling ◽

Research Reactors ◽

Margin Analysis ◽

Type Fuel ◽

Plate Type

Download Full-text

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

Science and Technology of Nuclear Installations ◽

10.1155/2018/9874196 ◽

2018 ◽

Vol 2018 ◽

pp. 1-17

Author(s):

Duvan A. Castellanos-Gonzalez ◽

João Manoel Losada Moreira ◽

José Rubens Maiorino ◽

Pedro Carajilescov

Keyword(s):

Steady State ◽

Pressure Drop ◽

Fuel Assembly ◽

Nuclear Reactors ◽

Nucleate Boiling ◽

Hydraulic Analysis ◽

Fuel Assemblies ◽

Thermal Hydraulic Analysis ◽

Type Fuel ◽

Plate Type

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.

Download Full-text

Studies on Production Planning of Dispersion Type U<sub>3</sub>Si<sub>2</sub>-Al Fuel in Plate-Type Fuel Elements for Nuclear Research Reactors

World Journal of Nuclear Science and Technology ◽

10.4236/wjnst.2016.64023 ◽

2016 ◽

Vol 06 (04) ◽

pp. 217-231

Author(s):

Miguel Luiz Miotto Negro ◽

Michelangelo Durazzo ◽

Marco Aurélio de Mesquita ◽

Elita Fontenele Urano de Carvalho ◽

Delvonei Alves de Andrade

Keyword(s):

Production Planning ◽

Nuclear Research ◽

Research Reactors ◽

Fuel Elements ◽

Type Fuel ◽

Plate Type

Download Full-text

Steady-State Safety Analysis of Ghana Research Reactor-1 With Low-Enriched-Uranium Core

Journal of Thermal Science and Engineering Applications ◽

10.1115/1.4046598 ◽

2020 ◽

Vol 12 (5) ◽

Author(s):

Edward Shitsi ◽

Prince Amoah ◽

Emmanuel Ampomah-Amoako ◽

Henry Cecil Odoi

Keyword(s):

Steady State ◽

Research Reactor ◽

Flow Instability ◽

Argonne National Laboratory ◽

Safety Analysis ◽

Nucleate Boiling ◽

Pressurized Water ◽

Safety Margins ◽

Water Reactors ◽

Enriched Uranium

Abstract Research reactors all over the world are expected to operate within certain safety margins just like pressurized water reactors and boiling water reactors. These safety margins mainly include onset of nucleate boiling ratio (ONBR), departure from nucleate boiling ratio (DNBR), and flow instability ratio (FIR) in addition to the maximum clad or fuel temperature and saturation temperature or boing point of the coolant inside the core of the reactor. This study carried out steady-state safety analysis of the Ghana Research Reactor-1 (GHARR-1) with low enriched uranium (LEU) core. Monte Carlo N-particle (MCNP) code was used to obtain radial and axial power peaking factors used as inputs in the preparation of the input file of plate temperature code of Argonne National Laboratory (PLTEMP/ANL code), which was then used to obtain the mentioned safety parameters of GHARR-1 with LEU core in this study. The data obtained on the ONBR were used to obtain the initiation of nucleate boiling boundary data with respect to the active length of the reactor core for various reactor powers. The obtained results for LEU core were also compared with that of the high enriched uranium (HEU) core. The results obtained show that the 34 kW GHARR-1 with LEU core is safe to operate just as the previous 30 kW HEU core was safe to operate.

Download Full-text

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

Journal of Nuclear Science and Technology ◽

10.1080/18811248.1998.9733966 ◽

1998 ◽

Vol 35 (12) ◽

pp. 943-951 ◽

Cited By ~ 11

Author(s):

Masanori KAMINAGA ◽

Kazuyoshi YAMAMOTO ◽

Yukio SUDO

Keyword(s):

Heat Flux ◽

Critical Heat Flux ◽

Research Reactors ◽

Type Fuel ◽

Plate Type

Download Full-text

ANALYSIS ON THE PERFORMANCE OF THE BANDUNG CONVERSION FUEL-PLATE TRIGA REACTOR IN STEADY STATE WITH CONSTANT COOLANT FLOW RATE

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA ◽

10.17146/tdm.2020.22.2.5843 ◽

2020 ◽

Vol 22 (2) ◽

pp. 41

Author(s):

Endiah Puji Hastuti ◽

Sudjatmi K. Alfa ◽

Sudarmono Sudarmono

Keyword(s):

Steady State ◽

Fuel Element ◽

Flow Rate ◽

Flow Instability ◽

Safety Margin ◽

Nucleate Boiling ◽

Coolant Flow ◽

Coolant Flow Rate ◽

Steady State Condition ◽

Triga Reactor

Bandung TRIGA2000 Reactor, a General Atomic (GA)-made research reactor used for training, research andiIsotope production, has been upgraded to operate at power of 2000 kW using TRIGA fuel rod type. Recently, the TRIGA reactor fuel element producers are going to discontinue the production of TRIGA fuel element. To overcome the unavailability of TRIGA fuel element, BATAN planned to modify TRIGA2000 fuel type from rod-type to U3Si2-Al plate-type fuel with 19.75% enrichment, similar to the domestically fabricated one used in RSG-GAS. The carried out design emphasized on the determination of operation condition limits for setting the reactor protection system in accordance to the reactor safety calculation results. The conceptual design of the innovative fuel plate TRIGA reactor cooling system is expected to remove heat generated by fuels with nominal power of 1 MW up to 2 MW. The design is developed through modelling and safety analysis using COOLOD-N2 validated code. The safety margin is set to its flow instability at transient condition of the fuel plate, which is ≥ 2.38; departure from nucleate boiling ratio ≥1.50; and no onset of nucleate boiling, ΔTONB ≥ 0oC. The primary coolant flow rate accommodating the existing Bandung TRIGA reactor capability is as high as 50 kg/s. The analysis results show that at power of 1 MW, the reactor can safely operate, while at power of 2 MW the safety margin is exceeded. In other words, the plate TRIGA reactor that employs forced convection mode operates safely at 1 MW with excess power 120% of its nominal power.Keywords: 1 MW, Thermalhydraulic design, Steady state condition, TRIGA plate, Constant flowrate

Download Full-text