Thermo-Hydraulic Calculation and Safety Analysis of the Tehran Research Reactor

2009 ◽  
Author(s):  
J. Jafari ◽  
M. K. Firuzjaee ◽  
S. Khakshournia ◽  
K. Sepanloo ◽  
F. D’Auria
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Coolant Flow ◽  
Severe Accident ◽  
Hydraulic Calculation ◽  
Pump Failure ◽  
Tehran Research Reactor ◽  
Safety Parameters

The aim of this study is to evaluate the thermal hydraulic parameters such as, Minimum DNBR, fuel, clad and coolant temperatures of the Tehran Research Reactor (TRR) core in order to assure that the thermal limitations are not exceeded at full power and steady-state and the loss-of-flow accident (LOFA) operating conditions and the existing design is adequate to assure that the consequences from this anticipated occurrence does not lead to a severe accident. The PARET and TRANS computer codes were used to analysis thermal-hydraulic and safety parameters under study steady-state conditions. Furthermore the LOFA as a consequence of two type of transients are analyzed; one related to main cooling pump failure accident (slow LOFA), and the other related to accidentally opening the flapper valve whereas the main cooling pump is in working (fast LOFA). The steady-state analysis involves the determination of the departure from nucleate boiling ratio (DNBR), the temperatures profile, heat flux across the hottest channel and etc. The LOFA analysis involves the power and coolant flow rate, the fuel centerline, clad surface and coolant outlet temperatures, and the occurring of coolant flow reversal.

scholarly journals Neutronic analysis of control rod effect on safety parameters in Tehran Research Reactor

2018 ◽  
Vol 50 (7) ◽  
pp. 1017-1023 ◽  
Author(s):  
Mina Torabi ◽  
A. Lashkari ◽  
Seyed Farhad Masoudi ◽  
Somayeh Bagheri
Keyword(s):  
Research Reactor ◽  
Control Rod ◽  
Tehran Research Reactor ◽  
Safety Parameters

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

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.

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

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

scholarly journals Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

2011 ◽  
Vol 26 (1) ◽  
pp. 45-49 ◽  
Author(s):  
Atta Muhammad ◽  
Masood Iqbal ◽  
Tayyab Mahmood
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Safety Margin ◽  
Nucleate Boiling ◽  
Hydraulic Analysis ◽  
The Core ◽  
Computer Codes ◽  
Neutronic Parameters ◽  
Thermal Hydraulic Analysis

The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

Neutronic Analysis for Tehran Research Reactor Upgrading

2002 ◽  
Author(s):  
Ebrahim Afshar ◽  
Alireza Shahidi
Keyword(s):  
Research Reactor ◽  
Three Dimensional ◽  
Operating Conditions ◽  
Monte Carlo Code ◽  
Radioisotope Production ◽  
The Core ◽  
Transport Code ◽  
Tehran Research Reactor ◽  
Fuel Elements ◽  
Good Agreement

This paper presents preliminary results of a study undertaken to investigate the possibility of raising the power of Tehran Research Reactor (TRR) from 5 to 10 MW (th), keeping the same core configuration and with minimum changes in the primary cooling circuit. The main aim of TRR upgrade is to increase the volume of radioisotope production. The neutronic analysis was carried out for a fresh core with 22 Standard Fuel Elements (SFE) under normal operating conditions. Two different calculational lines were used to simulate the neutronic behavior in the core and perform the necessary neutronic calculations. First, combination of cell calculation transport code WIMS-D/4 [1] and three dimensional core calculation diffusion code CITATION [2] were used to and next a Monte Carlo code MCNP-4B [3] together with point depletion code ORIGEN-2 [4] were used. The results obtained show good agreement between these two different schemes.

scholarly journals Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

2017 ◽  
Vol 67 (1) ◽  
pp. 69-76
Author(s):  
Jakub Jakubec ◽  
Juraj Paulech ◽  
Vladimír Kutiš ◽  
Gabriel Gálik
Keyword(s):  
Steady State ◽  
Temperature Distribution ◽  
Pressure Drop ◽  
Fuel Assembly ◽  
Nuclear Reactor ◽  
Software Tool ◽  
Operating Conditions ◽  
Coolant Flow ◽  
Cfd Modelling ◽  
Ansys Cfx

AbstractThe paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

Inspection of domestic nuclear fuel rods using neutron radiography at the Tehran Research Reactor

2016 ◽  
Vol 58 (9) ◽  
pp. 763-766 ◽  
Author(s):  
Mohammad Hosein Choopan Dastjerdi ◽  
Hossein Khalafi ◽  
Yaser Kasesaz ◽  
Amir Movafeghi
Keyword(s):  
Nuclear Fuel ◽  
Research Reactor ◽  
Neutron Radiography ◽  
Fuel Rods ◽  
Tehran Research Reactor ◽  
Nuclear Fuel Rods
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