Thermal-hydraulic analysis for the upgraded MIT Nuclear Research Reactor

1998 ◽  
Vol 45 (3) ◽  
pp. 1040-1044
Author(s):  
Lin Wen Hu ◽  
J.A. Bernard
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Hydraulic Analysis ◽  
Nuclear Research Reactor ◽  
Thermal Hydraulic Analysis

scholarly journals Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

2021 ◽  
Vol 2021 ◽  
pp. 1-10
Author(s):  
Kien-Cuong Nguyen ◽  
Vinh-Vinh Le ◽  
Ton-Nghiem Huynh ◽  
Ba-Vien Luong ◽  
Nhi-Dien Nguyen
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Nucleate Boiling ◽  
Fuel Cladding ◽  
Hydraulic Analysis ◽  
The Core ◽  
Nuclear Research Reactor ◽  
Calculation Results ◽  
Enriched Uranium ◽  
Thermal Hydraulic Analysis

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Thermal-hydraulic analysis of a research reactor on personal computer with TARR code

1992 ◽  
Vol 14 (3) ◽  
pp. 1-5
Author(s):  
Ngo Huy Can ◽  
Nguyen Manh Lan ◽  
Tran Van Tran
Keyword(s):  
Personal Computer ◽  
Research Reactor ◽  
Reactor Core ◽  
Nuclear Research ◽  
Stationary Regime ◽  
Hydraulic Calculation ◽  
Hydraulic Analysis ◽  
Fuel Assemblies ◽  
Nuclear Research Reactor ◽  
Thermal Hydraulic Analysis

The code has been created for thermal-hydraulic calculation of stationary regime of nuclear research reactor, using personal computer. The main objective of the code is to compute the thermal parameters in the reactor core in order to avoid any accident. The code can be applied for many fuel assemblies available in research reactors.

scholarly journals ANALYSIS OF AN EXTREME LOSS OF COOLANT IN THE IPR-R1 TRIGA REACTOR USING A RELAP5 MODEL

2013 ◽  
Vol 12 (2) ◽  
pp. 46
Author(s):  
P. A. L. Reis ◽  
A. L. Costa ◽  
C. Pereira ◽  
M. A. F. Veloso ◽  
H. V. Soares ◽  
...  
Keyword(s):  
Research Reactor ◽  
Technology Development ◽  
Nuclear Research ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Transient Case ◽  
Fuel Elements ◽  
Thermal Hydraulic Analysis ◽  
Pool Type

The RELAP5/MOD3.3 code has been applied for thermal hydraulic analysis of power reactors as well as nuclear research reactors with good predictions. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA have been validated for steady state and transient situations. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. In this work, an extreme transient case of loss of coolant accident (LOCA) has been simulated. For this type of analysis, the automatic scram of the reactor was not considered because the main aim was to verify the evolution of the fuel elements heating in the absence of coolant. The temperature evolutions are presented as well as an analysis about the temperature safety limits.

scholarly journals MCNP model of L-54M nuclear research reactor: validation by preliminary graphite radiological characterization

2019 ◽  
Vol 322 (3) ◽  
pp. 1341-1350
Author(s):  
Eros Mossini ◽  
Luca Codispoti ◽  
Giorgio Parma ◽  
Filippo Maria Rossi ◽  
Elena Macerata ◽  
...  
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Nuclear Research Reactor ◽  
Radiological Characterization

Thermal-Hydraulic Analysis of the 3-MW TRIGA MARK-II Research Reactor under Steady-State and Transient Conditions

2001 ◽  
Vol 135 (1) ◽  
pp. 51-66 ◽  
Author(s):  
M. Q. Huda ◽  
S. I. Bhuiyan ◽  
T. K. Chakrobortty ◽  
M. M. Sarker ◽  
M. A. W. Mondal
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Hydraulic Analysis ◽  
Transient Conditions ◽  
Triga Mark Ii ◽  
Thermal Hydraulic Analysis
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