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 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.

scholarly journals Confocal chromatic sensor for displacement monitoring in research reactor

2021 ◽  
Vol 253 ◽  
pp. 04021
Author(s):  
Marion Agoyan ◽  
Gary Fourneau ◽  
Guy Cheymol ◽  
Ayoub Ladaci ◽  
Hicham Maskrot ◽  
...  
Keyword(s):  
Small Volume ◽  
Research Reactor ◽  
Reactor Core ◽  
Nuclear Research ◽  
Displacement Monitoring ◽  
Sensor Performance ◽  
Fuel Rod ◽  
Displacement Thickness ◽  
Nuclear Research Reactor ◽  
Mitigation Techniques

Confocal chromatic microscopy is an optical technique allowing measuring displacement, thickness, and roughness with a sub-micrometric precision. Its operation principle is based on a wavelength encoding of the object position. Historically, the company STIL based in the south of France has first developed this class of sensors in the 90’s. Of course, this sensor can only operate in a sufficiently transparent medium in the used spectral domain. It presents the advantage of being contactless, which is a crucial advantage for some applications such as the fuel rod displacement measurement in a nuclear research reactor core and in particular for cladding-swelling measurements. The extreme environmental conditions encountered in such experiments i.e. high temperature, high pressure, high radiations flux, strong vibrations, surrounding turbulent flow can affect the performances of this optical system. We then need to implement mitigation techniques to optimize the sensor performance in this specific environment. Another constraint concerns the small volume available in the irradiation rig next to the rod to monitor, implying the challenge to conceive a miniaturized sensor able to operate under these constraints.

Neutronic and thermal–hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core

2016 ◽  
Vol 116 ◽  
pp. 178-184 ◽  
Author(s):  
E. Chham ◽  
T. El Bardouni ◽  
K. Benaalilou ◽  
H. Boukhal ◽  
B. El Bakkari ◽  
...  
Keyword(s):  
Research Reactor ◽  
Reactor Core ◽  
Hydraulic Analysis ◽  
Triga Mark Ii ◽  
Thermal Hydraulic Analysis

scholarly journals Design of a FPGA-based controller for power and period measurement in the start range of Dalat Nuclear Research Reactor

2021 ◽  
Vol 10 (3) ◽  
pp. 41-48
Author(s):  
Vo Van Tai ◽  
Nguyen Van Kien ◽  
Nguyen Nhi Dien ◽  
Trinh Dinh Hai ◽  
Le Van Diep
Keyword(s):  
Neutron Flux ◽  
High Speed ◽  
Research Reactor ◽  
Full Range ◽  
Reactor Core ◽  
Nuclear Research ◽  
Reactor Power ◽  
Processing Unit ◽  
Upward Movement ◽  
Nuclear Research Reactor

This paper introduces a new controller module based on a high-speed field-programmable gate array (FPGA) and digital signal processing (DSP) using moving average (MA) filters for calculation of the reactor power and period at the start range of the Dalat nuclear research reactor (DNRR). The reactor power is proportional to the neutron flux in the reactor core, and the reactor period is the time that the reactor power changes by a factor of 2.718. In the control and protection system (CPS) of the DNRR, the reactor power and period have been monitored by the 8-bit microprocessor controller named BPM-107R. There are two main functions of the BPM-107R controller including 1) measurement and determination of reactor power and period and 2) generation of warning and emergency protection signals by reactor power or/and by reactor period. Those discrete signals will access to the logical processing unit of the CPS to prohibit the upward movement of control rods or to shut down the reactor. The CPS has three BPM-107R controllers corresponding to three independent neutron flux measurement equipment (NFME) channels working by logic voting “2 out-of 3”. Each NFME channel was designed for detection of neutron flux density in the full range from 1×100 to 1.2×1010 n/cm2 ×s, which is divided into two sub-ranges named start range (SR) and working range (WR). The designed FPGA-based controller module was tested using simulated signals as well as signals from the CPS in comparison with the original controller BPM-107R. The experimental results show that the characteristics and functions of the two controllers are equivalent.

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.

Development of Computer Program for Whole Core Thermal-Hydraulic Analysis of Fast Reactors

2002 ◽  
Author(s):  
Hiroyuki Ohshima ◽  
Masahiko Ohtaka
Keyword(s):  
Natural Circulation ◽  
Reactor Core ◽  
Heat Removal ◽  
Hydraulic Analysis ◽  
Reactor Operation ◽  
Fast Reactors ◽  
Operation Condition ◽  
Operation Conditions ◽  
Fuel Assemblies ◽  
Thermal Hydraulic Analysis

A whole core thermal-hydraulic analysis program ACT was developed for the purpose of evaluating detailed in-core thermal-hydraulic phenomena of sodium cooled fast reactors under various reactor operation conditions. ACT consists of four kinds of calculation modules, i.e., fuel-assembly, inter-wrapper gap (core barrel), upper plenum and heat transport system modules. The latter two modules give proper boundary conditions for the reactor core thermal-hydraulic analysis. These four modules are coupled with each other by using MPI and calculate simultaneously on a cluster workstation. ACT was applied to analyzing a sodium experiment performed at JNC, which simulated the natural circulation decay heat removal under PRACS and DRACS operation condition. In the experiment, not only inter-wrapper flows but also reverses flows in the fuel assemblies were observed. ACT succeeded in simulating such complicated phenomena.

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