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.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient

2006 ◽  
Author(s):  
A. Martin ◽  
S. Benhamadouche ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Fluid Temperature ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA (Small Break Loss of Coolant Accident) transients. This paper presents the Research and Development program started at E.D.F on the Computational Fluid Dynamic (CFD) determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermal-hydraulic tool Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local Thermal-hydraulic analysis of a Small Break Loss of Coolant Accident transient, the paper presents mainly a parametric study which helps to understand the main phenomena that can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global Thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. The main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margin factors.

Small Break LOCA Analysis of ACR-700 NPP

2006 ◽  
Author(s):  
Limin Zheng ◽  
Sen Shen ◽  
David Wright
Keyword(s):  
Nuclear Power ◽  
Sensitivity Study ◽  
Hydraulic Analysis ◽  
Reactor Outlet ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Subsequent Loss ◽  
Safety Systems ◽  
Limiting Case ◽  
Thermal Hydraulic Analysis

A small break loss of coolant accident (SB-LOCA) analysis to assess a preliminary conceptual design of the ACR-700 PHWR nuclear power plant (NPP) developed by AECL has been performed with CATHENA MOD 3.5d, a PHWR system thermal-hydraulic analysis code. The limiting break size has been found by performing a sensitivity study for three different break locations [i.e. reactor inlet header (RIH), HTS pump suction (PS) pipe and reactor outlet head (ROH)] under the limiting case (i.e. SB-LOCA with subsequent loss of class IV power with all safety systems available). The analysis results indicate that the SB-LOCA acceptance criteria are satisfied.

CFD-Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Temperature Safety Injection and Fluid-Structure Thermal Coupling

2005 ◽  
Author(s):  
A. Martin ◽  
D. Monfort ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Fluid Temperature ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

Integrity evaluation methods for nuclear Reactor Pressure Vessels (RPVs) under Pressurised Thermal Shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during SBLOCA (Small Break Loss of Coolant Accident) transients. This paper presents the Research and Development program started at E.D.F on the Computational Fluid Dynamic (CFD) determination of the cooling phenomena of a PWR vessel during a Pressurised Thermal Shock. The numerical results are obtained with the thermal-hydraulic tool Code_Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local Thermal-hydraulic analysis of a Small Break Loss of Coolant Accident transient, the paper presents mainly a parametric study whose purpose is to understand the main phenomena which can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global Thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. On the whole, the main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the down comer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation which will subsequently assess the associated RPV safety margin factors.

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.

CFD Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient

2011 ◽  
Vol 133 (3) ◽  
Author(s):  
A. Martin ◽  
S. Benhamadouche ◽  
G. Bezdikian ◽  
F. Beaud ◽  
F. Lestang
Keyword(s):  
Heat Transfer ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Hydraulic Analysis ◽  
Loss Of Coolant Accident ◽  
Pressurized Thermal Shock ◽  
Loss Of Coolant ◽  
Thermal Hydraulic Analysis ◽  
Reactor Pressure

Integrity evaluation methods for nuclear reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) loading are applied by French Utility. They are based on the analysis of the behavior of relatively shallow cracks under loading PTS conditions due to the emergency cooling during small break loss of coolant accident (SBLOCA) transients. This paper presents the research and development program started at EDF on the computational fluid dynamics (CFD) determination of the cooling phenomena of a PWR vessel during a pressurized thermal shock. The numerical results are obtained with the thermal-hydraulic tool Code̱Saturne, in combination with the thermal-solid code SYRTHES to take into account the coupled effect of heat transfer between the fluid flow and the vessel. Based on the global and local thermal-hydraulic analysis of a small break loss of coolant accident transient, this paper presents mainly a parametric study that helps to understand the main phenomena that can lead to better estimating the margin factors. The geometry studied represents a third of a PWR pressure vessel, and the configuration investigated is related to the injection of cold water in the vessel during a SBLOCA transient. Conservative initial and boundary conditions for the CFD calculation are derived from the global thermal-hydraulic analysis. Both the fluid behavior and its impact on the solid part formed by cladding and base metal are considered. The main purpose of the numerical thermal-hydraulic studies is to accurately estimate the distribution of fluid temperature in the downcomer and the heat transfer coefficients on the inner RPV surface for a fracture mechanics computation, which will subsequently assess the associated RPV safety margin factors.

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