Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor

2018 ◽  
Author(s):  
Yuchuan Guo ◽  
Guanbo Wang ◽  
Dazhi Qian ◽  
Heng Yu ◽  
Bo Hu
Keyword(s):  
Fuel Assembly ◽  
System Analysis ◽  
Research Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Regulation System ◽  
Lower Power ◽  
Emergency Shutdown ◽  
Flow Blockage

The case of flow blockage of a single fuel assembly in the JRR-3 20MW open-pool-type research reactor is investigated without taking into account the effect of the power regulation system. The coolant system and multi-channel reactor core are modeled in detail using thermal hydraulic system analysis code RELAP5/MOD3.4. MDNBR (Minimum Departure From Nucleate Boiling Ratio) and the maximum fuel central temperature are investigated to assess the integrity of fuels. The fuel plates in blocked assembly are not damaged until the blockage ratio exceeds 70%. In addition, the mitigative effect of the assumed 18 MW lower power emergency shutdown operation on the accident is also discussed qualitatively. Results indicate that although the assumed lower power emergency shutdown operation cannot avoid the most severe operating condition, it can obviously mitigate the consequences of the accident. The reactor eventually remains in the long-term safe state when natural circulation is established.

Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

2005 ◽  
Vol 32 (15) ◽  
pp. 1679-1692 ◽  
Author(s):  
Martina Adorni ◽  
Anis Bousbia-Salah ◽  
Tewfik Hamidouche ◽  
Beniamino Di Maro ◽  
Franco Pierro ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Research Reactor ◽  
Reactor Core ◽  
Total Flow ◽  
Flow Blockage

scholarly journals Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

2015 ◽  
Vol 2015 ◽  
pp. 1-10 ◽  
Author(s):  
Patrícia A. L. Reis ◽  
Antonella L. Costa ◽  
Claubia Pereira ◽  
Maria Auxiliadora F. Veloso ◽  
Amir Z. Mesquita
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Reactor Core ◽  
Hydraulic Parameters ◽  
The Core ◽  
Steady State Operation ◽  
Transient Simulations ◽  
Abnormal Operation ◽  
Flow Blockage ◽  
Relap5 Code

Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.

Partial flow blockage analysis of the hottest fuel assembly in SNCLFR-100 reactor core

2018 ◽  
Vol 29 (1) ◽  
Author(s):  
Kang-Li Shi ◽  
Shu-Zhou Li ◽  
Xi-Lin Zhang ◽  
Peng-Cheng Zhao ◽  
Hong-Li Chen
Keyword(s):  
Fuel Assembly ◽  
Reactor Core ◽  
Flow Blockage

Analysis of a Total Flow Blockage of a Fuel Assembly in a Typical MTR Research Reactor by RELAP5/MOD3.3

2004 ◽  
Author(s):  
Franco Pierro ◽  
Beniamino Di Maro ◽  
Martina Adorni ◽  
Anis Bousbia Salah ◽  
Francesco D’Auria
Keyword(s):  
Fuel Assembly ◽  
Research Reactor ◽  
Reactor Power ◽  
Severe Accident ◽  
Total Flow ◽  
Flow Area ◽  
Forced Circulation ◽  
Flow Blockage ◽  
Pool Type ◽  
Extreme Scenario

The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool Type Research Reactor [1]. The reactor is core cooled and moderated by downward forced circulation of light water. The transient herein considered is the related to partial and total obstruction of a single Fuel Assembly (FA) cooling channel. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FA integrity. Two cases are analysed to emphasize the severity of the accident. The first one is a partial blockage of a single FA considering four different obstruction levels: 50%, 75%, 85% and 95% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FA. This study constitutes the first step of a larger work which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation.

Research on the Local Neutronics Thermal-Hydraulic Coupled Response of a PWR During Hypothetical Blockage Scenarios

2014 ◽  
Author(s):  
Xiaorong Li ◽  
Shinian Peng
Keyword(s):  
Fuel Assembly ◽  
Reactor Core ◽  
Feedback Effect ◽  
Thermal Hydraulics ◽  
Coupled Response ◽  
Flow Blockage

The phenomenon of the partial flow blockage of a fuel assembly in a reactor core is investigated with a coupled 3D neutronics/thermal-hydraulics code in order to account for the space reactivity feedback effect which is of great importance during hypothetical blockage scenarios. This paper identifies the neutronics thermal-hydraulics coupled response in the blocked assembly during the transient and analyzes the details of the phenomenon.

Thermo-Hydraulic Design of Westinghouse 4-Loop Reactor Core by COBRA-EN Code

2009 ◽  
Author(s):  
J. Jafari ◽  
B. Kalagar ◽  
E. Abdi Aghdam ◽  
F. D’Auria
Keyword(s):  
Heat Flux ◽  
Fuel Assembly ◽  
Nuclear Power ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Control Rod ◽  
Hydraulic Design ◽  
Fuel Assemblies

The Westinghouse 4-Loop PWR is a 3411MWth Nuclear Power Plant (NPP). The reactor core consists of 193 fuel assemblies within the core shroud. Each fuel assembly is arranged in 17×17 arrays and includes 264 fuel rods, 24 control rod guide tubes and one instrument tube. The objective of thermal and hydrodynamic design is to safely remove of the generated heat in the fuel without producing excessive fuel temperatures or steam void formations and without approaching the critical heat flux under steady-state operating conditions. This paper presents reactor core and fuel assembly modeling of the Westinghouse 4-Loop NPP using the thermo hydraulic subchannel analysis COBRA-EN code. The results of this modeling are compared with the VIPRE-01 thermal hydraulic code. The study involves the determination of the departure from nucleate boiling ratio (DNBR) in the hot channel of the reactor core, the temperature profiles, heat flux and pressure drop across the hottest channel of the hot assemblies. The obtained results shows that the good agreements are exist between the COBRA-EN and VIPRE-01 thermal hydraulic codes.

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