Insulation Material Deposition and Distribution in a PWR Fuel Assembly Cluster

2012 ◽  
Author(s):  
Stefan Renger ◽  
Sören Alt ◽  
Wolfgang Kästner ◽  
André Seeliger ◽  
Frank Zacharias
Keyword(s):  
Fuel Assembly ◽  
Reactor Core ◽  
Visual Observation ◽  
Test Facility ◽  
Insulation Material ◽  
Control Rod ◽  
Free Jet ◽  
Laser Sensors ◽  
Material Deposition ◽  
Loss Of Coolant Accident

Background of experimental and methodical work is the loss of coolant accident (LOCA) with release of fibrous pipe insulation material. Latest investigations were focused on material deposition and distribution (cross mixing) in the reactor core. Therefore, a 2×2 PWR fuel assembly (FA) cluster was constructed. Four shortened PWR-FA-dummies are provided with separated in- and outlets. Every 16×16 fuel rod dummy consists of 20 control rod simulators, two spacers, FA-head and FA-bottom with a 3.5×3.5 mm integrated debris-screen filter (IDF). The cluster is encased in an acrylic housing for visual observation. It is connected with the test facility “Zittau Flow Tray” (ZFT), a simplified sump model, which allows inclusion and investigation of complex phenomena like material sedimentation in the sump and strainer blockages. A well mixing of air in the fluid was also considered by free jet expansions and flows through full cone-nozzles as well as marginal air entrainments. This Paper includes descriptions of applied measuring techniques (digital image processing, thrubeam laser sensors etc.) and an overview of all considered boundary conditions. Experimental results, aiming at the development, implementation and verification of multiphase flow and strainer models, are presented.

Test and Evaluation of a Filtering System for Retention of Fibres in a Coolant Flow After Loss-of-Coolant Accident

2012 ◽  
Author(s):  
T. Gocht ◽  
W. Kästner ◽  
A. Kratzsch ◽  
M. Strasser
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Test Facility ◽  
Experimental Investigations ◽  
Time Interval ◽  
Insulation Material ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

In case of an accident the safe heat removal from the reactor core with the installed emergency core cooling system (ECCS) is one of the main features in reactor safety. During a loss-of-coolant accident (LOCA) the release of insulation material fragments in the reactor containment can lead to malfunctions of ECCS. Therefore, the retention of particles by strainers or filtering systems in the ECCS is one of the major tasks. The aim of the presented experimental investigations was the evaluation of a filtering system for the retention of fiber-shaped particles in a fluid flow. The filtering system consists of a filter case with a special lamellar filter unit. The tests were carried out at a test facility with filtering units of different mesh sizes. Insulation material (mineral rock wool) was fragmented to fiber-shaped particles. To simulate the distribution of particle concentration at real plants with large volumes the material was divided into single portions and introduced into the loop with a defined time interval. Material was transported to the filter by the fluid and agglomerated there. The assessment of functionality of the filtering system was made by differential pressure between inlet and outlet of the filtering system and by mass of penetrated particles. It can be concluded that for the tested filtering system no penetration of insulation particles occurred.

Scaling Analysis of the Pressurizer Surge Line in Nuclear Plant Integral Test Facilities

2012 ◽  
Author(s):  
Cheng-Cheng Deng ◽  
Hua-Jian Chang ◽  
Ben-Ke Qin ◽  
Han Wang ◽  
Lian Chen
Keyword(s):  
Reactor Core ◽  
Nuclear Plant ◽  
Test Facility ◽  
Scaling Analysis ◽  
Two Phase ◽  
Integral Test ◽  
Loss Of Coolant Accident ◽  
Initial Injection ◽  
Surge Line ◽  
Core Cooling

During small break loss of coolant accident (SBLOCA) of AP1000 nuclear plant, the behavior of pressurizer surge line has an important effect on the operation of ADS valves and the initial injection of IRWST, which may happen at a time when the reactor core reaches its minimum inventory. Therefore, scaling analysis of the PRZ surge line in nuclear plant integral test facilities is important. Four scaling criteria of surge line are developed, which respectively focus on two-phase flow pattern transitions, counter-current flow limitation (CCFL) behavior, periodic draining and filling and maintaining system inventory. The relationship between the four scaling criteria is discussed and comparative analysis of various scaling results is performed for different length scale ratios of test facilities. The results show that CCFL phenomenon and periodic draining and filling behavior are relatively more important processes and the surge line diameter ratios obtained by the two processes’ scaling criteria are close to each other. Thus, an optimal scaling analysis considering both CCFL phenomenon and periodic draining and filling process of PRZ surge line is given, which is used to provide a practical reference to choose appropriate scale of the surge line for the ACME (Advanced Core-cooling Mechanism Experiment) test facility now being built in China.

Fast Reactor Core Seismic Analysis for Verification of Assessment Model of Control Rod

2018 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Shinichiro Matsubara ◽  
Kazuteru Kawamura ◽  
Hidenori Harada ◽  
Tomohiko Yamamoto
Keyword(s):  
Fuel Assembly ◽  
Seismic Analysis ◽  
Reactor Core ◽  
Assessment Model ◽  
Guide Tube ◽  
Analysis Method ◽  
Control Rod ◽  
Support Plate ◽  
The Core ◽  
Core Elements

The control rod guide tube self-stands on the core support plate. The control rod is inserted in the control rod guide tube, and the control rod hangs from the upper structure of the reactor. At scrum in case of an earthquake, the control rod is detached and it sits on the seating structure in the control rod guide tube (Fig.1). In a vertical earthquake, the control rod guide tube is raised from the core support plate, and the control rod is also raised from the control rod guide tube. Therefore, drawing out may arise. During the earthquake after scrum, the rising behavior is different from the other core elements because the control rod and the control rod guide tube rise interfering each other. The control rod guide tube is raised more easily than the fuel assembly by the vertical differential pressure of the core during operation, because the control rod guide tube is lighter than the fuel assembly. Therefore, it is necessary to restrain the rising of the control rod guide tube. The sleeve dashpot structure, in which a sleeve is attached on the upper surface of the receptacle tube, is employed. Moreover, the control rod guide tube is equipped with the control rod dashpot in order to restrain the rising displacement of the control rod. This paper summarizes the analysis method of the rising behavior of the single control rod guide tube and the rising behavior of the control rod and the control guide tube after the control rod is inserted.

Direct ECC Bypass Phenomena During LBLOCA Reflood Phase Observed in the MIDAS Test Facility

2002 ◽  
Author(s):  
B. J. Yun ◽  
T. S. Kwon ◽  
D. J. Euh ◽  
I. C. Chu ◽  
C.-H. Song ◽  
...  
Keyword(s):  
Reactor Core ◽  
Flow Distribution ◽  
Test Facility ◽  
Test Results ◽  
Paper Test ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Test ◽  
Core Cooling ◽  
Direct Bypass

One of the advanced design features of the APR-1400, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood phase of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is in progress. In this paper, test results of a direct ECC bypass performed in the steam-water test facility called MIDAS (Multi-dimensional Investigation in Downcomer Annulus Simulation) is presented. The test condition is determined, based on the preliminary analysis of TRAC code, by applying the ‘modified linear scaling method’ with the 1/4.93 length scale. From the tests, ECC direct bypass fraction, steam condensation rate and information on the flow distribution in the upper annulus downcomer region is obtained.

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.

Preliminary Design of Unloading Device for MNSR LEU Conversion

2018 ◽  
Author(s):  
Hao Qian ◽  
Li Yiguo ◽  
Peng Dan ◽  
Wu Xiaobo ◽  
Lu Jin ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Reactor Vessel ◽  
Cover Plate ◽  
Spent Fuel ◽  
Long Distance ◽  
Control Rod ◽  
Small Cover ◽  
Sealing Performance ◽  
Safety And Reliability ◽  
Application Requirements

In order to solve the problem that the current unloading operation will destroy the sealing performance of Miniature Neutron Source Reactor (MNSR) reactor vessel and the tightness can’t be restored, and to meet the application requirements that the original reactor vessel will be reloaded and operated after MNSR LEU conversion, the new unloading device is designed, which can be used without separation of reactor vessel. There has only one fuel assembly in MNSR. When the fuel assembly are unload for MNSR LEU conversion, the cover plate of the pool is removed, the cadmium string is put in, and the neutron detector is placed at first. After removing the drive mechanism and the control rod, and opening the small cover plate at the top of reactor vessel, the fuel assembly can be grabbed and unloaded by unloading tool only through the opening of the small top cover plate. The MNSR spent fuel has very high radioactivity. The auxiliary mechanical device can be used with unloading tools to realize operation in a long distance by lifting and level motion, which is convenient to shield and can reduce the works’ irradiation dose level effectively. Through calculation and analysis, the results show that the structure strength of unloading device is much larger than the actual load to ensure operation safety and reliability. The unloading device is easy to process and operate, and can be used in the practical operation of MNSR LEU conversion or decommissioning at home and abroad to simplify the operation steps and improve the working efficiency.

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.

scholarly journals Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

scholarly journals Hydrodynamic Modeling of Mineral Wool Fiber Suspensions in a Two-Dimensional Flow

2009 ◽  
Author(s):  
Gregory M. Cartland Glover ◽  
Alexander Grahn ◽  
Eckhard Krepper ◽  
Frank-Peter Weiss ◽  
So¨ren Alt ◽  
...  
Keyword(s):  
Numerical Models ◽  
Mineral Wool ◽  
Wool Fiber ◽  
Turbulent Dispersion ◽  
Insulation Material ◽  
Fiber Suspensions ◽  
Loss Of Coolant Accident ◽  
Wool Fibers ◽  
Two Dimensional Flow ◽  
Mineral Wool Fiber

A consequence of a loss of coolant accident is that the local insulation material is damaged and maybe transported to the containment sump where it can penetrate and/or block the sump strainers. An experimental and theoretical study, which examines the transport of mineral wool fibers via single and multi-effect experiments is being performed. This paper focuses on the experiments and simulations performed for validation of numerical models of sedimentation and resuspension of mineral wool fiber agglomerates in a racetrack type channel. Three velocity conditions are used to test the response of two dispersed phase fiber agglomerates to two drag correlations and to two turbulent dispersion coefficients. The Eulerian multiphase flow model is applied with either one or two dispersed phases.

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