Optimum Channel Inclination for Gas Venting Under Countercurrent Flow Limitations

2006 ◽  
Author(s):  
Y. Liao ◽  
K. Vierow
Keyword(s):  
Cooling System ◽  
Superficial Velocity ◽  
Countercurrent Flow ◽  
Pressurized Water ◽  
Inclination Angles ◽  
Surge Line ◽  
Water Reactors ◽  
Optimum Channel ◽  
Accident Conditions ◽  
Core Cooling

Countercurrent flow limitation (CCFL) in the pressurizer surge line of future Pressurized Water Reactors (PWR) with passive safety systems is an important phenomenon in reactor safety analysis. The pressurizer surge line is typically comprised of several sections with various inclination angles. Under certain accident conditions, countercurrent flow takes place in the surge line with liquid flowing down from the pressurizer and steam flowing up from the hot leg. The steam venting rate as well as the liquid draining rate may affect the Emergency Core Cooling System (ECCS) actuation. The objective herein is to develop a physics-based model for evaluating the effect of inclination angle on CCFL. For a given liquid superficial velocity in the countercurrent flow system of the pressurizer surge line, the gas superficial velocity should be as large as possible at the onset of flooding, so that the steam can vent as fast as possible without inhibiting the pressurizer drain rate. Thus the system could depressurize in a timely manner to initiate the ECCS actuation. As indicated by CCFL experiments, for a given liquid superficial velocity, the gas superficial velocity attains a greatest value at a certain channel inclination, which is defined as the optimum channel inclination. In the present work, an analytical model is proposed to predict the optimum channel inclination under simplified conditions. The model predictions compare favorably with experimental data.

Emergency Response Guidance for Reactor Vessel Pressurized Thermal Shock Events

1986 ◽  
Vol 108 (3) ◽  
pp. 346-351
Author(s):  
W. T. Kaiser ◽  
B. S. Monty
Keyword(s):  
Thermal Shock ◽  
Temperature Limit ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Plant Safety ◽  
Pressurized Thermal Shock ◽  
Water Reactors ◽  
Accident Conditions ◽  
Core Cooling ◽  
Emergency Operating

The operational concern of pressurized thermal shock (PTS) can be minimized by proper operator guidance. This paper presents a method for calculating a pressure temperature limit curve for reactor vessel integrity which can be used to identify an ongoing potential PTS event. This method has been developed for use and is applicable to all pressurized water reactors. The curve is used in emergency operating procedures developed to prioritize various plant safety concerns including PTS and core cooling to ensure proper operator action during accident conditions. This paper emphasizes the development of the pressure-temperature limit and how it is used within the emergency operating procedures.

scholarly journals Experimental investigation of PWR accident scenarios at the PKL test facility

2019 ◽  
Vol 137 ◽  
pp. 01016 ◽  
Author(s):  
Rafał Bryk ◽  
Lars Dennhardt ◽  
Simon Schollenberger
Keyword(s):  
Severe Accident ◽  
Test Facility ◽  
Pressurized Water ◽  
Water Reactors ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Core Cooling ◽  
Safety Systems ◽  
Reactor Pressure ◽  
Additional Loss

PKL is the only test facility in Europe that replicates the entire primary side and the most important parts of the secondary side of western-type Pressurized Water Reactors (PWR) in the scale of 1:1 in heights. It is also worldwide the only test facility with 4 identical reactor coolant loops arranged symmetrically around the Reactor Pressure Vessel (RPV) for simulation of nonsymmetrical boundary conditions between the reactor loops. Thermal-hydraulic phenomena observed in PWRs are simulated in the PKL test facility for over 40 years. The analyses carried out in these years encompass a large spectrum of accident scenario simulations and corresponding cool-down procedures. The overall goal of the PKL experiments is to show that under accident conditions - even for extreme and highly unlikely assumptions as additional loss of safety systems - the core cooling can be maintained or re-established by automatic or operator- performed procedures and that a severe accident e.g. a core melt-down can be avoided under all circumstances. Another goal of the tests performed in the PKL facility is the provision of data for validation of thermal-hydraulic system codes. This paper presents recent modifications of the PKL facility, applied in order to adapt the facility to the latest western-type designs currently built in the world. The paper discusses also important results obtained in the last years.

Use of Engineered Materials to Reduce Both Strainer Head Loss and Fiber Bypass for Emergency Core Cooling Systems

2019 ◽  
Author(s):  
Alan J. Bilanin ◽  
Andrew E. Kaufman ◽  
Warren J. Bilanin
Keyword(s):  
Large Scale ◽  
Cooling System ◽  
Head Loss ◽  
Small Scale ◽  
Pressurized Water ◽  
Cooling Systems ◽  
Large Scale Testing ◽  
Engineered Materials ◽  
Water Reactors ◽  
Core Cooling

Abstract Testing has shown that the use of engineered materials that can be combined with Loss of Cooling Accident generated debris has the ability to reduce debris head loss for boiling water and pressurized water reactors on Emergency Core Cooling System strainers. This engineered material has also been shown to reduce the amount of fiber that penetrates a strainer and continues downstream toward the fuel. Large scale testing is described that demonstrates that engineered materials can reach the strainers and reduce head loss. Small scale testing is described that demonstrates that engineered material can reduce the amount of fiber that can penetrate a strainer.

scholarly journals Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach

Nukleonika ◽  
2015 ◽  
Vol 60 (2) ◽  
pp. 339-345 ◽  
Author(s):  
Tomasz Bury
Keyword(s):  
Hydrogen Generation ◽  
Cooling System ◽  
Component Mixture ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Issues ◽  
Loss Of Coolant Accident ◽  
Lumped Parameter ◽  
Core Cooling ◽  
Parameter Approach

Abstract The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

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.

Break Size Effects on CET Response in an Upper Head SBLOCA Transient

2012 ◽  
Author(s):  
S. Gallardo ◽  
A. Querol ◽  
G. Verdú
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Upward Direction ◽  
Water Reactors ◽  
Accident Conditions

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced. This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS). The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.

Sensitivity Study of LOCA-Generated Debris on the Pressure Drop of Fuel Assembly

2017 ◽  
Author(s):  
Hammad Aslam Bhatti ◽  
Zhangpeng Guo ◽  
Weiqian Zhuo ◽  
Shahroze Ahmed ◽  
Da Wang ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Fuel Assembly ◽  
Cooling System ◽  
Safety Issue ◽  
Cooling Water ◽  
Sensitivity Study ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Core Cooling

The coolant of emergency core cooling system (ECCS), for long-term core cooling (LTCC), comes from the containment sump under the loss-of-coolant accident (LOCA). In the event of LOCA, within the containment of the pressurized water reactor (PWR), thermal insulation of piping and other materials in the vicinity of the break could be dislodged. A fraction of these dislodged insulation and other materials would be transported to the floor of the containment by coolant. Some of these debris might get through strainer and eventually accumulate over the suction sump screens of the emergency core cooling systems (ECCS). So, these debris like fibrous glass, fibrous wool, chemical precipitates and other particles cause pressure drop across the sump screen to increase, affecting the cooling water recirculation. As to address this safety issue, the downstream effect tests were performed over full-scale mock up fuel assembly. Sensitivity studies on pressure drop through LOCA-generated debris, deposited on fuel assembly, were performed to evaluate the effects of debris type and flowrate. Fibrous debris is the most crucial material in terms of causing pressure drop, with fibrous wool (FW) debris being more efficacious than fibrous glass (FG) debris.

Spherical steel containments of pressurized water reactors under accident conditions

1984 ◽  
Vol 82 (1) ◽  
pp. 77-87 ◽  
Author(s):  
R. Krieg ◽  
F. Eberle ◽  
B. Göller ◽  
W. Gulden ◽  
J. Kadlec ◽  
...  
Keyword(s):  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Water Reactors ◽  
Accident Conditions

A New Method to Weaken Thermal Stratification Phenomena in Pipes of Nuclear Power Plant

2013 ◽  
Author(s):  
Shengfei Wang ◽  
Yuxin Pang ◽  
Xiaojing Li ◽  
Dandan Fu ◽  
Yang Li ◽  
...  
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Thermal Stresses ◽  
Heat Pipes ◽  
New Method ◽  
Simple Experiment ◽  
Pressurized Water ◽  
Piping Systems ◽  
Surge Line ◽  
Water Reactors

Thermal stratification phenomena are observed in piping systems of pressurized water reactors, especially in the pressurizer surge line. As a result of the thermal stratification induced thermal stresses, fatigue problems can occur in the pipework. US NRC requirements have also identified flow stratification in surge lines as a phenomenon that must be considered in the design basis of surge lines. In this paper, a new method to reduce thermal stratification is proposed. As we all know, heat pipe is a simple device with no moving parts and can transfer large quantities of heat over fairly large distance. The new method is that using heat pipes to weaken the thermal stratification. In order to validate the new method, a simple experiment and theoretical analysis was taken. The results show that, the temperature difference of thermal stratification with heat pipes is smaller than the stratification without heat pipes. A design scheme was also given at the end of paper.

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