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.

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.

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.

Experimental Investigation of Lab-Scale, Heat Exchanger Prototypes Designed to Provide Refugia for Trout

2021 ◽  
Author(s):  
Rajib Uddin Rony ◽  
Adam Gladen ◽  
Sarah LaVallie ◽  
Jeremy Kientz
Keyword(s):  
Heat Exchanger ◽  
Temperature Difference ◽  
Large Scale ◽  
Cold Water ◽  
Cooling System ◽  
Tube Bundle ◽  
Small Scale ◽  
Scaled Model ◽  
Average Temperature ◽  
Average Temperature Difference

Abstract In recent years Spring Creek in South Dakota, a popular fishing location, has been experiencing higher surface water temperatures, which negatively impact cold-water trout species. One potential solution is to provide localized refugia of colder water produced via active cooling. The present work focuses on the design and testing of a small-scale prototype heat exchanger, for such a cooling system. Various prototypes of the heat exchanger were tested in a 1/10th-scaled model of a section of the creek. A staggered, tube-bundle heat exchanger was used. The prototypes consisted of just the heat exchanger placed directly in the scaled-stream model and of the heat exchanger placed inside an enclosure with an aperture. The results show that, without the enclosure, the average temperature difference is 0.64 °C, with a corresponding heat transfer requirement of 1.63 kW/°C of cooling. However, with an enclosure, the average temperature difference is 1.95 °C, which required 0.59 kW/°C of cooling. Modifications to the enclosure decrease the average temperature difference but also decrease the standard deviation of the temperature difference. Thus, the cooling effect is more evenly spread throughout the water in the enclosure. This indicates that the enclosure design can be used to balance the requirements of obtaining a desired temperature difference with a relatively low spatial variation in that temperature difference. These results will be used to guide the design of the large-scale heat exchanger prototype.

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.

Recapturing Net Positive Suction Head Margins in Boiling Water Reactor Emergency Core Cooling Systems

2017 ◽  
Author(s):  
Alan J. Bilanin ◽  
Andrew E. Kaufman ◽  
Warren J. Bilanin
Keyword(s):  
Cooling System ◽  
Cooling Water ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Cooling Systems ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Reactor Pressure

Boiling Water Reactor pressure suppression pools have stringent housekeeping requirements, as well as restrictions on amounts and types of insulation and debris that can be present in the containment, to guarantee that suction strainers that allow cooling water to be supplied to the reactor during a Loss of Coolant Accident remain operational. By introducing “good debris” into the cooling water, many of these requirements/restrictions can be relaxed without sacrificing operational readiness of the cooling system.

Effect of Coolant Water Temperature of ECCS on Failure Probability of RPV

2019 ◽  
Author(s):  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Kai Lu ◽  
Tadashi Watanabe ◽  
Yinsheng Li
Keyword(s):  
Water Temperature ◽  
Thermal Shock ◽  
Failure Probability ◽  
Structural Integrity ◽  
Cooling System ◽  
Pressure Vessels ◽  
Coolant Temperature ◽  
Pressurized Water ◽  
Coolant Water ◽  
Core Cooling

Abstract For reactor pressure vessels (RPVs) of pressurized water reactors, temperature of the coolant water in the emergency core cooling system (ECCS) may influence the structural integrity of the RPV during pressurized thermal shock (PTS) events. By focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. The analysis results show that the failure probability of RPV decreased dramatically when the coolant temperature in accumulator as well as in the high- and low-pressure injection systems (HPI/LPI) was increased, although the increase in coolant temperature in the HPI/LPI only did not lead to a decrease in the failure probability.

PRELIMINARY METHODOLOGY FOR THE ANALYSIS OF THE NATIONAL RESEARCH UNIVERSAL REACTOR USING INTEGRATED SEVERE ACCIDENT MODELLING CODES

2015 ◽  
Vol 4 (1) ◽  
pp. 53-65 ◽  
Author(s):  
A.C. Morreale ◽  
M.J. Brown ◽  
S.M. Petoukhov
Keyword(s):  
Large Scale ◽  
Aluminum Matrix ◽  
Heat Sinks ◽  
Severe Accident ◽  
Accident Analysis ◽  
Pressurized Water ◽  
Fuel Design ◽  
Water Reactors ◽  
Accident Case ◽  
Accident Modelling

The National Research Universal (NRU) Reactor is a multi-purpose research reactor located at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories. The severe accident case for the NRU has been explored through deterministic and probabilistic safety analysis (PSA) including multi-level PSAs that detail the progression and consequences of a severe accident in the NRU. These previous calculations lack the interconnected and comprehensive features of a full severe accident modelling code that is now the standard for severe accident analysis of power reactors. It was of interest within AECL to evaluate modern severe accident modelling codes to the NRU reactor case to enhance the understanding of accident progression and predict the system damage and radiation release consequences of a severe accident, which is a very low probability event. The NRU is smaller and operates at a lower power than the large scale power reactors (e.g., pressurized heavy water reactors, pressurized water reactors, and boiling water reactors) that these codes were designed to analyze. Additionally, the NRU has a unique design different from the power reactors and several features relevant to severe accidents including filtered venting, large passive heat sinks, and a dispersion fuel design of uranium-silicide in an aluminum matrix. The major severe accident analysis codes available to AECL and their applicability to the NRU are explored in this paper. In addition, a preliminary strategy for employing the most applicable codes to the NRU for the purposes of severe accident modelling is proposed.

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