Integrated Chemical Effects Head Loss Experiments Using Multiconstituent Fibrous Debris Beds

2016 ◽  
Vol 3 (1) ◽  
Author(s):  
Daniel LaBrier ◽  
Amir Ali ◽  
Kerry J. Howe ◽  
Edward D. Blandford
Keyword(s):  
Cooling System ◽  
Product Formation ◽  
Head Loss ◽  
University Of New Mexico ◽  
Chemical Effects ◽  
Loss Of Coolant Accident ◽  
Chemical Conditions ◽  
Core Cooling ◽  
The University

The chemical head loss experiment (CHLE) program has been designed to acquire realistic material release and product formation in containment under post-loss of coolant accident (LOCA) conditions and their impact on the measured head loss through the use of modified debris beds developed at the University of New Mexico (UNM). A full-scale water chemistry test was conducted under Vogtle containment chemistry conditions to determine the release of these materials and the resulting head loss response of the formed products within the emergency core cooling system (ECCS) under prototypical chemical conditions. The test was designed to investigate material corrosion with the presence of excess aluminum and a nonprototypical temperature profile (80 °C for 120 h) to promote the production of aluminum precipitates. The head loss measured within the first 72 h of the test either surpassed the operational limits of the equipment or caused a failure within the system. The increase in head loss is not attributed to the formation of in situ precipitates but to a physical reaction of the epoxy used in constructing the debris beds to the local chemistry during the early stages of the test.

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 PARAMETRIC STUDY OF LOCA IN TRIGA-2000 USING RELAP5/SCDAP CODE

2017 ◽  
Vol 19 (2) ◽  
pp. 59 ◽  
Author(s):  
Anhar Riza Antariksawan ◽  
Surip Widodo ◽  
Hendro Tjahjono
Keyword(s):  
Cooling System ◽  
Fuel Temperature ◽  
Blow Down ◽  
Beam Tube ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Dry Out

A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered.  Keywords: safety analysis, LOCA, TRIGA, RELAP5 STUDI PARAMETRIK LOCA DI TRIGA-2000 MENGGUNAKAN RELAP5/SCDAP. Kecelakaan kehilangan air pendingin (LOCA) harus dianalisis untuk menjamin keselamatan suatu reaktor riset. Analisis LOCA dapat dilakukan menggunakan perhitungan best-estimate seperti RELAP5. Penelitian ini menekankan pada analisis LOCA di TRIGA-2000 akibat pecahnya pipa dan tabung berkas. Tujuan penelitian adalah memahami efek ukuran kebocoran dan waktu aktuasi sistem pendingin teras darurat (ECCS) pada sekuensi kejadian dan mengkaji keselamatan reaktor. Analisis dilakukan menggunakan program perhitungan RELAP/SCDAPSIM. Tiga ukuran kebocoran dan waktu aktuasi ECCS berbeda dipilih sebagai parameter dalam studi ini.  Hasil perhitungan mengonfirmasi bahwa semakin besar ukuran kebocoran, semakin cepat pengosongan tangki reaktor. Lubang siphon breaker dapat mencegah air terkuras dalam hal kebocoran pada pipa. Sedang dalam hal kebocoran pada beam tube, ECCS mampu memperlambat kenaikan temperatur bahan bakar. Dari studi ini dapat disimpulkan bahwa keselamatan reaktor dapat terjaga pada kejadian LOCA, namun pendinginan jangka panjang perlu dipertimbangkan untuk menjaga integritas bahan bakar.Kata kunci: analisis keselamatan, LOCA, TRIGA, RELAP5

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.

The Plant Feature and Performance of DMS (Double MS: Modular Simplified and Medium Small Reactor)

2008 ◽  
Author(s):  
Yukiko Kawabata ◽  
Masayoshi Matsuura ◽  
Shizuka Hirako ◽  
Takashi Hoshi
Keyword(s):  
Economic Efficiency ◽  
Cooling System ◽  
Natural Circulation ◽  
Circulation System ◽  
Surface Separation ◽  
Power Company ◽  
Loss Of Coolant Accident ◽  
Core Cover ◽  
And Performance ◽  
Core Cooling

The Japan Atomic Power Company has initiative in developing the DMS concept as a 400MWe-class light water reactor. The main features of the DMS relative to overcoming the scale demerit are the miniaturization and simplification of systems and equipment, integrated modulation of construction, standardization of equipment layouts and effective use of proven technology. The decrease in primary containment vessel (PCV) height is achieved by reducing the active fuel length of the DMS core, which is about two meters compared with 3.7 meters in the conventional BWR. The short active fuel length reduces the drop in core pressure, and overcomes the natural circulation system. And by using the lower steam velocity in the upper plenum in the reactor pressure vessel (RPV), we can adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator and thus helps minimize the RPV and PCV sizes. In order to improve safety efficiency, developing an Emergency Core Cooling System (ECCS) for the DMS was considered. The ECCS configuration in the DMS was examined to achieve core coverage and economic efficiency from the following. 1: Eliminating high-pressure injection systems. 2: Adopting passive safety-related systems. 3: Optimizing distribution for the systems and power source for the ECCS. In this way the configuration of the ECCS for the DMS was established, providing the same level of safety as the ABWR and the passive systems. Based on the results of Loss of Coolant Accident (LOCA) analysis, core cover can be achieved by this configuration. Therefore, the plant concept was found to offer both economic efficiency and safety.

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.

scholarly journals The sensitivity analysis for APR1400 nodalization under Large Break LOCA condition based on mars code

2017 ◽  
Vol 32 (1) ◽  
pp. 10-17
Author(s):  
Hyung-Wook Jang ◽  
Sang-Yong Lee ◽  
Seung-Jong Oh ◽  
Woong-Bae Kim
Keyword(s):  
Power Reactor ◽  
Cooling System ◽  
Cross Flow ◽  
Accident Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Flow Phenomena ◽  
Long Time ◽  
Core Cooling ◽  
D Model

The phenomena of loss of coolant accident have been investigated for long time and the result of experiment shows that the flow condition in the downcomer during the end-of-blowdown were highly multi-dimensional at full-scale. However, the downcomer nodalization of input deck for large break loss of coolant accident used in advanced power reactor 1400 analyses are made up with 1-D model and improperly designed to describe realistic coolant phenomena during loss of coolant accident analysis. In this paper, the authors modified the nodalization of MARS code LBLOCA input deck and performed LBLOCA analysis with new input deck. From original LBLOCA input deck file, the nodalization of downcomer and junction connections with 4 cold legs and direct vessel injection lines are modified for reflecting the realistic cross-flow effect and real downcomer structure. The analysis results show that the peak cladding temperature of new input deck decreases more rapidly than previous result and that the drop of peak cladding temperature was advanced by application of momentum flux term in cross-flow. Additionally, the authors developed a new input deck with multi-dimensional downcomer model and ran MARS code with multi-dimensional input deck as well. By using the modified input deck, the Emergency core cooling system by-pass flow phenomena is better characterized and found to be consistent with both experimental report and regulatory guide.

Investigation of Chemical Effects on Emergency Core Cooling Filtration Head Loss After Loss of Coolant Accident

2009 ◽  
Author(s):  
Jong Woon Park ◽  
Byung Gi Park ◽  
Chang Hyun Kim
Keyword(s):  
Calcium Silicate ◽  
Sodium Phosphate ◽  
Building Materials ◽  
Head Loss ◽  
Aluminum Surface ◽  
Loss Of Coolant Accident ◽  
Filter Screen ◽  
Loss Of Coolant ◽  
Reactor Building ◽  
Core Cooling

Integrated tests of head loss through an emergency core cooling filter screen are conducted, simulating reactor building environmental conditions for thirty days after a loss of coolant accident. A test apparatus with five individual loops each of whose chamber is established to test chemical product formation and measure the head loss through a sample filter. The screen area at each chamber is 78.54cm2 and reactor building materials can be scaled down according to specific plant condition. A series of tests have been performed to investigate the effects of reactor building spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the filter screen is strongly affected by spray duration and the head loss increase is rapid at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKON™ and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

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.

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