Evaluation of the Scaling Analysis Model of the EU-APR1400 Core Catcher Cooling System Test Facility

2014 ◽  
Author(s):  
Bo W. Rhee ◽  
K. S. Ha ◽  
R. J. Park ◽  
J. H. Song
Keyword(s):  
Performance Test ◽  
Cooling System ◽  
Reactor Core ◽  
Test Facility ◽  
Scaling Analysis ◽  
Analysis Model ◽  
System Pressure ◽  
Test Loop ◽  
Core Catcher ◽  
The Eu

This paper describes the basic design features of the EU-APR1400 reactor core catcher cooling system and its test facility, and the associated scaling analysis model. An assessment of the validity of the scaling analysis using the preliminary performance test result of the test facility is described. This includes comparison of the predicted mass flow rate of the test loop as a function of the heat load to the facility, inlet flow subcooling and system pressure to the experimental results.

Design of 1/4 Scale Performance Test Facility for the PMR200 RCCS

2014 ◽  
Author(s):  
Y. Y. Bae ◽  
B. H. Cho ◽  
J. H. Kim ◽  
M. H. Kim ◽  
Y. W. Kim
Keyword(s):  
Performance Test ◽  
Cooling System ◽  
Test Facility ◽  
Maximum Temperature ◽  
Scaling Analysis ◽  
Reactor Safety ◽  
Key Technologies ◽  
Scale Test ◽  
Demonstration Plant ◽  
Cavity Cooling

One of the key technologies for the development of the PMR200, a VHTR demonstration plant, is a verification of the reactor cavity cooling system (RCCS) performance, which ensures reactor safety by passively removing heat from the reactor cavity. A preliminary numerical analysis of the RCCS showed that the maximum temperature in RCCS reached up to 700°C. Since radiation dominates the heat transfer at such a high temperature, it should be considered in both the design and associated numerical works for the test facility. For a verification of the RCCS performance, a 1/4 scale test facility has been constructed, and a performance test is being carried out. As the first step for the design of the test facility, a scaling analysis has been performed; and the ratio of variables between the model and prototype were determined. Numerical calculations using a CFD code were also performed to support the scaling analysis. It was confirmed that the scaling analysis was reasonably correct.

Vertical Tube Passive Condenser Under Secondary Pool Water Level Transient

2010 ◽  
Author(s):  
Shripad T. Revankar ◽  
Wenzhong Zhou
Keyword(s):  
Water Level ◽  
Cooling System ◽  
Vertical Tube ◽  
Test Facility ◽  
Flow Mode ◽  
Scaling Analysis ◽  
Water Level Change ◽  
Level Change ◽  
Pool Water ◽  
System Pressure

An experimental work was carried out on a passive containment cooling system (PCCS) test facility where the effect of PCCS pool water level change on the PCCS heat transfer characteristics was investigated. The specific design of condensing tube was based on scaling analysis from the PCCS design of Economic Simplified Boiling Water Reactor (ESBWR). The annulus between the primary condensing tube and the secondary boiling tube is filled with water and serves as water pool. During the test, steam generated in the pool is discharged through three steam exit nozzles located symmetrically at the top of the secondary boiling tube. Transient tests carried out with secondary pool water level change show that the system pressure for complete condensation mode increases with decrease in water level, however rate of condensation is almost constant. However, if the PCCS is operated in through flow mode the system pressure (primary side pressure) is constant, however, the condensate rate decreases indicating that some of the steam does not condense.

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.

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.

Westinghouse Advanced Loop Tester (WALT) Update

2008 ◽  
Author(s):  
G. Wang ◽  
W. A. Byers ◽  
M. Y. Young ◽  
Z. E. Karoutas
Keyword(s):  
Nuclear Power ◽  
Hot Spot ◽  
Test Facility ◽  
Hydraulic Test ◽  
Pressurized Water ◽  
System Pressure ◽  
Test Loop ◽  
Zinc Addition ◽  
Water Reactors ◽  
The Impact

In order to understand crud formation on the fuel rod cladding surfaces of pressurized water reactors (PWRs), a crud Thermal-Hydraulic test facility referred to as the Westinghouse Advanced Loop Tester (WALT) was built at the Westinghouse Science and Technology Department Laboratories in October 2005. Since then, a number of updates have been made and are described here. These updates include heater rod improvements, system pressure stabilization, and more effective protection systems. After these updates were made, the WALT system has been operated with higher stability and fewer failures. In this test loop, crud can be deposited on the heater rod surface and the character of the crud is similar to what has been observed in the PWRs. In addition, chemistry in the WALT loop can be varied to study its impact on crud morphology and associated parameters. The WALT loop has been successful in generating crud and measuring its thermal impact as a function of crud thickness. Currently, this test facility is supporting an Electric Power Research Institute (EPRI) program to assess the impact of zinc addition to PWR reactor coolant. Meanwhile, the WALT system is also being utilized by Westinghouse to perform dry-out and hot spot tests. These tests support the industry goal of 0 fuel failures by 2010 set by Institute of Nuclear Power Operations (INPO). Another major goal of the Westinghouse tests is to gain a better understanding of unexpected changes in core power distributions in operating reactors known as crud induced power shifts (CIPS) or axial offset anomalies (AOA).

Construction of a High Temperature Grade 91 Sodium Component Test Loop

2010 ◽  
Vol 654-656 ◽  
pp. 416-419
Author(s):  
Hyeong Yeon Lee ◽  
Jae Han Lee ◽  
Tae Ho Lee ◽  
Jae Hyuk Eoh Lee ◽  
Tae Joon Kim ◽  
...  
Keyword(s):  
Heat Exchanger ◽  
Fast Reactor ◽  
Large Scale ◽  
Performance Test ◽  
Heat Removal ◽  
Test Facility ◽  
Decay Heat ◽  
Test Loop ◽  
Intermediate Heat Exchanger ◽  
Grade 91

A large scale sodium test facility of ‘CPTL’(Component Performance Test Loop) for simulating thermal hydraulic behavior of the Korean demonstration fast reactor components such as IHX(Intermediate Heat Exchanger), DHX(Decay Heat Removal Heat Exchanger) and sodium pump under development by KAERI is to be constructed. The design temperature of this test loop is 600°C and design pressure is 1MPa. The three heat exchangers are made of Grade 91 steel. Another sodium test facility of the ‘STEF’(Sodium Thermal-Hydraulic Experimental Facility) will be constructed next to the CPTL facility to simulate the passive decay heat removal behavior in the sodium cooled fast reactor. In this paper, the overall facility features of the CPTL and STEF are introduced and preliminary conceptual design of the facilities are carried out.

Scaling of Passive Condenser System Separate Effect Facility

2008 ◽  
Author(s):  
Shripad T. Revankar ◽  
Seungmin Oh ◽  
Wenzhong Zhou ◽  
Gavin Henderson
Keyword(s):  
Cooling System ◽  
Forced Flow ◽  
Test Facility ◽  
Prototype System ◽  
Scaling Analysis ◽  
Reactor Accident ◽  
Cooling Period ◽  
Separate Effect ◽  
Suppression Pool

The Passive Containment Cooling System (PCCS) of the Simplified Boiling Water Reactor (SBWR) is a passive condenser system designed to remove energy from the containment for long term cooling period after a postulated reactor accident. Depending on pressure condition and noncondensable (NC) gas fraction in drywell (DW) and suppression pool (SP), three different modes are possible in the PCCS operation namely the forced flow, cyclic venting and complete condensation modes. The prototype SBWR has total of six condenser units with each units consist of hundreds of condenser tubes. Simulation of such prototype system is very expensive and complex Hence a scaling analysis is used in designing an experimental model for the prototype PCCS condenser system. The motive for scaling is to achieve a homologous relationship between an experiment and the prototype which it represents. A scaling method for separate effect test facility is first presented. The design of the scaled test facility for PCCS condenser is then given. Data from the test facility are presented and scaling approach to relate the scaled test facility data to prototype is discussed.

Modification to Shield Tank Overpressure Protection in a CANDU Reactor for Beyond Design Basis Event

2019 ◽  
Author(s):  
Michael Huang ◽  
Khurram Khan ◽  
Ali Etedali-Zadeh ◽  
Jefferson Tse ◽  
Bing Li
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Lower Probability ◽  
Damage State ◽  
Core Damage ◽  
Design Basis ◽  
System Pressure ◽  
Severe Event ◽  
Large Heat

Abstract The Shield Tank and End Shield Cooling System in the CANDU reactor contains a large volume of light water surrounding the Calandria and circulates water to remove heat that arises from the reactor core and Moderator. In a beyond design basis event that results in a severe event, progression in the absence of mitigating cooling actions could result in a large heat load being transferred to the water inside the shield tank from the calandria wall causing shield tank failure due to over pressurization. Following the 2011 events at Fukushima Daiichi Nuclear Power Plant, the adequacy of system pressure relief was assessed against severe events. Emergency mitigating equipment tie-ins for water make-up will likely limit the core damage state and prevent the need to protect the shield tank. However, Shield Tank Overpressure Protection (STOP) has been installed against severe event conditions pursuant to the CANDU defense-in-depth safety philosophy. A larger open vent line has been installed at some CANDU units on the top of the shield tank outside containment. This design routes the vent piping high enough to preclude any venting under any operational configuration and discharges back into the containment through an existing spare penetration. Vent piping is designed as Nuclear Class 2 in accordance with ASME BPVC Section III. Assessment of stresses in the modification piping was also completed for BDBEs including for a lower probability seismic event, steam venting and corresponding higher pressure and temperature conditions.

Experimental Study of Effects of Non-Condensable on Condensation in a Vertical Tube Bundle

2009 ◽  
Author(s):  
Wenzhong Zhou ◽  
Gavin Henderson ◽  
Shripad T. Revankar
Keyword(s):  
Heat Transfer ◽  
Cooling System ◽  
Tube Bundle ◽  
Condensation Heat Transfer ◽  
Forced Flow ◽  
Test Facility ◽  
Flow Mode ◽  
Scaling Analysis ◽  
Condenser Tube ◽  
Air Concentration

One of the engineered safety systems in the advanced boiling water reactor is a passive containment cooling system (PCCS) which is composed of a number of vertical heat exchanger. After a loss of coolant accident, the pressurized steam discharged from the reactor and the noncondensable (NC) gases mixture flows into the PCCS condenser tube. The PCCS condenser must be able to remove sufficient energy from the reactor containment to prevent containment from exceeding its design pressure. The efficient performance of the PCCS condenser is thus vital to the safety of the reactor. In PCCS condenser tube three flow conditions are expected namely the forced flow, cyclic venting and complete condensation modes. The PCCS test facility consists of steam generator (SG), instrumented condenser with secondary pool boiling section, condensation tank, suppression pool, storage tank, air supply line, and associated piping and instrumentation. The specific design of condensing tube is based on scaling analysis from the PCCS design of ESBWR. The scaled PCCS is made of four tubes of diameter 52.5mm and height 1.8 m arranged in square pitch. Steam air mixture condensation tests were carried out in a through flow mode of operation where the mixture flows through the condenser tube with some steam condensation. Data on condensation heat transfer were obtained for two nominal pressures, 225 kPa and 275 kPa and for air concentration fraction from 0 to 13%. Test results showed that with increase in pressure the condensation heat transfer increased. The presence of the air in the steam decreased the condensation heat transfer coefficient from 10 to 45% depending on air fraction in the steam.

scholarly journals KARAKTERISASI TERMOKOPEL TIPE K PADA FASILITAS SIMULASI SISTEM FASSIP-02

POROS ◽  
2021 ◽  
Vol 16 (2) ◽  
pp. 127
Author(s):  
Sumantri Hatmoko Hatmoko ◽  
Kussigit Santosa Santosa ◽  
Giarno Giarno Giarno ◽  
Dedy Haryanto Haryanto ◽  
Mulya Juarsa Juarsa ◽  
...  
Keyword(s):  
Computer Programming ◽  
Cooling System ◽  
Reactor Core ◽  
Comparison Method ◽  
Small Error ◽  
Test Facility ◽  
Fixed Temperature ◽  
Labview Software ◽  
Type K ◽  
National Instrument

In the activities of the Pratama Insinas, Ministry of Higher Education technology research in 2018, PTKRN BATAN built a testing facility that simulates a passive cooling system on the reactor core when there is a loss of outside power. The test facility is the Passive-02 System Simulation Facility (FASSIP02).In FASSIP-02 there are several parameters that need to be measured, one of which is temperature. In the measurement of temperature using a K type Thermocouple Connected to the National Instrument 9178 and 9213 modules that use computer programming with LabVIEW software. Temperature measurements need to be characterized.Characterization of type K thermocouples was carried out using thermobaths, 30 type K thermocouples, standard thermocouples,National Instrument modules 9178 and 9213 with computer programming displays using LabVIEW software. The method used for characterization oftype K thermocouples is a fixed temperature comparison method where the results of the temperature control of thermobath is 30-90 ͦC compared with the results of measurements from the type Kthermocouple and standard thermocouple. From the difference of the copper-wrapped junction tip thermocouple without the copper-wrapped and standard thermocouple produces a small error value, so the use of copper as a thermocouple junction end wrapper can be used as a temperature measurement FASSIP-02.

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