Integral Effect Test for Performance Evaluation of the PAFS (Passive Auxiliary Feedwater System) During a SBO (Station Blackout) Transient

2014 ◽  
Author(s):  
Byoung-Uhn Bae ◽  
Seok Kim ◽  
Yu-Sun Park ◽  
Yun-Je Cho ◽  
Kyoung-Ho Kang
Keyword(s):  
Steam Generator ◽  
Flow Rate ◽  
Water Levels ◽  
Experimental Result ◽  
Test Facility ◽  
Decay Heat ◽  
The Core ◽  
Integral Effect ◽  
Station Blackout ◽  
Effect Test

Station blackout (SBO) accident is considered as one of the most significant design extension conditions (DECs), which has been extensively focused after the Fukushima Dai-chi accident. When the SBO accident occurs in the APR+ (Advance Power Reactor Plus), the PAFS (Passive Auxiliary Feedwater System), which is an advanced safety feature adopted in the APR+, should play a significant role to cool down the core decay heat without any operation of active safety systems. This study focuses on validation of the cooling and operational performance for the PAFS during the SBO transient with utilizing an integral effect test facility, ATLAS-PAFS. In order to simulate the SBO transient of the APR+ as realistically as possible, a pertinent scaling approach was taken into account. The initial steady-state conditions and the sequence of event in the SBO scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. In the transient simulation, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the break flow rate, and the condensate flow rate at the return-water line were measured and investigated. Following the reactor trip at the initiation of the transient, the coolant inventory of the secondary system of the steam generator was reduced by the repeated opening and closing of the MSSV. When the collapsed water level reached 25% of wide range, the PAFS was actuated to cool down the primary system by the condensation heat transfer at the PCHX (Passive Condensation Heat Exchanger). The pressure and the temperature of the reactor coolant system continuously decreased during the heat removal by the PAFS operation. It points out that the PAFS can supply auxiliary feedwater to the steam generator and remove the core decay heat without any active system. From the present experimental result, it could be concluded that the APR+ has the capability of coping with the hypothetical SBO scenario with adopting the PAFS and proper set-points of its operation. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS, RELAP5 as well as the SPACE code and to identify any code deficiency for a SBO simulation with an operation of the PAFS.

Integral Effect Test on Operational Performance of the PAFS (Passive Auxiliary Feedwater System) for a SGTR (Steam Generator Tube Rupture) Accident

2013 ◽  
Author(s):  
Yusun Park ◽  
Byoung-Uhn Bae ◽  
Seok Kim ◽  
Yun-Je Cho ◽  
Kyoung-Ho Kang
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Operational Performance ◽  
Experimental Result ◽  
Test Facility ◽  
Experimental Program ◽  
Primary System ◽  
Steam Generator Tube ◽  
Integral Effect ◽  
Effect Test

The PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace a conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism; i.e., condensing steam in nearly-horizontal U-tubes submerged inside the PCCT (Passive Condensation Cooling Tank). With an aim of verifying the operational performance of the PAFS, the experimental program of an integral effect test is in progress at KAERI (Korea Atomic Energy Research Institute). The test facility, ATLAS-PAFS was constructed to experimentally investigate the thermal hydraulic behavior in the primary and secondary systems of the APR+ during a transient when the PAFS is actuated. Since the ATLAS-PAFS facility simulates a single train of the PAFS, the anticipated accident scenarios in the experiment include FLB (Feedwater Line Break), MSLB (Main Steam Line Break), and SGTR (Steam Generator Tube Rupture). Among them, SGTR was considered as one of the design basis accidents having a significant impact on safety in a viewpoint of radiological release. Therefore, the SGTR test was determined to be the integral effect test item in the frame of the ATLAS-PAFS experimental program. In this study, the PAFS-SGTR-HL-02 test was performed to simulate a double-ended rupture of a single U-tube in the hot side of the steam generator of the APR+. The three-level scaling methodology was taken into account to determine the test conditions of the steady-state and the transient. The pressures and temperatures of the system and the data related to the PAFS operation were collected with the measurement of the break flow. The initial steady-state conditions and the sequence of event of SGTR scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. And it was shown that the pressure and the temperature of the primary system were continuously decreased during the heat removal by the PAFS operation. The water pool in the PCCT was heated up to the saturation condition and the evaporation of the water made a decrease of the PCCT water level. It could be concluded from the present experimental result that the APR+ has the capability of coping with the hypothetical SGTR scenario with adopting the PAFS and the proper set-points of its operation.

Experimental Study on the Safety Injection Pump (SIP) Failure Accompanied by the Steam Generator Tube Rupture (SGTR)

2018 ◽  
Author(s):  
Yusun Park ◽  
Byoung Uhn Bae ◽  
Jongrok Kim ◽  
Jae Bong Lee ◽  
Hae Min Park ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Water Levels ◽  
Experimental Result ◽  
Hydraulic Test ◽  
Steam Generator Tube ◽  
Injection Pump ◽  
Management Procedure ◽  
Test Loop ◽  
Integral Effect

The integral effect test to simulate the safety injection pump (SIP) failure accompanied by a steam generator tube rupture (SGTR), named a SGTR-SIP-01 test, was performed to investigate the thermal hydraulic phenomena during a multiple failure accident. In this study, a thermal-hydraulic integral effect facility, ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) was utilized to simulate thermal hydraulic phenomenon which can be occurred in the nuclear power plant, as realistically as possible. In this SGTR-SIP-01 test, a rupture of five steam generator u-tubes on the steam generator hot-side was simulated. Due to the initiation of SGTR, a reactor was tripped by high steam generator level (HSGL) signal. During the transient simulation for SGTR-SIP-01 test, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the flows in the primary loops, and the fluid temperatures, were measured and analyzed. Through this experimental result, insights about the accident management procedure can be provided in the case of the multiple failure accident, such as a SGTR accident with a total failure of SIPs. In addition to that, for improvement of the system code which are now on developing such SPACE or MARS-KS code, this test data can be utilized for validation and verification work.

Integral effect test on station blackout (SBO) scenario with steam generator tube rupture (SGTR) in ATLAS facility

2018 ◽  
Vol 328 ◽  
pp. 107-114 ◽  
Author(s):  
Byoung-Uhn Bae ◽  
Yu-Sun Park ◽  
Jong-Rok Kim ◽  
Kyoung-Ho Kang ◽  
Ki-Yong Choi
Keyword(s):  
Steam Generator ◽  
Steam Generator Tube ◽  
Integral Effect ◽  
Station Blackout ◽  
Effect Test

System-Core Thermal Hydraulic Interaction During a Reflood Phase of a Cold-Leg LBLOCA in the Atlas Integral Effect Test Facility

2009 ◽  
Author(s):  
Seok Cho ◽  
Hyun-Sik Park ◽  
Ki-Yong Choi ◽  
Kyong-Ho Kang ◽  
Yeon-Sik Kim ◽  
...  
Keyword(s):  
Experimental Data ◽  
Boundary Condition ◽  
Water Flow Rate ◽  
Experimental Tests ◽  
Test Facility ◽  
Hydraulic Behavior ◽  
The Core ◽  
Analysis Methodology ◽  
Integral Effect ◽  
Effect Test

A series of experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed by using the ATLAS facility. This paper describes related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of cold-leg LBLOCA condition. The present descriptions will be focus on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed by using the ATLAS. The LB-CL-09 is the integral effect test with conservative boundary condition, and LB-CL-11, and -14 are the integral effect tests with realistic boundary condition, and the LB-CL-15 is the separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase, and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.

Experimental Study on the Steam Line Break (SLB) Accident With the Steam Generator Tube Rupture (SGTR)

2018 ◽  
Author(s):  
Yusun Park ◽  
Byoung Uhn Bae ◽  
Jongrok Kim ◽  
Jae Bong Lee ◽  
Hae Min Park ◽  
...  
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Water Levels ◽  
Experimental Result ◽  
Secondary System ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Management Procedure ◽  
Test Loop ◽  
Integral Effect

The integral effect test to simulate a steam line break (SLB) accident accompanied by a steam generator tube rupture (SGTR), named SLB-SGTR-02 test, was performed to investigate the thermal hydraulic phenomena during a multiple failure accident. In this study, a thermal-hydraulic integral effect facility, ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) was utilized to simulate thermal hydraulic phenomena which can occur in the nuclear power plant, especially for the pressurized water reactor, as realistically as possible. The SLB is simulated as a guillotine break on the main steam line and, in the SGTR, five steam generator tubes on the steam generator hot-side are modeled. Due to the initiation of SLB, a reactor was tripped by low steam generator pressure (LSGP) signal and the SGTR was initiated when the water inventory in the secondary system of the affected steam generator was dried-out. During the transient simulation for SLB-SGTR-02 test, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the flows in the primary loops and the fluid temperatures, were measured and analyzed. Through this experimental result, insights about the accident management procedure can be provided in the case of the multiple failure accident, such as an SLB with a SGTR.

Development of a pump performance model for an integral effect test facility

2008 ◽  
Vol 238 (10) ◽  
pp. 2614-2623 ◽  
Author(s):  
Ki-Yong Choi ◽  
Yeon-Sik Kim ◽  
Sung-Jae Yi ◽  
Won-Pil Baek
Keyword(s):  
Performance Model ◽  
Test Facility ◽  
Pump Performance ◽  
Integral Effect ◽  
Effect Test

ROSA/LSTF Experiment on a PWR Station Blackout Transient With AM Measures and RELAP5 Post-Test Analysis

2014 ◽  
Author(s):  
Takeshi Takeda ◽  
Iwao Ohtsu ◽  
Taisuke Yonomoto
Keyword(s):  
Flow Rate ◽  
Large Scale ◽  
Heat Removal ◽  
Sensitivity Analyses ◽  
Test Facility ◽  
Primary System ◽  
Energy Agency ◽  
Coolant Injection ◽  
Station Blackout ◽  
Secondary Side

An experiment on a PWR station blackout transient with the TMLB’ scenario and accident management (AM) measures was conducted using the ROSA/large scale test facility (LSTF) at Japan Atomic Energy Agency under an assumption of non-condensable gas inflow to the primary system from accumulator (ACC) tanks. The AM measures proposed in this study are steam generator (SG) secondary-side depressurization by fully opening the safety valves in both SGs with the start of core uncovery and coolant injection into the secondary-side of both SGs at low pressures. The LSTF test revealed the primary pressure started to decrease when the SG primary-to-secondary heat removal resumed soon after the coolant injection into the SG secondary-side. The primary depressurization worsened due to the gas accumulation in the SG U-tubes after the completion of ACC coolant injection. The RELAP5 code predicted well the overall trend of the major phenomena observed in the LSTF test, and indicated remaining problems in the predictions of SG U-tube collapsed liquid level and primary mass flow rate after the gas ingress. The SG coolant injection flow rate was found to affect significantly the peak cladding temperature and the ACC actuation time through the RELAP5 sensitivity analyses.

Experimental Investigation of Non-Condensable Gases Effect on Novovoronezh NPP-2 Steam Generator Condensation Power Under the Condition of Passive Safety Systems Operation

2009 ◽  
Author(s):  
I. I. Kopytov ◽  
S. G. Kalyakin ◽  
V. M. Berkovich ◽  
A. V. Morozov ◽  
O. V. Remizov
Keyword(s):  
Steam Generator ◽  
Flow Rate ◽  
Computational Simulation ◽  
Heat Removal ◽  
Test Facility ◽  
Passive Safety ◽  
Primary Circuit ◽  
Water Radiolysis ◽  
Passive Safety Systems ◽  
Safety Systems

The design substantiation of the heat removal efficiency from Novovoronezh NPP-2 (NPP-2006 project with VVER-1200 reactor) reactor core in the event of primary circuit leaks and operation of passive safety systems only (among these are the systems of hydroaccumulators of the 1st and 2nd stages and passive heat removal system) has been performed based on computational simulation of the related processes in the reactor and containment. The computational simulation has been performed with regard to the detrimental effect of non-condensable gases on steam generator (SG) condensation power. Nitrogen arriving at the circuit with the actuation of hydroaccumulators of the 1st stage and products of water radiolysis are the main sources of non-condensable gases in the primary circuit. The feature of Novovoronezh NPP-2 passive safety systems operation is that during the course of emptying of the 2nd stage hydroaccumulators system (HA-2) the gas-steam mixture spontaneously flows out from SG cold headers into the volume of HA-2 tanks. The flow rate of gas-steam mixture during the operation of HA-2 system is equal to the volumetric water discharge from hydroaccumulators. The calculations carried out by different integral thermal hydraulic codes revealed that this volume flow rate of gas-steam mixture from SG to the HA-2 system would suffice to eliminate the “poisoning” of SG piping and to maintain necessary condensation power. In support of the calculation results, the experiments were carried out at the HA2M-SG test facility constructed at IPPE. The test facility incorporates a VVER steam generator model of volumetric-power scale of 1:46. Steam to the HA2M-SG test facility is supplied fed from the IPPE heat power plant. Gas addition to steam coming to the SG model is added from high pressure gas cylinders. Nitrogen and helium are used in the experiments for simulating hydrogen. The report presents the basic results of experimental investigations aimed at the evaluation of SG condensation power under the inflow of gas-steam mix with different gases concentration to the tube bundle, both under the simulation of gas-steam mixture outflow from SG cold header to the HA-2 system and without outflow. As a result of the research performed at the HA2M-SG test facility, it has been substantiated experimentally that in the event of an emergency leak steam generators have condensation power sufficient for effective heat removal from the reactor provided by PHR system.

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