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.

RELAP5 Code Study of ROSA/LSTF Experiment on a PWR Station Blackout (TMLB’) Transient

2013 ◽  
Author(s):  
Takeshi Takeda ◽  
Hideo Nakamura
Keyword(s):  
Reverse Flow ◽  
Natural Circulation ◽  
Sensitivity Analyses ◽  
Test Analysis ◽  
Coolant Injection ◽  
Post Test ◽  
Station Blackout ◽  
Core Cooling ◽  
Secondary Side ◽  
Relap5 Code

RELAP5 code post-test analysis was performed on one of abnormal transient tests conducted with the ROSA/LSTF simulating a PWR station blackout (SBO) transient with the TMLB’ scenario in 1995. 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 reverse flow U-tubes in steam generator (SG) during long-term single-phase liquid natural circulation. Sensitivity analyses were done further to clarify effectiveness of depressurization of and coolant injection into SG secondary-side as accident management measures to maintain core cooling, based on the LSTF post-test analysis. SG secondary-side depressurization was initiated by fully opening the safety valve in one of two SGs with the incipience of core uncovery. Coolant injection was done into the secondary-side of the same SG at low pressures considering availability of fire engines. The SG depressurization with the coolant injection was found to well contribute to maintain core cooling by the actuation of accumulator system during a PWR SBO (TMLB’) transient.

scholarly journals ROSA/LSTF Tests and Posttest Analyses by RELAP5 Code for Accident Management Measures during PWR Station Blackout Transient with Loss of Primary Coolant and Gas Inflow

2018 ◽  
Vol 2018 ◽  
pp. 1-19 ◽  
Author(s):  
Takeshi Takeda ◽  
Iwao Ohtsu
Keyword(s):  
Flow Rate ◽  
Injection System ◽  
Test Facility ◽  
Relief Valve ◽  
Nitrogen Gas ◽  
Primary Coolant ◽  
Gas Accumulation ◽  
Accident Management ◽  
Station Blackout ◽  
Secondary Side

Three tests were carried out with the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated accident management (AM) measures during station blackout transient with loss of primary coolant under assumptions of nitrogen gas inflow and total failure of high-pressure injection system in a pressurized water reactor. As the AM measures, steam generator (SG) secondary-side depressurization was done by fully opening the relief valves in both SGs, and auxiliary feedwater was injected into the secondary-side of both SGs simultaneously. Conditions for the break size and the onset timing of the AM measures were different among the three LSTF tests. In the three LSTF tests, the primary pressure decreased to a certain low pressure of below 1 MPa with or without the primary depressurization by fully opening the relief valve in a pressurizer as an optional AM measure, while no core uncovery took place through the whole transient. Nonuniform flow behaviors were observed in the SG U-tubes under natural circulation (NC) with nitrogen gas depending probably on the gas accumulation rate in the two LSTF tests that the gas accumulated remarkably. The RELAP5/MOD3.3 code predicted most of the overall trends of the major thermal hydraulic responses observed in the three LSTF tests. The code, however, indicated remaining problems in the predictions of the primary pressure, the SG U-tube collapsed liquid levels, and the NC mass flow rate after the nitrogen gas ingress as well as the accumulator flow rate through the analyses for the two LSTF tests, where the remarkable gas accumulation occurred.

Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

2008 ◽  
Author(s):  
Mitsuhiro Suzuki ◽  
Takeshi Takeda ◽  
Hideo Nakamura
Keyword(s):  
Time Delay ◽  
Large Scale ◽  
Simulation Experiment ◽  
Three Dimensional ◽  
Test Facility ◽  
Large Temperature ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Accident Management

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary sides in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.

Axial Turbine Tip Desensitization by Injection From a Tip Trench: Part 1 — Effect of Injection Mass Flow Rate

2004 ◽  
Author(s):  
Nikhil M. Rao ◽  
Cengiz Camci
Keyword(s):  
Mass Flow Rate ◽  
Mass Flow ◽  
Flow Rate ◽  
Total Pressure ◽  
Large Scale ◽  
Dynamic Pressure ◽  
Coolant Injection ◽  
Gap Height ◽  
Tip Desensitization ◽  
Injection Rates

An experimental study of a turbine tip desensitization method based on tip coolant injection was conducted in a large-scale rotating turbine rig. One of twenty-nine rotor blades was modified and instrumented to have a tip trench with discrete injection holes directed towards the pressure side. Time accurate absolute total pressure was measured 0.3 chord lengths downstream of the rotor exit plane using a fast response dynamic pressure sensor in a phase-locked manner. The test cases presented include results for tip gap heights of 1.40% and 0.72% of the blade height, and coolant injection rates of 0.41%, 0.52%, 0.63%, and 0.72% core mass flow rate. At a gap height of 1.40% the leakage vortex is large, occupying about 15% blade span. A reduction in gap height causes the leakage vortex to reduce in size and move towards the blade suction side. The minimum total pressure measured, for the reduced gap, in the leakage vortex is about 4% greater. Coolant injection from the tip trench is successful in filling in the total pressure defect originally resulting from the leakage vortex without injection. At the higher tip injection rates the leakage vortex is also seen to have moved towards the blade tip. The high momentum associated with the tip jets affects the total pressure distributions in the neighboring passages.

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.

Recent Knowledge From an Experimental Investigation on Self-Leveling Behavior of Debris Bed

2013 ◽  
Author(s):  
Songbai Cheng ◽  
Hidemasa Yamano ◽  
Tohru Suzuki ◽  
Yoshiharu Tobita ◽  
Yuya Nakamura ◽  
...  
Keyword(s):  
Flow Rate ◽  
Large Scale ◽  
Gas Flow ◽  
Particle Density ◽  
The Self ◽  
Gas Flow Rate ◽  
Energy Agency ◽  
Experimental Parameters ◽  
The Status ◽  
Series Of Experiments

Studies on the self-leveling behavior of debris bed are crucial in the assessment of core-disruptive accident (CDA) that could occur in sodium-cooled fast reactors (SFR). To clarify the mechanisms of this behavior, several series of experiments were elaborately designed and performed in recent years under the collaboration between Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). This paper presents the recent knowledge obtained from the newly developed large-scale experiments using gas-injection to simulate coolant boiling. Compared to previous investigations, it can cover a much wider range of gas velocities (presently up to a flow rate of around 300L/min). Based on the experimental data obtained, influence of various experimental parameters, including gas flow rate (∼ 300 L/min), water depth (180 mm and 400mm), bed volume (5L, 7L), particle size (2 ∼ 6 mm), particle density (beads of alumina, zirconia and stainless steel) along with particle shape (spherical and irregularly-shaped) on the leveling was checked and compared. In addition, the status of developing empirical model to predict the self-leveling over current setup was also presented. This work, which gives a large palette of favorable data for a better understanding and an improved estimation of CDAs in SFRs, is expected to benefit future analyses and verifications of computer models developed in advanced fast reactor safety analysis codes.

RELAP5 Code Analysis of LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization as an Accident Management Procedure: Investigation on Base Case Result Appropriate for the Best Estimate Plus Uncertainty (BEPU) Application

2014 ◽  
Author(s):  
Ikuo Kinoshita ◽  
Toshihide Torige ◽  
Michio Murase ◽  
Yoshitaka Yoshida ◽  
Takeshi Takeda ◽  
...  
Keyword(s):  
Steam Generator ◽  
Large Scale ◽  
Model Simulation ◽  
Test Facility ◽  
Base Case ◽  
Best Estimate ◽  
Management Procedure ◽  
Accident Management ◽  
Characteristic Features ◽  
Secondary Side

The application of the Best Estimate Plus Uncertainty (BEPU) method is made to analysis of the “Intentional depressurization of steam generator secondary side” which is an accident management procedure in a small-break loss-of-coolant accident (SBLOCA) with high pressure injection (HPI) system failure. RELAP5/MOD3.2 is used as the analysis code. By applying the BEPU method, the uncertainties of the analysis results can be estimated quantitatively. However, the accuracy of the analysis results depends primarily on the base case result predicted by the best estimate code. In this study, in order to investigate the appropriate base case model, simulation analyses using the RELAP5/MOD3.2 were carried out for the ROSA Large Scale Test Facility (ROSA/LSTF) secondary-side depressurization tests. It was found that the code predicted well the major event progressions such as pressure responses, core liquid level responses, and rod surface temperatures, as well as important phenomena such as formation and clearing of loop seals, accumulation of water from condensation, and countercurrent flow limitation (CCFL) at the inlet of the U-tubes, which are characteristic features of this accident scenario.

ROSA/LSTF Test on Nitrogen Gas Behavior During Reflux Cooling in PWR and RELAP5 Post-Test Analysis

2017 ◽  
Author(s):  
Takeshi Takeda ◽  
Iwao Ohtsu
Keyword(s):  
Large Scale ◽  
Flow Behavior ◽  
Liquid Level ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Nitrogen Gas ◽  
Energy Agency ◽  
Test Analysis ◽  
Pressurized Water ◽  
Post Test

An experiment focusing on nitrogen gas behavior during reflux cooling in a pressurized water reactor (PWR) was performed with the rig of safety assessment/large scale test facility (ROSA/LSTF) at Japan Atomic Energy Agency. The test conditions were made such as the constant core power of 0.7% of the volumetric-scaled PWR nominal power and the primary pressure of lower than 1 MPa, unlike a previous related test with the LSTF. The steam generator (SG) secondary-side collapsed liquid level was maintained at a certain liquid level above the SG tube height. Nitrogen gas was injected stepwise into each SG inlet plenum at a certain constant amount. The primary pressure and the SG U-tube fluid temperatures were greatly dependent on the amount of nitrogen gas accumulated in the SG U-tubes. Nitrogen gas accumulated from the outlet towards the inlet of the SG U-tubes. Non-uniform flow behavior was observed among the SG U-tubes with nitrogen gas ingress. The RELAP5/MOD3.3 code indicated remaining problems in the predictions of the primary pressure and the SG U-tube fluid temperatures after nitrogen gas inflow.

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.

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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