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.

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.

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.

Simulation of JAERI Downcomer Effective Water Head Experiments With WCOBRA/TRAC-TF2

2018 ◽  
Author(s):  
Jeffrey R. Kobelak ◽  
Jun Liao ◽  
Katsuhiro Ohkawa
Keyword(s):  
Hydraulic System ◽  
Large Scale ◽  
Reactor Vessel ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
System Pressure ◽  
Water Head ◽  
Effective Water

During the reflood phase of a postulated large break loss-of-coolant accident (LBLOCA), the liquid head in the reactor vessel downcomer provides the driving force to reflood the core. Since the reflood rate is a function of the downcomer inventory, the calculation of the downcomer liquid inventory is critical in simulating the reflood phase of a postulated LBLOCA accident in a pressurized water reactor. Since the reactor coolant system pressure decreases rapidly after the onset of a LBLOCA transient, the walls surrounding the downcomer become superheated for the duration of the transient. The Japan Atomic Energy Research Institute (JAERI) downcomer effective water head test facility was designed to study boiling and steam-water interaction in the reactor vessel downcomer under prototypical reflood conditions. A number of tests were conducted at this facility with varying degrees of wall superheating (among other things) that cover the expected degree of superheating in a pressurized water reactor. The wall superheating achieved at the JAERI facility is greater than that of other large-scale facilities that are typically simulated to validate thermal-hydraulic system codes. WCOBRA/TRAC-TF2 is the thermal-hydraulic system code utilized in the FULL SPECTRUM™ LOCA (FSLOCA™) evaluation model (EM). The ability of the WCOBRA/TRAC-TF2 code to predict phenomena occurring in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA has been previously validated. However, only limited wall superheating was present in the existing validation basis. As such, two experiments conducted at the JAERI downcomer effective water head test facility are simulated to provide additional information on the capability of WCOBRA/TRAC-TF2 to predict the liquid inventory in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA. The code captured all the trends observed in the experimental data for both Run 115 and Run 121. The various collapsed liquid levels tended to be well-predicted or under-predicted by the code after the initial simulated accumulator injection period.

scholarly journals Major Achievements and Prospect of the ATLAS Integral Effect Tests

2012 ◽  
Vol 2012 ◽  
pp. 1-18 ◽  
Author(s):  
Ki-Yong Choi ◽  
Yeon-Sik Kim ◽  
Chul-Hwa Song ◽  
Won-Pil Baek
Keyword(s):  
Large Scale ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Hydraulic Test ◽  
Validation Data ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Integral Effect

A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R&D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

Feasibility Study on the Natural Circulation Cooling Procedure at the Loop Unbalanced Condition

2017 ◽  
Author(s):  
Wataru Sakuma ◽  
Shinya Miyata ◽  
Manabu Maruyama ◽  
Junto Ogawa
Keyword(s):  
Large Scale ◽  
Natural Circulation ◽  
Test Facility ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Scale Test ◽  
Safety Enhancement

In typical pressurized water reactor (PWR) plant, in case that one steam generator (SG) is dried out and cannot be credited for the primary cooldown, at least one reactor coolant pump (RCP) has to be operated in order to homogenize the primary coolant temperature distribution among loops when the plant is cooled down to the cold shutdown state. For example, an accident such as steam line break (SLB) and feedwater line break (FLB) leads to this situation. If the natural circulation condition is established due to unavailability of all the RCPs, the natural circulation in the primary loop connected to the affected SG would be interrupted in the plant cooldown phase. In this situation, the continuous cooldown disturbs the smooth depressurization because it leads to void generation at the top of the affected SG tube where the high temperature coolant is left. In addition, there is a possibility that all RCPs cannot be operated in case of the earthquake or the fire if the RCPs are not earthquake-proof and fire-resistant. Therefore the establishment of the cooldown procedure without RCPs operation under the temperature unbalanced condition among the primary loops can contribute to the safety enhancement for typical PWR plants. The several experiments have been already performed to observe the natural circulation phenomena under the temperature unbalanced condition. It has been reported that the plant can be continuously cooled down with smooth depressurization by stepwise cooling manner using MSRVs of the intact SGs. In this study, Mitsubishi Heavy Industries, Ltd. (MHI) performed the transient analyses to simulate the natural circulation cooldown test under the temperature unbalanced condition among loops performed by Large Scale Test Facility (JAEA ROSA/LSTF) using M-RELAP5, which was a modified plant system transient code by MHI based on RELAP5-3D. Based on the analysis results, the thermal hydraulic phenomena of natural circulation cooldown under the temperature unbalanced condition were investigated. As a result, the mechanism of natural circulation interruption was clarified, and this paper shows the outline of the cooldown procedure under the temperature unbalanced condition which could be applied to the PWR plants.

ATHLET Simulation of OECD/NEA ROSA-2 Test-2 About Cold Leg Intermediate-Break LOCA

2013 ◽  
Author(s):  
Nan Yu ◽  
Xiaoliang Fu ◽  
Zheng Du ◽  
Lifang Liu ◽  
Zhen Cao ◽  
...  
Keyword(s):  
Test Data ◽  
Large Scale ◽  
Liquid Level ◽  
Test Facility ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Post Test ◽  
Scale Test

Experiment about intermediate-break loss-of coolant accident with 17% break at cold leg was performed in OECD/NEA ROSA-2 project on Large Scale Test Facility (LSFT). Safety injection was assumed single failure and only injected into intact loop. Before the loop seal clearing, the liquid level dropped obviously and the core dryout took place. ATHLET Mod 2.1 Cycle A was used to do the post-test calculations of this test. The major calculated parameters were compared with the test data. The trend of the prediction results fit well with that of the test data, and the cause of the deviations was analyzed.

Post Test Analysis of LB LOCA Transient in PSB-VVER by CATHARE2 Code

2010 ◽  
Author(s):  
Eugenio Coscarelli ◽  
Alessandro Del Nevo ◽  
Francesco D’Auria
Keyword(s):  
Experimental Data ◽  
Fourier Transform ◽  
Integral Representation ◽  
Hydraulic System ◽  
Large Scale ◽  
Test Facility ◽  
Test Analysis ◽  
Post Test ◽  
Base Method ◽  
Quantitative Evaluations

The OECD/NEA PSB-VVER project provided unique and useful experimental data from the large-scale PSB-VVER test facility for code validation. This facility represents the scaled down layout of the Russian designed PWR reactors, namely VVER-1000. The fifth experiment is a LB LOCA in cold leg. The objectives of the test are the investigation of the thermal-hydraulic response of the VVER-1000 following a LB LOCA accident as well as the investigation of the phenomena involved in the transient by the availability of experimental data which are useful for the validation of thermal-hydraulic system codes. The paper discusses the achievements of the assessment of CATHARE2 code from the qualitative and the quantitative evaluations of the results. The results are compared with the analysis carried out by RELAP5-3D code. The quantification of the accuracy is based on the Fast Fourier Transform Base Method, developed at University of Pisa, which provides an integral representation of the accuracy quantification in the frequency domain.

RELAP5/MOD3 Code Verification Through PWR Pressure Vessel Small Break LOCA Tests in OECD/NEA ROSA Project

2008 ◽  
Author(s):  
Hideo Nakamura ◽  
Tadashi Watanabe ◽  
Takeshi Takeda ◽  
Hideaki Asaka ◽  
Masaya Kondo ◽  
...  
Keyword(s):  
State Of The Art ◽  
Simulation Models ◽  
Predictive Capability ◽  
Experimental Conditions ◽  
Energy Agency ◽  
Test Analysis ◽  
Post Test ◽  
Complex Phenomena ◽  
Reactor Accidents ◽  
Relap5 Code

The Japan Atomic Energy Agency (JAEA) started OECD/NEA ROSA Project in 2005 to resolve issues in the thermal-hydraulic analyses relevant to LWR safety through the experiments of ROSA/LSTF in JAEA. More than 17 organizations from 14 NEA member countries have joined the Project. The ROSA Project intends to focus on the validation of simulation models and methods for complex phenomena that may occur during DBEs and beyond-DBE transients. Twelve experiments are to be conducted in the six types. By utilizing the obtained data, the predictability of codes is validated. Nine experiments have been performed so far in the ROSA Project to date. The results of two out of these experiments; PV top and bottom small-break (SB) LOCA simulations are studied here, through comparisons with the results from pre-test and post-test analyses by using the RELAP5/MOD3.2 code as a typical and well-utilized/improved best estimate (BE) code. The experimental conditions were defined based on the pre-test (blind) analysis. The comparison with the experiment results may clearly indicate a state of the art of the code to deal with relevant reactor accidents. The code predictive capability was verified further through the post-test analysis. The obtained issues in the utilization of the RELAP5 code are summarized as well as the outline of the ROSA Project.

Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

2009 ◽  
Author(s):  
Yuriy V. Parfenov ◽  
Oleg I. Melikhov ◽  
Vladimir I. Melikhov ◽  
Ilya V. Elkin
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Evolutionary Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tube Bundles ◽  
Horizontal Steam Generator ◽  
Calculation Results ◽  
Steam Separation

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

Experimental Research on Non-Condensable Gases Effects in Passive Decay Heat Removal System

2016 ◽  
Author(s):  
Yang Liu ◽  
Haijun Jia ◽  
Li Weihua
Keyword(s):  
High Pressure ◽  
Heat Removal ◽  
Volume Ratio ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Decay Heat ◽  
Optimizing Design ◽  
Heat Removal System ◽  
Removal System

Passive decay heat removal (PDHR) system is important to the safety of integral pressurized water reactor (IPWR). In small break LOCA sequence, the depressurization of the reactor pressure vessel (RPV) is achieved by the PDHR that remove the decay heat by condensing steam directly through the SGs inside the RPV at high pressure. The non-condensable gases in the RPV significantly weaken the heat transfer capability of PDHR. This paper focus on the non-condensable gas effects in passive decay heat removal system at high pressure. A series of experiments are conducted in the Institute of Nuclear and New Energy Technology test facility with various heating power and non-condensable gas volume ratio. The results are significant to the optimizing design of the PDHR and the safety operation of the IPWR.

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