Analysis of the Core Exit Temperature and the Peak Cladding Temperature During a SBLOCA: Application to a Scaled-Up Model

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Andrea Querol ◽  
Sergio Gallardo ◽  
Gumersindo Verdú
Keyword(s):  
Time Delay ◽  
Large Scale ◽  
Test Facility ◽  
The Core ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Exit Temperature ◽  
Scale Test ◽  
Versus Peak ◽  
Physical Phenomena

During loss-of-coolant accidents (LOCAs), operators may start accident management (AM) actions when the core exit temperature (CET) measured by thermocouples exceeds a certain value. However, a significant time delay and temperature discrepancy in the superheat detection were observed in several facilities. This work is focused on clarifying CET thermocouple responses versus peak cladding temperature (PCT) and studying if the same physical phenomena are reproduced in two TRACE5 models with different geometry (a large-scale test facility (LSTF) and a scaled-up LSTF) during a pressure vessel (PV) upper head small break LOCA (SBLOCA). Results obtained show that the delay between the core uncover and the CET excursion is reproduced in both cases.

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.

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.

ATHLET Analysis of OECD/NEA ROSA-2 Test 7 Experiments on Intermediate-Break LOCA at Cold Leg

2013 ◽  
Author(s):  
Zheng Du ◽  
Xiaoliang Fu ◽  
Nan Yu ◽  
Lifang Liu ◽  
Zhen Cao ◽  
...  
Keyword(s):  
Large Scale ◽  
Flow Model ◽  
Test Facility ◽  
One Dimension ◽  
Primary Coolant ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Post Test ◽  
Scale Test

Test 7 intermediate-break loss-of-coolant accident (IBLOCA) with 13% break at cold leg was conducted in OECD/NEA ROSA-2 Project using Large Scale Test Facility (LSFT). In this test, auxiliary feedwater was assumed to fail and all safety injection was injected only into the intact loop. Core started to dryout when break valve opened. Liquid level in the core dropped rapidly before loop seal clearing (LSC). ATHLET Mod 2.1 Cycle A was used in the post-test analyses of this LSTF experiment. A basis model with two primary coolant loops, one group steam generator U-tube, and three channels in core was built to simulate this test. One dimension finite critical flow model was employed to simulate a nozzle type break with an over predicted result. The major calculated parameters were compared with the test data, and the overall trend of the test was well calculated by the code, it reveals that ATHLET model could predict such IBLOCA with reasonable results.

scholarly journals Validation of TRACE Code Against ROSA/LSTF Test for SBLOCA of Pressure Vessel Upper-Head Small Break

2013 ◽  
Author(s):  
Mian Xing ◽  
Xiao Hu ◽  
Yaodong Chen ◽  
Liangxing Li ◽  
Weimin Ma
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Calculation Model ◽  
Test Facility ◽  
Flow Area ◽  
Cross Sectional ◽  
Loss Of Coolant Accident ◽  
Exit Temperature ◽  
Almost All ◽  
Physical Phenomena

OECD/NEA ROSA/LSTF project tests are performed on the Large Scale Test Facility (LSTF). LSTF is a full-height, full-pressure and 1/48 volumetrically-scaled two-loop system which aims to simulate Japanese Tsuruga-2 Westinghouse-type 4-loop PWR. ROSA-V Test 6-1 simulates a pressure vessel (PV) upper-head small break loss-of-coolant accident (SBLOCA) with a break size equivalent to 1.9% of the volumetrically scaled cross-sectional area of the reference PWR cold leg. By building a TRACE calculation model of LSTF and PV upper-head, the paper dedicated to assess the effect of different modeling options and parameters on simulating thermal hydraulic behaviors of TRACE code. The results show that TRACE code well reproduce the physical phenomena involved in this type of SBLOCA scenarios. Almost all the events in the experiment are well predicted by the model based on TRACE code. In addition, the sensitivity of different models and parameters are investigated. For example, the code slightly overestimated the break mass flow from upper head which could affect the accuracy of the results significantly. The rising of core exit temperature (CET) is significantly influenced by the bypass flow area between downcomer and hot leg.

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.

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