A Preliminary 3D Steam Flow Analysis for CET Behavior During LSTF SBLOCA Experiment Using FLUENT Code

2013 ◽  
Author(s):  
Dwi Irwanto ◽  
Akira Satou ◽  
Takeshi Takeda ◽  
Hideo Nakamura
Keyword(s):  
Large Scale ◽  
Flow Analysis ◽  
Steam Flow ◽  
Test Facility ◽  
Ansys Fluent ◽  
Steady State Condition ◽  
The Core ◽  
Flow Mixing ◽  
Calculation Results ◽  
Core Height

A 3D steam flow within simulated fuel bundle of Large Scale Test Facility (LSTF), a PWR system simulator, has been investigated by Computational Fluid Dynamics (CFD) analysis with Ansys Fluent code to clarify influences of the steam flow on Core Exit Temperature (CET) response. A LSTF SBLOCA experiment with 1.5% hot leg break as the OECD/NEA ROSA-2 Project Test 3 was simulated by the CFD code to clarify relation between CET and fuel rod surface temperature. A portion of the LSTF core above the mixture level up to around CET sensors was modeled by taking into account high, medium and low heat-zone heater rod bundle, including internal structures such as end-box and upper core plate (UCP). Simulation of steady-state condition at a certain time when mixture level lowered to a certain position at around half of the core height (post-5) was carried out by considering relevant boundary conditions which were developed based on the LSTF Test 3 results. The calculation results revealed that inner structures of the core such as core spacer, end box and UCP indeed affect the CET due to heat transfer from hot steam to these cool structures. 3D flow mixing may also contribute to the final CET values and the delayed increase in the CET relative to the Peak Cladding Temperature (PCT) in the core.

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.

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.

Influence of Trailing Edge Geometry on the Condensing Steam Flow in Low-Pressure Steam Turbine

2015 ◽  
Author(s):  
Giteshkumar Patel ◽  
Yogini Patel ◽  
Teemu Turunen-Saaresti
Keyword(s):  
Turbine Blades ◽  
Steam Flow ◽  
Nucleation Theory ◽  
Classical Nucleation Theory ◽  
Condensation Process ◽  
Trailing Edge ◽  
Ansys Fluent ◽  
Flow Angle ◽  
Flow Mixing ◽  
Model Flow

The paper describes the influence of trailing edge geometries on the non-equilibrium homogeneously condensing steam flow in the stationary cascade of turbine blades. The computational fluid dynamics (CFD) simulations were performed with the ANSYS Fluent CFD code using the Eulerian-Eulerian approach. The condensation phenomena were simulated on the basis of the classical nucleation theory, and the steam properties were calculated with the real gas model. Flow turbulence was solved by employing the modified version of the shear-stress transport (SST) k-ω turbulence model. For this study, three trailing edge profiles; that is, conic, semicircular and square were considered. The influence of the trailing edge shapes were discussed together with experimental data available in the literature. The presented results show that the trailing edge geometries influence on the nucleation process, the droplet size, wetness fraction, the shock waves structure generated at trailing edge and its angles, the flow angle, the entropy generation and flow mixing in the wake. The cascade loss coefficients were calculated for the low inlet superheat case and for the high inlet superheat case. The presented results demonstrated that the losses that occur due to the irreversible heat and mass transfer during the condensation process were also influenced due to the trailing edge shapes.

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.

Geometry Survey on the Convex Shaped Core for Recriticality Prevention Against CDA in Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Keiko Chitose ◽  
Yoshiaki Tachi ◽  
Toshio Wakabayashi ◽  
Naoyuki Takaki
Keyword(s):  
Molten Pool ◽  
Large Scale ◽  
Inner Core ◽  
Outer Core ◽  
Core Region ◽  
Energy Agency ◽  
The Core ◽  
Negative Reactivity ◽  
Core Height ◽  
Reactivity Insertion

In sodium-cooled fast reactors, the core is not arranged in its most reactive configuration. In this case, when the fuel melts to form a molten pool, the recriticality may occur by positive reactivity insertion due to core compaction. To prevent such recriticality, special devices of the fuel subassembly structure for discharging the molten fuel from the core region, have been investigated by the Japan Atomic Energy Agency (JAEA). On the other hand, the inherent feature of core geometry and the neutron characteristics may provide the similar effect to prevent such recriticality. The purpose of this study is to design the core specification its deformation in CDA causes negative feedback to subcritical condition, without any fuel discharge device. The convex shaped core has the longer fuel length in the inner-core region and the shorter fuel in the outer-core region. Therefore, the core geometry as intact status has a lower neutron leakage effect. When the fuel melts in CDA, the core height is compacted and negative reactivity insertion is expected during molten pool formation. The convex shaped core is based on the large-scale cylindrical homogeneous core (3,600 MWth, 4.95m in core diameter, and 0.75m in core height). The calculation showed that the compaction of cylindrical core leads to a reactivity gain, whereas the convex shaped core results in negative reactivity effect. In this geometry, both inner-core and outer-core are divided into two regions. Furthermore, we introduced the smaller diameter pin for inner-core and keep uniform Pu enrichment for all regions. The smaller diameter pins in high importance region are effective for flat-distribution. Through pin diameter survey, we confirmed the advantages of smaller diameter pin, such as reducing pressure loss of core coolant and decreasing the height of molten pool.

Soil/Tire Interaction Analysis Using FEM and FVM

2005 ◽  
Vol 33 (1) ◽  
pp. 38-62 ◽  
Author(s):  
S. Oida ◽  
E. Seta ◽  
H. Heguri ◽  
K. Kato
Keyword(s):  
Large Scale ◽  
Interaction Analysis ◽  
Yield Criterion ◽  
Surrounding Medium ◽  
Flow Analysis ◽  
Measurement Data ◽  
Traction Force ◽  
Elastoplastic Material ◽  
The Road ◽  
Soil Flow

Abstract Vehicles, such as an agricultural tractor, construction vehicle, mobile machinery, and 4-wheel drive vehicle, are often operated on unpaved ground. In many cases, the ground is deformable; therefore, the deformation should be taken into consideration in order to assess the off-the-road performance of a tire. Recent progress in computational mechanics enabled us to simulate the large scale coupling problem, in which the deformation of tire structure and of surrounding medium can be interactively considered. Using this technology, hydroplaning phenomena and tire traction on snow have been predicted. In this paper, the simulation methodology of tire/soil coupling problems is developed for pneumatic tires of arbitrary tread patterns. The Finite Element Method (FEM) and the Finite Volume Method (FVM) are used for structural and for soil-flow analysis, respectively. The soil is modeled as an elastoplastic material with a specified yield criterion and a nonlinear elasticity. The material constants are referred to measurement data, so that the cone penetration resistance and the shear resistance are represented. Finally, the traction force of the tire in a cultivated field is predicted, and a good correlation with experiments is obtained.

Measurements and Modeling of Ingress in a New 1.5-Stage Turbine Research Facility

2016 ◽  
Vol 139 (1) ◽  
Author(s):  
Marios Patinios ◽  
James A. Scobie ◽  
Carl M. Sangan ◽  
J. Michael Owen ◽  
Gary D. Lock
Keyword(s):  
Theoretical Model ◽  
Gas Turbines ◽  
Test Facility ◽  
Radial Clearance ◽  
Operating Life ◽  
The Core ◽  
Wide Range ◽  
Purge Flow ◽  
Rotor Disk ◽  
Engine Components

In gas turbines, hot mainstream flow can be ingested into the wheel-space formed between stator and rotor disks as a result of the circumferential pressure asymmetry in the annulus; this ingress can significantly affect the operating life, performance, and integrity of highly stressed, vulnerable engine components. Rim seals, fitted at the periphery of the disks, are used to minimize ingress and therefore reduce the amount of purge flow required to seal the wheel-space and cool the disks. This paper presents experimental results from a new 1.5-stage test facility designed to investigate ingress into the wheel-spaces upstream and downstream of a rotor disk. The fluid-dynamically scaled rig operates at incompressible flow conditions, far removed from the harsh environment of the engine which is not conducive to experimental measurements. The test facility features interchangeable rim-seal components, offering significant flexibility and expediency in terms of data collection over a wide range of sealing flow rates. The rig was specifically designed to enable an efficient method of ranking and quantifying the performance of generic and engine-specific seal geometries. The radial variation of CO2 gas concentration, pressure, and swirl is measured to explore, for the first time, the flow structure in both the upstream and downstream wheel-spaces. The measurements show that the concentration in the core is equal to that on the stator walls and that both distributions are virtually invariant with radius. These measurements confirm that mixing between ingress and egress is essentially complete immediately after the ingested fluid enters the wheel-space and that the fluid from the boundary layer on the stator is the source of that in the core. The swirl in the core is shown to determine the radial distribution of pressure in the wheel-space. The performance of a double radial-clearance seal is evaluated in terms of the variation of effectiveness with sealing flow rate for both the upstream and the downstream wheel-spaces and is found to be independent of rotational Reynolds number. A simple theoretical orifice model was fitted to the experimental data showing good agreement between theory and experiment for all cases. This observation is of great significance as it demonstrates that the theoretical model can accurately predict ingress even when it is driven by the complex unsteady pressure field in the annulus upstream and downstream of the rotor. The combination of the theoretical model and the new test rig with its flexibility and capability for detailed measurements provides a powerful tool for the engine rim-seal designer.

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