scholarly journals Temperature Profile Measurement in Simulated Fuel Assembly Structure with Wire-Mesh Technology

2018 ◽  
Vol 2018 ◽  
pp. 1-9
Author(s):  
Hiroki Takiguchi ◽  
Masahiro Furuya ◽  
Takahiro Arai
Keyword(s):  
Temperature Distribution ◽  
Flow Field ◽  
Fuel Assembly ◽  
Temperature Profile ◽  
Wire Mesh ◽  
Calibration Line ◽  
Temperature Calibration ◽  
Water Reactor ◽  
Simulated Fuel ◽  
Assembly Structure

When light water reactor (LWR) is subject to a cold shutdown, it needs to be cooled with pure water or seawater to prevent the core melting. To precisely evaluate the cooling characteristics in the fuel assembly, a measurement method capable of installing to the fuel assembly structure and determining the temperature distribution with high temporal resolution, high spatial resolution, and in multidimension is required. Furthermore, it is more practical if applicable to a pressure range up to the rated pressure 16 MPa of a pressurized water reactor (PWR). In this study, we applied the principle of the wire-mesh sensor technology used in the void fraction measurement to the temperature measurement and developed a simulated fuel assembly (bundle) test loop with installing the temperature profile sensors. To investigate the measurement performance in the bundle test section, it was confirmed that a predetermined temperature calibration line with respect to time-average impedance was calculated and became a function of temperature. To evaluate the followability of measurement in a transient temperature change process, we fabricated a 16 × 16 wire-mesh sensor device and measured the hot-water jet-mixing process into the cold-water pool in real time and calculated the temperature profile from the temperature calibration line obtained in advance from each measurement point. In addition, the sensors applied to three-dimensional temperature distribution measurement of a complex flow field in the bundle structure. The axial and cross-sectional profiles of temperature were quantified in the forced flow field with nonboiling when the 5×5 bundle was heated by energization.

Performances Analyses of Crud of EPR Fuel Assembly

2017 ◽  
Author(s):  
Xuming Wang ◽  
Cenxi Yuan ◽  
Chen Ye
Keyword(s):  
Risk Assessment ◽  
Fuel Assembly ◽  
Low Risk ◽  
Third Generation ◽  
Pressurized Water Reactor ◽  
Worst Case ◽  
Pressurized Water ◽  
Water Reactor ◽  
Advanced Fuel ◽  
New Type

Taishan European Pressurized Water Reactor (EPR) is a third generation advanced pressurized water reactor (PWR), which adopts the third generation advanced fuel assembly (AFA-3G-LE) from AREVA for the first time. As suggested by American Electric Power Research Institute (EPRI), an EPRI level III crud risk assessment is necessary for new type of plants. Because crud induced power offset (CIPS) and crud induced local corrosion (CILC) can lead to axial offset anomaly (AOA) and fuel cladding failure, respectively. A EPRI level III CIPS/CILC risk assessment for Taishan EPR is performed with a new framework of simulation by using sub-channel code FLICA, crud code BOA, and Monte Carlo transport code Tripoli-4. Such framework enables a self-consistent calculation, including a detailed description on neutronics contributed by boron. The validation of present work is confirmed because of the good agreement with the experienced data of EPRI. The results show that AFA-3G-LE has a good performance on crud risk assessment. Even in the worst case, the boron-10 deposition (2.6 g) and the maximum thickness of crud (59 μm) are lower than the low risk threshold, 31.33 g and 75 μm, respectively. Hence, It is expected that Taishan EPR has a very low risk on CIPS and CILC.

Effect of Inlet Configuration on Moderator Velocity and Temperature Distribution Inside the Calandria of a Heavy Water Reactor

2008 ◽  
Author(s):  
Abhijeet Mohan Vaidya ◽  
Naresh Kumar Maheshwari ◽  
Pallippattu Krishnan Vijayan ◽  
Dilip Saha ◽  
Ratan Kumar Sinha
Keyword(s):  
Temperature Distribution ◽  
Impact Velocity ◽  
Heavy Water ◽  
Computational Study ◽  
Water Reactor ◽  
Heavy Water Reactor ◽  
The Impact

Computational study of the moderator flow in calandria vessel of a heavy water reactor is carried out for three different inlet nozzle configurations. For the computations, PHOENICS CFD code is used. The flow and temperature distribution for all the configurations are determined. The impact of moderator inlet jets on adjacent calandria tubes is studied. Based on these studies, it is found that the inlet nozzles can be designed in such a way that it can keep the impact velocity on calandria tubes within limit while keeping maximum moderator temperature well below its boiling limit.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

2013 ◽  
Vol 444-445 ◽  
pp. 411-415 ◽  
Author(s):  
Fu Cheng Zhang ◽  
Shen Gen Tan ◽  
Xun Hao Zheng ◽  
Jun Chen
Keyword(s):  
Temperature Distribution ◽  
Fluid Dynamic ◽  
Characteristic Temperature ◽  
Reactor Core ◽  
Flow Distribution ◽  
Sensitivity Study ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

Experimental and Numerical Modeling of Transition Matrix From Momentum to Buoyancy-Driven Flow in a Pressurized Water Reactor

2008 ◽  
Author(s):  
Thomas Ho¨hne ◽  
So¨ren Kliem ◽  
Roman Vaibar
Keyword(s):  
Mass Flow ◽  
Transition Matrix ◽  
Turbulence Models ◽  
Wire Mesh ◽  
Test Facility ◽  
Pressurized Water ◽  
Flow Rates ◽  
Water Reactor ◽  
Ansys Cfx ◽  
Mass Flow Rates

The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the cold leg and downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. This paper presents a matrix of ROCOM experiments in which water with the same or higher density was injected into a cold leg of the reactor model with already established natural circulation conditions at different low mass flow rates. Wire-mesh sensors measuring the concentration of a tracer in the injected water were installed in the cold leg, upper and lower part of the downcomer. A transition matrix from momentum to buoyancy-driven flow experiments was selected for validation of the CFD software ANSYS CFX. A hybrid mesh with 4 million elements was used for the calculations. The turbulence models usually applied in such cases assume that turbulence is isotropic, whilst buoyancy actually induces anisotropy. Thus, in this paper, higher order turbulence models have been developed and implemented which take into account for that anisotropy. Buoyancy generated source and dissipation terms were proposed and introduced into the balance equations for the turbulent kinetic energy. The results of the experiments and of the numerical calculations show that mixing strongly depends on buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation with lower mass flow rates and/or higher density differences. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor directly below the inlet nozzle. Therefore, density effects play an important role during natural convection with ECC injection in PWR and should be also considered in Pressurized Thermal Shock (PTS) scenarios. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

2D Natural Convection and Radiation Heat Transfer Simulations of a PWR Fuel Assembly Within a Constant Temperature Support Structure

2006 ◽  
Author(s):  
Pablo E. Araya Go´mez ◽  
Miles Greiner
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Fuel Assembly ◽  
Heat Generation ◽  
Temperature Difference ◽  
Wall Temperature ◽  
Radiation Heat Transfer ◽  
Radiation Heat ◽  
Fuel Rods ◽  
Water Reactor

Two-dimensional simulations of steady natural convection and radiation heat transfer for a 14×14 pressurized water reactor (PWR) spent nuclear fuel assembly within a square basket tube of a typical transport package were conducted using a commercial computational fluid dynamics package. The assembly is composed of 176 heat generating fuel rods and 5 larger guide tubes. The maximum cladding temperature was determined for a range of assembly heat generation rates and uniform basket wall temperatures, with both helium and nitrogen backfill gases. The results are compared with those from earlier simulations of a 7×7 boiling water reactor (BWR). Natural convection/radiation simulations exhibited measurably lower cladding temperatures only when nitrogen is the backfill gas and the wall temperature is below 100°C. The reduction in temperature is larger for the PWR assembly than it was for the BWR. For nitrogen backfill, a ten percent increase in the cladding emissivity (whose value is not well characterized) causes a 4.7% reduction in the maximum cladding to wall temperature difference in the PWR, compared to 4.3% in the BWR at a basket wall temperature of 400°C. Helium backfill exhibits reductions of 2.8% and 3.1% for PWR and BWR respectively. Simulations were performed in which each guide tube was replaced with four heat generating fuel rods, to give a homogeneous array. They show that the maximum cladding to wall temperature difference versus total heat generation within the assembly is not sensitive to this geometric variation.

Temperature Distribution in the Duct Wall and at the Exit of a 19-Rod Simulated LMFBR Fuel Assembly (FFM Bundle 2A)

1974 ◽  
Vol 24 (2) ◽  
pp. 176-200 ◽  
Author(s):  
M. H. Fontana ◽  
R. E. MacPherson ◽  
P. A. Gnadt ◽  
L. F. Parsly ◽  
J. L. Wantland
Keyword(s):  
Temperature Distribution ◽  
Fuel Assembly ◽  
Duct Wall
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