atw 2018-07

More documents

Recommendations

Info

atw Vol. 63 (2018) | Issue 6/7 ı June/July ENVIRONMENT AND SAFETY 400 | | Fig. 5. Radial cladding temperature distribution after 13.5 h (exp. I). | | Fig. 6. Axial temperature differences of the hottest and coldest rod after 6 h, 9 h and 13.5 h at 20 W/rod (exp. I). 3.2 Comparison of the experimental results of experiment I with those of experiment II For a better understanding of the heat transfer mechanisms, the experiment I with the opened bottom connection is compared to experiment II with the flow blockade. The cladding temperatures of the hottest rod inside the rod bundle and the temperatures of the outer surface of the inner channel were analyzed in Figure 7. The rod bundle is shorter than the inner channel and ends at a height of 4,035 mm. Hence, there is no measurement point above this elevation for the rod bundle. In the lower and in the upper region, differences between the temperature profiles of both experiments can be seen. In experiment I, the rod temperatures in the region from bottom to 2,500 mm are lower compared to the rod temperatures of experiment II. Conversely, the temperatures in the upper part are higher in experiment I compared to experiment II. This is evidence for a chimney effect inside the bundle. Air heated inside the bundle flows upward, and colder air of the outer channel is forced to flow through the bottom connection inside the rod bundle. In experiment II, a stagnation region is formed in the lower part. Ambient air could only interact in the upper and middle region of the facility. 4 Conclusion and outlook In the special case of a complete or a partial drain-down scenario, decay heat has to be removed from the FA by air-cooling. The flow characteristics and temperature progression in the assemblies are not investigated in detail especially not experimentally. Therefore, experimental tests are under way within the joint project SINABEL. Experiments with a rod power half a year after shutdown are presented. The investigation shows that there are huge radial temperature differences inside the bundle itself. This indicates, that a sufficient heat transfer inside a BWR FA is not possible although the cooling outside is high enough. Heat can only be transferred in axial direction. In this event, a blockade at the bottom of a FA has an influence on the convection flows inside the rod bundle. However, these convection flows only lead to opposite temperature changes in the lower and upper part of the rod bundle and has no significant effect on the maximum temperature in the investigated range up to cladding temperatures of 450 °C. The analysis of the heat transfer mechanism of BWR FA shortly after reactor shutdown up to half a year are under way and will give further information of these effects and their dependencies. A grid sensor for combined temperature and flow velocity measurement will be inserted in a quarter of the rod bundle to detect the flow velocities and their changes over the test time and at different rod powers. By modelling the entire FA structure in original scale with additional heaters to adjust nearly adiabatic boundary conditions it will be possible to make improved predictions about the cladding temperature development of a FA in case of air-cooling conditions. In combination with numerical investigations with CFD methods and integral codes, a significant improvement in this field of research can be expected. For this purpose, further publications with additional insights will follow. | | Fig. 7. Axial temperature profiles of the “inner channel” and “rod” after 13.5 h and axial temperature differences (exp. I/II). Acknowledgement This work is part of the research project “SINABEL” and is sponsored by the German Federal Ministry of Education and Research (BMBF) under the contract number 02NUK027A. Responsibility for the content of this paper lies with the authors. Environment and Safety Thermal Hydraulic Analysis of the Convective Heat Transfer of an Air-cooled BWR Spent Fuel Assembly ı Christine Partmann, Christoph Schuster and Antonio Hurtado
atw Vol. 63 (2018) | Issue 6/7 ı June/July References [1] Smith, C.W.: Calculated fuel perforation temperatures: commercial power reactor fuels, NEDO-10093, September 1969. [2] Nourbakhsh, H.P. et al.: Analysis of spent fuel heatup following loss of water in a spent fuel pool, NUREG/ CR-6441, Upton: Brookhaven National Laboratory, March 2002. [3] Benjamin, A. S. et al.: Spent fuel heatup following loss of water during storage, NUREG/CR–0649, US Nuclear Regulatory Commission, Washington DC, 1979. [4] Schuster, C. et al.: Experimental investigation of the rod load in an evaporating spent fuel pool, Proceedings on Annual Meeting on Nuclear Technology, Berlin, Germany, 2007. [5] Schulz, S.; Schuster, C.; Hurtado, A.: Convective heat transfer in a semiclosed BWR-fuel assembly in absence of water, Nuclear Engineering and Design, Volume 272, 2014, Pages 36-44, ISSIV 0029-5493 Authors Further Development of a Thermal- Hydraulics Two-Phase Flow Tool Verónica Jáuregui Chávez, Uwe Imke, Javier Jiménez and V.H. Sánchez-Espinoza Dipl.-Ing. Christine Partmann Dr.-Ing. Christoph Schuster Prof. Dr.-Ing. habil. Antonio Hurtado Technische Universität Dresden Institute of Power Engineering Chair of Hydrogen and Nuclear Energy 01062 Dresden, Germany 401 OPERATION AND NEW BUILD The numerical simulation tool TWOPORFLOW is under development at the Institute for Neutron Physics and Reactor Technology (INR) of the Karlsruhe Institute of Technology (KIT). TWOPORFLOW is a thermal-hydraulics code that is able to simulate single- and two-phase flow in a structured or unstructured porous medium using a flexible 3-D Cartesian geometry. It has the capability to simulate simple 1-D geometries (like heated pipes), fuel assemblies resolving the sub-channel flow between rods or a whole nuclear core using a coarse mesh. The code uses six conservation equations in order to describe the coupled flow of steam and liquid. Several closure correlations are implemented to model the heat transfer between solid and coolant, phase change, wall friction as well as the liquid-vapor momentum coupling. Originally, TWOPORFLOW was used to calculate the flow and heat transfer in micro-channel heat exchangers. The main purpose of this work is the extension, improvement and validation of TWOPORFLOW in order to simulate the thermal-hydraulic behavior of Boiling Water Reactor (BWR) cores. For that aim, the code needs some additional empirical models. In particular, a turbulent lateral mixing model, and a void drift model have been implemented, tested and validated, adopting relevant tests found in the literature. Regarding reactor conditions, the BFBT critical power bundle experiments were selected for the validation. 1 Introduction TWOPORFLOW is a thermal- hydraulic code based on a porous media approach to simulate single and twophase flow in 3D Cartesian coordinates. The time dependent mass, momentum and energy conservation equations for each fluid are solved with a semi-implicit continuous Eulerian type method. TWOPOR- FLOW was originally developed for the simulation of thermal-hydraulic phenomena inside micro-channels [1] [2]. However, the code has been recently modernized and adapted to be able to describe thermal-hydraulic phenomena occurring in Light Water Reactors (LWRs), specifically BWRs [3]. 2 TWOPORFLOW capabilities and main features TWOPORFLOW is capable to solve transient or steady state problems in reactor cores or RPV with a flexible 3D Cartesian geometry which can be used to represent sub-channels, fuel assemblies, or even the whole core. The rod centered and the coolant centered approaches are available for sub- channel simulations. TWOPOR- FLOW uses a system of six conservation equations. The mass conservation equations of the two phases are given by the following equations: (1) (2) (3) The source term Γ I describes the rate of evaporation or condensation at the liquid-vapor interface. The momentum equations are used in non-conservative form as follows: (4) (5) For energy conservation equations, the internal energy (e) is used as the main variable: (6) (7) TWOPORFLOW has additional models to close the system of conservation equations, like solid-coolant heat transfer, interphase heat exchange, empirical correlations for wall friction, empirical correlations for Operation and New Build Further Development of a Thermal- Hydraulics Two-Phase Flow Tool ı Verónica Jáuregui Chávez, Uwe Imke, Javier Jiménez and V.H. Sánchez-Espinoza
Page 1 and 2: nucmag.com 2018 6/7 369 AMNT 2018:
Page 3 and 4: atw Vol. 63 (2018) | Issue 6/7 ı J
Page 5 and 6: Kommunikation und Training für Ker
Page 43: atw Vol. 63 (2018) | Issue 6/7 ı J
Page 67 and 68: Kommunikation und Training für Ker

atw 2018-07

Create successful ePaper yourself

Delete template?

Save as template?