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Session Overview
Session
4.01: MSR and Advanced Reactors – I
Time:
Monday, 16/Mar/2020:
1:30pm - 3:00pm

Session Chair: Li (Emily) Liu, Rensselaer Polytechnic Institute-RPI, United States of America
Location: L-2012

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Presentations

Steady State and Transient Analysis of the Molten Salt Fast Reactor Using Coupled Serpent and OpenFOAM

Ashkhen Nalbandyan1, Esben Bryndt Klinkby1, Bent Lauritzen1, Jacob Groth-Jensen2

1Technical University of Denmark; 2Seaborg Technologies

Molten Salt Reactors (MSRs) potentially exhibit a number of advantages compared to conventional reactors based on solid fuels; these include inherent safety features, enhanced fuel utilization and reduced costs. Among several conceptual designs, the Molten Salt Fast Reactor (MSFR) concept has been selected as one of six main concepts of the Generation IV Reactors [1]. The circulating fuel in the MSFR induces a strong coupling between neutronics and thermal hydraulics, with the added complication of delayed neutrons being affected by the fuel velocity fields. Since the fuel salt is also the primary coolant, turbulence effects have direct impact on the neutronics performance of the reactor and have to be captured correctly to ensure safe reactor operation. As conventional reactor modeling tools are intended for solid fuel, substantial efforts are applied to develop new approaches for modeling of reactors with circulating fuel, in particular the MSFR. Mainly these efforts focus on development of in-house software with dedicated neutronics and thermal hydraulics solvers, or on incorporating a deterministic neutronics solver into a computational fluid dynamics (CFD) software. In this paper we apply an external coupling scheme between the Serpent Monte Carlo code and the OpenFOAM software in order to model the MSFR behavior with the emphasis on investigating transient scenarios such as a step reactivity insertion (RIA) or an unprotected loss-of-coolant-accident (ULOC). The unique features of Serpent and OpenFOAM allow for setting up a coupling interface between two codes without modifying the source code [2]. Specific approach is applied in this work to model the drift of delayed neutron precursors, based on the point-wise delayed neutron precursor tracking available in Serpent. The impact of different turbulence models on fuel flow field is also investigated. The results are compared to the simulation results using other coupling approaches published in literature [3].



Instrumentation and Control Design in the Versatile Test Bay, a Modular Platform for Scaled Separate Effects Tests

Omar Ashraf Alzaabi1, Christopher Forsyth1, James Kendrick2, Per F. Peterson1

1University of California, Berkeley; 2Kairos Power LLC

The application of surrogate fluids for scaled Separate Effects Tests (SETs) broadens the scope of experiments that can be conducted in a university setting by allowing for relatively low-temperature operation at reduced height and area. Benefits include reduced cost and increased options for instrumentation. The Pebble-Bed Heat Transfer Experiment (PB-HTX) at the University of California, Berkeley was designed to use DOWTHERM A oil, which, between the temperatures of 50°C to 120°C, can be made to simultaneously match the Prandtl, Reynolds, and Grashof numbers of molten fluoride salt coolants at prototypical reactor conditions. The PB-HTX was chosen for conversion into the Versatile Test Bay (VTB) due to its modular construction. The VTB is a SET facility capable of quick analog control of hardware power supplies for simulating transient conditions in the installed test-section. This test-section can be easily swapped out for future scaled experiments. It is capable of sending a variety of power profile inputs into the system through the heater, which enables frequency response experiments. Utilizing the VTB, the authors have demonstrated collection of pebble-bed heat transfer data using sinusoidal power inputs. The measured Nusselt numbers matched heat transfer correlation predictions from literature, in addition to previous PB-HTX experiments. This paper serves as a detailed technical guide to the Instrumentation and Control in the VTB.



Preliminary Design Study of Heat pipe for Space Nuclear Reactor Application

Ye Yeong Park, Kyung Mo Kim, In Cheol Bang

Ulsan National Institute of Science and Technology

A design study of the wick structure of a heat pipe was performed herein to suggest the optimal design for space nuclear reactor application and evaluate the thermal performance of the suggested wick design by comparing the heat transfer capacity and operation limits. The advantages of using heat pipes in space nuclear reactors include passive and continuous heat removal from the reactor core after shutdown, high power output, self-containment, low inventory of working fluid, and the light weight of the pipes. The factors governing the heat transfer capacity of the heat pipe are latent heat and capillary force, and because the capillary pumping force is the driving force of the working fluid circulation which determines the maximum heat transfer capacity, the wick is a key component that significantly influences the performance of the heat pipe. There are several operating conditions that must be satisfied to operate heat pipe in space, such as heat transfer under zero gravity where the working fluid returns from the condenser to the evaporator by capillary forces in the wick, or under gravity-assisted condition in ground testing, where the working fluid must be transferred by overcoming the gravitational forces. However, a homogeneous wick is not suitable for this application due to its lack of capillary pumping force ability or permeability to operate in zero-gravity or gravity-assisted conditions. Therefore, a combined wick structure was suggested considering the benefits of a wick having small pore size, which can lead to high capillary limit and liquid permeability. In this study, various combinations of the capillary structure were considered, such as screen wire mesh, sintered porous wick, arterial wick, and groove wick. A series of experiments was performed to estimate the thermal performances and deduce the optimal design of the heat pipe for space nuclear reactor application.



 
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