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Title: CFD Analysis of Upper Plenum Flow in Sodium-Cooled Fast Reactors

Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE Office of Science (SC)
OSTI Identifier:
DOE Contract Number:
Resource Type:
Resource Relation:
Conference: 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16), 08/30/15 - 09/04/15, Chicago, IL, US
Country of Publication:
United States

Citation Formats

Kraus, Adam R., and Hu, Rui. CFD Analysis of Upper Plenum Flow in Sodium-Cooled Fast Reactors. United States: N. p., 2015. Web.
Kraus, Adam R., & Hu, Rui. CFD Analysis of Upper Plenum Flow in Sodium-Cooled Fast Reactors. United States.
Kraus, Adam R., and Hu, Rui. 2015. "CFD Analysis of Upper Plenum Flow in Sodium-Cooled Fast Reactors". United States. doi:.
title = {CFD Analysis of Upper Plenum Flow in Sodium-Cooled Fast Reactors},
author = {Kraus, Adam R. and Hu, Rui},
abstractNote = {},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2015,
month = 1

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  • Gas coolant at low pressure exhibits poor heat transfer characteristics. This is an area of concern for the passive response targeted by the Generation IV GCFR design. For the first 24 hour period, the decay heat removal for the GCFR design is dependent on an actively powered blower, which also would reduce the temperature in the fuel during transients, before depending on the passive operation. Natural circulation cooling initiates when the blower is stopped for the final phase of the decay heat removal, as under forced convection the core decay heat is adequately cooled by the running blower. The abilitymore » of the coolant to flow in the reverse direction or having recirculation, when the blowers are off, necessitates more understanding of the flow behavior characteristics in the upper plenum. The work done here focuses primarily on the period after the blower has been turned off, as the core is adequately cooled when the blowers are running, thus there was no need to carry out the analysis for the first 24 hours. In order to understand the plume behavior for the GCFR upper plenum several cases were run, with air, helium and helium-air mixture. For each case, the FLUENT was used to characterize the steady state velocity vectors and corresponding temperature in the upper plenum under passive decay heat removal conditions. This study will provide better insight into the plume interaction in the upper plenum at low flow and low pressure conditions.« less
  • Upper plenum flow behavior is important for many operational and safety issues in sodium fast reactors. The Prototype Gen-IV Sodium Fast Reactor (PGSFR), a pool-type, 150 MWe output power design, was used as a reference case for a detailed characterization of upper plenum flow for normal operating conditions. Computational Fluid Dynamics (CFD) simulation was utilized with detailed geometric modeling of major structures. Core outlet conditions based on prior system-level calculations were mapped to approximate the outlet temperatures and flow rates for each core assembly. Core outlet flow was found to largely bypass the Upper Internal Structures (UIS). Flow curves overmore » the shield and circulates within the pool before exiting the plenum. Cross-flows and temperatures were evaluated near the core outlet, leading to a proposed height for the core outlet thermocouples to ensure accurate assembly-specific temperature readings. A passive scalar was used to evaluate fluid residence time from core outlet to IHX inlet, which can be used to assess the applicability of various methods for monitoring fuel failure. Additionally, the gas entrainment likelihood was assessed based on the CFD simulation results. Based on the evaluation of velocity gradients and turbulent kinetic energies and the available gas entrainment criteria in the literature, it was concluded that significant gas entrainment is unlikely for the current PGSFR design.« less
  • An innovative sodium-cooled fast reactor has been investigated in a feasibility study of fast breeder reactor cycle systems in Japan. A compact reactor vessel and a column-type upper inner structure with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates are set in the reactor vessel below the free surface to prevent gas entrainment. We performed a one-tenth-scaled model water experiment for the upper plenum of the reactor vessel. Gas entrainment was not observed in the experiment under the same velocity condition as the reactor. Three vortex cavitations were observed near the hot-leg inlet.more » A vertical rib on the reactor vessel wall was set to restrict the rotating flow near the hot leg. The vortex cavitation between the reactor vessel wall and the hot leg was suppressed by the rib under the same cavitation factor condition as in the reactor. The cylindrical plug was installed through the hole in the dipped plates for the FHM to reduce the flow toward the free surface. It was effective when the plug was submerged into the middle height in the upper plenum. This combination of two components had a possibility to optimize the flow in the compact reactor vessel.« less
  • An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heatmore » from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor inlet temperature. The peak cladding, hot pool, cold pool and reactor inlet temperatures were calculated during LOFC. The results indicate that there are two phases during LOFC transient – the initial thermal equilibration phase and the long term decay heat removal phase. The initial thermal equilibration phase occurs over a few hundred seconds, as the system adjusts from forced circulation to natural circulation flow. Subsequently, during long-term heat removal phase all temperatures evolve very slowly due to the large thermal inertia of the primary and buffer pool systems. The results clearly show that passive safety PRACS can effectively transfer decay heat from the primary system to the buffer pool by natural circulation. The DRACS system in turn can effectively transfer the decay heat to the environment.« less
  • Analysis of the proposed GCFR upper and lower plenum flow-through shields has been performed using both discrete ordinate (DOT) and Monte Carlo (MORSE) methods. Several shields having one change of direction in the coolant path (chevron) and two changes of direction (herringbone) were investigated. The shields were modeled as unit cells with periodic boundary conditions. From plenum fluence calculations and design constraints at the reactor vessel liner, it was determined that all the shield configurations analyzed should be adequate for the necessary radiation attenuation.