Validation of the finned sodium–air heat exchanger model in MARS-LMR
Introduction
The Korea Atomic Energy Research Institute (KAERI) has designed a pool type sodium-cooled fast reactor, which is called the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) (Kim et al., 2013). Fig. 1 shows a schematic of the PGSFR. The PGSFR has a decay heat removal system (DHRS) has independently passive and active loops to satisfying a system diversity. The decay heat removal system (DHRS) is a designated safety grade component providing a sufficient decay heat removal capability during abnormal conditions, such as a loss of heat sink (LOHS) accident. The passive DHRS (PDHRS) relies exclusively on a natural convection heat transfer, i.e., natural circulation on the sodium side and natural draft on the air side. And its loop is equipped with one sodium-to-sodium decay heat exchanger (DHX), natural-draft sodium-to-air heat exchanger (AHX), and connecting pipes. The active DHRS (ADHRS) is operated by a EM-pump in a loop-side and a blower in the air-side. And its loop is integrated with same components in the PDHRS, except a forced-draft sodium-to-air heat exchanger (FHX) for an air-side heat exchanger. The AHX is a helical type air-sodium heat exchanger in the PDHRS, and the FHX is a finned serpentine-type air-sodium heat exchanger in the ADHRS. The DHX located in a cold pool, removes heat from a primary side and transfers heat to the loop side in the DHRS. And heat in the loop is transferred to ultimate heat sink of ambient air by capability of the AHX or FHX, which depends on the opening area of the damper at the inlet region. Under normal operation, a small amount of heat is removed by the AHX and FHX through a small open area of the damper. It was designed to prevent the solidification of sodium, and to ensure operability when it fully operates during accident conditions. The current heat removal capacities are about 0.3 MW and 2.5 MW for normal and accident conditions, respectively.
The MARS-LMR code has been used for a safety analysis of the PGSFR. The code is based on the MARS which was developed for the transient analysis of a light water reactor. The MARS code basically employs the three-dimensional transient two-fluid model for the two-phase flow system (Jeong et al., 1999). Also the point kinetics equation and the heat conduction equation are modeled to calculate the neutron behavior in the reactor core and the heat transfer from the heat structure to fluid. This code was modified to simulate the sodium thermal–hydraulics and neutronic behaviors in a transient condition for a liquid metal cooled fast reactor (Ha and Jeong, 2010). Recently, the heat transfer models for the heat exchangers in the DHRS were added in the MARS-LMR. Therefore, the validation of a new heat transfer model is essential to analyze the characteristics of the DHRS during an accident.
A large-sodium thermal test program called a sodium test loop for a safety simulation and assessment (STELLA) is being progressed by KAERI. As the first step of the program, the sodium component test loop called STELLA-1 has been completed, which is to be used for demonstrating the thermal–hydraulic performance of major components such as heat exchangers and a mechanical sodium pump and providing validation and verification (V&V) data for their design and safety analysis codes. As a second step of the program an integral effect test loop, called STELLA-2, will be constructed to demonstrate the plant safety and support the design approval for the prototype reactor (Eoh et al., 2013). The performance test for the FHX will be conducted in the STELLA-2 facilities. Fig. 2 shows the test component of the FHX in the STELLA-2 facilities. Before obtaining the experimental data for the FHX from the STELLA-2 facilities, the validation test of MARS-LMR code for the FHX is conducted using existing experimental data.
Section snippets
Heat transfer models for a finned heat exchanger
In order to simulate a sodium–air heat exchanger bundle, correlations for each of tube- and shell-sides are necessary. The convective heat transfer correlation in the tube-side was already added in MARS-LMR with Aoki’s correlation (Aoki, 1973). A major concern of this study is a convective heat transfer in the shell-side of a cross-flow bundle, since Zukauskas’ correlations for shell-side air convective heat transfer are recently added in MARS-LMR (Zukauskas and Karni, 1989). The tube rows of a
Air-to-sodium heat exchanger (AHX) in the PFBR
A prototype fast breeder reactor (PFBR) also has four loops of decay heat removal system called a safety grade decay heat removal system (SGDHRS). The decay heat removal rate is 24 MW, which is approximately 2% of the nominal thermal power of the PFBR. The sodium-to-air heat exchanger (AHX) in the PFBR is quite similar to the FHX in the PGSFR. A heat transfer performance test for the AHX is conducted in the steam generator test facility (SGTF) at the Indira Gandhi Centre for Atomic Research (
Dump heat exchanger (DHX) in JOYO
Nanashima et al. and Doi et al. conducted a JOYO start-up test to confirm whether the heat transfer characteristics of the intermediate heat exchanger (IHX) and dump heat exchanger (DHX) satisfy the design values (Nanashima et al., 1979, Doi et al., 1980). Fig. 9 shows a schematic of the JOYO cooling system, which has two intermediate heat exchangers (IHX) and four dump heat exchangers (DHX). Thus, two DHXs and one IHX connected to a single loop, and there are two loops named A and B. The
Conclusions
The DHRS in the PGSFR has two kinds of sodium–air heat exchangers: the one is helical type (AHX) and the other is finned serpentine type (FHX). The DHRS is an important system for a heat removal during an accident. For example, the DHRS is an only heat removal system under a loss of heat sink accident, which has failure of heat removal through all steam generators. Therefore, its accuracy of the modeling in MARS-LMR is critical to estimate the safety evaluation. Recently, the heat transfer
Acknowledgments
This work was supported by the Nuclear Research & Development Program of the National Research Foundation (NRF) grant funded by the Korean government MSIP (Ministry Science, ICT and Future Planning).
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