Prediction of nucleate boiling heat transfer on horizontal U-shaped heat exchanger submerged in a pool of water using MARS code
Introduction
In the field of nuclear engineering, a horizontal U-shaped heat exchanger (HX) submerged in a pool is under development as a key equipment of a passive heat removal system. As representative passive safety systems with this type of HX, there are passive containment cooling system (PCCS) of economic simplified boiling water reactor (ESBWR), the emergency condenser system (ECS) of the siede wasser reaktor-1000 (SWR-1000), and the passive auxiliary feedwater system (PAFS) of the advanced power reactor plus (APR+). During an accident condition of a nuclear power plant (NPP), these passive safety systems mitigate accidents by cooling the nuclear system effectively via the heat transfer through the steam condensation inside the U-shaped tube and the water boiling outside the U-shaped tube. By the transfer of the core decay heat to the cold water in the pool as a final heat sink passively, the integrity of the NPP can be ensured. Therefore, the accurate prediction of the heat transfer performance of the HX has been an important issue for the reliable design of the HX and the safety analysis of NPP installed with these passive safety systems.
At present, the design and the safety analysis of the passive safety systems are performed mainly using the best-estimate thermal-hydraulic analysis codes (BE codes) such as RELAP5 and MARS. However, those codes under-predict the heat removal performance of the HX significantly (Kim et al., 2013, Cho et al., 2013, Jeon et al., 2014) because the present BE codes do not have the suitable models for both the condensation heat transfer in the horizontal tube and the nucleate boiling heat transfer on the horizontal tube, both of which ultimately determining the heat transfer performance of the HX. This is because, in the thermal-hydraulic (TH) field of previous nuclear engineering, most researches predicting condensation and boiling heat transfer have focused on vertical tubes. There are few models developed for the horizontal U-shaped HX in the passive safety systems. Therefore, it is required to develop and secure suitable models for both the condensation heat transfer in the horizontal tube and the nucleate boiling heat transfer on the horizontal tube for the accurate prediction of the heat transfer performance of the HX using the BE codes.
Recently, for the condensation heat transfer model in the horizontal tube, Jeon et al., 2013a, Jeon et al., 2013b proposed the applicable models to the HX of the passive safety systems by assessing the previous condensation models for the annular and stratified flows using the MARS code. Using the recommended condensation models depending on the flow regime, it is expected that the predictive capability of the BE codes can be improved for the horizontal in-tube condensation heat transfer of the HX in the passive safety system. Now, the nucleate boiling heat transfer on the horizontal tube remains to be an issue. In order to complete the heat transfer modeling for the horizontal U-shaped HX submerged in a pool, this study aims to improve the prediction capability of the BE codes for the nucleate boiling heat transfer on the horizontal tube.
Nucleate boiling heat transfer on the horizontal tube is a frequent and important phenomenon encountered in many industrial applications such as evaporators in a refrigeration and air-conditioning systems, reboilers in chemical process industries, and many components in power engineering and other thermal processing plants because of its high heat transfer capability. In various engineering fields, over the past few decades, there have been many experimental and analytical researches to understand the nucleate boiling heat transfer on the horizontal tubes. A few correlations (Polley et al., 1980, Hwang and Yao, 1986, Webb and Chien, 1994, Gupta et al., 1995) have been proposed to predict the nucleate boiling heat transfer on the horizontal HX tubes. Most correlations were developed based on the superposition model by Chen (1966) established for two-phase flow in vertical tubes, and predicted their own data for the horizontal tubes with reasonable accuracy. However, there is no agreement on which correlation is the best. All correlations, based on the use of limited database, show a considerable deviation from the experimental data produced by different authors using different HX geometries, fluids and flow conditions.
In the TH field of nuclear engineering, there are few nucleate boiling models developed for the horizontal U-shaped HX of the passive safety systems. Furthermore, there have been no researches conducted on the applicability of the previous nucleate boiling models, based on the design and research experience of the reboilers or evaporators, to the HX of the passive safety systems. Most heat transfer analyses with BE codes for the HX have been performed using the Chen (1966) correlation (Arai et al., 2003, Schaffrath et al., 1999, Cho et al., 2013, Jeon et al., 2014). Moreover, it is not known how to physically model the HX pool, and which correlations are suitable among the pool boiling and forced convective boiling correlations. Therefore, in order to predict the nucleate boiling heat transfer on the horizontal tube using BE codes, it is essential to assess the prediction capability of the previous nucleate boiling correlations and to establish the prediction method. Then, if necessary, it is required to develop an appropriate heat transfer model in consideration of the main heat transfer mechanism on the horizontal U-shaped HX submerged in a pool.
In this study, a target passive safety system is a PAFS. It has the U-shaped HX deeply submerged in a pool of water. To obtain a reliable prediction of the heat transfer coefficient (HTC) on the horizontal U-shaped HX in the PAFS using the BE code, this study performed the followings: (1) From a literature survey, various nucleate boiling heat transfer correlations ranging from 7 pool boiling correlations to 8 forced convective boiling correlations on the horizontal tubes were reviewed. Then, all boiling correlations were incorporated into the MARS. (2) The water boiling data were collected from PASCAL (Kang et al., 2012, Kim et al., 2013, Bae et al., 2013) and ATLAS-PAFS (Kang et al., 2012, Bae et al., 2013) experiments. Those experiments were conducted by Korea Atomic Energy Research Institute (KAERI) to evaluate the performance of the horizontal U-shaped HX in the PAFS. They provided the detailed heat transfer data. (3) For the PASCAL experiment, MARS simulations were performed to assess the prediction capability of the previous nucleate boiling correlations. The MARS predictions by each correlation were compared with the experimental data. Then, the prediction capability of each correlation was evaluated. (4) The main mechanisms for the boiling heat transfer on the horizontal U-shaped HX submerged in a pool were investigated, taking into account the PASCAL experimental data, MARS simulations and literature survey comprehensively. Then, the prediction method was proposed. (5) Based on the prediction method, a new boiling model for the horizontal U-shaped HX submerged in a pool was developed with MARS. Then, the proposed model was validated with the experimental data from the PASCAL and ATLAS-PAFS.
Section snippets
Review of previous nucleate boiling heat transfer correlations
Various nucleate boiling heat transfer correlations on the horizontal tubes are available in the literature. Most of these correlations can be categorized as pool boiling (see Table 1) or forced convective boiling correlations (see Table 2). In this section, nucleate boiling heat transfer correlations to be assessed are presented. The understanding of the characteristics of each correlation provides a basis of the development of new boiling model described later in this paper.
Analysis code
In order to predict the nucleate boiling HTCs, this study used MARS (KAERI, 2009) as a BE code. It has been developed by KAERI with the objective of producing a state-of-art realistic TH systems analysis code with multi-dimensional analysis capability (Jeong et al., 1999, Lee et al., 2002). The main structures of MARS are based on the consolidated version of RELAP5/MOD3.2 (USNRC, 1988) and COBRA-TF codes (Thurgood et al., 1983). In addition to them, for a 3D simulation, MARS incorporates a
Results
This section contains five parts: (A) assessments of the prediction capability of the previous nucleate boiling correlations; (B) investigation of the main heat transfer mechanisms on the horizontal U-shaped HX submerged in a pool; (C) establishment of the prediction method with the BE code; (D) development of the nucleate boiling model; (E) validation of the proposed boiling model.
Low HTC at position 2 in PASCAL experiment
Given that the heat transferred from the tube-side by the steam condensation decreases along the tube length, it is expected that the nucleate boiling HTC might be highest at the tube inlet region. However, in PASCAL experiment, it is observed that the local HTC at position 2 in Fig. 4 is significantly smaller than that at positions 3, 4, and 5 and it is slightly larger than the local HTC at position 7 (see Fig. 12). This is possibly because the forced convection effect from the lower part of
Summary
This article reported the first systematic heat transfer analysis with the BE code for the nucleate boiling heat transfer on the horizontal U-shaped HX submerged in a pool of water with the multi-dimensional flow. To obtain a reliable prediction of the nucleate boiling heat transfer on the horizontal parts of the U-shaped tubes, following researches were performed, namely: (1)a comprehensive review of the characteristics of the previous nucleate boiling correlations ranging from 7 pool boiling
Acknowledgments
This work was supported by the Korea Institute of Energy Technology Evaluation and Planning (KETEP) and the Korea Radiation Safety Foundation (KORSAFE) grant funded by the Korean government (the Nuclear Research & Development program of the Ministry of Trade, Industry & Energy (grant number 20131510101620) and the Nuclear Safety Research Center Program of the Nuclear Safety and Security Commission (grant number 1305011)).
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