Cooling capacity of plate type research reactors during the natural convective cooling mode Daeseong Jo * , Suki Park, Jonghark Park, Heetaek Chae, Byungchul Lee Korea Atomic Energy Research Institute, 1045 Daeduk-Daero, Dukjin-Dong, Yuseong-Gu, Daejeon 305-353, Republic of Korea article info Article history: Received 29 June 2011 Received in revised form 15 December 2011 Accepted 15 December 2011 Keywords: Plate type research reactor Natural convective cooling Onset of nucleate boiling margin Rectangular channel abstract Numerical investigations on the cooling capacity of plate type research reactors during the natural convective cooling mode are fulfilled in the present study. The cooling capacity, as a thermal core design criterion, is determined by the onset of nucleate boiling (ONB) temperature margin. A simplified model, consisting of the lower plenum, the core, the chimney, the flap valve, and the pool, is simulated by RELAP5/MOD3 and NATCON. The axial and radial peaking factors for the power distribution are applied to investigate the cooling capacity on the hot spot in a fuel assembly. Several convective heat transfer correlations developed by Dittus and Boelter (1930), Petukhov (1970), and Sudo et al. (1985) are implemented into the simulations; then the coolant and surface temperatures and ONB temperature margin, as a function of core power, are obtained from the simulations. The comparisons between RELAP5/MOD3 and NATCON simulations are in good agreement, and the lowest cooling capacity, which limits the permissible core power during the natural convection cooling, is found with Sudo heat transfer correlation. Ó 2012 Elsevier Ltd. All rights reserved. 1. Introduction Thermal hydraulic design ensures that adequate cooling of the core is provided and the core integrity is maintained during normal operations, abnormal operational transients, and accidents. In general, pool type research reactors use a forced convection cooling mode as well as a natural convection cooling mode for core cooling (IATE TECDOC-1474, 2005). In the natural convection cooling mode, the core flow is an upward flow because of the difference of the fluid density between the core (source) and the pool (sink). It is important to know the cooling capacity of reactors under natural convective cooling for reactor safety. The cooling capacity is usually determined by the ONB temperature margin, which is defined as a difference between the fuel surface temperature at the starting point of a nucleate boiling and the fuel surface temperature at a local cooling condition. If the nucleate boiling is prevented, a large thermal margin can be also ensured for the onset of flow instability (OFI) and departure from nucleate boiling (DNB). IAEA guidebook on research reactor conversions recommends BergleseRohsenow correlation, as shown in Eq. (1), for the deter- mination of ONB heat flux in plate type research reactor coolant channels (IATE TECDOC-233, 1980). q 00 ONB ¼ 1082p 1:156 ½1:8ðT W  T S Þ 2:16 p 0:0234 (1) Bergles and Rohsenow (1964) proposed the correlation of heat flux for incipient boiling in a tube. The tube had an inner diam- eter of 2.387 mm. The ONB experiments were performed at different inlet temperatures. The range of inlet velocity was from 3.74 to 19.42 m/s and the corresponding Re range was from 9  10 3 to 2  10 5 . The predicted incipient boiling was compared with boiling data obtained by Rohsenow and Clark (1951). Although the correlation was developed based on observations in tubes, it is still applicable to narrow rectangular channels. This is because ONB is a localized phenomenon on a heating surface. Hence, the fuel surface temperature at the starting point of nucleate boiling is determined by BergleseRohsenow correlation, and the fuel surface temperature at a local cooling condition is determined by convective heat transfer correlations. There are many convective heat transfer correlations that have been developed with various geometric configurations and operational conditions (Dittus and Boelter, 1930; Sieder and Tate, 1936; Sudo et al., 1985; Vliet and Liu, 1969). However, only a few correlations are applicable to narrow rectangular vertical channels heated from both sides. A common and simple correlation used extensively for many applications is DittuseBoelter convective heat transfer correlation for fluids in turbulent flow as * Corresponding author. E-mail address: djo@kaeri.re.kr (D. Jo). Contents lists available at SciVerse ScienceDirect Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene 0149-1970/$ e see front matter Ó 2012 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2011.12.018 Progress in Nuclear Energy 56 (2012) 37e42