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Thermal Hydraulics Problems in Nuclear Reactors IJISET - International Journal of Innovative Science, Engineering & Technology, Vol. 2 Issue 9, September 2015. www.ijiset.com ISSN 2348 – 7968 Thermal Hydraulics Problems in Nuclear Reactors: A Review Deepak Sharma1, K.M.Pandey2 1 Research Scholar, Department of Mechanical Engineering, NIT Silchar, Assam-788010, India 2 Professor, Department of Mechanical Engineering, NIT Silchar, Assam-788010, India [email protected] Abstract water, liquid metals like molten sodium, and gases like helium. In water-cooled reactors, water picks up reactor The energy released in nuclear fission appears as kinetic heat and leaves the core at a temperature high enough to energy of fission reaction products and finally as heat generate steam in a heat exchanger or directly in the core. generated in the nuclear fuel elements. This heat must be In a pressurized water reactor, the system pressure (~14 removed from the fuel and reactor core and used to MPa) is high enough to avoid any coolant boiling in the produce electrical power. The primary goals of thermal core. In a boiling water reactor, the system pressure (~4-7 core design include achieving a high power density (to MPa) corresponds to saturation at the reactor operating minimize core size), a high specific power (to minimize temperature causing the steam generation to directly take fuel inventory) and high coolant exit temperatures (to place in the reactor core. Although cheap and safe to maximize thermodynamic efficiency). An important aspect handle, water is corrosive at high temperatures and poses of nuclear reactor core analysis involves the determination significant safety challenges since it requires pressurized of the optimal coolant flow distribution and pressure drop primary coolant systems. As a more attractive coolant with across the core. On the one hand, higher coolant flow rates superb heat transfer characteristics and a much wider will lead to better heat transfer coefficients and higher margin between melting and boiling points, the liquid CHF limits. On the other hand, higher flow rates will also sodium does not require a pressurized system, but a result in larger pressure drops across the core, hence larger sodium-cooled fast spectrum reactor often includes an required pumping powers and larger dynamic loads on the intermediate loop and requires special sodium handling core components. Thus, the role of the hydrodynamic and technology since it reacts with water and air at elevated thermal-hydraulic core analysis is to find proper working temperatures. Despite their low heat transfer capacity, with conditions that assure both safe and economical operation their non-reactive properties and low neutron absorption of the nuclear power plant. characteristics, the gas coolants are also considered in Keywords: Heat transfer, Thermal hydraulics, Fuel some reactor types assembly, Reactor core etc. Literature Review Introduction Ralph Nelson and Cetin Unal [1], A phenomenological In a power reactor, the energy produced in fission reaction model of the thermal hydraulics of convective boiling manifests itself as heat to be removed by a coolant and during the quenching of hot rod bundles, Nuclear utilized in a thermodynamic energy conversion cycle to Engineering and Design, Volume 136, Issue 3, 2 August produce electricity. Although this process is essentially the 1992, Pages 299-318--- In this paper, a phenomenological same as in any other steam plant configuration, the power model of the thermal hydraulics of convective boiling in density in a nuclear reactor core is typically four orders of the post-critical-heat-flux (post-CHF) regime is developed magnitude higher than a fossil fueled plant and therefore it and discussed. The model was implemented in the TRAC- poses significant heat transfer challenges. Maximum PF1/MOD2 computer code (an advanced best-estimate power that can be obtained from a nuclear reactor is often computer program written for the analysis of pressurized limited by the characteristics of heat transport system water reactor systems). The model was built around the rather than nuclear considerations. Various factors that determination of flow regimes downstream of the quench influence the power level include the coolant type, core front. The regimes were determined from the flow-regime configuration to maximize heat transfer surface area, map suggested by Ishii and his coworkers. Heat transfer in coolant flow rate, thermo-physical properties of coolant the transition boiling region was formulated as a position- and core materials, and consideration of material dependent model. The propagation of the CHF point was compatibilities. The most common coolant choices include strongly dependent on the length of the transition boiling 789 IJISET - International Journal of Innovative Science, Engineering & Technology, Vol. 2 Issue 9, September 2015. www.ijiset.com ISSN 2348 – 7968 region. Wall-to-fluid film boiling heat transfer was February 1994, Pages 241-252--- In this paper the considered to consist of two components: first, a wall-to- transient behavior of natural circulation for boiling two- vapor convective heat-transfer portion and, second, a wall- phase flow was investigated by simulating normal and to-liquid heat transfer representing near-wall effects. Each abnormal start-up conditions to research the feasibility of contribution was considered separately in each of the natural circulation BWRs such as the SBWR. It was found inverted annular flow (IAF) regimes. The interracial heat that the instabilities, which are out-of-phase geysering, in- transfer was also formulated as flow-regime dependent. phase natural circulation oscillation and out-of-phase The interfacial drag coefficient model upstream of the density wave instability, may occur during the start-up CHF point was considered to be similar to flow through a when the vapor generation rate is insufficient. In this roughened pipe. A free-stream contribution was calculated paper, the mechanism of in-phase natural circulation using Ishii's bubbly flow model for either fully developed oscillation induced by hydrostatic head fluctuation in sub cooled or saturated nucleate boiling. For the drag in steam separators, which has never been understood well the smooth IAF region, a simple smooth-tube correlation enough, is experimentally clarified. Next, the effect of for the interracial friction factor was used. The drag system pressure on the occurrences of the geysering and coefficient for the rough-wavy IAF was formulated in the the natural circulation oscillation are investigated. Finally, same way as for the smooth IAF model except that the from the results, a recommendation is provided to roughness parameter was assumed to be proportional to establish the rational start-up procedure and reactor liquid droplet diameter entrained from the wavy interface. configuration for natural circulation BWRs. The drag coefficient in the highly dispersed flow regime M. aritomi, T. miyata, M. horiguchi and A. sudi [4], considered the combined effects of the liquid droplets Thermo-hydraulics of boiling two-phase flow in high within the channel and a liquid film on wet unheated conversion light water reactors (thermo-hydraulics at low walls. The heat-transfer and interfacial drag models used velocities), International Journal of Multiphase Flow, were based on the flow-regime map noted above with Volume 19, Issue 1, February 1993, Pages 51-63--- In this length averaging of the flow-regime length if more than paper the aim of obtaining a fundamental data base with one regime existed in a given hydraulic cell. regard to HCLWR (high conversion light water reactor) core thermo-hydraulics, the effect of channel gaps on Nicholas J. Morley, Mohamed S. E1-Genk [2], Thermal- thermo-hydraulic behavior is investigated, particularly at hydraulic analysis of the pellet bed reactor for nuclear low velocities as these may be assumed to occur under thermal propulsion, Institute for Space Nuclear Power various abnormal operating conditions in HCLWRs. Studies, Nuclear Engineering and Design, Volume 149, Boiling heat transfer and frictional loss are investigated Issues 1–3, 1 September 1994, Pages 387-400--- In this experimentally, and the applicability of the drift flux paper a two-dimensional steady-state thermal-hydraulics model to narrow channels is examined. Finally, the analysis of the pellet bed reactor for nuclear thermal reflooding behavior is investigated experimentally in propulsion is performed using the NUTHAM-S thermal- relation to a postulated loss-of-coolant accident in hydraulic code. The effects of axial heat and momentum HCLWRs. It has been made clear from the results that the transfers on the temperature and flow fields in the core are incidence of a pressure drop during the reflooding phase is investigated. In addition, the porosity profile in the hot frit much higher than that in current LWRs. is optimized to avoid the development of a hot spot in the S.Jewer, A.Thompson, A.Hoeld, P.A.Beeley [5], Initial reactor core. Finally, a sensitivity analysis is performed version of an integrated thermal hydraulics and neutron using the optimized hot frit porosity profile to determine kinetics 3D code X3D, Nuclear Engineering and Design, the effects of varying the propellant and core parameters Volume 236, Issues 14–16, August 2006, Pages 1533- on the peak fuel temperature and pressure drop across the 1546 --- In this paper a theoretical concept for the core. These parameters include the inlet temperature and description of
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