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 , 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 and leaves the core at a high enough to energy of fission reaction products and finally as heat generate in a 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 , 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 (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 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

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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 . 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, 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 3D nuclear kinetics, heat conduction and the mass flow rate of the hydrogen propellant, average thermal-hydraulic single and two-phase flow phenomena porosity of the core bed, the porosity of the hot frit, and in a 3D light water reactor core simulated by parallel local hot frit blockage. The peak temperature of the fuel is channels is presented. The heat generation within each shown not to exceed its melting point as a result of fuel element is described by a 3D core kinetics code changing any of these parameters from the base case, with employing a two-group diffusion theory model. Finite the exception of hot frit blockage greater than 60% over a differences are used to approximate the diffusion 0.12 m axial segment of the hot frit. equations for fast and thermal neutron fluxes. The Jing-Hsien Chiang, Masanori Aritomi, Ryuichi Inoue, resulting algebraic equations are solved by an iterative Michitsugu Mori [3], Thermo-hydraulics during start-up in procedure. The heat transport out of a fuel element is natural circulation boiling water reactors, Nuclear determined from a finite difference version of the Fourier Engineering and Design, Volume 146, Issues 1–3,

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IJISET - International Journal of Innovative Science, Engineering & Technology, Vol. 2 Issue 9, September 2015. www.ijiset.com ISSN 2348 – 7968 heat conduction equation applied in the radial direction including experimental results and blind benchmark only. An implicit solution for the exercises. Temperature in each radial cell is found using a tri- Giacomino Bandini, Paride Meloni, Massimiliano Polidori diagonal algorithm. The heat transfer rate into the coolant [7] ,Thermal-hydraulics analyses of ELSY lead fast channel is calculated from the clad and coolant reactor with open square core option, Nuclear Engineering temperatures. The thermal-hydraulic part of the code is and Design, Volume 241, Issue 4, April 2011, Pages based on the generally applicable and self-contained 1165-1171--- This paper deals with the development of ‘coolant channel module’ the ELSY thermal-hydraulic and point kinetic model for (CCM) developed by Hoeld [Hoeld, A., 2004a, A RELAP5 focusing the attention at core assembly level to theoretical concept for a thermal-hydraulic 3D parallel verify that the temperatures at nominal power conditions channel core model. PHYSOR 2004, April 25–29, stay within the safety limits both in Beginning-of-Cycle Chicago, USA; Hoeld, A., 2004b, Are separate-phase (BOC) and in End-of-Cycle (EOC) conditions. Moreover, thermal-hydraulic models better than mixture-fluid in order to have a first evaluation of the system behavior approaches. It depends. Rather not. International in accidental conditions, an Unprotected Loss-of-Flow Conference on ‘Nuclear Engineering for New Europe Accident (ULOF) simulation at BOC has been analyzed 2004’. September 6–9, Portoroz, Slov; Hoeld, A., 2005, A and discussed. Among the next generation of nuclear thermal-hydraulic drift-flux based mixture-fluid model for reactors, well-known as Gen-IV, the LFR (Lead Fast the description of single- and two-phase flow along a Reactor) is one of the most promising advanced reactor general coolant channel. NURETH-11, October 2–6, able to comply the principles of sustainability, economics, Avignon, France, which offers a new approach for the and safety and proliferation resistance. The ELSY project simulation of steady state and transient behavior of single- (European Lead-cooled System), funded by the 6th or two-phase flow within a general coolant channel. The European Framework Programme, aims at investigating concept is demonstrated in the experimental code X3D, the technical/economic feasibility of a high power lead concentrating in a first stage on the steady state behavior fast reactor with waste transmutation capability. Several of a hypothetical BWR or PWR core which consists of a innovative design solutions have been proposed at system central channel and four quadrants. Initial results are level and some of them regard the core region with open- presented to demonstrate the viability of the proposed assemblies that represents the reference option. To support concept. This code therefore tests a different theoretical the design phase and the safety assessment of ELSY, the approach of how to determine the essential 3D thermal- thermal-hydraulic code RELAP5, modified to treat heavy hydraulic mass flow distribution into parallel channels by liquid metals, has been taken into account. . maintaining equal pressure decrease terms over all F. D’Auria, S. Soloviev, V. Malofeev, K. Ivanovd, C. channels. Parisi [8], The three-dimensional neutron kinetics coupled E.Studer, J.P.Magnaud, F.Dabbene, I.Tkatschenko [6], with thermal-hydraulics in RBMK accident analysis, International standard problem on containment thermal– Nuclear Engineering and Design, Volume 238, Issue 4, hydraulics ISP47, Nuclear Engineering and Design, April 2008, Pages 1002-1025---This paper deals that the Volume 237, Issue 5, March 2007, Pages 536-551--- In RBMK core is a complex ensemble of high-pressure high- this paper the understanding of hydrogen distribution temperature tubes, graphite bricks, low-pressure low- during severe accidents in nuclear reactor containment is temperature control rod tubes, graphite interstitial gas still an open issue. Several containment thermal– passages. An about 7MPa boiling light water crosses the hydraulics international standard problems (ISP) have around 19m long vertical tubes (7m active length). The been conducted to address this topic. However, the lattice consisting of graphite columns and hydraulic predictions made by the available lumped parameter or channels is bounded by the reactor cavity whose resistant CFD computer codes were generally not satisfactory. elements are the metal cylindrical tank and thick circular Therefore, a new exercise was launched in 1999 using top and bottom plates with proper holes for the passage of new state-of-theart experimental facilities TOSQAN, tubes. Related to a typical water cooled reactor, the MISTRA and Thai that included sophisticated 3D peculiarities of the RBMK core can be summarized as instrumentation and well-controlled boundary conditions. follows: (a) large dimensions – the overall core volume is Predictive capabilities of important and still uncertain by far the largest for a nuclear power plant (NPP) phenomena such as wall condensation, natural circulation producing electricity; (b) use of separate moderator and and gas stratification are assessed. In addition, comparison coolant constituted by graphite and light boiling water, between lumped parameter (LP) and CFD codes and respectively – the boiling water mostly absorbs neutrons in assessment of the capability of CFD codes to deal with this environment leading to the (small) positive void scaling effects are performed. This article reports on the reactivity coefficient; (c) presence of water channels very part of the exercise which concerns the MISTRA facility close to each other containing coolant at different

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IJISET - International Journal of Innovative Science, Engineering & Technology, Vol. 2 Issue 9, September 2015. www.ijiset.com ISSN 2348 – 7968 temperatures (543–557K and 350K for fuel channels (FC) accuracy and reliability in the solution fields than constant and control and protection system (CPS) channels, time stepping methods, even for transients with rapid and respectively); (d) presence of core-wide radial, core-wide discontinuous variations. axial and local temperature gradients in the graphite bricks Chang Oha, Eung Kim, Richard Schultz, Mike Patterson, with temperature values in the range 330–650K with the David Petti, Hyung Kang [10], Comprehensive thermal high-temperature values justified by the neutron hydraulics research of the very high temperature gas moderation and gamma-heating processes. Owing to the cooled reactor, Nuclear Engineering and Design, Volume above peculiarities, the development and the use of a 240, Issue 10, October 2010, Pages 3361-3371--- In this three-dimensional neutron kinetics code (3D NK) coupled paper The Idaho National Laboratory (INL), under the with a one dimensional thermal-hydraulic (TH) code is auspices of the U.S. Department of Energy, is conducting essential in RBMK safety analyses. Two approaches have research on the Very High Temperature Reactor (VHTR) been used within the present context, i.e. use of coupled design concept for the Next Generation Nuclear Plant 3D NK-TH codes to support the accident analysis in the (NGNP) Project. The reactor design will be graphite RBMK as discussed in the first of the companion papers in moderated, thermal neutron spectrum reactor that will this journal volume: application of Korsar-Bars making produce electricity and hydrogen in a highly efficient use of the Unk code to derive -matrices needed for Bars manner. The NGNP reactor core will be either a prismatic and of Relap5/3D-Nestle making use of the Helios code to graphite block type core or a pebble bed core. The NGNP derive the macroscopic cross-sections. Bounding transient will use very high-burn up, low-enriched uranium, analyses of accident scenarios including control rod TRISO-coated fuel, and have a projected plant design withdrawal, various Loss of Coolant Accident (LOCA) service life of 60 years. The VHTR concept is considered and discharge of the control rod circuit, have been to be the nearest-term reactor design that has the capability completed. In all of the analyzed cases, starting from to efficiently produce hydrogen. The plant size, reactor nominal operating conditions, modest fission power time thermal power, and core configuration will ensure passive gradients have been found, i.e. characterized by time decay heat removal without fuel damage or radioactive derivative values for local and global power changes material releases during reactor core accidents. The substantially smaller than current values accepted in safety objectives of the NGNP Project are to: Demonstrate a full- analyses of light water reactors. scale prototype VHTR that is commercially licensed by Jean C. Ragusa, Vijay S. Mahadevan [9], Consistent and the U.S. Nuclear Regulatory Commission, and accurate schemes for coupled neutronics thermal- Demonstrate safe and economical nuclear-assisted hydraulics reactor analysis, Nuclear Engineering and production of hydrogen and electricity. The DOE Design, Volume 239, Issue 3, March 2009, Pages 566-579 laboratories, led by the INL, perform research and --- In this paper conventional coupling paradigms development (R&D) that will be critical to the success of currently used to couple different physics components in the NGNP, primarily in the areas of: high temperature gas reactor analysis problems can be inconsistent in their reactor fuels behavior, high temperature materials treatment of the nonlinear terms due to the operator-split qualification, design methods development and validation, (OS) strategies employed. This leads to the usage of small hydrogen production technologies energy conversion. This time steps to maintain accuracy requirements, thereby paper presents current R&D work that addresses increasing the overall computational time. This paper fundamental thermal hydraulics issues that are relevant to proposes some remedies to OS techniques that can restore a variety of possible NGNP design. consistency in the coupling of the nonlinear terms and explores high-order mono-block nonlinearly consistent Heat Transfer Issues in Reactors techniques with time step control. The performance of the Several generations of reactors are commonly methods was studied for several transient scenarios using a distinguished- 0D point-kinetics/thermal-hydraulics lumped model and a Generation I reactors were developed in 1950-60s and 1D neutronics/heat conduction/enthalpy balance model. very few are still running today. They mostly used natural The results prove that consistent approximations can be uranium fuel and used graphite as moderator. made to enhance the overall accuracy in conventional Generation II reactors are typified by the present US fleet codes with simple nonintrusive techniques. Additionally, and most in operation elsewhere. They typically use an analysis of a mono-block coupling strategy (without enriched uranium fuel and are mostly cooled and having recourse to an OS strategy) is carried out to assess moderated by water. automated time stepping control using higher order Generation III are the advanced reactors, the first few of Implicit Runge–Kutta (IRK) schemes. The conclusions which are in operation in Japan and others are under from these results indicate that nonlinearly consistent adaptive time stepping methods can provide better

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IJISET - International Journal of Innovative Science, Engineering & Technology, Vol. 2 Issue 9, September 2015. www.ijiset.com ISSN 2348 – 7968 construction and ready to be ordered. They are of the structural integrity as well as stability during startup second generation with enhanced safety. /shutdown and load variations. This leads one to Generation IV designs are still on the drawing board and investigate the following issues in heat transfer and will not be operational before 2020 at the earliest, associated fluid flow: probably later. They will tend to have closed fuel cycles Supercritical heat transfer – Current understanding of and burn the long live actinides now forming part of spent (a) fuel, so that fission products are the only high level waste. heat transfer from heated surface to a supercritical fluid is Many will be the fast neutron reactors. quite limited. Empirical correlations have been developed by few investigators over the last two decades (e.g. Advanced Water Reactor System Bishop, Jackson) but agreement is disappointing. This will The major innovation for the ESBWR and the AP1000 require fundamental heat transfer studies under over currently operating light water reactors is the supercritical conditions. Stimulants fluid may be useful simplification of the safety system, while retaining the and reduces the need for prototypic testing under many familiar technologies for BWR and PWR operational conditions. concepts. These simplifications are achieved by using passive safety systems that do not need AC power for activation and gravity driven flows when possible to (b) Supercritical flow stability- The ability to vary the reduce operator/Mechanical actuation. reactor power and flow for startup and shutdown as well as for load variations is required for power plant Advanced Heat Transfer Issues for the ESBWR and operation. Even though the supercritical fluid does not AP1000 undergo phase changes, the large change in density can This advanced reactor containment design has unique cause unstable power flow behavior. The conditions under features, which require us to consider multiphase heat which this occurs needs to be studied. transfer and fluid flow phenomena that had not been considered as crucial in current reactor systems: Ongoing R&D Activities in Reactors (a) Multiphase flows under buoyancy driven and natural Several inherent and passive systems have been adopted in circulation conditions - Current light water reactor Indian innovative reactor, AHWR. Analyses have been design utilize active cooling with pumps and high performed to prove design concepts of these systems. pressure tanks, whereas these advanced designs rely on Experiments and further analyses of these systems are flows generated by modest pressure changes and this being carried out rigorously. Several major areas of R & D have been identified for detailed study and the required can be complicated by boiling phenomena. development activities are in progress. Extensive work has (b) Condensation with non condensibles – The AP been carried out in the area of natural circulation. A containment primarily relies on steam condensation in transparent rectangular loop has been installed to study the presence of non condensibles (air and possibly natural circulation. The stability of natural circulation with hydrogen) on the inner containment walls as well as different heater and cooler orientations has been studied in water evaporation off the outer surface as a means to the loop. Start-up procedure and instability studies are remove the nuclear decay heat from the fuel. being carried out at high pressure natural circulation loop (HPNCL). Flow pattern transition instability studies using (c) Lower Head Cooling – The RCS and IRWST water neutron radiography have been conducted at APSARA can be discharged into containment and flood the reactor in flow pattern transition instability loop (FPTIL). reactor cavity above the lower head of the reactor Integral behavior experiments are being conducted in an vessel; this downward facing heat transfer phenomena integral test loop (ITL). The ITL was commissioned in is key to preserving the vessel integrity. 2005 in BARC. The performance of isolation condensers for reactor decay heat removal is being evaluated in the Advanced Heat Transfer Issues for the SCWR ITL. A parallel channel experimental facility is set up in concept BARC to investigate parallel channel instability. Void coefficient of reactivity has been simulated in the loop. The SCWR reactor concept is unique in that it will require The pre-test single channel stability and parallel channel designers to demonstrate that the reactor and fuel design stability analyses by RELAP 5 and other in house codes limit can be met under supercritical water conditions. have been carried out for the ITL and parallel channel These design limits are the fuel rod cladding and fuel loop. Two-phase low flow pressure drop experiments are

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IJISET - International Journal of Innovative Science, Engineering & Technology, Vol. 2 Issue 9, September 2015. www.ijiset.com ISSN 2348 – 7968 being conducted in 3 MW boiling. Passive containment • Mini symposium on Thermal-hydraulics of non-unity isolation system. 49 water loop (BWL) across the various components of coolant channel. Earlier single phase and Prandtl number flows two phase (air-water) pressure drop experiments were • Coherent Large Scale Structures in the Gaps of Rod- performed on simulated full scale fuel bundle of AHWR in flow test facility at low pressure. Thermal stratification Bundles inside a water pool is being investigated. One dimensional (B) Code Developments and Applications theoretical model and computer code for solving two dimensional Navier-Stokes equations have been developed Computational Fluid Dynamics and to study the stratification phenomena in the water pool. Verification/Validation/Applications (DNS, LES, RANS, Other generalized computer codes available are also being etc.) used for this purpose. Computational Multi-Fluid Dynamics and Validation/Verification/Applications Future Scope The objective of our research is to develop a better • Core Thermal-Hydraulics and Sub channel Analysis understanding of flow and heat transfer phenomena in • Plant System Codes Development and Assessment nuclear reactor systems and to improve their prediction • Nuclear reactors Thermal-Hydraulics capability. We tackle problems encountered in reactors • Containment Analysis that are currently under operation, as well as performing fundamental research towards the design of new • Uncertainties Analysis generation reactors. A focus of our fundamental research • Experiments and Data Bases for Assessment and is the study into mixed . Effects of fluid Verification of 3D Models properties, buoyancy and flow acceleration have been studied extensively. Such effects become particularly • Mini-symposium on Pressure Surges in Nuclear Power significant for fluids at supercritical pressure, a primary Plants operating condition in the Supercritical Water-cooled (C) Severe Accidents and Fires Reactor which is one of the Generation IV candidate • reactors. In dealing with the issues encountered by the Molten Core and Physico- industry, we have been studying the flow induced Chemical Phenomena, Modeling and Experiments vibration related to Pin-Brace interaction in AGRs, effect • Natural Convection and Mixing phenomena, Modeling of reactor core cross flow and modeling of flow in fuel and Experiments bundles. We also study instability issues in reactors. It is a • unique opportunity for researchers and practitioners in the Fuel Coolant Interaction, Modeling and Experiments field to present results of their work and to discuss • Direct Containment Heating by Dispersed Molten Fuel challenges and new ideas with attendees. • Advanced Design Features for Severe Accident Mitigation (A) Two-Phase Flow and Heat Transfer Fundamentals (D) Advanced Code Developments • Multi field Two-Phase Flow Modeling • Fast Transient Modeling and Experiments • Mini Symposium on Flow-induced vibration in nuclear • Enhanced Near–Wall Flow and Heat Transfer components Modeling • Supercritical fluids thermal hydraulics • Fluid and Structures Mechanical Interactions • Interfacial Area Transport (data base, modeling, • Multi-scale multi-physics couplings measurement techniques) • CASL-Thermal-hydraulics Activities in the • Micro and Nano-Scale Basic Phenomena, Fluid Flow Consortium for Advanced Simulation of LWRs and Heat Transfer

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IJISET - International Journal of Innovative Science, Engineering & Technology, Vol. 2 Issue 9, September 2015. www.ijiset.com ISSN 2348 – 7968

(E) Operation and Safety of Existing Reactors • Mini-symposium on Pressure Surges in Nuclear Power • Plant life extension and power up-rating Plants; • Instabilities and Nonlinear Dynamics • Thermal Hydraulics Activities in the Consortium for • NPP Transients and Accidents Analysis Advanced Simulation of LWRs (CASL). • Safety of Sodium cooled RBMK and VVER Reactors Conclusion • Natural Circulation Phenomena and Passive Safety Systems Passive systems are widely considered in ‘innovative’ or advanced nuclear reactor designs and are adopted for (F) Experimental Thermal-Hydraulics coping with critical safety functions. The spread and the • Transient Heat Transfer in reactors variety of related configurations are outlined in the present document. Twenty ‘innovative’ nuclear reactors are • CHF and Post CHF Heat Transfer, Flooding and CCFL described, specially giving emphasis to the passive safety • Instrumentation Technique systems, in the annexes and distinguished in two groups • Integral Testing (a) Advanced water cooled nuclear power plants, (b) Integral reactor systems. • Flow Visualization The levels of development or even the actual deployment (G) Advanced Reactors Thermal-Hydraulics (Gen of the concerned reactor designs (i.e. equipped with III+, - IV, Inpro and Fusion) passive systems) for electricity production are very • Sodium Cooled Fast Reactors Design and Safety different, and the range of maturity of these extend from reactors already in operation to preliminary reactor • Small and Medium Reactors with/without On-Site designs which are not yet submitted for a formal safety Refueling review process. A dozen different passive system types, • Advanced PWRs, Advanced BWRs, Advanced having a few tens of reactor specific configurations, suitable to address safety functions in primary loop or in CANDU Reactors containment have been distinguished, as in Table 2. These • Gas Cooled Fast Reactors and Very High Temperature include systems like the core make-up tanks, the containment spray cooling and the isolation condenser. Reactors The thermal-hydraulic performance of the passive systems • Lead and Lead-Bismuth Cooled Reactors has been characterized by less than a dozen key phenomena at their time characterized through specific • Supercritical Water Reactors descriptions including a few tens of relevant thermal- (H)Waste Management Thermal- Hydraulics hydraulic aspects. There is the need to demonstrate the understanding of the key thermal-hydraulic phenomena (J) Special Topics that are selected for characterizing the performance of • Thermal Hydraulics and Structural Integrity in passive systems: this implies the identification of parameter ranges, the availability of proper experimental Connection to Aging and Life Extension; programs and the demonstration of suitable predictive • BEPU (Best Estimate code Plus Uncertainty) method, capabilities for computational tools. Comprehensive CSAU, Statistical Methods; experimental and code development research activities have been conducted, also very intensely at an • Study of Pressurized Thermal Shock; international level, in the past three to four decades in • OECD/NEA Analysis and Management of Accidents relation to the understanding of thermal-hydraulic phenomena and for establishing related code predictive (BEMUSE, CFD Studies); capabilities for existing nuclear power reactors. In the • Development, Assessment and Applications of same context, research activities also addressed some of the phenomena for passive systems. However, a TRACE; systematic effort for evaluating the level of understanding • Radiological Hazard Related Thermal Hydraulics; of thermal-hydraulic phenomena for passive systems and • Mini-symposium on Flow Induced Vibration in connected code capabilities appears to be limited and in general lacking. Nuclear Components;

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IJISET - International Journal of Innovative Science, Engineering & Technology, Vol. 2 Issue 9, September 2015. www.ijiset.com ISSN 2348 – 7968 Design, Volume 239, Issue 3, March 2009, Pages 566- References- 579.

[1] Ralph Nelson and Cetin Unal, A phenomenological [10]Chang Oha, Eung Kim, Richard Schultz, Mike model of the thermal hydraulics of convective boiling Patterson, David Petti, Hyung Kang, Comprehensive during the quenching of hot rod bundles, Nuclear thermal hydraulics research of the very high temperature Engineering and Design, Volume 136, Issue 3, 2 August gas cooled reactor, Nuclear Engineering and Design, 1992, Pages 299-318. Volume 240, Issue 10, October 2010, Pages 3361-3371

[2]Nicholas J. Morley, Mohamed S. E1-Genk, Thermal- hydraulic analysis of the pellet bed reactor for nuclear thermal propulsion, Institute for Space Nuclear Power Studies, Nuclear Engineering and Design, Volume 149, Issues 1–3, 1 September 1994, Pages 387-400.

[3]Jing-Hsien Chiang, Masanori Aritomi, Ryuichi Inoue, Michitsugu Mori, Thermo-hydraulics during start-up in natural circulation boiling water reactors, Nuclear

Engineering and Design, Volume 146, Issues 1–3, February 1994, Pages 241-252

[4]M. aritomi, T. miyata, M. horiguchi and A. sudi, Thermo-hydraulics of boiling two-phase flow in high conversion light water reactors (thermo-hydraulics at low velocities), International Journal of Multiphase Flow, Volume 19, Issue 1, February 1993, Pages 51-63.

[5]S.Jewer, A.Thompson, A.Hoeld, P.A.Beeley, Initial version of an integrated thermal hydraulics and neutron kinetics 3D code X3D, Nuclear Engineering and Design, Volume 236, Issues 14–16, August 2006, Pages 1533- 1546.

[6]E.Studer, J.P.Magnaud, F.Dabbene, I.Tkatschenko, International standard problem on containment thermal– hydraulics ISP47, Nuclear Engineering and Design, Volume 237, Issue 5, March 2007, Pages 536-551.

[7]Giacomino Bandini, Paride Meloni, Massimiliano Polidori ,Thermal-hydraulics analyses of ELSY lead fast reactor with open square core option, Nuclear Engineering and Design, Volume 241, Issue 4, April 2011, Pages 1165-1171

[8]F. D’Auria, S. Soloviev, V. Malofeev, K. Ivanovd, C. Parisi, The three-dimensional neutron kinetics coupled with thermal-hydraulics in RBMK accident analysis, Nuclear Engineering and Design, Volume 238, Issue 4, April 2008, Pages 1002-1025.

[9]Jean C. Ragusa, Vijay S. Mahadevan, Consistent and accurate schemes for coupled neutronics thermal- hydraulics reactor analysis, Nuclear Engineering and

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