Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment

Hydrodynamic studies of post dryout two-phase downflow in narrow channels

An experimental study of the hydrodynamics of a narrow channel was performed in order to obtain the heat transfer mechanisms and influences contributing to the flow regime transition from inverted annular to inverted slug flows for post dryout downflow. The experimental series consisted of both adiabatic and diabatic visualization tests over a wide range of fluid and thermal parameters. The system inlet gas velocities ranged from 0 to 14 meters per second while the inlet fluid velocities ranged from 1 to 3 meters per second. Full extent visualization of the flow regime was possible due to a quartz tube in tube construction with a clear heating fluid. Constant temperature heating of the freon was accomplished at bulk fluid temperatures above the critical heat flux temperature. For each hydrodynamic flow condition, one to three minuets of VHS-video filming was performed to acquire both flow regime and break-up length data. In addition to this the flow field parameters were recorded simultaneously with the filming

Heat transfer characteristics of an ex-vessel molten core cooling system based on two-phase natural circulation

Passive cooling based on natural circulation is utilized in ex-vessel core catcher system of an advanced nuclear reactor to handle severe accident scenario. The core catcher coolant channel has a unique geometry which consists of heated downward-facing slightly inclined and vertical surfaces. A full height experimental facility with natural circulation driven flow to model ex-vessel core catcher system was designed using scaling analysis. In this study, the cooling capability and heat transfer characteristics of the ex-vessel core catcher system was carried out. Two-phase flow parameters and wall temperatures were measured under a uniform heat flux condition. Two-phase flow structures were identified by high-speed camera visualization along with measurements of two local parameters, void fraction and re-wetting time. The wall temperature and local heat transfer coefficient distribution along the cooling channel were obtained by direct measurements of the heater surface and liquid temperature. The cooling performance of the core catcher system was analyzed based on the experimental results. The results indicated that the core catcher coolant system provides adequate cooling and maintains the integrity of the core catcher plate for prototype heat flux conditions.11Nsciescopu

Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004)

Thoria-based cermet nuclear fuel : neutronics fuel design and fuel cycle analysis.

Cermet nuclear fuel has been demonstrated to have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle produces intriguing options for advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) Evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) Develop enabling technologies required for the economic application of this new fuel form. This paper will first describes the fuel thermal performance model developed for the analysis of dispersion metal matrix fuels. The model is then applied to the design and analysis of thorium/uranium/zirconium metal-matrix fuel pins for light-water reactors using neutronic simulation methods

Interfacial area and interfacial transfer in two-phase systems. DOE final report

In the two-fluid model, the field equations are expressed by the six conservation equations consisting of mass, momentum and energy equations for each phase. The existence of the interfacial transfer terms is one of the most important characteristics of the two-fluid model formulation. The interfacial transfer terms are strongly related to the interfacial area concentration and to the local transfer mechanisms such as the degree of turbulence near interfaces. This study focuses on the development of a closure relation for the interfacial area concentration. A brief summary of several problems of the current closure relation for the interfacial area concentration and a new concept to overcome the problem are given

Thoria-based cermet nuclear fuel : cermet fabrication and behavior estimates.

Cermet nuclear fuels have been demonstrated to have significant potential to enhance fuel performance because of low internal fuel temperatures and low stored energy. The combination of these benefits with the inherent proliferation resistance, high burnup capability, and favorable neutronic properties of the thorium fuel cycle produces intriguing options for advanced nuclear fuel cycles. This paper describes aspects of a Nuclear Energy Research Initiative (NERI) project with two primary goals: (1) evaluate the feasibility of implementing the thorium fuel cycle in existing or advanced reactors using a zirconium-matrix cermet fuel, and (2) develop enabling technologies required for the economic application of this new fuel form. Critical elements in the demonstration of this new fuel form include developing low-cost fabrication methods and characterizing the cermet properties and important performance parameters. A powder-in-tube drawing and heat treatment process is being evaluated as an alternative to hot extrusion. In this method, zirconium metal and ceramic microspheres are mixed, poured into a Zircaloy shell, and compacted into simulated fuel pins. Important processing variables being evaluated include the amount of compaction required to achieve a desired matrix density and the inter-drawing thermal treatment temperature required to achieve adequate matrix fusion and grain growth

Study on hydrodynamically induced dryout and post dryout important to heavy water reactors

Recently, the safety of low pressure liquid cooled nuclear reactors has become a very important issue with reference to the operation of the heavy water reactors at Savannah River Plant. Under accident conditions such as loss-of-flow or loss-of-coolant, these reactors typically encounter unstable two-phase flow which may lead to the occurrence of dryout and subsequent fuel failure. An analytical study using the one-dimensional drift flux model was carried out to investigate the two-phase flow instability for Westinghouse Savannah River Site reactor. The analysis indicates that the first and higher order instabilities exist in the possible transient operational conditions. The instabilities are encountered at higher heat fluxes or lower flow rates. The subcooling has a stabilizing effect except at very low subcooling. An experimental loop has been designed and constructed. A study was conducted on the CHF induced by various flow instabilities. Details of this test loop are presented

Experimental study on two-phase flow natural circulation in a core catcher cooling channel for EU-APR1400 using air-water system

Ex-vessel core catcher cooling system driven by natural circulation is designed using a full scaled air-water system. A transparent half symmetric section of a core catcher coolant channel of a pressurized water reactor was designed with instrumentations for local void fraction measurement and flow visualization. Two designs of air-water top separator water tanks are studied including one with modified ��super-step�� design which prevents gas entrainment into down-comer. In the experiment air flow rates are set corresponding to steam generation rate for given corium decay power. Measurements of natural circulation flow rate, spatial local void fraction distribution and re-wetting time near the top wall are carried out for various air flow rates which simulate boiling-induced vapor generation. Since heat transfer and critical heat flux are strongly dependent on the water mass flow rate and development of two-phase flow on the heated wall, knowledge of two-phase flow characteristics in the coolant channel is essential. Results on flow visualization showing two phase flow structure specifically near the high void accumulation regions, local void profiles, rewetting time, and natural circulation flow rate are presented for various air flow rates that simulate corium power levels. The data are useful in assessing the cooling capability of and safety of the core catcher system. ? 2017 Elsevier B.V.111sciescopu