[en] DOWTHERM A oil is being considered for use as a heat transfer fluid in experiments to help in the design of heat transfer components for the Next Generation Nuclear Plant (NGNP). In conjection with the experiments RELAP5-3D/ATHENA will be used to help design and analyzed the data generated by the experiments. Inorder to use RELAP5-3D the thermophysical properties of DOWTHERM A were implemented into the fluids package of the RELAP5-3D/ATHENA computer propgram. DOWTHERM A properties were implemented in RELAP5-3D/ATHENA using thermophysical property data obtain from a Dow Chemical Company brochure. The data were curve fit and the polynomial equations developed for each required property were input into a fluid property generator. The generated data was then compared to the orginal DOWTHERM A data to verify that the fluid property data generated by the RELAP5-3D/ATHENA code was representitive of the original input data to the generator.

[en] In the development of reactor systems, if cost penalties are not too great, direct testing of reactor components is usually undertaken. When the financial burden of such direct testing becomes too severe, ensuring the adequate performance of reactor components entails the use of proven design methods. The establishment of such methods involves: a) the carrying out of basic generic work to determine the physical phenomena and physical principles involved in component performance. b) the use of this generic data in the generation of mathematical models by means of which reactor situations can be assessed. c) the validation of these models against experimental data (produced at a reasonable cost) in fairly complex situations representative of the reactor configurations. This paper describes the present position, in the UK, in respect of the development of such design methods for the CDFR above sodium environment

[en] Nanofluid has a potential to become a promising coolant in many diverse industrial processes. However, that opportunity faces several challenges that need to be solved through a long road of nanofluid research programs. Three kinds of the challenges that will be studied in this paper are: 1) determination of nanofluid thermophysical properties, 2) heat transfer characteristics of nanofluid, and 3) the stability factor of nanofluid. This paper also assesses the issue that must be addressed when nanofluid is utilized in nuclear technology applications. The radiation safety aspect of nanofluid utilization in nuclear reactor technology must be taken into account. The comprehensive and multidisciplinary research and assessment are crucial to be carried out in order to ensure the practical applications of nanofluid as new and potential heat transfer fluid. (paper)

[en] Direct Numerical Simulation (DNS) can be used for the description of turbulent heat transfer in the fluid at low Reynolds numbers. DNS means precise solving of Navier-Stoke's equations without any extra turbulent models. DNS should be able to describe all relevant length scales and time scales in observed turbulent flow. The largest length scale is actually dimension of system and the smallest length and time scale is equal to Kolmogorov scale. In the present work simulations of fully developed turbulent velocity and temperature fields were performed in a turbulent flume (open channel) with pseudo-spectral approach at Reynolds number 2670 (friction Reynolds number 171) and constant Prandtl number 5.4, considering the fluid temperature as a passive scalar. Two ideal thermal boundary conditions were taken into account on the heated wall. The first one was an ideal isothermal boundary condition and the second one an ideal isoflux boundary condition. We observed different parameters like mean temperature and velocity, fluctuations of temperature and velocity, and auto-correlation functions.(author)

No abstract available

[en] Two-phase heat transfer for the slug to annular flow regimes. Successful design of heat exchanger such as evaporator requires accurate prediction of the local heat transfer coefficient for boiling regime. Investigations of two-phase heat transfer in horizontal pipe flow have led to a new correlation for the transfer coefficient. The proposed correlation is Nu_tp=2,90(Re)^0,82(Pr)^0,78(Bo)^0,62. This correlation was tested and experimental data obtained on covering the slug to annular flow regimes. The correlation produces satisfactory result. (author). 14 refs.; 7 figs

[en] The AGREE code was originally developed as a multi physics simulation code to perform design and safety analysis of Pebble Bed Reactors (PBR). Currently, additional capability for the analysis of Prismatic Modular Reactor (PMR) core is in progress. Newly implemented fluid model for a PMR core is based on a subchannel approach which has been widely used in the analyses of light water reactor (LWR) cores. A hexagonal fuel (or graphite block) is discretized into triangular prism nodes having effective conductivities. Then, a meso-scale heat transfer model is applied to the unit cell geometry of a prismatic fuel block. Both unit cell geometries of multi-hole and pin-in-hole types of prismatic fuel blocks are considered in AGREE. The main objective of this work is to verify and validate the heat transfer model newly implemented for a PMR core in the AGREE code. The measured data in the HENDEL experiment were used for the validation of the heat transfer model for a pin-in-hole fuel block. However, the HENDEL tests were limited to only steady-state conditions of pin-in-hole fuel blocks. There exist no available experimental data regarding a heat transfer in multi-hole fuel blocks. Therefore, numerical benchmarks using conceptual problems are considered to verify the heat transfer model of AGREE for multi-hole fuel blocks as well as transient conditions. The CORONA and GAMMA+ codes were used to compare the numerical results. In this work, the verification and validation study were performed for the heat transfer model of the AGREE code using the HENDEL experiment and the numerical benchmarks of selected conceptual problems. The results of the present work show that the heat transfer model of AGREE is accurate and reliable for prismatic fuel blocks. Further validation of AGREE is in progress for a whole reactor problem using the HTTR safety test data such as control rod withdrawal tests and loss-of-forced convection tests

No abstract available

[en] Heat from a high temperature heat pipe is transferred through a vacuum or a gap filled with electrically nonconducting gas to a cooler heat pipe. The first heat pipe is used to cool the nuclear reactor while the second heat pipe is connected thermally and electrically to a thermionic converter. If the receiver requires greater thermal power density, geometries are used with larger heat pipe areas for transmitting and receiving energy than the area for conducting the heat to the thermionic converter. In this way the heat pipe capability for increasing thermal power densities compensates for the comparatively low thermal power densities through the electrically non-conducting gap between the two heat pipes

No abstract available