STUDIES OF IMPROVEMENT OF POWER DENSITY IN ORR LOOPS

Using a simplified model, calculations were made of the possible effects of voids upon the power density in ORR loop experiments. It is concluded that the power density may be markedly increased if voids and channels are plugged with moderator material such as graphite or beryllium. (auth

Nuclear Characteristics of Cylindrical Breeder Reactors Containing Thorium in Pellet Form

The results of nuclear calculations for aqueous homogeneous reactors containing blanket thorium in pellet form are presented. Critical core fuel concentrations, gross breeding ratios, power densities, and multiplication as dependent on core fuel concentration and neutron balances are included. Critical U/sup 233/ core concentration for a 24in.-diam. core was 8.4 gm/l when U/Th was 0.003 in the blanket and the relative power density at the core wall was 1.04 kw/ l/Mw. A 48-in.-diam. core required 1.6 g U/sup 233//l in the core and the relative power density at the core wall was 0.095 kw/l/Mw. The reactors had gross breeding ratios of 1.12 and 1.09, respectively. (auth

Studies of Improvement of Power Density in Orr Loops

Using a simplified model, calculations were made of the possible effects of voids upon the power density in ORR loop experiments. It is concluded that the power density may be markedly increased if voids and channels are plugged with moderator material such as graphite or beryllium. (auth

Pebble Bed Reactor: core physics and fuel cycle analysis

The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes

Neutronic calculations for a new high flux reactor

The Oak Ridge National Laboratory has begun the design of a new high flux reactor to be used for basic research, isotope production, and material irradiation. One of the principal goals of the design is the production of a thermal flux peak in the reflector larger than 5 x 10/sup 15//cm/sup 2/sec. A theoretical analysis of the slowing-down and diffusion of neutrons produced by a spherical fission source immersed in a moderator shows that the flux per unit power is maximized by combining a very undermoderated core with a very low absorbing reflector. The theoretical model interrelates total power, power density and transport properties with the thermal flux allowing very inexpensive scoping calculations. Full scale and detailed calculations were made with a numerical model which uses the Bold Venture code system. Calculations show that a highly enriched /sup 235/U reactor with D/sub 2/O as moderator and reflector would produce the desired peak flux, and the reactor would have a reasonable core life

Thermal reactor design for a neutron scattering source facility

The purpose of this paper is to present conclusions drawn from preliminary neutronic calculations and to present a possible reactor design configuration capable of meeting the above thermal flux level, flux location, and lifetime requirements. (JDB

Fission-product afterheat: a review of experiments pertinent to the thermal-neutron fission of sup235sup 235U

Experimental data relating to beta and gamma energy release rates following thermal-neutron fission of /sup 235/U are reviewed. An afterheat function based on these experiments, for after-shutdown times from 1 s to 5 x 10/ sup 5/ s, is proposed and compared with the proposed ANS Standard ANS-5.1; and the uncertainty in the function proposed here is estimated. This proposed function varies between 0.98 and 1.08 times the proposed standard function; the standard deviation is estimated to be between 10% and 15%. (31 figures, 17 tables, 48 references) (auth

VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport, version II. [LMFBR]

The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently