MULTIGROUP DIFFUSION THEORY FORMULATION OF THE CALCULATION OF THE MEAN SQUARE SLOWING DOWN DISTANCE IN AN INFINITE MEDIUM

Simple expressions for the mean square distance from a point fission source for slowing down past a given energy and for the mean square distance of neutrons that belong to a given energy group are derived within the framework of multigroup diffusion theory. The expressions may be applied to systems having arbitrary group transfer cross sections. (auth

THE COUPLED ASPECTS OF A FAST-THERMAL CRITICAL: ZPR-V

ZPR-V is a zero-power, coupled fast-thermal critical assembly involving a dilute fast core and a H/sub 2/O-moderated, enriched U thermal annulus. The two regions are coupled through a natural U isolation blanket. This facility is useful in studying some of the properties of dilute fast reactor systems as well as those of coupled fast-thermal power breeders. Recent emphasis has been on the latter aspect of the problem and an investigation of the coupled aspects of the system has been made. An important parameter in the coupling concept is the division of reactivity. This can be determined by means of fast fuel substitutions and 1/v poisoning of the thermal annulus in conjunction with theoretical calculations. Auxiliary information obtained in such a procedure is the fast fuel worth throughout the core and the prompt neutron lifetime of the whole system. A second important pnrameter, S/sub 1/S/sub 2/, is the relative number of neutrons emitted per unit time due to fast and slow fissions. Information regarding this quantity can be obtained with the aid of U/sup 235/ and U/sup 238/ fission distributions. Measurenments of thc above parameters were made for the case of a U/sup 238/:U/sup 235/ atomic ratio of 5:1 in the fast core and a 1-in. isolation blanket. The fractions of reactivity in the fast core, blanket, and annulus are, respectively, 24, 14, and 61%; the prompt lifetime is 39.0 plus or minus 3.0 mu sec; the ratio S/sub 1/S/sub 2/ is 0.43. (auth

ETOE-2/MCsup2sup 2-2/SDX multigroup neutron cross-section processing

The ETOE-2/MC/sup 2/-2/SDX code system provides a comprehensive neutron cross-section processing capability. The three components of the system include an ENDF/B library processor (ETOE-21, an ultra-fine-group fundamental mode calculation (MC/sup 2/-2), and an intermediate-group space-dependent capability (SDX). Brief descriptions of the types of calculations are discussed. Selected results from typical problem executions are presented. (auth

AN EVALUATION OF REACTOR CONCEPTS FOR USE AS SEPARATE STEAM SUPER-HEATERS

Various reactor concepts were compared for use as a separate superheater which could be added on to an advanced 300-Mwe reactor producing saturated steam. Fossil steam plant superheat temperatures were used as a criterion for selecting nuclear superheat temperatures. Therefore, the performance specified for the superheater was a minimum exit steam temperature of 566 deg C (1050 deg F) when supplied with saturated steam at either 7l atm (l050 psia) or l67 atm (2450 psia). A preliminary screening of ten different reactor concepts resulted in the selection of two for a detailed evaluation. These are a direct-cycle, watermoderated reactor, and an indirect-cycle, sodium-cooled reactor. The steam- cooled, water-moderated system is judged to have the best chance for initially reaching 566 deg C (l050 deg F), whereas, the indirect-cycle, sodiumcooled system is considered best for subsequent advances to exploit the more efficient, high- pressure steam reheat cycles. A design concept was selected for each of the reactors to establish a basis for the detailed evaluations and comparisons. The technical evaluation of the two concepts shows that the sodium-cooled, indirect- cycle superheater can realize a significant reductlon in the gross plant heat rate STA0.6097 to 0.5404cal/watt-sec (87l0 to 7720 Btu/kwh)! by use of the l64- atm (2400 psig) reheat steam cycle. This is not feasible with the water- moderated superheater, because of an assumed limitation of 71 atm (l050 psia) on large-size pressure vessels for reactors. This restricts the direct-cycle superheater to the use of 69-atm (l000psig) turbines without reheat. On the other hand, to reach the specified steam exit temperature of 566 deg C (l050 deg F), the direct-cycle concept requires less extrapolation of current technology. Oxide fuel is suitable for the direct-cycle reactor but carbides are necessary to reach sufficiently high sodium temperatures with the indirect-cycle reactor. Both concepts need fuel cladding of high creep strength; however, the problems of corrosion and erosion are alleviated in a sodium-cooled system. Cost estimates give an advantage to the water-moderated, direct-cycle system although both show a net reduction in power costs when added as separate superheaters to a boiling water reactor. The significance of a size factor for plants in the 500-Mwe range is illustrated by comparisons made in this study. Power costs for the direct- cycle superheater, boiling water reactor combination were almost identical with those of an integraltype water-moderated direct-cycle system. This is because two of the integral-type units are required to match the power (575 Mwe) of the separate superheater-boiler combination. Conversely, with the sodium system, a separate sodium-cooled superheater, sodium-graphite boiler combination gives higher power costs than an integral-type sodium system because only one reactor is required for the latter in the 500-Mwe range. This study was centered on an approach to cheaper nuclear power through higher over-all thermal efficiency. To reach the ceiling now imposed by steam turbine technology, higher temperatures are indicated for the sodium-cooled superheater, since it is already amenable to the use of a high-pressure, reheat steam cycle. Alternatively, the direct-cycle, steam-cooled superheater (in the light of current development programs) is closer to the current limit to turbine temperature so that the major degree of freedom remaining is increased pressure for improved cycle efficiency. (auth