Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG

Low-energy electron transport with the method of discrete ordinates

The one-dimensional discrete ordinates code ANISN was adapted to transport low energy (a few MeV) electrons. Calculated results obtained with ANISN were compared with experimental data for transmitted electron energy and angular distribution data for electrons normally incident on aluminum slabs of various thicknesses. The calculated and experimental results are in good agreement for a thin slab (0.2 of the electron range), but not for the thicker slabs (0.6 of the electron range). Calculated results obtained with ANISN were also compared with results obtained using Monte Carlo methods

Sensitivity and uncertainty investigations for Hiroshima dose estimates and the applicability of the Little Boy mockup measurements

This paper describes sources of uncertainty in the data used for calculating dose estimates for the Hiroshima explosion and details a methodology for systematically obtaining best estimates and reduced uncertainties for the radiation doses received. (ACR

Analysis of gas-cooled fast reactor shield designs

In its shielding program for the Gas-Cooled Fast Reactor (GCFR) as conceived by General Atomic, Oak Ridge National Laboratory has developed an advanced shielding analysis system that incorporates the latest analysis techniques for converging to a shield design compatible with other design parameters such as cooling and structural requirements or material compatibility. Basically the system consists in applying the various techniques in a logical sequence to a given design, thereby generating a large body of data to serve as an information base for subsequent redesign. As an illustration, this system is applied to successive typical models for the GCFR, resulting in a reduction in the thickness of the radial shield and redesign of the lower shield region. In principle, the design-analysis-redesign iterations would continue until they converge upon an acceptable configuration

Preliminary cross-section sensitivity analysis for an air-over-ground environment

Two-dimensional sensitivity calculations were made for an air-over- ground geometry to determine the effect of air and ground cross-section perturbations on the total neutron and gamma ray dose near the air/ground interface and 415 meters above the ground. ENDF/B 22 neutron and 18 gamma group Version IV cross-sections were used in all computations. (auth

Accelerator breeder nuclear fuel production: concept evaluation of a modified design for ORNL's proposed TME-ENFP

Recent advances in accelerator beam technology have made it possible to improve the target/blanket design of the Ternary Metal Fueled Electronuclear Fuel Producer (TMF-ENFP), an accelerator-breeder design concept proposed by Burnss et al. for subcritical breeding of the fissile isotope /sup 233/U. In the original TMF-ENFP the 300-mA, 1100-MeV proton beam was limited to a small diameter whose power density was so high that a solid metal target could not be used for producing the spallation neutrons needed to drive the breeding process. Instead the target was a central column of circulating liquid sodium, which was surrounded by an inner multiplying region of ternary fuel rods (/sup 239/Pu, /sup 232/Th, and /sup 238/U) and an outer blanket region of /sup 232/Th rods, with the entire system cooled by circulating sodium. In the modified design proposed here, the proton beam is sufficiently spread out to allow the ternary fuel to reside directly in the beam and to be preceded by a thin (nonstructural) V-Ti steel firThe spread beam mandated a change in the design configuration (from a cylindrical shape to an Erlenmeyer flask shape), which, in turn, required that the fuel rods (and blanket rods) be replaced by fuel pebbles. The fuel residence time in both systems was assumed to be 90 full power days. A series of parameter optimization calculations for the modified TMF-ENFP led to a semioptimized system in which the initial /sup 239/Pu inventory of the ternary fuel was 6% and the fuel pebble diameter was 0.5 cm. With this system the /sup 233/Pu production rate of 5.8 kg/day reported for the original TMF-ENFP was increased to 9.3 kg/day, and the thermal power production at beginning of cycle was increased from 3300 MW(t) to 5240 MW(t). 31 refs., 32 figs., 6 tabs

GCFR radial blanket and shield experiment: objectives, preanalysis, and specifications

An integral experiment has been designed for the verification of radiation transport methods and nuclear data used for the design of the radial shield for the proposed 300 MW(e) gas-cooled fast breeder reactor (GCFR). The scope of the experiment was chosen to include a thorium oxide radial blanket mockup as well as several shield configurations in order to reduce the uncertainties in the calculated source terms for the radial shield, and to reduce the uncertainties in the calculated radiation damage to the prestressed concrete reactor vessel (PCRV). Additionally, the measurements are intended to bound the uncertainties in calculated gamma-ray heating rates within the blanket and shield. Although designed specifically for the GCFR, the experiment will provide generic data regarding deep penetration in ThO/sub 2/ and common shield materials, which should also benefit LMFBR designers