The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.

Work has continued on the development of accurate nuclear fuel burnup analysis. Work performed during the fourth quarter is summarized here.

BOUNCE III is a program which was written for the IBM-704 as part of a study of the parameters of the neutron distribution in a large thermal column. The program calculates the eigenvalues and corresponding eigenvectors of the matrix resulting from a diffusion-theory, multigroup description of the thermal neutron spectrum.

Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.

The raw materials and fabrication procedures employed in preparing thirty two zirconium alloy compositions for evaluation as described in GEAP-3979 are reported. Considerations involved in the extension of reported laboratory procedures to larger scale production are discussed.

The effect of the (n, 2n) and (n, o<) reactions on the neutron economy of a beryllium moderated reactor is investigated.

This document presents the results of investigations concerned with the conceptual design of a 300 KWe "Capsule" nuclear power plant.

The primary objective of this program is to obtain improved data on the changes in nuclear characteristics with burnup of uranium oxide fuel in a boiling water reactor.

This report covers the S-I-5-B-M fuel irradiation in the GETR Maritime Loop during the first quarter of fiscal year 1962. The data are summarized in Section II.

A design and economic analysis is presented for a 44,000 DWT tanker utilizing a 27,300 shp direct cycle boiling water reactor propulsion system. Included are design descriptions and drawings, estimates of operating and capitol costs for the reference ship, estimates of long range potential, descriptions of required research and development programs, a preliminary safeguards analysis, an analysis of ship's motion on boiling water reactor performance, a summary reactor physics analysis, and a discussion of direct cycle accessibility considerations. Also included are summary descriptions for various design alternatives which were considered.

It is proposed in this report to describe a new two-phase flow mixing length model. It consists of treating the two-phase system as a continuous medium and applying to it the methods and assumptions so widely accepted in single-phase turbulent flow. Because it reduces a two-phase problem to an equivalent one in single-phase flow, the proposed model can be readily used to predict the effects of flow rate, channel geometry, and channel direction in two-phase flow.

Information obtained in a survey of control rod applications and operating experience is reported. This survey covered the Army Nuclear Power Program, the Organic Moderated Reactor Experiment, the Yankee Atomic Power Plant, the Shippingport Pressurized Water Reactor and the Argonne Experimental Boiling Water Reactor.

This report supersedes the similarly titled report dated December 31, 1958. It describes briefly features of the Dresden plant and proposes technical specifications.

Technical report describing that the pressure drop along an annular channel with dimensions D(1) = 0.375 inch; D(2) = 0.875 inch, L = 70 inches. Flow was vertical and upward, and only the internal surface was heated. Subcooled conditions existed at the inlet, with two-phase conditions at the exit. Groups of three radial spacer pins on 18-inch centers along the channel, held the inner surface concentric with the outer surface. The single phase loss coefficient for each spacer group is K(8) = 0.21. The single phase friction factor for the annual channel is given by f = 0.16 N(R)(-0.16). The two phase pressure drop increases as the quality increases for G [over] 10(6) = 0.5 ;b/hr ft(2). The effect of heat flux on the pressure drop is very is very slight over the range of fluxes tested (0.55 less than or equal to Q over 10(6).\ less than or equal to 0.8). The two-phase pressure drop gradient in the same annulus, with no heat addition is qualitatively the same as for a 1/4-inch by 1-3/4 inches rectangular channel but is quantitatively greater than for the rectangular channel.

This is the third quarterly report on the work done under Contract AT(04-30-189, Project Agreement No.27, and covers the period August 12 to November 12, 1962.