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.

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.

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.

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 by the third quarter of 1962 is summarized.

Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.

Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.

Technical report describing a UO2-Mo cermet fuel assembly fabricated for long-term irradiation performance testing in the Vallecitos Boiling water Reactor. The design and fabrication histories of this assembly are described and pre-irradiation data on each individual rod are presented. Molybdenum was added to improve the bulk thermal conductivity of the fuel, so that fuel temperatures would remain comparatively low during high-power level operation of the fuel element. The molybdenum was incorporated into the compacts either as fibers or as a thin coating on individual UO2 particles. Fuel pellets were produced from these materials by vacuum hot pressing. The distribution of the molybdenum in both types of cermet fuels appeared favorable to good heat transfer. The fibers were oriented predominantly in the radial planes of the pellet as a result of the uni-directional compaction during the hot-pressing operation. In the pellets made from the coated particles, a continuous network of molybdenum occurred as a result of the coating welding together during the hot-pressing operation. The test assembly contains eight fuel rods; three contain UO2-Mo cermet, three contain the cermet produced from the coated particles, and two are for reference and contain the conventional sintered UO2 pellet fuel. The nominal outside diameter …

Quarterly progress report discussing activities related to the Vallecitos Boiling Water Reactor (VBWR) and related facilities

Commonwealth Edison Company for the purpose of supplementing the license application for the Dresden Nuclear Power Station submits herewith Amendment No.1 to the portion of the Final Hazards Summary Report entitled Preliminary Hazards Summary Report.

GE is amending its application of 6/5/57 to construct and operate the Nuclear Test Reactor in order to incorporate changes in procedure and equipment.

This report is the second amendment to the Preliminary Hazards Summary Report for the Dresden Nuclear Power Station (GEAP-1044) submitted to the United States Atomic Energy Commission on September 3, 1957.

This report is an amendment to the Preliminary Hazards Summary Report (1) and the Operating Procedures and Emergency Plans (5) for the Dresden Nuclear Power Station, submitted to the United States Atomic Energy Commission on September 3, 1957, and June 5, 1958, respectively.

Summary: Data on equlaxed UO2 grain growth from "out of pile" experiments have been gathered from all known sources and analyzed to determine the relationship between the grain size developed and annealing temperature and between grain size and the time at temperature. On the basis of the analysis, an equation relating gain size to time and temperate has been selected that appears to best describe the data considered as a whole. The coefficients in this grain growth equation have been evaluated to indicate the variance between different investigations and/or different UO2 sinters. The general applicability and limitations of "out of pile" grain growth data for the determination of temperatures in the microstructures of irradiated UO2 are discussed. Specific recommendations are made for the conditions under which grain size can be reliably employed as a temperature indicator. The objective in undertaking this analysis on UO2 grain growth was to obtain a working relationship between UO2 grain size and annealing time and/or temperature, and also a measure of the potential variation in the relationship. The intended use of the results was the determination of temperatures based on the grain sized observed in the post-irradiation metallographic examination of AEC-Euratom High Performance UO2 Program …

Previously reported results on the testing of small samples of boron carbide dispersed in nickel by electrolytic codeposition were adequately encouraging to lead to the development of a prototype control rod for operation in the Vallecitos Boiling Water Reactor. The operation of the control rod has been entirely satisfactory.

A potential performance limitation of superheat fuel is the susceptibility of the fuel cladding to low cycle fatigue failure. Two simplified analytical methods are presented to estimate the cyclic lifetime of circular superheat fuel cladding. One failure relation is based on a displacement method. The other failure relation is based on a stress method. These relations were compared with data from the literature, and with data involving damage obtained by Reynolds. A recommended design procedure involving the relations is presented. The technique was applied to the SADE 4B experiment with moderate success. These cycling relations involve only mechanical damage imposed by cycling, with a modification for additional damage caused by radiation; they do not include any other potential performance limiting mechanisms, such as stress corrosion, which are normally factored into the over-all fuel design. This work work done under Task C (Materials Development) of the Nuclear Superheat Project, AEC Contract AT(04-3)-189 - Project Agreement 13.

From introduction: "This report sets forth the results of a technical and economic analysis of automatic propulsion system control as a possible design improvement in the direct cycle boiling water reactor propulsion system in a T7 tanker."

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.

Tests were run at the General Electric Company, Atomic Power Equipment Department, to determine the burnout conditions for a non-uniformly heated rod in an annular geometry.

This report contains a listing of the bound-proton kernels for neutron scattering which were calculated in conjunction with the USAEC Control Rod Materials Program.

Methods for analyzing chemical impurities in sodium metal samples are presented. Chemical analysis was used to determine the following impurities: Calcium, Carbon, Chromium, Iron, Lithium, Nickel, Oxygen, Potassium, and Zirconium. Spectrographic analysis was used to determine many other impurities. Sodium samples obtained from experimental apparatus operated as part of the work being conducted for Atomics International were analyzed by these methods.

The reason for initiating the compact drive study for the T7 tanker was to investigate control rod drive size, location, and removal space requirement factors and select the control rod drive mechanism which would allow optimization of the over-all size of the containment vessel. Approximately twelve mechanical/hydraulic control rod drive arrangements were considered during this study.

Equipment and techniques for remote examination of irradiated fuel assemblies applicable to the Maritime Program are described. The following subjects are covered: visual and photographic examination, dimensional measurements, gamma activity scanning, fission gas release and fuel rod void volume determinations, density measurements, metallographic examination, and radiochemical burnup analysis.