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- Accurate Nuclear Fuel Burnup Analyses: Fifth Quarterly Progress Report December 1962-February 1963
- Work has continued on the development of Accurate Fuel Burnup Analysis. Work performed during the fifth quarter is summarized here.
- Accurate Nuclear Fuel Burnup Analyses: Fourth Quarterly Progress Report September-November, 1962
- Work has continued on the development of accurate nuclear fuel burnup analysis. Work performed during the fourth quarter is summarized here.
- Accurate Nuclear Fuel Burnup Analyses: Third Quarterly Progress Report June - August, 1962
- Work has continued on the development of accurate nuclear fuel burnup analysis. Work performed by the third quarter of 1962 is summarized.
- Amendment No. 1 to Preliminary Hazards Summary Report For The Dresden Nuclear Power Station
- 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.
- Amendment No. 2 To License Application For Nuclear Test Reactor
- 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.
- Amendment No. 2 To Preliminary Hazards Summary Report For The Dresden Nuclear Power Station
- 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.
- Amendment No. 4 to Hazards Summary Report For The Dresden Nuclear Power Station
- 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.
- Application of Boron Carbide Nickel Dispersion to a Prototype Control Rod
- 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.
- Bounce III
- 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.
- Burnout Conditions for Nonuniformly Heated Rod in Annular Geometry, Water at 1000 PSIA
- 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.
- Calculated Scattering Kernels For Light Water at 23C, 42C, 61C, and 82C
- 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.
- Collected Methods for Analysis of Sodium Metal
- 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.
- Compact Control Rod Drive Study For a Boiling Water Reactor in a T7 Tanker
- 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.
- Compilation of Techniques Used By Vallecitos Radioactive Materials Laboratory
- 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.
- Consumers Baffle Two-Phase Air-Water Flow Tests in a One-Fifth Scale Model
- Tests in a one-fifth scale clear plastic model were conducted to investigate the flow characteristics of the asymmetrical riser configuration in the Consumers Big Rock power plant.
- Consumers Big Rock Point Nuclear Power Reactor Stability Analysis
- This report presents the results of an analysis which was undertaken to investigate the power stability of the Consumers Big Rock Point Nuclear Power Reactor.
- Control Worth of B4C Rods
- This report considers the theoretical evaluation of a system for gaining increased control strength and increased control lifetime and presents a theoretical model which is applicable to conventional multigroup diffusion theory.
- Design Report: Superheat Strain - Cycle Capsule
- In order to investigate the low frequency strain cycle fatigue for tubular sheath geometries an apparatus was designed and fabricated for laboratory and reactor experiments. The design of this apparatus is described herein.
- The Determination of Fission Product Gamma Doses
- In this paper arbitrary limits of the general fission source gamma problem are set. Then, by assuming cooling of at least one day, it is shown that only twelve different fission product gamma sources need ever be considered.
- The Development of a Scheduling Computer For The Big Rock Plant
- The basic work for development of a scheduling computer for the Consumers' Big Rock Plant is outlined in this report. The computer's purpose is to make feasible higher power densities by operation closer to fuel element burnout limits, and to maximize fuel burnup.
- Development Program For Increased Output In The Garigliano Nuclear Reactor : Quarterly Progress Report No. 1 September-December, 1962
- Authorization from the U.S. Atomic Energy Commission has been received to begin Pressure Vessel Sample Irradiation and Data logging and Computer System of the Development Program for Increased Output in the Garigliano Nuclear Reactor. The work described herein was performed in the period of September 1 to December 31, 1962.
- Economic Evaluation of Control Rod Materials and Fabrication Processes
- Control rod materials, designs, and fabrication processes are compared for their relative economies. Control rod lifetime data are calculated with a simple approximation of nuclear worth depreciation. These data are used in conjunction with the estimated fabrication costs to determine the cost of using several absorber materials in a typical power reactor on a cost-per-year basis. The effect the control system has on core power density and fuel lifetime is included.
- Enclosure Pressure Calculation Method
- A method of determining enclosure pressure in the event of a reactor rupture is presented and a sample calculation is shown. This method was used in calculating the design pressure of the Dresden Nuclear Power Station enclosure.
- Enclosure Section of the Hazards Summary Report for the Dresden Nuclear Power Station
- The General Electric Company is designing and building a 180,000 kilowatt nuclear power plant for the Commonwealth Edison Company at a site near the confluence of the Kankakee and Des Plaines Rivers in Grundy County, Illinois, about 47 miles southwest of Chicago. The plant will be known as the Dresden Nuclear Power Station, and will employ a nuclear reactor of the dual-cycle boiling water type.
- Erosion Experiments of Powder Compacted Uranium Dioxide Under Dynamic Steam Flow (Preliminary Report)
- Experiments were carried out to determine the erosion, oxidation and dimensional characteristics of purposely defected fuel elements containing unsintered UO2 powder prepared by the swaging technique. The experiments were conducted in an out-of-reactor loop under superheat conditions of pressure, temperature, flow velocity and steam chemical composition.
- Experimental Fast Ceramic Reactor Design Status Report as of October 31, 1961
- The design status of the Experimental Fast Ceramic Reactor (EFCR) is described for the period up to October 31, 1961. The primary purpose of the facility is to study the dynamic behavior of a fast ceramic reactor, including the experimental demonstration of the effectiveness of the Doppler coefficient in limiting the power excursion following a rapid insertion of reactivity.
- Experimental Investigations of the Removal of Sodium Oxide From Liquid Sodium
- Experimental investigations were conducted to obtain additional information on the growth and characteristics of sodium oxide deposits in liquid sodium which could lead to system plugging. These investigations included the removal of sodium oxide from molten sodium by the cold trap method.
- Fabrication of Zirconium Alloys For Specific Zirconium Alloy Design Program
- 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 Fast Effect in a Beryllium Moderated Reactor
- The effect of the (n, 2n) and (n, o<) reactions on the neutron economy of a beryllium moderated reactor is investigated.
- Final Report: 300 KWe Capsule Nuclear Power Plant Study
- This document presents the results of investigations concerned with the conceptual design of a 300 KWe "Capsule" nuclear power plant.
- Final Report Chicago Operations Office Steam Separation Program, February-March, 1960
- A program of free-surface steam-water separation at 600 psig was carried out in support of the design for operation of the Experimental Boiling Water Reactor at 100 MW thermal. Reduced test data are presented without evaluation since the ANL technical representative received an original copy of the data for evaluation with respect to the EBWR design.
- Final Summary Safeguards Report For The General Electric Test Reactor
- This report is submitted to the U. S. Atomic Energy Commission as a final summary safeguards and hazards evaluation of a proposed test reactor at its Vallecitos Atomic Laboratory in Alameda County of California. It is the purpose of this report to provide sufficient data to obtain an AEC facility license for the reactor.
- FORE - A Computational Program For The Analysis Of Fast Reactor Excursions
- A digital computer program, FORE, which calculates transients for a fast reactor, has been coded for the TRANSAC 2000 computer. Its purpose is to provide understanding of the dynamics of fast reactors with particular emphasis on large ceramic-fueled fast reactors. The program calculates reactor power and temperatures of fuel, coolant, clad, and structure.
- Frequency Response of Weighted Voids VS. Power
- A method for calculating the frequency response of weighted voids (proportional to reactivity of steam voids) as a function of reactor power is presented.
- Fuel Cycle Program, A Boiling Water Reactor Research and Development Program Eighth Quarterly Progress Report April 1962 - June 1962
- The Fuel Cycle Program is an integrated program of investigation in the Vallecitos Boiling Water Reactor (VBWR) and other facilities to improve the technological limits of boiling water reactors in several areas. This report presents updates on tasks related to those areas.
- Fuel Cycle Program: A Boiling Water Reactor Research and Development Program - First Summary Report, March 1959-July 1960
- The Fuel Cycle Development Program is a basic development program of boiling and other water reactor technology. This is the first report to the U.S. Atomic Energy Commission of activities carried out under this program. It constitutes a review of the progress from April 27, 1959 through July 31, 1960.
- Fuel Cycle Program - A Boiling Water Reactor Research and Development Program: Ninth Quarterly Progress Report July 1962 - September 1962
- The Fuel Cycle Program is an integrated program of investigation in the Vallecitos Boiling Water Reactor (VBWR) and other facilities to improve the technological limits of boiling water reactors.
- Fuel Cycle Program - A Boiling Water Reactor Research and Development Program: Seventh Quarterly Progress Report January 1962 - March 1962
- The Fuel Cycle Program is an integrated program of investigation in the Vallecitos Boiling Water Reactor and other facilities to improve the technological limits of boiling water reactors in several areas. Progress is reported here.
- Fuel Cycle Program - A Boiling Water Reactor Research and Development Program: Tenth Quarterly Progress Report October-December 1962
- The Fuel Cycle Program is an integrated program of investigation in the Vallecitos Boiling Water Reactor and other facilities to improve the technological limits of boiling water reactors in several areas.
- Fuel Cycle Program - A Boilng Water Reactor Research and Development Program. Fifth Quarterly Report, July 1961-September 1961
- This report summarizes progress on investigation into improving the technological limits of boiling water reactors, Vallecitos Boiling Water Reactor and other facilities.
- Further Experimental Results on Natural Circulation Loop Performance at 1000 psia Under Periodic Accelerations
- Experimental results on the effect of periodic acceleration on a natural circulation, 1000 psia, two-phase flow loop are presented, and related to the topic of marine reactor design. The initial results of this work were given previously in GEAP 3397.
- General And Stress Corrosion Of High Nickel Alloys In Simulated Superheat Reactor Environment
- It is the purpose of this report to summarize the results of the evaluation program carried out to-date on the high nickel alloys Inconel, Incoloy and Hastelloy-X in the out-of-pile superheat facilities as part of Task E of the Atomic Energy Commission sponsored Superheat Program.
- General And Stress Corrosion Of High Nickel Alloys In Simulated Superheat Reactor Environment
- It is the purpose of this report to summarize the results of the evaluation program carried out to-date on the high nickel alloys Inconel, Incoloy and Hastelloy-X in the out-of-pile superheat facilities as part of Task E of the Atomic Energy Commission sponsored Superheat Program.
- The Half-Life and Gamma Ray Abundance of Cs-137
- The nuclide Cs-137 is a fission product commonly used for measurement of uranium burnup in irradiated uranium fuel by the fission product to uranium ratio method. In the application of this method, the largest single error introduced in the measurement of burnup is the uncertainty in the half-life of Cs-137. Because of the uncertainty in this value and its importance in nuclear fuel burnup analysis, a reinvestigation was undertaken to obtain a more accurate value using the mass spectrometric method.
- Heavy Element Isotopic Analysis of UO2 Fuel Irradiated In The VBWR: Report #1
- 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.
- High Performance UO2 Program First Quarterly Progress Report: April-June 1961
- A better understanding of the maximum operating characteristics that can be achieved with the use of UO2 as a reactor fuel is the primary purpose of this program for Euratom and the Atomic Energy Commission. During this program work will be undertaken in two areas that have been of concern to the reactor core designer for a long time, viz. fission gas release and central melting in fuel rods.
- High Performance UO2 Program Fourth Quarterly Progress Report
- This phase of the program is concerned with irradiation of fuel assemblies to determine central temperature limitations.
- High Performance UO2 Program Quarterly Progress Report Number 2: July-September 1961
- The primary purpose of this joint USAEC-Euratom program is to obtain a better understanding if the maximum achievable operating characteristics of UO2 as a reactor fuel. During the program work will be performed in two areas that have been of concern to reactor core designers for a long time, namely fission gas release and central melting in fuel rods.
- High Performance UO2 Program Quarterly Progress Report No.5: April-June 1962
- Work performed during the quarter is summarized by direct measurement of fission gas pressure, loop modifications, UO2 fuel performance
- High Performance UO2 Program Quarterly Progress Report No. 6 : July-September 1962
- Work performed during the quarter is summarized by direct measurement of fission gas pressure, loop operations, performance of UO2 fuel.