Methods and procedures are described for removing a radioactive primary sodium pump or a primary cold trap with handling casks from the Hallam Reactor. (D.L.C.)

The hazards evaluation was modified to reflect certain changes made to the equipment as a result of operating experience. These changes included: the addition of a startup interlock circuit; the modification of a startup interlock circuit; several minor modifications to the control rod actuators; and the addition of the tube-sheet cooling system. (M.C.G.)

The nonspherically symmetric solutions to the Bardeen-Cooper-Schrieffer theory are given a physical interpretation in terms of an anisotropic fluid model. These solutions have been used previously to predict a phase transition in liquid by He{sup 3} by Emery and Sessler and Anderson, Morel, Brueckner, and Soda. An investigation of the flow properties of such systems is made that involves the calculation of the effective mass for flow in a straight channel and the moment of inertia of a cylindrical container of the liquid. The angular dependent energy-gap characteristic of this type of theory leads to an effective mass for flow that depends on the angle between the axis of symmetry of the fluid and the direction of flow. It also vanishes as the absolute temperature tends to zero, although not as rapidly as for a spherically symmetric gap. The moment of inertia, when the symmetry direction for the fluid and the rotation axis are the same, is simply related to the mass for flow.

A number of organisms have been examined for their ability to produce electron-spin-resonance signals at low temperatures in response to illumination. The efficiency of the response is of the order of not less than 5%, and the wavelength for maximum response is generally slightly on the longer side of the wavelength of maximum absorption, with a minimum appearing at the wavelength of maximum absorption.

The following report is the final technical report on a series of reports under contract AT(11-1)-515, titled "Niobium Phase Diagrams", covering the period from April 15, 1957 to April 16, 1960.

The results of a multi-group, diffusion type code computation for the fast neutron flux distribution in C, K, and N reactors are presented. Relative Ni activations in conjunction with the calculated neutron spectra are used to obtain values for the integrated fast flux per MWD/AT for C and K reactors. The relative, theoretical, integrated fast flux values at the same adjacent fuel powers in C, K, and N reactors in the mid-plane of the filler layers are: (1) C = 1.0; (2) K = 1.3; and (3) N = 2.6. For C reactor, the best estimate of the integrated fast flux per MWD/AT is: 5.0 {times} 10{sup 16} nvt (> 1 MeV) = 1 MWD/AT.

The high pressure loop installed in the 3X3 reflector position of the ETR and the associated instrumentation to detect and study failure mechanisms handled the rupture tests without difficulty. Failure of the elements was initiated by shearing off a projection on the fuel elements. The first test of the series used previously unirradiated seven-rod clusters. After the projection was sheared off the fuel elements were operated for seven hours with no failure. Failure is defined as having occurred when sufficient uranium oxide has formed to split open the cladding and release large amounts of fission products into the loop water. The second and third tests used fuel which had been irradiated to 2400 MWD/T at Hanford prior to insertion into the ETR. The second test was operated for 14 hours after the projection was sheared off--again with no failure. The third test was operated for only 33 minutes after the projection was sheared off before fission product activity in the loop water caused the test to be terminated.

The objective of this test authorization is to increase reactivity and thus reduce short term enrichment requirements by increasing graphite temperature limits during low exposure operation following full central zone discharge.

Under simplifying assumptions, an expression is obtained relating the integrated kinetic energy transferred to moderator atoms to the thermal neutron exposure in adjacent fuel. Reactor-dependent factors are explicit in the expression which make it possible to compare graphite contraction data from different reactors or to estimate graphite damage in a new reactor based oil observed damage in existent reactors. Graphite temperature, which may be an important factor in graphite contraction, is not considered as one of the variables in this report. Also, the damage dependency on neutron energy is not dealt with in great detail because of the relatively unknown facts in this regard. A particular application of concern at the present is to determine the pertinent factors relating to fast neutron exposure in the NPR to assist in determining the integrated damage over the life of the pile.

Preliminary engineering evaluations indicate that piping in the 105 B, D, F, DR, and H reactors has deteriorated to the extent that an increasing rate of component failure can be expected. In view of this, a budget submission was made in the FY-1962 P. A. and C budget and has been included in the I.P.D. Plant Improvement Program. The purpose of this report is to substantiate the need for this program and to review information generated during the past three years concerning the condition of rear face piping and hardware. This review includes the history of rear face piping and hardware problems, study activities undertaken to date to ascertain the condition of the components, action taken to correct actual component failures, programs recommended to correct deficiencies which operating experience and engineering judgement indicate are necessary, and programs to accumulate additional information to support design of new piping and hardware components.

This production test is designed to demonstrate that the K Reactors can be operated. with a higher graphite temperature limit than stipulated in present standards without a significant increase in the rate of either burnout or contraction of the graphite moderator stack. It is intended that the increase in allowable maximum graphite temperature will be utilized to permit operation with slightly higher reactor power levels with substantially the same enrichment level of the total uranium fuel charge as is currently in practice at the K Reactors.

This monthly document details activities of the Hanford Operations Office during the month of April 1960. (FI)

This document details activities of the Fuels Preparation Department during the month of April 1960. (FI)

From introduction: A thorough investigation, both experimental and theoretical was carried out on plaque type radiators. Five phases of the work are reported: preliminary design analysis, experimental measurements, computational program and analytical studies, followed by a general discussion of source design principles. The results are summarized in graphs and formulae which may be applied to other design requirements.

The objective of this supplement is to determine the effects of operation with neutral pH water on crud deposition on fuel surfaces. This supplement authorizes testing designed to show the effects of coolant pH on the thickness of the crud layer deposited on fuel elements. The supplement will not require additional reactor down time nor will it introduce any hazards to the reactor.

Abstract: A loop has been constructed to study the removal of water and highly volatile materials from Organic Moderated Reactor coolant by vacuum degasification. An analysis of the process was made to determine the most important parameters for study during the experimental program.

Abstract: Accurate measurement of OMRE fuel-element surface temperatures is important for evaluation of the performance of organic compound as moderator and coolant materials in nuclear reactors; a mixture of polyphenyls has been used in reactor operation to-date.

Abstract: The influence of plutonium on the nuclear characteristics of organic moderated cores is studied.

Abstract: Equipment has been developed to handle the fuel elements and control rods for the Piqua (OMR) reactor.With the handling machine, which consists of a shielded cask mounted on a gantry, a fuel element can be replaced in the core in about 27 minutes.

This document presents experimental data obtained during a series of tests which were completed at 105-D and DR Reactors in February and March, 1960. No analysis of the data is included in this document. The tests were: PT-301-I, II -- Reactor cold, full flow, BPA power failure; PT-278-A, III B -- 1170 MW, full flow, BPA power failure; and PT-278-A, II -- 1190 MW, full flow, poison push causing bulk surge and scram.

The neutrons in an operating nuclear reactor are generated primarily by the fission events which are taking place. The great bulk of these relatively high energy or fast neutrons are slowed down or thermalized by a series of elastic collisions with the moderator nuclei which comprise the bulk of the volume of the reactor core. Once slowed down, the neutrons diffuse through the reactor core until they are absorbed or eliminated by some other process. Each of these three groups of neutrons, i.e., the fast or source neutrons, the intermediate or slowing down neutrons, and the slow or thermal neutrons, has a characteristic energy distribution. At a constant power level or rate of fissioning, an equilibrium is soon established among the groups at any point in the reactor. If it is assumed that a smooth transition exists between the different energy groups, it is possible to evaluate the entire neutron spectrum at a point in the reactor by determining the parameters which characterize each of the three groups. This has been done in the F Reactor Quickie Facility using radioactivants.

In the Hanford production reactors, failure of fuel element cladding is known as subcooled burnout since the bulk coolant temperature is below the saturation temperature at time of failure or burnout of the fuel cladding. The heat generation rate at which subcooled burnout occurs, establishes a limit to allowable reactor power level. This document compares present and potential power level limits at the Hanford production reactors with power level limits established by subcooled burnout.

Introduction: Development of a process for manufacturing fuel elements for the Yankee reactor and fabrication and testing of a prototype assembly is reported.

A resume of Health Physics activities for January, February, and March, 1960 is presented. Discussions and tabulations which summarize results of field surveys, bioassay, personnel monitoring, and environmental surveys are included.