Drag coefficients were determined for various irregular objects such as glass fragments, stones, steel fragments, and spheres by means of drop tests for use in a mathematical model to correlate nuclear explosion blast experiments. Drop tests were also made on small laboratory animals and extrapolated to estimate the drag properties of man. A method was developed to estimate the average drag properties of man from his total surface area. (D.L.C.)

The authors have previously reported that deuterium oxide in the drinking water of either male or female mice produces sterility. An investigation of some of the conditions--with particular reference to time--of deuterium oxide treatment to produce sterile C{sub 57} male mice indicated that the sensitive phase of sperm production centered around the late prophase of meiosis. In some experiments, although D{sub 2}O was almost completely absent during maturation of the sperm, and when the mice mated, these sperm exhibited the effects of their much earlier contact with D{sub 2}O. No viable offspring were obtained from these matings. They concluded that the presence of D{sub 2}O during the late prophase and meiotic divisions interfered with the normal construction or division (or both) of genetic material. It was suggested that changes in the forces, principally hydrogen bonds, in macromolecules affected their structural characteristics and resulted in abnormal division. The objective of the experiments reported here was to determine the phases of embryonic development of the mouse at which the lethal action of deuterium oxide on sperm is manifested. These investigations on embryonic growth initiated by sperm developed in D{sub 2}O have yielded additional evidence that D{sub 2}O severely damages the genetic material …

Angular distribution and kinetic-energy spectra of fragments, and cross sections for fission of gold with 68- to 124-Mev C{sup 12} ions have been obtained by observation of the fragments in two types of detectors, gas scintillation chambers and silicon p-n junctions. From the parameters used to fit the angular distributions to the theoretical curves of Halpern and Strutinski, we have obtained the average excitation energy of the fissioning nucleus at the time of fission. This quantity is approximately 25 Mev, which is nearly independent of bombarding energy, suggesting that fission is preceded by the emission of several particles from the compound nucleus. The fission cross section increases from a value of 100 mb at 68 Mev to 1.28 b. at 124 Mev. Over this range of bombarding energies, the total fragment kinetic-energy release rises from 142 {+-} 6 to 146 {+-} 6 Mev. At all bombarding energies, the variation of laboratory-system kinetic energy of the fragments with laboratory-system angle indicates full momentum transfer by the bombarding particle to the fissioning system.

To permit the irradiation of one dummy fuel element assembly for one operating period and to permit, during a subsequent operating period, the irradiation of one defected, four-rod-cluster UO{sub 2} fuel element assembly, in a KE front-to-rear test hole. The fuel material is natural UO{sub 2} of 95 per cent theoretical density; the cladding is zircaloy. The defect in the assembly is artificial and will be made before irradiation by drilling a .005in. diameter hole through the cladding near the mid-point of two of the rods.

Failure of reactor rear face connectors is a problem at the reactors. The vibration has been studied at H and F Reactors to assist in the development of a permanent replacement for these connectors. Vibration data were otained from rear nozzles, connectors, and crossheaders within the process tube pattern for the following operating conditions: full cold water flow prior to reactor startup, normal operating power level, and during transition from shutdown to normal operating power level. The vibration patterns at 105-F and 105-H were similar in magnitude. Oscillograph data are presented.

This report documents analysis of radionuclides from reactor effluent water samples. Separation from the sample matrix is accomplished by successive carbonate, hydroxide and fluoride precipitations. Experimental data is provided.

The cap-spire pulsing technique of preheating the cap-spire portion of the fuel assembly does significantly improve the brazing of the cap-spire assembly. The air pocket at the spire wafer junction is fully removed. The cap side wafer is essentially 100% wetted with brazing alloy. Destructive tests show that a 4 to 1 improvement in most quality measurements is achieved over present cap preheating techniques, without using additional cleaning step of spire etching. The pulsing is accomplished by a cammed drive system, using a stroke of three-fourths of an inch with a spring return. The system is driven by an electrical gear reduction motor at a rate of 1.4 pulses per second. A preheating cycle of 21 {plus_minus} 2 seconds is used for the current I&E cap designs. The cap-spire assembly does not require any special treatment other than the normal chemical cleaning.

Goal exposure and power level planning and administration on either a short-term or long-term basis, is heavily dependent upon assessment of the level of metal performance. It is the purpose of this document to present a method of monitoring the metal performance level employing the concept of control charts, and to propose adoption of this method as the basis for detecting changes in the metal performance level of the three main types of production fuel elements, i.e., C-II-N, O-II-N, and K-III-N material. Supplemented with predetermined decision rules for evaluating the extent to which the metal performance level has improved or declined, these control charts become valuable tools for use in making IPD operating decisions.

Ex-reactor high temperature, high pressure recirculating water loop defect-tests have been made on SPR size coextruded fuel tubes. The behavior in terms of fuel corrosion loss and fuel shape distortion for ``pin-hole`` type defects has been determined. Effects of annular spacing on the defect-test behavior and the influence on the mechanical and metallurgical conditions of adjacent components has been observed.

Enclosed is the data that was collected during the recent thermal cycle of KER-1. This data includes the RTD readings and information taken from the thermocouple fuel element. In conjunction with obtaining data on film build-up, the test provided an excellent opportunity to study the effect of temperature on measured tube power. It has been noticed that the KER reactor tubes appear to produce more power (greater {Delta}T) at lower operating temperatures, but the reason for this power increase is not known. With the additional thermocouples to measure water temperatures at the center of the tube, it should be possible to determine if this increase is from the graphite or from the fuel elements.

Changes in operation and corresponding changes in the reactivity status of Hanford reactors are the result of a continuing effort to improve operating efficiency. Trends data related to these changes in operation and reactivity have been published previously for the periods from 1950 through 1958. The purpose of this report is to include trends data for 1959. All included data were selected from the sources listed as references on page 48. Bar graphs in the first part of the report show yearly averages of selected data, and tables in the last part of the report show maximum, average, and minimum values. This document presents trends data for C, H, KE, and KW reactors while a second document, HW-64931, presents trends data for B, D, DR, and F reactors. Data included in past years which have not been included in this report are trends in pile power level at shutdown omitted due to a security status change regarding power levels. Potential non-equilibrium gains and potential equilibrium gains,data were added in this report. Notice that all reactivity values are listed in the unit per cent excess k.

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

None

The objective of the production test detailed in this report is to evaluate Zr-5 w/o Be brazed end closures on enriched tube-and-tube elements irradiated in high temperature coolant in the KER Loops.

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

None

This report presents provisional specifications for the modifications and additions to the Lead Dip Canning Process for Heat Treated Uranium for the production test fabrication of projection fuel elements by resistance spot welding.

The purpose of this test is to establish the reactor hydraulic flow vs pressure demand curve in the low-regions of around 4--10,000 gpm.

The DR-1 Loop, located in the C test hole of the DR Reactor, provides a high temperature, recirculating gas-cooled facility for the irradiation of experimental fuel elements. The loop is being utilized currently by General Atomic, a Division of General Dynamics, to evaluate fuel elements in support of their work on the Maritime Gas-Cooled Reactor Program, a program which is directed at the development of a ship propulsion unit consisting of a gas-cooled reactor driving a closed cycle gas turbine. The loop irradiations for this program require that the experimental fuel elements be maintained at specific test conditions. It is also necessary that all of the loop components be kept within certain operating limits. Therefore, the power generation rate of each experimental fuel element must be evaluated and established as accurately as possible prior to insertion in the loop. One method of establishing the enrichment required to obtain a required heat generation rate in an experimental element is to irradiate a nuclear mock-up of the assembly in the Hanford Test Reactor to determine the relative neutron density within the assembly and the reactor. This report presents the results of such irradiations using the MGCR-III mock-up.

The objective of this test is to correct the fuel depression caused by 3X balls lodged in graphite by charging compensating enrichment in the depression region.

The feasibility of contoured I&E cans for production use has been demonstrated using our present flat base I&E sleeve (HW-37187). Studies by Process Engineering and Quality Control have shown that only a material savings would result from the use of only the contoured I&E can. Consideration was then given to the use of contoured sleeves (H-3-16879) to improve the contact areas and the resulting heat transfer during fuel assembly.

The rupture model presently being used by IPD Research and Engineering is the Weibull cumulative distribution function which in the region of interest, the model can be expressed approximately as: F(X) = failure rate in number of ruptures per total number of tubes, P = average tube power, T = average tube outlet temperature, M = exposure a{sub o}, a{sub 1}, a{sub 2}, a{sub 3} = parameters determined from rupture experience. F(X) = a{sub o}P{sup a1}T{sup a2}M{sub a3}. The parameters a{sub 1} = 3.3, a{sub 2} = 8.7, and a{sub 3} = 5 were estimated from experience with solid slug side-other and side-hot spot failures occurring from the spring of 1956 to the end of 1957, and were found to hold quite well for experience through 1958 as well. The model using the aforementioned estimates of the a{sub 1}, a{sub 2} and a{sub 3} parameters is being used in numerous ways, including fuel performance measurement (estimating values of a{sub o}, in essence), reactor optimization, project justification, production scheduling and forecasting, and fuel element test design. From time to time, doubt has been expressed about the appropriateness of applying these estimates of the a{sub 1}, a{sub 2} and a{sub 3} parameters, …

To permit the irradiation of one dummy fuel assembly for one operating period and to permit, during a subsequent operating period, the irradiation of one defected, four-rod-cluster, UO{sub 2} fuel element assembly, in a KE front-to-rear test hole. The fuel material is natural UO{sub 2} of 95 per cent theoretical density; the cladding is zircaloy. The defect in the assembly is artificial and will be made before irradiation by drilling a .005in. diameter hole through the cladding near the mid-point of two of the rods.

Until recently, bond adherence between the Al-Si braze layer and the uranium core was estimated by a method known as the ``chisel test.`` Fuel element jackets were removed from the core using a hammer and chisel, and the relative bond strength was estimated on a comparison basis. Bonds were classified as normal, below normal, or better than normal, depending upon the ease with which the jacket was removed. The weakest bonds were usually found one-to-two inches below the fuel element cap, but on occasions extended as far as five inches below the cap prior to making process changes in the duplex canning bath to improve bonding. After duplex furnace conditions were changed (silicon concentration, core submersion depth, and temperature) during September 1958, the weak bond area was reduced to a band 1/4 inch to 1 inch below the cap end of the piece. Recently, the stud-pull method for determining bond strength was developed providing a better method for measuring bond quality. Bond strength measured by the stud-pull method provides actual tensile strength measurements, which are valuable for test purposes and detecting in-process variations. This report summarizes tests made using the Instron Tensile Strength Machine for determining fuel element bond strength.