A tabulation is presented of values for selected properties of lithium hydride. (M.C.G.)

The intermediate energy neutron flux existing outside the biological shielding of reactors has not been studied to any great extent previous to this time, because of the lack of an instrument capable of detecting neutrons in the intermediate energy range. The instrument used at the MTR utilizes polyethylene spheres of various sizes to give different amounts of moderation and absorption to the impinging neutrons. A procedure for the approximate determination of the relative number of intermediate energy and fast neutrons is given. By graphical methods the following information is obtained: (1) fraction of intermediate neutrons, (2) fraction of fast neutrons, and (3) the approximate average energy of the fast neutrons. Since the instrument described can be used to determine the thermal neutron flux independent of intermediate and fast fluxes, only one instrument is required to determine the neutron flux in all three energy ranges. Dose calculations indicate the intermediate range neutrons give a dose greater than the dose delivered by fast neutrons around the MTR-ETR reactors under normal operating conditions. (auth)

A method is described for continuously varying the Van de Graaff ion beam energy while the count rate from the decay of the compound nucleus is monitored and recorded. The method utilizes a synchronous motor system to synchronize the recorder and the magnetic field control of the Van de Graaff analyzing system. Use of this method results in significant time saving in calibration of the Van de Graaff accelerator and the determination of target thicknesses. (auth)

Studies the prototype and production air pulse units.

Several safety parameters are calculated for a large, fast, Na-cooled, oxide-fueled reactor. The Doppler coefficient is studied as regards its effect on the reactivity during a power excursion, and its dependence on the neutron spectrum, on the concentrations of U/sup 238/, Pu /sup 239/ , a nd Pu/sup 240/, on spatial temperature and power distributions, and on temperature. Other parameters studied include the Na temperature coefficient of reactivity, and the reactivity insertion caused by total Na loss; reactivity coefficients caused by thermal expansion of fuel and steel cladding; the excess operating reactivity; and the reactivity caused by a fuel slump. The effects of a Be reflector on neutron lifetimes are determined. The safety aspects are balanced against economic considerations. (T.F.H.)

The heat exchanger and steam generator for the U. S. Atomic Energy Commission Sodium Components Project will be constructed entirely of type 316 stainless steel. Because of the susceptibility of this alloy to stress corrosion cracking, it is proposed to clad all areas of the steam generator with Inconel where the stainless steel will be exposed to water and steam. A discussion is presented of the work to justify the selection of Inconel for this service. A discussion of Inconel type welding alloys is also included. (auth)

None

This is the monthly report for the Hanford Laboratories Operation, February 1961. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

The objective of this production test was to evaluate the effects of Zr-2 jacket uniformity, thickness, and temperature on non-uniform clad straining. Test details are provided.

This report summarizes the evaluations of the maximum monetary liability associated with cesium-137 shipments in the STT Casks and strontium-90 shipments in the HAPO-IA Cask. These evaluations are of most immediate importance since these shipments are planned for the month of March 1961. These liability evaluations concern the direct consequences of a release of the fission product shipment in each case, but do not cover the probability of such events occurring. The liability evaluation of the STT shipment of cesium-137, 90,000 curies per car, was based upon the following premises. Cesium is a relatively volatile material, and the entire shipment quantity can be vaporized and released as a cloud during involvement of the car in a fire of sufficient magnitude and duration to rupture the cask. This mechanism would create, by far, the greatest potential exposure of personnel and property to the fission product.

This report presents an analysis of rupture experience at C Reactor over the past two years. The purpose of the study was to provide a basis for revising, if necessary, the current metal performance level multipliers for C-II-N and C-II-E material.

The objective of this production test is to evaluate the behavior of large diameter tubular UO{sub 2} fuel elements during high temperature irradiation. Eighteen inch long tubular UO{sub 2} fuel elements 1.804 inch OD, 0.544 inch ID, in 0.060 inch Zircaloy-2 jackets will be irradiated in the KER loops either alone or in conjunction with other tests to exposures up to 3500 MWD/T of contained uranium.

This document provides the final report on the research project of chemical milling of counterbore recesses in the uranium wall of Zircaloy-2 clad uranium tubes. The scope of work, results and conclusions, and experimental procedures are provided.

C Reactor was shut down on a scheduled basis at 8:30 a.m. March 6, 1961 for the purpose of overboring 17 process channels. this report will cover that outage and discuss problems encountered in completing the tasks involved in overboring.

This report contains questions and answers on topics in reactor physics.

This report summarizes the IPD predicted vs. CPD measured values from twelve two-ton test batches. Two sets of comparisons are provided: CPD measured vs. IPD predicated based on techniques used since January 1, 1960 and CPD measured vs. IPD predicted based on techniques used prior to January 1, 1960.

This test is designed to obtain up-to-date information on heat decay values following a reactor scram. More specifically its purpose is to determine the effect of increased outlet water temperatures upon this heat output value. A recent review of the ``last ditch`` cooling system at the Hanford reactors indicated that at some reactors this backup coolant supply did not meet our coolant adequacy criteria at certain critical times after the loss of electric and steam pumping power. On the basis of this review, a high priority program is underway to bring the adequacy of reactor backup cooling systems into conformance with the criteria. An important part of this program lies in improving the accuracy of certain basic elements of reactor cooling conditions following a power loss. Primarily these are: Accurate reactor low flow values; and accurate reactor heat output values. A program is already in progress to improve both the amount of last-ditchcooling flow and to improve the accuracy with which this flow is determined.

During the post-irradiation measuring of fuel elements from PT-IP-262A by, Component Testing Operations, Irradiation Processing Department, one element exhibited 0.069 inch of diametric ellipticity. The element had been fabricated from natural uranium, clad in X-8001 aluminum alloy, and was internally and externally cooled during irradiation. The element was selected for detailed examination to determine the cause of the distortion and was transferred to the Radiometallurgy Laboratory in January 1961. Growth in the radial direction over the total length in one quadrant of the fuel was the cause of the distortion. Maximum ellipticity of the uranium was 0.037 inch. Corrosion of the cladding corresponding to the minor axis of ellipticity made the distortion appear more severe. The uranium had away from the spire braze layer over a three-inch length in the central portion of the element and at the point of maximum ellipticity approximately 110{degree}of the angular distance around the spire contained a gap. Grains in the uranium were larger in the area adjacent to the spire gap than in the remainder of the piece. Two possible causes are: (1) A pocket of large grained, oriented uranium existed in the element when it was canned. (2) Oriented uranium caused growth which …

The estimated achieved reliability of SNAP 10A space nuclear power units will be relatively low at the timeof the first SNAPSHOT flight test in April 1963 and the existing R&D program does not provide a significant reliabiity growth thereafter. The total costs of an 8-satellite network using SNAP 10A units over a 5-year period has been approximated for the case where the total cost of a single satellite launched is 8 million dollars.

Developments at the Bettis Laboratory in the area of reactor technology are reported. Seven papers are included for which separate abstracts are prepared. (J.R.D.)

ABS>A test was made on the Radioactive Waste Disposal (RWD) system to determine the acceptability of procedures used in containing, processing, and disposing the wastes received from the plant during a steady-state operation. The RWD system was found to be adequate and to have 60,000 gallons available at all times in the Surge and Decay tanks for safety injection. (D.L.C.)

A bibliography comprising 208 unclassified references is presented on nuclear direct energy conversion devices. Major emphasis is placed on auxiliary power devices suitable for use in satellites including reports on nuclear batteries, thermoelectric cells, thermionic conversron and aspects of the SNAP program. (J.R.D.)

None

None