EWR No. VTE-188--Tubes processed by flashing getters immediately prior to seal-off from vacuum systems are compared for total residual gas pressure to tubes processed by flashing getters after tubes were sealed off vacuum systems. Comparisons of residual pressures determined from current flows in the cold cathode ion gauge.

A micrometer that utilizes eddy current techniques is described. The gage is capable of measuring nominal 0.5000-in. aluminum rods to an accuracy of plus or minus 0.00005 in., and is unaffected by residual nonconductive surface filins such as oxides or corrosion products. (auth)

An estimate was made of the fission product activity which would result in the HFIR primary coolant system following a meltdown of the fuel element within the reactor. The rare gases and the halogens appear to be the main contributors to the gamma activity in the coolant system imnmediately after the meltdown, and iodine appears to be the main contributor 24 hours after the meltdown. (auth)

The 9 single-crystal elastic moduli pertaining to principal crystallographic axes of alpha U at 25 deg C were determined from measurements of high-frequency wave velocities for 21 modes in seven single-crystal specimens, using the phase-comparison method of McSkimin. From the results the elastic compliances, compressibilities, and Poisson"s ratios were computed for the principal axes. Th variations with crystal direction of the stiffness moduli, Young's moduli, and rigidity moduli were plotted. The nature of the anisotropy for the different moduli indicated that the nearest neighbor interatomic bonds are considerably stiffer than the next nearest bonds, which are only slightly larger in interatomic distance. (auth)

As a result of atomic shelter tests and field experiments condueted over the past nine years, it has been conclusively shown that shelters provide the only promising means of civilian protection in the event of a nuclear war. Design details are presented for a group shelter to accommodate 100 persons of all age groups and both sexes. The shelter structure is a multiplate corrugated- steel arch set on a concrete slab with end walls of bridge plate sheathing. The entire structure is covered with a minimum of 3 feet of earth. The shelter combines outstanding protection against radioactive fall-out with good protection against blast and thermal radiation. Drawings are included. General operating procedures are outlined. (C.H.)

A process for separating americium and curium from rare earths by anion exchange based on selective chloride complexing was developed and tested on a laboratory scale. The separation is accomplished by sorption of americium, curium, and rare earths on Dowex 1-10X resin from a solution of 8 M LiNO/dub 3/ followed by selective elution of rare earths with 10 M LiCl and americium-curium elution with 1 M LiCl. In a laboratory demonstration of this process, greater than 99.5% of americium tracer containing no detectable amounts of rare earths was recovered. (auth)

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.

A simple laboratory apparatus for the study of fissiongas release from neutron-activated fueled graphite was developed. Xenon-133 released from a heated specimen is carried in a helium sweep gas to a charcoal trap, where the accumulated activity is monitored continuously by a scintillation detector, ratemeter, and pen recorder. The maximum specimen temperature (2500 deg F) is achieved in 10 min with an induction heater. All instrumentation is commercially available. Data for several neutron-activated fueled-graphite specimens heated in the range from 800 to 2500 deg F are presented to illustrate the typical results obtained with the apparatus. (auth)

As part of an earlier program of investigation in this laboratory, studies were made of the gamma ray spectrum and the beta ray spectrum of cerium-144. In the present work, seme coincidence studies were made on one of the beta groups appearing in the cerium-144 decay and on the gamma rays appearing in the deexcitations from the energy levels of praseodymium-144. Sources of cerium-144 were prepared frcm carrier free radioactive cerium-144 as supplied by the Oak Ridge National Laboratory. The sample material was more than two years old at the time of preparation of sources. No additional chemical purification was attempted. Sources for use in the beta crystal spectrometer were mounted on thin Formvar film on spectrometer ring mounts. The gamma ray spectrum of cerium-144 in the energy range 20 kev to 180 kev is shown in Figure 1. This spectrum was determined using a 2-inch by 2-inch NaI(Tl) crystal. The pulse spectrum was analyzed by a Radiation Instrument Development Laboratory (RIDL) 200 channel analyzer. The spectrum gives clear evidence of gamma ray peaks at 34 {+-} 3 kev and 134 {+-} 2 kev. A rather broad peak at 80 kev is observed. An indication of a gamma ray group …

Studies were made to ascertain what useful satellite and/ or space missions may be accomplished with the SNAP 2, 8, and 10 power systems within the 1960 to 1970 time period, to delineate useful satellite and/or space missions which would most profitably employ nuclear power sources, to ascertain what restrictions are imposed on various payloads as a result of having a nuclear power system on board, and to investigate the general vehicle integration and installation requirements of the SNAP 2 and 10 power units. Space flight programs, characteristics and limitations of SNAP, applications of nuclear power supplies. design integration, and system integration are discussed. (M.C.G.)

An increase in the flow of water from the lower depths of the Grand Coulee Reservoir was used to lower the river temperature at Hanford Atomic Products Operation (HAPO). A net average daily reduction of over 1 C resulted. The average for one day varied from 1.7 to 0.2 C. Before the dam control period, the Bonneville Power Administration transferred load from Grand Coulee Dam to other dams in order to conserve cold water. The authors calculate that there was more value to HAPO from this transfer than the combined effects of river bed losses and/or any variance in scheduled or base loads.

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A description is given of bubble chambers in use and those in the design stages. Safety factors in the design and operation of a bubble chamber are discussed. Data are presented on fatige and rupture tests on glass. Data are contained on the effects of liquid helium on the tensile properties of various stainless steels. (C.J.G.)

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.

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Production of Pu nitrate from separations plants during May was below forecast. A Np recovery campaign in Purex yielded 1.5 kg. Production and shipments of UO{sub 3} met schedules. Unfabricated Pu metal production was below forecast, but all shipments were on schedule. Decontamination efficiency was low in Purex solvent extraction around the time of the Np recovery. The damaged Redox B-2 dissolver is being restored; processing of enriched metal in A and C dissolvers was continued. A spectrograph for inclusions in Pu metal was installed. 4 kg Pu oxide was produced in a continuous direct calciner. Scope design on Purex Np recovery and purification facilities was completed. Other design and contracts are discussed.

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.

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A design concept is presented for an electrical system using two SNAP III type generators fueled with cerium. In the modified SNAP III generator, a capsule of Haynes 25 contains 9725 curies of cerium oxide pellets, which will provide 67 thermal watts at time of launch. Sufficient void volume and capsule strength ensure containment of the oxygen evolved through isotops decay during the operational life of the generator. Thermal converter configuration in the conceptual generator is identical to that of the SNAP III except that the shell is stainless steel. Two methods of biological shielding are considered. The first uses mercury contained in a sphere surrounding the generator. In the second concept, a lead cask shields the unit until its installation in the launch vehicle. A remote installation procedure and an equipment arrangement are proposed. Generator output predictions were based on actual test data. The output of a single unit would be 3.8 watts at launch, decreasing to 1.9 watts in the course of a 6-month mission. A groundhandling procedure and conceptual designs of the equipment are included. (auth)

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 control rod drive mechanisms and associated instrumentation in the Shippingport PWR were in satisfactory operating condition. Deficiencies that were observed during the test and subsequently corrected were: no positive indication of movement in rods 11, 82, 62, 53, and 14, as shown on the rod position indicating lights; and the bottom indicator coil for rod 81 was connected improperly. After the deficiencies were corrected, the test was rerun for the rods in question and all operated satisfactorily. (auth)

Current results indicate probable over-all rates of about 0.2 mils/month for titanium vs. 1.5 to 3.0 mils/month for Ni-o-nel. Tests are not 100% comparable due to changes made in flowsheet conditions, but they have been of sufficient variation and length as to allow good predictions to be made. Both metals show some tendency toward local attack in Thorex solutions. These tendencies are increased by poor welding techniques. There is a possibility that a Ni-onel dissolver could be used for interim Zirflex processing although the rates obtained in scouting tests are relatively high (6 to 8 mils/month). Titanium shows no promise of usefulness in either Zirflex or Sulfex on the basis of scouting tests. It can be used for dissolution of U-Al fuels in 8M HNO/sub 3/- Hg(N0/sub 3/)/sub 2/ (rate < 1 mil/month). If the choice between these two processes were made on the basis of present corrosion results alone, Darex-Thorex would be chosen. (auth)

Corrosion screening tests were carried out on potential materials of construction for use in a fused salt-liquid metal extraction process plant. The corrodents of interest were NaCl--KCl-- MgCl/sub 2/ eutectic, LiCl--KCl eutectic, Bi-- U fuel, and BiCl/sub 3/, either separately or in various combinations. Screening tests to determine the resistance of a wide range of commercial alloys to the corrodents were performed in static and tilting-furnace capsules. Some ceramic materials were tested in static capsules. Largerscale tests of metallic materials were conducted in thermal convection loops and in a forced circulation loop. Some of the tests were conducted isothermally at 500 deg C, and others were performed under 40 to 50 deg C temperature differences at roughly the same teinperature level. On the basis of metallographic examination of exposed test tabs and chemical analyses of corrodents, it was found that the binary and ternary eutectics by themselves produced little attack on any of the materials tested. A wide variety of materials including 1020 mild steel, 2 1/4 Cr--1 Mo alloy steel, types 304 (ELC), 310, 316, 347, 430, and 446 stainless steel, 16-1 Croloy, Inconel, Hastelloy C, Inor-8, Mo, and Ta is, therefore, available for further study. Corrosion by the …