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.

Three E-N loaded tubes were dissolved, sampled and analyzed, starting November 9, 1960. The results of these analyses and an explanation of the methods used are the subject of this report. Each tube loading received an identification code in each facility in which it was processed. All of these codes are listed for future reference. Each batch of slugs was dissolved in a preflushed dissolver. When complete solution was indicated by a leveling off of the specific gravity, two 1 ml samples were taken and analyzed for uranium, specific gravity, and excess nitric acid. The dissolver charge was digested an additional four hours. At the end of the digestion period, two 1 ml samples and one 20 ml pig sample were taken. The three samples were analyzed for U, SpG, and HNO{sub 3}. Agreement between these samples and the previous samples was taken as confirmation of complete dissolution and representative sampling. If agreement was not obtained, sampling was continued. After verification of the pig sample, six 1 ml aliquots were taken for analysis by the Analytical Control group. The remainder was aliquoted to provide material for mass analysis and for analysis by the Process Chemistry group.

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.)

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Experiments showed that 30% tributyl phosphate will not extract acid- deficient species of uranyl nitrate. Flooding throughputs for the Mark I stacked- clone contactor ranged from 600 cc/min organic at zero aqueous to 60 cc/min organic at 1950 cc aqueous. A large electronic vibrator of 5000-lb thrust was found somewhat inferior to pneumatic vibration for compacting oxide fuels into stainless-steel tubes. Tests were started on the use of fixed-bed CuO oxidizers for removing hydrogen contamination from helium gas streams. None of the variables studied within this period effected an increase in particle size in the denitration of thorium nitrate to produce ThO/sub 2/. The rate of uranyl sulfate loading on nitrate equilibrated Dowex 21K appears to be essentially independent of the loading solution sulfate concentration. The nitric acid concentrations corresponding to maximum UO/sub 2/-ThO/sub 2/ pellet dissolution rate were 15.5 M for Thorex solution and 13 M for the adjusted Darex solution. Two additional Semicontinuous Sulfex declad and Thorex core dissolutions of prototype Consolidated Edison fuel assemblies were made to complete the series of runs. The effective area of cylindrical UO/sub 2/ pellets dissolving in nitric acid was estimated from experimental rate measurements as a function of the fraction dissolved. …

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.

This report details the proposed E-N, tritium-plutonium producing reactor loading is made up of striped columns of lithium-aluminum (mint) target slugs and enriched uranium (.947 w/o U{sup 235}) slugs. Both target and uranium slugs are of the I & E geometry. The ratio of mint (N) to uranium (E) is determined by the requirement of sufficient k-excess to sustain an operable reactor. The designer of the lattice loading has a choice between natural lithium (7-5 w/o Li{sup 6}) or lithium enriched to {approximately} 38.5 w/o Li{sup 6} for the mint slugs, assuming enriched or ``black`` mint is available. It is possible to show at least 0.8% increase in total conversion ratio for an E-N load for enriched mint (38.5 w/o Li{sup 6}) over natural mint. The basis of the calculations rests on measured E-N length ratios for equivalent neutron multiplicative properties for both types of mint (in a dry pile) plus analytical calculations. It is shown that both increased Pu and H{sup 3} production are obtained by using blacker mint. The basic phenomena are (1) increased resonance capture in U{sup 238} due to more uranium volume in the black mint E-N lattice and (2) more efficient utilization of neutrons in …

Solid-solubility limits at 1500, 1000, snd 500/sup o/C for carbon, nitrogen, and oxygen in high-punity tantalum were determined by x-ray lattice- parameter methods. For carbon, the solubility was found to be 0.17 at. % at 1500/sup o/C and less than 0.07 at. % at l00/sup o/C. A nitrogen solubility of 3.70 at. % at l500/sup o/C decreased linearly with temperature to 2.75 at. % at 1000/sup o/C and 1.8 at. % at 500/sup o/C. In the case of oxygen, the solubility was found to be 3.65 at. % at 1500/sup o/C, 1.95 at. % at l0O0/sup o/ C, and 2.5 at. % at 500/sup o/C. The phases Ta/sub 2/, the lowtemperature modificstion of Ta/sub 2/O/sub 5/, and Ta/sub x/N of unknown coznposition hut which has a superlattice structure based upon the oniginsl body-centered-cubic tantalum lattice were identified is the initisl precipitates in the respective systems. (auth)

It is the purpose of this document to provide assistance to the Manufacturing Section personnel in determining their future operating plans. In general, the inter-relationship of such engineering parameters as flow, reactor orificing, rupture performance, etc. has been considered. The effect of these engineering parameters are summed up in our recommendations for {open_quotes}Operating Plans{close_quotes} shown graphically in this document. It is to be emphasized that these plans do not reflect operational considerations which may modify the desirability of the indicated level increases nor has allowance been made for ability of the indicated level increases nor has allowance been made for major projects, major maintenance outages, or major changes in pile loadings. Many factor, which only Manufacturing personnel are capable of evaluating, may make it desirable to operate below or above these {open_quotes}Operating Plans.{close_quotes} These {open_quotes}plans{close_quotes} are based on incremental metal cost and burnout cost estimates obtained recently from L. W. Lang. A change in these assumed costs would require a revision to be made to these {open_quotes}plans.{close_quotes} It is also to be noted that many of the engineering parameters and basic assumptions which have been factored into these {open_quotes}plans{close_quotes} are subject to continual re-evaluation and revision. Thus, in a strict …

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The Purex Plant was originally designed to operate at a instantaneous rate of 8-1/3 tons of uranium per 24-hour day on a three-cycle flowsheet. Subsequently, the rate was increased to 26-2/3 T/day and the process changed to a two-cycle flowsheet. Recently, flowsheet changes made for neptunium recovery and replacements of certain critical equipment items have altered the plant`s capacity. The status of the plant is discussed in the Continuity of Operations, Plant Improvement Program, Reference 1; wherein, the need for expansion to a capacity factor of 4.0 by January 1, 1962, was envisioned. Reference 2 was prepared to implement that program. A new chemical flowsheet, Reference 3, was prepared to serve as a basis for the Purex Expansion Program. The purpose of this engineering study was to determine the changes and capital expenditure needed to obtain a 33-1/3 T/day instantaneous production rate through the Purex Plant. This report discusses equipment additions and modifications that are required for the Purex Plant to operate at an instantaneous rate of 33-1/3 tons of uranium per day on the flowsheet, as described in Reference 3.

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The compatibility of carbon and graphite matrices with UO/sub 2/ particles coated with Al/sub 2/O/sub 3/, BeO, nickel, niobium, and nickel- chromium alloy was investigated at several temperatures up to 3OOO deg F in flowing helium. Two different carbonaceous fillers and binders were used. As expected, the 2 to 8- mu metal coatings were badly damaged by reaction with carbon at temperatures as low as 17OO deg F. Both oxide coatings were completely destroyed after 6 hr at 3000 deg F. Considerable reaction between the matrices and.Ai/sub 2/O/sub 3/ and BeO coatings occurred during 1000 hr at 2500 deg F. Coating damage was more severe in surface particles than in particles located inside the specimens. The graphite filler and pitch binder combination used in this study was less reactive than combinations containing coke filler or resin binder. (auth)

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The scintillation counter has proven to be a very valuable research tool, but urfortunately, its ability to meet necessary stability requirements has restricted its use in industrial applications. Several techniques are being investigated which cancel out reasonable variations in detector sensitivity, resulting in improved stability. The general technique consists of alternately measuring the intensity transmitted through the sample and through a calibrated absorber, and difference in intensity causing the calibrated wedge to re-position itself. A comparison of commutating and noncommutating systems is made and other applications of scintillation counter systems are discussed. (For preceding period see ARF-1152-6.) (W.D.M.)

Since startup in January 1945 through December 1959, 4 {times} 10{sup 9} gallons of radioactive liquid wastes have been discharged to cribs and trenches at HAPO by the Chemical Processing Department facilities in the 200 Areas. These wastes contained approximately 2.5 {times} 10{sup 6} curies of beta emitters. The Scavenged Waste Recovery Program was completed in 1957 and Redox Plant process changes were made during the latter part of 1958. These changes resulted in significant reduction in the amount of radioactive materials that have been discharged to the ground in subsequent years, from a maximum of 8.2 {times} 10{sup 5} curies in 1955 to 9 {times} 10{sup 3} curies in 1959. Although these large amounts of radioactive materials have been discharged to the ground, periodic waste report inventories include no reduction due to radioactive decay. The estimated depletion by radioactive decay is the basis of this report.

Boron-impregnated polyethylene tape was irradiated in the Engineering Test Reactor Critical Facility to study the effect of boron as a burnable poison in reactor fuel. Isotopic analysis of the boron was performed with a conventional CEC Model 21-103 mass spectrometer. The tape was distilled off and the residual boron was converted to trimethyl borate. The reaction mixture was analyzed without separation. Good precision was obtained with samples containing less than 0.5 mg. boron. Features of the mass spectrum of trimethyl borate are discussed. Other applications of the method are suggested. (auth)

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.

From Tamalpais debris crushed to <53 mu , 5.5, 19.6, and 12.5 f the alpha, beta, and gamma activities, respectively, were leached in 72 hr at room temperature by a 100 to 1 weight excess of ground water from the Nevada test site. The extracted alpha-activity material was> 97% Pu/sup 239/ and < 3% Am/sup 241/, and the gamma was 92% Ru/sup106/, 4.7% ZrNb/sup 95/, and 3.1% Cs/137/. The beta activities could not be identified. The ground water leached 10 to 20 times as high a percentage of activity from Tamalpais debris as from Rainier debris in a previous study, but on a count rate basis the total activity released to the ground water was a factor of 2 greater for Rainier than for Tamalpais. Increasing the leaching temperature from room temperature to boiling doubled the amount of active material extracted. increasing the particle size clsssification from <53 to 5901190 mu decreased the extraction efficiently approximately 3- fold. (auth)

Supplement F to Production Test IP-314-A authorized an upstream thermocouple train in KER-1. This document provides fuel element loadings, operating temperatures, and trip settings for two possible charges in KER-1, one using KSE-3 heater elements and the other using KSN-3 heater elements. For a charge containing KSE-3 elements, the loading and the operating conditions are given. For a charge containing KSN-3 elements, the loading and the operating conditions are given. The desired operating temperature and trip settings needed to maintain the maximum uranium temperature below 660 C are given. One of the thermocouples in the thermocouple element, however, indicates the maximum uranium temperature directly. Based on the thermocouple readings, the outlet temperature and trip settings may be set to exceed or fall short of the values designated by the dashed lines provided. The temperatures do not exceed the limitations set by the solid lines which are based on the requirements that boiling will not occur on the fuel elements during normal operation and burnout will not occur at the limiting trip conditions. The maximum uranium temperature shall not exceed 660 C.

The reference design is an unmoderated, sodium-cooled reactor using a paste fuel of uranium monocarbide in sodium. The core is a cylinder 5 ft in diameter and 5 ft in height. An 18-in. thick breeding blanket surrounds the core, and an 18-in. thick graphite reflector surrounds the blanket. Various changes were made in the reference core to uncover any possible modifications for cost reductions and to evaluate the consequences of certain design modifications which might occur. Cases were studied for variations in: fuel volume fraction in the core from 0.2 to 0.6; fertile material volume fraction in the blanket from 0.2 to 0.6; blanket thickness 3 in. to 24 in.; fuel materials of UC, U metal, UC/ sub 2/, PuC-- UC, Pu-- U metal, and PuO/sub 2/-- UC/sub 2/; and liquid carrier in the paste of Na, Sn, or Pb. (auth)

This report contains a flowsheet which was prepared to serve as a guide for an engineering study to be made to determine the costs and revisions required to increase the capacity of the Purex Plant to a 4.0 capacity factor. The needs for such a study are outlined in documents HW-62952 (Ref. 1) and HW-63927 (Ref. 2), which include increased 100 Area production forecasts and maximizing Palm recovery by continuous recovery equipment as primary factors. The flowsheet, shown graphically on the flowsketch and specifically in tabular form under Table 1, is a ``best estimate`` of process conditions and modifications which will be in operation by the third quarter of FY 1962. The modifications included are: Formaldehyde Treatment of Concentrated Wastes -- IWW (Ref. 5 and 6); A Continuous Palm Recovery Cycle; A Palm Ion Exchange Purification and Loadout Facility (Ref. 3); Alkaline-Permanganate Washing of the No. 2 Organic System Solvent; and Rough-cut Fission Product Recovery and Shipping (Ref. 4). Since the fission product recovery and shipping operations do not effect the equilibrium conditions of the plant, they are not included in the flowsheet.

In 1957 three of the four high temperature recirculating loops in KE Reactor were put into service and in 1958 the fourth loop was started up. The primary purpose of the loops has been high temperature irradiation testing of fuel elements in support of the NPR fuel element development program, although tests have been run to provide information for PRTR fuel development and high temperature aluminum corrosion programs. Coolant technology studies relating to activity build-up, crude formation, corrosion, failure detection, and decontamination are carried on in conjunction with the fuel element tests. The purpose of this document is to provide a resume of all fuel element test irradiations in the KER Loops from the time they started up to the end of CY-1959. A description of each test, its operating conditions, a chronological history of each loop, and a bibliography of documents pertaining to the tests are included. Supplements to this report will be issued periodically to bring the information up to date.