The positional tolerance method of dimensioning is described. The emphasis is placed on the elimination of ambiguities and increase in tolerances provided by this method as compared to the older, bilateral method. (auth)

This document provides the required definitive scope design for dissolver equipment capable of nuclear safe processing by geometry of E-Metal and certain other fuels enriched to a maximum of one per cent U-235 equivalent. Using only a single dissolver installation of this design, it will be possible to process the current E-Metal monthly commitment in less than two weeks (five-day week). The proposed dissolver system is to incorporate design flexibility required to permit conversion to Zirflex processing of NPR fuels.

This report discusses research and engineering work on the purex and redox processes at the Hanford Engineering Works in May, 1959. (JL)

This is the monthly report for the Hanford Laboratories Operation, April, 1959. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology financial activities. Biology operation, physics and instrumentation research, employee relations, operations research and synthesis operation programming, radiation protection, and laboratory auxiliaries operation are discussed.

The objectives of the test described in this report are to evaluate the integrity of I & E MINT fuel elements canned by the ``C`` process to determine the conversion ratio of I & E geometry E-N ``striped loadings,`` and to provide a demonstration loading for a full reactor loading incorporating a ``striped`` center loading and a fringe ``blanket`` loading. This test will involve two portions a 26 tube block loading of I & E enriched uranium and MINT fuel elements arranged in a striped array, and a 10 tube fringe blanket loading of MINT fuel elements supported by the required enriched uranium fuel columns, and accompanied by controls. Either portion may be charged independently.

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Last year (1948) during the flood period the flow at Grand Coulee fluctuated widely. 2 PM, June 8, 543000 c.f.s.; 4 AM, June 9, 568000 c.f s.; 2 PM, June 9, 543000 c.f.s.; 2 AM, June 10, 573000 c.f.s. A total instantaneous fluctuations of 37,500 c.f.s. was reported. Now there is installed a new control. This control can keep downstream variation within 500 c.f.s. By lowering the lake level prior to the crest period, the drum gates could be used as flood control (1948 high water basis) the drum gate control plus the water turbine discharge (if the lake level had been reduced) could have dropped the crest at Richland three feet. a. Drop in crest at Richland one foot: Electrical loss nominal, b. Drop in crest at Richland two feet: Electrical loss 1 megawatt/foot for six generators. Loss Max possible 13,310 KW each generator, 79,860 KW total (7 days). Capacity 1,170,000 KW Max Loss 6.8% for 7 days to 10 days. c. Drop in crest at Richland three feet: Electrical loss 1 megawatt/foot for 6 generators Max possible 30,100 KW each generator 180,600 KW total 8 days. Capacity 1,170,000 KW Maximum loss 15.4% for 8 to 12 days. Actual …

Sections of the dummies furnished by Radiological Engineering, Process Reactor Development Operation were machined into 1/4 inch diameter by 1 inch long cylinders and irradiated in the Quickie Facility at F area. The pieces were discharged directly into a holder one foot from the Beckman chamber. The transient time from in-pile to the chamber is approximately 30 seconds. The readings were taken using a Beckman chamber, Beckman Micro-Micro Ammeter and Recorder. This system has been calibrated with Co{sup 60} sources obtained from the Oak Ridge National Laboratory. We are including data taken from a sample of 61-S and 99.998 per cent aluminum which may be of interest.

This document details activities of the Fuels Preparation Department during the month of April 1959. (FI)

Areas covered in this report are as follows: nuclear safety in the Fuel Preparation Dept.; studies related to the present production reactors; studies related to future production reactors; studies related to separation plants; studies form the neutron cross section program; reactor development (gas cooled reactor program and test reactor operations); and biology and medicine.

In early July 1958, it appeared that Columbia River temperatures at HAPO would be near 24--50{degree}C by the end of August. River temperatures were averaging 40 to 50{degree}C above 1957 figures and were 3{degree} to 4{degree} above the ten year highs. It seemed desirable to examine the problem to determine if any corrective measure could be taken, since it was apparent that production losses were imminent. The large storage of cold water behind Grand Coulee Dam, normally untapped, was a source of possible relief. A plan for use was proposed for the peak high temperature period and agreed to by the Bureau of Reclamation.

Considerable production gains might be attained if each reactor discharged its entire flattened region during one scheduled outage instead of utilizing several outages for this purpose. Several of the older reactors are now discharging a high percentage of their flattened zones in a single outage and could be put into this type of operation with relatively little difficulty. Production gains may be possible through better flattening efficiency, a more favorable rupture rate effect, fewer non-equilibrium losses, higher conversion ratio, and more efficient usage of outage work. Since this document is written Primarily from the Operational Physics standpoint, some gains and pitfalls which must be evaluated by other affected groups will only be mentioned here as possibilities. The purpose of this document is simply to point out the potential gains in flattening efficiency from this method. Potential gains from improved fuel performance have been described in another document.

Production of Pu from separations plants was only 58% of April commitment because of Purex difficulties. UO{sub 3} production, shipments met schedules. Pu shape production and shipments exceeded forecast by 14%. Purex HS column, repaired Oct 1958, developed another leak and was bypassed April 18, resulting in Pu and U that required reprocessing. A Palm recovery run at Purex with all- reducing flowsheet, resulted in 87% recovery and excellent decontamination of product. The prototype dual-pass silver reactor in Purex C-cell plugged with offgases. Processing of unclarified feed through Purex solvent extraction continued. Redox dissolver batch sizes for E-metal processing were increased from 1.75 to 2.0 tons. Testing of first extraction cycle acidic flowsheet at Redox continued, with Np losses to HAW being below detection limit. Ru in 1AFS stream increased 10-fold F.P. activity but was removed in acid deficient U cycles. A sulfamic acid process is being explored for dissolving Pu metal. Scope design of Redox dissolver and RMA line replacement at Z plant was completed. Shielding effectiveness of medium and high density x-ray lead glass was compared.

This report addresses the chemical technology divisions' monthly report for chemical development.

The demand plot has a 5-set, modified pump decay curve; it shows that 20,000 gpm emergency flow would be required within 80 seconds of complete pump power failure. Bases for the demand curve are constant bulk inlet temperature of 2 C, constant bulk outlet temperature of 95 C, K-3 I&E fuel elements, and initial reactor flow of 188,000 gpm.

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The temperatures of the Columbia River was reduced 1 to 30 Centigrade with beneficial effects at HAPO. It is reasonable to expect that future benefits may be possible. It is desirable that the temperature of the river be controlled each year to the maximum extent possible. Instrumentation improvements requested to effect optimum savings. Records of river temperatures and flows should continue to be maintained by IPD as a necessary part of temperature optimization. Where possible, the coincident use of the river cooling technique should be made for the benefit of anadromous fish.

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The proposed operating conditions for fuel elements to be charged into the NPR require the fuel to be of an extended surface geometry and maintain adequate strength and corrosion resistance in 300 C water. A contract was let to Nuclear Metals Inc. to produce by co-extrusion lengths of fuel rod containing both natural and 1.6% enriched uranium of irradiation quality for fabrication into fuel elements. The fuel rods used in the irradiation test represent the first enriched uranium rods coextruded in 0.030 inches of Zircaloy-2 to be irradiated and examined at Hanford. The rods used for this test were fabricated into four, 4 rod cluster fuel elements thus allowing adequate space between individual rods for expansion in the case of a fuel rod failure. This rod was of particular interest since it contained an irregular uranium-Zircaloy-2 interface. The purpose of the irradiation was to determine the dimensional stability of coextruded fuel rods and to determine whether the irregularity in the bond interface had any effect upon the irradiation performance of the fuel. Fuel elements were irradiated in 200 C water in the KER Loop 2 facility to an exposure of 0.28 a/o burnup (2,200 MWD/T). Post irradiation examination showed that …

The Purpose of the work described in this report has been to determine experimentally the rate of production of tritiun in fringe lithium-aluminum alloy loadings with the degree of precision necessary for economic analyses of such a method of isotope production. These results are provided for use in such an analysis.

The technique of high-frequency titrimetry was applied to the determination of thorium, uranium, sulfate, and free acid. In Part I, the reproducibility of the method for the titration of standard solutions that contained 50 rag of thorium in the absence of interferences is established. Under these conditions, the coefficient of variation of the method was <1%. In Part II, the effect of uranium on the highfrequency titration of thorium, as well as the application of the method to actual samples, is discussed. Uranium in a ratio of 5 to 1 to thorium can be tolerated. When the method is applied to the analysis of representative samples, the coefficient of variation is 1%. Attempts to determine uranium by high-frequency titration with 8-quinolinol were unsuccessful. Tests on this titration and possible reasons for its failure are discussed in Part III. The application of the high-frequency titration method to the determination of sulfate in solutions of uranyl sulfate is described in Part IV. The coefficient of variation, on titrating 40 to 70 mg of sulfate with barium chloride after the uranium is masked with citrate or fluoride, is 2%. In Part V, the high-frequency titration of free acid in solutions of uranyl sulfate …

After two batches ( approximately 340 kg) of fluoride salt from the ARE were reprocessed, a pilot plant operations were terminated because of a leak through which an estimated 780 g of uranium (as UF/sub 6/ escaped. Of the 21 kg of highly enriched uranium in the feed, 93.12% was collected as UF/sub 6/ product, 0.13% represented measured losses, and 3.72% was unaccounted for (leak). An additional 3.03% was reclaimed from NaF beds and equipment washes. The produce met both chemical purity and activity specifications for product level UF/ sub 6/. Decontamination from fission products was essentially complete. A gross gamma decontamination factor was apparently limited by the low activity of the feed salt. (auth)

The basic theory of operation of the triggered spark gap is established, and qualitative and quantitative engineering design data are given. From the basic twoelectrode gap, a three-electrode or triggered gap model is established with its static and dynamic triggering characteristics shown. Several geometry conditions such as gap spacings trigger electrode hole sizes and insulator effects are discussed, showing their influence upon the triggering mechanism. A suggested trigger mechanism is given based on that proposed by Sletten and Lewis for the trigatron and modified to fit the present analysis. (auth)