The effects of the maximum credible accident relative to the design of the containment shell are discussed. The maximum credible accident is defined. The thermal and hydraulic effects of the maximum credible accident on the reactor system were analyzed. The extent to which fuelrod cladding will melt was estimated. The amount of energy released from the reactor system by the escaping steam and water and by a possible chemical reaction was calculated along with the corresponding pressure rise inside the containment shell. The kinds, amounts, and total radioactivity of fission products released to the atmosphere of the containment shell after the core melts were predicted. (M.C.G.)

This report, from the Chemical Processing Department at HAPO for October 1960, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

This report, from the Chemical Processing Department at HAPO for November 1960, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

This document discusses the test program of evaluating the temperature effect of crud film build-up on fuel element heat generating surfaces in a carbon steel system. This program has three phases: Measurement of the effect of film build-up during normal high pH equilibrium operation; measurement of the temperature effect of film build-up subsequent to a loop decontamination; and measurement of the effect of film build-up in the event of loss of pH control.

In response to specific requests of the AEC and as part of a Combined Operations over-all review, a number of engineering studies have been made of alternative methods for increasing availabiltiy of neptunium at Hanford. The report updates an earlier study in which recycling of both natural and enriched uranium was considered for Hanford. The earlier study showed that recycled natural uranium would provide appreciable gains in neptunium availability but at an excessive cost. Recycle of the slightly enriched uranium streams proved a more economical means of realizing smaller but still significant gains in neptunium production. Subsequent to the earlier report, a feasible and immediately applicable scheme for UFI blending has been conceived demonstrating further advantages for recycling the enriched uranium. Approximately 34 kilograins of supplementary neptunium could be produced at Hanford during the next seven to eight years by upgrading irradiated E-metal and NPR uranium through a blending operation at Oak Ridge (rather than in the diffusion cascades) and then recycling the material through the Hanford reactors. (cf Table 1) Such a scheme would conserve uranium-236 for use as a source of neptunium-237 in the reactors without incurring major capital costs. Oak Ridge Operations personnel have estimated that capital …

This document is the third of a series of question and answer lists issued as a review of the material discussed in the Reactor Processing Fundamentals Course. Each document represents the material covered by the Reactor Specialists during a typical three-month training program. Each question is discussed individually and the entire list completed during the three-month session. Each three-month-training period is devoted to the complete discussion of a single chapter of the IPD Physics Primer series. The questions are compiled in a logical sequence with the material as presented in Chapter V of the Primer series control. As the course progresses, a certain amount of recall of earlier chapters is essential to a thorough knowledge of the specifics of reactivity and distribution control. Therefore the first half of this document consists of material previously presented in a slightly different manner. Basic points are stressed; the intent is, of course, that informative discussions lead to a better understanding of the material presented in the Primer.

The purpose of this memorandum is to transmit revisions to the goal exposure plans for C-II-N, C-II-E, K-III-N, K-III-E, O-II-N and O-II-E material. Inherent in these plans is an assumption that the only restraint on metal usage is that imposed by the economics of the plutonium production process and associated uranium cycle. In the case when metal throughput is limited, by ex-reactor physical capabilities, to a level lower than the unrestrained reactor requirements, exposures higher than those recommended in this document would be indicated.

Data have recently been obtained on a number of C-Reactor process tubes. These data make is possible to reassess the current conditions of the tubes in the C-Reactor and hence to re-evaluate the probable data when retubing of the reactor will have to be started. The results of this reassessment are presented below.

The objective of this document is to authorize large-scale irradiations of alloyed, low hydrogen, dingot uranium fuel elements in order to evaluate their suitability as a reactor fuel. Although purer than standard ingot uranium, the dingot uranium produced in the past by Mallinckrodt Chemical Plant has demonstrated inferior dimensional stability during irradiation. This instability has been manifest as warp and as swelling at the center of the fuel elements. To reduce this instability 150 ppM iron and 100 ppM silicon haze been added to refine the grain structure of the uranium. Destructive examination of grain structure and nondestructive examination of apparent core quality have indicated that this material is equivalent or superior to ingot material. In light of forecasted metal requirements, the Mallinckrodt Chemical Plant must be brought on stream as rapidly as possible. This PITA schedule will meet the programmed metal delivery requirements with a minimum of risk potential while accelerating the evaluation of dingot uranium as to its suitability as a reactor fuel. Completed irradiation performance data from small scale testing will not be available before dingot material is charged under this PITA. Therefore, the possibility exists that unforeseen dingot uranium properties will result in poor performance. Low …

The objective of this production test is to compare the dimensional and structural behavior of fuel elements with various heat treatments during irradiation at conditions similar to those anticipated for the NPR. Zircaloy-2 jacketed unalloyed natural uranium fuel elements of the same diametral dimensions an those anticipated for the N Reactor inner fuel tube, incorporating various heat treatments, will be irradiated to about 2000 MWD/T in the KER Loops. The elements will be enclosed in a Zircaloy-2 sleeve to provide a more balanced flow distribution through and around the pieces. An optimum heat treatment for N Reactor fuel elements providing minimum fuel distortion through control of uranium grain size and orientation has not been determined. This production test provides for the irradiation of fuel elements nearly equivalent to N Reactor inner tubes in size, geometry, and temperature conditions during irradiation. These elements have been subjected to varying heat treatments providing a range of grain sizes and degrees of orientation. Post-irradiation examination of the elements will show the fuel element distortion associated with each heat treatment and provide information on which to base further irradiation tenting in the development of a heat treating procedure for N Reactor fuel elements.

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.

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This report, for August 1960 from the Chemical Processing Department at HAPO, discusses the following: Production operation: Purex and Redox operation; finished products operation; maintenance; Financial operation; facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

The results of examinations of UC compounds having nominal compositions of uranium-4.6, -4.8, and 5.0 wt. % carbon after irradiation to burnups of from 400 to 15,000 Mwd/t of uranium are given. Density changes were small. varying from a minimum of 0.3% to a maximum of about 2.5%. Cracking occurred in all specimens; however. it can probably be largely attributed to thermal stresses and to oxidation. after decapsulation. of NaK entrapped in microcracks. Depletion of carbon appears to be occurring in the specimens having the nominal uranium-5 wt.% carbon composition. Metallographic examination shows that the UC/sub 2/ phase disappears at high temperature and high burnup. The fission-gas-retention properties of the compounds appear quite good. In all cases. the amount of fission gas released was comparable with the calculated amount released by recoil. (auth)

The effects of strain rate on the stress-strain characteristics of copper and lead were studied by measuring both stress and strain as functions of time using short cylindrical specimens supported at one end on a modified Hopkinson pressure bar and impacted at the other end by a steel projectile. Corresponding stresses and strains were computed according to an elementary nonstrain-rate theory (sometimes referred to as the von Karman theory) in which the dynamic stress-strain curve is assumed to be the same as the static stress- strain curve. Stresses and strains were also computed according to an elementary strain-rate theory (sometimes referred to as the Malvern theory) in which the dynamic stress may exceed the static stress for a given strain by an amount which depends upon the strain rate. It was found that the predictions of the nonstrain- rate theory agreed with measured values only for low impact velocities and for points at least two diameters from the impact end of the specimen. By proper choice of the flow or relaxation constant in the elementary strain-rate theory, measured and computed values of strain, or of stress, but not both simultaneously, could be brought into agreement. In the more general exponentialtype, …

Studies were made to determine revised moderator coolant flow requirements in the EGCR core for the latest design. The temperature distribution in the graphite columns was also determined. The total moderator coolant flow was calculated to be 24,024 lb/hr and 31,020 lb/hr for full power and maximum anticipated power, respectively. It was concluded that the maximum moderator temperature (1220 deg F average over the cross section), at full power operation, occurs near the peak heat flux at a location 6.25 to 7.25 ft from the bottom of the active core. The temperature in most of the graphite columns varies from 555 to 1150 deg F over the lower half of the column and 1150 to 990 deg F over the top half of the column. The maximum surface temperature is less than 1110 deg F. Temperature distributions are shown on graphs. (M.C.G.)

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This report for June 1960, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance: Financial operations; facilities engineering; research; and employee relations.

Production Pu nitrate and unfabricated Pu metal during Feb. was below forecast; however FY output is above forecast. Production of UO{sub 3} exceeded commitments; shipments met schedule. Decontamination performance of Purex solvent extraction system was subnormal. Pu nitrate solutions were concentrated. A fire occurred in Purex N Cell during conversion of ion exchange prototype to production facility.

This document details activities of the Irradiation Processing Department during the month of March, 1960. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

Production of Pu nitrate from separations plants exceeded commitments. Two Np recovery runs were made, a routine one at Purex and a demonstration run at Redox. Total Np recovered in CY 1959 was 5.7 kg. The UO{sub 3} plant production and shipments also met commitments. Production of unfabricated Pu metal continued. Process studies are being conducted in Purex to resolve the problems in the Np recovery effort. A new solvent (hexone) condensate decanter was operated in Redox. Reprocessing of nonproduction reactor fuels is being reviewed. (DLC)

This document details activities of the chemical processing department as compiled by the operation managers during the month of March, 1960.

The turbulent mixing of a cold liquid jet with a stream of vapor bubbles and saturated liquid is analyzed for a single axial liquid jet. The cases where the issuing jet velocity is very much greater than and nearly equal to the free stream velocity are considered. Transport of axial momentum and of scalar quantities in the stream are calculated and the results applied to the transport of enthalpy in a bubbly mixture. (D.L.C.)

An in-pile corrosion test loop is described which is used to study the effect of reactor radiation on the corrosion of materials of construction and the chemical stability of fuel solutions of interest to the Aqueous Homogeneous Reactor Program at ORNL. Aqueous solutions of uranyl sulfate are circulated in the loop by means of a 5-gpm canned-rotor pump, and the pump loop is designed for operation at temperatures to 300 ts C and pressures to 2000 psia while exposed to reactor radiation in beam-hole facilities of the LITR and ORR. Operation of the first loop in-pile was begun in October 1954, and since that time 17 other in-pile loop experiments were completed. Design criteria of the pump loop and its associated auxiliary equipment and instrumentation are described. In-pile operating procedures, safety features, and operating experience are presented. A cost summary of the design, fabrication, and installation of the loop and experimental facillties is also included. (auth)