Potential cladding materials for a flat-plate fuel element containing a dispersion of UC or UC/sub 2/ in U--15 wt.% Mo alloy were surveyed on the bases of compatibility with the fissile compounds, matrix material, protective cover materials, and liquid sodium as well as the feasibility of fabricating fuel plates by roll cladding. Radiative-capture cross sections, thermodynamic data, eutectic and intermediate compound formation, mechanical properties, and corrosion by 1000 tained F Na are reported for austenitic stainless steels, chromium, nickel, niobium, molybdenum, tantalum, vanadium, and zirconium. It was recommended that "A" nickel (molybdenum barrier), Zr-3 wt.% Al. Nb--2 wt.% Cr, and Fansteel 82 be relected for investigation. (auth)

This document details activities of the Fuels Preparation Department during the month of March 1960. (FI)

A calculation was made to determine the fuel plate heat fluxes resulting from after shutdown fission product heating. Fission product source strengths were obtained via the IBM Internuc code. Slab geometry was assumed. The results indicated that the maximum heat flux would occur slightly inboard of the center of the fuel annulus, with the heat flux at the inner annulus radius running about 8% below the maximum, and the outer radius heat flux 10% below the maximum. For decay times of 1.0, 10, 10/sup 2/, 10/sup 3/, 10/sup 4/, and 10/sup 5/ seconds the maximum calculated fuel plate heat fluxes were 42.0, 30.0, 18.0, 9.5, 4.0, and 1.3 x 10/sup 3/ Btu/hr-ft/sup 2/, respectively. The core coolant gamma heating rate during reactor operation was also calculated using the same techniques, but including the fission and capture gamma sources. Average coolant gamma heat generation rate was about 33 watts/cc at the start of the fuel cycle, and 57 watts/cc after the fission products built up. (auth)

A program consisting of two related phases is described Ia Phase I, a study was made to find the relationship between the amount and nature of radioactivity, particle size distribution, and weight of particulate matter present in the lower troposphere. Emphasis was placed on the distribu tion of strontium-90 and total beta activity. Results of a limited number of analyses indicate that strontium-90 and total beta activity is associated primarily with particles below approximately 0.1 micron diameter. Phase II consisted of experimental studies on scavenging of solid particulate matter by water droplets. It was found that water vapor gradient around a condensing droplet promotes scavenging of particles of 1.3 micron and 0.3 micron diameters. The effect of water vapor gradient around an evaporating droplet is not well defined. (auth)

Changes in operation and corresponding changes in the reactivity status of Hanford reactors are the result of a continuing effort to improve operating efficiency. Trends data related to these changes in operation and reactivity have been published previously for the periods from 1950 through 1958. The purpose of this report is to include trends data for 1959. Bar graphs in the first part of the report show yearly averages of selected data, and tables in the last part of the report show maximum, average, and minimum values. This document presents trends data for B, D, DR, and F reactors while a second document, HW-64932, presents trends data for C, H, KE, and KW reactors. Data included in past years which have not been included in this report are trends in pile power level at shutdown omitted due to a security status change regarding power levels, and number of temporary poison columns per startup omitted due to virtual elimination of temporary poison startups at B, D, DR, and F Reactors; added were potential non-equilibrium gains and potential equilibrium gains. Notice that all reactivity values are listed in the unit per cent excess k.

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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This report discuss the design of the Plutonium Reclamation Facility the capacity of which will be 300 kilograms per month or 3600 kilograms per year or plutonium. The subject facility, as the name implies, must be extremely flexible in its ability to handle a wide variety of feed materials. The new facility will be operating on a three-shift day, five-day week, 40-week year with an overall efficiency of 75 percent; twelve weeks per year will be required for ``turnaround`` time to enable campaign operation for segregation of feed plutonium by isotopic content.

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The intent of this document is to establish criteria for equipment for: Receiving, handling and component inspection; miscellaneous handling and storage; final inspection; and finished storage in the new fuel cladding facility.

As power levels have been raised at the K Reactors, the graphite temperature limit has caused significant production losses. These losses have come from three sources: (1) Occasions vhen pile power was directly limited because the graphite temperatures were limiting; (2) Losses in ECT because control rod movements to control graphite temperatures are not necessarily the best rod movements for pile, flattening; and (3) Losses due to lack of control in the final stages of an operating period because COp could not be used as a means of controlling long-term reactivity gains. This document attempts to establish the magnitude of these losses and shows the justification for increasing the graphite temperature limit. The data which is presented here is based upon operating data from December, 1959) through March, 1960, at KW. The results apply to KE as well as KW.

Lengths of pistillate inflorescences of sedges (Carex spp.) growing in the contaminated soils of White Oak Lake bed were measured in relation to radiation fields. Carex Frankii Kunth and Carex vulpinoidea Michx. populations were sampled from areas with air dose rates of 0, 10, 20, and 40 mr/hr. Analyses of variance showed that mean lengths of inflorescences of C. vulpinoidea were simllar to each other in these areas but that those of C. Frankii were significantly different. However, the sites differ in such factors as soil moisture, fertility, and alkalinity, so that these differences in the length of inflorescence may not be accounted for primarily by the exposure dose rate of the radiation field. (auth)

Corrosion specimens of the five major aluminum alloys used in the construction of the Oak Ridge Research Reactor have been exposed to the high- purity primary cooling water in the ORR core and in the extennal portion of the primary cooling loop to determine their corrosion rates under actual operating conditions. These alloys, 1100, 3003, 5052, 5154. and 8061, exhibited average corrosion rates of less than 2.6 mpy during the first 500-hr test period and less than 0.5 mpy for a 4032-hr test. Very superficial pitting was observed. and no case of intergranular corrosion was found. (auth)

A description is given of the Advanced Sodium Graphite Reactor Power Plant, including the reactor, heat transfer systems, generatirg plant, control systems, and the economics of producing 256 Mw(e). The safety of this design is due to its unusually low operating pressure, absence of chemically incompatible materials in the core, and excellent stability under atatic and dynamic conditions. The reactor is being constructed at Hallam, Nebraska, at a probable cost of 0 to 0/kw, exclusive of the first core costs. The 151 fuel elements of uranium carbide are enriched to 2.75 at.% U/sup 235/ and clad in stainless steel. The average thermal neutron flux in the fuel is 8 x 10/sup 13/ n/cm/sup 2/sec. (B.O.G.)

A study was made to determine the coolant flow required to prevent the temperature of the graphite moderator and reflector blocks from exceeding 1100 deg F at full-power operating conditions. The temperature distribution in the graphite blocks was also determined. Tha moderator was primarily cooled by coolant flowing in the 1/8-inch annulus formed by the fuel-assembly sleeves and the fuel-channel holes in the moderator blocks. Coolant flow in the controlrod channels and in the experimental-tube coolant annuli also cooled the moderator, and this effect was considered in setting moderator-annulus coolant flow. The coolant flows required for each of four zones in the reactor core were determined. The total moderator-annulus coolant flow (excluding 16 unfueled channels) was 15,200 lb/hr. The moderator-block temperature distributions for fullpower reactor operation after 20 years of operation are given. The maximum temperature (1100 deg F average over the cross section) occurred near the top of the graphite block. The temperature in the majority of the graphite columns varied from 600 to 1000 deg F over the lower half of the column and from 1000 to 1100 deg F over the top half of the column. The maximum graphite surface temperature was less than 1100 deg F. …

This report is an interpretation of data (flow, pressure, temperatures within a process tube during events affecting single tubes) as applied to the most severe (rapid) K-reactor transients which are credible. Analyses indicate that no fuel channel burnout will result from a BPA power loss to the process pumps.

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

plate and to calculate its critical dimensions. Calculations were made which show that the plate and positioning keys meet the design requirements. Work sheets pertaining to the evaluation are included along with references to drawings. (J.R.D.)

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.

Tests were performed on a thoria slurry flowing in a pipe to determine the magnitude of the possible error involved in the sampling process. Evidence indicates that a correct sample is obtained by withdrawing the sample isokinetically (i.e., by facing the sampler into the flow and adjusting the sampler velocity to match the ambient velocity) provided that the sampler is larger than some minimura diameter that is dependent on the mean eddy length and/ or the mean particle size. (auth)

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.