Basic surface-water and precipitation data collected from June 1961 through June 1965 at 10 stations in Lower Little Creek basin of Lamar and Marion Counties, Mississippi, are presented. The U.S. Atomic Energy Commission's Tatum dome site of Project Dribble is in Lower Little Creek basin. The data show a 3.80-inch deficiency in precipiation in the Tatum dome are from April 1964 through June 1965 and an 11.72-inch deficiency from June 1961 through June 1965. The mean monthly discharge at Lower Little Creek (4892.40) was less than the median monthly value during 11 of the 15 months from April 1964 through June 1965. Chemical analyses of water samples collected from five streams in the Tatum Dome area indicate that the water is of generally good quality; for example, dissolved-solids content ranged from 16 to 23 parts per million for samples collected during the period from October 31, 1961, through October 23, 1964. 4 figures, 4 tables.

The current specifications for the cladding on {open_quotes}N{close_quotes} fuels were established early in the course of process development and were predicted on several basic considerations. Among these were: (a) a desire to provide an adequate safety factor in cladding thickness to insure against corrosion penetration and rupture from uranium swelling stresses; (b) an apprehension that the striations in the zircaloy cladding of the U/zircaloy interface and on the exterior surface might serve as stress-raisers, leading to untimely failures of the jacket; and (c) then existing process capability - the need to maintain a specified ratio between zircaloy and uranium in the billet assembly to effect satisfactory coextrusion. It now appears appropriate to review these specifications in an effort to determine whether some of them may be revised, with attendant gains in economy and/or operating smoothness.

Daily river temperature cycling has existed at reactor area river pumphouses from the beginning of operations. Of small account at first and due only to the daily air and solar radiation cycle, the phenomenon became significant with the initiation of daily river flow cycling at priest Rapids Dam to meet changing power generation requirements. The problem was compounded by the construction of the N reactor, placing a water intake and outfall at approximately half the previous distance between the K and D reactors. The decidedly non-uniform effluent distributions in the river at the shorter distances has caused apparent daily temperature swings at 181-D. This distribution was determined by the use of rhodamine B dye. Estimates of the relative contributions over a range of river flow to the temperature cycling at 181-N were made. A similar evaluation of the relative contributions at low river flow only at 181-D from K and N reactor operations has been requested by Douglas United Nuclear, Inc. This report is made to satisfy that request.

This monthly report details activities of the 200 Area for the month of April 1966.

This document provides details of the Report to the FEDC Working Committee for the period of January to April 1966. N-Reactor fuel development and production is discussed.

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Daily check of 325 stack filters on 5/9/66 revealed contamination released to the atmosphere within the operating area. Investigation indicated that about 0.4 Ci (mostly Ce-144) had been released as a result of steam jet transfer of some Ce-141-144 from a cask to a storage vault in 325-A. on 5/9/66. A further release of about 0.2 Ci vas indicated by survey of stack filters 5/12/66 following completion ot the transfer. Actions taken include vehicle smears, ground surveys around the 300 area fence line, and smears of the 329 loading dock. Daily check or 325 stack filters on 5-9-66 shoved an increase from about 8000 c/m (GM) Friday afternoon to about 450 mrads/hr (CP) Monday afternoon. Vehicle smears on 4-12 shift 5/9/66, ground surveys around the 300 Area fence line, and smears of the 329 loading dock have not revealed any contamination spread. Inventory in 325A (5/10/66) includes about 3 x 10{sup 6} Ci Ce-141-144 and about 3 x 10{sup 6} Ci Pm-147.

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This report has the following sections: Summary of Activities; Civilian Power Reactors; Applied and Reactor Physics; Reactor Fuels and Materials; Engineering Development; Plutonium Recycle Program; Advanced Systems; and Nuclear Safety.

This report describes the procedures and techniques involved in area reduction measurements of fractured tensile specimens.

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This document provides the report to the working committee of the fuel element development committee for small and K reactor production fuels. Topics discussed are: Uranium core production data; uranium specification; future planning -- five year R&D program; thoria development; heat treating; UO{sub 2} irradiation; and alternate process development.

Isotope production calculations have been prepared in the past to indicate present and potential reactor capability; however, the cases evaluated generally considered production of only one or possibly two isotopes. With the wide variety and potential demand for numerous isotopes, the question arises as to what the production capability would be if several isotopes were being produced at the same time. To study this problem in more detail and determine which of several alternates may be used in meeting production requirements, the Case Generator and Evaluator System has been developed. This document presents a production matrix and reviews the assumptions used to derive the matrix. This matrix should be useful in planning future CAGE isotope cases.

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