Procurement and installation are proposed of a 50-kW, high-frequency induction heating unit for can-sleeve preheating prior to dip brazing to the uranium core.

Production test IP-536-A authorizes the irradiation of N Reactor inner fuel tubes in KER-1 and KER-2. Supplement B, PT-IP-536-A, authorizes the irradiation of N Reactor inner fuel tubes prepared from a U-2 w/o Zr alloy. Both tests require that specific operating conditions and trip settings for particular loadings be approved by the managers of the Process and Reactor Development Sub-Section and the Process Technology Sub-Section prior to charging. The purpose of this document is to provide the process tube loading, high and low temperature operating limits, and water shutoff times for a charge of twelve 17.3-inch N inner fuel elements or twelve 18.5-inch U-2 w/o Zr elements in KER-1 and KER-2. The operating conditions were prepared for the 17.3-inch elements and with the same downstream spacer column, are conservative for the slightly longer U-2 w/o Zr charge.

The purpose of this study was to determine the maximum forces that can be imposed by the graphite moderator on prospective VSR channel sleeves. In order to do this, both the origins and modes of transmission of the forces were determined. Forces in the moderator stack that are capable of acting on a block or group of blocks may originate from any of the following primary effects: Contraction of graphite due to irradiation; thermal expansion of graphite; frictional resistance to motion; resistance from keys; gravity; and other.

Three conditions representing various quantities of 10-inch and/or 12-inch fuel elements were presented by the Industrial Engineering Group, IPD, as a basis for initial estimates of economic incentive to partially convert the Production Fuels Section AlSi process to longer fuel elements for FY 1964. An earlier study (Ref. 1) pertained to 10-inch instead of 6-inch enriched fuels. The comparatively larger quantities of the 4 in. longer fuel for full E-N loading showed a larger payout than for the requirements shown in this study. Since the estimates have been prepared without the benefit any process development or equipment design, the validity of the results of this study rest directly with the assumptions. These assumptions are itemized in some detail to explain the basis from which payout was calculated. At best, the study can be used as a preliminary estimate, subject to changes as more detailed analysis are made.

Progress is reported on the thorium fuel cycle development under two topics: sol-gel process development and fabrication and material development. Separate abstracts were prepared for the two topics. (M.C.G.)

3 8 ; 9 4 5 7 5 C 9 E ; < : 7over a 1456-day period. The half life was calculated to be 12.355 plus or minus 0.0l0 years, based on measurements in three calorimeters. The yields of /sup 208/Po and /sup 209/Po from the irradiation of multicurie quantities of /sup 210/Po were calculated. Production of small quantities of /sup 208/Po and /sup 209/Po is possible by the (n,xn) reaction; however, large quantities of /sup 210/Po would be present in the /sup 208/Po--/sup 209/Po sample. On the basis of these calculations the maximum /sup 208/Po content would occur after 466 days' irradiation in the materials testing reactor. An experimental decay scheme was drawn for /sup 208/Po on the basis of experimental data. Experiments are in progress to determine whether levels at about 1.4 Mev in /sup 208/Bi are populated in the electron capture decay of /sup 208/Po The known half-life method for determining the resolution time of a counting instrument was found to yield only an average value for the counting range in the computation. To correct the counting error a series of computations should be made with data covering only a short segment of the decay curve …

The diffusion of lithium in aluminum was measured at various temperatures with diffusion couples of aluminum-LiAl. The activation energy, E, is 33.3 kcal/mol, and the diffusion factor, Do, is 4.5 cm{sup2}/sec. (auth)

This report describes some present practices in regard to power calibration techniques and goes on to propose an automatic calibrator based on thermal power.

A detailed radiation survey of the Shippingport Reactor vessel head was made, and the magnitude of the radiation in the vicinity of the vessel head after extended power operation and the effect of normal rod motion and scrams on this magnitude are reported. (L.T.W.)

In two previous reports nucleon-meson cascade calculations were carried out for several cases of interest in the design of the transverse shield for the proposed 45-Bev linear electron accelerator at Stanford University. Results are now given for two additional cases. (auth)

An investigatiori of the annealing of the radioinduced carrier concentration change in Sb-doped Ge in the range 370 to 455 l K was made. The irradiations were conducted at liquid nitrogen temperature using Co/ sup 60/ gamma irradiation. A model that explains the observed behavior is presented. On the basis of the model, the observed annealing consists of vacancy diffusion simultaneously to impurity sites and annihilation centers. Analysis of the activation energy for the annealing process yields values of 0.8 to 1.4 ev in agreement with the range of energies that were attributed to vacancy motion but that cannot be resolved into unique components. The complex activation energy is explained by the model in terms of the impurity concentration. It was observed that the change in carrier concentration saturates before complete annealing is achieved. The saturation, which is stable for further annealing at higher temperatures, is also explained in terms of the model. The vacancies are considered to diffuse to annihilation centers, such as dislocation lines, and to the site adjacent to an Sb atom. Those that go to an Sb are trapped. The Sb- vacancy complex can break up to supply a vacancy back to the system or can …

Declassified 24 Sep 1973. The purpose of this work was to develop a process for recovering the uranium from spent Rover fuels. Only one reactor is used, and the process involves a 4-hr combustion of the fuel in oxygen at about 800 deg C, a 4-hr chlorination of the U/sub 3/O/sub 8/-Nb/ sub 2/O/sub 5/ ash in refluxing hexachloropropene at 180 deg C, dissolution-extraction of the UCl/sub 4/ and NbCl/sub 5/ products at room temperature by dilute nitric acid, and extraction of the uranium from the resulting acid solution with 30% TBP in Amsco diluent. The results indicate that an extract containing 50 g of uranium per liter can be produced in seven or eight extraction stages, with total uranium losses of less than 0.02%. Corrosion rates of several possible construction materials during chlorination are less than 0.1 mil/month. Problems in the process involve handling about 10% of the niobium as a solid during the liquid- liquid separations, and handling solutions containing chloride. The results of this laboratory-scale work indicate that the liquid-phase chlorination and subsequent extraction operations are reducible to large-scale practice, since these operations resemble the liquid-phase operations typically performed in radiochemical separation plants. (auth)

In a previous report nucleon-meson cascade calculations were carried out for several cases of interest in the design of the transverse shield for the proposed 45-Bev linear electron accelerator at Starford University. In this report results are given for several additional cases. Muon, neutron, pion, and proton intensities as function of energy and distance for varying angles are given. Corresponding doses are also included. (D.C.W.)

The Experimental Beryllium Oxide Reactor is described. The configurations and properties of BeO blocks are discussed. Various BeO shapes were fabricated and tested. BeO specimens of selected compositions and densities were irradiated. Dimensional changes due to radiation were in the range below 1.4% as determined from micrometer measurements. Some of the specimens in the highest exposure region of the capsule exhibited radial fractures. (M.C.G.)

ABS>A detailed description of the printed output sheets from the intranuclear cascade calculation described in ORNL3383 is given. The three analysis codes --analysis codes I and II and an evaporation code --that were written to organize the raw data of the output tape are considered. (auth) A tabulation of the neutron total cross section of U/sup 233/ as a function of neutron energy from 0.07 to 10,000 ev, measured with the ORNL fast chopper time- of-flight neutron spectrometer is given. (auth)

The anisotropy of mechanicai properties in Zircaloy-2 associated with preferred orientation developed during manufacture has caused many difficulties in the fabrication and utilization of mill products. A preliminary examination of the development and utilization of texture preparatory to an experimental study of anisotropy in Zircaloy-2 tubing is presented. The development of preferred orientation during manufacture of Zircaloy-2 plate, strip, and tubing is qualitatively analyzed in terms of the plastic strain undergone, the modes of deformation operative, the existing anisotropy of flow strengths, and the effects of forming forces of the rolls or dies. The problem of forming structural shapes from anisotropic mill products is discussed, and the utilization of the anisotropy of mechanical properties to ease manufacture and to strengthen structures is considered. (auth)

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Distortion of reactor-grade graphites was studied at varying neutron exposures ranging up to 14 x 10/sup 21/ neutrons per cm/sup 2/ (nvt)/sup */ at temperatures of irradiation ranging from 425 to 800 deg C. This exposure level corresponds to approximately 100,000 megawatt days per adjacent ton of fuel (Mwd/ At) in a graphite-moderated reactor. A conventionalcoke graphite, CSF, and two needle-coke graphites, NC-7 and NC-8, were studied. At all temperatures of irradiation the contraction rate of the samples cut parallel to the extrusion axis increased with increasing neutron exposure. For parallel samples the needle- coke graphites and the CSF graphite contracted approximately the same amount. In the transverse direction the rate of cortraction at the higher irradiation temperntures appeared to be decreasing. Volume contractions derived from the linear contractions are discussed. (auth)

The use of anion exchange for final Pu purification in the Redox facility requires the adjustment of Pa(VI) to Pu(IV). Two methods of reduction were investigated: ferrous sulfamate and nitrite. The ferrous sulfamate method was adopted. Data on reduction of Pu(VI) to Pu(IV) and Pu(III) by nitrite are presented. The reduction by nitrite was found to be dependent on nitrite concentration, temperature, acidity, ferric ion concentration, and time, and it is possible to obtain &gt;90% reduction to Pu(HI) by controlling the above conditions. (D.L.C.)

A fuel element hold-down device, which is an integral part of the fuel element assembly, was developed to lock reactor fuel to the bottom grid plate. Concurrently with the hold-down latch development, a fuel bandling machine grapple capable of setting, releasing, and checking the hold-down device was developed. (auth)

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The weights and thicknesses of two-layer W-- LiH shadow shields were calculated for SNAP-2, -8, and (assumed) -50 reactors. The shields are shaped as spherical sectors with haif-angles up to 30 deg and are sized so that no part of the reactor is visible within the volume subtended by the shield. Weights and thicknesses are presented for both manned and unnaanned vehicles over a wide range of reactor thermal powers, operating durations, reactorpayload separation distances, and allowable radiation dosages. The thicknesses of the W and LiH layers are optimized to provide a minimum weight shield. (auth)

This study was undertaken to answer questions asked regarding the required rod stroke for control of modified Tory II-C reactors. All problems described were solved with the Angie code and based on Tory-II-C design problems, RZ 501 and RZ 502, representing hot and cold reactors respectively.