This report summarizes experimental work completed on Subcontract AT-147.

Foreword: This report describes results of a series of evaluation and design studies conducted between July 1, 1960 and June 30, 1961, related to heavy water moderated power reactor plants.

Abstract: This technical report contains a description and results of a second series of critical experiments performed on the SM-2 core mock-up, as additional to the first series of experiments reported in APAE No. 54. The SM-2 core mock-up contains 36.4 kg U-235 and and estimated 67.9 gm B-10. The equivalent diameter and the active height are about 22 in.; the metal-to-water volume ration is 0.344. Data is presented on activation, reactivity, and stuck rod measurements. All measurements were conducted on the open seven control rod array employing 38 stationary fuel elements. Activation measurements consisted of neutron flux measurements using uranium fission foils for relative power distribution studies, the effect of flux suppressors on reducing power peaks, blocked coolant channel measurements, and gamma ray dose distribution. Reactivity measurements were performed to determine the effect f flow divider, flux suppressors and stimulated high temperature and pressure operation; b-10 loading in the SM-2 core; and core material coefficients. For the later, the worth in cents per gm or cents per cc was determined at simulated temperature of 510 degree F for B-10, U-235, stainless steel, and void. Stuck rod measurements were made to obtain an indication of the criticality margin in the …

A detailed analysis was made of the power distributions in the SM-2 experimental core at 68 deg F and the SM-2 reference core at 510 deg F. This analysis supersedes the power distribution calculations presented in APAE No. 69. The calculated distributions for the experimental core were normalized to measured data wherever possible in order to obtain corrections factors for application to the reference core. The over-all power distributions were calculated by synthesis of one-dimensional axial calculations with twodimensional radial calculations. The variation of the power distribution with fuel burnup is also presented. In order to improve the agreement between measured and calculated axial power distributions, flux-weighting the nuclear parameters in the rods-in and rods-out regions was investigated. (auth)

Plastic Research. The tensile strength of Dacron-filled diallyl phthalate was determined to average 4377 psi. Composition and stress-strain data are tabulated for ten adhesive films. Analytical studies of an adhesive exudate are reported. Radioelements. Results of analysis of ioniumbearing raffinates and residues for Th/sup 232/ and of aged>s Ra/sup 223/ for Ac/sup 2/2/sup >/s7>s are given. Progress on Pa recovery from raffinates and residues and separation from Nb is reported. Isotope Separation and Purification. Proposed work on gas centrifugal and photochemical separation of uranium isotopes is discussed. Progress on xenon and helium isotopes separation and purification is outlined. Reactor Fuels and Materials Development. The density of liquid La at 945 to 1000 deg C was determined. The performances of an oscillating Cup viscometer with La and Bi and of a high-temperature calorimeter with Po/sup 210/ are described. A study of the compatibility of Haynes 25 alloy with Pu at 9O0 deg C indicated that very little penetration took place after the first hour at 900 deg C. The efficiency of escape of alpha particles from a glass fiber containing 10 wt% pu oxide was determined to be approximately 72%. Eight glass compositions were evaluated for their ability to dissolve 15 …

None

A test was carried out to determine the adequacy of storage capacity and operating procedures of the radioactive waste disposal system during a normal reactor plant warmup. The capacity and operating procedures were found to be adequate. It was impossible to perform a complete material balance based on existing level instrumentation and using the data required by the test procedure. Approximately 21,290 gal. of waste were received in the system and 13,210 gal. were discharged to the river with a total activity of 1200 mu c. A quantity of 6,670 gal. of reactor coolant effluent was processed. Approximately 634 lb of combustible waste were incinerated. (M.C.G.)

The plant design for PL-2, a 1000-kw net electric direct cycle boiling water nuclear power plant, is presented. The design includes all buildings, foundations, and structures required for the installation of the plant in a snow tunnel. (M.C.G.)

Plant drawings for the final design for the Army Reactor (PL-2) are presented. Two hundred and twenty-eight figures are included. (M.C.G.)

Specifications for the plant equipment for a P L-2 nuclear power plant are given. (M.C.G.)

Drawings andd specifications for the PL reactor core are given. The requirements for the material procurement, fabrication, testing, and inspection of one P L-2 reactor core and spares are listed. (M.C.G.)

The physical and mathematical reactor models which are used in Program ODD are discussed. In addition, the FORTRAN II source program listings, decimal data input sheets, and input and output for a sample case are given. Program ODD was designed to raake use of the Revised Nuclear Data System at ANPD which consists of twenty-five energy group cross-section data including high energy inelastic scattering matrices, resonance parameters for the resolved resonances, and thermalization scattering matrices for the near thermal energy region. The most unique aspect of the program is the mathematical technique employed for eliminating inner iterations and slow convergenc rates occasioned by the up- scattering'' in the thermalization region of the energy lattice. Direct inversion of the energy matrix coupling the thermal and last four epitherma groups provides simultaneous consistent solutions for thes groups within each power iteration. (auth)

The study of meteorite Murray has been reported in previous Quarterly Reports. This report gives further results with Murray, and information on another meteorite, Orgueil. A sample of Orgueil was sent from the Museum National d Histoire Naturelle, Paris. It fell in several pieces over an area of 2 square miles near Orgueil, France, in 1864. The elemental analysis of this meteorite is shown in Table 1. They extracted a 10.07-g sample of this meteorite with water, using the same procedure as that for Murray. The water extracted 1.32 g, which is at least twice as much material as was water-extracted from Murray. The elemental analysis of the water extract is given in Table II and its uv spectrum is shown in Figure 1. From an x-ray diffraction pattern it was determined that the water extract contained mostly MgSO{sub 4} {center_dot} 6H{sub 2}O with some calcium sulfate. Their spectrum (Figure 2) shows a strong SO{sub 4} band at 1100 cm{sup -1}, = strong H{sub 2}O bands at 1650 cm{sup -1} and 3200-3600 cm{sup -1}, and some unidentified peaks at 2300, 1400, and 980 cm{sup -1}. The approximately 8 g of Orgueil left after the water extraction was then extracted with …

Activities in a program to obtain UC pellets of maximum density for irradiation testing are described. A study was made of the effects of the UC powder thermal history, sintering time, sintering temperature, and additives on the UC sinterability. Preliminary results indicate that the UC powder synthesized under the most severe conditions of temperature and time (1740 deg C, 107 min) was less sinterable than powders prepared at lower temperatures, or at the same temperature for shorter time. It was also concluded that although both temperature and time of sintering pellets affect density, temperature is more important. Additions of Fe (0.1 to 1.0%) improve sinterability and promote grain growth. The final density obtained on sintering in a vacuum was about the same as that resulting from sintering in He at atmospheric pressure. (J.R.D.)

This report details model studies performed as required by the design, development, and research contract between the General Electric Company and Washington State University. These studies provide analytical considerations for K-Downcomer and Bellows.

Current NPR fuel production plans call for a thick layer (0.030 inch or greater) of 5% Be + 95% Zry-2 braze alloy in the closure region. This requirement was imposed to eliminate many of the production welding problems brought about by the presence of the low melting braze alloy between two surfaces of Zircaloy. All GEH-4 irradiations in the past have involved very thin braze lines (0.015 inch or less). As a part of the fuel evaluation program it is essential to run a comparative irradiation to determine what effect the braze line thickness has on the stability of the fuel closure. For this purpose three fuel elements were prepared, two with a braze thickness of 0.030 inch and one with a braze thickness of 0.015 inch. To provide a more severe thermal stress, the I&E geometry was used. Five MTR cycles should be sufficient to test this fuel concept. This report details this test proposal.

Report containing ongoing projects and experiments undertaken by the Oak Ridge National Laboratory's Molten-Salt Reactor Program.

Activities are discussed for work done on the design, components development, and engineering analysis of the MSRE, and materials development studies including metallurgy, in-pile tests, chemistry, engineering research, and fuel processing. (B.O.G.)

Document issued by the Office of the Attorney General of Texas in Austin, Texas, providing an interpretation of Texas law. It provides the opinion of the Texas Attorney General, Will Wilson, regarding a legal question submitted for clarification: Whether bonuses received by a corporation as consideration for the execution of mineral leases constitute surplus for purposes of calculating the franchise tax levied by Chapter 12 of Title 122A, Vernon's Civil Statutes.

The KSE-3 element, a 1.6% enriched Zr-2 Jacketed tubular element nominally 1.74 inch O.D. by 1.05 inch I.D., was designed for irradiation in the KER loops to simulate the behavior of an N-reactor outer fuel tube. Three charges of these fuel elements have been irradiated under PT-IP-363-A: one in KER-2 to 1985 MWD/T, one KER-3 to 3555 MWD/T, and one in KER-4 to 1.195 MWD/T. This document provides the calculated powers and temperatures for each fuel element during the time it was irradiated.

An experimental investigation of the interdiffusion and forced-flow behavior of helium and argon in Speer Moderator No. 1 graphite was performed. The data were employed to determine a mutual diffusion coefficient and to verify certain superposed-flow equations. In addition, two series of experiments at high values of the forced-flow component were conducted to investigate contributions of the backdiffusion mechanism of those pores whose diameters are equal to or smaller than the mean free path of the gas molecules, approaching Rhudsen or free-molecule difiusion. At small forced-flow rates, normal diffusion was the controlling diffusion mechanism, while Knudsen effects were negligible. Flow equations employed previously are applicable to these data. Experiments conducted at high forced-flow rates show the contribution of small channels, which appears to follow the Knudsen diffusion mechanism. A critical value of sweep rate was determined. If the sweep rate is lower than the critical, the contamination will increase, whereas sweep rates greater than this would require large reprocessing capacities without additional decrease in contamination. (auth)

None

Uranium ore concentrates from seventeen world sources were compared to a standard to determine variations in isotopic content. A spread of about 0.06% in U/sup 235/ content was indicated for the concentrates analyzed. Domestic sources showed much wider variations than those from other parts of the world. (auth)

This report consists of the description, drawings, connections, and schematics of the various control elements that make up the control system of the Engineering Test Reactor Critical Facility (ETRC).