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.)

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.)

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)

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.)

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)

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

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.

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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.)

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 …

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

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 …