Apparatus, field layout, miscellaneous testing, and the method of analysis of a 100-ton test at Trinity.

Experimental combination hand and shoe counter redesign as a compact, simplified, register readout instrument no larger than the present Five-Fold called Model 2.

From introduction: "The development of one or more nuclear steam supply system concepts, with certain trade-off studies to aid in the definition of these concepts. The selection of a reference concept for further study."

Abstract: "This report describes a flame photometric system which continuously monitors a process water stream for ppm quantities of alkali metals, and automatically diverts the stream when the contamination exceeds a pre-determined level."

Fourier refinement analysis of the uranium parameters in uranium tetrafluoride.

From introduction: "The work described in this report represents the Joint United States - European Atomic Energy Community effort which is in keeping with the spirit of cooperation in contributing to the common good by the sharing of scientific and technical information and minimizing the duplication of effort by the limited pool of technical talent available in Western Europe and the United States."

Describes the current prototype model of a linear amplifier, two-channel pulse-height analyser, two-channel count-rate meter, and count-rate difference detector for a gamma-ray spectrometer monitor developed to approach the above design criteria.

Report describing a 300-Mw(e) Supercritical Pressure Power Reactor's facilities, physics, economics, and problems encountered during its development. Appendix begins on page 103.

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).

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 …

Information on experimental facilities and reactors.

A test program conducted to develop high integrity piping and seal components for Plutonium Recycle Test Reactor installation. Methods of preventing leakage of costly D2O coolant and maintaining operational safety were investigated.

Progress report discussing the activities in "a program of studies of nutron-deficient nuclides using the techniques of gamma-ray spectroscopy" (abstract).

Abstract: This technical report describes the changes incurred in the PM-2A by the planned insertion of PM-2A Core II and the replacement of the startup and check sources. PM-2A Core II components were fabricated to specifications very nearly identical to those of PM-2A Core I. The essential difference in the cores is the boron loading which permits PM-2A Core II to meet a "one-stuck rod criteria" at beginning of life. This core has been subjected to a zero power experiment and loading procedures have been developed at the Alco Critical Facility. The nuclear and thermal and hydraulic characteristics are essentially identical to those of Core I and the replacement of the startup and check sources represent no increase in the potential for or magnitude of a hazardous situation.

A series of calculations are described which evaluated the heat transfer and fluid flow aspects of the Plutonium Recycle Test Reactor Pressurized Gas-Loop Facility.

From abstract: "Describes Heat Transfer Reactor Experiment No. 3, a solid-moderated nuclear power plant with a horizontal reactor."

Evaluation of the conditions under which the interaction effects of two or more sub-critical quantities of enriched uranium may be safely arranged.

Abstract; SM-1 Reactor Core I Rearranged and Spiked, and Core II with Special Components were analyzed under various off-design conditions to induce nucleate boiling. The steady state code, STDY-3, written for the thermal analysis of pressurized water cores, was employed for the analysis. The code performs a complete steady state parallel channel thermal analysis for both nominal and hot channels. Thermal characteristics of individual elements were investigated while changing the parameters of primary pressure or inlet temperature to introduce the phenomenon of nucleate boiling in the the core. Reduction of system pressures to 1000, 800, and 600 psia and increasing core inlet temperatures to 465 and 500 degree F were studied as the means to induce boiling in the core. This analysis indicates that SM-1 Core I Rearranged and Spiked can be safely operated at the reduced pressure of 910 psia without introducing extensive boiling in the core. SM-1 Core II with Special Components can be operated at 800 psia or at an inlet temperature of 500 degree F at 1200 psia.

From foreword: "This is a documentation of the experiments performed and the data accumulated during the early period of the Experimental Gas-Cooled Reactor (EGCR) fuel element design."

Work required, proposed schedule, and an estimate of the costs for complete program to build first LMFR power plant.

Series of isothermal transformation studies for improvement in fabrication and reactor performance of uranium fuel elements.

Detailed study of the mathematically developed end-point method and its application to the Milne kernel.

This purpose of this report is to establish a means of determining the mean effective temperature at which any specimen of Al-Si had been annealed.

Abstract: A cyclic stress analysis of the SM-1 primary system was carried out. Problems encountered in the fabrication of PM-2A Core II and SM-lA Core II are described, and the results of an examination of damaged SM-lA Core I stationary fuel elements reported. A preliminary study of the radiation damage to SM-1 reactor vessel was made and the possibility of annealing the vessel discussed. Performance analyses are presented for five cores: SM-1 Core, SM-1 Core 1 rearranged and spiked, SM-1 Core II with special components, PM-2A Core 1, and SM- 1A Core 1. Preliminary critical experiments were made with SM-2 elements in a SM- 1 core configuration and nuclear and thermal analyses of the use of SM-2 elements in SM-1, SM-1A, and PM-2A completed. A throttling steam calorimeter was selected for measuring moisture carry-over on the PM-2A steam generator. Test procedures for evaluating the shielding of the SM-1, SM-lA, and PM-2A plants are summarized. Radiochemical and chemical analyses of SM-1 coolant and crud are summarized, and methods of activity control discussed. Preliminary results of studies of the properties of reactor pressure vessels under irradiation and no irradiation conditions are summarized briefly.