The pressure drop of a new fuel element design concept of spacing rods by means of helical wire springs was investigated experimentally and analytically.

"The quantum theory problem of constructing explicitly a local Lorentz invariant model field theory which has an asymptotic particle interpretation and gives rise to an S matrix different from unity is discussed. It is proved that if a local covariant field H(x) has a complete current, the S matrix associated with this field is identically equal to unity. Results show that in order to get an interaction, the current is not allowed to have a terminating expansion in terms of a free field. The absence of a finite connection between a free field and an interacting field tended to confirm the belief that in local quantum field theories with particle interpretation the requirement of the existence of scattering forces in the physical states are such that it is not pos sible to identify them with a linear space spanned by ""free'' particle states."

(p,xp) reactions are those nuclear reactions induced by high energy protons in which both the mans and the charge of the target nucleus are reduced by x-1 units. The most common type of these reactions to be expected in the 100 to 400 Mev region involves the exclusive emission of protons as a results of the interactions. The specific (p,xp) reactions investigated in this study were those where 2 is less than or equal to x where x is less than or equal to 5.

The development of swaging and vibratory compaction process for fabrication of clad UO2 fuel rods is described. The cost is less than 50% of that for an equivalent core fabricated from pellets. The irradiation testing of vibratory filled and swaged UO2 rods is reported for burn-ups from 2,000 to 12,000 Mwd/t UO2; the results indicate that the rods are capable of excellent performance power reactors.

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.

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.

Technical report outlining the use of circular weirs in solvent extraction between liquid phases of nuclear fuel elements. From Abstract : "A correlation of the experimental coefficients was developed to allow calculation of general weir performance."

Technical report. From Abstract : "A semicontinuous process was developed for dissolving the solid residues that are produced in the calcium reduction process for converting plutonium fluoride to plutonium metal. The slag and crucible residues are charged to a column-type dissolver, and are there exposed to a flowing solution of hot nitric acid followed by a nitric acid - aluminum nitrate solution. The resultant solution is suitable as feed for an ion exchange process for recovering plutonium."

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 …

Abstract: The boron-carbon equilibrium diagram has been determined by X-ray, metallographic, and thermal analysis of sintered and arc-cast alloys. A single carbide having a range of solubility from approximately 9 to 20 a/o carbon and melting congruently exists in the system. The terminal solubility of carbon in boron is 0.1-0.2a/o. The freezing reaction at the composition and melting temperature of elemental boron; there is a eutectic reaction at 29a/o carbon and 2375 degrees Celcius. No allotropy of boron was observed.

This is the fourth quarterly report that documents a system of mass flow that can record measurements of homogeneous flow, slurries, highly corrosive fluids and multiphase fluids, additionally considering pressure drops, measuring external to the flow, ruggedness and reliability.

The following report follows geological investigations made on the 410 Area of the Nevada Test Site.

The following report covers the study of a small composite stock at Oak Spring, Nevada, known as Climax granite to study the seismic effect of a deep nuclear explosion contained in a large volume of unfractured rock.

From introduction: This summary report of the geology of the Appalachian basin was prepared as part of the Radioactive Disposal Program being conducted by the USGS on behalf of the Division of Reactor Development of the AEC.

Introduction: The Laboratory follows the recommendations of the National Committee on Radiation Protection concerning permissible exposure limits for those working with radiation. A system of controls and regulations has been established which should ensure that no person at the Laboratory will receive more than the permissible exposure. This manual is intended to provide the minimum background necessary for intelligent cooperation with the regulations.

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.

From foreword: This report contains the test results and analysis of test results of flow studies and reactor vessel internals. Tests were performed at, below, and above the PM-1 design flow rate to observe flow distribution and pressure drops within the reactor vessel, as well as the overall reactor vessel pressure drop.

From summary: The SNAP 7 program is being conducted for the purpose of developing four radioisotope-fueled thermo-electric power generation systems. An important phase of this program is the processing of Strontium-90 into heat sources for these systems. The current effort involves the design and engineering analysis of two of the four thermo-electric generators.

This volume describes the final results from completed investigational programs intended to test the performance of various concrete materials exposed to weathering.

The Stationary Low Power Reactor No. 1 (SL-1) was a boiling water reactor demonstration plant for remote military bases. The plant had been operated by Combustion Engineering, Inc. from February, l959 until January 3, 1961 at which time a nuclear excursion rendered the plant inoperable, This report summarizes the of operations of maintenance, test, health and safety, and administrative activities for the year February, 1960 until the incident.

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).

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

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.

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.