The development of nuclear reactors and other sources of nuclear particles has provided the analyst with a new method which has been successfully applied to the determination of microgram and even submicrogram quantities of many elements. In this method, known as "radioactivation analysis," the element to be determined is "activated" through some type of nuclear reaction which produces a radioactive isotope of the element. Since the radioisotope produced decays with its own characteristic radiation and half-life, it is possible, through radiochemical measurements following radioactivation, to devise a method of analysis which is very specific for particular elements. Whenever necessary, the radioisotope is separated by chemical means and its radioactivity is then measured by some type of radiation counter.

Tests in a one-fifth scale clear plastic model were conducted to investigate the flow characteristics of the asymmetrical riser configuration in the Consumers Big Rock power plant.

The second and third law methods of thermodynamics are used to obtain the enthalpies of vaporization of the halides of Be, Mg, Ca, Sr, Ba, Zn, Cd, Hg, Sn, Pb, Ti, V, Cr, Mn, Fe, Co, Ni and Cu. Molecular and free energy data necessary for such calculations are presented. The structure of bonding in these molecules is briefly discussed.

"The design, analysis, and selection of system components for functions related to secondary shielding and associated auxiliary systems for the Antarctic PM-3A Nuclear Power Plant are described. Backfill material characteristics, shielding and activation analyses, heat transfer analysis, backfill cooling system, water exclusion and removal system, and instrumentation and control are discussed."

"A comparison of grafting to two thicknesses of polypropylene film is made using isotope and electron accelerator initiation. The results indicate that the grafting of methacrylic acid: styrene to polypropylene is mainly a surface reaction."

The extraction of beryllium by triisooctylamine is measured over the pH range from 1 to 10, in the presence of malonic, maleic, succinic, phthalic, and salicylic acids. Values are obtained for the stability constants of the various metal-acid complexes by potentiometric titration. The degree of extraction is dependent on the abundance of the anionic complex present in the aqueous phase and this is in turn dependent on the strength of the corresponding acid and the stability of the complex. (auth).

Abstract: This technical report describes the as-constructed SM-1 penetrated gasket designed for SM-1 Core and Flow Instrumentation (Task XIV). This report supplements APAE No. 79, The Summary Hazards Report for Task XIV, and evaluates the effects of a postulated failure of this gasket. The effects of failure on the Maximum Credible Accident are determined and conclusions and recommendations for the use of this gasket are made.

Abstract: This technical report describes the Martin Co. PM-1-M-2 test element and analyzes the potential hazard incurred by its inclusion in the SM-1 Core II. A nuclear analysis develops power distributions and reactivity effects. Hydraulic and thermal analyses develop anticipated burnout heat flux ratios. An evaluation of the risk involved with the inclusion of this element is presented. In view of the narrow margin by which the PM-1-M-2 test element meets the minimum burnout ratios as defined by Alco Products, Inc., it is recommended that if time permits that critical facility design verification be accomplished. The PM-1-M-2 test element meets the minimum requirements for insertion in SM-1 Core II and in view of the importance of this element to the PM-1 and PM-3A program, should be considered for insertion.

Results of zero power tests to confirm the nuclear design of the Consolidated Edison Thorium Reactor Core are presented. The maximum core reactivity, control rod worths, and power distribution were measured.

Abstract: Thermal characteristics of the SM-2 core were analyzed at steady state and loss of flow conditions. For steady state operation, the steady state code STDY-3 was used. For transients during a loss of flow accident, ART-02, a one-dimensional code, was used. This analysis indicates the SM-2 core is safe from burnout under steady state operation at design power level (28 tMW) because (1) no nucleate boiling exists, and (2) the minimum burnout ratio is above 2.0. The core is safe from burnout under loss of flow transient because the minimum burnout ratio in the hottest element channel of 1. 82 is above the minimum design criteria of 1. 5.

Abstract: This technical report presents design criteria to prevent the brittle fracture of ferritic reactor vessels that cold occur as a result of the rise in NDT caused by fast neutron irradiation. The criteria require that maximum principal stress in the vessel does not exceed 18 percent of yield stress at temperatures below NDT + 60 degree F. Under certain conditions the allowable stress may be based on the irradiated yield stress. A discussion of brittle fracture and an explanation of the criteria are included.

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This report is divided into: production (fuel elements), engineering (present reactor fuels, N Reactor development), and general (AlSi shop). (DLC)

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Abstract: "Thirty-three specific landing sites were investigated in the ice-free land area of East Greenland between Scores by Sund and Loch Fyne. Eight of these are considered suitable for emergency landings in summer by heavy cargo planes, and several more for light cargo planes. Several sites were investigated for the Royal Greenland Trade Department in the Scorebysund - Kap Tobin area. A 1550-foot airstrip was located on a gravel terrace in the Jaettedal, eight miles northwest of Kap Tobin, and a short strip requiring some construction work was located near Kap Tobin. An 11,500-foot airstrip was tentatively laid out on a gravel terrace at Storelv, near Moskusoksefjord. Utilization of several of these sites can add a significant safety factor to commercial or military aircraft operations in East Greenland. Reconnaissance observations verify the presence of abundant emergency sources of fresh water in East Greenland; analyses of 36 samples indicate water of good to excellent chemical quality."

It is the purpose of this document to provide assistance to the Manufacturing Section personnel in determining their future operating plans. In general, the inter-relationship of such engineering parameters as projected flow rates, reactor orificing pattern, fuel element performance, and process limits have been considered. Based on these engineering parameters and related process economics, suggested reactor ``Operating Plans`` are graphically presented in this document. It is emphasized that these plans do not reflect operational considerations which may modify the desirability of the indicated power level increase nor has allowance been made for major projects, major maintenance outages, etc. Many factors which only manufacturing personnel are capable of evaluating may make it desirable to operate above or below these operating plants. These plans are designed to present reasonably achievable but perhaps optimistic power levels together with process limits which will be approached or will possibly limit reactor power levels unless limit revisions can be effected. It should be noted that the engineering parameters and basic assumptions which have been factored into these plans are subject to continual re-evaluation and revision. In a strict sense, these plans are out-dated even as they are published. However, their value will lie primarily in illustrating …

Low flow calibration tests have now been completed at all old reactors as a part of the program to determine the adequacy of the old reactor last ditch water backup system more accurately. In addition, the high tank discharge line strainer gates have been removed at B, D, and F reactors in order to increase the emergency high tank flow rates to the reactors and thereby increase high tank water backup adequacy. The purpose of this document is to list the results of the last two low flow calibration tests at DR and C reactors (calibration test results at B, D, F, and H reactors are listed in reference 7) and to list the increased high tank flow rates that were obtained at B, D, and F reactors when the strainer gates were removed.

At the request of IPD Personnel, calculations on neutron and gamma attenuation were made for the NPR shield. The calculations were made using a new shielding computer code developed for the IBM 7090. The calculations show the thermal neutron flux, total neutron dose rate, and gamma dose rate distribution through the entire shield assembly. The calculations show that the side and top primary shield design is adequate to reduce the radiation level below design tolerances. The radiation leakage through the front shield was higher than the design tolerances. Two alternate biological shield materials were studied for use on the front face. These two materials were iron serpentine concrete mixtures with densities of 245 lb/ft{sup 3} and 265 lb/ft{sup 3} (designated by I-S-245-P and I-S-265-P, respectively). Both of these concretes reduced the radiation below design tolerances. It is recommended that the present front face biological shield be changed from I-S-220-P to I-S-245-P. With this change the NPR shield is adequate according to these calculations. The calculations reported here do not include leakage through penetration in the shield.

The purpose of the production test was to {open_quotes}investigate the reactivity and temperature effects and the associated operating problems resulting from the use of nitrogen instead of carbon dioxide as a constituent of the K Reactor atmosphere.{close_quotes} This report summarizes the reactivity and reactivity-associated graphite temperature effects observed during the first four months of the test which was initiated at KE Reactor on December 5, 1960.

This report provides operational worksheets detailing charge/discharge, maintenance, fuel element ruptures, leaks, tube replacements, and production data.

Two Hanford hot-pressed fuel elements were irradiated in the Materials Test Reactor to an approximate exposure of 1000 MWD/T (0. 11 a/o burnup or 5. 8 {times} 10{sup 19} fissions/CM{sup 3}). One element represented the best bond quality, the other element represented a poor bond due to unoutgassed uranium. After irradiation, radiometallurgical examination revealed extensive bond separation in the unoutgassed fuel element. The Ni{sub 2}Al{sub 3} layer thickened slightly and cracks were observed in this layer throughout both elements. Any cracks observed in the uranium terminated at the nickel layer. The double-worked diffusion closure showed no sign of corrosion, pitting or intergranular attack.

This document from the principal engineer of the manufacturing section at Hanford calls attention to irreconcilable costs for two fuel fabrication studies. Specifically, the major trouble areas appear to be in refinery, green salt and machining steps.

Nickel plated aluminum was considered as a jacketing material for nuclear fuel elements as early as 1954, and both static and dynamic corrosion tests were carried out by Argonne National Laboratories and by Atomic Energy of Canada Ltd., employing demineralized water at temperatures of from 260 to 316{degree}C. Results generally indicated that the nickel had excellent corrosion resistance; however, difficulties were experienced in achieving satisfactory continuity and adhesion of the plate; subsequent work emphasized Ni-Aluminum alloy development. At Hanford, our earliest experience employed Ni plate on aluminum-jacketed fuel elements primarily to minimize mechanical damage to the jacket surface during an irradiation test. The appearance of these fuel elements after discharge suggested that the nickel plate might also be a highly satisfactory coating for corrosion and abrasion resistance. Incentives were manifold, including reducing the incidence of in-reactor fuel element failures and permitting reduction of the aluminum jacket thickness with a concomitant increase in space available for uranium or for cooling water passage. A program has been carried out for the past three years aimed at determining various methods of employing nickel plated aluminum jacket material and testing the capabilities of high quality commercially adequate plate. Almost exclusively chemically deposited plate has …

Available pressure data on the B and D downcomers and approach piping is tabulated herein. All readings were taken with an air bubbler system; readings given in psi were taken on a Bourdon-type pressure-vacuum gauge, readings given in inches of water were read on an oil-filled manometer. In each case, fluctuations occurred and the mean value was estimated by eye. Data are presented as a function of flow rate and temperature.