The Fuel Cycle Program is an integrated program of investigation in the Vallecitos Boiling Water Reactor (VBWR) and other facilities to improve the technological limits of boiling water reactors.

The rf particle separator, proposed in 1959 by W.K.H. Panofsky and now in preparation for the Brookhaven Alternating Gradient Synchrotron, required a rf structure which gives a transverse impulse to a passing relativistic particle. In order to produce an accumulative transverse deflection of a traveling charged particle with an electromagnetic field, it is necessary that the field contains a synchronous component and in principle, waveguides and cavities are equivalent with respect to the particle dynamics. It was pointed out by H.G. Hereward, that the electric and magnetic deflection of a transverse electric mode (i.e., with no electric field component parallel to the direction of the particle velocity) cancel exactly at all particle velocities. The deflecting force of a transverse magnetic mode on a synchronous particle with the velocity v is proportional to the factor 1-(v/c)2 and vanishes therefore in the case of relativistic particles.

It has been observed that energetic gaseous ions mark their region of impingement on certain metals in color. Under identical conditions of bombardment, multicharged ions of the same gas mark their impact areas with different colors. The colors are sensitive to the type and energy of the ion as well as to the metal bombarded. These colors may result from reflections from thin films formed by reactions at the target. Because the colors indicate that thicker films are produced with increased ion energy, ion penetration depth as predicted by theory is compared with the depth indicated by the colors observed.

Technical report. From Abstract : "The silver-1,10-phenanthroline system has been studied and the stability constants determined for the 1:1 and 2:1 complexes using the silver electrode. This known system was then used to study the complexes of 1,10-phenanthroline with other metals. The log βn values found for the over-all stoichiometric stability constants are as follows: Ag, 5.02, 12.07; Mn, 3.88, 7.04, 10.11; Cd, 5.93, 10.52, 14.30; Zn, 6.83, 12.05, 16.92. The stoichiometric stability constants determined for copper are log k2 = 6.42 and log k2k3 = 11.05."

This technical report discusses an infrared study of AlH3-2N(CH3)3 while the compound was in its gas phase. This investigation was then extended to include the Raman spectrum observed in liquid trimethylamine.

Abstract: The most adverse power distribution and power peaking factors have been developed for Type 3 replacement cores for SM-1, SM-1A, and PM-2A. Also included is the 37 element Type 3 core for SM-1, which is known as SM-1 Core III and is a prototype of a Type 3 core for PM-2A. The power distributions were compared with measured distributions obtained on type 3 fuel elements in SM-1, SM-1A and PM-2A core configurations. Using the measured data as a standard, correction factors were obtained and applied to the analytical predictions of the most adverse power distribution.

Abstract: Progress is reported on research and development tasks under the Program Plan for Engineering Support and Development of Army Pressurized Water Reactor Power Plants, Contract AT(30-1)-2639, during the six months' period October 1, 1061 to March 31, 1962.

Abstract: "Production of [rho]'s and [omega]'s in [pi]- + p and p + [letter p with a bar over it] interactions is discussed. The [rho] is presumably a J = 1-, 1 = 1 state and thus can decay into 2[pi]'s via a strong coupling. The [omega] is probably J = 1-, 1 = 0 which can also decay into 2[pi]'s via electromagnetic coupling. A theoretical description of this situation is presented in terms of coupled time-dependent Schroedinger equations which mix [rho] and [omega] states that are assumed to be isotopically pure at the time of production. The interference between the two resonances was found to have the effect of extending the 2[pi] spectrum toward the [omega] mass."

This volume is one of twenty-one summarizing the Aircraft Nuclear Propulsion Program of the General Electric Company.

The General Electric Company, as prime contractor to the AEC at Hanford, is proposing to modify the lattice characteristics of the 8 3/8-inch lattice reactors for the purposes of improving the conversion ratio of these reactors. The proposed overbore modification of the reactors would remove the existing aluminum process tubes, enlarge the diameters of the graphite channels by about one-half inch, insert smooth-bore Zircaloy-2 process tubes and refuel the reactor with larger size, self-supported fuel elements. The overbore fuel will remain the internally-and-externally-cooled cylindrical type, but the weight per foot will be about twice that of the present fuel element. The removal of the inlet and outlet piping connections which would be required in the overboring process will permit the replacement of the existing fittings with ones of improved design. Furthermore, new orifices and venturis which are compatible with the hydraulic characteristics of the overbore tube and fuel geometry and the pumping system will be installed. No basic changes are proposed in the pumping system though the reactor flaw rate may be increased 5--10 percent by changes in hydraulic characteristics depending on the water plant flow capacity.

The New Production Reactor, as compared to existing Hanford reactors, employs new and unique design concepts. To properly evaluate these design concepts and their effects on reactor operations, nuclear safety, and reactor life, a comprehensive testing program is planned; this program, with its objectives and restrictions, is discussed in this report. It has been developed along the same line as the C and K Reactors test, programs, and is expected to require a total time of 6--8 weeks of round-the-clock testing. This estimate includes fuel loading time, but does not include time allotments for engineering acceptance tests prior to power operation nor does it include any of the time necessary for engineering and physics tests during the extensive power ascension program. The main body of this report is presented in three parts. The first section describes startup hazards and restrictions, reactor and component safety provisions prior to loading, and the itemized listing of quantities to be measured. The second includes preliminary material and plant conditions and a brief description of the individual tests. The third section (the Appendix), written in procedure language, comprises a rather detailed description of each individual test on a tentative basis; final test details and procedures, …

These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

One of the early candidate fuel elements for N Reactor use was the coextruded, Zircaloy-2-clad seven-rod cluster. As part of the program of evaluating the seven-rod cluster geometry, three-rod cluster fuel elements, two and three feet long, were irradiated. These long cluster fuel elements were irradiated to determine the distortion (or sag) which might occur at the center of the unsupported length during irradiation. Two three-rod clusters made up of 0.630 inch diameter rods, containing natural uranium cores were irradiated in KER Loop 3. The rods of one cluster were three feet long; the rods of the other were two feet long. The three-feet long rods were supported at their ends and at their midlengths, the two-feet long rods only at their ends. During the irradiation, the maximum core temperature was 435 C. The fuel elements were discharged from the loop after they had reached an exposure of 1400 MWD/T. Following the discharge, the fuel elements were visually examined in the KE view pit. No sag was observed in any of the rods. The test demonstrated that two- and three-feet long rod cluster fuel elements can be irradiated without appreciable sag occurring in the rods.

One of the early candidate fuel elements for the N Reactor was the seven-rod cluster fuel element. An objective of the program to determine the suitability of the seven-rod cluster fuel element for N Reactor use was to evaluate the irradiation performance of coextruded, Zircaloy-2-clad, seven-rod cluster fuel elements over a range of exposures from low exposures to high exposures. This report describes the irradiation testing of an enriched seven-rod cluster fuel element which was irradiated to 520 MWD/r.

River temperature traverse data within and just the plant boundary are quite limited. For this report, traverse temperatures were calculated by first averaging the depth measurements vertically at each traverse point, and then weighing the average for the cross section by the depth at each point and the span between points. Calculated {Delta} T`s were determined by dividing the total reactor power level above the traverse line by the river flow, with the appropriate factor to obtain the correct units. No attempt has been made to apply corrections for atmospheric effects or the diurnal cycle.

On September 19, 1961, the No. 3 horizontal control rod at KW reactor was removed after an apparent failure of the wall separating the boron carbide powder from the coolant water. The instrument indications that were interpreted as a rod failure include: high radiation alarms in the exhaust air, unexplained gain in reactivity, increasing coolant outlet temperature on No. 3 HCR, and high radiation readings of the outlet water from No. 3 HCR. The possible reactor safety aspects of such a failure made it necessary to obtain a thorough examination of the rod and inner coolant tube. A complete borescope examination of the rod and partial visual examination of the inner coolant tube have recently been completed. This document is intended to summarize the inspection results, discuss the safety and costs aspects of a horizontal rod failure, and suggest courses of action for the remaining rods at KE and KW reactors.

None

S>The compatibility of type 304L stainless steel in intimate contact with graphite is being studied as a function of temperature and contact pressure. This study is an outgrowth of materials compatibility problems in present and advanced gas-cooled reactors, where structural members in direct contact with graphite provide the possibility of both carburization and self-welding. Initial studies were concerned with surface reactions in the absence of gaseous contaminants under a vacuum of 10/sup -6/ mm Hg at 540 to 705 deg C. Stainless steel specimens are pretreated to provide three surface conditions: H/sub 2/- fired, preoxidized, and Cu-plated. Surface contact pressures ranged from 0 to 10,000 psi. Test results are presented which establish the lower temperature limit for significant diffusion between graphite and stainless steel at approximately 60O deg C. Above this temperature, diffusion between untreated or H2-fired stainless steel surfaces was found to effect complete bonding of the two materials at contact pressures as low as 500 psi. Bonding was effectively prevented by the presence of either an oxide film or a Cu plate at temperatures up to 700 deg C. Where bonding occurred, diffusion rates measured for C in stainless steel were comparable with those reported for stainless steel …

The specification contains information concerning the material, process, and product requirements to be met by the fabricator. (J.R.D.)

Run 25 was the final period of power operational of the HRT. The reactor was operated for periods of 62, 8, 52, and 80 hours at 5 Mw with no outward indication of fuel and core and blanket average temperatures of 270 and 230 deg C, respectively. The uranium concentration in the was 1.7 to 2.0 g U/kg D/sub 2/O. Longer periods of operation were prevented by mechanical difficulties, notably with the fuel feed pump. While the reactor was subcritical after the last of the above runs, the upper patch in the core tank wall became dislodged, allowing greater core-to-blanket mixing. The resultant blanket uranium concentration was 2.9 g U/kg D/sub 2/O. The reactor was subsequently operated at April 28, 1961. The experiment was operated at high temperature for a total of 10,866 hours. The system was critical for a total of 8,841 hours and produced 16,295 Mwhours of power. The fuel, heavy water, and some corrosion specimens were recovered, and the reactor was stored in an assembled state. (auth)

A seven- to twenty-fold volume reduction can be obtained from fluidized bed calcination of aqueous aluminum nitrate wastes, depending on the operating conditions employed and their effect on the intra-panticle porosity and absolute density of the calcined alumina. Among the calcining variables, only the bed temperature and the fuel aluminum concentration had a significant effect on the intra-particle porosity of alumina generated during studies conducted primarily in a two-foot-square fluidized bed calciner. A quantitative correlation of the effect of these variables is presented. Alumina with an intra-particle porosity as low as five per cent can be generated by employing a suitable combination of low bed temperature and dilute aluminum feed concentration. Feed sodium concentration and product alpha alumina content were found to have minor effect on intra-particle porosity. Results also show that an inverse relationship exists between the nitrate content of the calcine and the calcination temperature. (auth)

An addition is presented for the IBM-7090 program MURGATROYD to produce a rough graph of reactor power versus time. A sample of output is included for the case given as an example. (J.R.D.)

The effect of modifying the dislocation structure by room-temperature prestraining on the subsequent yielding behavior at 77 deg K is being studied. Results on tantalum single crystals indicated that a considerable decrease in the early yield stress for a given strain is effected by prestraining at room temperature. Twinning was observed in the tantalum at 77 deg K and normal rates of strain. It is shown that the potential barrler to dislocation motion in crystalline solids can be measured in relatively pure bcc transition metals such as Nb, Ta, Mo, and W. Work is being carried out to extend the method of differential calorimetry to measurements of stored-energy-release spectra in deformed bcc metals. Measurements for copper are discussed. (M.C.G.)

Tests were carried out on the self-actuated reactor relief valves and the self-actuated pressurizer steam relief valves in order to insure their reliable operation. The valves operated properly, and the final set pressures were all within the specified limits. (D.L.C.)