Abstract: This technical report contains the description and results of an experimental program to evaluate the effect of utilizing Type 3 (SM-2) replacement cores in existing Army field plants SM-1, SM-1A and PM-2A. This program, conducted at the Alco Products Critical Facility, employed SM-2 mockup fuel elements similar in composition to Type 3 fuel elements to determine start-up characteristics of Type 3 cores in SM-1, SM-1A and PM-2A core configurations. measurements include comprehensive power distribution, temperature coefficients, initial critical bank positions, control rod calibrations, critical rod configuration and material coefficients, all obtained under cold, clean, core conditions. The 45 element SM-1 and SM-1A configuration with SM-2 mockup fuel elements contain 36.4 Kg U-235 and an estimated 67.9 gm B-10, while the 37 element PM-2A configuration with SM-2 mockup elements contains 30.0 Kb U-235 and an estimated 56 gm B-10.

This technical report describes the development, design, operation, and performance of a full-circle, radial-vane steam separator for the boiling water section of a nuclear superheat reactor. Steam-water tests of this model have demonstrated that is has vane capacity in excess of that required for the 300-Mx(e) separate superheat reactor and for the 300-Mw mixed spectrum superheat reactor. It is proposed that the vane capacity requirement of the 600 Mw(e) separate superheat reactor may be attained by increasing the nozzle length.

From introduction: "This is the eighteenth in a series of quarterly reports which cover the progress and results from the conceptual designs, economic evaluations and research and development work performed by the General Electric Company as part of Contract AT(01-3)-189, Project Agreement No. 13."

From introduction: "This is the seventeenth in a series of quarterly reports which cover the progress and results from the conceptual designs, economic evaluations and research and development work performed by the General Electric Company as part of Contract AT(01-3)-189, Project Agreement No. 13."

From introduction: "This is the sixteenth in a series of quarterly reports which cover the progress and results from the conceptual designs, economic evaluations and research and development work performed by the General Electric Company as part of Contract AT(01-3)-189, Project Agreement No. 13."

In regard to the production of ozone and various nitrogen oxides in air by action of x-rays or gamma rays, thre seems to be no direct experimental data which can be used. There is some data on the action of alpha particles on air and some on the action of cathode rays.

Summary: The Mixed Spectrum Superheater (MSSR) is an integral superheater reactor in which boiling occurs in an annular Boiling Water Reactor section and steam in superheated in an unmoderated fast section in the center. A Mixed Spectrum Critical Assembly (MSCA) to be operated at the Vallecitos Atomic Laboratory has been designed to mock up a 75-150 MWe prototype MSSR. The principal experimental measurements aimed at proving the feasibility of the MSSR concept include power distribution, Doppler effect, flooding effects, distribution of reactivity, control rod worths, and the effect of the control system on the power distribution.

Abstract: This technical report summarizes the results of research and development tasks performed during Phase I of this contract and presents recommendations for future development work. Work is reported in the areas of plant assembly and relocation, housings and footings, waste heat dissipation, instrumentation, refueling systems, waste disposal, shielding, core nuclear thermal and hydraulic studies, gaseous waste processing (BWR), and critical experiments on a 5 x 5 array of Type 3 fuel elements.

Thirty gram patches of plutonium dioxide can be readily chlorinated by reaction with phosgene or carbon tetrachloride at temperatures of 350 and 450 C respectively. Plutonium trichloride prepared by either method can be reduced to the metal in a hermetically sealed bomb by reaction with calcium. On a twenty gram scale yields of approximately 97 per cent are obtained when a calcium-iodine booster is employed. It has been demonstrated that a method for reduction of plutonium trichloride to the metal without the use of a booster can be development.

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.

Technical report of a proposal of an analytical model to predict two-phase critical flow rate. The model is based upon thermal equilibrium, a "lumped" treatment of the two-phase velocity (each phase is represented by a single mean velocity), and upon the neglect of frictional and hydrostatic pressure losses. A comparison, of the proposed predictions with available test results and previous analyses shows that: (1) The present model agrees very well with the published test data. (2) In contrast to all other analyses, the model requires no assumption about the gas void fraction.

Abstract: This preliminary technical report covers design of the SM Type 3 element, its use .in.. SM-1 Core III, and planned use in the Type 3 Core for PM-2A. The Type:3 element characteristics are compared with Type 1 (SM-1 Core I) element nuclear, metallurgical, thermal and hydraulic characteristics and fabrication. The effect of using the Type 3 element in SM-1 Core III and. its planned use in a Type 3 Core for PM-2A is discussed with regard to operation, shielding, reactor safety and all conceivable special problems.

Summary: An analytical study of the stability of the Big Rock Nuclear Reactor has been performed for the plant as built, and supplements a previous design stability study. The plant has been determined by this analysis to be very stable under every mode of operation anticipated during Phase I of the development program testing. Even under conservative assumptions of system parameters the minimum calculated gain and phase margins do not go below 13.0 db and 46 degrees, respectively. (Nor are these both reached simultaneously for the same operating condition.) These are characteristics of a very stable, well-behaved system. In addition to this analysis, a second, less conservative series of computations was performed to provide expected realistic closed loop data for comparison with Phase I test results. The most responsive test thus predicted occurs at 60 percent power, 1500 psia, minimum flow, and maximum subcooling. For this case the closed loop peak response of power to reactivity occurs at a frequency of 0.90 cycles per second with an amplitude of 3.90 db. This corresponds to an expected open loop gain margin of 16.5 db and a phase margin of 63 degrees. Although knowledge of reactor transfer function is to be determined …

A buncher and pre-accelerator for A-12 is described that has 360 degree acceptance and could handle a current of one ampere from a low voltage source.

Abstract: The pressurized water reactor preliminary design presented in this volume is the preferred design developed under Phase I of the PL-3 contact. This technical report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and a training program outline.

This is the sixteenth quarterly report of the research and development activities of the Radiological Sciences Department, Hanford Atomic Products Operation, as before, includes some items charged to control but included for general interest. Such are identified as "not charged to research."

Abstract: A core design study of a 10 Mwe fast ceramic reactor is presented. Local reactivity coefficients, safety criteria, accident analyses, and economics are considered. An attempt is made to find a new balance of characteristics by purely geometric devices, i.e., by exploring the sodium : fuel ratio and varying the height : diameter ratio of the core. The use of BeO in the core was also investigated.

The first of two general methods are discussed based on the removal of the thin layer of aluminum to which graphite adheres. Two electro-polishing techniques, an electrolytic etch, an anodization-deanodication cycle and two chemical etches are described.

This report contains the portion of the Substitute Materials Program dealing with research and development on the production and application of actinium-227.

Abstract: Residual and fission gas release from UO2 were studied in the laboratory and in in-reactor experiments. Arc-fused powder and sintered pellets were used to determine the rate of evolution and types of residual gases as a function of temperature. Fission gas release was related to the average UO2 temperature and fission gas release calculations were made using the latest thermal conductivity values, isotopic half lives, and branching ratios available in the literature. The results obtained were compared with those available in the literature, and a satisfactory agreement was found among the groups of comparable data.

This report describes the work carried out in 321 Building semiworks equipment, to define the factors contributing to high corrosion-product contamination and presents recommendations for reducing the impurity level to meet current specifications (maximum of 200 parts total metals per million parts U).

Abstract: Neutron flux and gamma radiation measurements through the SM-1 primary shield were made at the startup of Core II in June 1961. They extend previous measurements (APAE-35) both vertically and horizontally in the primary shield and in the rod drive pit. Dose rate measurements on spent fuel elements under water are also reported.

Abstract: This technical report covers development of an AG-Cd-In alternate absorber section for Army Type SM reactors. It describes the absorber material composition and the geometric configuration. It gives the nuclear and thermal analyses supporting this configuration and a detailed description of the manufacturing practice employed in fabricating the final design component.

Abstract: This technical report is concerned with the program and results of the SM-1 reactor vessel head removal and core inspection at Fort Belvoir, Virginia from the period March 7, 1969 to May 17, 1959. It covers the operating procedures in detail and records the conditions found and problems encountered in order to make a record for reference in future work of this nature. The major objective of the program, to obtain irradiation data on the SM-1 type core, has been met. The boron control rod elements were found unsatisfactory for full core life, and europium oxide elements were placed in the core for future irradiation stability data. A major problem was experienced with the cracked pressure vessel head studs. The methods developed for the removal of the broken studs are presented. The complete metallurgical study of the stud failure from stress corrosion is included as Appendix B.