Graphite reacts with such gases as CO{sub 2}, O{sub 2}, or water vapor to form gaseous oxides of carbon. In the case of CO{sub 2}-graphite interaction, the reaction rate is not significant until about 550 C. Water oxidizes graphite, very roughly, three times faster than CO{sub 2}. Air will oxidize graphite appreciably at temperatures below 500 C. Graphite removal from Hanford reactors is very important, since graphite is used both as a structural support and a moderator for neutrons. Griggs has shown that small graphite samples oxidized to 10 per cent weight loss had only about one-half their original compression strength. Hence, the longevity of the reactors depends to a great extent on maintaining a low graphite oxidation rate. A means of monitoring the extent of graphite loss, i. e., the burnout rate, is necessary to establish future reactor operational standards. Presently, weighed samples of reactor grade graphite are placed along the length of an empty process channel in each reactor. Thus, a sample is exposed to the reactor`s ambient conditions of power level, moderator temperature, and gas composition. This program was initiated in the vicinity of June, 1953 by Woodley. This report presents data on graphite burnout obtained from …

Report documenting the progress of the Army Gas-Cooled Reactor Systems Program to develop a mobile nuclear power plant for military field operation.

A research and development program to investigate the feasibility of eliminating vapor closure for the Pathfinder Reactor was conducted. The major inquiry under the feasibility study involved an analysis of the complex transient conditions occurring in the reactor cavity, the pump rooms, and the entire reactor building following a primary system rupture. To solve the simultaneous nonlinear set of equations evolving from the heat, mass, and force balances in the system, an IBM-704 digital computer program was developed. The program has a very general input and can therefore be used with other containment designs. Input parameters include: initial pressures in reactor and throughout the containment building, hent capacity of vessel, decay heat, feedwater flow rate, enthalpy of feedwater, and volume of primary coolant system. (auth)

A study of sodium fires was performed to obtain detailed information on their characteristics and behavior in order to develop techniques for preventing, containing, and combatting them. lnvestigation was made of the technology of sodium fires, design criteria for improving the fire resistance of equipment and installations using sodium as a coolant, extinguishing materials and procedures for fighting sodium fires, and the evaluation of protective equipment. (auth)

"This is the first quarterly report on Task II of the subject Utilization of Radioactive Isotopes in Coal Process Research. There were two separate projects on Tank I; I. methods Development for Tritium Labeling of Coal Product Hydro-carbons, and II. Applications of Radio-Tracer Techniques to the Study of Fluidized Particle Mechanics. The following extended work on Task II concerns only Methods Development of Radio-Tracing With Tritium."

"With careful attention to details, tritium assays of satisfactory accuracy have been achieved on low activity standard samples by dry combustion and liquid scintillations counting of the product water.This method is faster, more accurate and dependable than the zinc fusion-ion chamber method fro low level tritium assays. At 380 degrees C, tritium loses from hydrocarbon tracers by isotopic exchange is significant and must be corrected for in tracer measurements at this elevated temperature."

Scope. This specification covers ingots of zirconium alloy designated as Zircaloy-4, and intended for the production of reactor components.

Scope. This specification covers Zircaloy-4 wire material for reactor use where high integrity and satisfactory corrosion resistance at elevated temperatures are required.

Scope. This specification covers Zircaloy-4 sheet and strip material for reactor use where high integrity and satisfactory corrosion resistance at elevated temperatures are required.

Absolute (p,n) cross sections have been measured for Ti47, V51, Cr52, Co59, and Cu63 at energies between 4 and 6.5 Mov. These data plus earlier measurements of the cross section for inelastic proton scattering have been used to estimate total proton absorption cross sections for V51 and Co59. An optical model calculation using parameters giving a good fit to elastic scattering measurements predicts an absorption cross section in good agreement with the measurements for Co59. For V51, some sets of parameters gave good agreement with the measured absorption cross section, but the fit to the elastic scattering data was only fair.

Abstract: An element by element reactivity check for SM-1 Core II fuel elements and control rod absorber sections was performed and the burnable nuclear poison loading in the SM-1 Core II stationary fuel elements was established. An approach to criticality of the SM-1 Core II was performed by the inverse multiplication method and the critical rod bank position obtained as a function of fuel loading up to the full SM-1 Core II loading. Maximum and minimum core reactivity measurements were obtained by selective loading of stationary fuel elements and the total "excess K" for the core established. Power distribution measurements in the region of the core-reflector interface and the fuel-absorber interface in the control rod assemblies were performed. The effectiveness of europium flux suppressors in the top of control rod fuel elements and the power peaking in stationary elements adjacent to water gaps in control rod assemblies were measured. Survey measurements established the worth of spiking cold clean SM-1 cores with SM-2 elements, and of water holes in the SM-1 core which might be utilized as flux traps for materials irradiation.

"A finite difference approximation to a non-linear set of parabolic differential equations arising in shallow water theory is given. These difference equations were used to determine the shape and rate of propagation of a hum of fluid down a channel of constant depth. The hump of fluid was found to spread instead of steepen, as is the case in the usual shallow water theory."

"The loss of particles through a cusp of a particular containment geometry utilizing cusped magnetic field lies is considered. A velocity space loss criterion analogous to the loss cone in the mirror machine is derived. The effect of a uniform longitudinal magnetic field perpendicular to the containing field is considered and a loss criterion is derived. The effect of the longitudinal field is to decrease cusp losses.

The fabrication processes applied to nuclear fuels are subject to severe limitations because of the conditions imposed by the reactor environment. The combined problems of neutrons fluxes, high heat fluxes, corrosion by the coolant , and embrittlement by hydriding or similar reactions may be minimized through establishing rigorous materials and fabrication specifications for fuel and cladding.

The result of a leak in a PRTR process tube between the flow monitor and the heated section would be to increase the flow through the monitor, but to decrease the flow through the heated section. The concern for the case of small leaks is whether the increase in flow through the flow monitor is sufficient to cause a high flow tip and a reactor scram for the condition where the flow through the heated section is reduced to the point to cause excessive fuel element temperatures.

"The problems of sequencing jobs through machines are discussed in a linear graph framework. The construction of feasible schedules from given technological orderings is related to the construction of transitive graphs from given component graphs. Methods of constructing transitive graphs are given and bounds on the number of different transitive graphs constructed from given components are determined. A recursive convex function defined on the transitive graphs-the job operation completion time and schedule time-is studied. Bounds on the number of different values that the schedule time can attain is obtained. Examples of multiprogramming, flow shop and machine shop scheduling are studied."

"An algorithm for the generation of feasible schedules and the computation of the completion times of the job operations of feasible schedule is presented. Using this algorithm, the distribution of schedule times over the set of feasible schedule—or a subset of feasible schedules—was determined for technological orderings that could occur in a general machine shop. These distributions are found to be approximately normal. Biasing techniques corresponding to “first come first serve,” random choice of jobs ready at each machine and combinations of these two extremes were used to compute distributions of schedule times."

"As an extension of an earlier paper the Green's function is evaluated for the Lundquist equation linearized about uniform magnetic field, constant matter density, and zero flow velocity. It is assumed that all quantities are functions of two space variables and time only. In the general magnetic field configuration considered here a pure Alfvén disturbance no longer exists; there is instead a wave with properties of both the Alfvén and fast‐slow disturbance."

One of the primary reasons for operating the high pressure KER loops is to obtain information concerning water quality control characteristics for recirculating water cooled reactors. The KER-1 loop is predominantly carbon steel and approximates the water quality conditions specified for the New Production Reactor (NPR).

In answer to the question, "How well do we know the composition of the fuel material for the PRTR start-up tests?", the analytical data on the PRTR fuel elements and other fuel elements which were fabricated by similar processes was gathered and analyzed. The results of this analysis are presented.

VARI solves on the IBM-650 a system of simultaneous, first-order, ordinary, differential equations. The program was written so that a large number of calculations could be done in a reasonable length of time. The program permits the consideration of the production of the isotope by absorption and/or decay of one or more of any of the other isotopes in the chain.

BOUNCE III is a program which was written for the IBM-704 as part of a study of the parameters of the neutron distribution in a large thermal column. The program calculates the eigenvalues and corresponding eigenvectors of the matrix resulting from a diffusion-theory, multigroup description of the thermal neutron spectrum.

This document presents the results of investigations concerned with the conceptual design of a 300 KWe "Capsule" nuclear power plant.

The problem of radiolytic gas diffusion into a bubble of fixed radius is solved. A constant source of radiolytic gas is assumed. The concentration of gas at the bubble surface is related to the pressure within the bubble by Henry's constant. (W. L.H.)