A report is presented to provide a tool for evaluating the relative economic incentives for changing reactor core parameters. The cost relations are shown in terms of differential cost in lieu of total cost. A total cost for each reactor described is included so that power costs for a specified set of parameters can be obtained. A description is also included concerning 5 reactor types considered along with a discussion of the effects on power costs of varying the significant core parameters. A listing of basic references is given. (J.R.D.)

Possible mechanisms for creep rupture during irradiation are examined. Evidence that the rupture occurs by grain boundary sliding alone, or by vacancy condensation, is compared. It is observed that vacancy condensation is the more probable mechanism, and that this mechanism predicts a reduction in creep rupture life for metals exposed to a fast neutron flux (neglecting effects of radiation annealing). (T.F.H.)

It is deemed feasible to store reactor fuel wastes in a salt dome cavity to a depth where the differential in pressure between the soil over-burden pressure and pressure of the fluid inside the cavity does not exceed 3000 psi, and the temperature is less than 400 deg F. Tests at pressure increments of 1000 psi were conducted on a 2" cylindrical cavity contained in a 6-in. long by 6-in. cylindrical salt core. Tests indicate that the cavity exhibited complete stability under pressures to 3000 psi and temperatures to 300 deg F. At temperatures of 100 to 400 deg F and pressures to 5000 psi continuous deformation of the cavity resulted. Initial movement of the salt was observed at all pressures. This was evidenced by vertical deformation and cavity size reduction. It was noted that a point of structural equilibrium was reached at lower temperatures when the pressure did not exceed 5000 psi. A literature study reveals that the most common type of cavity utilized in liquefied petroleum gas storage is either cylindrical or ellipsoidal. A few are pear or inverted cone shaped. There was no indication of leakage for cavities when pressure tested for as long as 72 hr. …

This document contains 3 reports dealing with the delayed effects of radiation on the eyes of survivors of the atomic explosions in Hiroshima and Nagasaki. In the first study, 1000 persons who were listed as having been in the open and within two kilometers of the hypocenter at the time of the explosion were selected at random from the census files of the Atomic Bomb Casualty Commission for study. In addition, 231 others, comprising the total available number of surviving persons listed at present in the census files as having been within one kilometer of the hypocenter, were examined, as were several hundred others who were contacted through newspaper publicity, referrals from local ophthalmologists, or through hearsay. The survey resulted in bringing in persons having, or having had, a variety of ocular conditions. Those connected with the atomic bomb included the following diagnoses; multiple injuries of eyes and eyelids; keratoconjunctivitis from ultraviolet and ionizing radiations; thermal burn of the cornea and of the retina; retinitis proliferans; and radiation cataracts. The cataracts were the only delayed manifestations of ocular injury from the atomic bomb. The second paper is a case report of a histopathologic study of atomic bomb radiation cataract. The …

This quarterly progress report discusses the ongoing work within the Solid State Division at the Oak Ridge National Laboratory for the period ending August 10, 1952. Projects discussed include radiation metallurgy, engineering properties, fused salts, crystal physics, and solid state reactions.

This document is a HAPO report dated January 23, 1959, and describes the plutonium tail-end anion exchange system, installed in Purex as a prototype unit. Although some modifications, those considered most needed, were made to the unit, additional changes and refinements were still needed to convert the prototype to a fully acceptable manufacturing facility. This document covers the scope design of these modifications. The purpose of this document is to provide scope design criteria for a project to convert the plutonium tail-end Purex Anion Exchange Prototype to a manufacturing facility.

The purpose of this report is to present the conclusions and recommendations obtained during the course of the CPD Control Study and the conceptual framework upon which they are based. Primary emphasis has been given to the control of product materials. In order to logically present the background for the definition and delineation of an appropriate CPD Control System, Section III of this report discusses a control system in the following manner; (1) the description of the control system information requirements, (2) the definition of the conceptual framework of product material control, (3) the discussion of the interrelationships of production scheduling, process control and accountability and (4) the methods for the effective utilization of control system information. Section IV utilizes this conceptual framework in order to enable a logical presentation of a proposed product material control system for the CPD. A summary of conclusions and recommendations is included in Section II. The Appendices consist of discussions of specific analysis conducted during the study. Other related reports that have been issued during the course of the study are included in the references.

This report discusses a proposal to measure the corrosion rate of aluminum-clad DOE elements in deionized water in one of the KER Loops at the highest obtainable surface temperatures under conditions where the bulk water adjacent to the elements is saturated with aluminum corrosion product. This will determine whether in-reactor corrosion may be reduced by controlling the aluminum concentration in the water and also provide corrosion data in deionized water at higher temperatures than is now available. A thermocouple slug will be included.

This document details activities of the irradiation processing department during the month of December 1958. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering operation; Employee Relations Operation; and Financial Operation.

This document details activities of the Fuels Preparation Department during the month of December 1958. (FI)

A solid spline made of 430 boron stainless steel was tested in tube 2451 at D Reactor. The test was conducted to compare the boron carbide filled hollow aluminum spline presently used with a solid steel spline which is being considered as a substitute. Usage of hollow splines has shown that a solid rugged spline would be highly desirable. Comparisons of spline radiation and reactivity effect were obtained both in-pile and in out-of-pile test facilities. Results of the production test are reported in this document.

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The purpose of this test was (1) to determine the conversion ratio of alternated enriched uranium fuel segments and lithium aluminum target slugs, and (2) to determine the stability of solid and cored enriched uranium with this type of load. The test consisted of an irradiation of two twelve-tube pile charges, each of them forming a sixteen tube square without corners. The first twelve tubes were a ``striped`` load containing solid fuel elements enriched to .95% U-235 alternated with lithium-aluminum target slugs. The fuel elements of the second array were 1/2 inch cored, but canned with the core plugged to give the same O.D. and external appearance as a solid element. To determine fuel element stability, the irradiation was essentially a run-to-rupture test. The central four-tubes of each square were designated for special extraction after the irradiation to determine conversion ratio. Neither solid nor cored enriched uranium of metal quality comparable to that used in this test has adequate stability to reach a routine goal exposure of more than 500 MWD/T at the specific powers of the test, 70--80 KW/foot. From post-irradiation measurements of diameter growth of the cored slugs, it might be concluded that many of the slugs were …

The objective of this production test is to determine the changes that occur in the uranium-Al-Si bond during irradiation at bond temperatures between 255 and 285 C. Twenty-five M-388 jacketed dip canned depleted uranium solid fuel elements will be irradiated to an exposure of 500 MWD/T in high temperature water. The location and size of unbonded areas on the fuel elements will be measured by ultrasonic mapping before and after irradiation to show the changes in bonding resulting from irradiation at high temperature.

The purpose of this report is to provide scope and justification for the installation of two new 16,500 gpm, 150-foot head pumps in the 181-C building in order to furnish sufficient additional water to increase production at 105-B and to supply the entire normal export demand from B Area. A minimum of 11,000 gpm will be required to be transferred from the 183-C reservoir to the 183-B clearwell via the cross-tie line during the adverse turbidity period when the I&E program is in effect at 105-B. It is contemplated to transfer 15,000 to 20,000 gpm so as to provide the necessary water for this increased flow at 105-B and to reduce the flow through the existing 183-B filters. It is apparent that the pumping capacity at 181-C is not adequate to supply this extra water at 183-C unless all export pumps are used to pump water to 183-C. If this increased demand were met by using the export pumps, the 100-D Area would have to supply most of the export demand to the 200 Areas which is anticipated to be 12,000 to 15,500 gpm. After the new expected I&E flows are achieved at 105-D and 105-DR, D Area will have …

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The following report describes the investigations of the microwave solution and the Astron solution, to produce high current electron beams.

We have previously transmitted interim recommendations of goal exposure plans for enriched I&E material at the C and K reactors and for natural I&E material at the K Reactors. We indicated in the references that corrections for changes in inlet temperature would be forthcoming. The purpose of this letter is to transmit revised goal exposure equations.

Seven enriched seven-rod cluster elements, three with thirty mil Zircaloy-2 jackets and four with twenty mil Zircaloy-2 jackets, will be irradiated at high temperature to an exposure of 4500 MWD/T. Coolant temperatures in the internal and external flow channels will be measured during irradiation.

The five year forecast for operation of the HAPO reactors calls for the achievement of increased process water flows in B, C, D, DR, F and H reactors. The Process Design Operation has initiated a study in support of this forecast whose objectives are: to determine present water plant and effluent system flow capabilities; to provide basic data for determining the ultimate economic optimum flow capability of these plants; and-to provide a basis for scope and development work preliminary to the initiation of any required project action. The present I&E slug program has pointed up the need for such a study of increased flows in order to take advantage of the lower system resistance of the I&E, slugs. Initial studies have indicated that considerable development work and testing is required in order to determine the most economical method of achieving increased process water flows. For this reason, CGI-815 ``Increased Water Capacity, 100-B, C, D, M, F and H`` was initiated. This interim report presents the information on the first goal of the study, namely the present capabilities of the existing water plant systems and equipment. The reactor study program has been reported separately in HW-57737. Conditions which may be encountered …

This document details activities of the Hanford Operations Office during the month of January 1958. (FI)

Neptunium-237 is currently produced in the Hanford reactors at a rate of approximately .003 gms/MWD via the following reactions: (a) U{sup 238} (n,2n) U{sup 237}{sup {beta}}{yields} Np{sup 237}. (b) U{sup 235} (n,{gamma}) U{sup 236} (n,{gamma}) U{sup 237}{sup {beta}}{yields} Np{sup 237}. In order to calculate the buildup of Np{sup 237} via reaction (a), which accounts for the greater share of the formation of Np{sup 237}, the n,2n cross section for U{sup 238} must be known. An old value quoted by Arnold of 5.2 millifermis for an {open_quotes}effective{close_quotes} 2200 m/s value is not large enough to account for the observed Np{sup 237} yield by about a factor of two. Recent n,2n cross section measurements for U{sup 238} permit a newer calculation and the result is 11.2 mF, effective 2200 m/s value.

Selection of new graphites for nuclear reactor moderator applications can be accomplished either by: (1) an evaluation of the behavior of untested materials processed by proven methods of manufacture when tested under conditions typical of the proposed application, or, (2) a development program exploring new materials, processes and environments which results in graphites meeting or exceeding design requirements. The latter approach is not easily adapted to construction schedules and therefore can at this date contribute little to the 105-N program. The program for selecting 105-N reactor graphite is governed basically by two factors: dimensional stability of graphite under the proposed operating conditions is the major design requirement to be met and the types of petroleum coke and processes used in manufacturing artificial graphites directly influence their behavior under irradiation. Since the evaluation and procurement of graphite for the 100-K reactors, carbon companies have established a number of new coke sources. Some major changes in processing graphite have also been developed; however, these have not reached full production scale and cannot be considered sources for 105-N graphite. Consequently, candidate graphites for the 105-N reactor are those representing the new coke sources plus Texas Lockport coke, the only previously evaluated coke now …

From introduction: The principal work described below is the development and tests of a slow neutron flux generator based on a generator developed at the University of California (UCRL 8359, W. Patterson, Roger Wallace, "A Method of Calibrating Slow Neutron Detectors"). It is shown that a tolerance flux density of slow neutrons results within a cubical cavity 15 inches on an edge with 4-inch-thick walls of water or paraffin wax when a plutonium-beryllium source emitting about 7 million neutrons per second is placed within the cavity according to a standard procedure. The principle on which the generator is based is that the fast neutrons from the source are slowed to near thermal velocities by scattering from the cavity walls. (Although the principal interest is in the slow flux, a primary fast flux is present at any point depending on the inverse square of the distance from the source.) The variation of slow neutron flux density over the walls of the cavity is unimportant for the calibration of thermal neutron detectors which follow a 1/v or 1/E response law (where v and E are the neutron velocity and energy), so that the detector is sensitive mainly to the slow neutron flux. …