The designer's concept of a test program for the 30-Mw prototype intermediate heat exchanger and steam generator designed and fabricated as part of the Sodium Components Development Program is presented. The performance data will serve to verify the thermal design, or allow application of improved techniques to future designs, give an improved basis for stress analysis in design of future units, and demonstrate the capability and limitations of the units in relation to the performance specifications for which they were designed. Welding techniques for type 316 stainless steel are described. The specifications and operating conditions of the units are given along with instrumentation drawings showing test equipment design and arrangement.

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The status of aqueous processing methods for thorium fuels is summarized, with principal emphasis on the stainless steel-clad ThO/sub 2/UO/sub 2/ type. Data were obtained principally from laboratory-scale experiments with fully irradiated fuel samples and engineering-scale tests with unirradiated fuel. Stainless steel cladding was easily dissolved with 4 to 6M H/sub 2/SO/sub 4/ (Sulfex process) or 5M HNO/sub 3/-2M HCl (Darex process) in LCNA (Nionel type) or titanium equipment, respectively, in semicontinuous or batch equipment. Uranium losses to the decladding solutions were approximates 0.3% and 3 to 5% for the Sulfex and Darex processes, respectively, with fuel irradiated to approximates 20,000 Mwd/ton of core. The uranium was readily recovered from the Darex decladding solution in the acid Thorex extraction process. The ThO/sub 2/UO/sub 2/ core was dissolved in 13M HNO/sub 3/ -0.04M NaF-0.1M Al(NO/sub 3/)sub 3/. Uranium and thorium can be recovered from graphite-base fuels by disintegration and leaching with 90% HNO/sub 3/, grinding and leaching with 70% HNO3, or combustion followed by dissolution in fluoridecatlyzed nitric acid. Uranium and thorium were recovered from nitric acid solutions and separated from fission products by extraction with 30% tributyl phosphate in Amsco in the acid Thorex process. The use of an acid …

Safe and economic methods of handling radioactive materials off-site are required for the successful operation of nuclear chemical plants. These occasions arise in the shipment of spent fuel, radioactive, isotopes, and liquid wastes. An unsolved problem exists in the development of techniques and sites for the final disposal of waste products. (auth)

During the course of the Christmas Island portion of Operation Dominic, a fairly simple, inexpensive method of making a rapid fireball yield determination was developed. In order to check the zero time alignment of the skysweeper antiaircraft mount on the burst point at zero time as provided by the Sandia 584 radar system, a pinhole camera camera was installed on the mount. This was helpful in determining whether or not the photographic and photoelectric diagnostic systems were observing the appropriate volume of air excited by prompt gamma and neutron radiation. After viewing the initial circular image on the Nambe event, it was decided that an attempt at a prompt fireball yield number would be made. Ad hoc scaling laws were used for the duration of the operation and the instant fireball yield number was included in several TWXs of early shot data. This report provides calculations, graphs, and illustrations which show that with the data now in hand, there is a simple method of determining a yield number to generally better than 10 percent. This number can be available within minutes after the detonation.

Normal Pile exposure variations and day-to-day operating reactivity adjustments must be compensated by a combination of in-reactor poison and enrichment. The calculation of current reactivity status and the prediction of future trends is an integral part of the Physicist`s responsibility. This document describes the computer program ``CALHOT,`` which is based upon one-group diffusion theory, and is intended to be used in routine reactivity-enrichment calculations. It offers the advantages over present hand calculations of greater accuracy, more flexibility, and, most important, much less calculational time.

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The interests of several departments at Hanford are involved in the planning, execution and evaluation of the results of the KER loop testing effort in support of the NPR fuel program. The varied interests and activities of the participating groups must be well-integrated if effective use of our limited testing capability is to be made. The purpose of this report is to help achieve this integration by summarizing the current thinking on the goals of the NPR fuel testing program and by presenting the current loop schedule.

ABS>Hot cell demonstration of the Zirflex decladding process coupled with a modified Purex solvent extraction process was completed using specimens of Zircaloy-clad UO/sub 2/ irradiated to levels of 6150-14,600 Mwd/TU. Soluble losses of uranium and plutonium to the decladding solutions were about 0.05%. Centrifugation of the decladding solution is probably necessary to remove up to 1% of the UO/sub 2/ present as fines resulting from the fracture of low (93 to 95%) density pellets; high (96%) density pellets produced few fines. Approximately 5 hours were required to dissolve the UO/sub 2/ core material (14,000 Mwd/TU) in 4M HNO/sub 3/ versus 6 to 7 hours for unirradiated pellets to produce a solvent extraction feed of 100 g U/l and 3M HNO/sub 3/. Gamma decontamination factors for uranium in the Purex CU stream and plutonium in the BP stream were increased by factors of 2 to 10 from the normal 1.3 x 10/sup 3/ and 2.1 x 10/sup 3/, respectively, by pretreatment of the solvent extraction feed with dincetyl monoxime or its degradation product, oxalic acid. Preliminary data indicate radiation damage degrades the solvent, 30% TBP diluted with Amsco 125- 82, upon one pass through the mixer-settler banks with feed solutions irradiated …

This document describes the objectives, targets and scope of the development programs to be sponsored and conducted by the Irradiation Processing Department in support of the Hanford production reactors during FY-1963 - 1964. it is anticipated that the basic funding level for the IPD Reactor and Metallurgy R&D programs for the next two years will total $6,930,000 annually including overhead. Approximately 70 per cent of these funds will be directed in support of the N Reactor, and the remainder in support of the older reactors. The funding levels for the various reactor and Metallurgy programs requiring technical support are reflected in HW-72790, {open_quotes}Budget Proposals for O2 Program Research and Development, FY-1963 and FY-1964, Irradiation Processing Department{close_quotes}, O. H. Greager, April 23, 1962. This latter document also contains Supplementary Budget Proposals to the Commission for work associated with the Plant Improvement Program for the older reactors and alternate production concedes for the N Reactor, which require funds in addition to the basic program. For current planning purposes, however, the technical efforts described within this Program Document have been restricted to those within the basic funding levels, and if Supplementary Programs are authorized, adjustments will be made accordingly.

By equating net radiation to space to I/sup 2/R heating in the SNAP-2 instrument compartments, and neglecting internal radiation between compartments, the panel surface equilibrium temperatures were computed for hot and cold temperature extremes. These extremes are defined by: (1) cold - pre-start phase in the shade, (2) hot - full power phase facing the sun. The results of the heat balances for hottest and coldest cases are presented graphically. These figures relate the equilibrium surface temperatures to the I/sup 2/R heat load dissipated by the panel surface for hot and cold orbits before and after startup. Included as parameters are effective panel area (dependent on Agena-interface design details) and the ..cap alpha../epsilon ratio for the surface coating. If ..cap alpha../epsilon = .3/.9 and A = 1 ft/sup 2/, the extremes of temperatures suffered are from -191/sup 0/F to +212/sup 0/F. This study shows that the normal R/C surface coating (..cap alpha../epsilon = .3/.9) is acceptable regarding allowable maximum surface temperatures, if the effective panel area is not less than 1 ft/sup 2/. It also indicates that further work is warranted regarding internal radiative distribution of heat in order to limit the lowest temperatures to -65/sup 0/F.

This document details activities of the Irradiation Processing Department during the month of August, 1962. 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; and NPR Project.

This document details activities of the Irradiation Processing Department during the month of November, 1962. 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 is the monthly report for the Hanford Laboratories Operation, November 1962. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

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The objective of this production test is to evaluate the effect of Zr-2 jacket uniformity, thickness, and temperature on non-uniform clad straining of coextruded Zr-2 jacketed rod samples.

From abstract: "This report describes work done on ARF Project C 216 from January 5, 1962 to September 15, 1962. The objectives of the program include evolution of an analytical research program to evaluate effects of environment contributed contamination on precise devices"

This technical report provides a design criteria for a technical support facility for the Lithium Cooled Reactor Experiment (LCRE) and SNAP-50-DR-1 Test Facilities. The support facility is adjacent to the LCRE Test Facility and is located completely within the existing Building 629 structure at the former ANP area of the National Reactor Testing Station (NRTS) near Idaho Falls, Idaho. The information and specifications presented establish the basis for the design of laboratories, shops and engineering areas required to support the installation, operation, maintenance and disassembly of the LCRE and SNAP-50 tests. The construction and modification required to adapt the building to reactor test support operations are described in detail in the following report.

This paper presents thrust calculations for low altitude trajectories for a Tory II-C type propulsion reactor.

During the normal operation of reactors ionic, particulate and colloidal substances in the coolant are activated as they pass through the neutron flux. The parent substances for the activated materials come from several sources: transient corrosion products from the coolant system piping and fuel element cladding, coolant additives and coolant impurities. Previous studies have shown that a major portion of these substances do not simply pass through the neutron flux field but undergo complex sorption phenomena that result in effluent radio-nuclide levels considerably in excess of what would be expected from activation on a once-through basis. This report is being written to outline studies that will provide activity inventory data that will aid in defining radionuclide retention reactions in reactor coolant systems.

"In Householder, using Weissinger's identity , obtained the necessary and sufficient conditions for the c onvergence of the Seidel method for the solution of finite matrix equations of the form (D-S-S*-F)u = f. The same coniditions were also obtained by Krein and Prozorovakaya for an analog of the Seidel method for the operator equations of the form (D+S+S*)u = f in a Hilbert space. The purpose of this article is to extend the result of the above authors to the generalized overrelaxation iterative method for the solution of a wider class of of operator equations in a Hilbert space."

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The stress analysis performed on the PM-2A reactor vessel and cover is discussed. The maximum combined stress (51,000 psi) occurred in the studs after reaching steady-state conditions. A fatigue analysis indicated that this stress could be safely applied 2500 times, and since the studs do not approach 2500 cycles from initial stud tightening to steady-state conditions, they should not suffer any fatigue damage. (auth)