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This document details activities of the Fuels Preparation Department during the month of October 1961. (FI)

This document details activities of the irradiation processing department during the month of October, 1961. 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.

A high expense item in the operation of the button line has been the RS-6 crucible which can be used only once. The 234-5 Research and Development Laboratory has done some preliminary development on a reusable crucible. Before continuing their program, they have requested that an economic study be made to determine what potential savings could be realized if RS-6 were replaced with a reusable crucible. For the purpose of this study, the costs involved were broken down into three areas: Button line costs, Recuplex costs, and Conversion costs. Button line costs include initial crucible costs, operational handling and radiation factors, and other material savings. Recuplex costs cover the effect the elimination of crucible fragments would have on the recovery operation. Conversion cost is an estimate of the expense required to convert the present button line to use a reusable crucible. The results of the study are very favorable and the total possible savings appear to be high enough to write off the cost of conversion of the button line to use the reusable crucible in one year. To write the conversion off in one year, the purchase cost per firing of the crucible must be equal to or less …

The object of the production test described in this report is to evaluate the operational aspects of a proposed method for reducing minimum downtime. The excess xenon poisoning, which occurs during the first 32--38 hours after the shutdown of a reactor from present equilibrium levels, will be partially overridden by a central enriched zone whose added reactivity contribution would be compensated during normal operation by means of poison splines.

The objective of this production test is to compare the irradiation behavior of N Reactor inner fuel tubes (NIE1 elements) having negligible warp after beta heat treatment to the behavior of NIE1 elements with extensive warp after beta heat treatment which were straightened and beta heat treated a second time. The information from this test will be used in the development of a program to evaluate the irradiation behavior of NIE1 elements in various conditions of pre-irradiation warp and straightening.

Accumulation of data on rear face piping strain, piping vibration, water pressures; and water temperatures at KE Reactor was recently completed. A preliminary review of this data indicates that much of the pressure and temperature data could be significant to the study regarding the capability of the K Reactor rear face piping to accommodate the proposed increased water flow. In view of this, a copy of the pressure-temperature data is attached. The purpose of the test was two-fold: (1) to establish mechanical and hydraulic conditions in the rear face piping accompanying operation at current flows and maximum bulk outlet temperatures; (2) to use this information to assess the feasibility of continued operation at these conditions. A preliminary review of the data brings forth the following observations: (1) a downcomer flow transition begins to occur at about the 85 C bulk outlet temperature; (2) the current flow conditions at the top of the downcomer appear to be slightly different from the original design concept; (3) the pressure recorded on the near side of the downcomer just above the orifice (P-10) appears doubtful, and instrumentation will be checked; (4) flow characteristics of the multiple parallel orifices in a single channel does not …

Tests in which aluminum-jacketed, Al-Si bonded uranium fuel elements were baked at various temperatures have shown there is a time-temperature relationship for Al-Si layer decomposition. For heat transfer and secondary coolant barrier considerations, the extent of bonding layer deterioration during fuel element irradiation is important. Currently, Al-Si bonded fuel elements show evidence of spire bond separation, and to a lesser degree, can-bond separation following irradiation. Such evidence has aroused concern for the ability of the currently produced Al-Si bonded fuel elements to withstand future reactor operating conditions. Several potential uranium fabrication and canning process improvements are being developed to further advance fuel element stability and performance. Optimization of process conditions based on these improvements may provide the necessary margin of safety for good bond layer integrity, but before a decision can be made to continue improvement of the present process or convert to a new canning process, more information on the stability of the present fuel element bond is needed. This report presents the irradiation phase of a test which was designed to more fully evaluate Al-Si bond integrity under anticipated future reactor conditions.

The purpose of this Process Improvement Transition Authorization (PITA) is to authorize full pile loading bumper fuel elements in the fringe zone will be reviewed and, if desirable, recommendations to curtail fringe loading may be made based on economic considerations.

The combination of a large reactor core with a sufficiently high flux level results in complications of the dynamic behavior of the core due to asymmetric xenon and temperature instabilities. In both phenomena, a local increase in neutron flux causes an increase in neutron multiplication in a surrounding zone. We consider here the instability associated with the delayed production of Xe{sup l35}, an isotope with a very high thermal neutron cross section. In order for this instability to result in flux oscillations., the neutron production in one part of the core must be independent of conditions in another part. Said in another way this means that the neutron migration area must be considerably less than the square of any core dimensions. Thus, even though the total reactor power is held constant, it may be possible that neutron leakage from one part of the reactor to another is insufficient to counteract the tendency of the local flux to continue increasing in the part of the reactor where it is already high. An asymmetry in the distribution of reactor power may thus tend to increase in After some initial confusion., it has been found that close monitoring of the local power trends …

The oxidation rate of the moderator in D-Reactor has been monitored from samples placed along the length of process tube channel 3478. During the interval from August 8, 1960, to August 10, 1961 were very high, up to 40%/KOD (1000 operating days). From the shape of the front-to-rear burnout profile, the oxidant appears to be oxygen and/or water.

This is the monthly report for the Hanford Laboratories Operation October 1961. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

The following data represents processing conditions used in fabricating prototypic (except for supports) natural and enriched NPR element assemblies and K-type I&E{sup 2} for the following production tests: ETR, 6{times}6 Loop; GEH-10, Nos. 52, 53, 54 and 55; KER Loops; PT-377; MTR Loop; GEH-4, Nos. 68, 69 and 70. The purpose in documenting the data is to provide a permanent record of processing conditions and dimensions which may be referred to for post irradiation analysis and possible future process work. Post irradiation results will be issued by the Fuels Development Operation, Hanford Laboratories Operation, and the test loop operating conditions will be issued by Process and Reactor Development Operation, Irradiation Processing Department, as outlined in the Production Test Procedure.

The purpose of this report is to present experimental data concerning the heat transfer and fluid flow conditions within a C-overbore geometry process channel for the cases of steady state operation, flow plugging incidents, and inlet piping failure incidents.

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The interim design and development status is reported. The scope of the work was limited to conceptual design studies supported by critical experiments plus heat transfer and hydraulic tests. Design criteria, facility and site, reactor, core geometry and composition, fuel elements, reflector, core and reflector support structure, reactor vessel, control and instruments, primary coolant systems, secondary coolant system, auxiliary systems, experimental facilities, building layout and construction, plant ventilation, heating and air conditioning, critical experiments, reactor physics, heat transfer studies, and shieldings are discussed. (M.C.G.)

A function to evaluate the cost of electricity produced by a nuclear power reactor was developed. The basic equation, revenue = capital charges + profit + operating expenses, was expanded in terms of various cost parameters to enable analysis of multiregion nuclear reactors with uranium and/or plutonium for fuel. A corresponding IBM 704 computer program, which will compute either the price of electricity or the value of plutonium, is presented in detail. (auth)

Solutions of the Boltzmann equation for monoenergetic neutron transport in spherical geometry are derived which are respectively singular and regular at the center of the sphere. A few specific partial singular solutions are presented. The regular solutions in spherical geometry are constructed by superposition of solutions in plane geometry which belong to the same k. Finally, the solutions are compared with their representations by a series of spherical harmonics. (D. L.C.)

Laboratory-scale tests on methods for recovering uranium and thorium from graphite-base reactor fuel elements are reported. The 90% HNO/sub 3/ process, which involves simultaneous disintegration and leaching in 21 M HNO/sub 3/, is applicable to all fuel elenments which do not contain coated fuel particles. Leaching of irradiated (0.001% burnup) fuels containing 3 and 12% uranlum recovered approximates 99.3 and 99.9%, respectively, of the uranium in two 4-hr leaches with boiling acid. The graphite residue retained > 50% of the long-lived fission products. Three successive leaches of fuel containing uranium and thorium recovered approximates 99% of both elements. Uranium recoveries by combustion in oxygen followed by dissolution of the ash hn nitric acid or fluorlde-catalyzed nitric acid are quantitative only when the fuel is not coated, does not contain Al/sub 2/O/sub 3/-coated fuel particles, and is free from impurities such as iron. During combustion up to 95% of the Ru-106 was volatilized from irradiated specimens. Recoveries, by leaching with 70% HNO/sub 3/, from fuel specimens containing Al/sub 2/O/sub 3/-coated fuel particles were greater than 99% when the specimens were ground finer than 200 mesh to ensure crushing of the fuel particles. (auth)

Preliminary design concepts are presented for a wire scanner for experimentally evaluating spatial variations of neutron flux in the SM-l reactor core. Results of a literature search and determination of optimum criteria for flux mapping the core in minimum time dictated requirements for design concepts and specifications. The utility of both manually instrumented and automatically instrumented wire scanners was analyzed with respect to rapidity of measurement, selectivity of detector location, cost, value of data, plant downtime, and additional factors. (auth)

A conceptual design and economic evaluation of 300 and 40 MW/.sub e/ steam-cooled fast breeder reactor power plants were performed. A reactor core composed of U-Pu oxide rod-type fuel elements clad with Inconel-X and surrounded by a blanket of depleted UO/sub 2/ fuel was studied in some detail. Reactor breeding ratios of from 1.27 to 1.5 and overall system doubling times of from 20 to 30 years are achievable. For the near term (1967) 300 MW/sub e/ plant, an energy cost of 7.6 mills/kwh is estimated, based on AEC ground rules for privately financed plants and utilities. This cost may go down to 5.7 mills/kwh by 1975. For the 40 MW/sub e/ plant corresponding energy costs are 19.5 and 13.7 mills/kwh, r -spectively. The R&D program required for this reactor concept is estimated at million with an additional million for improvements leading to the 1975 reactor. Investigation of the operational and safety aspects of the reactor indicated that satisfactory procedures can be used for startup, shutdown, and emergency cooling of the reactor. An increase in reactivity upon flooding can be prevented by incorprating small amounts of high resonance absorption material in the core. Preliminary calculations indicate a substantial increase in …

The development of procedures for the vacuum induction casting of U--Mo alloys into both thin (0.010 to 0.100-in. thick) plates and long (36 in.) plates is described. Melting and casting cycles were developed, and casting techniques established, which resulted in sound, integral plates. These plates were evaluated by radiographic and metallographic examination, and by chemical analysis. The results indicated the feasibility of the process for the fabrication of fuel plates for nuclear reactors. The process promises a potential reduction in fabrication costs, by eliminating waste. (auth)