The alpha phase extrusion of dingot bars to rods for core blank heat treatment at MCW offers a three-fold incentive, including: (1) establishment of an independent alternate source of uranium cores, (2) a slight cost improvement of MCW extrusion over FMPC rolling of dingot and (3) potential improvement in core quality and performance. The purpose of this test is two-fold: (1) to evaluate gross dimensional stability and (2) to evaluate the gross irradiation performance of alpha phase extruded dingot uranium fuel cores. The ultimate objective of this test program is to establish an independent alternate uranium fuel core feed source. The irradiation test program will be conducted at B-Reactor.

Chemical engineering analysis and stress analysis for design of the 3 megawatt intermediate heat exchanger and steam generator for service with a liquid sodium heat transfer fluid.

Abstract: The natural circulation, direct cycle, boiling water reactor reference design presented in this technical report is the alternate to the preferred preliminary design developed under Phase I of the PL-3 contract. The 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 training program outline.

Gaseous waste handling systems are compared for direct and indirect boiling water and pressurized water reactors for PL-3 application. Areas that are common to the various concepts are not discussed since they do not enter into a comparison study. The major differences present are in the handling of active gases released to or held in the primary system coolant. These gases which could be present, their possibIe release from the system, and the necessary processing requirements are discussed in detail. Sufficient information is presented to enable one to analyze the methcds required to remove gaseous activity present in other size and type pressurized and boiIing water reactors. (auth)

Three types of frozen-seal valves'' were tested for possible use in the MSRE. The seal was melted by direct resistance heat, by induction heat, and by clamp-on Calrod heat. The frozen seal was made in a preformed restriction section of a standard piece of pipe by a cooling-gas jet stream directed at the restriction. All three valves performed satisfactorily through 100 test cycles. The Calrodheated valve was selected for MSRE use on the basis of simplicity of design and of operation. Two of the valves are successfully undergoing further tests on the MSRE Engineering Test Loop. (auth)

Abstract: Test procedures for special tests involving in-core SM-1 temperature and flow instrumentation are described (Task XIV Package Tests). These tests involve in-core steady state flow and temperature measurements, loss of flow transients, load transients, reduced primary system pressure operations and reduced element flow. The thermal and hydraulic conditions prevailing in these tests, including steady state and transient burnout rations, are developed. The effects of reduced system pressure and flow on the burnout ratios are determined as are the expected stuck rod conditions when Task XIV test elements are installed. The effect on the maximum credible accident is included and a recommendation to conduct these Task XIV package tests is made.

From introduction: This report is one of three volumes of the first annual progress report on the GE-NMPO high-temperature materials program conducted during calendar year 1961.

Annual technical report describing the progress and development of improved zirconium alloys for service in superheated water and steam. This report covers the period between February 1, 1961 to February 28, 1962 and was conducted by the United States and the European Atomic Energy Community (EURATOM).

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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.

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The loading operation for SM-1 Core II is described. Results of startup physics measurements (Test A-300 (Series) and fission product iodine monitoring in the primary coolant are given. The SM-1 Core II initial loading progressed satisfactorily, fulfilling the predictions of the zero power experiment performed at the Alco Criticality Facility. The initial cold clean five rod bank position was 6.53 in.; the initial hot, no xenon, five rod bank position was 9.62 in.; the initial hot, equilibrium xenon, five rod bank position was 11.41 in.; and the initial hot, peak xenon, five rod bank position was 12.14 in. Rods A and B were 19.00 in. in all four measurements. Stuck rod measurements indicated that an adequate shutdown margin was available with 20% of the rods fully withdrawn. All rod calibrations indicated a distinct shift and broadening of the peaks when compared with similar Core I calibrations. The temperature coefficient for Core II was 3.5 cents/ deg F at 440 deg F. Equilibrium xenon was worth approximately - 00 while peak xenon was worth - 43, both relative to the hot, no xenon core condition. During the period June 2, 1961 through September 30, 1961, the reactor operated at a total …

The ICPP processed Al fuel, prtncipally of the MTR-ETR type, durtng this quarter. Newly designed and installed processing equipment exhibited excellent operating performance. This included direct-air-pulsed extraction, stripping and scrub columns, and a cascade-controlled continuous evaporator in first cycle product concentration service. Aqueous zirconium fuel processing studies continued with the objective of adapting the hydrofluoric acid process to continuous dissolution-complexing to increase the capacity of the LCPP process while using as much existing equipment as possible to minimize costs. Six variations of hydrofluoric acid flowsheets were tested in a smallscale continuous dissolvers. Dissolution rates were found to be high in all cases and dissolution was easily initiated at temperatures as low as 36 deg C. Monel and Carpenter 2O(Nb) were found to be satisfactory construction materials, from the standpoint of corroston, if oxidizing conditions were carefully controlled. Additional studies are reported on the stability of blended Zr and Al process raffinates and on the nature of the solids which result from the sodium formate headend precipitation process. Electrolytic dissolution studies, dtrected at fundamentals of current utiltzation in a series''-type dissolver, demonstrated that high current utilization is obtained when the polarization resistance is small compared with the solution resistance. Factors affecting …

Technical report outlining the nature and effects of energy loss and radiation in biological materials during radiobiological processes.

Technical report outlining the effects of chromosomal abnormalities in tetraploid Argentine flint maize pollination, and subsequent generations.

The enrichment inventory required in the Hanford reactors is greatly dependent on the location of the radial enrichment ring. Present and past philosophy has dictated that the radial enrichment be located as close to the periphery of the reactors as possible, consistent with total control criteria, to obtain maximum flattening efficiency. As long as individual tube power limits dictate total pile power this philosophy of maximum flattening is consistent with a goal of maximum production. For the past year the total pile power at the K reactors has been restricted by bulk outlet water temperature limit or administrative total power level limits. During this time fuel ruptures have been negligible due to improved metal quality and axial flux shaping. If present operating conditions continue the relocation of the enrichment ring could be desirable to decrease the enrichment inventory. Moving the radial enrichment toward the center of the reactor to decrease the inventory would result in lower flattening efficiency and require higher tube power limits for the same total pile power level.

Lactates of all the rare earths except lanthanum, cerium, and praseodymium were prepared and studied. In the case of lanthanum, cerium, and praseodymium the laciates are apparently so soluble that they cannot be crystallized directly from aqueous solution. Products obtained by addition of anhydrous ethanol to sirupy solutions of the iight-rare-earth lactates were not well characterized and were deliquescent. Neodymium lactate trinydrate was prepared only by the addition of alcohol to a sirupy solution. Since it was not as deliquescent as the lactates of lanthanum, cerium, and praseodymium, some of its properties could be studied. The lactates of all the other rare earths, samarium through lutetium, as well as the lactate of yttrium, separated readily from aqueous media as trinydrates---the solubility of the hydrated lactate decreasing as the atomic number of the rare earth increased. Thermal decomposition of the crystalline lactates, except those of neodymium and samarium, proceeded through dihydrated and anhydrous states to a composition corresponding roughly to LnlacCOa at a temperature of about 290 to 300 deg C. Ln/ sub 2/(CO/sub 3/)/sub 3/ or Ln/sub 2/O(CO/sub 3/)/sub 2/ was obse rved. In order to make a comparison, thermal decomposition of the rareearth glycolates was studied also. Solubilities in …

From abstract: "Lactates of all the rare earths except lanthanum, cerium, and praseodymium were prepared and studied. In the case of lanthanum, cerium, and praseodymium the lactates are apparently so soluble that they cannot be crystallized directly from aqueous solution. Products obtained by addition of anhydrous ethanol to sirupy solutions of the light-rare-earth lactates were not well characterized and were deliquescent. Neodymium lactate trihydrate was prepared only by the addition of alcohol to a sirupy solution. Since it was not as deliquescent as the lactates of lanthanum, cerium, and praseodymium, some of its properties could be studied. The lactates of all the other rare earths, samarium through lutetium, as well as the lactate of yttrium, separated readily from aqueous media as trihydrates--the solubility of the hydrated lactate decreasing as the atomic number of the rare earth increased. Thermal decomposition of the crystalline lactates, except those of neodymium and samarium, proceeded through dihydrated and anhydrous states to a composition corresponding roughly to LnlacCO3 at a temperature of about 290-300°C. Ln2(CO3)3 or Ln2O(Co3)2 were observed. In order to make a comparison, thermal decomposition of the rare-earth glycolates was studied also. Solubilities in water at both 20 and 60°C have been tabulated."

Declassified 27 Nov 1973.<><DSN>29:020169<ABS>Declassified 27 Nov 1973.

Reproducibly high density UC-PuC pellets containing minor amounts of (U,Pu)/sub 2/C/sub 3/ were made with nickel sintering aid. Average densities of 95% and maximum densities of 96.6% of theoretical were obtained. High density, singlephase UC-PuC pellets were made by sintering at high temperatures. Densities of 94% of theoretical appear to be attainable. Improved identification of microstructural constituents was made by means of metallography, x-ray diffraction, and microprobe analysis. Development of specialized chemical analysis techniques continued. The post-irradiation examination of the UC specimens was completed. Mass spectrometer analysis showed that 0.4% of the fission gases was released. Metallographic examination of the UC did not reveal any channges in structure. Metallographic examination of the metal specimen parts that were in contact with UC was completed. No evidence for interaction with niobium and slight evidence for interaction with stainless steel was found. The latter was not deleterious. Perturbation factors were calculated for the UC- PuC experiments. Negotiations continued, in order to select the most suitable test reactor. (auth)

From abstract: "Electrical measurements have been made on CaB6, SrB6, BaB6, YB2, YB4, YB6, and YB12 for the purpose of testing models of the electronic structure of the boron atoms in the four boride types represented. The compounds CaB6, SrB6, and BaB6 were found to be semiconductors. The Hall coefficients of YB4 was also measured, but could not be interpreted because the free electron theory does not provide a theoretical value for comparison with experiment. The results are in agreement with the models proposed."

A large number of irradiation tests have been conducted with UO{sub 2} fuel elements to characterize the behavior of Zircaloy cladding when cooling water is allowed to penetrate the cladding to the fuel core. The test element, is designed so that cooling water will be in contact with the surface of the UO{sub 2} fuel core during irradiation. The irradiation behavior of the UO{sub 2} will be compared to the behavior of similarly irradiated, non-defected UO{sub 2} fuel-elements. Several characteristics are of particular interest: (1) Previous investigations have shown that fission fragments formed in UO{sub 2} fuel cores tend to migrate from their point of formation. The fission fragment distribution in this element will be compared with that observed in other irradiated UO{sub 2} fuel cores. (2) This test will provide a direct measure of fission products released from UO{sub 2} during irradiation. (3) Sintered UO{sub 2} is quite resistant to erosion and is, in general, chemically stable in a water environment. This test will provide data concerning the dissolution rate of UO{sub 2} in water during irradiation. (4) Since the coolant will be in direct contact with the fuel, the average temperature of the fuel will be comparatively low. …

A reinforced-concrete steel-plate-lined vault and steel vault door were exposed to a nuclear detonation (shot Priscilla, about 37 kt, balloon suspended, at 700 ft) at the predicted 75-psi peak incident pressure level (1150 ft from Ground Zero). The vault was designed by the ultimate-strength theory to utilize the additional strain energy available in the elastoplastic and plastic ranges. Flexural and thrust capacities were determined, and shear capacity was computed. It was established that above-ground structures can be designed and constructed to survive the pressure levels experienced by this structure. The vault and door provided adequate blast and thermal protection for normal usage, and the structure was adequate to resist overturning and excessive sliding under conditions of the test.

From introduction: "This brief paper deals with the subject of radial temperature differences for heat transfer to fluids flowing through concentric annuli under the following conditions and assumptions: a. constant heat flux; b. fully-established turbulent flow; c. heat transfer through inner wall only; d. physical properties independent of temperature."