This report talks about the Aerojet- general corporation report no. 2325

Report summarizing ongoing nuclear work performed by the ORNL for the United States Army.

; 8 7 < 8 < : : : 6 9 9 = < 9 < : 5 < > ;" icipation in the program continued to include review, inspection, and support in various areas of reactor technology. An advanced fuel irradiation test program was established that is to be conducted in the pressurized-water loop in the Oak Ridge Research Reactor. Review of the design of the MH-1A reactor was initiated. This reactor, a pressurized-water system fueled with low- enrichment bulk UO/sub 2/ clad with stainless steel, is being designed as a floating plant to furnish electrical energy to shore installations. Studies of the out-of-pile corrosion resistance of stainless steel brazed joints were completed. T-joint specimens of type 304 stainless steel were brazed together with 18 different alloys. Initial testing resulted in the selection of five of these alloys for extended testing, which was carried out in autoclaves with O/sub 2/ or H/sub 2/-O/sub 2/ added to the autoclave water. These alloys, General Electric alloys Nos. 81 and 75, Coast Metals alloy NP, low-melting Nicrobraz, and a Pdbase alloy, were satisfactory. Coast Metals alloy NP was selected as the reference braze material for the SM-1 fuel elements because it was …

Irradiation of Np-237 in Hanford reactors and recovery of the Pu-238 product in a close-coupled separations plant is currently of interest. Such a concept has the potential of increasing in Pu-238 production rates. The results of initial calculations on the subject are presented herein to aid further study and evaluation. Much of the information is presented in terms of the aqueous target system proposed in earlier work (i.e., irradiation and processing of an aqueous neptunium solution). However, most of the information can be converted for evaluation of a solid target system.

This report addresses the compilation of contract year 1964 preliminary tes plans of the WANL - radiation effects program.

Report compiling the melting points of 70 metal oxides published prior to January 1963. Both the original melting point and the equivalent value based on the International Practical Temperature Scale of 1948 are presented. Included in the survey is information on pertinent experimental details such as the method of temperature measurement, purity, furnace type, and environmental conditions.

Technical report. From Abstract : "The stoichiometric formula of the lower chloride in the BiCl3-Bi system has been established as Bi12Cl14 (BiCl1.167) through a dingle crystal, X-ray determination of its structure. ... Previous studies of the Bi-Cl3-Bi system are re-examined in the light of the structural results."

Measurements were made of the total energy deposited at various points within a 42-cm-dia spherical water-filled lucite phantom by the secondary particles resulting from 160-Mev proton reactions with various targets. Target materials were water, aluminum, carbon, copper, and bismuth. Detectors were small lucite-walled ionization chambers filled with 97% A--3% CO/sub 2/ or ethylene gas. Data were taken both with the lucite phantom on the beam axis and with the phantom offset approximately 54 deg -43' from the beam axis. The proton beam energy determined from a part of these results, 160-162 Mev, is in good agreement with published values. The energy deposited by secondary particles was found to increase with Z, as expected. The depth-dose curves obtained have a steeply negative slope over the region near the surface of the phantom and a more gentle slope at greater depths. The magnitude of the dose in the region of the initial slope decreases with increasing target thickness. The dose in this region is presumably due to secondary protons. The magnitude of the dose at greater depths increases with increasing target thickness. At the greater depths the slope of the depth-dose curves, presumabiy controlled by secondary neutron interactions, is similar to that …

Criteria provided in this report delineate the objective, bases, and functional requirements that shall govern the preparation of detail design of the bauxite-sulfuric acid feed facilities to be installed in the 183-KE and KW Buildings. These facilities will produce the chemical coagulant used in the treatment of Columbia River water in the water plants and thus replace the existing liquid alum feed systems used for this purpose. The objective of this document is to define the operational and technical requirements of the new process and to outline the functional requirements of the proposed facilities for the purpose of detail design. The criteria below define the requirements for a single K Area water plant. Unless otherwise stated they shall apply for both K Area water plants.

Estimates are made relating radionuclide movement to ground water velocity as part of the safety program for a proposed experiment to detonate nuclear devices within the Tatum salt dome. The estimates are based on distribution coefficients obtained from laboratory studies. Core samples obtained from hydrologic test well HT-3, Tatum salt dome, Lamar County, Mississippi, were equilibrated with radionuclides in solutions simulating aquifer waters found in the area. The combinations of Cenozoic sand and silty clay, and quality of water of the area were studied and summarized. The distribution coefficients obtained for different radionuclides were tested and indicate retardation factors from 1.3 to 857 for the travel time of these radionuclides when compared to the travel time of water in the aquifer system. Laboratory results indicate that migration of any radioisotope inadvertently introduced to the aquifers in the vicinity of the dome as a result of proposed nuclear test explosions would be extremely slow. Revised estimates of the rate of dissolved radioisotope movement will be made on the basis of further laboratory studies utilizing chromatographic adsorption columns of 0.5 to 4.0 feet in length.

Hot die sizing (HDS) is a process being considered at Hanford to replace or supplement the existing AlSi brazing process. Hot die sizing consists of passing a preheated core-component fuel assembly through a cold the to bond the aluminum jacket to the core while passing a die plug through the internal bore to form the internal bond. Fuel end bonding is accomplished in a following step by applying heat and pressure to the sized fuel element. This report summarizes the fabrication of fuel elements for irradiation testing of hot die sized fuel elements as authorized by ``Production Test IP-546-A, Irradiation of Hot Die Size Diffusion Bonded Fuel Elements,`` HW-75465.

This is a summary report of various technical projects relating to uranium.

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On the basis of 4800 hr exposure to 680 l F water, a number of ternary compositions were shown to have corrosion resistance superior to Zircaloy-2. In addition, the strength and hydrogen pickup properties of the alloys were generally improved over Zircaloy-2. The promising alloys were based on the binary materials Zr- 1Sb, Zr- 1Cr, Zr0.5Nb, and Zr-0.5Sn to which small percentages of Te, Ge, Cr, or Fe were added. At present, 680 l F water corrosion data for an exposure time of 5000 hr are available for the modified ternary alloys. Development of materials for potential service in 750 and 900 l F steam proceeded in a manner similar to that for 680 l F water application. On the basis of corrosion resistance and strength, the alloys Zr3Cr- 1Fe, Zr-3Cr- 0.25Te, and Zr- 1V- 1Fe were considered highly promising and initially acceptable. However, hydrogen pickup properties, which were about the same as Zircaloy-2, were judged as unacceptable. In an attempt to improve this characteristic as well as to further enhance corrosion resistance and strength, an additional series of ternary alloys was prepared. The compositions studied in 900 l F steam did not exhibit satisfactory corrosion resistance. (P.C.H.)

A number of ternary compositions were developed for service in 680 deg F water which meet the objectives of a zirconium alloy improvement program on the basis of corrosion resistance, strength, and hydrogen pickup. These are compositions based on Zr-- 1Sb, Zr-- 1Cr, Zr--0.5Nb, and Zr--0.5Sn with minor additions of Te, Ge, Cr, or Fe. In 750 deg F steam, the materials Zr--3Cr-1Fe, Zr--3Cr--0.25Te, and Zr-- 1V--1Fe are initially acceptable on the basis of corrosion resistance and strength; however, hydrogen pickup is excessive. The current work is for the study of ternary alloys which are intended to show an optimum combination of corrosion, strength, and hydrogen-pickup properties. (auth)

In-pile loop experiments L-2-15 and L-4-16 were designed to test the radiation corrosion of Zircaloy-2 and other possible reactor construction materials in UO/sub 2/SO/sub 4/ solutions. The solutions employed were 0.17 m UO/ sub 2/SO/sub 4/, 0.015 m CuSO/sub 4/, and 0.03 m H/sub 2/SO/sub 4/ in H/sub 2/O for experiment L-2-15, and 0.17 m UO/sub 2/SO/sub 4/, 0.015 m CuSO/sub 4/, and 0.025 m H/sub 2/SO/sub 4/ in H/sub 2/O for experiment L-4-16. The mainstream temperature in the experiments ranged from 278 to 280 deg C. Construction material for the loops was type 347 stainless steel. Specimens of types 347 and 309SCb stainless steels titanium-55A and -110AT, platinum, Zircaloy-2, crystalbar zirconium, and a variety of other zirconium alloys were tested. The power density at core specimens ranged from 19.8 to 4.6 w/ml in L-2-15 and from 5.7 to 1.3 w/ml in L-4-16. For loop L-2-15, the total time of hightemperature operation with UO/sub 2/SO/sub 4/ was 792 hr, during in-pile exposure, and the reactor energy was 1632 Mwh; for loop L-4-16, 1032 hr and 2325 Mwh. During both experiments most of the reactor energy was accumulated at 3-Mw power level. In general, stainless steel corrosion results from these experiments …

In-pile loop experiments L-2-15 and L-4-16 were two of a series designed to test the radiation corrosion of Zircaloy-2 and other possible reactor construction materials in UO2SO4 solutions under various conditions of radiation intensities, temperatures, solution compositions, and velocity flow past specimens.

Visual examination from 600 fuel elements., 300 discharged from the near side and 300 from the far side, showed the following results: (1) Primary ledge corrosion vas evident on 26 per cent of the fuel pieces, 21 per cent on the near side and 30 per cent on the far side. (2) Primary groove corrosion was evident on 3 per cent of the fuel pieces, 6 per cent on the near side and 1 per cent on the far side. (3) None of the fuel pieces exhibited severe localized corrosion. These results agree with previous studies, indicating little change in corrosion environment.

Vapor volume fraction (void fraction) experiments using a gamma attenuation technique were planned. The experimental program required designing a test section and a steam generator. While the design of the test section was straightforward, the steam generator required considerably more attention because of the large flow and power requirements. Both designs were completed. All components of the gamma attenuation detection system were assembled and checked to be sure they were in proper operating condition. An IBM-1620 computer program was written to reduce the count rate data, taken during the experiments, to void fractions. A calibration of the void detection system was accomplished by scanning a Lucite-air mock-up of a water-steam flow pattern. The measured void fraction distribution was in good agreement with the known void distribution of the Lucite-air geometry. (auth)

A literature survey was performed in order to provide an adequate basis for selecting models for comparison with the data. A computer program was written that allows calculation of void profiles based on an exponential model. A variable exponent is input data for the program, which also computes an integrated average void fraction. After the test section installation was completed, shakedown tests were performed to assure that all equipment was operating properly. All water and all air reference traverses were run with the gamma attenuation equipment. The first twophase run was accomplished at a superficial liquid velocity of one foot per second. (auth)

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Declassified 5 Sep 1973. Information is presented concerning LCRE specifications, primary coolant circuit, aaxiliary systems, fuel elements, instrumentation, materials development, and fabrication; and SNAP-50DR-1 specifications, fuel elements, pumps, steam generator, and materials development. (DCC)

A scavenging-precipitation ion-exchange process using phenolic resins was developed to decontaminate lowradioactivity-level process water waste prior to discharge to the environment. In laboratory and small engineeringscale tests, greater than 99.9% of the cesium and strontium, the principal biological hazards were removed from ORNL low-level waste, and the total activity level was lowered to less than the maximum permissible concentration recommended for populations in the neighborhood of atomic energy installations. The water was treated by a scavenging-precipitation with sodium hydroxide, pH 11.7, and ferrous sulfate, copperas-5 ppm Fe, to remove suspended solids and soluble hardness, clarified, and then passed through a carboxylic-phenolic ion-exchange resin to sorb the remaining radionuclides. After passage of 1,500 to 2,000 resin-bed volumes, the resin was eluted with 10 volumes of 0.5 M HNO/sub 3/. Sodium carbonate can be added in the precipitation step to aid the quantitative precipitation of calcium for wastes that contain small amounts of phosphates, or alternatively, an extra ion-exchange column of carboxylic resin can be used to remove calcium and thus conserve the capacity of the phenolic resin for cesium and strontium. Three kinds of studies were made: batch laboratory-scale studies, continuous nonradioactive runs at 15 liters/hr, and runs with radioactive waste …

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