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.

Fourteen NAE1 fuel elements were discharged from KER Loop-4, after irradiation to an average exposure of 1250 MWD, at prototype N-Reactor coolant temperature and pressure. The elements were disassembled and measured in the KE fuel examination facility. This report includes all measurements, except the profilometer data.

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.

A study was made to determine operating temperatures of old pile HCR sheaths. The calculations were performed with the aid of a FORTRAN coded program for the IBM 7090. The difficulty of determining the correct value for the contact coefficient between the HCR and the graphite channel at any point in the channel length resulted in performing the calculations in a parametric style. The independent system parameters were varied during the calculations. These parameters are diametral spacing between the poison rings and the outer aluminum sheath, heat generation rate or reactor power level, the heat transfer contact coefficient between the rod and the graphite channel, and the cooling water temperature or cooling flow rate. For contact coefficients between 200 and 700 Btu/hr -- ft{sup 2} -- {degree}F and for diametral clearances around the poison of 5 mils to 20 mils, the maximum sheath temperature was calculated to vary between 350{degree}F and 650{degree}F. The dependence of sheath temperature upon the rod coolant temperature or coolant flow rate was found to be small enough to be neglected in the normal flow range of 12 gpm to 15 gpm. For a given increase of the average coolant temperature, the maximum sheath temperature increased …

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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.

Irradiation of process tubes containing appropriate parent materials has been proposed by Manufacturing as a method for obtaining new products from the Hanford Reactors. The process tubes would be removed at appropriate intervals and shipped to separations plants for recovery of the products. The tube residence in the reactor could be determined by the optimum irradiation period for isotope production rather than by the period required to corrode tubes of current design to the minimum permissible wall thickness at replacement. This paper, presents an analysis to determine the benefits from red reducing the initial wall thickness of the process tubes below the current 65 mils when the desired residence for isotope production is shorter than the residence based on maximum permissible internal corrosion for tubes of current design.

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.

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.

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

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 …

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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)

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 …

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)

Calibration of the liquid-level detector for steam-water facility is complete from 600 to 2000 psig. Permissible steam release rates for 24 and 36 inches above the interface are established for the test pressures. The correlalation equation which relates steam void fraction to basic system parameters is further refined to include ail the data from this program as well as from the low-pressure steam-water facility. By addition of another dimensionless term to the equation it is possible to extend the equation to include the radial gradient in the void fraction. Collection of carryunder in the downcomer region of the natural separation reactor mockup is complete. From this data it is possible to establish an equation for the downcomer slip ratio. The development of two types of upcomer separators is discussed: a small separator to fit over a fuel element or a group of elements, and a large separator to get over an entire core. The initial small separator under investigation is a centrifugal axial flow model. Results from this program are encouraging. The large upcomer models tested are of the double-helix type. Results with these separators are not encouraging, but data from the small separator program should provide useful information to …

Installation of the test equipment and checkout of the steam-water test facility's controls for the first series of tests was completed with favorable results. Tests to determine the relation between steam void fraction and superficial steam velocity are complete through the pressure range of 600 to 2,000 psig. A correlation equation which relates steam void fraction to basic system parameters ia reported. The liquid level detector is in operation, and preliminary tests were performed. Detector performance is as predicted. Void fraction measurements in the downcomer region of the reactor mock-up were completed. Results show that by using a reduced area riser, carryunder in a natural separation system can be greatly reduced. The prediction equation which relates riser geometric parameters and fluid properties to downcomer voids is refined to include results of the large diameter tests. The resulting equation more accurately describes large diameter risers. Development and testing of the 8- and 10-inch diameter centrifugal downflow separators were completed. The best separator tested to date has a flow capacity of 2800 gpm with 0.75% carryunder. Development of an analytical approach to design of this type of separator for a given set of pressure and flow conditions is in progress. (auth)

; 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 …

Nuclear cross section data for gadolinium, Gd/sup 155/, and Gd/sup 157/ were compiled. Average absorption cross sections were computed assuming Maxwellian thermal spectra for moderator temperatures from 70 to 600 deg F. Poison demand curves were computed for a boiling water reactor with an H/sub 2/O- to-UO/sub 2/ ratio of 2.6, U/sup 235/ enrichments from 1.8 to 3.8 wt%, zirconium or stainless steel as clad, and fuel depletions of 10,000 to 25,000 Mwd/t. Based on the above poison demand curves, poison lumps in the shape of cylinders are being investigated analytically. A calculational method is being developed for detailed depletion studies of cylinders. Gadolinia, alumina, and Al/sub 2/O/sub 3/- Gd/sub 2/O/sub 3/ rods wer e formed and sintered at temperatures to 2700 deg F, in an air atmosphere for two to six hours. Investigations of reaction between Gd/sub 2/O/sub 3/ and Al/sub 2/O/sub 3/ were begun. (auth)

A semicontinuous ion exchange column was used for the pilot-scaie production of purified uranyl chloride solutions from scrap leach liquor. Continuous box mixer-settler tests demonstrated that an Aliquot-336--Solvesso 100 system will satisfactorily extract uranium from scrap leach liquor. An airless shot blaster is being used to clean slag from debries using uranium shot as the blasting medium. Plasma-spraying equipment was installed for coating graphite crucibles and molds with refractory materials. Production-size uranium ingots having fine, evenly dispersed inclusions were produced by consumableelectrode arc meiting under vacuum. Variations in uranium billet temperatures prior to rolling were found to affect the amount of recrystallization and the texture coefficients in the alpha-rolled rods. Four HAPO I and E fuel cores that were beta treated and water quenched were examined for texture gradients. G/sub 2/, G/sub 3/, and J data are tabulated. Uranium phosphide inclusions were identified in productiongrade derby metal of high phosphorus content. (auth)

For purposes of analysis and experiment the centrifugal type downflow separator was divided into the inlet nozzle, separating zone, and outlet nozzle. The analysis and experiments have resulted in a new outlet design, a method of determining separating length, and a more effective inlet nozzle. The results have caused a reduction in pressure loss from 5 ft of water for the reference design to 1.5 ft of water for the new design at a flow rate of 1400 gpm. A reactor core riser and downcomer region was mocked-up in the large air-water tank. Void fraction in the downcomer region was measured as a function of water velocity, water temperature, inlet gas flow rate, and riser geometry. Results show that the void fraction in the downcomer is essentially zero until a threshold downcomer velocity is reached. The void fraction then rises rapidly with increasing water velocity to approximately l1% and then appears to remain constant. Test data from this experiment are being correlated using a dimensional analysis technique. An initial prediction equation was developed. (auth)