The Fuel Cycle Program is an integrated program of investigation in the Vallecitos Boiling Water Reactor (VBWR) and other facilities to improve the technological limits of boiling water reactors in several areas. This report presents updates on tasks related to those areas.

Technical report. From Abstract : "A system has been developed to program a digital computer in much the same manner as an analog computer is programmed. Properties of a general purpose analog computer and a digital differential analyzer are combined to yield a program which employs, as input, standard data cards prepared from a diagram analogous to an analog computer diagram. The capacity of the system is much greater than that of most analog computers, making it applicable to large simulation problems. No scaling is required. The system is compatible with the Fortran symbolic language and may be used as part of a larger digital computer program. It may be applied to the more general types of boundary value problems arising in process simulation in addition to the initial value problems ordinarily solved on analog computers. It is particularly well adapted to non-linear problems and to control problems involving large transport delays. The program operates in basic machine language and, when used by itself, requires no compilation time."

Abstract: "Geodimeter surveys were made on four traverses near the east scarp of Rainier Mesa between October 27, 1967 and April 26, 1968. These measurements were made in order to determine if an changes in distance between survey stations had resulted from downslope creep of the rocks behind the scarp, or if any displacements had resulted from nuclear explosions beneath the mesa and in adjacent areas. A persistent small but possibly significant, slopeward displacement was measured along one traverse and may represent detachment of a large block; monitoring for an additional period of time is recommended."

Irradiation of the initial test in this program involving ten tubes of alternately charged nickel-plated C-64 alloy clad test elements and X-8001 alloy control elements has been successfully completed. The test indicated that the nickel-plate spalling problem has been resolved as no significant spalling or flaking was observed during the post-irradiation examination. The second test in this program will be to verify the performance of nickel-plate with a pilot loading (up to 100 charges) of fuel elements which have been nickel-plated on a production basis. The objectives of this test are to demonstrate with a larger scale test that nickel-plate performs satisfactorily and that reducing the nominal plate thickness from .6 mil to .2 mil will not affect the performance of the nickel-plate fuel element. This test authorizes the irradiation of up to 100 columns of OIIN, chemically nickel-plated, C-64 alloy jacketed fuel elements to 200% of normal goal exposure to extend the evaluation of nickel-plated fuel elements on a pilot scale at DR Reactor. Seventy columns will be plated to a nominal thickness of .6 mil and thirty columns to a nominal .2 mil thickness. Twenty measured columns, ten representing each plate thickness, will be charged to monitor the …

The programmed long outage which started May 1 was concluded on June 6. During this outage, the sixth heat exchanger cell was tied in, the auxiliary steam boiler was completely overhauled, four steam generators were equipped with isolation valves, extensive replacements of the process tube resistance temperature detectors were made, the reactor was re-orificed, a leaking reactor gas seal was repaired, extensive preventive maintenance was done, and numerous smaller pieces of work were accomplished. After startup, testing of the WPPSS generating station equipment was resumed. Electric power up to 320 Mw was generated, generator trip-offs were executed, and data on the reactor plant response were taken. Both the generating station and the reactor plant behaved excellently. The sixth cell installations performed as designed. Electrical power generation is now limited to about 400 Mw by the unavailability of a step down transformer on the BPA system at Vantage. Elevation of the generated power level to the rated 800 Mw will be possible in the early autumn when transformer delivery is expected by BPA. Three scrams from relatively low power levels occurred during the June 6 startup, which may be regarded as being associated with returning the plant to service after a …

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.

The raw water export system forms the last ditch water supply system to the ``O`` and ``C`` type reactors; in the event of electrical and steam power failure, the export system is designed to supply enough raw water coolant. After the original export orifice was modified twice, the export system was retested.

Neptunium-237 production has been-calculated according to the equations documented in Reference 1 for use in 1964-AEC study. cases. The assumptions made in the forecasts are as follows: Only the 947 stream is recycled. The blend enrichment for the recycle maternal is 1.2 per cent for F-Y 64 and 65. For FY 66 -73, a blend enrichment of 1.4 per cent is assumed which is better value according to the optimization model of C. W. Showalter. The 947 stream recycle time is two years. In cases A and B where the 947 stream volume is materially increased, virgin 947 material is introduced to make up the material balance. The U-236 content of the blend material- is 66 ppm{sup 2}. The U-236 input level in the 947 stream for the input of FY 64 is 200 ppm{sup 1}. Separations recovery is 85 per cent for FY 64--65 and 90 per cent thereafter. The material throughput data are given in Table I. These values are consistent with those being assumed for the AEC study cases. The Kgs of Np-237 by fiscal year are given in Table II. It must be remembered that these forecasts are based on a particular set of idealized assumptions, …

The purpose of this review is to report the operability status of the reactor control instrumentation. The status of the instruments was determined twice during the first six months of this year, April 1 and June 1. The information contained in this report is not intended to be a complete description of the control instruments, but only as they apply to reactor control. The assigned Pile Physicist at each reactor reported the status of instrumentation at his reactor. Chart I summarizes the operability status of the various instruments. Chart II shows the relative range of reactor power over which these control instruments apply. Appendix II contains a functional description of the instruments and Appendix III lists how each instrument is used during reactor operation.

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.

Paper presented at American Nuclear Society Meeting, June I8-21, 1962, Boston, Mass. Design and operating problems of unclad and ceramic gas-cooled reactor fuels in high temperature circulating gas systems will be studied using a test facility now nearing completion at the Oak Ridge Research Reactor. A shielded air-tight cell houses a closed circuit gas system equipped for dealing with fission products circulating in the gas. Experiments can be conducted on fuel element performance and stability, fission product deposition, gas clean up, activity levels, component and system performance and shielding, and decontamination and maintenance of system hardware. (auth)

Computations of proton recoil energy losses from neutron irradiation of hydrogenous systems possessing plane symmetry were made for the purpose of fast neutron counter design. A digital computer code was written to carry out the calculations for two irradiation geometries: the neutron beam is assumed to be normally incident upon the system, and the system is exposed to an isotropic flux of neutrons. Numerical results are presented for neutrons of various energies incident on plane radiators of different thicknesses for these two irradiation geometries. These results were used in the conceptual design of fast neutron dosimeters employing a generalized dosimetry principle. (auth)

None

This report provides a bibliography of sources published between June 1967 and June 1969 related to organized crime in the United States divided by material type (books and documents, Congressional Record, magazine articles, and newspaper articles).

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 …

None

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)

Uranium-bearing thermoelectric compounds are now being prepared by tantalum bomb melting and by the hydride process. Tests of devices made up from these compounds indicate that the main fabrication problems are densification and contact bonding. Data from a hot-swaged pellet and a swaged device of US/sub 2/ indicate some promise for that compound. Improvements in techniques of thermoelectric parameter measurements include programming of automatic test data recording at desired intervals around the clock; increased accuracy and versatility of measurements through use of a newly-constructed adjustable precision resistor; and a method for measuring which should lead to an experimental means for determining the thermoelectric figure of merit, Z. Potential profile studies on PbTe pelleta are yielding important information on contact resistance parameters. A fission-fired thermoelectric generator is being prepared for the next in-pile test. (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 …

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)

None

None

A generalized least-squares technique can be used to calculate equilibrium constants in solvent extraction equilibria by minimizing the difference between observed and calculated distribution ratios. The method was successfully applied to the extraction of uranyl nitrate and nitric acid by TBP dissolved in an inert diluent. (auth)

A brief review of boiling for ordinary fluids with emphasis on the wall superheat required for bubble initation from different surfaces at varying pressures is presented. Calculations based on an equation valid for normal fluids indicate that superheats from two to eight times that for water are required to initiate a vapor bubble in some of the alkali metals. The high superheat and the high thermal conductivity peculiar to liquid metals are combined to explain the unstable liquid-metal boiling phenomenon in natural- and forced-circulation loops. Forced-convection boiling-potassium heat-transfer coefficients correlated by the Locknart-Martinelli two-phase flow parameter are in reasonable agreement with the forced-convection boiling-water data of Dengler and Addoms. Some recomraendations on bubble initiation sites and future work with liquid metals are also included. (auth)