System curves for each portion of the C Area Water Plant were obtained from referenced work and are presented in figures. Field test data, corroborating the calculated curves, are presented as singular points on the same graphs. Present maxima capacity of the C Area Filter Plant was 121,000 gpm with 118,000 gpm available for use as primary reactor coolant. Modifications to the filter effluent piping would increase this available flow to about 180,000 gpm. Of the 118,000 gpm available for C Reactor use, 10,000 to 12,000 gpm was demanded by B Area through the 183 BC intertie. The maximum flow that the intertie line could handle, without reducing the filter capacity of the C Area filters, is about 21,000 gpm.

A low-background facility is described for rneasuring lung burdens of U, Th, and other nuclides in vivo. Problems associated with this method of radiation measurement are discussed. A mathennatical, computer-oriented simulation was devised to gain insight into the relation of the net observed radiation spectrum to the burden of radioactivity in the body or its organs. Chest cavities for persons of three sizes were synthesized in a three-coordinate space comprised of one-inch cubes and including a 9-in.-diameter crystal detector. Data, describing the tissue composition of each cube in the body and the characteristic radiation attenuation for each tissue-type, were coded for use with a program on a high-speed digital computer. Efficiencies for measuring radiation emitted by numerous point sources of enriched uranium were calculated. Data on in vivo measurement efficiency were obtained assuming uniform distribution of radioactive material throughout the lungs and also for nonuniform deposits. The effects of individual size and geometry, and of detector position on the measurement efficiency were determined for these twvo categories and radiation flux distributions on the detector face were computed in some cases. Data are appended and a flow diagram of the computer program is included. (C.H.)

None

A mass flow measurement technique was devised which meets many requirements for mass flow measurement systems. In this technique. fluid is made to pass through a pivoted S-shaped tube. which tends to rotate due to the angular momentum of the fluid. A torque motor and torsion spring maintain zero displacement. the restoring force being a measure of the angular momentum, and density is measured by absorption of radiation passing through a window in the S tube. From the measured angular momentum and density, the mass flow can be computed. Two experimental mass gages using this technique have been constructed. (D.L.C.)

In connection with an evaluation of the operability of a 36-inch diameter remote evaporator at the Idaho Chemical Processing Plant that was to be packed with a corrosionresistant neutron-poison packing for criticality control, an investigation in a 30-inch diameter vessel proved that air sparging effectively mixes solutions. The data showed that at similar spar;e rates the presence of the packing caused an increase in the time needed for complete mixing. The investigation showed that solutions are readily evaporated in spite of the presence of packing in the tank. (auth)

Conversion of the ORR from 20 to 30 Mw operating level was achieved during July 1960 after a scheduled shutdown for completion of a new cooling system. Operating time reached a high of 82% during the last quarter of 1960. The first quarter of 1961 showed an operating time of 80% despite 2 shutdowns and some additional down time to repair mechanisms associated with the shim rods. Changes were made in the south engineering test facility for GCR test loops. Detailed data taken from quarterly operational reports are included. (J.R.D.)

Interim results of analytical investigation of nuclear and thermal characteristics of SM-2 type fuel elements in SM-1, SM-lA, and PM-2Ae reactor cores are reported. Utilizing modified two-proup diffusion theory, predictions of power distribution and core and rod reactivity were performed. The calculations indicate that use of SM-2 fuel elemente in SM-1, SM-lA, and PM-2A is feasible from the nuclear standpoint. A steady state thermal analysis of each plant utilizing SM-2 elements was carried out. This analysis showed that the minimum departure from nuclear boiling ratios were considerably abovs the minimum value ftom the design criteria standpoint at both the operating and scram power levels. The investigation indicated that SM-2 elements can successfully be employed in SM-1, SM-lA, and PM-2A as replacement core elements. (auth)

None

Experiments using a plastic-moderated assembly as a radiation source are described, and the hazards attendant to these experiments are evaluated. The critical assembly, designated Battelle's Plastic Reactor Facility, is constructed in the form of a cube. A portion of the plastic and fuel will be removed from the center of this reactor to provide a radiation environment for exposing fission-chamber devices. This central void region will approximate a flux trap in which a bigh neutron-to-gamma ratio is expected. The fuel-element assemblies are composed of strips of aluminum and Teflon-coated uranium sandwiched with plastic and encased in aluminum boxes. One portion of the core is on a movable table, while the other part is on a fixed table. The core is assembled remotely by driving the movable table against the fixed table. Primary control and safety of the assembly are achieved by insenting or withdrawing regulating and safety elements and by increasing or decreasing the distance between the two core halves. For safety, the maximum normal rate of reactivity addition has been limited to 0.04% DELTA k/k per sec for control-element withdrawal and table closure. The system is interlocked so that only one control element can be withdrawn at a time …

The Birch Plant Improvement Program is described in this document. This program was instituted to fully support the AEC program for maximum recovery of neptunium from current separations feed material and other sources as the AEC may consider appropriate to their needs for increasing the availability of neptunium.

The types of tests performed in SPERT-I, and the tests proposed for SPERT-II and -III, are described. These reactors are described, and factors influencing their dynamic behavior are discussed. The tests are classed as static, step, ramp, and oscillatory. The correlation between the test results and the reactor dynamic safety characteristics (stability, self-shutdown under excursion conditions, etc.) is investigated. (T.F.H.)

Interest in neptunium recovery became active in early 1958. At this time neptunium chemistry was sketchy, the path of neptunium through the Purex Plant was not clearly defined nor were the effects of operating variables on the path. An intense research and development program carried out in Hanford Laboratories Operation, research and Engineering Operation (CPD), and Purex Operation led to the first recovery attempt in June 1958. Continuing efforts on flowsheet improvement resulted in further neptunium recovery from Purex. In April 1961 a routine neptunium isolation run was undertaken. During the course of this run approximately 75 per cent of the accumulated neptunium was lost to the high level waste storage tanks. It is the purpose of this report to document the facts surrounding this occurrence.

The effect of very high heat fluxes on the corrosion of 1100 and 6061 aluminum alloys by water was investigated. The test conditions generally simulated those expected to exist during operation of the High-Flux lsotope Reactor. At heat fluxes between 1 and 2 x l0/sup 6/ Btu/hr-ft/sup 2/ and with coolant temperatures and velocities in the ranges of 13l to 250 deg F and 3l to 51 fps, respectively, a layer of boehmite ( alpha Al/sub 2/O/sub 3/- H/sub 2/0), which has low thermal conductivity, formed on the water-cooled aluminum surfaces during test. When only relatively thin films formed, the boehmite adhered tightly to the aluminum, but in those cases where relatively thick films formed, some boehmite spontaneously spalled from the surface. The rate at which the boehmite formed on the surface (and consequently the rate at which the aluminum temperature increased) was a function of the temperature at the specimen-water interface and the pH of the coolant. The lower the temperature and the lower the pH (in the range of 5.0 to 6.5 with HNO/sub 3/), the lower the rate of corrosion- product formation. Within the ranges investigated, pressure and flow rate were without effect, and the same results …

Activities are reported on development work on a mass flow system capable of measuring externally the properties of homogeneous flow, slurries, highly corrosive fluids, and multi-phase fiuids. In the proposed system, the fluid passes through an S-shaped tube wherein measurements of angular momentum and density yield mass flow directly. (B.O.G.)

This report, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; and employee relations.

White Oak Lake, Oak Ridge, Tennessee, received lowlevel radioactive wastes from Oak Ridge National Laboratory for 12 years prior to draining in 1955. Studies on the insects inhabiting the vegetation on White Oak Lake bed revealed 401 species present during 1956 and 1957. Most numerous were members of the insect Orders Hymenoptera, Diptera, and Coleoptera. In the summer of 1956, the first summer following draining of the lake, there were relatively fewer species of insects represented by large numbers of individuals. In 1957, there were relatively more species of insects but fewer individuals were present. By the end of the summer of 1957, only two years after the lake was drained, the vegetation supported a rich and varied insect fauna. (C.H.)

The purpose of this document is to present the results of an analog simulation of the behavior of a Hanford reactor upon the simultaneous loss of primary coolant and introduction of a reactivity ramp during normal shutdown.

This incident report deals with mechanical damage (caused by falling rocks) to the power line supplying station power, Priest Rapids Dam lost all generating flow at 4:23 p.m., cutting discharge from 71,700 cfs to about 12,000 cfs. Within five minutes, spillway gates were opened, bringing river flow back to greater than 36,000 cfs in about 10 minutes. The flow at 181-B dropped from 72,000 cfs to a minimum of 56,000 cfs at about 5:25 p.m. Priest Rapids generators returned to service at 4:45 p.m., the indicated flow at the gauge reaching 71,700 cfs again at about 8:00 p.m. River temperatures at the gauge increased 0.5 C following the interruption, but not at 181-B. Prompt HAPO notification of the flow reduction as provided for in the agreement between the PUD and the AEC was not made on this occasion; the first notice came from the 251 Substation.

The effects of certain pulse-shaping networks on the signal-to-noise ratio of a nuclear pulse amplifier were considered. The shaping networks discussed are: equal RC-integrating and RC-differentiating time constant, single- delay-line clipper and RC integrator, and doubledelay-line clipper and RC integrator. The effects of these networks on the signal, when high count rates and overload pulses are present, were also considered. Equations and curves were developed for the energy resolution (signal-tonoise ratio) and resolving time (related to the ability to operate at high counting rates) of the networks. Experimental results are shown for the energy resolution of the types of pulse- shaping networks considered. (auth)

The Expansion Joint Test are detailed in this report are part of the work accomplished under Design, Development and Research Contract DDR-111 between General Electric Company, Hanford Atomic Products Operation and Washington State University. The equipment and instrumentation used for the K-Downcomer tests were arranged to permit installation and testing of the expansion joints.

An experimental evaluation of the shield design for the ML-1 mobile reactor was made at the Lid Tank Shielding Facility. Thermal-neutron fluxes, fast-neutron dose rates, and gamma-ray dose rates were measured behind slab mockups of the basic shield design and a number of possible variations. The designs embodied various combinations of lead, Hevimet, stainless steel, boral, water, and aqueous solutions of ammonium pentaborate at two concentrations. The after-shutdown decay characteristics of the basic design were determined, and data were obtained from a fairly accurate mockup of the stainless steel plenum and gas duct typical of the top and bottom regions of the ML-1 shield. Analysis of results and application to the final shield design are not reported. (auth)

The adequacy of the EGCR steam and feedwater systems for continued heat removal from the core was assessed for the following accidents: steam line failures, steam generator internal failures, and feedwater system failures. The reaction of the reactor coolant blowers and vessel cooling compressors to steam- helium mixtures was evaluated. The steam generator isolation systems are described. The containment system was found to be adequate to prevent activity release to the atmosphere in the event of all credible steam line or feedwater system failures. The design of the drum water level automatic isolation system does not prevent activity release during certain credible steam generator failures, however, this activity is released at the top of the plant stack and does not result in a dose which exceeds the acceptable yearly exposure. (M.C.G.)

The fission products radiostrontium and radiocesium are being recovered for sale, and there is concern about the possibility of revealing irradiation conditions through their compositions. This report presents results of preliminary studies which indicate some possible security problems.

This report presents a planned reactor test program in support of the research and development study to investigate the formation of radioisotopes in reactor cooling water and to devise means of reducing the amounts of those radioisotopes formed which result in significant exposures to downstream users of the Columbia River water. The effects of water treatment process changes and reactor operation on the formation of radioisotopes of radiological interest will be studied utilizing four of the 1706-KE single-pass tubes together with special demineralizer and water treatment facilities to be obtained in 1961. Specially treated water containing no salts, single salts, mixtures of salts, special additives, and complexing agents will be investigated. The effects of varying water composition and pH, of using aluminum and zirconium reactor tubes, and of varying reactor conditions of tube outlet temperature, water flow rate, flux, and power level will be tested. 1 ref.