This article presents data on the effects of gamma radiation on the physical properties of a number of different classes of plastics materials of construction which were screened to determine their capabilities and limitations for service in applications where ionizing radiation exists. The irradiations were carried out in air at ambient temperature at normal atmospheric pressure to various gamma radiation doses up to 1*10^9 r depending upon the stability of the materials to radiation exposure. The different classes of materials and materials within a class are compared.

Abstract. Some recent analysis of certain Lid Tank fast neutron dose rates measured in oil and water indicates that there are some basic inconsistencies with reported oxygen and carbon removal cross sections and the reported data. These inconsistencies may be explained in several ways: (1) The reported carbon removal cross section is wrong. (2) The reported oil composition is wrong. (3) The reported oxygen removal cross section is wrong since it is based on an assumed rather than a measured oil composition. (4) Some of the experimental data are wrong. It is not possible to determine which of the above is most likely on the basis of analysis alone but the possibilities are pointed out and, based on the assumption that all the experimental data are correct, it appears most likely that the oil composition assumed in ORNL 2197 was in error.

Abstract; SM-1 Reactor Core I Rearranged and Spiked, and Core II with Special Components were analyzed under various off-design conditions to induce nucleate boiling. The steady state code, STDY-3, written for the thermal analysis of pressurized water cores, was employed for the analysis. The code performs a complete steady state parallel channel thermal analysis for both nominal and hot channels. Thermal characteristics of individual elements were investigated while changing the parameters of primary pressure or inlet temperature to introduce the phenomenon of nucleate boiling in the the core. Reduction of system pressures to 1000, 800, and 600 psia and increasing core inlet temperatures to 465 and 500 degree F were studied as the means to induce boiling in the core. This analysis indicates that SM-1 Core I Rearranged and Spiked can be safely operated at the reduced pressure of 910 psia without introducing extensive boiling in the core. SM-1 Core II with Special Components can be operated at 800 psia or at an inlet temperature of 500 degree F at 1200 psia.

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 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.

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.

In connection with current updating of the NPR testimony, calculations of NPR production parameters have been made for a variety of assumed operating conditions. All cases assume operation at full design power of 4,000 MW. Initial and final enrichments and plutonium production per MWD of exposure have been calculated for a nominal 22 lb/ft. metallic uranium fuel element at exposures of 1,000, 1,400, and 2,000 MWD/T; 23 lb/ft. fuel element at 1,000 MWD/T; 24 lb/ft. element at 1,000 and 2,000 MWD/T; a UO{sub 2} fuel element of about 12.9 lb/ft. at 2,000 and 3,000 MWD/T; and a 22 lb/ft. uranium fuel element at 1,400 MWD/T with the Zr-2 process tube replaced with a stainless steel tube of 170 mil thickness. The fuel element geometry in all cases is a tube-in-tube. All calculations for this review were made with R.O. Gumprecht`s MOFDA fuel design code for the IBM-7090. The dimensions of the basic 22 lb/ft. fuel element were Gumprecht`s current best values. For all other cases, the ratio of inner tube volume to outer tube volume was held the same as in the 22 lb/ft. case and the dimensions were adjusted by the program to obtain a balance of water temperatures …

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.

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.

Reported herein are the results of laboratory heat transfer experiments. These experiments were conducted to investigate fuel element temperatures which could result from coolant flow loss following a failure of the inlet piping to a process tube at a K reactor. The failure of the inlet coolant piping between the front header and the process tube on a reactor would cut off the flow of cooling water to the fuel elements but should immediately initiate a reactor scram by causing a low trip on the Panellit pressure monitor. However, the reactor power reduction would not be immediate nor absolute and would be dictated by the time required to insert the emergency control rods and by post-scram delayed fission and fission product decay heating. The only means of heat removal from the affected tube and fuel elements during the post-scram period would be by reverse flow of hot water from the rear header. The objective of the subject experiments was to determine what rear header pressures would be required to achieve adequate cooling of a K reactor fuel assembly during such a post-scram period. Such information is of value in updating of reactor hazards evaluation reports. Experimental studies were previously reported …

This report presents the results of heat transfer experiments which were conducted to determine the ability of a K-reactor fuel channel assembly to undergo various degrees of coolant supply header pressure decrease without resulting in fuel jacket melting. The experiments were conducted with an electrically heated test section in the single tube prototype Heat Transfer Apparatus of Thermal Hydraulics Operation.

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.

The purpose of this report is to present some experimental data concerning boiling burnout and to discuss briefly the significance of these data in relation to the Hanford production reactors.

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.

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.

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.

This is the fifth quarterly report that documents a system of mass flow that can record measurements of homogeneous flow, slurries, highly corrosive fluids and multiphase fluids, additionally considering pressure drops, measuring external to the flow, ruggedness and reliability.

This is the third quarterly report that documents a system of mass flow that can record measurements of homogeneous flow, slurries, highly corrosive fluids and multiphase fluids, additionally considering pressure drops, measuring external to the flow, ruggedness and reliability.

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)

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 …

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

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)

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.)