This report recommends developing a statistical model to characterize the three-dimensionall pattern of thermal plumes, It indicates the danger of drawing detailed conclusions on the basis of a small number of observations, and emphasizes the need for proper interpretation of empirical measurements, Two reports, on the discharges (1) from the Dresden Power Station into the Illinois River and (2) from Waukegan Station into-Lake Michigan are analyzed. In conclusion, a plan for interpretation of temperature measurements is recommended.

This memorandum provides a discussion of the K Reactor dichromate test results. Recommendations as to sodium dichromate concentration control are provided. (FL)

At the present stage of reactor coolant studies it is possible to say that radiation has an effect on the deposition behavior of corrosion products in the reactor. A quantitative relation between flux ad deposition rate or amount is not available, nor is there information as to exactly which component, or combination thereof, of the radiation is of most importance. To determine these effect conducted and those of other process parameters, Battelle-Northwest (BNW) conducted an experimental program for the Knolls Atomic Power Laboratory (KAPL) to evaluate the behavior of corrosion products in pressurized water-cooled nuclear reactor systems. Testing procedures and results are detailed in this report.

This final report details production test-176, the primary purpose of which was to provide a quantitative comparison of fuel element 8001 aluminum alloy cladding corrosion in cooling water containing 0.5 and 1.0 ppm sodium dichromate inhibitor. Secondary purposes were to provide information on cladding corrosion as a function of temperature and as a function of time at both inhibitor concentrations. The primary purpose was well fulfilled: Equations were developed to describe and predict fuel element weight losses at both dichromate concentrations and the effect of dichromate concentration on localized corrosion was clearly established. Little effect of dichromate concentration on localized corrosion was observed when fuel cladding temperatures were below 115--120 C but above these temperatures the localized corrosion definitely was more severe in coolant containing 0.5 ppm inhibitor. The bottom rows of fuel element supports rapidly corroded away at the higher temperatures which caused a further increase in local temperatures due to flow imbalances and this in turn accelerated the localized corrosion. Because the higher dichromate concentration reduced the amount of support corrosion, 1.0 ppm dichromate should be employed to control localized corrosion whenever cladding temperatures are expected to be above 120 C for long periods of time. These tests …

In this semiannual issue the single pass reactor and water plant data are shown for the six-month period January through June, 1970, together with data for the three preceding six-month periods.

This report summarizes the chemical disposal data from facilities operated by Douglas United Nuclear, Inc., during Fiscal Year 1970. The total quantity of chemicals discharged to the Columbia River from the Single-Pass Reactor facilities is detailed in Section I and from the N Reactor facilities in Section II. In addition to the gross quantities shown, supplemental information has been included to note the manner and normal rate of discharge. Section III summarizes the disposal frequencies and estimated quantities of chemicals discharged to the 300 Area process ponds from DUN facilities. These materials are discharged to the process sewer and flow to the 300 Area process ponds. A portion of the chemicals reach the Columbia River through underground seepage. Section III also includes a report of the former Plutonium Recycle Test Reactor water plant. Douglas United Nuclear, Inc., placed this facility in operation in FY 1970 to supply a portion of the area water needs formerly supplied by the City of Richland. As the process sewer system in the 300 Area serves other companies as well, the controls established by DUN include routine composite sampling of the 300 Area process ponds. The automatic sampler collects integrated samples of pond inlet water …

This report presents calculations to reduce the quantity of tritium released to the environment from the Purex Plant.

This report discusses the causes and effects of dust storms in the Tri-Cities area.

This report details activities of Pacific Northwest Laboratory`s Division of Production and Hanford Plant Assistance Programs for the month of March 1970.

This report presents a review of current 100-K Plant liquid waste disposal practices. The report s divided into two sections with Section I consisting of a general description of the Plant facilities to dispose of liquid wastes by ground disposal or by direct discharge to the Columbia River. Section II further appraises liquid waste disposal practices by examining each material that is used in any significant degree in Plant operations. For ease of presentation in Section II, the materials are grouped into general categories of usage with two primary considerations reviewed for each material, i.e., (a) the storage and supply arrangement with emphasis on provisions or safeguards to prevent accidental discharge and (b) the specific application of the material in the process, manner of usage, and final disposition. The report should be viewed as only an informal source of data to aid in the future preparation (by others) of a comprehensive environmental control plan document. Suggestions as to refinements, additions, or corrections are solicited.

None

The purpose of this study is to experimentally determine the thermal hydraulic characteristics of the One-Inch Overbore fuel geometry under both single and two-phase flow conditions. Results of these experiments demonstrate the relation between pressure drop and flow at various tube power levels and will be utilized to establish process tube operating limits and trip margins.

The work described in this volume of ``S`` Projects History is an extension of the type of work described in Volume I. E.I. du Pont de flemours & Company had entered into Contract AT (07-2)-l with the United States Atomic Energy Commission to develop, design, construct, install, and operate facilities to produce heavy water, fissionable materials, and related products. Under this contract,, Du Pont constructed and operated the Savannah River Plant. The engineering, design, and construction for most of the larger ``S`` projects was performed by the Engineering DeDartment. For some of the large and many of the smaller projects the Engineering Department was responsible only for the construction because the Atomic Energy Division (AED) of the Explosives Department handled the other phases. The Engineering Department Costruction Division also performed the physical work for many of the plant work orders. This volume includes a general description of the Du Pont Engineering Department activities pertaining to the engineering, design, and construction of the ``S`` projects at the Savannah River Plant; brief summaries of the projects and principal work requests; and supplementary informaticn on a few subjects in Volume I for which final data was not available at the closing date. Projects …

Per your request, we have summarized for both the single-pass reactors and N Reactor the amount of radioisotopes released in DUN reactor liquid effluents. Wherever possible, the sources of the radioisotope and the important parameters which influence the release of the radioisotope have been identified. Because of the great dissimilarities between the two types of reactors, the data are discussed on an individual basis.

An increase in scandium-46 activity has been suggested as a possible cause of K Reactor rear face dose rates. The first investigations were initiated on {sup 46}Sc in 1968 after Environmental Evaluation, BNW, reported the increase in the {sup 46}Sc activity in the Columbia River. The first investigation was to verify whether or not there had been any increase in parent material ({sup 45}Sc) in the Columbia River. The resulting neutron activation data for river samples taken at Vernita during the latter part of 1968 and the first part of 1969 are compared with data obtained in 1962 by Silker. These data show that there was no increase in parent scandium since 1962. There also appears to be no seasonal increase in parent scandium as occurs with several other elements such as zinc. This latter observation (based primarily on the 1962 data) is quite puzzling since the BNW river data from samples taken near Richland show a definite seasonal trend. The BNW data are shown. Scandium-46 analyses of the reactor effluent water were initiated in March 1969 and these data also fail to show any significant seasonal trend of the magnitude shown by the BNW data. The reactor effluent data …

This reference contains information which details experiments done at the Hanford Reservation on communities of periphyton. These autoradiographic experiments examined the uptake of radioisotopes and diffusion of radioisotopes in the communities. Arguments were presented discussing the presentation of the results on an areal or gravimetric basis.

Sonic tagged, adult chinook and steelhead were tracked in 1967 during upstream migration past cooling water discharges from nuclear reactors. All fish migrated near shorelines, showing preference for the river bank opposite reactors. Clear responses to local temperature differences were exhibited by few fish, these responses being to small shoreline seepages rather than to main center-channel outfalls.

Lattice physics subsystems of the JOSHUA system are described. Information is included on data management, geometry modules, nuclear data modules, and integral transport modules. The JOSHUA system is a set of computer codes for the design of heterogeneous reactors.

Fault tree techniques have been used to identify possible failure paths within the NPTR which could lead to core disassembly. The analysis o f the various faults has led to formulation of design requirements, protective system requirements, and administrative restraints required to prevent accidents from these faults. Transient analyses were performed using the heat transfer-nuclear kinetics codes, Nutiger-II, FORE-II, and MELT-II . To verify results, intercomparison studies were made between the codes. The codes were i n good general agreement. Each code was found to exhibit different advantages and disadvantage. Inherent reactivity feedback effects were assessed in the analysis. With the assumed core parameters, there appears to be sufficient Doppler to prolong a nuclear transient to allow protective action to prevent fuel from melting. The use of average values of the feedback coefficients smeared over the entire core does not appear to be an acceptable method with spacially dependent temperatures. In the thermal analysis, the fuel pin gap coefficient and sodium film coefficient do not appear to be highly sensitive parameters for transient analysis. Power transients resulting from reactivity insertions of from 2$/sec to 20$/sec have been examined in detail. Sodium will be molten before fuel melting occurs for accidents …

The objective of the Reactor Physics Quarterly Report is to inform the scientific community in a timely manner of the technical progress made on the many phases of reactor physics work within the laboratory. The report contains brief technical discussions of accomplishments in all areas where significant progress has been made during the quarter.

Metallurgical examinations and evaluations were performed on a group of four target tubes from Atomic Power Development Associates (APDA) Rig 10. This group included two 2/sup 1///sub 4/ Cr--1 Mo ferritic steel tubes, one Type 304 SS, and one Incoloy 800 tube. These tubes had been exposed to small leaks of water in 600/sup 0/F sodium. Metallographic data indicate that the material loss (wastage) was primarily due to impact (or erosion) of liquid particles. No evidence of corrosion was found in the wasted areas of any target tubes. Minor corrosion was observed on the outside diameter of the austenitic stainless-steel tubes only in areas away from impact; no corrosion was detected on the outside diameter of the ferritic steel tubes. Hardness measurements and microstructural analyses revealed that the maximum temperature may have reached 1600/sup 0/F in the wasted areas. This localized temperature increase appeared to enhance the wastage process for all target tubes regardless of composition. Surface grain elongation and plastic deformation lines were observed in wasted areas, as well as in other areas. However, their cause(s) cannot be ascertained with respect to deformation due to water-jet impact, or deformation due to fabrication and handling, or both. Based on the …

Information on the JOSHUA system for lattice physics measurements is presented concerning leakage modules, edit modules, control modules, lattice depletion modules, lattice Monte Carlo modules.

Iodoacetamide has been shown by others to be a radiation sensitizer for bacteria and for certain mammalian cells tested in vitro. This work describes an examination of the effectiveness of iodoacetamide used in vivo. Survival of ascites tumor cells maintained in the peritoneal cavity of mice was used as an indicator of sensitization. Survival was assessed using TD{sub 50} and total tumor cell population determination methods. A comparison of results obtained by these methods is made. The effects of oxygen tension and radiation dose rate upon results was examined. Iodoacetamide was found to be effective as a radiation sensitizer under all conditions although to a lesser degree than that reported by others for in vitro experiments with bacteria. Radioactive tracer studies indicate that iodoacetamide has rapid and total access to most if not all tissues of the body. This fact coupled with the observation of a sensitization in an in vivo system where the anoxia so prevalent in well developed tumors was present, suggests the possibility of clinical usefulness of iodoacetamide in cancer radiation therapy. Certain observations are reported on the effect of various cell and host treatment procedures upon cell population growth kinetics seen subsequent to inoculation of hosts …

Magnetic susceptibility and neutron diffraction measurements were made on a high purity, polycrystalline sample of ..cap alpha..-phase /sup 242/Pu. The susceptibility results show the existence of weak ferromagnetism below 50/sup 0/K. This suggests that the ..cap alpha..-phase is antiferromagnetic, the observed ferromagnetic moment (9.5 x 10/sup -4/ emu per gram of sample at 4.35/sup 0/K) originating within antiferromagnetic domain walls. No evidence for antiferromagnetic ordering was observed in the neutron diffraction study suggesting that the moment per Pu atom is too small to observe with this technique.