A series of critical experiments was completed with mixed plutonium-uranium solutions having a Pu/(Pu + U) ratio of approximately 0.22 in a boiler tube-type lattice assembly. These experiments were conducted as part of the Criticality Data Development Program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. A complete description of the experiments and data are included in this report. The experiments were performed with an array of mixed oxide fuel pins in aqueous plutonium-uranium solutions. The fuel pins were contained in a boiler tube-type tank and arranged in a 1.4 cm square pitch array which resembled cylindrical geometry. One experiment was perfomed with the fuel pins removed from the vessel. The experiments were performed with a water reflector. The concentration of the solutions in the boiler tube-type tank was varied from 4 to 468 g (Pu + U)/liter. The ratio of plutonium to total heavy metal (plutonium plus uranium) was approximately 0.22 for all experiments.

A community health survey was conducted during the first five months of 1987 in Kalapana, Hawaii. Some 676 residents were interviewed during the study, which represents some 82% of all households in the community. The goal was to obtain base-line data on the health status of all community residents and ambient air quality, in order to evaluate any changes in health status of residents after geothermal development in the area.

The TMI-2 accident resulted in approximately 40% of the reactor's core melting and collecting on the lower head of the reactor pressure vessel. The severity of the accident has raised questions about the margin of safety against rupture of the lower head in this accident since all evidence seems to indicate no major breach of the vessel occurred. Scoping heat transfer analyses of the relocated core debris and lower head have been made based upon assumed core melting scenarios and core material debris formations while in contact with the lower head. This report describes the structural finite element creep rupture analysis of the lower head using a temperature transient judged most likely to challenge the structural capacity of the vessel. This evaluation of vessel response to this transient has provided insight into the creep mechanisms of the vessel wall, a realistic mode of failure, and a means by which margin to failure can be evaluated once examination provides estimated maximum wall temperatures. Suggestions for more extensive research in this area are also provided. 6 refs., 15 figs.

This report covers the last 5 months of the second year of this grant and the first 8 months of the third year. The research thrust of this grant has been directed into two areas. The principal effort has been spent in an investigation of the use of anomalous small angle x-ray scattering (ASAXS) to observe changes in the microstructure of a relatively complex alloy produced by high temperature deformation or aging. The second effort involves a study of the high temperature behavior of several ferritic steels. During this past year we have been examining the effect of environment (air vs vacuum) on the high temperature strength of Fe9Cr1Mo modified by the addition of small amounts of V and Nb.

The US Department of Energy (DOE) has requested that all DOE facilities handling defense transuranic (TRU) waste develop and implement a program whereby all TRU waste will be contained, stored, and shipped to the Waste Isolation Pilot Plant (WIPP) in accordance with the requirements set forth in the DOE certification documents WIPP-DOE-069, 114, 120, 137, 157, and 158. The program described in this report describes how Oak Ridge National Laboratory (ORNL) intends to comply with these requirements and the techniques and procedures used to ensure that ORNL TRU wastes are certifiable for shipment to WIPP. This document describes the program for certification of newly generated (NG) contact-handled transuranic (CH-TRU) waste. Previsions have been made for addenda, which will extend the coverage of this document to include certification of stored CH-TRU and NG and stored remote-handled transuranic (RH-TRU) waste, as necessary. 24 refs., 11 figs., 4 tabs.

Containment plumes at Uranium Mill Tailings Remedial Action (UMTRA) Project sites were ranked to assist in Subpart B (i.e., restoration requirements of 40 CFR Part 192) compliance strategies for each site, to prioritize aquifer restoration, and to budget future requests and allocations. The rankings roughly estimate hazards to the environment and human health, and thus assist in determining for which sites cleanup, if appropriate, will provide the greatest benefits for funds available. The rankings are based on the scores that were obtained using the US Department of Energy's (DOE) Modified Hazard Ranking System (MHRS). The MHRS and HRS consider and score three hazard modes for a site: migration, fire and explosion, and direct contact. The migration hazard mode score reflects the potential for harm to humans or the environment from migration of a hazardous substance off a site by groundwater, surface water, and air; it is a composite of separate scores for each of these routes. For ranking the containment plumes at UMTRA Project sites, it was assumed that each site had been remediated in compliance with the EPA standards and that relict contaminant plumes were present. Therefore, only the groundwater route was scored, and the surface water and air …

Augmented Fish Health Monitoring Contract DE-AI79-87BP35585 was implemented on July 20, 1987. First year highlights included remodeling of the Olympia (WA) Fish Health Center to provide laboratory space for histopathological support services to participating state agencies, acquisition of gas monitoring equipment for hatchery water systems, expanded disease detection work for bacterial kidney disease and erythrocytic inclusion body syndrome in fish stocks at 13 Columbia River Basin National Fish Hatcheries and advancements in computerized case history data storage and analysis. This report summarizes the health status of fish reared at Service facilities in the Columbia River basin, briefly describes work being done to meet contract requirements for fish disease surveillance at those hatcheries and provides a summary of case history data for calendar years 1984, 1985, 1986 and 1987. 1 ref.

This paper briefly describes the following topics: an overview of graphite wall pumping effects; wall pumping experiments in PISCES-A; and non-saturable walls in toroidal devices. 11 figs., 2 tabs. (LSP)

Pacific Northwest Laboratory (PNL) is conducting ground-water monitoring at the Hanford Site, near Richland, Washington. This document contains a data listing of results for monitoring by PNL and Westinghouse Hanford Company (WHC) during January through March 1987. 4 refs.

In response to Section 7(b) of the Low-Level Radioactive Waste Policy Amendments Act of 1985 (Public Law 99-240), this report summarizes the progress of states and low-level radioactive waste compacts in 1987 in establishing new low-level waste disposal facilities. It also reports the volume of low-level waste received for disposal in 1987 by commercially operated low-level waste disposal facilities.

H/sub ..cap alpha../ emission is a useful indicator of hydrogen ionization since the energy required to excite hydrogen to the n = 3 level is approximately the same as to ionize the hydrogen atom. Other aspects of recycling, such as plasma scrape-off parameters, wall particle fluxes, and wall retention rates can be inferred from spatial distributions of H/sub ..cap alpha../ emission. There are two H/sub ..cap alpha../ diagnostics on TFTF observing the inner limiter: a relatively calibrated wide-angle TV camera, and the absolutely calibrated HAIFA diagnostic, both viewing the inner limiter. H/sub ..cap alpha../ observations by these instruments have been analyzed using the DEGAS neutral transport code. The core recycling rate, and ion and neutral particle limiter fluxes, have been deduced from the calculations. The results are very sensitive to the ion flux distribution, and therefore also provide information on the plasma scrape-off conditions. A survey was made of over 500 TFTR discharges. H/sub ..cap alpha../ emission appears to increase in proportion to /bar n//sub e//sup 2/. The ratio of core ionization to average H/sub ..cap alpha../ emission was calculated to be roughly constant over a large range of discharges. 9 refs., 4 figs., 1 tab.

The ATM-WV/205 glass was produced in accordance with PNL's QA Manual for License-Related Programs, MCC technical procedures, and MCC QA Plan that were in effect during the course of this work. The method and procedure to be used in the fabrication and characterization of the ATM-WV/205 glass were specified in two run plans for glass preparation and a characterization plan. The ATM-WV/205 glass meets all specifications. The elemental composition and oxidation state of the glass are within the sponsor's specifications. Visually, the ATM-WV/205 glass bars appear uniformly glassy and generally without exterior features. Microscopic examination and x-ray diffraction revealed low (about 0.5 wt %) concentrations of 3-..mu..m iron chrome spinel crystals and 1-..mu..m ruthenium inclusions scattered randomly throughout the glassy matrix. Closed porosity, with pores ranging in diameter from 20 to 135 ..mu..m, was observed in all samples. 3 refs., 10 figs., 21 tabs.

The primary objective of this study was to determine the effects of different contact gases and experimental conditions on product yields from fluid-bed wood pyrolysis. The yield of condensible hydrocarbons was of particular interest. The contact gases employed included carbon dioxide (CO/sub 2/), methane (CH/sub 4/), and hydrogen (H/sub 2/). Reaction temperatures ranged from 650/degree/C to 800/degree/C, and gas contact time was between 1 and 2 s. Tests with CO/sub 2/ contact gas used a secondary vessel of catalyst for cracking of condensibles generated in the primary gasifier. Results of earlier studies at PNL showed the effectiveness of various catalysts for destruction of condensible hydrocarbons. Analyses of results on the effects of contact gas, reaction temperature, gas contact time, and wood feed rate are presented in this report, along with a description of experimental equipment and methods. Data were obtained from single tests at each set of experimental conditions. Conditions were not repeated to determine the amount of experimental error. Material balance closures were generally quite good for the experimental tests, within 10%. 9 refs., 15 figs., 16 tabs.

Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of high-level radioactive-waste disposal containers. The waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The copper-based alloy materials are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The austenitic materials are Types 304L and 316L stainless steels and Alloy 825. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr, and they must be retrievable from the disposal site during the first 50 yr after emplacement. The containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This volume surveys the available data on the phase stability of both groups of candidate alloys. The austenitic alloys are reviewed in terms of the physical metallurgy of the iron-chromium-nickel system, martensite transformations, carbide formation, and intermetallic-phase precipitation. The copper-based alloys are reviewed in terms of their phase equilibria and the possibility of precipitation of the minor alloying constituents. For the austenitic …

This report is compiled by the Nuclear Waste treatment Program and the Hanford Waste Vitrification Program at Pacific Northwest Laboratory to describe the progress in developing, testing, applying and documenting liquid-fed ceramic melter vitrification technology. Progress in the following technical subject areas during the quarter of FY 1987 is discussed: melting process chemistry and glass development; feed preparation and transfer systems, melter systems; and canister filling and handling systems. 4 refs., 20 figs., 18 tabs.

This report presents an analytical method for determining wellbore heat transmission during liquid or gas flow along the tubing. The mathematical model describes the heat transfer between the flowing fluid in the wellbore and in the surrounding formation as one whole physical system. The transient heat transfer equations in the two regions with coupling at the sandface are solved simultaneously. Non-homogeneous formations are treated which consist of several layers with different physical properties and arbitrary initial temperature distributions in the vertical direction. Closed form analytical solutions are obtained in real space and in Laplace space, which can be used to calculate the temperature distribution along the wellbore and in the formation, and to evaluate heat transfer rate and cumulative heat exchange between wellbore and formation. A more accurate formula is given for the widely-used transient heat conduction function f(t{sub D}) of thermal resistance. 14 refs., 12 figs., 1 tab.

This report presents the results of studies addressing several problems in the head-end processing (decladding, metathesis, and core dissolution) of N Reactor fuel elements in the Hanford PUREX plant. These studies were conducted over 2 years: FY 1986 and FY 1987. The studies were divided into three major areas: 1) differences in head-end behavior of fuels having different histories, 2) suppression of /sup 106/Ru volatilization when the ammonia scrubber solution resulting from decladding is decontaminated by distillation prior to being discharged, and 3) suitability of flocculating agents for lowering the amount of transuranic (TRU) element-containing solids that accompany the decladding solution to waste. 16 refs., 43 figs.

This report describes the modifications that have been made to enhance the original Allowable Residual Contamination Level (ARCL) method to make it more applicable to site-specific analyses. This version considers the mixture of radionuclides present at the time of site characterization, the elapsed time after decommissioning when exposure to people could occur, and includes a calculation of the upper confidence limit of the potential dose based on sampling statistics that are developed during the site characterization efforts. The upper confidence limit of potential exposure can now be used for comparison against applicable radiation dose limits (i.e., 25 mrem/yr at Hanford). The level of confidence can be selected by the user. A wide range of exposure scenarios were evaluated; the rationale used to select the most limiting scenarios is explained. The radiation dose factors used for the inhalation and ingestion pathways were also updated to correspond with the radiation dosimetry methods utilized in the International Commission of Radiological Protection Publications 26 and 30 (ICRP 1977; 1979a,b, 1980, 1981, 1982a,b). This simplifies the calculations, since ''effective whole body'' doses are now calculated, and also brings the dosimetry methods used in the ARCL method in conformance with the rationale used by DOE in …

We propose a new method for modeling the sawtooth instability and other MHD activity in axisymmetric tokamak transport simulations. A hyper-resistivity (or current viscosity) term is included in the mean field Ohm's law to describe the effects of the three-dimensional fluctuating fields on the evolution of the inverse transform, q, characterizing the mean fields. This term has the effect of flattening the current profile, while dissipating energy and conserving helicity. A fully implicit MHD transport and 2-D toroidal equilibrium code has been developed to calculate the evolution in time of the q-profile and the current profile using this new term. The results of this code are compared to the Kadomtsev reconnection model in the circular cylindrical limit. 17 refs., 8 figs.

Six alloys are being considered as possible materials for the fabrication of containers for the disposal of high-level radioactive waste. Three of these candidate materials are copper-based alloys: CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The other three are iron- to nickel-based austenitic materials: Types 304L and 316L stainless steels and Alloy 825. Radioactive waste will include spent-fuel assemblies from reactors as well as waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, the containers must be retrievable from the disposal site. Shortly after emplacement of the containers in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This radiation will promote the radiolytic decomposition of moist air to hydrogen. This volume surveys the available data on the effects of hydrogen on the six candidate alloys for fabrication of the containers. For copper, the mechanism of hydrogen embrittlement is discussed, and the effects of hydrogen on the mechanical properties …

Photoinduced ir absorptions with much longer lifetimes and lower excitation energies than previously reported were observed after the photolysis of polyacetylene with upsilon > 5000 cm/sup -1/. These absorptions are similar to those previously reported to have millisecond lifetimes, and are attributed to charged bond-alternation defects. The persistent photoinduced absorptions could be depleted by photoexcitation of a transition with an absorption maximum near 4000 cm/sup -1/. Unpaired spins were observed to develop and be depleted at the same photolysis frequencies as the ir activity. These spins may be due to the charged defects. This would mean that the defects are charged polarons. The temperature required to dope polyacetylene chemically was measured for various dopants. This was taken to be a measure of the activation energy for the doping reaction. The activation energies were found to be much lower than predicted by stepwise reaction mechanisms. Photolysis with above-band-gap light did not alter the temperature required for reaction. The chemical stabilization of photoinduced defects does not occur at a perceptible rate at lower temperatures than the chemical doping reaction. 35 refs., 26 figs., 14 tabs.

A comprehensive evaluated neutronic data file for elemental cobalt is described. The experimental data base, the calculational methods, the evaluation techniques and judgments, and the physical content are outlined. The file contains: neutron total and scattering cross sections and associated properties, (n,2n) and (n,3n) processes, neutron radiative capture processes, charged-particle-emission processes, and photon-production processes. The file extends from 10/sup /minus/5/ eV to 20 MeV, and is presented in the ENDF/B-VI format. Detailed attention is given to the uncertainties and correlations associated with the prominent neutron-induced processes. The numerical contents of the file have been transmitted to the National Nuclear Data Center, Brookhaven National Laboratory. 143 refs., 16 figs., 5 tabs.

This study was conducted to determine the parameters governing factors of scale for liquid-fed ceramic melters (LFCMs) in order to design full-scale melters using smaller-scale melter data. Results of melter experiments conducted at Pacific Northwest Laboratory (PNL) and Savannah River Laboratory (SRL) are presented for two feed compositions and five different liquid-fed ceramic melters. The melter performance data including nominal feed rate and glass melt rate are correlated as a function of melter surface area. Comparisons are made between the actual melt rate data and melt rates predicted by a cold cap heat transfer model. The heat transfer model could be used in scale-up calculations, but insufficient data are available on the cold cap characteristics. Experiments specifically designed to determine heat transfer parameters are needed to further develop the model. 17 refs.

An accurate tracing of the thyristor thermal response is investigated. This paper offers several alternatives for thermal modeling and analysis by using an electrical circuit analog: topological method, convolution integral method, etc. These methods are adaptable to numerical solutions and well suited to the use of the digital computer. The thermal analysis of thyristors was performed for the 1000 MVA converter system at the Princeton Plasma Physics Laboratory. Transient thermal impedance curves for individual thyristors in a given cooling arrangement were known from measurements and from manufacturer's data. The analysis pertains to almost any loading case, and the results are obtained in a numerical or a graphical format. 6 refs., 9 figs.