The Morgantown Energy Technology Center (METC) is developing the technology required to recover the deep thin seam Eastern bituminous coal resource by gasification in-situ. The approach is to prove concepts through field tests and to support field testing-with theoretical modeling. METC is currently fielding Pricetown I, the first of two tests scheduled for the Pricetown, West Virginia, underground coal gasification field test. Pricetown I is a small-scale test designed to provide information concerning the in-situ characteristics of the Pittsburgh coal seam; to gain additional experience in the in-situ combustion and gasification of bituminous coal; and to evaluate the functional applicability of the linked vertical concept to recover the Eastern resources. Mound Facility is participating with METC in the design and the implementation of the instrumentation necessary to monitor the surface and subsurface process and product gas stream; and acquire real-time gas analysis and subsurface thermal data. The principal objective of this effort is to provide an integrated instrumentation system that will permit rapid automatic monitoring of subsurface and surface variables and to ensure data storage, retrieval and reduction for process monitoring and results interpretation. Mound also will support METC with the manpower and technical assistance necessary to operate the field …

The Savannah River Laboratory (SRL) is currently involved as the focal point of an inter-laboratory effort to estimate the enviornmental effects of operating a nuclear preprocessing facility. As a part of this effort SRL has assumed the responsibility of providing estimates of population exposure resulting form radioactive discharges to the atmosphere. This responsibility does not include contributions from drinking water and food ingestion which are being estimated by Hanford Engineering Development Laboratory (HEDL) and Lawrence Livermore Laboratory (LLL), two of the participating laboratories for the overall environmental effects study. The SRL effort in exposure estimation is confined to contributions resulting from immersion in contaminated air and external beta and gamma dose from exposure to contaminated ground. Population exposure resulting from immersion in contaminated air traditionally have been given more attention. A primary objective of this report is to develop a technique to provide an adequate estimation of population dose resulting from gound depositions and to establish the relative importance of the contributing factors. This report was written in 1978, and has been identified as information for the Dose Reconstruction Project.

Monsanto Research Corporation operates Mound Facility, a government-owned facility of the U.S. Department of Energy (DOE) in Miamisburg, Ohio. Mound Facility is an integrated research, development, and production facility performing work in support of DOE weapon and nonweapon programs with emphasis on explosive and nuclear technology. This report provides information on the environmental monitoring of the Mound plant for 1977.

This report discusses various aspects of the Mound Laboratory including; who they are, what they do, their technological capabilities, and their resources.

Using the transient method, we determine the permeability of high-quartz Eleana argillite from the Nevada Test Site as a function of effective pressure. By comparing calculated and observed pressure decay in the upstream reservoir, we have determined the permeability of intact and fractured specimens at effective pressures ranging from 1.0 to 24.0 MPa. Over this pressure range, Eleana argillite has a low permeability (10{sup -16} to 10{sup -19} cm{sup 2}) when intact and a higher permeability (10{sup -12} to 10{sup -17} cm{sup 2}) with one induced through-going fracture.

Results of preliminary study are presented which was performed under subtask 2.7 of the NTS Terminal Waste Storage Program Plan for 1978. Subtask 2.7 examines the feasibility of locating a nuclear waste repository in a granitic stock or pluton in southern Nevada near the Nevada Test Site (NTS). It is assumed for the purposes of this study that such a repository cannot be located at NTS. This assumption may or may not be correct. This preliminary report does not identify a particular site as being a suitable location for a repository. Nor does it absolutely eliminate a particular site from further consideration. It does, however, answer the basic question of probable suitability of some of the sites and present a systematic method for site evaluation. Since the findings of this initial study have been favorable, it will be followed by more exhaustive and detailed studies of the original 30 sites and perhaps others. In future studies some of the evaluation criteria used in the preliminary study may be modified or eliminated, and new criteria may be introduced.

The interim draft report containing the first stage of CSC`s work on prediction of subsurface ground motion was completed. Twenty-four stations of the seismic monitoring network are now operational. The location for the first exploratory hole in Calico Hills was changed based on interpretation of magnetic and electrical geophysical data. Two core holes were completed in the Climax Stock in the Pile Driver tunnel complex. Drilling on the first exploratory hole at Yucca Mountain commenced on July 30, 1978. Field reconnaissance of granitic rocks in southern Nevada continued, including locations in Esmeralda, Nye, and White Pine Counties. The modeling of the Eleana Heater Experiment showed good agreement with field temperature data for conduction energy transfer. A rough draft of the tuff scoping report was completed. Review of the LASL quality program plan for their activities on the NTS Terminal Waste Storage Program was completed by Sandia Quality Assurance. A geological reconnaissance of the region near the Yucca Mountain drill site suggested a high probability that large, reasonably unfaulted blocks of tuff exist in the area.

The technology and data base required to license a nuclear repository in a crystalline rock medium, located at or near the Nevada Test Site are being developed. The program consists of three related project areas: field and laboratory studies of the availability and migration of radionuclides in ground water; thermomechanical response of granite, through heater tests at the Climax stock of the Nevada Test Site; and laboratory measurements of physical properties of rocks at elevated temperatures and pressures, including physical/chemical factors that inhibit water transport in deep silicate rocks. Work accomplished in these areas is report. (LK)

The interactions of a quartz monzonite, an argillite, an alluvium, and several tuffs with various radionuclides in selected phreatic waters have been studied. The sorption--desorption hehavior of Sr, Tc(VII), Cs, Ba, Ce, Eu, U(VI), Pu, and Am under ambient and 70{sup 0}C temperature conditions has been measured.

Volume 1 contains the following papers: Solution Species of {sup 239}Pu in Oxidizing Environments; Solution Species of {sup 239}Pu in the Environment; Theoretical and Experimental Evaluation of Waste Transport in Selected Rocks; Studies of Radionuclide Availability and Migration at the Nevada Test Site Relevant to Radioactive Waste Disposal; Systematic Study of Metal Ion Sorption on Selected Geologic Media; Chromatographic K/sub d/ values of Radionuclides; Effects of Redox Potentials on Sorption of Radionuclides by Geologic Media; and Transport Properties of Nuclear Waste in Geologic Media. Individual papers were processed.

Full-scale near-surface heater experiments are presently being conducted by Sandia Laboratories in the Conasauga Formation at Oak Ridge, Tennessee, and in the Eleana Formation on the Nevada Test Site, Nevada. The purposes of these experiments are: (1) to determine if argillaceous media can withstand thermal loads characteristic of high level waste; (2) to provide data for improvement of themomechanical modeling of argillaceous rocks; (3) to identify instrumentation development needed for further in situ testing; and (4) to identify unexpected general types of behavior, if any. The basic instrumentation of these tests consists of a heater in a central hole, surrounded by arrays of holes containing various instrumentation. Temperatures, thermal profiles, vertical displacements, volatile pressurization, and changes in in situ stresses are measured in each experiment as a function of time, and compared with pretest modeling results. Results to date, though in general agreement with modeling results assuming conductive heat transfer within the rock, indicate that the presence of even small amounts of water can drastically affect heat transfer within the heater hole itself, and that small amounts of upward convection of water may be occurring in the higher temperature areas of the Conasauga experiments.

This report contains a discussion of an instrumentation and rock mechanics program recommended for consideration as part of the overall Lawrence Livermore nuclear waste storage program at NTS. It includes a discussion of (1) rationale for the heater tests, spent fuel facility evaluation, heated room tests, (2) recommended instrumentation types together with estimated delivery schedules, (3) recommended instrumentation layouts, (4) other proposed rock mechanics tests both laboratory and in situ, and (5) data acquisition and reduction requirements.

Sandia`s Particle Beam Fusion Program is investigating several driver options, based on pulsed power technology, with the goal of demonstrating a practical ignitor for Inertial Confinement Fusion (ICF) Reactors. The interrelated aspects of power conditioning and compression, beam-target interaction, and target ignition are being studied. The issues of efficiency, reliability and multiple pulse capability are being integrated into the program to provide a viable approach to an experimental power reactor. On a shorter time scale the authors expect to derive important military-related benefits from attendant research and facility development. The two most important advantages of pulsed power driven fusion are the inherent low cost and high efficiency of high current particle accelerators. However, comparison of the relative merits of particle beams and focused laser beams must include many other factors such as beam transport, and target coupling, as well as target design and fabrication. These issues are being investigated to determine if the perceived practical benefits of particle beam fusion can indeed be realized. The practical considerations are exemplified in a comparison of the leading ICF drivers. The plan being followed by Sandia involves using the Electron Beam Fusion Accelerator (EBFA) to meet three objectives by 1985: significant burn using …

A unique feature of the shear system is the requirement for a removable master tool module and tool submodule assembly. Each tooling submodule is designed specifically to meet the compaction, gagging, and shearing requirements for specific spent fuel assemblies. These submodules are interchangeable within a common master tool module housing. The cross section of the spent fuel assemblies range from a 4.575-in. hexagonal shrouded to a 8.75-in. square grid-unshrouded. A number of interrelated initial design problems had to be solved: a hydraulic force system arrangement with minimum spacing between cylinders; readily removable couplings between the force system and associated tooling submodule; couplings with maximum stiffness and minimum geometry; a roller bearing system for each tool submodule and hydraulic stem assembly; and long life tool operation under high loads, wear, high temperaure, and corrosive conditions. It was established that the cylinder arrangement should consist of three 200-ton, 10-in. dia tandem cylinders, and one 100-ton, 10-in. dia standard cylinder at 3000 psi operating pressure. The shear would be operated by two - 200 ton tandem cylinders in a vertical straddle mode about the horizontal center line of the tooling arrangement. The compactor would be operated by one - 200 ton tandem cylinder …

A series of exploratory hot-cell tests was made to determine the effects of voloxidation on tritium, fission product removal, and on subsequent steps of the light water reactor fuel cycle. The 100-g scale experiments indicated that >99% of the tritium, 50% of the {sup 14}C, 6% of the {sup 85}Kr, and smaller amounts of other elements were volatilized and collected when the UO{sub 2} was roasted in air at 480{sup 0}C for {similar_to} h. There was little effect on the solubility of the uranium and plutonium in HNO{sub 3}. The experiments indicated that the first 2-h leach usually dissolved >99.9% of the uranium and plutonium, and varying amounts of fission products, in voloxidized or unvoloxidized fuel. Two additional HNO{sub 3} leaches on the unvoloxidized fuel dissolved all but 0.004% of the plutonium; the additional leaches on the voloxidized material left {similar_to}.02% of the plutonium. (28 figs., 21 tables)

This is the ninth in a series of quarterly progress reports on Fuel Cycle Technologies/Thorium Fuel Cycle Technologies/International Spent Fuel Receipt and Storage (AFCT/TFCT/ISFS) program. This program provides information needed by industry to close the back end of the power reactor fuel cycle. Included in the program are activities supporting specific design studies, as well as activities for general fuel cycle technology. Studies were conducted in the following tasks: thorium resource price analyses; investigation of air cleaning processes for removing TBP from off-gas streams; study of iodine chemistry in process solutions; high-level waste treatment; electropolishing to decontaminate metallic waste from alternate and thorium converter fuel cycles; U.S. scale transport, dispersion and removal model comparison; safety criticality experiments; and criticality research in support of thorium fuel cycle technology program. (21 figs., 7 tables) (DLC)

This is the last in a series of quarterly progress reports on research and development studies performed for the Alternate Fuel Cycle Technologies/Thorium Fuel Cycle Technologies/International Spent Fuel Receipt and Storage (AFCT/TFCT/ISFS) Program. This program provided information needed by industry to close the back end of the power reactor fuel cycle. During the last quarter. studies were conducted on the following task: Thorium Resource Price Analyses; Investigation of Air Cleaning Processes for Removing Tributyl Phosphate from Off-Gas Streams; Study of Iodine Chemistry in Process Solutions; High-Level Waste Treatment; Electropolishing to Decontaminate Metallic Waste from Alternate and Thorium Converter Fuel Cycles; US Scale Transport, Dispersion and Removal Model Comparison; Safety Criticality Experiments; and Criticality Research in Support of Thorium Fuel Cycle Technology Program.

An exploratory development contract was undertaken on December 23, 1977 which had as its purpose the development and demonstration of a flexible armored blanket design suitable for providing ballistic protection to nuclear weapons during shipment. Objectives were to design and fabricate a prototype blanket which will conform to the weapon shape, is troop-handleable in the field, and which, singly or in multiple layers, can defeat a range of kinetic energy armor piercing (AP) ammunition potentially capable of damaging the critical portion of the nuclear weapon. Following empirical testing, including the firing of threat ammunition under controlled laboratory and field test conditions, materials were selected and assembled into two blanket designs, each weighing approximately 54 kg/m{sup 2} (11 lbs/ft{sup 2}) and estimated to cost from $111 to $180 per ft{sup 2} in production. A firing demonstration to evidence blanket performance against terrorist/light infantry weapons, heavy infantry weapons, and aircraft cannon was conducted for representatives of the DOD and interested Sandia employees on April 12, 1978. The blankets performed better than anticipated defeating bullets up to 7.62 mm x 51 mm AP with one layer and projectiles up to 23 mm HEI with two layers. Based on these preliminary tests it is …

The basis processes for reprocessing thorium based light water reactor type fuels are defined for use in identifying criticality data needs. The Reference Thorium Fuel Cycle is used as the primary fuel cycle. Material forms and compositions are described for each major processing step. These forms consist of nitrates and oxides of Th--U--Pu combinations. Fuel fabrication and fuel pool storage facilities are also defined to the extent they interact with fuel reprocessing.

Voloxidation and dissolution studies: rotary-kiln heat-transfer tests are under way using a small rotary kiln along with the development of a mathematical model to determine kiln-heat-flux profiles necessary to maintain a desired temperature gradient. The erosion/corrosion test for evaluating materials of construction is operational. Fuel from a BWR (Big Rock Point) yielded more fine solid residue on dissolution than in previous tests with PWR fuel. Two additional parametric voloxidation tests with H.B. Robinson fuel compared air vs pure oxygen atmospheres at 550{sup 0}C; overall tritium release and subsequent fuel dissolution were equivalent. Thorium dissolution studies: the dissolution rate of thoria in fluoride-catalyzed 8 to 14 M HNO{sub 3} (100{sup 0}C) was max between 0.04 to 0.06 M HF; at higher fluoride concentrations, ThF{sub 4}.5H{sub 2}O precipitated. The rate of zircaloy dissolution continued to increase with increasing fluoride concentration. Stainless-steel-clad (Th,U)0{sub 2} fuel rods irradiated in the NRX reactor were sheared, voloxidized, and dissolved. {le}10% of the tritium was released during voloxidation in air at 600{sup 0}C. Carbon-14 removal from off-gas and fixation: carbon dioxide removal with Linde 13X molecular sieves to less than 100 ppB was experimentally verified using 300 ppM CO in air. Decontamination factors from 3000 to 7500 …

This is the tenth in a series of quarterly progress reports on studies performed for the Alternate Fuel Cycle Technologies/Thorium Fuel Cycle Technologies/International Spent Fuel Receipt and Storage (AFCT/TFCT/ISFS) Program. This program provides information needed by industry to close the back end of the power reactor fuel cycle. During the past quarter, studies were conducted in the following tasks: thorium resource price analyses; spent fuel receipt and storage; investigation of air cleaning processes for removing TBP from off-gas streams; study of iodine chemistry in process solutions, high-level waste treatment; electropolishing to decontaminate metallic waste from alternate and thorium converter fuel cycles; U.S. scale transport, dispersion and removal model comparison; safety criticality experiments; and criticality research in support of thorium fuel cycle.

The Plutonium(IV) Oxalate Precipitation and Calcination Process for converting plutonium nitrate to plutonium oxide is described for a 100-kg plutonium per day (Pu/day/ throughput facility. Block flow diagrams, equipment flowsheets, and stream material balances are included. Advantages and disadvantages of the process, additional research and development necessary, and history of the process are also discussed. This report is one of a series describing various processes for converting plutonium nitrate to oxide. This information in this report should be used when comparing the various processes, and as a starting point for development of a prototype or plant-scale facility.

This is the eighth in a series of quarterly progress reports on studies performed for the Alternate Fuel Cycle Technologies/Thorium Fuel Cycle Technologies/International Spent Fuel Receipt and Storage (AFCT/TFCT/ISFS) Program, formerly the LWR Fuel Recycle Program. This program is designed to provide information needed by industry to close the back end of the power reactor fuel cycle. During the past quarter, studies were conducted in the following tasks: Survey of Current Technology of Fuel Handling Techniques; Investigation of Air Cleaning Processes for Removing Tributyl Phosphate (TBP) from Off-Gas Streams; Study of Iodine Chemistry in Process Solutions; Electropolishing to Decontaminate Metallic Waste from Alternate and Thorium Converter Fuel Cycles; and U.S. Scale Transport, Dispersion and Removal Mode Comparison Safety Criticality Experiments. (11 figs., 7 tables)

This report is a summary of work performed for the U.S. Arms Control and Disarmament Agency under Contract AC7NC122. The work is directly related to the Agency effort to examine potential alternative fuel cycles that might enhance uranium resource utilization, minimize plutonium production, and reduce the need for reprocessing of spent light water reactor (LWR) fuel. Two alternative fuel cycles are examined in this report: (1) the possible use of metallic fuel in LWR fuel elements, and (2) the use of mixture of light water and heavy water as coolant/moderator in a pressurized water reactor. In both cases, calculations of survey nature were performed to determine the basic characteristics of the two cycles -- reactivity, fuel depletion isotopics, uranium resource requirements and plutonium production. Results of these calculations are presented, together with a discussion of the state-of-the-art and design considerations relative to the utilization of metallic fuel.