The purpose of this project is threefold. First to provide a reliable and more continuous source of electrical power to the Operational Support Building-West (OSW) in order to better serve Mound financial and CAD computer facilities. This increased reliability shall be accomplished by the installation of a tie breaker that will connect the 480 volt secondary of the OSW substation with the 480 volt secondary of B substation, within 200 feet of the building. The OSW substation is in the penthouse of the four (4) story OSW building. The B substation is outside along the plant roadway. Installation of the tie breaker will also permit work to be done on the primary (12,470 volt) side of the transformer without a shutdown of the building. Secondly, the replacement of the OSW substation PCB transformer shall be done in order to eliminate the risk of a fire spreading PCB vapors to the building ventilation system. Thirdly, the replacement of the switchboard on the 480 volt secondary of the OSW substation with more reliable breakers. Also, the existing switches do not allow expansion of the 480 volt supply to building equipment. One of the new breakers shall supply power to a new power …

The Mt. Taylor volcanic field and Lucero uplift of west-central New Mexico occur in a transitional-boundary zone between the tectonically active Basin-and Range province (Rio Grande rift) and the less tectonically active Colorado plateau. The general geomorphology and Cenozoic erosional history has been discussed primarily in terms of a qualitative, descriptive context and without the knowledge of lithospheric processes. The first discussion of geomorphic surfaces suggested that the erosional surface underlying the Mt. Taylor volcanic rocks is correlative with the Ortiz surface of the Rio Grande rift. In 1978 a study supported this hypothesis with K-Ar dates on volcanic rocks within each physiographic province. The correlation of this surface was a first step In the regional analysis of the boundary zone; however, little work has been done to verify this correlation with numerical age dates and quantitatively reconstruct the surface for neotectonic purposes. Those geomorphic surfaces inset below and younger than the ``Ortiz`` surface have been studied. This report provides a summary of this data as well as unpublished data and a conceptual framework for future studies related to the LANL ISR project.

We have studied the solubilities of neptunium, plutonium, and americium in J-13 groundwater from Yucca Mountain (Nevada) at three temperatures and hydrogen ion concentrations. They are 25{degree}, 60{degree}C, and 90{degree}C and pH 5.9, 7.0, and 8.5. The results for 25{degree}C are from a study which we did during FY 1984. We included these previous results in the tables to give more information on the solubility temperature dependence; they were, however, done at only one pH (7.0). The solubilities were studied from oversaturation. The nuclides were added at the beginning of each experiment as NpO{sub 2}{sup +}, Pu{sup 4+}, and Am{sup 3+}. The neptunium solubility decreased with increasing temperature and with increasing pH. The soluble neptunium did not change oxidation state at steady state. The pentavalent neptunium was increasingly complexed by carbonate with increasing pH. All solids were crystalline and contained carbonate, except the solid formed at 90{degree}C and pH 5.9. We identified this solid as crystalline Np{sub 2}P{sub 5}. The 25{degree}C, pH 7 solid was Na{sub 3}NpO{sub 2}(CO{sub 3}){sub 2} {center_dot} nH{sub 2}O. Plutonium concentrations decreased with increasing temperature and showed no trend with pH. Pu(V) and Pu(VI) were the dominant oxidation states in the supernatant solution; as the amount …

The Yucca Mountain site in Nevada has been designated by the Nuclear Waste Policy Act of 1982, as amended, for detailed study as the candidate site for the first US geologic repository for spent nuclear fuel and high-level radioactive waste. The detailed study --- called ``site characterization`` --- will be conducted by the Department of Energy (DOE) to determine the suitability of the site for a repository and, if the site is suitable, to obtain from the Nuclear Regulatory Commission authorization to construct the repository. As part of the site characterization study, DOE has prepared a Site Characterization Plan (SCP) for the Yucca Mountain site. The Site Characterization Plan is a nine-volume document, approximately 6300 pages in length, which describes the activities that will be conducted to characterize the geologic, hydrologic, and other conditions relevant to the suitability of the site for a repository. Part 1 of this Handbook explains what site characterization is and how the Site Characterization Plan (Plan) relates to it. Part 2 tells how to locate subjects covered in the Plan. Another major purpose of this Handbook is to identify opportunities for public involement in the review of the Site Characterization Plan. DOE wants to be …

The OSIRIS program is an imaging optical simulation program which has been developed to predict the output of space-borne sensor systems. The simulation is radiometrically precise and includes highly realistic laser, atmosphere, and earth background models, as well as detailed models of optical components. This system was developed by Rockwell Power Services for the Los Alamos National Laboratory. It is based upon the LARC (Los Alamos Radiometry Code, also by Rockwell), and uses a similar command structure and 3d coordinate system as LARC. At present OSIRIS runs on the Cray I computer under the CTSS operating s stem, and is stored in the OSIRIS root directory on LANL CTSS mass storage.

The objective of this project is to construct and operate an integrated test circuit for the Molten-Caustic-Leaching (Gravimelt) process for desulfurization and demineralization of coal to prove process economics assumptions, deliver product coal and to test process conditions aimed at significantly lowering costs. The test circuit consists of six unit operations which together provide a continuous system for leaching coal and regenerating the reactant. These units are: (a) a kiln for reacting molten caustic with coal; (b) a seven stage water washing section for recovering caustic from the coal; (c) a three-stage acid washing section for removing the last traces of metals and alkali and providing an ultra pure coal product; (d) a water treatment section to provide either dischargeable or recyclable water; (e) a regeneration section to provide purified aqueous caustic; and (f) an evaporator section to provide molten-caustic for recycle to the kiln reactor. The integrated test circuit facility contains more than 160 pieces of equipment including filters, centrifuges, tanks, reactors, feeders and the kiln and rising film evaporator. It occupies 3700 square feet and is fitted with more than 6000 feet of piping, 425 valves, 80 instruments and controls as well as a control room with computer …

The following report marks the third quarter of the third phase of the centrifugal slurry pump improvement program. The program was begun in 1982 for the purpose of improving the operating life of centrifugal slurry pumps for coal liquefaction service. This phase of work will verify the design of a pump at higher speed operation. Eventual scale-up of the prototype slurry pumps to full-scale synthetic fuel generation plants could require ten times the flow. The higher speed will allow pumps to be smaller with respectable efficiencies. Conversely, without increasing the specific speed of the pump design, the eventual size would be more than triple that of the prototype slurry pump. The prototype slurry pump during this phase of the program incorporated all the features proven in the earlier phases of the program. This new, higher specific speed pump will be tested for the ability of the hydraulic design to inhibit wear. It will be tested and compared to the previous optimum prototype slurry pump of this program.

Electron beam weld development studies were performed for both the platinum frit vent-to-vent cap weld and also the vent cap-to-body weld for the LWRHU Project using a Hamilton Standard EBW-6 Electron Beam Welder. A total of six (6) development welds each was performed to establish welding parameters and procedures which would produce satisfactory and acceptable welds. The relatively small size of the platinum frit vent dictated that the frit-to-vent cap weld would have to be limited as to depth of penetration and also to minimize the reduction of the porous frit areas.

Brief summaries are given for DIII-D Research Operations covering the following areas: beta and stability; confinement; boundary physics; electron cyclotron heating; ion Bernstein wave heating; current drive; tokamak operations; neutral beam operations; ECH operations; ICH operations; computer data systems; program development; and hardware development. The progress summaries on the International Cooperation task are given for the Tora Supra, HIDEX -- Nagoya Tokamak Experiment, ASDEX, JET, JFT-2M, CHS, and JT-60. Finally a brief summary of progress on the CIT physics task is given.

This report of the Science Panel supplements an earlier document (included), which summarized the Panel`s views on the September, 1984, draft report of the National Institutes of Health Ad Hoc Working Group to Develop Radioepidemiological Tables. Although the Orphan Drug Act requires that tables be produced for a range of dose from one millirad to 1000 rad, the Working Group chose to make no estimates below one rad. The Science Panel is in agreement with this judgment because in the low-dose range there is little empirical evidence of a carcinogenic effect in humans. In order for the Tables to be applicable, a person must have both one of the cancers listed in the Tables and a previous exposure to ionizing radiation. Therefore, the Tables should not be applied to determine the likelihood that a person having received a specific radiation dose will have such a cancer. The Panel`s previous recommendations were primarily concerned with three issues, use of Quality Factors, use of tables for exposures from internally deposited radionuclides, and treatment of uncertainties. In the final report, the Working Group has dispensed with the use of Quality Factors in treating the risks from internally deposited alpha emitters and has substantially …

The Guidance for UMTRA Project Surveillance and Maintenance describes the procedures that will be used to verify that Uranium Mill Tailings Remedial Action (UMTRA) Project disposal sites continue to function as designed. The approach of this guidance document is to identify surveillance requirements and maintenance procedures that will be used to comply with NRC license requirements. This document addresses five primary activities: Definition and characterization of final site conditions. Site inspections; Ground-water monitoring; Aerial photography; and Custodial maintenance and contingency repair. Final site conditions will be defined and characterized prior to the completion of remedial actions at a site. As-built drawings will be compiled, a final topographic survey will be performed, a vicinity map will be prepared, and ground and aerial photographs will be taken. Survey monuments, site markers, and signs will be established as will a network of monitoring wells.

This radiologic characterization of the inactive uranium millsite at Mexican Hat, Utah, was conducted by Bendix Field Engineering Corporation foe the US Department of Energy (DOE), Grand Junction Project Office, in response to and in accord with a Statement of Work prepared by the DOE Uranium Mill tailings Remedial Action Project (UMTRAP) Technical Assistance Contractor, Jacobs Engineering Group, Inc. the objective of this project was to determine the horizontal and vertical extent of contamination that exceeds the US Environmental Protection Agency (EPA) standards at the Mexican Hat site. The data presented in this report are required for characterization of the areas adjacent to the Mexican Hat tailings piles and for the subsequent design of cleanup activities. Some on-pile sampling was required to determine the depth of the 15-pCi/g Ra-226 interface in an area where wind and water erosion has taken place.

The ZPPR-17 assembly was part of the JUPITER-III cooperative program between the US DOE and PNC of Japan. The assembly was designed to study the neutronic behavior of a large, axially heterogeneous, liquid-metal-cooled reactor. The unique feature of the assembly was an internal blanket region in the axial center of the core extending two-thirds of the core radius. Assembly variants with 25 control rod positions and with 13 half-inserted control rods were built. This report describes in detail the results of measurements and analyses of ZPPR-17. The measurements emphasized reaction rate distributions, and gamma dose measurements were included. Additional measurements were control worth, sodium void worth, and reactivity coefficients associated with small material motions in assembly expansion and bowing.

Further results are presented from the JUPITER-III program. Calculation models and k-effective results are given for the three configurations of the large, homogeneous assembly ZPPR-18. Reaction rate results, including sodium activation are given for ZPPR-18A. Also included are spatial decoupling results from ZPPR-18. As a successor to JUPITER-III, the Io program investigated the effects of uranium fuel distribution in mixed uranium and plutonium fueled assemblies. The ZPPR-19A and 19B assemblies which made up the Io program are described.

This document provides the modifications to the 1988 Remedial Action Plan (RAP) of the contaminated materials at the Monument Valley, Arizona, and Mexican Hat, Utah. The text detailing the modifications and attachments 1 through 6 are provided with this document. The RAP was developed to serve a two-fold purpose. It presents the activities proposed by the Department of Energy (DOE) to accomplish long-term stabilization and control of the residual radioactive materials (RRM) from Monument Valley, Arizona, and Mexican Hat, Utah, at the Mexican Hat disposal site. It also serves to document the concurrence of both the Navajo Nation and the Nuclear Regulatory Commission (NRC) in the remedial action. This agreement, upon execution by DOE and the Navajo Nation and concurrence by the NRC, becomes Appendix B of the Cooperative Agreement. This document has been structured to provide a comprehensive understanding of the remedial action proposed for the Monument Valley and Mexican Hat sites. It includes specific design and construction requirements for the remedial action. Pertinent information and data are included with reference given to the supporting documents.

Various magnetic optics have been considered for PEP storage ring which can be used depending on the operational circumstances. The storage ring for example is operated around 14.5 GeV when high energy investigations are carried out in which the positron and electron beams collide, the colliding-beam optics (CBO) mode. The low-emittance optics (LEO) has been tested at 8 GeV which is very useful for numerous synchrotron radiation studies. In addition, a new lattice with damping wigglers has been proposed which can provide very low emittance. This is referred to as very low emittance optics (VLEO). These lattices also provide straight sections with different lengths varying from 6 m in the symmetry straight, to 14 or 117 meters in the interactions regions which in principle would permit installation of very long undulator insertion devices for special applications. The ability to profitably utilize the radiation from these undulators proposed for PEP is determined by their performance in the different operating modes and by whether the design tolerance required for acceptable operation of the device can be met with available technology. The purpose of this paper is to provide spectral characteristics for some typical devices calculated using a Monte-Carlo algorithm in which the …

There is some concern that the FB-Line D1 dissolver may have been pressurized above the liquid head pressure by a reaction that occurred in the dissolver. The D1 Pu limit of 8.8 kg is based on a tank with an inner slab thickness of 3.34 inches and a wall thickness of 0.40 inches for a total slab thickness of 4.14 inches. When the incident occurred there was less than 1,500 gm Pu (1,323 gm) in the dissolver. Some calculations were made to determine the safe slab thickness for 1,500 gm Pu in the D1 dissolver. Calculations show that 1,500 gm Pu can be safely contained in a 6 inch thick D1 dissolver provided the cabinet panels are in place.

This radiologic characterization of the inactive uranium millsite at Mexican Hat, Utah, was conducted by Bendix Field Engineering Corporation for the US Department of Energy (DOE), Grand Junctions Project Office in response to and in accord with a Statement of Work prepared by the DOE Uranium Mill Tailings Remedial Action Project (UMTRAP) Technical Assistance Contractor, Jacobs Engineering Group, Inc. The objective of this project was to determine the horizontal and vertical extent of contamination that exceeds the US Environmental Protection Agency (EPA) standards at the Mexican Hat site. The data presented in this report are required for characterization of the areas adjacent to the Mexican Hat tailings piles and for the subsequent design of cleanup activities. Some on- pile sampling was required to determine the depth of the 15-pCi/g Ra- 226 interface in an area where wind and water erosion has taken place.

This document discusses the results of this evaluation and a recommendation to retain the graded fuel cycle in which one-half of the fuel elements are exchanged at each refueling. This recommendation is based on the better performance of the graded cycle relative to the evaluation criteria of both economics and control margin. A choice to retain the graded cycle and a power density of 5.9 MW/m{sup 3} for the upcoming conceptual design phase was deemed prudent for the following reasons: the graded cycle has significantly better economics, and essentially the same expected availability factor as the batch design, when both are evaluated against the same requirements, including water ingress; and the reduction in maximum fuel pin power peaking in the batch design compared to the graded cycle is only a few percent and gas hot streaks are not improved by changing to a batch cycle. The preliminary 2-D power distribution studies for both designs showed that maximum fuel pin power peaking, particularly near the inner reflector, was high for both designs and nearly the same in magnitude. 10 figs., 9 tabs.

The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in accordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site;to describe the conceptual designs for the repository and the waste package;and to present the plans for obtaining the geologic information necessary to demonstrate the suitability of the site for repository, to design the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and developing a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next;it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing principles, and organization of this site characterization plan are outlined, …

We have established a methodology to determine the time dependence of strength and transport properties of tuff under conditions appropriate to a nuclear waste repository. Exploratory tests to determine the approximate magnitudes of thermomechanical property changes are nearly complete. In this report we describe the capabilities of an apparatus designed to precisely measure the time-dependent deformation and permeability of tuff at simulated repository conditions. Preliminary tests with this new apparatus indicate that microclastic creep failure of tuff occurs over a narrow strain range with little precursory Tertiary creep behavior. In one test, deformation under conditions of slowly decreasing effective pressure resulted in failure, whereas some strain indicators showed a decreasing rate of strain.

The Nevada Nuclear Waste Storage Investigation (NNWSI) has developed a strategy to license a nuclear waste repository in tuff. This strategy, which is currently circulating in draft form within the Department of Energy`s Office of Civilian Radioactive Waste Management, has important implications for DWPF waste form qualification activities, design of the DWPF process, and DWPF operations. In this report, the strategy and its implications for the DWPF are presented. 2 refs.

The objective of the waste package program is to insure that a package is designed suitable for a repository in tuff that meets performance requirements of the NRC. In brief, the current (draft) regulation requires that the radionuclides be contained in the engineered system for 1000 years, and that, thereafter, no more than one part in 10{sup 5} of the nuclides per year leave the boundary of the system. Studies completed as of this writing are thermal modeling of waste packages in a tuff repository and analysis of sodium bentonite as a potential backfill material. Both studies will be presented. Thermal calculations coupled with analysis of the geochemical literature on bentonite indicate that extensive chemical and physical alteration of bentonite would result at the high power densities proposed (ca. 2 kW/package and an area density of 25 W/m{sup 2}), in part due to compacted bentonite`s relatively low thermal conductivity when dehydrated ({similar_to}0.6 +- 0.2 W/m{sup 0}C). Because our groundwater contains K{sup +}, an upper hydrothermal temperature limit appears to be 120 to 150{sup 0}C. At much lower power densities (less than 1 kW per package and an areal density of 12 W/m{sup 2}), bentonite may be suitable.

The objective of the EMMP is to document compliance with the NWPA. To do so, a summary description of site characterization activites is provided, based on the consultation draft of the SCP. Subsequent chpaters identify those technical areas having the potential to be impacted by site characterization activities and the monitoring plans proposed to identify whether those impacts acutally occur. Should monitoring confirm the potential for significant adverse impact, mitigative measures will be developed. In the context of site characterization, mitigation is defined as those changes in site characterization activities that serve to avoid or minimize, to the maximum extent practicle, any significant adverse environmental impacts. Although site characterization activies involve both surface and subsurface activities, it is the surface-based aspect of site characterization that is addressed in detailed by the EMMP. The schedule and duration of these activities is given in the consultation draft of the SCP. A breif summary of all proposed activities is given in the EMMP. 10 refs., 8 figs.