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The Radiation Exposure Compensation Act (RECA): Compensation Related to Exposure to Radiation from Atomic Weapons Testing and Uranium Mining

Description: This report discusses the Radiation Exposure Compensation Act (RECA), which provides one-time benefit payments to persons who may have developed cancer or other specified diseases after being exposed to radiation from atomic weapons testing or uranium mining, milling, or transporting.
Date: March 24, 2015
Creator: Szymendera, Scott D.
Item Type: Report
Partner: UNT Libraries Government Documents Department

Closing Yucca Mountain: Litigation Associated with Attempts to Abandon the Planned Nuclear Waste Repository

Description: Passed in 1982, the Nuclear Waste Policy Act (NWPA) was an effort to establish an explicit statutory basis for the Department of Energy (DOE) to dispose of the nation's most highly radioactive nuclear waste. Congress amended the NWPA's site selection process in 1987, however, and designated Yucca Mountain, Nevada, as the sole candidate site for the repository by terminating site specific activities at all other sites. This report discusses the Obama Administration and the DOE's steps to terminate the Yucca Mountain project, and the subsequent opposition to their efforts.
Date: June 4, 2012
Creator: Garvey, Todd
Item Type: Report
Partner: UNT Libraries Government Documents Department

Effects of Radiation from Fukushima Dai-ichi on the U.S. Marine Environment

Description: The massive Tohoku earthquake and tsunami of March 11, 2011, caused extensive damage in northeastern Japan, including damage to the Fukushima Dai-ichi nuclear power installation, which resulted in the release of radiation. Concerns arose about the potential effects of this released radiation on the U.S. marine environment and resources.
Date: April 2, 2012
Creator: Buck, Eugene H. & Upton, Harold F.
Item Type: Report
Partner: UNT Libraries Government Documents Department

Effects of Radiation from Fukushima Daiichi on the U.S. Marine Environment

Description: The massive Tohoku earthquake and tsunami of March 11, 2011, caused extensive damage in northeastern Japan, including damage to the Fukushima Dai-ichi nuclear power installation, which resulted in the release of radiation. This report discusses concerns about the potential effects of this released radiation on the U.S. marine environment and resources.
Date: April 5, 2011
Creator: Buck, Eugene H.; Upton, Harold F. & Folger, Peter
Item Type: Report
Partner: UNT Libraries Government Documents Department

Effects of Radiation from Fukushima Daiichi on the U.S. Marine Environment

Description: The massive Tohoku earthquake and tsunami of March 11, 2011, caused extensive damage in northeastern Japan, including damage to the Fukushima Dai-ichi nuclear power installation, which resulted in the release of radiation. This report discusses concerns which have arisen about the potential effects of this released radiation on the U.S. marine environment and resources.
Date: April 15, 2011
Creator: Buck, Eugene H. & Upton, Harold F.
Item Type: Report
Partner: UNT Libraries Government Documents Department

Characteristics of KE Basin Sludge Samples Archived in the RPL - 2007

Description: Samples of sludge were collected from the K East fuel storage basin (KE Basin) floor, contiguous pits (Weasel Pit, North Load Out Pit, Dummy Elevator Pit, and Tech View Pit), and fuel storage canisters between 1995 and 2003 for chemical and radionuclide concentration analysis, physical property determination, and chemical process testing work. Because of the value of the sludge in this testing and because of the cost of obtaining additional fresh samples, an ongoing program of sludge preservation has taken place with the goals to track the sludge identities and preserve, as well as possible, the sludge composition by keeping the sludge in sealed jars and maintaining water coverage on the sludge consistent with the controlling Fluor Hanford (FH) Sampling and Analysis plans and FH contracts with the Pacific Northwest National Laboratory (PNNL). This work was originally initiated to provide material for planned hydrothermal treatment testing in accordance with the test plan for the Sludge Treatment Project (STP) corrosion process chemistry follow on testing (Delegard et al. 2007). Although most of the planned hydrothermal testing was canceled in July 2007 (as described in the forward of Delegard et al. 2007), sample consolidation and characterization was continued to identify a set of well-characterized sludge samples that are suited to support evolving STP initiatives. The work described in the letter was performed by the PNNL under the direction of the Sludge Treatment Project, managed by Fluor Hanford.
Date: November 22, 2011
Creator: Delegard, Calvin H.; Schmidt, Andrew J. & Chenault, Jeffrey W.
Item Type: Report
Partner: UNT Libraries Government Documents Department

THE AEC PROGRAM OF ATMOSPHERIC RADIOACTIVITY AND FALLOUT RESEARCH

Description: From Health Physics Society 8th Annual Meeting, New York, June 1963. The Atomic Energy Commission's research program on atmospheric radioactivity and fallout is reviewed. The main purpose of the research program is to provide an increasingly complete and sound scientific basis for the prediction of radiation doses to man that may result from the introduction of radioactive material into the atmosphere. Techniques and computation models for making such predictions are discussed. Emphasis is placed on studies of world-wide transport and distribution phenomena associated with fallout from weapons tests and low- altitude local problems arising from industrial and laboratory operations. The program is conducted through research contracts with universities, industrial organizations, and interagency agreements with other government agencies and AEC facilities. A list is appended of current participants in the program. (C.H.)
Date: January 1, 1963
Creator: Klement, A.W. Jr. & Holland, J.Z.
Item Type: Report
Partner: UNT Libraries Government Documents Department

CHARACTERISTICS OF RADIOACTIVITY PRODUCED BY NUCLEAR EXPLOSIVES

Description: The amounts and kinds of radioactivities produced by detonation of a nuclear explosive are dependent upon the specific design of the explosive. The two design extremes are a pure fission device, which will produce about 1.4 x 10/ sup 23/ fissions, and therefore about 2.8 x 10/sup 23/ fission products, per kiloton of energy released, and a pure thermonuclear device which would produce no fission products, but would produce approximately 10/sup 23/ atoms of tritium per kiloton. In both cases, interaction of escaping neutrons with the materials of the device itself, and with the surrounding media, could lead to further radioactivities. The behavior and ultimate fate of the activities produced by the explosion depend on the composition of the medium in which the detonation occurs, the nature of the detonation, and the chemical species involved. Some typical cases are described. (auth)
Date: January 1, 1961
Creator: Miskel, J A
Item Type: Report
Partner: UNT Libraries Government Documents Department

CHEMICAL ENGINEERING DIVISION SUMMARY REPORT FOR JANUARY, FEBRUARY, AND MARCH 1957

Description: Fluoride Volatilization Separations Process. Development of a fused fluoride process for dissolution of uranium-- zirconium fuel alloys continued. In corrosion tests to find a suitable container material, Ni was found to be susceptible to a sulfur-type attack. Hastelloy B showed promise, and graphite offers excellent chemical resistance but poor mechanical strength. The dissolution rate of Zr in NaF-- ZrF as affected by impingement of the HF sparge was studied. Production of UF/sub 6/ by fluidized bed fluorination of UF/sub 4/ from ore concentrates was studied. The preparation, melting point, vapor pressure, and vapor density of UF/sub 5/ are given. Preliminary dissolution and recovery runs in semi-works equipment are discussed. Fluidization. Fluidized- bed techniques have been applied to conversion of UO/sub 2/(NO/sub 3/)/sub 2/ to UF/sub 4/ and to calcination of radioactive liquid wastes. Activities of the Green Salt Pilot Plant and shakedown runs of the shielded waste calciner are described. Reactor Chemistry. Studies continued on the kinetics and mechanism of oxidation of U, Th, and Zr. Data are given for oxidation of U in oxygen from 125 to 295 tained C and 20 to 800 mm pressure, and for Zr from 400 to 900 tained C and 200 nan O/sub 2/ pressure. The ratio of capture to fission cross sections for U/sup 233/ and U/sup 238/ in EBR-I have been determined as a function of position. ChemicalMetallurgical Separations Processes. Development of pyrometullurgical processing of spent reactor fuels continued. Work is repcrted on: melt refining and casting of U--Pu; iodine volatility problem; the system U--B-- Ta; the distribution coefficients for Pu between U--Cr and Mg and U and Mg; extraction of Pu from U by liquid Mg; Ce removal by dross refining; adsorption of volatilized metuls on surface active materials; and fractional crystallization of U with Zn. Analytical Research. A study ...
Date: October 31, 1960
Item Type: Report
Partner: UNT Libraries Government Documents Department

The Distribution of Radioactivity in the Mouse Following Administration of Dibenzanthracene Labeled in the 9 and 10 Positions with Carbon Fourteen

Description: Dibenzanthracene, labeled in the 9 and 10 positions with carbon fourteen has been administered to mice intravenously and by stomach tube as an aqueous colloid, and intraperitoneally, subcutaneously, and by stomach tube in tricaprylin solution. The distribution of radioactivity in the mice at various time intervals after administration of the carcinogen has been determined. The radioactivity is rapidly eliminated, largely through the feces, and ordinarily very little is absorbed. The distribution and rate of elimination depends upon the mode of administration. There is an appreciable quantity of radioactivity in tumors produced several months after a single subcutaneous injection of dibenzanthracene. There appear to be no detectable effects from the radiation of the labeled carcinogen.
Date: January 30, 1948
Creator: Heidelberger, Charles & Jones, Hardin, B.
Item Type: Article
Partner: UNT Libraries Government Documents Department

APPARATUS FOR THE STUDY OF FISSION-GAS RELEASE FROM FUELS DURING POSTIRRADIATION HEATING AT TEMPERATURES UP TO 1600 C

Description: An apparatus to study rare-gas fission-product release from nuclear fuel materials during postirradiation heating was developed. Xenon and krypton fission gases escaping from a small specimen during heating at constant temperature are measured using a continuous radioactivity monitor and charcoal adsorption traps. The rhodium-wound furnace is capable of operation at 1600 deg C. Helium carrier gas is purified by activated alumina, copper, and zirconium traps, and the oxygen and moisture contents of the gas are monitored continuously. The operating procedure and data are presented for a typical heating experiment in which fused uranium dioxide was studied. (auth)
Date: July 22, 1960
Creator: Barnes, R. H. & Sunderman, D. N.
Item Type: Report
Partner: UNT Libraries Government Documents Department

A TRACER STUDY OF THE TRANSPORT OF CHROMIUM IN FLUORIDE FUEL SYSTEMS

Description: An experimental study was made of the mass transport of chromium in poly- thermal Inconel-fluoride fuel systems. The transport of chromium was followed by toe technique of adding radioactive Cr/sup 51/ to the system as either CrF/sub 2/ , in the salt or as elemental chromium in the solid phase. The rates of diffusion of chromium in Inconel at 600, 700, 800, and 900 deg C were determined by an electropolishing technique. Polythermal studies were carried out by three methods, tilting capsules, thermal-convection loops, and pumping loops. Tilting- capsule experiments indicated that the preferred location for chromium deposition on the wall was in the region of maximum temperature but the conclusions were not clear cut. Thermal convection loops operated for 125 and 288 hr showed radioactivity profile which could be attributed to simple exchange, with some distortion in the 288 hr case. The duration of these experiments was evidently insufficient io allow equilibrium to be reached in the salt. A thermal- convection loop operated for 400 hr showed distortion in the exchange radioactivity profile which indicated a favorable position for chromium deposition at a point about 100 deg F below the maximum wall temperature, and on the upstream side of the flow. A pumping loop of Inconel and salt mix gave an activity profile which was very similar to that of the 400-hr thermalconvection loop, indicating a favorable deposition point 100 deg F below maximuim temperature on the upstream side. One hypothesis advanced is that the long-term corrosion rate of chromium in the Inconel-salt system is controlled by the rate of diffusion of chromium into the wall at a temperature about 100 deg F below the maximum temperature on the upstream side. (auth)
Date: June 18, 1957
Creator: Price, R.B.; Sunderman, D.N.; Pobereskin, M. & Calkin, G.D.
Item Type: Report
Partner: UNT Libraries Government Documents Department

RADIOMETRIC METHODS FOR THE DETERMINATION OF MAGNESIUM AND CALCIUM IN PORTLAND CEMENT

Description: Radiometric methods of analysis for magnesium and calcium have been developed as part of a program for the U. S. Atomic Energy Commission. Office of Isotopes Development, which are applicable to the determination of these elements in portland cement Both methods employ, as a precipitant, a standard solution of (NH/sub 4/)/sub 2/HPO/sub 4/ labeled with phosphorus-32. In the presence of NH/ sub 4/OH, this reagent precipitate; MgNH/sub 4/PO/sub 4/ or Ca/sub 3/(PO/sub 4/)/ sub 2/ from a solution of magnesium or calcium ions. The reduction in the radioactivity level of the labeled phosphate solution after precipitation serves as a measure of the phosphate reacted and thus a measure of the quantity of magnesium or calcium present. Studies have been made of the effects of reagent concentration, NH/sub 4/OH concentration, and other experimental variables. The interference of other elements present normally in portland cement and its raw materials has been determined. The concentration ranges for highest accuracy have been found to be 5 to 15 mg of MgO per 100 ml and 15 to 30 mg of CaO per 50 ml. (auth)
Date: February 18, 1960
Creator: Brown, C.T.; Howes, J.E. Jr.; Elleman, T.S.; Townley, C.W. & Sunderman, D.N.
Item Type: Report
Partner: UNT Libraries Government Documents Department

Benchmark Studies of Induced Radioactivity Produced in LHC Materials, Pt I: Remanent Dose Rates

Description: Samples of materials which will be used in the LHC machine for shielding and construction components were irradiated in the stray radiation field of the CERN-EU high-energy reference field facility. After irradiation, the specific activities induced in the various samples were analyzed with a high-precision gamma spectrometer at various cooling times, allowing identification of isotopes with a wide range of half-lives. Furthermore, the irradiation experiment was simulated in detail with the FLUKA Monte Carlo code. A comparison of measured and calculated specific activities shows good agreement, supporting the use of FLUKA for estimating the level of induced activity in the LHC.
Date: April 12, 2006
Creator: Brugger, M.; Mayer, S.; Roesler, S.; Ulrici, L.; /CERN; Khater, H. et al.
Item Type: Article
Partner: UNT Libraries Government Documents Department

Benchmark Studies of Induced Radioactivity Produced in LHC Materials, Pt II Specific Activities

Description: A new method to estimate remanent dose rates, to be used with the Monte Carlo code FLUKA, was benchmarked against measurements from an experiment that was performed at the CERN-EU high-energy reference field facility. An extensive collection of samples of different materials were placed downstream of and laterally to a copper target, intercepting a positively charged mixed hadron beam with a momentum of 120 GeV/c. Emphasis was put on the reduction of uncertainties such as careful monitoring of the irradiation parameters, the use of different instruments to measure dose rates, detailed elemental analyses of the irradiated materials and detailed simulations of the irradiation experiment. Measured and calculated dose rates are in good agreement.
Date: April 12, 2006
Creator: Brugger, M.; Mayer, S.; Roesler, S.; Ulrici, L.; /CERN; Khater, H. et al.
Item Type: Article
Partner: UNT Libraries Government Documents Department

Analysis of Radionuclide Migration through a 200-m Vadose Zone Following a 16-year Infiltration Event

Description: The CAMBRIC nuclear test was conducted beneath Frenchman Flat at the Nevada Test Site on May 14, 1965. The nuclear device was emplaced in heterogeneous alluvium, approximately 70 m beneath the ambient water table, which is itself 220 m beneath the ground surface. Approximately 10 years later, groundwater adjacent to the test was pumped steadily for 16 years to elicit information on the migration of residual radionuclide migration through the saturated zone. The pumping well effluent--containing mostly soluble radionuclides such as tritium, {sup 14}C, {sup 36}Cl, {sup 85}Kr, {sup 129}I, and {sup 106}Ru--was monitored, discharged to an unlined ditch, and allowed to flow towards Frenchman Lake over one kilometer away. Discharged water and radionuclides infiltrated into the ground and created an unexpected second experiment in which the migration of the effluent through the unsaturated zone back to the water table could be studied. In this paper, the pumping and effluent data are being utilized in conjunction with a series of geologic data, new radionuclide measurements, isotopic age-dating estimates, and vadose zone flow and transport models to better understand the movement of radionuclides between the ditch and the water table. Measurements of radionuclide concentrations in water samples produced from a water table monitoring well 100 m away from the ditch indicate rising levels of tritium since 1993. The detection of tritium in the monitoring well occurs approximately 16 years after its initial discharge into the ditch. Modeling and tritium age dating have suggested 3 to 5 years of this 16-year transit time occurred solely in the vadose zone. They also suggest considerable recirculation of the pumping well discharge back into the original pumping well. Notably, there have been no observations of {sup 14}C or {sup 85}Kr at the water table, suggesting their preferential retention or volatilization during transit to the water ...
Date: September 21, 2004
Creator: Tompson, A B; Hudson, G B; Smith, D K & Hunt, J R
Item Type: Article
Partner: UNT Libraries Government Documents Department

DISPOSAL CONTAINER HANDLING SYSTEM DESCRIPTION DOCUMENT

Description: The Disposal Container Handling System receives and prepares new disposal containers (DCs) and transfers them to the Assembly Transfer System (ATS) or Canister Transfer System (CTS) for loading. The system receives the loaded DCs from ATS or CTS and welds the lids. When the welds are accepted the DCs are termed waste packages (WPs). The system may stage the WP for later transfer or transfer the WP directly to the Waste Emplacement/Retrieval System. The system can also transfer DCs/WPs to/from the Waste Package Remediation System. The Disposal Container Handling System begins with new DC preparation, which includes installing collars, tilting the DC upright, and outfitting the container for the specific fuel it is to receive. DCs and their lids are staged in the receipt area for transfer to the needed location. When called for, a DC is put on a cart and sent through an airlock into a hot cell. From this point on, all processes are done remotely. The DC transfer operation moves the DC to the ATS or CTS for loading and then receives the DC for welding. The DC welding operation receives loaded DCs directly from the waste handling lines or from interim lag storage for welding of the lids. The welding operation includes mounting the DC on a turntable, removing lid seals, and installing and welding the inner and outer lids. After the weld process and non-destructive examination are successfully completed, the WP is either staged or transferred to a tilting station. At the tilting station, the WP is tilted horizontally onto a cart and the collars removed. The cart is taken through an air lock where the WP is lifted, surveyed, decontaminated if required, and then moved into the Waste Emplacement/Retrieval System. DCs that do not meet the welding non-destructive examination criteria are transferred to ...
Date: June 30, 2000
Creator: Loros, E. F.
Item Type: Report
Partner: UNT Libraries Government Documents Department

Errors Associated with the Direct Measurement of Radionuclides in Wounds

Description: Work in radiation areas can occasionally result in accidental wounds containing radioactive materials. When a wound is incurred within a radiological area, the presence of radioactivity in the wound needs to be confirmed to determine if additional remedial action needs to be taken. Commonly used radiation area monitoring equipment is poorly suited for measurement of radioactive material buried within the tissue of the wound. The Lawrence Livermore National Laboratory (LLNL) In Vivo Measurement Facility has constructed a portable wound counter that provides sufficient detection of radioactivity in wounds as shown in Fig. 1. The LLNL wound measurement system is specifically designed to measure low energy photons that are emitted from uranium and transuranium radionuclides. The portable wound counting system uses a 2.5cm diameter by 1mm thick NaI(Tl) detector. The detector is connected to a Canberra NaI InSpector{trademark}. The InSpector interfaces with an IBM ThinkPad laptop computer, which operates under Genie 2000 software. The wound counting system is maintained and used at the LLNL In Vivo Measurement Facility. The hardware is designed to be portable and is occasionally deployed to respond to the LLNL Health Services facility or local hospitals for examination of personnel that may have radioactive materials within a wound. The typical detection levels in using the LLNL portable wound counter in a low background area is 0.4 nCi to 0.6 nCi assuming a near zero mass source. This paper documents the systematic errors associated with in vivo measurement of radioactive materials buried within wounds using the LLNL portable wound measurement system. These errors are divided into two basic categories, calibration errors and in vivo wound measurement errors. Within these categories, there are errors associated with particle self-absorption of photons, overlying tissue thickness, source distribution within the wound, and count errors. These errors have been examined and can cause ...
Date: March 2, 2006
Creator: Hickman, D. P.
Item Type: Report
Partner: UNT Libraries Government Documents Department

A Science-Based Approach to Understanding Waste Form Durability in Open and Closed Nuclear Fuel Cycles

Description: There are two compelling reasons for understanding source term and near-field processes in a radioactive waste geologic repository. First, almost all of the radioactivity is initially in the waste form, mainly in the spent nuclear fuel (SNF) or nuclear waste glass. Second, over long periods, after the engineered barriers are degraded, the waste form is a primary control on the release of radioactivity. Thus, it is essential to know the physical and chemical state of the waste form after hundreds of thousands of years. The United States Department of Energy's Yucca Mountain Repository Program has initiated a long-term program to develop a basic understanding of the fundamental mechanisms of radionuclide release and a quantification of the release as repository conditions evolve over time. Specifically, the research program addresses four critical areas: (a) SNF dissolution mechanisms and rates; (b) formation and properties of U{sup 6+}-secondary phases; (c) waste form-waste package interactions in the near-field; and (d) integration of in-package chemical and physical processes. The ultimate goal is to integrate the scientific results into a larger scale model of source term and near-field processes. This integrated model will be used to provide a basis for understanding the behavior of the source term over long time periods (greater than 10{sup 5} years). Such a fundamental and integrated experimental and modeling approach to source term processes can also be readily applied to development of advanced waste forms as part of a closed nuclear fuel cycle. Specifically, a fundamental understanding of candidate waste form materials stability in high temperature/high radiation environments and near-field geochemical/hydrologic processes could enable development of advanced waste forms ''tailored'' to specific geologic settings.
Date: June 22, 2006
Creator: Peters, M.T. & Ewing, R.C.
Item Type: Article
Partner: UNT Libraries Government Documents Department