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Barrier Attenuation of Air-Scattered Gamma Radiation

Description: Report of a study that "was conducted to determine the attenuation provided by vertical and horizontal barriers exposed only to skyshine radiation from cobalt-60 and cesium-137 sources. Materials of steel, aluminum, concrete, and wood were used as barriers" (p. 5).
Date: December 1964
Creator: Burson, Z. G. & Summers, R. L.
Partner: UNT Libraries Government Documents Department

Long term bleaching of optical glasses darkened by Co60 Ionizing radiation

Description: Typical camera designs include optical glass elements that may be affected by the ionizing radiation present in the natural space environment. Ordinary optical glasses darken at low (10(to the 3rd power) rad) dose levels when exposed to ionizing radiation. This darkening decreases the sensitivity of optical sensors. Optical glass flats of FK 51, LaK 0, PK 51A, and ZK Ny were exposed to a 10.6 krad dose of ionizing radiation. Spectrophotometer traces determined the transmittance of the samples as a function of wavelength in the range 350 to 850 nm before and at various time intervals after the irradiation. These measured values were then use to evaluate the rate of recovery or ``bleaching`` of the exposed samples. To prevent accelerated bleaching, the samples were kept at room temperature and away from light, except during measurement. Tables of the measured data and plots of the transmissivity vs. wavelength at various times after irradiation are presented.
Date: May 1, 1997
Creator: Wirtenson, G.R. & White, R.H.
Partner: UNT Libraries Government Documents Department


Description: >An examination was made to observe the extent and location of corrosion, crud deposits, and defects in components of the PWR pnimary fluid system and its auxiliaries. A gamma spectrum of the four-inch line upstream from the two pressurizer self-actuated relief valves showed the presence of Mn/sup 54/ and Co/sup 60/. From the gamma spectrum and the gross gamma activity, the Co/sup 60/ was found to be 5.03 x 10/sup 3/ dpm/mg or about 70 per cert of the gross gnmma activity. (J.R.D.)
Date: January 20, 1961
Partner: UNT Libraries Government Documents Department

LDRD 140639 final report : investigation of transmutation claims.

Description: The Proton-21 Laboratory in the Ukraine has been publishing results on shock-induced transmutation of several elements, including Cobalt 60 into non-radioactive elements. This report documents exploratory characterization of a shock-compressed Aluminum-6061 sample, which is the only available surrogate for the high-purity copper samples in the Proton-21 experiments. The goal was to determine Sandia's ability to detect possible shock-wave-induced transmutation products and to unambiguously validate or invalidate the claims in collaboration with the Proton-21 Laboratory. We have developed a suitable characterization process and tested it on the surrogate sample. Using trace elemental analysis capabilities, we found elevated and localized concentrations of impurity elements like the Ukrainians report. All our results, however, are consistent with the ejection of impurities that were not in solution in our alloy or were deposited from the cathode during irradiation or possibly storage. Based on the detection capabilities demonstrated and additional techniques available, we are positioned to test samples from Proton-21 if funded to do so.
Date: November 1, 2009
Creator: Reich, Jeffrey E.; Van Devender, J. Pace; Mowry, Curtis Dale; Grant, Richard P. & Ohlhausen, James Anthony
Partner: UNT Libraries Government Documents Department

Final Removal Action Report of the CPP-603A Basin Facility

Description: This Final Removal Action Report describes the actions that were taken under the non-time-critical removal action recommended in the Action Memorandum for the Non-Time Critical Removal Action at the CPP-603A Basins, Idaho Nuclear Technology and Engineering Center, as evaluated in the Engineering Evaluation/Cost Analysis for the CPP-603A Bason Non-Time Critical Removal Action, Idaho Nuclear Technology and Engineering Center. The Removal Action implemented consolidation and recording the location of debris objects containing radioactive cobalt (cobalt-60), removal and management of a small high-activity debris object (SHADO 1), the removal, treatment, and disposal of the basin water at the Idaho CERCLA Disposal Facility (ICDF) evaporation ponds, and filling the basins with grout/controlled low strength material.
Date: January 4, 2007
Creator: Croson, D. V.
Partner: UNT Libraries Government Documents Department


Description: S>Of primary interest in a military reactor is the desirability of quick and easy maintenance of as much of the system as possibie throughout the lifetime of the power plant. It is therefore desirable that no serious amounts of long- lived activity build up in equipment outside the core. An investigation was made to determine the activity due to small amounts of cobalt and tantalum in the stainless steel core of the APPR-1. The investigation indicates that equal weight percentages of Ta and Co would build up about equal activities in the core after 1.5 years of operation. The use of 347 stainless steel containing a nominal 0.2 wt.% Co would therefore produce a more serious activation problem than is now being experienced wlth the present 304L core. A desire to reduce the production of long-lived activities would then require specifying a low tantalum as well as a low cobalt content if 347 stainless steel were used for the second APPR-1 core. Fabrication of a low cobalt 304L core is therefore an attempt to reduce and to simplify the activation problemn (A.C.)
Date: February 10, 1958
Creator: Gross, E.E.
Partner: UNT Libraries Government Documents Department


Description: Casting a U shield for a kilocurie Co/sup 60/ source is described. The melting equipment is described, and data on the castings are tabulated. Examination of the casting revealed no large blow holes or cracks, and the method was considered adequate. (J.R.D.)
Date: August 1, 1959
Creator: Dunworth, R. J. & Macherey, R. E.
Partner: UNT Libraries Government Documents Department

Periodic Radiation Survey of Reactor Plant Container and Components After Shutdown. Core I, Seed 1. Section 3. Test Results T-612076

Description: A periodic survey was conducted to determine changes in radiation level in the hairpin loops resulting from continued operation of the reactor at power. Results indinate that Co/sup 66/ was the major contributor to the gamma activity present on the 1 BD Hairpin Loop. Loose scale in this area amounted to 0.104 mg/ cm/sup 2/. (J.R.D.)
Date: January 17, 1961
Partner: UNT Libraries Government Documents Department

Treatment of spent electropolishing solution for removal of cobalt-60

Description: The Irradiated Materials Examination and Testing (IMET) Facility at Oak Ridge National Laboratory electropolishes various types of irradiated metal specimens prior to examination of metallurgical and mechanical properties. The standard electropolishing solution used at IMET for most specimens consists of a 7:1 methanol/sulfuric acid mixture, with smaller amounts of a 3:1 methanol/nitric acid solution and a 10:6:1 methanol/2-butoxyethanol/perchloric acid solution also being used. Cobalt-60 is the primary source of gamma radiation in the spent solutions, with lesser amounts from manganese-54 and iron-59. A treatment method is needed to remove most of the Co-60 from these solutions to allow the waste solutions to be contact-handled for disposal. A wide range of adsorbents was tested for removing cobalt from the electropolishing solutions. No adsorbent was found that would treat full strength solution, but a complexing ion exchange resin (Chelex 100, BioRad Labs, or Amberlite IRC-718, Rohm and Haas Co.) will remove cobalt and other heavy metals from partially neutralized (pH=3) solution. A 5 wt% sodium hydroxide solution is used for pH adjustment, since more concentrated caustic caused sodium sulfate precipitates to form. Lab-scale column tests have shown that about 10 bed volumes of methanol/sulfuric acid solution, 30 bed volumes of methanol/nitric acid solution or 15 bed volumes of methanol/2-butoxyethanol/perchloric acid solution can be treated prior to initial Co-60 breakthrough.
Date: February 1, 1996
Creator: Taylor, P.A.; Youngblood, E.L. & Macon, R.J.
Partner: UNT Libraries Government Documents Department

Rate of long term bleaching in FK 51 optical glass darkened by Co60 ionizing radiation at dose rates of 10 krad/hr and 7 rad/hr

Description: A previous paper presented long term bleaching data on various glasses exposed to 10.6 krad of ionizing radiation. All the glasses reported except FK 51 have readily available `G` glass equivalents that are stabilized to the natural space environment. Yet, FK 51, because of its location on the Abbe diagram is extremely useful in certain lens design applications. To more fully explore the bleaching of FK 51, after the initial dose of 10.6 krad at 11.8 krad/hour, we irradiated three more samples at a similar dose rate but to different total doses. Since the dose rate for this study was significantly higher than the dose rate anticipated for glasses in as shielded space-based lens system (tilde 3 rad/day), additional data were obtained at a lower rate of 7 rad/hour. While this dose rate is still higher than the anticipated operational rate, it is more than 1000 times lower than the dose 011 011 011 rate used for our initial studies. The bleaching rate for the samples exposed at the lower dose rate is considerably less than for the samples exposed at the higher rate.
Date: July 1997
Creator: Wirtenson, G. R. & White, R. H.
Partner: UNT Libraries Government Documents Department

Gamma Irradiation Facility at Sandia National Laboratories, Albuquerque, New Mexico. Final environmental assessment

Description: The US Department of Energy (DOE) has prepared an environmental assessment (EA) on the proposed construction and operation of a new Gamma Irradiation Facility (GIF) at Sandia National Laboratories/New Mexico (SNL/NM). This facility is needed to: enhance capabilities to assure technical excellence in nuclear weapon radiation environments testing, component development, and certification; comply with all applicable ES and H safeguards, standards, policies, and regulations; reduce personnel radiological exposure to comply with ALARA limits in accordance with DOE orders and standards; consolidate major gamma ray sources into a central, secured area; and reduce operational risks associated with operation of the GIF and LICA in their present locations. This proposed action provides for the design, construction, and operation of a new GIF located within TA V and the removal of the existing GIF and Low Intensity Cobalt Array (LICA). The proposed action includes potential demolition of the gamma shield walls and removal of equipment in the existing GIF and LICA. The shielding pool used by the existing GIF will remain as part of the ACRR facility. Transportation of the existing {sup 60}Co sources from the existing LICA and GIF to the new facility is also included in the proposed action. Relocation of the gamma sources to the new GIF will be accomplished by similar techniques to those used to install the sources originally.
Date: November 1, 1995
Partner: UNT Libraries Government Documents Department


Description: The Building 830 Gamma Irradiation Facility (GIF) at Brookhaven National Laboratory (BNL) was decommissioned because its design was not in compliance with current hazardous tank standards and because its cobalt-60 sources were approaching the end of their useful life. The facility contained 354 stainless steel encapsulated cobalt-60 sources in a pool, which provided shielding. Total cobalt-60 inventory amounted to 24,000 Curies (when the sources were shipped for disposal). The decommissioning project included packaging, transport and disposal of the sources and dismantling and disposing of all other equipment associated with the facility. Worker exposure was a major concern in planning for the packaging and disposal of the sources. These activities were planned carefully according to ALARA (As Low As Reasonably Achievable) principles. As a result, the actual doses experienced during the work were lower than anticipated. Because the sources were sealed, most of the remaining equipment was not contaminated; therefore disposal was straightforward, as scrap metal and construction debris. However, disposal of the pool water involved addressing environmental concerns, since the planned method was to discharge the slightly contaminated water to the BNL sewage treatment plant.
Date: February 24, 2001
Partner: UNT Libraries Government Documents Department

Demonstration recommendations for accelerated testing of concrete decontamination methods

Description: A large number of aging US Department of Energy (DOE) surplus facilities located throughout the US require deactivation, decontamination, and decommissioning. Although several technologies are available commercially for concrete decontamination, emerging technologies with potential to reduce secondary waste and minimize the impact and risk to workers and the environment are needed. In response to these needs, the Accelerated Testing of Concrete Decontamination Methods project team described the nature and extent of contaminated concrete within the DOE complex and identified applicable emerging technologies. Existing information used to describe the nature and extent of contaminated concrete indicates that the most frequently occurring radiological contaminants are {sup 137}Cs, {sup 238}U (and its daughters), {sup 60}Co, {sup 90}Sr, and tritium. The total area of radionuclide-contaminated concrete within the DOE complex is estimated to be in the range of 7.9 {times} 10{sup 8} ft{sup 2}or approximately 18,000 acres. Concrete decontamination problems were matched with emerging technologies to recommend demonstrations considered to provide the most benefit to decontamination of concrete within the DOE complex. Emerging technologies with the most potential benefit were biological decontamination, electro-hydraulic scabbling, electrokinetics, and microwave scabbling.
Date: December 1, 1995
Creator: Dickerson, K.S.; Ally, M.R.; Brown, C.H.; Morris, M.I. & Wilson-Nichols, M.J.
Partner: UNT Libraries Government Documents Department

Using ytterbium-169 for safe and economical industrial radiography

Description: Safety has become an issue of paramount importance for industrial radiography. Many NDE facilities and suppliers are finding the cost of performing radiography Prohibitive due to heightened safety concerns for radiation area protection. The most common sources used in radiography, Iridium-192 and Cobalt-60, result in high radiation fields over a large area. Even when collimators are used large radiation fields can result from multicurie source radiography. Radiographic operations are being forced to find alternative test methods and techniques to the use of the old stand-by sources. These alternate methods are not always as comprehensive a test as full volumetric examination with radiography. Since Iridium and Cobalt are in such wide spread use, they are sometimes called upon to perform test of materials which are not in their optimum sensitivity range.
Date: January 1, 1994
Creator: Dowalo, J.A.
Partner: UNT Libraries Government Documents Department