WorldWideScience

Sample records for actinium 233

  1. Application of partition chromatography method for separation and analysis of actinium radionuclides

    International Nuclear Information System (INIS)

    Sinitsina, G.S.; Shestakova, I.A.; Shestakov, B.I.; Plyushcheva, N.A.; Malyshev, N.A.; Belyatskij, A.F.; Tsirlin, V.A.

    1979-01-01

    The method of partition chromatography is considered with the use of different extractants for the extraction of actinium-227, actinium-225 and actinium-228. It is advisable to extract actinium-227 from the irradiated radium with the help of D2FGFK. The use of 2DEGFK allows us to separate actinium-227 from alkaline and alkaline-earth elements. Amines have a higher radiative stability. An express-method has been developed for the identification of actinium-227 with TOA by its intrinsic α-emission in nonequilibrium preparations of irradiated radium-226 of small activity. Actinium-225 is extracted from uranium-233 with due regard for the fact that U, Th, and Ac are extracted differently by TBP from HNO 3 solutions. With the help of the given procedure one can reach the purifying coefficient of 10 4 . Actinium-228 is extracted from the radiummesothorium preparations by a deposition of decay products, including polonium-210 on the iron hydroxyde. Actinium-228 extraction from the mixture of radium radionuclides is performed by the partition chromatography method on D2EGFK. All the procedures for separation of actinium isotopes by the above methods are described

  2. Actinium

    International Nuclear Information System (INIS)

    Keller, C.

    1977-01-01

    There are only very few investigations dealing with the chemical and physical properties of actinium, the lanthanum homologue in the actinide series, 227 Ac, the only long-lived isotope can be produced in gram amounts only by neutron irradiation of 226 Ra, the amounts occurring in nature are too low for isolation (about 1 μg 227 Ac/1 uranium ore). Experimental work with 227 Ac gives rise to a lot of problems due to the radiation characteristics of the 227 Ac daughter nuclides. Therefore, the metal and the only ten solid compounds, prepared up to now, have been isolated in the microgram scale. Due to the high specific activity of 227 Ac, the preparation of a lot of compounds, e.g. metal-organic compounds seems to be very difficult, if not impossible. The properties of actinium in aqueous solutions have been deduced from experiments in the tracer scale only. The present investigations on actinium show that only the oxidation state + 3 exists - only radiopolarographic studies indicate the possibility of a lower valancy state (Ac 2+ ). - This review will give a critical and comprehensive description on the present knowledge about this element. The presently decreasing interest in the development of thermionic batteries using 227 Ac 2 O 3 radionuclide also implies that there will be only small progress in the chemistry of this radio-element in the near future. (orig.) [de

  3. Production of Actinium-225 via High Energy Proton Induced Spallation of Thorium-232

    Energy Technology Data Exchange (ETDEWEB)

    Harvey, James T.; Nolen, Jerry; Vandergrift, George; Gomes, Itacil; Kroc, Tom; Horwitz, Phil; McAlister, Dan; Bowers, Del; Sullivan, Vivian; Greene, John

    2011-12-30

    V protons available at Fermi National Accelerator Laboratory. Targets will be processed at Argonne National Laboratory to separate and purify the actinium-225 that will subsequently be transferred to NorthStar laboratory facilities for product quality testing and comparison to the product quality of ORNL produced actinium-225, which is currently the industry standard. The test irradiations at FNAL will produce 1-20 mCi per day which is more than sufficient for quantitative evaluation of the proposed production process. The beneficial outcome of this effort will be a new production route for actinium-225 that does not use or require any uranium-233 materials owned by DOE or use any radium-226 as an irradiation target but can supply the medical community's needs for actinium-225 now and in the future.

  4. Separation of protactinum, actinium, and other radionuclides from proton irradiated thorium target

    Science.gov (United States)

    Fassbender, Michael E.; Radchenko, Valery

    2018-04-24

    Protactinium, actinium, radium, radiolanthanides and other radionuclide fission products were separated and recovered from a proton-irradiated thorium target. The target was dissolved in concentrated HCl, which formed anionic complexes of protactinium but not with thorium, actinium, radium, or radiolanthanides. Protactinium was separated from soluble thorium by loading a concentrated HCl solution of the target onto a column of strongly basic anion exchanger resin and eluting with concentrated HCl. Actinium, radium and radiolanthanides elute with thorium. The protactinium that is retained on the column, along with other radionuclides, is eluted may subsequently treated to remove radionuclide impurities to afford a fraction of substantially pure protactinium. The eluate with the soluble thorium, actinium, radium and radiolanthanides may be subjected to treatment with citric acid to form anionic thorium, loaded onto a cationic exchanger resin, and eluted. Actinium, radium and radiolanthanides that are retained can be subjected to extraction chromatography to separate the actinium from the radium and from the radio lanthanides.

  5. Separation of actinium-227 from its daughter products by cationic resins technique

    International Nuclear Information System (INIS)

    Nastasi, M.J.C.

    1976-01-01

    A method for separating actinium-227 from its daughter products based on ion exchange principle is shown. Radionuclides mixture in perchloric acid 8,5 N and chloridric acid 0,5 N medium pass by a cationic resin column. Thorium-227 and actinium-227, which are retained by the resin, are eluted with nitric acid 6 N which releases actinium-227 while oxalic acid 7% is used for thorium-227 elution [pt

  6. Separation of Actinium 227 from the uranium minerals

    International Nuclear Information System (INIS)

    Martinez-Tarango, S.

    1991-01-01

    The purpose of this work was to separate Actinium 227, whose content is 18%, from the mineral carnotite found in Gomez Chihuahua mountain range in Mexico. The mineral before processing is is pre-concentrated and passed, first through anionic exchange resins, later the eluate obtained is passed through cationic resins. The resins were 20-50 MESH QOWEX and 100-200 MESH 50 X 8-20 in some cased 200-400 MESH AG 50W-X8, 1X8 in other cases. The eluates from the ionic exchange were electrodeposited on stainless steel polished disc cathode and platinum electrode as anode; under a current ODF 10mA for 2.5 to 5 hours and of 100mA for .5 of an hour. it was possible to identify the Actinium 227 by means of its descendents, TH-227 and RA-223, through alpha spectroscopy. Due to the radiochemical purity which the electro deposits were obtained the Actinium 227 was low and was not quantitatively determined. A large majority of the members of the natural radioactive series 3 were identified and even alpha energies reported in the literature with very low percentages of non-identified emissions were observed. We conclude that a more precise study is needed concerning ionic exchange and electrodeposit to obtain an Actinium 227 of radiochemical purity. (Author)

  7. Short history of radioactivity. No. XIII. The actinium and thorium series

    Energy Technology Data Exchange (ETDEWEB)

    Chalmers, T W

    1950-06-16

    Discussions of the actinium disintegration series (about 1905), the /sup 235/U or actinium series (as it is accepted today), the disintegration of thorium (about 1905), the thorium series in the modern form, and the 4n, 4n + 1, 4n + 2, and 4n + 3 series are presented.

  8. The sorption of polonium, actinium and protactinium onto geological materials

    International Nuclear Information System (INIS)

    Baston, G.M.N.; Berry, J.A.; Brownsword, M.; Heath, T.G.; Ilett, D.J.; McCrohon, R.; Tweed, C.J.; Yui, M.

    1999-01-01

    This paper describes a combined experimental and modeling program of generic sorption studies to increase confidence in the performance assessment for a potential high-level radioactive waste repository in Japan. The sorption of polonium, actinium and protactinium onto geological materials has been investigated. Sorption of these radioelements onto bentonite, tuff and granodiorite from equilibrated de-ionized water was studied under reducing conditions at room temperature. In addition, the sorption of actinium and protactinium was investigated at 60 C. Thermodynamic chemical modeling was carried out to aid interpretation of the results

  9. The sorption of polonium, actinium and protactinium onto geological materials

    Energy Technology Data Exchange (ETDEWEB)

    Baston, G.M.N.; Berry, J.A.; Brownsword, M.; Heath, T.G.; Ilett, D.J.; McCrohon, R.; Tweed, C.J.; Yui, M.

    1999-07-01

    This paper describes a combined experimental and modeling program of generic sorption studies to increase confidence in the performance assessment for a potential high-level radioactive waste repository in Japan. The sorption of polonium, actinium and protactinium onto geological materials has been investigated. Sorption of these radioelements onto bentonite, tuff and granodiorite from equilibrated de-ionized water was studied under reducing conditions at room temperature. In addition, the sorption of actinium and protactinium was investigated at 60 C. Thermodynamic chemical modeling was carried out to aid interpretation of the results.

  10. Application of ion exchange and extraction chromatography to the separation of actinium from proton-irradiated thorium metal for analytical purposes.

    Science.gov (United States)

    Radchenko, V; Engle, J W; Wilson, J J; Maassen, J R; Nortier, F M; Taylor, W A; Birnbaum, E R; Hudston, L A; John, K D; Fassbender, M E

    2015-02-06

    Actinium-225 (t1/2=9.92d) is an α-emitting radionuclide with nuclear properties well-suited for use in targeted alpha therapy (TAT), a powerful treatment method for malignant tumors. Actinium-225 can also be utilized as a generator for (213)Bi (t1/2 45.6 min), which is another valuable candidate for TAT. Actinium-225 can be produced via proton irradiation of thorium metal; however, long-lived (227)Ac (t1/2=21.8a, 99% β(-), 1% α) is co-produced during this process and will impact the quality of the final product. Thus, accurate assays are needed to determine the (225)Ac/(227)Ac ratio, which is dependent on beam energy, irradiation time and target design. Accurate actinium assays, in turn, require efficient separation of actinium isotopes from both the Th matrix and highly radioactive activation by-products, especially radiolanthanides formed from proton-induced fission. In this study, we introduce a novel, selective chromatographic technique for the recovery and purification of actinium isotopes from irradiated Th matrices. A two-step sequence of cation exchange and extraction chromatography was implemented. Radiolanthanides were quantitatively removed from Ac, and no non-Ac radionuclidic impurities were detected in the final Ac fraction. An (225)Ac spike added prior to separation was recovered at ≥ 98%, and Ac decontamination from Th was found to be ≥ 10(6). The purified actinium fraction allowed for highly accurate (227)Ac determination at analytical scales, i.e., at (227)Ac activities of 1-100 kBq (27 nCi to 2.7 μCi). Copyright © 2014 Elsevier B.V. All rights reserved.

  11. Amides with nitrogenous heterocyclic substituent, their manufacturing process and their use to draw out selectively Actinium series (III) and to separate them in particular from Lanthanides (III)

    International Nuclear Information System (INIS)

    Cuillerdier, C.; Musikas, C.

    1993-01-01

    Present invention is concerned with new amides with nitrogenous heterocyclic substituent utilizable to separate trivalent actinium series from trivalent lanthanides. In these molecules, it is possible to obtain particularly covalent liaison which has more affinity with 5f series, that is to say actinium series; included a manufacturing process for these amides with nitrogenous heterocyclic substituent

  12. Analysis of the gamma spectra of the uranium, actinium, and thorium decay series

    International Nuclear Information System (INIS)

    Momeni, M.H.

    1981-09-01

    This report describes the identification of radionuclides in the uranium, actinium, and thorium series by analysis of gamma spectra in the energy range of 40 to 1400 keV. Energies and absolute efficiencies for each gamma line were measured by means of a high-resolution germanium detector and compared with those in the literature. A gamma spectroscopy method, which utilizes an on-line computer for deconvolution of spectra, search and identification of each line, and estimation of activity for each radionuclide, was used to analyze soil and uranium tailings, and ore

  13. Thermal Stabilization of 233UO2, 233UO3, and 233U3O8

    International Nuclear Information System (INIS)

    Thein, S.M.; Bereolos, P.J.

    2000-01-01

    This report identifies an appropriate thermal stabilization temperature for 233 U oxides. The temperature is chosen principally on the basis of eliminating moisture and other residual volatiles. This report supports the U. S. Department of Energy (DOE) Standard for safe storage of 233 U (DOE 2000), written as part of the response to Recommendation 97-1 of the Defense Nuclear Facilities Safety Board (DNFSB), addressing safe storage of 233 U

  14. Excited levels of Pa-233; Niveles excitados del Pa-233

    Energy Technology Data Exchange (ETDEWEB)

    Vara Cuadrado, J M

    1969-07-01

    A study of Pa-233 excited levels from the alpha decay of Np-237 and from beta decay of Th-233 has been performed. The alpha decay spectrum was measured with a semiconductor spectrometer of 18 keV effective resolution (FWHM). Over 13 new lines were identified. The gamma ray spectra of Np-237 and Th-233 were obtained with a Ge-Li detector low and medium range energy lines, and with Si-Li detector for the low energy region. A continuous purification method of Np-237 from its comparatively short-lived daughter Pa-233 was applied. A high number of new lines were identified in both spectra. The gamma-gamma coincidence spectra were obtained with INa(T{sub 1}) detectors. (Author) 54 refs.

  15. Actinium-225 and Bismuth-213 Alpha Particle Immunotherapy of Cancer

    International Nuclear Information System (INIS)

    Scheinberg, D.

    2013-01-01

    Nuclides with appropriate half-lives and emission characteristics that would be potent enough to kill neoplastic cells in the small quantities that reach targets in vivo, include the high linear energy transfer (LET) alpha emitters such as Actinium-225 and Bi-213. We developed methods for the attachment of radiometals via bifunctional chelates to monoclonal antibodies (mAb) without loss of immunoreactivity. We developed alphaemitting Bi-213 lintuzumab constructs, characterized and qualified them in preclinical models, and took them into human clinical trials in patients with AML. Safety, anti-leukemic activity, and complete responses (CR’s) have been demonstrated through phase 2 trilas. Bi-213 is produced in a portable small generator device based on Ac- 225 in the hospital nuclear medicine lab. The isotope is then purified, attached to the antibody, and the product is qualified and processed. Despite this success, the major obstacle to the widespread use of these drugs remains the short 213 Bi half-life (46 minutes), which poses a large logistical hurdle before injection and limits its delivery to only the most accessible cancer cells after injection

  16. 15 CFR 23.3 - Plan.

    Science.gov (United States)

    2010-01-01

    ... 15 Commerce and Foreign Trade 1 2010-01-01 2010-01-01 false Plan. 23.3 Section 23.3 Commerce and... MISSING CHILDREN § 23.3 Plan. (a) The Department of Commerce will supplement and expand the national... biannual meetings of departmental representatives to discuss the current plan and recommendations for...

  17. 49 CFR 233.9 - Reports.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Reports. 233.9 Section 233.9 Transportation Other... TRANSPORTATION SIGNAL SYSTEMS REPORTING REQUIREMENTS § 233.9 Reports. Not later than April 1, 1997 and every 5 years thereafter, each carrier shall file with FRA a signal system status report “Signal System Five...

  18. Computational Approach in Determination of 233U and 233Th Fermi Energy

    International Nuclear Information System (INIS)

    Kurniadi, R.; Perkasa, Y. S.; Waris, A.

    2010-01-01

    There are several methods to get Fermi energy such as hermit polynomial expansion and Wigner-Kirkwood expansion, these are analytical method. In this paper will be discussed numerical approach of calculating Fermi energy of 233 Th and 233 U nuclei. Our work demonstrates the simple technique of determining Fermi energy.

  19. 40 CFR 233.31 - Coordination requirements.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Coordination requirements. 233.31 Section 233.31 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING 404 STATE PROGRAM REGULATIONS Program Operation § 233.31 Coordination requirements. (a) If a proposed...

  20. Excited levels of Pa-233

    International Nuclear Information System (INIS)

    Vara Cuadrado, J. M.

    1969-01-01

    A study of Pa-233 excited levels from the alpha decay of Np-237 and from beta decay of Th-233 has been performed. The alpha decay spectrum was measured with a semiconductor spectrometer of 18 keV effective resolution (FWHM). Over 13 new lines were identified. The gamma ray spectra of Np-237 and Th-233 were obtained with a Ge-Li detector low and medium range energy lines, and with Si-Li detector for the low energy region. A continuous purification method of Np-237 from its comparatively short-lived daughter Pa-233 was applied. A high number of new lines were identified in both spectra. The gamma-gamma coincidence spectra were obtained with INa(T 1 ) detectors. (Author) 54 refs

  1. 7 CFR 58.233 - Skim milk.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 3 2010-01-01 2010-01-01 false Skim milk. 58.233 Section 58.233 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (Standards... Materials § 58.233 Skim milk. The skim milk shall be separated from whole milk meeting the requirements as...

  2. 14 CFR 23.3 - Airplane categories.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Airplane categories. 23.3 Section 23.3... STANDARDS: NORMAL, UTILITY, ACROBATIC, AND COMMUTER CATEGORY AIRPLANES General § 23.3 Airplane categories. (a) The normal category is limited to airplanes that have a seating configuration, excluding pilot...

  3. Preserving Ultra-Pure Uranium-233

    International Nuclear Information System (INIS)

    Krichinsky, Alan M.; Goldberg, Steven A.; Hutcheon, Ian D.

    2011-01-01

    Uranium-233 ( 233 U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium ( 232 Th). At high purities, this synthetic isotope serves as a crucial reference material for accurately quantifying and characterizing uranium-bearing materials assays and isotopic distributions for domestic and international nuclear safeguards. Separated, high purity 233 U is stored in vaults at Oak Ridge National Laboratory (ORNL). These materials represent a broad spectrum of 233 U from the standpoint of isotopic purity - the purest being crucial for precise analyses in safeguarding uranium. All 233 U at ORNL is currently scheduled to be disposed of by down-blending with depleted uranium beginning in 2015. This will reduce safety concerns and security costs associated with storage. Down-blending this material will permanently destroy its potential value as a certified reference material for use in uranium analyses. Furthermore, no credible options exist for replacing 233 U due to the lack of operating production capability and the high cost of restarting currently shut down capabilities. A study was commissioned to determine the need for preserving high-purity 233 U. This study looked at the current supply and the historical and continuing domestic need for this crucial isotope. It examined the gap in supplies and uses to meet domestic needs and extrapolated them in the context of international safeguards and security activities - superimposed on the recognition that existing supplies are being depleted while candidate replacement material is being prepared for disposal. This study found that the total worldwide need by this projection is at least 850 g of certified 233 U reference material over the next 50 years. This amount also includes a strategic reserve. To meet this need, 18 individual items totaling 959 g of 233 U were identified as candidates for establishing a lasting supply of certified reference materials (CRM), all

  4. 49 CFR 233.11 - Civil penalties.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Civil penalties. 233.11 Section 233.11..., DEPARTMENT OF TRANSPORTATION SIGNAL SYSTEMS REPORTING REQUIREMENTS § 233.11 Civil penalties. Any person (an... subject to a civil penalty of at least $650 and not more than $25,000 per violation, except that...

  5. Disposition Options for Uranium-233

    International Nuclear Information System (INIS)

    Beahm, E.C.; Dole, L.R.; Forsberg, C.W.; Icenhour, A.S.; Storch, S.N.

    1999-01-01

    The U.S. Department of Energy (DOE) Fissile Materials Disposition Program (MD), in support of the U.S. arms-control and nonproliferation policies, has initiated a program to disposition surplus weapons-usable fissile material by making it inaccessible and unattractive for use in nuclear weapons. Weapons-usable fissile materials include plutonium, high-enriched uranium (HEU), and uranium-233 (sup 233)U. In support of this program, Oak Ridge National Laboratory led DOE's contractor efforts to identify and characterize options for the long-term storage and disposal of excess (sup 233)U. Five storage and 17 disposal options were identified and are described herein

  6. Preserving high-purity 233U

    International Nuclear Information System (INIS)

    Krichinsky, Alan; Giaquinto, Joe; Canaan, Doug

    2016-01-01

    The MARC X Conference hosted a workshop for the scientific community to communicate needs for high-purity 233 U and its by-products in order to preserve critical items otherwise slated for downblending and disposal. Currently, only small portions of the U.S. holdings of separated 233 U are being preserved. However, many additional kilograms of 233 U (>97 % pure) still are destined to be disposed, and it is unlikely that this material will ever be replaced due to a lack of operating production capability. Summaries of information conveyed at the workshop and feedback obtained from the scientific community are presented herein. (author)

  7. 40 CFR 233.4 - Conflict of interest.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Conflict of interest. 233.4 Section 233.4 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN DUMPING 404 STATE PROGRAM REGULATIONS General § 233.4 Conflict of interest. Any public officer or employee who has a direct...

  8. 10 CFR 600.233 - Supplies.

    Science.gov (United States)

    2010-01-01

    ... supplies exceeding $5,000 in total aggregate fair market value upon termination or completion of the award... 10 Energy 4 2010-01-01 2010-01-01 false Supplies. 600.233 Section 600.233 Energy DEPARTMENT OF... Supplies. (a) Title. Title to supplies acquired under a grant or subgrant will vest, upon acquisition, in...

  9. 45 CFR 233.52 - Overpayment to aliens.

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Overpayment to aliens. 233.52 Section 233.52... ELIGIBILITY IN FINANCIAL ASSISTANCE PROGRAMS § 233.52 Overpayment to aliens. A State Plan under title IV-A of the Social Security Act, shall provide that: (a) Any sponsor of an alien and the alien shall be...

  10. 233S Decommissioning Project Environmental Control Plan

    International Nuclear Information System (INIS)

    Zoric, J.P.

    2000-01-01

    This Environmental Control Plan is for the 233S Decommissioning activities conducted under the removal action report for the 233S Decontamination and Demolition Project. The purpose of this ECP is to identify environmental requirements for the 233S project. The ECP is a compilation of existing environmental permit conditions, regulatory requirements, and environmental requirements applicable to the specific project or functional activity

  11. Developments towards in-gas-jet laser spectroscopy studies of actinium isotopes at LISOL

    International Nuclear Information System (INIS)

    Raeder, S.; Bastin, B.; Block, M.; Creemers, P.; Delahaye, P.; Ferrer, R.; Fléchard, X.; Franchoo, S.; Ghys, L.; Gaffney, L.P.; Granados, C.; Heinke, R.; Hijazi, L.

    2016-01-01

    To study exotic nuclides at the borders of stability with laser ionization and spectroscopy techniques, highest efficiencies in combination with a high spectral resolution are required. These usually opposing requirements are reconciled by applying the in-gas-laser ionization and spectroscopy (IGLIS) technique in the supersonic gas jet produced by a de Laval nozzle installed at the exit of the stopping gas cell. Carrying out laser ionization in the low-temperature and low density supersonic gas jet eliminates pressure broadening, which will significantly improve the spectral resolution. This article presents the required modifications at the Leuven Isotope Separator On-Line (LISOL) facility that are needed for the first on-line studies of in-gas-jet laser spectroscopy. Different geometries for the gas outlet and extraction ion guides have been tested for their performance regarding the acceptance of laser ionized species as well as for their differential pumping capacities. The specifications and performance of the temporarily installed high repetition rate laser system, including a narrow bandwidth injection-locked Ti:sapphire laser, are discussed and first preliminary results on neutron-deficient actinium isotopes are presented indicating the high capability of this novel technique.

  12. Developments towards in-gas-jet laser spectroscopy studies of actinium isotopes at LISOL

    Energy Technology Data Exchange (ETDEWEB)

    Raeder, S., E-mail: s.raeder@gsi.de [KU Leuven, Instituut voor Kern- en Stralingsfysica, Celestijnenlaan 200D, B-3001 Leuven (Belgium); Helmholtz-Institut Mainz, 55128 Mainz (Germany); GSI Helmholtzzentrum für Schwerionenforschung GmbH, Planckstraße 1, 64291 Darmstadt (Germany); Bastin, B. [GANIL, CEA/DSM-CNRS/IN2P3, B.P. 55027, 14076 Caen (France); Block, M. [Helmholtz-Institut Mainz, 55128 Mainz (Germany); GSI Helmholtzzentrum für Schwerionenforschung GmbH, Planckstraße 1, 64291 Darmstadt (Germany); Institut für Kernchemie, Johannes Gutenberg Universität, 55128 Mainz (Germany); Creemers, P. [KU Leuven, Instituut voor Kern- en Stralingsfysica, Celestijnenlaan 200D, B-3001 Leuven (Belgium); Delahaye, P. [GANIL, CEA/DSM-CNRS/IN2P3, B.P. 55027, 14076 Caen (France); Ferrer, R. [KU Leuven, Instituut voor Kern- en Stralingsfysica, Celestijnenlaan 200D, B-3001 Leuven (Belgium); Fléchard, X. [LPC Caen, ENSICAEN, Université de Caen, CNRS/IN2P3, Caen (France); Franchoo, S. [Institute de Physique Nucléaire (IPN) d’Orsay, 91406 Orsay, Cedex (France); Ghys, L. [KU Leuven, Instituut voor Kern- en Stralingsfysica, Celestijnenlaan 200D, B-3001 Leuven (Belgium); SCK-CEN, Belgian Nuclear Research Center, Boeretang 200, 2400 Mol (Belgium); Gaffney, L.P.; Granados, C. [KU Leuven, Instituut voor Kern- en Stralingsfysica, Celestijnenlaan 200D, B-3001 Leuven (Belgium); Heinke, R. [Institut für Physik, Johannes Gutenberg Universität, 55128 Mainz (Germany); Hijazi, L. [GANIL, CEA/DSM-CNRS/IN2P3, B.P. 55027, 14076 Caen (France); and others

    2016-06-01

    To study exotic nuclides at the borders of stability with laser ionization and spectroscopy techniques, highest efficiencies in combination with a high spectral resolution are required. These usually opposing requirements are reconciled by applying the in-gas-laser ionization and spectroscopy (IGLIS) technique in the supersonic gas jet produced by a de Laval nozzle installed at the exit of the stopping gas cell. Carrying out laser ionization in the low-temperature and low density supersonic gas jet eliminates pressure broadening, which will significantly improve the spectral resolution. This article presents the required modifications at the Leuven Isotope Separator On-Line (LISOL) facility that are needed for the first on-line studies of in-gas-jet laser spectroscopy. Different geometries for the gas outlet and extraction ion guides have been tested for their performance regarding the acceptance of laser ionized species as well as for their differential pumping capacities. The specifications and performance of the temporarily installed high repetition rate laser system, including a narrow bandwidth injection-locked Ti:sapphire laser, are discussed and first preliminary results on neutron-deficient actinium isotopes are presented indicating the high capability of this novel technique.

  13. Disposition of Uranium -233 (sup 233U) in Plutonium Metal and Oxide at the Rocky Flats Environmental Technology Site

    International Nuclear Information System (INIS)

    Freiboth, Cameron J.; Gibbs, Frank E.

    2000-01-01

    This report documents the position that the concentration of Uranium-233 ( 233 U) in plutonium metal and oxide currently stored at the DOE Rocky Flats Environmental Technology Site (RFETS) is well below the maximum permissible stabilization, packaging, shipping and storage limits. The 233 U stabilization, packaging and storage limit is 0.5 weight percent (wt%), which is also the shipping limit maximum. These two plutonium products (metal and oxide) are scheduled for processing through the Building 371 Plutonium Stabilization and Packaging System (PuSPS). This justification is supported by written technical reports, personnel interviews, and nuclear material inventories, as compiled in the ''History of Uranium-233 ( 233 U) Processing at the Rocky Flats Plant In Support of the RFETS Acceptable Knowledge Program'' RS-090-056, April 1, 1999. Relevant data from this report is summarized for application to the PuSPS metal and oxide processing campaigns

  14. Dicty_cDB: SHH233 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available SH (Link to library) SHH233 (Link to dictyBase) - - - Contig-U11264-1 - (Link to Or...c08.g1 Strongyloides ratti whole genome shotgun library (SRAAGSS 004) Strongyloides ratti genomic...iginal site) - - SHH233Z 563 - - - - Show SHH233 Library SH (Link to library) Clone ID SHH233 (Link to dicty...631_5( AY458631 |pid:none) Uncultured marine bacterium 159 cl... 84 3e-15 CU92816...86F1 NIH_MGC_58 Homo sapiens cDNA clone IMAGE:4069772 5', mRNA sequence. 46 0.86 1 AP008210 |AP008210.1 Oryza sativa (japonica culti

  15. 39 CFR 233.12 - Civil penalties.

    Science.gov (United States)

    2010-07-01

    ... 39 Postal Service 1 2010-07-01 2010-07-01 false Civil penalties. 233.12 Section 233.12 Postal... Civil penalties. False representation and lottery orders— (a) Issuance. Pursuant to 39 U.S.C. 3005, the... be liable to the United States for a civil penalty in an amount not to exceed $11,000 for each day...

  16. 233U Assay A Neutron NDA System

    International Nuclear Information System (INIS)

    Hensley, D.C.; Lucero, A.J.; Pierce, L.

    1998-01-01

    The assay of highly enriched 233 U material presents some unique challenges. Techniques which apply to the assay of materials of Pu or enriched 235 U do not convert easily over to the assay of 233 U. A specialized neutron assay device is being fabricated to exploit the singles neutron signal, the weak correlated neutron signal, and an active correlated signal. These pieces of information when combined with γ ray isotopics information should give a good overall determination of 233 U material now stored in bldg. 3019 at the Oak Ridge National Laboratory

  17. 233U Assay A Neutron NDA System

    Energy Technology Data Exchange (ETDEWEB)

    Hensley, D.C.; Lucero, A.J.; Pierce, L.

    1998-11-17

    The assay of highly enriched {sup 233}U material presents some unique challenges. Techniques which apply to the assay of materials of Pu or enriched {sup 235}U do not convert easily over to the assay of {sup 233}U. A specialized neutron assay device is being fabricated to exploit the singles neutron signal, the weak correlated neutron signal, and an active correlated signal. These pieces of information when combined with {gamma} ray isotopics information should give a good overall determination of {sup 233}U material now stored in bldg. 3019 at the Oak Ridge National Laboratory.

  18. 12 CFR Appendix A to Part 233 - Model Notice

    Science.gov (United States)

    2010-01-01

    ... FUNDING OF UNLAWFUL INTERNET GAMBLING (REGULATION GG) Part 233, App. A Appendix A to Part 233—Model Notice... 12 Banks and Banking 3 2010-01-01 2010-01-01 false Model Notice A Appendix A to Part 233 Banks and... that your institution processed payments through our facilities for Internet gambling transactions...

  19. 40 CFR 86.233-94-86.234-94 - [Reserved

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 18 2010-07-01 2010-07-01 false [Reserved] 86.233-94-86.234-94 Section 86.233-94-86.234-94 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR... New Medium-Duty Passenger Vehicles; Cold Temperature Test Procedures §§ 86.233-94—86.234-94 [Reserved] ...

  20. 27 CFR 24.233 - Addition of spirits to wine.

    Science.gov (United States)

    2010-04-01

    ... wine. 24.233 Section 24.233 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE BUREAU, DEPARTMENT OF THE TREASURY LIQUORS WINE Spirits § 24.233 Addition of spirits to wine. (a) Prior to the addition of spirits. Wine will be placed in tanks approved for the addition of spirits. The...

  1. 40 CFR 233.12 - Attorney General's statement.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Attorney General's statement. 233.12... STATE PROGRAM REGULATIONS Program Approval § 233.12 Attorney General's statement. (a) Any State that seeks to administer a program under this part shall submit a statement from the State Attorney General...

  2. 48 CFR 852.233-70 - Protest content/alternative dispute resolution.

    Science.gov (United States)

    2010-10-01

    .../alternative dispute resolution. 852.233-70 Section 852.233-70 Federal Acquisition Regulations System... Provisions and Clauses § 852.233-70 Protest content/alternative dispute resolution. As prescribed in 833.106, insert the following provision: Protest Content/Alternative Dispute Resolution (JAN 2008) (a) Any protest...

  3. History of Uranium-233(233U)Processing at the Rocky Flats Plant. In support of the RFETS Acceptable Knowledge Program

    International Nuclear Information System (INIS)

    Moment, R.L.; Gibbs, F.E.; Freiboth, C.J.

    1999-01-01

    This report documents the processing of Uranium-233 at the Rocky Flats Plant (Rocky Flats Environmental Technology Site). The information may be used to meet Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC)and for determining potential Uranium-233 content in applicable residue waste streams

  4. 20 CFR 410.233 - Cancellation of a request for withdrawal.

    Science.gov (United States)

    2010-04-01

    ... 20 Employees' Benefits 2 2010-04-01 2010-04-01 false Cancellation of a request for withdrawal. 410.233 Section 410.233 Employees' Benefits SOCIAL SECURITY ADMINISTRATION FEDERAL COAL MINE HEALTH AND... Entitlement; Filing of Claims and Evidence § 410.233 Cancellation of a request for withdrawal. Before or after...

  5. 17 CFR 201.233 - Depositions upon oral examination.

    Science.gov (United States)

    2010-04-01

    ... examination. 201.233 Section 201.233 Commodity and Securities Exchanges SECURITIES AND EXCHANGE COMMISSION... upon oral examination. (a) Procedure. Any party desiring to take the testimony of a witness by.... Examination and cross-examination of deponents may proceed as permitted at a hearing. The witness being...

  6. Spectral shift controlled reactors, denatured U-233/thorium cycle

    International Nuclear Information System (INIS)

    1978-05-01

    This paper presents technical and economic data on the SSCR which may be of use in the International Fuel Cycle Evaluation Program to intercompare alternative nuclear systems. Included in this paper are data on the denatured U-233/thorium cycle. This cycle shows a proliferation advantage over more classical thorium fuel cycle (e.g., highly-enriched U-235/thorium or plutonium/thorium) due to the elimination of chemically-separable, concentrated fissile material from unirradiated nuclear fuel. The U-233 is denatured by mixing with depleted uranium to a concentration no greater than 12 w/o. An exogenous source of U-233 is assumed in this paper, since U-233 does not occur in nature and only a limited supply has been produced to date for research and development work

  7. 33 CFR 136.233 - Proof.

    Science.gov (United States)

    2010-07-01

    ... SOURCE; AND ADVERTISEMENT Procedures for Particular Claims § 136.233 Proof. In addition to the... must be established. (d) Whether alternative employment or business was available and undertaken and...

  8. 45 CFR 233.51 - Eligibility of sponsored aliens.

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Eligibility of sponsored aliens. 233.51 Section... CONDITIONS OF ELIGIBILITY IN FINANCIAL ASSISTANCE PROGRAMS § 233.51 Eligibility of sponsored aliens... affidavit(s) of support or similar agreement on behalf of an alien (who is not the child of the sponsor or...

  9. Plutonium and U-233 mines

    International Nuclear Information System (INIS)

    Milgram, M.S.

    1983-08-01

    A comparison is made among second generation reactor systems fuelled primarily with fissile plutonium and/or U-233 in uranium or thorium. This material is obtained from irradiated fuel from first generation CANDU reactors fuelled by natural or enriched uranium and thorium. Except for plutonium-thorium reactors, second generation reactors demand similar amounts of reprocessing throughput, but the most efficient plutonium burning systems require a large prior allocation of uranium. Second generation reactors fuelled by U-233 make more efficient use of resources and lead to more flexible fuelling strategies, but require development of first generation once-through thorium cycles and early demonstration of the commercial viability of thorium fuel reprocessing. No early implementation of reprocessing technology is required for these cycles

  10. 48 CFR 1852.233-70 - Protests to NASA.

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 6 2010-10-01 2010-10-01 true Protests to NASA. 1852.233... 1852.233-70 Protests to NASA. As prescribed in 1833.106-70, insert the following provision: Protests to NASA (OCT 2002) Potential bidders or offerors may submit a protest under 48 CFR part 33 (FAR part 33...

  11. Uses for Uranium-233: What Should Be Kept for Future Needs?

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Lewis, L.C.

    1999-01-01

    Since the end of the cold war, the United States has been evaluating what fissile materials to keep for potential uses and what fissile materials to declare excess. There are three major fissile materials: high-enriched uranium (HEU), plutonium, and uranium-233 ( 233 U). Both HEU and plutonium were produced in large quantities for use in nuclear weapons and for reactor fuel. Uranium-233 was investigated for use in nuclear weapons and as a reactor fuel; however, it was never deployed in nuclear weapons or used commercially as a nuclear fuel. Uranium-233 has limited current uses, but it could have several future uses. Because of (1) the cost of storing 233 U and (2) arms control considerations, the U.S. government must decide how much of the existing 233 U inventory should be kept for future use and how much should be disposed of as waste. The objective of this report is to provide technical and economic input to make a use-or-dispose decision

  12. Measurement of 233U/234U ratios in contaminated groundwater using alpha spectrometry

    International Nuclear Information System (INIS)

    Harrison, Jennifer J.; Payne, Timothy E.; Wilsher, Kerry L.; Thiruvoth, Sangeeth; Child, David P.; Johansen, Mathew P.; Hotchkis, Michael A.C.

    2016-01-01

    The uranium isotope 233 U is not usually observed in alpha spectra from environmental samples due to its low natural and fallout abundance. It may be present in samples from sites in the vicinity of nuclear operations such as reactors or fuel reprocessing facilities, radioactive waste disposal sites or sites affected by clandestine nuclear operations. On an alpha spectrum, the two most abundant alpha emissions of 233 U (4.784 MeV, 13.2%; and 4.824 MeV, 84.3%) will overlap with the 234 U doublet peak (4.722 MeV, 28.4%; and 4.775 MeV, 71.4%), if present, resulting in a combined 233+234 U multiplet. A technique for quantifying both 233 U and 234 U from alpha spectra was investigated. A series of groundwater samples were measured both by accelerator mass spectrometry (AMS) to determine 233 U/ 234 U atom and activity ratios and by alpha spectrometry in order to establish a reliable 233 U estimation technique using alpha spectra. The Genie™ 2000 Alpha Analysis and Interactive Peak Fitting (IPF) software packages were used and it was found that IPF with identification of three peaks ( 234 U minor, combined 234 U major and 233 U minor, and 233 U major) followed by interference correction on the combined peak and a weighted average activity calculation gave satisfactory agreement with the AMS data across the 233 U/ 234 U activity ratio range (0.1–20) and 233 U activity range (2–300 mBq) investigated. Correlation between the AMS 233 U and alpha spectrometry 233 U was r 2  = 0.996 (n = 10). - Highlights: • Describes a technique for deconvoluting the combined 233 U and 234 U multiplet in alpha spectra. • Enables 233 U and 234 U activities and 233 U/ 234 U ratios to be quantified without requiring additional analysis and measurement. • Applicable to an environmental matrix (groundwater) using standard alpha spectrometry counting equipment, operation and set-up.

  13. 233-S Plutonium Concentration Facility data quality objectives

    International Nuclear Information System (INIS)

    Encke, D.B.

    1996-08-01

    This document is a summary of the decision-making associated with the Data Quality Objective process that pertains to the characterization activities in the 233-S Plutonium Concentration Facility at the Hanford Site in Richland, Washington. The 233-S Plutonium Concentration Facility is located adjacent to, and north of, the REDOX Plant. The facility was used to concentrate the plutonium nitrate product solution from the REDOX facility. The 233-S Pipe Gallery, Control Room, SWP Change Room, Toilet, Equipment Room and the Electrical Cubicle are currently scheduled for decontamination and cleanout to support future demolition (D and D). Identification of the radiological contamination and presence of hazardous materials is needed to allow for disposal of the D and D debris

  14. 49 CFR 233.5 - Accidents resulting from signal failure.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Accidents resulting from signal failure. 233.5... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION SIGNAL SYSTEMS REPORTING REQUIREMENTS § 233.5 Accidents resulting... by toll free telephone, number 800-424-0201, whenever it learns of the occurrence of an accident...

  15. Strategy for the future use and disposition of uranium-233: Technical information

    International Nuclear Information System (INIS)

    Bereolos, P.J.; Forsberg, C.W.; Kocher, D.C.; Krichinsky, A.M.

    1998-04-01

    This document provides a summary of technical information on the synthetic radioisotope 233 U. It is one of a series of four reports that map out a national strategy for the future use and disposition of 233 U. The technical information on 233 U in this document falls into two main areas. First, material characteristics are presented along with the contrasts of 233 U to the more well known strategic fissile materials, 235 U and plutonium (Pu). Second, information derived from the scientific information, such as safeguards, waste classifications, material form, and packaging, is presented. Throughout, the effects of isotopically diluting 233 U with nonfissile, depleted uranium (DU) are examined

  16. Water-Reflected 233U Uranyl Nitrate Solutions in Simple Geometry

    International Nuclear Information System (INIS)

    Elam, K.R.

    2001-01-01

    A number of critical experiments involving 233 U were performed in the Oak Ridge National Laboratory Building 9213 Critical Experiments Facility during the years 1952 and 1953. These experiments, reported in Reference 1, were directed toward determining bounding values for the minimum critical mass, minimum critical volume, and maximum safe pipe size of water-moderated solutions of 233 U. Additional information on the critical experiments was found in the experimental logbooks. Two experiments utilizing uranyl nitrate (UO 2 (NO 3 ) 2 ) solutions in simple geometry are evaluated in this report. Experiment 37 is in a 10.4-inch diameter sphere, and Experiment 39 is in a 10-inch diameter cylinder. The 233 U concentration ranges from 49 to 62 g 233 U/l. Both experiments were reflected by at least 6 inches of water in all directions. Paraffin-reflected uranyl nitrate experiments, also reported in Reference 1, are evaluated elsewhere. Experiments with smaller paraffin reflected 5-, 6-, and 7.5-inch diameter cylinders are evaluated in U233-SOL-THERM-004. Experiments with paraffin reflected 8-, 8.5-, 9-, 10-, and 12-inch diameter cylinders are evaluated in U233-SOL-THERM-002. Later experiments with highly-enriched 235 U uranyl fluoride solution in the same 10.4-inch diameter sphere are reported in HEU-SOL-THERM-010. Both experiments were judged acceptable for use as criticality-safety benchmark experiments

  17. 27 CFR 46.233 - Payment of floor stocks tax.

    Science.gov (United States)

    2010-04-01

    ... tax. 46.233 Section 46.233 Alcohol, Tobacco Products and Firearms ALCOHOL AND TOBACCO TAX AND TRADE...) Electronic funds transfer. If the dealer pays any other excise taxes collected by TTB by electronic funds transfer, then the dealer must also send the payment for the floor stocks tax by an electronic funds...

  18. Fast and thermal data testing of 233U critical assemblies

    International Nuclear Information System (INIS)

    Wright, R.Q.; Jordan, W.C.; Leal, L.C.

    1999-01-01

    Many sources have been used to obtain 233 U benchmark descriptions. Unfortunately, some of these are not reliable since a thorough and complete benchmark evaluation often has not been done. For 24 yr a principal source for 233 U benchmarks has been the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications. The CSEWG specifications included only two fast benchmarks and three thermal benchmarks. The thermal benchmarks were H 2 O-moderated thorium-oxide exponential lattices. Since the thorium-oxide lattices were exponential experiments, they have not been widely used. CSEWG has also used the 233 U Oak Ridge National Laboratory (ORNL) spheres for many years. One advantage of the CSEWG fast benchmarks, JEZEBEL-23 and FLATTOP-23, is that experiments were done for central-reaction-rate ratios. These reaction-rate ratios provide very valuable information to data testers and evaluators that would not otherwise be available. In recent years the International Handbook of Evaluated Criticality Safety Benchmark Experiments has, in general, been a very useful and reliable source. The Handbook does not include central-reaction-rate ratio experiments, however. A new set of 233 U benchmark experiments has been added to the most recent release of the Handbook, U233-SOL-THERM-004. These are paraffin-reflected cylinders of 233 U uranyl-nitrate solutions. Unfortunately, the estimated benchmark uncertainties are on the order of 0.9 to 1.0% in k eff . Benchmark testing has been done for some of these U233-SOL-THERM-004 experiments. The authors have also discovered that the benchmark specifications for the Thomas uranyl-nitrate experiments given in Ref. 5 are incorrect. One problem with the Ref. 5 specifications is that the excess acid was not included. As part of this work, the authors developed revised specifications that include an excess acid correlation based on information from the experimental logbook

  19. 28 CFR 23.3 - Applicability.

    Science.gov (United States)

    2010-07-01

    ... Administration DEPARTMENT OF JUSTICE CRIMINAL INTELLIGENCE SYSTEMS OPERATING POLICIES § 23.3 Applicability. (a) These policy standards are applicable to all criminal intelligence systems operating through support...-647). (b) As used in these policies: (1) Criminal Intelligence System or Intelligence System means the...

  20. 45 CFR 233.32 - Payment and budget months (AFDC).

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Payment and budget months (AFDC). 233.32 Section... CONDITIONS OF ELIGIBILITY IN FINANCIAL ASSISTANCE PROGRAMS § 233.32 Payment and budget months (AFDC). A State... period used to determine that payment (budget month) and whether it adopts (a) a one-month or two-month...

  1. Initial ORNL site assessment report on the storage of 233U

    International Nuclear Information System (INIS)

    Bereolos, P.J.; Yong, L.K.; Sadlowe, A.R.; Ramey, D.W.; Krichinsky, A.M.

    1998-03-01

    The 233 U storage facility at ORNL is Building 3019. The inventory stored in Building 3019 consists of 426.5 kg of 233 U contained in 1,387.1 kg of total uranium. The inventory is primarily in the form of uranium oxides; however, uranium metal and other compounds are also stored. Over 99% of the inventory is contained in 1,007 packages stored in tube vaults within the facility. A tank of thorium nitrate solution, the P-24 Tank, contains 0.13 kg of 233 U in ∼ 4,000 gal. of solution. The facility is receiving additional 233 U for storage from the remediation of the Molten Salt Reactor Experiment (MSRE) at ORNL. Consolidation of material from sites with small holdings is also adding to the 233 U inventory. Additionally, small quantities ( 233 U are in other research facilities at ORNL. A risk assessment process was chosen to evaluate the stored material and packages based on available package records. The risk scenario was considered the failure of a package (or a group of similar packages) in the Building 3019 inventory. The probability of such a failure depends on packaging factors such as the age and material of construction of the containers. The consequence of such a failure depends on the amount and form of the material within the packages. One thousand seven packages were categorized with this methodology resulting in 859 low-risk packages, 147 medium-risk packages, and 1 high-risk package. This initial assessment also documents the status of the evaluation of the Building 3019 and its systems for safe storage of 233 U. The final assessment report for ORNL storage of 233 U is scheduled for June 1999. The report will document the facility assessments, the specific package inspection plan, and the results of initial package inspections

  2. Isotopic dilution requirements for 233U criticality safety in processing and disposal facilities

    International Nuclear Information System (INIS)

    Elam, K.R.; Forsberg, C.W.; Hopper, C.M.; Wright, R.Q.

    1997-11-01

    The disposal of excess 233 U as waste is being considered. Because 233 U is a fissile material, one of the key requirements for processing 233 U to a final waste form and disposing of it is to avoid nuclear criticality. For many processing and disposal options, isotopic dilution is the most feasible and preferred option to avoid nuclear criticality. Isotopic dilution is dilution of fissile 233 U with nonfissile 238 U. The use of isotopic dilution removes any need to control nuclear criticality in process or disposal facilities through geometry or chemical composition. Isotopic dilution allows the use of existing waste management facilities, that are not designed for significant quantities of fissile materials, to be used for processing and disposing of 233 U. The amount of isotopic dilution required to reduce criticality concerns to reasonable levels was determined in this study to be ∼ 0.66 wt% 233 U. The numerical calculations used to define this limit consisted of a homogeneous system of silicon dioxide (SiO 2 ), water (H 2 O), 233 U, and depleted uranium (DU) in which the ratio of each component was varied to determine the conditions of maximum nuclear reactivity. About 188 parts of DU (0.2 wt% 235 U) are required to dilute 1 part of 233 U to this limit in a water-moderated system with no SiO 2 present. Thus, for the US inventory of 233 U, several hundred metric tons of DU would be required for isotopic dilution

  3. 47 CFR 90.233 - Base/mobile non-voice operations.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 5 2010-10-01 2010-10-01 false Base/mobile non-voice operations. 90.233... SERVICES PRIVATE LAND MOBILE RADIO SERVICES Non-Voice and Other Specialized Operations § 90.233 Base/mobile non-voice operations. The use of A1D, A2D, F1D, F2D, G1D, or G2D emission may be authorized to base...

  4. 5 CFR 532.233 - Preparation for full-scale wage surveys.

    Science.gov (United States)

    2010-01-01

    ... presence on the job, and the prudent management of available financial and human resources. Employing.... 532.233 Section 532.233 Administrative Personnel OFFICE OF PERSONNEL MANAGEMENT CIVIL SERVICE... and jobs to be covered in the wage survey. (2) Shall prepare a summary of the hearings and submit it...

  5. 9 CFR 2.33 - Attending veterinarian and adequate veterinary care.

    Science.gov (United States)

    2010-01-01

    ... veterinary care. 2.33 Section 2.33 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE... adequate veterinary care. (a) Each research facility shall have an attending veterinarian who shall provide adequate veterinary care to its animals in compliance with this section: (1) Each research facility shall...

  6. 48 CFR 5452.233-9001 - Disputes: Agreement To Use Alternative Dispute Resolution (ADR).

    Science.gov (United States)

    2010-10-01

    ... Alternative Dispute Resolution (ADR). 5452.233-9001 Section 5452.233-9001 Federal Acquisition Regulations... of Provisions and Clauses 5452.233-9001 Disputes: Agreement To Use Alternative Dispute Resolution... Alternative Dispute Resolution (ADR) (APR 2001)—DLAD (a) The parties agree to negotiate with each other to try...

  7. 19 CFR 10.233 - Articles eligible for preferential tariff treatment.

    Science.gov (United States)

    2010-04-01

    ... control of the customs authority of the intermediate country; (ii) Did not enter into the commerce of the... 19 Customs Duties 1 2010-04-01 2010-04-01 false Articles eligible for preferential tariff treatment. 10.233 Section 10.233 Customs Duties U.S. CUSTOMS AND BORDER PROTECTION, DEPARTMENT OF HOMELAND...

  8. 48 CFR 2852.233-70 - Protests filed directly with the Department of Justice.

    Science.gov (United States)

    2010-10-01

    ... with the Department of Justice. 2852.233-70 Section 2852.233-70 Federal Acquisition Regulations System DEPARTMENT OF JUSTICE Clauses and Forms SOLICITATION PROVISIONS AND CONTRACT CLAUSES Text of Provisions and Clauses 2852.233-70 Protests filed directly with the Department of Justice. As prescribed in 2833.102(d...

  9. 25 CFR 23.3 - Policy.

    Science.gov (United States)

    2010-04-01

    ... AFFAIRS, DEPARTMENT OF THE INTERIOR HUMAN SERVICES INDIAN CHILD WELFARE ACT Purpose, Definitions, and Policy § 23.3 Policy. In enacting the Indian Child Welfare Act of 1978, Pub. L. 95-608, the Congress has declared that it is the policy of this Nation to protect the best interests of Indian children and to...

  10. Criticality considerations for 233U fuels in an HTGR fuel refabrication facility

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1978-01-01

    Eleven 233 U solution critical assemblies spanning an H/ 233 U ratio range of 40 to 2000 and a bare metal 233 U assembly have been calculated with the ENDF/B-IV and Hansen-Roach cross sections. The results from these calculations are compared with the experimental results and with each other. An increasing disagreement between calculations with ENDF/B and Hansen-Roach data with decreasing H/ 233 U ratio was observed, indicative of large differences in their intermediate energy cross sections. The Hansen-Roach cross sections appeared to give reasonably good agreement with experiments over the whole range; whereas the ENDF/B calculations yielded high values for k/sub eff/ on assemblies of low moderation. It is concluded that serious problems exist in the ENDF/B-IV representation of the 233 U cross sections in the intermediate energy range and that further evaluation of this nuclide is warranted. In addition, it is recommended that an experimental program be undertaken to obtain 233 U criticality data at low H/ 233 U ratios for verification of generalized criticality safety guidelines. Part II of this report presents the results of criticality calculations on specific pieces of equipment required for HTGR fuel refabrication. In particular, fuel particle storage hoppers and resin carbonization furnaces are criticality safe up to 22.9 cm (9.0 in.) in diameter providing water or other hydrogenous moderators are excluded. In addition, no criticality problems arise due to accumulation of particles in the off-gas scrubber reservoirs provided reasonable administrative controls are exercised

  11. Experimental 233U nondestructive assay with a random driver

    International Nuclear Information System (INIS)

    Goris, P.

    1979-01-01

    Nondestructive assay (NDA) of 233 U in quantities up to 15 grams containing 7 ppM 232 U age 2 years was investigated with a random driver. A passive singles counting technique showed a reproducibility within 0.2% at the 95% confidence level. This technique would be applicable throughout a process in which all of the 233 U had the same 232 U content at the same age. Where the 232 U content varies, determination of 233 U fissile content would require active NDA. Active coincidence counting utilizing a 238 Pu, Li neutron source and a plastic scintillator detector system showed a reproducibility limit within 15% at the 95% confidence limit. The active technique was found to be very dependent on the detector system resolving time in order to make proper random coincidence corrections associated with the high gamma activity from the 232 U decay chain

  12. An extraction method of uranium 233 from the thorium irradiates in a reactor core; Une methode d'extraction de l'uranium-233 a partir du thorium irradie dans une pile

    Energy Technology Data Exchange (ETDEWEB)

    Chesne, A; Regnaut, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    Description of the conditions of separation of the thorium, of the uranium 233 and of the protactinium 233 in hydrochloric solution by absorption then selective elution on anion exchange resin. A precipitation of the thorium by the oxalic acid permits the recuperation of the hydrochloric acid which is recycled, the main, raw material consumed being the oxalic acid. (authors) [French] Description des conditions de separation du thorium, de l'uranium 233 et du protactinium 233 en solution chlorhydrique par absorption puis elution selective sur resine echangeuse d'anions. Une precipitation du thoriun par l'acide oxalique permet la recuperation de l'acide chlorhydrique qui est recycle, la principale matiere premiere consommee etant l'acide oxalique. (auteurs)

  13. Sources of neutronics data involving thorium of 233U and light water moderation

    International Nuclear Information System (INIS)

    Davenport, L.C.

    1978-11-01

    A literature search has been conducted to locate sources of neutronics data for light water moderated systems which contain thorium and/or uranium-233. It is concluded that insufficient data is currently available to validate neutronics design methods for licensing the 233 UO 2 -ThO 2 fuel cycle in light water reactors. A summary of the neutronics data sources found is reported in this document. These sources include critical and exponential experiments with lattices of fuel rods containing 233 U + Th or 235 U + Th. A few experiments using homogeneous aqueous solutions of 233 UO 2 (NO 3 ) 2 or 233 UO 2 F 2 are also included. The only critical lattice data using both 233 U and Th came from the LWBR program. All these experiments were zoned radially and in most cases axially also. Geometrically clean lattice critical data were measured for the CETR and TUPE programs. Both series used 235 UO 2 -ThO 2 pellets. A series of 21 exponential experiments using 3% 233 UO 2 - 97% ThO 2 fuel vibratory compacted to 92% of theoretical density in Zircaloy-2 tubing was performed at BNL using both unpoisoned and boric acid poisoned H 2 O moderator. For completeness, homogeneous systems are listed in which basic neutronics data have been measured. However, it is expected that most data concerning homogeneous systems will be applied to criticality safety problems rather than neutronics methods validation

  14. Th/U-233 multi-recycle in PWRs

    International Nuclear Information System (INIS)

    Yun, D.; Kim, T.K.; Taiwo, T.A.

    2010-01-01

    The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle including: (1) its use in a once-through fuel cycle to replace non-fissile uranium or to extend fuel burnup due to its attractive fertile material conversion, (2) its use for fissile plutonium burning in limited recycle cores, and (3) its advantage in limiting the transuranic elements to be disposed off in a repository (if only Th/U-233 fuel is used). The possibility for thorium utilization in multirecycle system has also been considered by various researchers, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this project is to evaluate the potential of the Th/U-233 fuel multirecycle in current LWRs, with focus this year on pressurized water reactors (PWRs). In this work, approaches for ensuring a sustainable multirecycle without the need for external source of makeup fissile material have been investigated. The intent is to achieve a design that allows existing PWRs to be used with minimal modifications. In all cases including homogeneous and heterogeneous assembly designs, the assembly pitch is kept consistent with that of the current PWRs (21.5 cm used). Because of design difficulties associated with using the same geometry and dimensions as a PWR core, the potential modifications (other than assembly pitch) that would be needed for PWRs to ensure a sustainable multirecycle system have been investigated and characterized. Additionally, the implications of the use of thorium on the LWR fuel cycle are discussed. In Section 2, background information on studies evaluating the use of thorium in the fuel cycle is provided, but focusing on

  15. Th/U-233 multi-recycle in PWRs.

    Energy Technology Data Exchange (ETDEWEB)

    Yun, D.; Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-09-07

    The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle including: (1) its use in a once-through fuel cycle to replace non-fissile uranium or to extend fuel burnup due to its attractive fertile material conversion, (2) its use for fissile plutonium burning in limited recycle cores, and (3) its advantage in limiting the transuranic elements to be disposed off in a repository (if only Th/U-233 fuel is used). The possibility for thorium utilization in multirecycle system has also been considered by various researchers, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this project is to evaluate the potential of the Th/U-233 fuel multirecycle in current LWRs, with focus this year on pressurized water reactors (PWRs). In this work, approaches for ensuring a sustainable multirecycle without the need for external source of makeup fissile material have been investigated. The intent is to achieve a design that allows existing PWRs to be used with minimal modifications. In all cases including homogeneous and heterogeneous assembly designs, the assembly pitch is kept consistent with that of the current PWRs (21.5 cm used). Because of design difficulties associated with using the same geometry and dimensions as a PWR core, the potential modifications (other than assembly pitch) that would be needed for PWRs to ensure a sustainable multirecycle system have been investigated and characterized. Additionally, the implications of the use of thorium on the LWR fuel cycle are discussed. In Section 2, background information on studies evaluating the use of thorium in the fuel cycle is provided, but focusing on

  16. 45 CFR 233.28 - Monthly reporting.

    Science.gov (United States)

    2010-10-01

    ... ELIGIBILITY IN FINANCIAL ASSISTANCE PROGRAMS § 233.28 Monthly reporting. (a) State plans specifying... information requested on the form, and provides a telephone number for this purpose; (4) Includes a statement, to be signed by the recipient, that he or she understands that the information he or she provides may...

  17. 40 CFR 233.11 - Program description.

    Science.gov (United States)

    2010-07-01

    ... organization and structure of the State agency (agencies) which will have responsibility for administering the... under § 233.10 shall include: (a) A description of the scope and structure of the State's program. The... will coordinate its enforcement strategy with that of the Corps and EPA; (h) A description of the...

  18. Compilation of criticality data involving thorium or 233U and light water moderation

    Energy Technology Data Exchange (ETDEWEB)

    Gore, B.F.

    1978-07-01

    The literature has been searched for criticality data for light water moderated systems which contain thorium or /sup 233/U, and data found are compiled herein. They are from critical experiments, extrapolations, and exponential experiments performed with homogeneous solutions and metal spheres of /sup 233/U; with lattices of fuel rods containing highly enriched /sup 235/UO/sub 2/ - ThO/sub 2/ and /sup 233/UO/sub 2/ - ThO/sub 2/; and with arrays of cyclinders of /sup 233/U solutions. The extent of existing criticality data has been compared with that necessary to implement a thorium-based fuel cycle. No experiments have been performed with any solutions containing thorium. Neither do data exist for homogeneous /sup 233/U systems with H/U < 34, except for solid metal systems. Arrays of solution cylinders up to 3 x 3 x 3 have been studied. Data for solutions containing fixed or soluble poisons are very limited. All critical lattices using /sup 233/UO/sub 2/ - ThO/sub 2/ fuels (LWBR program) were zoned radially, and in most cases axially also. Only lattice experiments using /sup 235/UO/sub 2/ - ThO/sub 2/ fuels have been performed using a single fuel rod type. Critical lattices of /sup 235/UO/sub 2/ - ThO/sub 2/ rods poisoned with boron have been measured, but only exponential experiments have been performed using boron-poisoned lattices of /sup 233/UO/sub 2/ - ThO/sub 2/ rods. No criticality data exist for denatured fuels (containing significant amounts of /sup 238/U) in either solution or lattice configurations.

  19. 39 CFR 233.3 - Mail covers.

    Science.gov (United States)

    2010-07-01

    .... For purpose of these regulations, the following terms are hereby defined. (1) Mail cover is the... criminal law. (3) When time is of the essence, the Chief Postal Inspector, or designee, may act upon an... furnish information as defined in § 233.3(c)(1) to any person, except as authorized by a mail cover order...

  20. Recovery of 233U from irradiated J rods - operating experience

    International Nuclear Information System (INIS)

    Lakshmanan, K.; Natarajan, D.; Muthukumar, M.; Halder, Surajit; Jayakrishnan, G.; Selvarasan, M.; Kuppusamy, V.; Ganesan, V.; Vijayakumar, V.

    2000-01-01

    The first campaign of reprocessing was completed in 1988 and 233 U was delivered for fabrication of fuel for KAMINI. After revamping the facility, the second campaign was started in Dec 1998 and has processed some of the high density thoria rods from CIRUS successfully. Currently the campaign is in progress and it is planned to process the irradiated thorium rods from Dhruva. Interim 23 process selective to 233 U is adopted for separation of uranium from thorium and fission products

  1. Interim assessment of the denatured 233U fuel cycle: feasibility and nonproliferation characteristics

    International Nuclear Information System (INIS)

    Abbott, L.S.; Bartine, D.E.; Burns, T.J.

    1979-12-01

    A fuel cycle that employs 233 U denatured with 238 U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured 233 U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured 233 U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured 233 U fuel and are based on the energy center concept are evaluated

  2. Uranium-233 waste definition: Disposal options, safeguards, criticality control, and arms control

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Storch, S.N.; Lewis, L.C.

    1998-01-01

    The US investigated the use of 233 U for weapons, reactors, and other purposes from the 1950s into the 1970s. Based on the results of these investigations, it was decided not to use 233 U on a large scale. Most of the 233 U-containing materials were placed in long-term storage. At the end of the cold war, the US initiated, as part of its arms control policies, a disposition program for excess fissile materials. Other programs were accelerated for disposal of radioactive wastes placed in storage during the cold war. Last, potential safety issues were identified related to the storage of some 233 U-containing materials. Because of these changes, significant activities associated with 233 U-containing materials are expected. This report is one of a series of reports to provide the technical bases for future decisions on how to manage this material. A basis for defining when 233 U-containing materials can be managed as waste and when they must be managed as concentrated fissile materials has been developed. The requirements for storage, transport, and disposal of radioactive wastes are significantly different than those for fissile materials. Because of these differences, it is important to classify material in its appropriate category. The establishment of a definition of what is waste and what is fissile material will provide the guidance for appropriate management of these materials. Wastes are defined in this report as materials containing sufficiently small masses or low concentrations of fissile materials such that they can be managed as typical radioactive waste. Concentrated fissile materials are defined herein as materials containing sufficient fissile content such as to warrant special handling to address nuclear criticality, safeguards, and arms control concerns

  3. Interim assessment of the denatured 233U fuel cycle: feasibility and nonproliferation characteristics

    International Nuclear Information System (INIS)

    Abbott, L.S.; Bartine, D.E.; Burns, T.J.

    1978-12-01

    A fuel cycle that employs 233 U denatured with 238 U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured 233 U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured 233 U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured 233 U fuel and are based on the energy center concept are evaluated. Under this concept, dispersed power reactors fueled with denatured or low-enriched uranium fuel are supported by secure energy centers in which sensitive activities of the nuclear cycle are performed. These activities include 233 U production by Pu-fueled transmuters (thermal or fast reactors) and reprocessing. A summary chapter presents the most significant conclusions from the study and recommends areas for future work

  4. TRASH TO TREASURE: CONVERTING COLD WAR LEGACY WASTE INTO WEAPONS AGAINST CANCER

    International Nuclear Information System (INIS)

    Nicholas, R.G.; Lacy, N.H.; Butz, T.R.; Brandon, N.E.

    2004-01-01

    As part of its commitment to clean up Cold War legacy sites, the U.S. Department of Energy (DOE) has initiated an exciting and unique project to dispose of its inventory of uranium-233 (233U) stored at Oak Ridge National Laboratory (ORNL), and extract isotopes that show great promise in the treatment of deadly cancers. In addition to increasing the supply of potentially useful medical isotopes, the project will rid DOE of a nuclear concern and cut surveillance and security costs. For more than 30 years, DOE's ORNL has stored over 1,200 containers of fissile 233U, originally produced for several defense-related projects, including a pilot study that looked at using 233U as a commercial reactor fuel. This uranium, designated as special nuclear material, requires expensive security, safety, and environmental controls. It has been stored at an ORNL facility, Building 3019A, that dates back to the Manhattan Project. Down-blending the material to a safer form, rather than continuing to store it, will eliminate a $15 million a year financial liability for the DOE and increase the supply of medical isotopes by 5,700 percent. During the down-blending process, thorium-229 (229Th) will be extracted. The thorium will then be used to extract actinium-225 (225Ac), which will ultimately supply its progeny, bismuth-213 (213Bi), for on-going cancer research. The research includes Phase II clinical trials for the treatment of acute myelogenous leukemia at Sloan-Kettering Memorial Cancer Center in New York, as well as other serious cancers of the lungs, pancreas, and kidneys using a technique known as alpha-particle radioimmunotherapy. Alpha-particle radioimmunotherapy is based on the emission of alpha particles by radionuclides. 213Bi is attached to a monoclonal antibody that targets specific cells. The bismuth then delivers a high-powered but short-range radiation dose, effectively killing the cancerous cells but sparing the surrounding tissue. Production of the actinium and

  5. Criticality safety validation: Simple geometry, single unit 233U systems

    International Nuclear Information System (INIS)

    Putman, V.L.

    1997-06-01

    Typically used LMITCO criticality safety computational methods are evaluated for suitability when applied to INEEL 233 U systems which reasonably can be modeled as simple-geometry, single-unit systems. Sixty-seven critical experiments of uranium highly enriched in 233 U, including 57 aqueous solution, thermal-energy systems and 10 metal, fast-energy systems, were modeled. These experiments include 41 cylindrical and 26 spherical cores, and 41 reflected and 26 unreflected systems. No experiments were found for intermediate-neutron-energy ranges, or with interstitial non-hydrogenous materials typical of waste systems, mixed 233 U and plutonium, or reflectors such as steel, lead, or concrete. No simple geometry experiments were found with cubic or annular cores, or approximating infinite sea systems. Calculations were performed with various tools and methodologies. Nine cross-section libraries, based on ENDF/B-IV, -V, or -VI.2, or on Hansen-Roach source data, were used with cross-section processing methods of MCNP or SCALE. The k eff calculations were performed with neutral-particle transport and Monte Carlo methods of criticality codes DANT, MCNP 4A, and KENO Va

  6. U-233 fuelled low critical mass solution reactor experiment PURNIMA II

    International Nuclear Information System (INIS)

    Srinivasan, M.; Chandramoleshwar, K.; Pasupathy, C.S.; Rasheed, K.K.; Subba Rao, K.

    1987-01-01

    A homogeneous U-233 uranyl nitrate solution fuelled BeO reflected, low critical mass reactor has been built at the Bhabha Atomic Research Centre, India. Christened PURNIMA II, the reactor was used for the study of the variation of critical mass as a function of fuel solution concentration to determine the minimum critical mass achievable for this geometry. Other experiments performed include the determination of temperature coefficient of reactivity, study of time behaviour of photoneutrons produced due to interaction between decaying U-233 fission product gammas and the beryllium reflector and reactor noise measurements. Besides being the only operational U-233 fuelled reactor at present, PURNIMA II also has the distinction of having attained the lowest critical mass of 397 g of fissile fuel for any operating reactor at the current time. The paper briefly describes the facility and gives an account of the experiments performed and results achieved. (author)

  7. Study of electrodeposition technique to prepare alpha-counting plates of uranium 233

    International Nuclear Information System (INIS)

    Mertzig, W.

    1979-01-01

    The electrodeposition technique to prepare alpha-counting plates of 233 U for its determination is presented. To determine the optimum conditions for plating 233 U the effects of such parameters as current density, pH of eletrotype, salt concentration, time of electrolysis and distance electrodes were studied. A carrier method was developed to attain a quantitative electrodeposition of 233 U from aqueous solutions into alpha counting paltes. A single and incremental addition of natural uranium and thorium as carrier were studied. All samples were prepared using a electrodeposition cell manufactured at the IPEN, especially for use in electroplating tracer actinides. This cell is made of a metal-lucite to contain the electrolyte, which bottom is a polished brass disk coated with a Ni film serving as the cathode. A Pt wire anode is fixed on the top of the cell. The electroplated samples were alpha-counted using a surface barrier detector. A recovery of more than 99% was obtained in specific conditions. The plating procedure produced deposits which were firmly distributed over the plate area. The method was applied to determine tracer amounts of 233 U from oxalate and nitrate solutions coming from chemical processing irradiated thorium. (Author) [pt

  8. Integral benchmark test of JENDL-4.0 for U-233 systems with ICSBEP handbook

    International Nuclear Information System (INIS)

    Kuwagaki, Kazuki; Nagaya, Yasunobu

    2017-03-01

    The integral benchmark test of JENDL-4.0 for U-233 systems using the continuous-energy Monte Carlo code MVP was conducted. The previous benchmark test was performed only for U-233 thermal solution and fast metallic systems in the ICSBEP handbook. In this study, MVP input files were prepared for uninvestigated benchmark problems in the handbook including compound thermal systems (mainly lattice systems) and integral benchmark test was performed. The prediction accuracy of JENDL-4.0 was evaluated for effective multiplication factors (k eff 's) of the U-233 systems. As a result, a trend of underestimation was observed for all the categories of U-233 systems. In the benchmark test of ENDF/B-VII.1 for U-233 systems with the ICSBEP handbook, it is reported that a decreasing trend of calculated k eff values in association with a parameter ATFF (Above-Thermal Fission Fraction) is observed. The ATFF values were also calculated in this benchmark test of JENDL-4.0 and the same trend as ENDF/B-VII.1 was observed. A CD-ROM is attached as an appendix. (J.P.N.)

  9. Narrowband tunable laser for uranium-233 cleanup process

    International Nuclear Information System (INIS)

    Singh, Sunita; Sridhar, G.; Rawat, V.S.; Kawde, Nitin; Sinha, A.K.; Bhatt, S.; Gantayet, L.M.

    2009-01-01

    Design, development and technology demonstration of proto type Single Longitudinal Mode pulsed tunable laser is reported in this work. The tunable laser has a narrow bandwidth less than 400 MHz required for isotopic clean up of 233 U. (author)

  10. Strategy for the future use and disposition of Uranium-233: History, inventories, storage facilities, and potential future uses

    International Nuclear Information System (INIS)

    Bereolos, P.J.; Lewis, L.C.

    1998-06-01

    This document provides background information on the man-made radioisotope 233 U. It is one of a series of four reports that map out potential national strategies for the future use and disposition of 233 U pending action under the National Environmental Policy Act (NEPA). The scope of this report is separated 233 U, where separated refers to nonwaste 233 U or 233 U that has been separated from fission products. Information on other 233 U, such as that in spent nuclear fuel (SNF), is included only to recognize that it may be separated at a later date and then fall under the scope of this report. The background information in this document includes the historical production and current inventory of 233 U, the uses of 233 U, and a discussion of the available facilities for storing 233 U. The considerations for what fraction of the current inventory should be preserved for future use depend on several issues. First, 233 U always contains a small amount of the contaminant isotope 232 U. The decay products of 232 U are highly radioactive and require special handling. The current inventory has a variety of qualities with regard to 232 U content, ranging from 1 to about 200 ppm (on a total uranium basis). It is preferable to use 233 U with the minimum amount of 232 U in all applications. The second issue pertains to other isotopes of uranium mixed in with the 233 U, specifically 235 U and 238 U. A large portion of the inventory has a high quantity of 235 U associated with it. The presence of bulk amounts of 235 U complicates storage because of the added volume needing safeguards and criticality controls. Isotopic dilution using DU may remove safeguards and criticality concerns, but it increases the overall mass and may limit applications that depend on the fissile nature of 233 U. The third issue concerns the packaging of the material. There is no standard packaging (although one is being developed); consequently, the inventory exists in a variety of packages. For some

  11. Production of 232,233Pa in 6Li+232Th Collisions in the Classical Trajectory Approach

    International Nuclear Information System (INIS)

    Aleshin, V.P.

    2000-01-01

    The semiclassical model of nuclear reactions with loosely bound projectiles (V.P. Aleshin, B.I. Sidorenko, Acta Phys. Pol. B29, 325 (1998)) is refined and compared with experimental data of Rama Rao et al. on the excitation function for the production of 232,233 Pa in 6 Li+ 232 Th collisions at E = 30-50 MeV. The main contribution to the production of 232 Pa is the 2 neutron emission from excited states of 234 Pa formed in the ( 6 Li,α) reaction. The main source of 233 Pa is the ( 6 Li,αp) reaction followed by γ transitions from excited states of 233 Th to 233 Th (g.s.) which transforms to 233 Pa through β - decay. The ground state of 6 Li regarded as a combination of n+p+α is modeled with the K=2, l x =l y =0 hyperspherical function. The calculation underpredicts the excitation function of 232 Pa by a factor of 0.6 and overpredicts the excitation function of 233 Pa by a factor of 2.3, on the average. With the more realistic wave function of 6 Li both factors are expected to be closer to 1. (author)

  12. Studies on the recovery of 233U from phosphate containing aqueous waste using DBDECMP as extractant

    International Nuclear Information System (INIS)

    Sagar, V.B.; Oak, M.S.; Pawar, S.M.; Sivaramakrishnan, C.K.; Patil, S.K.

    1990-01-01

    A method for the recovery and purification of 233 U from phosphate containing analytical waste is developed. Extraction studies with Di-butyl N,N-diethylcarbamoylmethylphosphonate (DBDECMP) in xylene were carried out to explore the feasibility of separation and purification of 233 U from such wastes. Based on the data obtained, optimum conditions for the recovery of 233 U are suggested. (author) 11 refs.; 1 fig.; 3 tabs

  13. Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Bartine, D.E.; Burns, T.J. (eds.)

    1978-12-01

    A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated. Under this concept, dispersed power reactors fueled with denatured or low-enriched uranium fuel are supported by secure energy centers in which sensitive activities of the nuclear cycle are performed. These activities include /sup 233/U production by Pu-fueled transmuters (thermal or fast reactors) and reprocessing. A summary chapter presents the most significant conclusions from the study and recommends areas for future work.

  14. Final Oak Ridge National Laboratory Site Assessment Report on the Storage of 233U

    International Nuclear Information System (INIS)

    Bereolos, P.J.; Yong, L.K.

    1999-01-01

    This assessment characterizes the 233 U inventories and storage facility at Oak Ridge National Laboratory (ORNL). This assessment is a commitment in the U.S. Department of Energy (DOE) Implementation Plan (IP), ''Safe Storage of Uranium-233,'' in response to the Defense Nuclear Facilities Safety Board's Recommendation 97-1

  15. Scoping studies of 233U breeding fusion fission hybrid

    International Nuclear Information System (INIS)

    Maniscalco, J.A.; Hansen, L.F.; Allen, W.O.

    1978-05-01

    Neutronic calculations have been carried out in order to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (greater than or equal to 1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approximately 4). Two hybrid blankets, a thorium and a uranium-thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The overall performance of the two laser fusion driven 233 U producers is discussed and estimates are given of (1) the number of equivalent thermal power fission reactors (LWR, HWR, SSCR and HTGR) that these fusion breeders can fuel, (2) their capital cost, and (3) the cost of electricity in the combined fusion breder-converter reactor scenario

  16. Scoping studies of 233U breeding fusion fission hybrid

    International Nuclear Information System (INIS)

    Maniscalco, J.A.; Hansen, L.F.; Allen, W.O.

    1978-01-01

    Neutronic calculations have been carried out in order to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (greater than or equal to 1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approx. 4). Two hybrid blankets, a thorium and a uranium--thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The overall performance of the two laser fusion driven 233 U producers is discussed and estimates are given of (1) the number of equivalent thermal power fission reactors (LWR, HWR, SSCR and HTGR) that these fusion breeders can fuel, (2) their capital cost, and (3) the cost of electricity in the combined fusion breeder-converter reactor scenario

  17. Induction of keratinocyte migration by ECa 233 is mediated through FAK/Akt, ERK, and p38 MAPK signaling.

    Science.gov (United States)

    Singkhorn, Sawana; Tantisira, Mayuree H; Tanasawet, Supita; Hutamekalin, Pilaiwanwadee; Wongtawatchai, Tulaporn; Sukketsiri, Wanida

    2018-03-13

    Centella asiatica is widely considered the most important medicinal plant for treating and relieving skin diseases. Recently developed standardized extract of Centella asiatica ECa 233 has demonstrated positive effects on wound healing of incision and burn wound in rats. However, knowledge associated with wound healing mechanism of ECa 233 was scare. Therefore, this study aimed to investigate the effect and underlying molecular mechanisms of ECa 233 on the migration of a human keratinocyte cell line (HaCaT) using scratch wound healing assay. Formation of filopodia, a key protein in cell migration as well as signaling pathways possibly involved were subsequently assessed. It was found that HaCaT cell migration was significantly enhanced by ECa 233 in a concentration- and time-dependent manner. The filopodia formations were accordingly increased in exposure to ECa 233 at concentrations of 0.1-100 μg/ml. Furthermore, ECa 233 was found to significantly upregulate the expression of Rac1 and RhoA and to induce phosphorylation of FAK and Akt as well as ERK and p38 MAPK. Taken all together, it is suggestive that ECa 233 induces cell migration and subsequently promotes wound healing activity, through the activation of FAK, Akt, and MAPK signaling pathways thereby supporting the role of ECa 233 to be further developed for the clinical treatment of wound. Copyright © 2018 John Wiley & Sons, Ltd.

  18. Neutron data evaluation of 233U

    International Nuclear Information System (INIS)

    Maslov, V.M.; Tetereva, N.A.; Kagalenko, A.B.; Kornilov, N.V.; Baba, M.; Hasegawa, A.

    2003-08-01

    Consistent evaluation of 233 U measured data base is performed. Hauser-Feshbach- Moldauer theory, coupled channel model and double-humped fission barrier model are employed. Total, differential scattering, fission and (n,xn) data are calculated using fission cross section data description as a major constraint. The direct excitation of ground state is calculated within rigid rotator model. Average resonance parameters are provided, which reproduce evaluated cross sections in the range of 0.6-40.5 keV. (author)

  19. High conversion Th-U{sup 233} fuel assembly for current generation of PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Baldova, D.; Fridman, E. [Reactor Safety Div., Helmholtz-Zentrum Dresden-Rossendorf, POB 510119, Dresden, 01314 (Germany)

    2012-07-01

    This paper presents a preliminary design of a high conversion Th-U{sup 233} fuel assembly applicable for current generation of Pressurized Water Reactor (PWRs). The considered fuel assembly has a typical 17 x 17 PWR lattice. However in order to increase the conversion of Th{sup 232} to U{sup 233}, the assembly was subdivided into the two regions called seed and blanket. The central seed region has a higher than blanket U{sup 233} content and acts as a neutron source for the peripheral blanket region. The latest acts as a U{sup 233} breeder. While the seed fuel pins have a standard dimensions the blanket fuel radius was increased in order to reduce the moderation and to facilitate the resonance neutron absorption in blanket Th{sup 232}. The U{sup 233} content in the seed and blanket regions was optimized to achieve maximal initial to discharged fissile inventory ratio (FIR) taking into account the target fuel cycle length of 12 months with 3-batch reloading scheme. In this study the neutronic calculations were performed on the fuel assembly level using Helios deterministic lattice transport code. The fuel cycle length and the core k{sub eff} were estimated by applying the Non Linear Reactivity Model. The applicability of the HELIOS code for the analysis of the Th-based high conversion designs was confirmed with the help of continuous-energy Monte-Carlo code SERPENT. The results of optimization studies show that for the heterogeneous seed and blanket (SB) fuel assembly the FIR of about 0.95 can be achieved. (authors)

  20. Engineering evaluation/cost analysis for the 233-S Plutonium Concentration Facility

    International Nuclear Information System (INIS)

    1997-01-01

    The deactivated 233-S Plutonium Concentration Facility (233-S Facility) is located in the 200 Area. The facility has undergone severe degradation due to exposure to extreme weather conditions. A rapid freeze and thaw cycle occurred at the Hanford Site during February 1996, which caused cracking to occur on portions of the building's roof. This has resulted in significantly infiltration of water into the facility, which provides a pathway for potential release of radioactive material into the environment (air and/or ground). The weather caused several existing cracks in the concrete portions of the structure to lengthen, increasing the potential for failed confinement of the radioactive material in the building. Differential settlement has also occurred, causing portions of the facility to separate from the main building structure thus creating a potential for release of radioactive material t the environment. An expedited removal action is proposed to ensure that a release from the 233-S Facility does not occur. The US Department of Energy (DOE), Richland Operations Office (RL), in cooperation with the EPA, has prepared this Engineering Evaluation/Cost Analysis (EE/CA) pursuant to CERCLA. Based on the evaluation, RL has determined that hazardous substances in the 233-S Facility may present a potential threat to human health and/or the environment, and that an expedited removal action is warranted. The purpose of the EE/CA is to provide the framework for the evaluation and selection of a technology from a viable set of alternatives for a removal action

  1. Uranium-233 analysis of biological samples

    International Nuclear Information System (INIS)

    Gies, R.A.; Ballou, J.E.; Case, A.C.

    1979-01-01

    Two liquid scintillation techniques were compared for 233 U analysis: a two-phase extraction system (D2EHPA) developed by Keough and Powers, 1970, for Pu analysis; and a single-phase emulsion system (TT21) that holds the total sample in suspension with the scintillator. The first system (D2EHPA) was superior in reducing background (two- to threefold) and in accommodating a larger sample volume (fivefold). Samples containing > 50 mg/ml of slats were not extracted quantitatively by D2EHPA

  2. 48 CFR 352.233-71 - Litigation and claims.

    Science.gov (United States)

    2010-10-01

    ... the action in good faith. The Government shall not be liable for the expense of defending any action... compensated by insurance which was required by law or regulation or by written direction of the Contracting... FORMS SOLICITATION PROVISIONS AND CONTRACT CLAUSES Texts of Provisions and Clauses 352.233-71 Litigation...

  3. Nuclear reactor for breeding 233U

    International Nuclear Information System (INIS)

    Bohanan, C.S.; Jones, D.H.; Raab, H.F. Jr.; Radkowsky, A.

    1976-01-01

    A light-water-cooled nuclear reactor capable of breeding 233 U for use in a light-water breeder reactor includes physically separated regions containing 235 U fissile material and 238 U fertile material and 232 Th fertile material and 239 Pu fissile material, if available. Preferably the 235 U fissile material and 238 U fertile material are contained in longitudinally movable seed regions and the 239 Pu fissile material and 232 Th fertile material are contained in blanket regions surrounding the seed regions. 1 claim, 5 figures

  4. Recovery of thorium along with uranium 233 from Thorex waste solution employing Chitosan

    International Nuclear Information System (INIS)

    Priya, S.; Reghuram, D.; Kumaraguru, K.; Vijayan, K.; Jambunathan, U.

    2003-01-01

    The low level waste solution, generated from Thorex process during the processing of U 233 , contains thorium along with traces of Th 228 and U 233 . Chitosan, a natural bio-polymer derived from Chitin, was earlier used to recover the uranium and americium. The studies were extended to find out its thorium sorption characteristics. Chitosan exhibited very good absorption of thorium (350 mg/g). Chitosan was equilibrated directly with the low level waste solution at different pH after adjusting its pH, for 60 minutes with a Chitosan to aqueous ratio of 1:100 and the raffinates were filtered and analysed. The results showed more than 99% of thorium and U 233 could be recovered by Chitosan between pH 4 and 5. Loaded thorium and uranium could be eluted from the Chitosan by 1M HNO 3 quantitatively. (author)

  5. Criticality safety validation: Simple geometry, single unit {sup 233}U systems

    Energy Technology Data Exchange (ETDEWEB)

    Putman, V.L.

    1997-06-01

    Typically used LMITCO criticality safety computational methods are evaluated for suitability when applied to INEEL {sup 233}U systems which reasonably can be modeled as simple-geometry, single-unit systems. Sixty-seven critical experiments of uranium highly enriched in {sup 233}U, including 57 aqueous solution, thermal-energy systems and 10 metal, fast-energy systems, were modeled. These experiments include 41 cylindrical and 26 spherical cores, and 41 reflected and 26 unreflected systems. No experiments were found for intermediate-neutron-energy ranges, or with interstitial non-hydrogenous materials typical of waste systems, mixed {sup 233}U and plutonium, or reflectors such as steel, lead, or concrete. No simple geometry experiments were found with cubic or annular cores, or approximating infinite sea systems. Calculations were performed with various tools and methodologies. Nine cross-section libraries, based on ENDF/B-IV, -V, or -VI.2, or on Hansen-Roach source data, were used with cross-section processing methods of MCNP or SCALE. The k{sub eff} calculations were performed with neutral-particle transport and Monte Carlo methods of criticality codes DANT, MCNP 4A, and KENO Va.

  6. Site specific health and safety plan, 233-S decontamination and decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    J. E. Fasso

    1997-12-31

    The deactivated 233-S Plutonium Concentration Facility, located in the 200 Area at the Hanford Site, is the subject of this Health and Safety Plan.The 233-S Facility operated from January 1952 until July 1967 at which time the building entered the U.S. Department of Energy`s Surplus Facility Management Program as a retired facility. The facility has since undergone severe degradation due to exposure to extreme weather conditions. Additionally, the weather caused existing cracks in concrete structures of the building to lengthen, thereby increasing the potential for failed confinement of the radioactive material in the building. Differential settlement has also occurred causing portions of the facility to separate from the main building structure, increasing the potential for release of radioactive material to the environment. An expedited response is proposed to remove this threat and ensure protection of human health and the environment. On this premise it is intended that the 233-S Facility removal action be performed as a Comprehensive Environmental Response, Compensation, and Liability Act of 1980 Time-Critical Project being conducted under the Pilot Hanford Environmental Restoration (ER) Initiative

  7. An extraction method of uranium 233 from the thorium irradiates in a reactor core

    International Nuclear Information System (INIS)

    Chesne, A.; Regnaut, P.

    1955-01-01

    Description of the conditions of separation of the thorium, of the uranium 233 and of the protactinium 233 in hydrochloric solution by absorption then selective elution on anion exchange resin. A precipitation of the thorium by the oxalic acid permits the recuperation of the hydrochloric acid which is recycled, the main, raw material consumed being the oxalic acid. (authors) [fr

  8. 33 CFR 110.233 - Prince William Sound, Alaska.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Prince William Sound, Alaska. 110... ANCHORAGES ANCHORAGE REGULATIONS Anchorage Grounds § 110.233 Prince William Sound, Alaska. (a) The anchorage grounds. In Prince William Sound, Alaska, beginning at a point at latitude 60°40′00″ N., longitude 146°40...

  9. A preliminary simulation of an ADS using MCNPX for U233 production

    International Nuclear Information System (INIS)

    Barros, Graiciany P.; Pereira, Claubia; Veloso, Maria A.F.; Costa, Antonella L.

    2009-01-01

    The code MCNPX is used to evaluate a simplified model of a hybrid system regenerator accelerator (ADS - Accelerator Driven Subcritical), for energy and 233 U production. The concept consists of coupling a high-energy particle accelerator with a sub-critical reactor core, using 232 ThO 2 + 233 UO 2 as initial composition. In this work, the spallation source definition used in MCNPX is a point source of 1000 MeV. The system consists in three coaxial cylinders. The internal cylinder is the spallation target that is a thick natural Pb. The intermediate cylinder is the core, composed by the mixture of fuel ( 232 ThO 2 + 9.5% 233 UO 2 ) and Pb coolant; and lead, as reflector, composes the external cylinder. The goal is to begin studies to evaluate the regenerator blanket composition when submitted to a neutron flux during a time step. The effective multiplication coefficient of the system and the variation of the composition of the regenerative layer are analyzed. The preliminary results show the possibility of utilization of this system. (author)

  10. Fabrication routes for Thorium and Uranium233 based AHWR fuel

    International Nuclear Information System (INIS)

    Danny, K.M.; Saraswat, Anupam; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun

    2011-01-01

    India's economic growth is on a fast growth track. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear Energy is best suited to meet this demand without causing undue environmental impact. Considering the large thorium reserves in India, the future nuclear power program will be based on Thorium- Uranium 233 fuel cycle. The major characteristic of thorium as the fuel of future comes from its superior fuel utilization. 233 U produced in a reactor is always contaminated with 232 U. This 232 U undergoes a decay to produce 228 Th and it is followed by decay chain including 212 Bi and 208 Tl. Both 212 Bi and 208 Tl are hard gamma emitters ranging from 0.6 MeV-1.6 MeV and 2.6 MeV respectively, which necessitates its handling in hot cell. The average concentration of 232 U is expected to exceed 1000 ppm after a burn-up of 24,000 MWD/t. Work related to developing the fuel fabrication technology including automation and remotization needed for 233 U based fuels is in progress. Various process for fuel fabrication have been developed i.e. Coated Agglomerate Pelletisation (CAP), impregnation technique (Pellet/Gel), Sol Gel Micro-sphere Pelletisation (SGMP) apart from Powder to Pellet (POP) route. This paper describes each process with respect to its advantages, disadvantages and its amenability to automation and remotisation. (author)

  11. Numerical simulations of groundwater flow and solute transport in the Lake 233 aquifer

    Energy Technology Data Exchange (ETDEWEB)

    Klukas, M H; Moltyaner, G L

    1995-05-01

    A three-dimensional numerical flow model of the Lake 233 aquifer underlying the site of the proposed Intrusion Resistant Underground Structure (IRUS) for low level waste disposal is developed. A reference hydraulic conductivity distribution incorporating the key stratigraphic units and field estimates of recharge from Lake 233 are used as model input. The model was calibrated against the measured hydraulic head distribution, the flowpath of a historic {sup 90}Sr plume in the aquifer and measured groundwater velocities. (author). 23 refs., 4 tabs., 31 figs.

  12. Numerical simulations of groundwater flow and solute transport in the Lake 233 aquifer

    International Nuclear Information System (INIS)

    Klukas, M.H.; Moltyaner, G.L.

    1995-05-01

    A three-dimensional numerical flow model of the Lake 233 aquifer underlying the site of the proposed Intrusion Resistant Underground Structure (IRUS) for low level waste disposal is developed. A reference hydraulic conductivity distribution incorporating the key stratigraphic units and field estimates of recharge from Lake 233 are used as model input. The model was calibrated against the measured hydraulic head distribution, the flowpath of a historic 90 Sr plume in the aquifer and measured groundwater velocities. (author). 23 refs., 4 tabs., 31 figs

  13. 45 CFR 233.101 - Dependent children of unemployed parents.

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Dependent children of unemployed parents. 233.101... unemployed parents. (a) Requirements for State Plans. Effective October 1, 1990 (for Puerto Rico, American... children of unemployed parents. A State plan under title IV-A for payment of such aid must: (1) Include a...

  14. A compact multi-plate fission chamber for the simultaneous measurement of 233U capture and fission cross-sections

    Directory of Open Access Journals (Sweden)

    Bacak M.

    2017-01-01

    Full Text Available 233U plays the essential role of fissile nucleus in the Th-U fuel cycle. A particularity of 233U is its small neutron capture cross-section which is about one order of magnitude lower than the fission cross-section on average. Therefore, the accuracy in the measurement of the 233U capture cross-section essentially relies on efficient capture-fission discrimination thus a combined setup of fission and γ-detectors is needed. At CERN n_TOF the Total Absorption Calorimeter (TAC coupled with compact fission detectors is used. Previously used MicroMegas (MGAS detectors showed significant γ-background issues above 100 eV coming from the copper mesh. A new measurement campaign of the 233U capture cross-section and alpha ratio is planned at the CERN n_TOF facility. For this measurement, a novel cylindrical multi ionization cell chamber was developed in order to provide a compact solution for 14 active targets read out by 8 anodes. Due to the high specific activity of 233U fast timing properties are required and achieved with the use of customized electronics and the very fast ionizing gas CF4 together with a high electric field strength. This paper describes the new fission chamber and the results of the first tests with neutrons at GELINA proving that it is suitable for the 233U measurement.

  15. Reaction rate constants of HO2 + O3 in the temperature range 233-400 K

    Science.gov (United States)

    Wang, Xiuyan; Suto, Masako; Lee, L. C.

    1988-01-01

    The reaction rate constants of HO2 + O3 were measured in the temperature range 233-400 K using a discharge flow system with photofragment emission detection. In the range 233-253 K, the constants are approximately a constant value, and then increase with increasing temperature. This result suggests that the reaction may have two different channels. An expression representing the reaction rate constants is presented.

  16. 45 CFR 233.100 - Dependent children of unemployed parents.

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Dependent children of unemployed parents. 233.100... unemployed parents. (a) Requirements for State Plans. If a State wishes to provide AFDC for children of unemployed parents, the State plan under title IV-A of the Social Security Act must: (1) Include a definition...

  17. Assessment of the available {sup 233}U cross-section evaluations in the calculation of critical benchmark experiments

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.; Wright, R.Q.

    1996-10-01

    In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U.S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the S{sub n} transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

  18. Assessment of the Available (Sup 233)U Cross Sections Evaluations in the Calculation of Critical Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.

    1993-01-01

    In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U. S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the Sn transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

  19. Process technology for the molten-salt reactor 233U--Th cycle

    International Nuclear Information System (INIS)

    Hightower, J.R. Jr.

    1975-01-01

    After a brief description of the design features of the molten-salt breeder reactor, fuel processing for removal of 233 Pa and fission products is examined. Some recent developments in processing technology are discussed

  20. ALARA review for the decontamination and decommissioning of the 233-S pipe trench

    International Nuclear Information System (INIS)

    Kornish, M.J.

    1998-01-01

    The 233-S Facility was completed in 1955 to expand plutonium production by further concentrating the plutonium nitrate product solution from the Reduction Oxidation (REDOX) Plant. The facility is radiologically contaminated because of operations and accidents. Isolation from REDOX and removal of the product transfer lines from the pipe trench is the second step in the decontamination and decommissioning of the entire 233-S Facility. The work scope is to isolate all piping from REDOX and then to remove all the piping/equipment from the pipe trench. The building is presently a Hazard Category 2 Nuclear Facility. A formal as low as reasonably achievable (ALARA) review is required by BHI-SH-02, Vol. 1, Procedure No. 1.22, Planning Radiological Work, when radiological conditions exceed trigger levels. The level of contamination inside the pipe trench and the process fluid piping is unknown. The potential exists to exceed the level of loose surface contamination, which requires a formal ALARA review when opening the pipe trench and cutting of piping commences. This ALARA review is for task instruction 1997-03-18-009 Revision 1, 233-S Pipe Trench Decon and Pipe Removal

  1. The Arg233Lys AQP0 mutation disturbs aquaporin0-calmodulin interaction causing polymorphic congenital cataract.

    Directory of Open Access Journals (Sweden)

    Shanshan Hu

    Full Text Available Calmodulin (CaM directly interacts with the aquaporin 0 (AQP0 C-terminus in a calcium dependent manner to regulate the water permeability of AQP0. We previously identified a missense mutation (p.R233K in the putative CaM binding domain of AQP0 C-terminus in a congenital cataract family. This study was aimed at exploring the potential pathogenesis of this mutation causative of cataract and mainly identifying how it influenced the binding of AQP0 to CaM. Wild type and R233K mutant AQP0 with EGFP-tag were transfected separately into Hela cells to determine the expression and subcellular localizations. The co-immunoprecipitation (CoIP assay was used to detect the interaction between AQP0 and CaM. AQP0 C-terminus peptides were synthesized with and without R233K, and the binding abilities of these peptides to CaM were assessed using a fluorescence binding assay. Localizations of wild type and R233K mutant AQP0 were determined from EGFP fluorescence, and the chimeric proteins were both localized abundantly in the plasma membrane. Protein expression levels of the culture cells showed no significant difference between them. The results from CoIP assay implied that R233K mutant presented more weakly in association with CaM than wild type AQP0. The AQP0 C-terminal mutant peptide was found to have 2.5-fold lower binding affinity to CaM than wild type peptide. These results suggested that R233K mutation did not affect the expression, location and trafficking of the protein but did influence the interaction between AQP0 and CaM. The binding affinity of AQP0 C-terminus to CaM was significantly reduced. Due to lack of the modulation of the Ca2+-calmodulin complex, the water permeability of AQP0 was subsequently augmented, which might lead to the development of this cataract.

  2. A method for the quantitative determination of uranium-233 in an irradiated thorium rod; Une methode de dosage de l'uranium 233 contenu dans un barreau de thorium irradie

    Energy Technology Data Exchange (ETDEWEB)

    Bathellier, A; Sontag, R; Chesne, A [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1961-07-01

    A rapid method for the quantitative determination of uranium-233 in irradiated thorium is described. A 30 per cent solution of trilaurylamine in xylene is used to extract the uranium from an aqueous hydrochloric acid solution and separate it from the thorium. This may be followed by {alpha} counting or fluorimetry. The practical operating conditions of the separation are discussed in detail. (author) [French] Une methode rapide de dosage de l'uranium-233 contenu dans le thorium irradie est decrite. Elle utilise la trilauryfamine a 30 pour cent dans le xylene pour extraire l'uranium d'une dissolution aqueuse chlorhydrique et le separer du thorium. Le comptage {alpha} ou la fluorimetrie sont alors possibles. Les conditions operatoires de la separation sont discutees et precisees. (auteur)

  3. Subcritical multiplication measurements with a BeO reflected 233U uranyl nitrate solution system

    International Nuclear Information System (INIS)

    Job, P.K.; Srinivasan, M.; Nargundkar, V.R.; Chandramoleshwar, K.; Pasupathy, C.S.; Das, S.; Mayankutty, P.C.

    1978-01-01

    A series of subcritical multiplication measurements were carried out in PURNIMA with 233 U uranyl nitrate solution contained in all 11 x 11 cm 2 square sectional tank and reflected by 30 cm thickness of BeO on all sides. The objective of these experiments was to determine the 'Minimum critical mass' of the system in rectangular parellelopiped geometry. The rectangular aluminium core tank was attached to the bottom of an alpha tight glove box. BeO reflector was arranged below the glove box outside the core tank. The system multiplication was measured as a function of solution concentration and core volume by means of neutron detectors placed outside the assembly. The extrapolated critical mass was obtained through conventional inverse counts plot. The maximum amount of 233 U used was 120 gms. The rectangular geometry was estimated to be 235 +- 10 gms, in the concentration range of 80 to 120 gms/litre of 233 U. The experimental set up, procedure adopted, method of analysis and the details of the results are described. (author)

  4. Los Alamos National Laboratory Site Integrated Management plan, uranium 233 storage and disposition. Volume 1: Project scope and description

    International Nuclear Information System (INIS)

    Nielsen, J.B.; Erickson, R.

    1997-01-01

    This Site Integration Management plan provides the Los Alamos Response to the Defense Nuclear Facility Safety Board (DNFSB) Recommendation 97-1. This recommendation addresses the safe storage and management of the Departments uranium 233 ( 233 U) inventory. In the past, Los Alamos has used 233 U for a variety of different weapons related projects. The material was used at a variety of sites in varying quantities. Now, there is a limited need for this material and the emphasis has shifted from use to storage and disposition of the material. The Los Alamos program to address the DNFSB Recommendation 97-1 has two emphases. First, take corrective action to address near term deficiencies required to provide safe interim storage of 233 U. Second, provide a plan to address long term storage and disposition of excess inventory at Los Alamos

  5. Engineering evaluation/cost analysis for the 233-S Plutonium Concentration Facility

    International Nuclear Information System (INIS)

    Rugg, J.E.

    1996-08-01

    The 100, 200, 300 and 1100 Areas of the Hanford Site were placed on the U. S. Environmental Protection Agency's National Priorities List in November 1989 under the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). Located in the 200 Area is the deactivated 233-S Plutonium Concentration Facility (used in the REDOX process). The facility has undergone severe degradation due to exposure to extreme weather conditions. An expedited response is proposed to ensure protection of human health and the environment. The Department of Energy, Richland Operations Office (RL) in cooperation with the Washington State Department of Ecology, has prepared this Engineering Evaluation/Cost Analysis pursuant to CERCLA. Based on the evaluation, RL has determined that hazardous substances in the 233-S Facility may present a potential threat to human health or the environment, and that an expedited removal action is warranted for decommissioning of the facility

  6. Contribution to the study of {sup 233}U production with MOX-ThPu fuel in PWR reactor. Transition scenarios towards Th/{sup 233}U iso-generating concepts in thermal spectrum. Development of the MURE fuel evolution code; Contribution a l'etude de la production d'{sup 233}U en combustible MOX-ThPu en reacteur a eau sous pression. Scenarios de transition vers des concepts isogenerateurs Th/{sup 233}U en spectre thermique. Developpement du code MURE d'evolution du combustible

    Energy Technology Data Exchange (ETDEWEB)

    Michel-Sendis, F

    2006-12-15

    If nuclear power is to provide a significant fraction of the growing world energy demand, only through the breeding concept can the development of sustainable nuclear energy become a reality. The study of such a transition, from present-day nuclear technologies to future breeding concepts is therefore pertinent. Among these future concepts, those using the thorium cycle Th/U-233 in a thermal neutron spectrum are of particular interest; molten-salt type thermal reactors would allow for breeding while requiring comparatively low initial inventories of U-233. The upstream production of U-233 can be obtained through the use of thorium-plutonium mixed oxide fuel in present-day light water reactors. This work presents, firstly, the development of the MURE evolution code system, a C++ object-oriented code that allows the study, through Monte Carlo (M.C.) simulation, of nuclear reactors and the evolution of their fuel under neutron irradiation. The M.C. methods are well-suited for the study of any reactor, whether it'd be an existing reactor using a new kind of fuel or a future concept altogether, the simulation is only dependent on nuclear data. Exact and complex geometries can be simulated and continuous energy particle transport is performed. MURE is an interface with MCNP, the well-known and validated transport code, that allows, among other functionalities, to simulate constant power and constant reactivity evolutions. Secondly, the study of MOX ThPu fuel in a conventional light water reactor (REP) is presented; it explores different plutonium concentrations and isotopic qualities in order to evaluate their safety characteristics. Simulation of their evolution allows us to quantify the production of U-233 at the end of burnup. Last, different french scenarios validating a possible transition towards a park of thermal Th/U-233 breeders, are presented. In these scenarios, U-233 is produced in ThPu moxed light water reactors. (author)

  7. Evaluation of cross sections of Th-232 and U-233

    International Nuclear Information System (INIS)

    Dias, A.M.

    1978-01-01

    The cross sections in multigroups of Th-232 and U-233 are evaluated by comparison of theoretical results and experimental data obtained through experiments on the fast reactors IBR-I, EBR-II, BR-I and AETR. The deviation between calculated values and experimental results is about 10%. They are therefore satisfatory for neutronic calculations [pt

  8. A novel hash based least significant bit (2-3-3) image steganography in spatial domain

    OpenAIRE

    Manjula, G. R.; Danti, Ajit

    2015-01-01

    This paper presents a novel 2-3-3 LSB insertion method. The image steganography takes the advantage of human eye limitation. It uses color image as cover media for embedding secret message.The important quality of a steganographic system is to be less distortive while increasing the size of the secret message. In this paper a method is proposed to embed a color secret image into a color cover image. A 2-3-3 LSB insertion method has been used for image steganography. Experimental results show ...

  9. ALARA review for the decontamination and decommissioning of the 233-S P.R. can loadout and decontamination

    International Nuclear Information System (INIS)

    Kornish, J.M.

    1998-01-01

    The 233-S Facility was completed in 1955 to expand plutonium production by further concentrating the plutonium nitrate product solution from the Reduction Oxidation (REDOX) Plant. The facility is radiologically contaminated because of operations and accidents. The building is presently a Hazard Category 2 Nuclear Facility. Disassembly of the loadout hood and its associated equipment may be done in parallel with the isolation of 233-S from REDOX via the pipe trench equipment removal. The work scope is to remove the entire loadout hood from the Product Receiver (P.R.) Can Loadout and Decon Room inside the 233-S Facility. A formal as low as reasonably achievable (ALARA) review is required by BHI-SH-02, Vol. 1, Procedure 1.22, Planning Radiological Work, when radiological conditions exceed trigger levels. The level of contamination inside the loadout hood and its associated equipment is unknown. The potential exists to exceed the level of loose surface contamination, which requires a formal ALARA review when opening the loadout hood and disassembly commences. This ALARA review is for the task instruction 1997-03-18-010 Revision 0, 233-S Loadout Hood Decon and Dismantlement

  10. 39 CFR 233.5 - Requesting financial records from a financial institution.

    Science.gov (United States)

    2010-07-01

    ... INSPECTION SERVICE AUTHORITY § 233.5 Requesting financial records from a financial institution. (a... Department of the U.S. Postal Service to request financial records from a financial institution pursuant to... authorized to request financial records of any customer from a financial institution pursuant to a formal...

  11. Influence of Contact Time on the Extraction of 233Uranyl Spike and Contaminant Uranium From Hanford Sediment

    International Nuclear Information System (INIS)

    Smith, Steven C.; Szecsody, James E.

    2011-01-01

    In this study 233Uranyl nitrate was added to uranium (U) contaminated Hanford 300 Area sediment and incubated under moist conditions for 1 year. It hypothesized that geochemical transformations and/or physical processes will result in decreased extractability of 233U as the incubation period increases, and eventually the extraction behavior of the 233U spike will be congruent to contaminant U that has been associated with sediment for decades. Following 1 week, 1 month, and 1 year incubation periods, sediment extractions were performed using either batch or dynamic (sediment column flow) chemical extraction techniques. Overall, extraction of U from sediment using batch extraction was less complicated to conduct compared to dynamic extraction, but dynamic extraction could distinguish the range of U forms associated with sediment which are eluted at different times.

  12. Alecto - results obtained with homogeneous critical experiments on plutonium 239, uranium 235 and uranium 233; Alecto - resultats des experiences critiques homogenes realisees sur le plutonium 239, l'uranium 235 et l'uranium 233

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, J G; Brunet, J P; Caizegues, R; Clouet d' Orval, Ch; Kremser, J; Tellier, H; Verriere, Ph [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    In this report are given the results of the homogeneous critical experiments ALECTO, made on plutonium 239, uranium 235 and uranium 233. After a brief description of the equipment, the critical masses for cylinders of diameters varying from 25 to 42 cm, are given and compared with other values (foreign results, criticality guide). With respect to the specific conditions of neutron reflection in the ALECTO experiments the minimal values of critical masses are: Pu239 M{sub c} = 910 {+-} 10 g, U235 M{sub c} = 1180 {+-} 12 g and U233 M{sub c} = 960 {+-} 10 g. Experiments relating to cross sections and constants to be used on these materials are presented. Lastly, kinetic experiments allow to compare pulsed neutron methods to fluctuation methods. [French] On presente dans ce rapport les resultats des experiences critiques homogenes ALECTO, effectuees sur le plutonium 239, l'uranium 235 et l'uranium 233. Apres avoir rappele la description des installations, on donne les masses critiques pour des cylindres de diametres variant entre 25 et 42 cm, qui sont comparees avec d'autres chiffres (resultats etrangers, guide de criticite). Dans les gammes des diametres etudies pour des cuves a fond plat reflechies lateralement, la valeur minimale des masses critiques est la suivante: Pu239 M{sub c} = 910 {+-} 10 g, U235 M{sub c} = 1180 {+-} 12 g et U233 M{sub c} 960 {+-} 10 g. Des experiences portant sur les sections efficaces et les constantes a utiliser sur ces milieux sont ensuite presentees. Enfin des experiences de cinetique permettent une comparaison entre la methode des neutrons pulses et la methode des fluctuations. (auteur)

  13. Review of thorium-U233 cycle thermal reactor benchmark studies (AWBA Development Program)

    International Nuclear Information System (INIS)

    Ullo, J.J.; Hardy, J. Jr.; Steen, N.M.

    1980-03-01

    A survey is made of existing integral experiments for U233 systems and thorium-uranium based fuel systems. The aim is to understand to what extent they give a consistent test of ENDF/B-IV nuclear data. A principal result is that ENDF/B-IV leads to an underprediction of neutron leakage. Results from testing alternate thorium data sets are presented. For one evaluation due to Leonard, the results depict a possible growing discrepancy between measured integral parameters such as rho 02 and I 232 and the differential data, which underpredicts these parameters. Sensitivities to other nuclear data components, notably the fission neutron spectrum, were determined. A new harder U233 spectrum significantly reduces a bias trend in K/sub eff/ vs leakage

  14. Contribution to the study of {sup 233}U production with MOX-ThPu fuel in PWR reactor. Transition scenarios towards Th/{sup 233}U iso-generating concepts in thermal spectrum. Development of the MURE fuel evolution code; Contribution a l'etude de la production d'{sup 233}U en combustible MOX-ThPu en reacteur a eau sous pression. Scenarios de transition vers des concepts isogenerateurs Th/{sup 233}U en spectre thermique. Developpement du code MURE d'evolution du combustible

    Energy Technology Data Exchange (ETDEWEB)

    Michel-Sendis, F

    2006-12-15

    If nuclear power is to provide a significant fraction of the growing world energy demand, only through the breeding concept can the development of sustainable nuclear energy become a reality. The study of such a transition, from present-day nuclear technologies to future breeding concepts is therefore pertinent. Among these future concepts, those using the thorium cycle Th/U-233 in a thermal neutron spectrum are of particular interest; molten-salt type thermal reactors would allow for breeding while requiring comparatively low initial inventories of U-233. The upstream production of U-233 can be obtained through the use of thorium-plutonium mixed oxide fuel in present-day light water reactors. This work presents, firstly, the development of the MURE evolution code system, a C++ object-oriented code that allows the study, through Monte Carlo (M.C.) simulation, of nuclear reactors and the evolution of their fuel under neutron irradiation. The M.C. methods are well-suited for the study of any reactor, whether it'd be an existing reactor using a new kind of fuel or a future concept altogether, the simulation is only dependent on nuclear data. Exact and complex geometries can be simulated and continuous energy particle transport is performed. MURE is an interface with MCNP, the well-known and validated transport code, that allows, among other functionalities, to simulate constant power and constant reactivity evolutions. Secondly, the study of MOX ThPu fuel in a conventional light water reactor (REP) is presented; it explores different plutonium concentrations and isotopic qualities in order to evaluate their safety characteristics. Simulation of their evolution allows us to quantify the production of U-233 at the end of burnup. Last, different french scenarios validating a possible transition towards a park of thermal Th/U-233 breeders, are presented. In these scenarios, U-233 is produced in ThPu moxed light water reactors. (author)

  15. Sampling and Analysis Plan for the 233-S Plutonium Concentration Facility

    International Nuclear Information System (INIS)

    Mihalic, M.A.

    1998-02-01

    This Sampling and Analysis Plan (SAP) provides the information and instructions to be used for sampling and analysis activities in the 233-S Plutonium Concentration Facility. The information and instructions herein are separated into three parts and address the Data Quality Objective (DQO) Summary Report, Quality Assurance Project Plan (QAP), and SAP

  16. Comparison of potential radiological impacts of 233U and 239Pu fuel cycles

    International Nuclear Information System (INIS)

    Meyer, H.R.; Little, C.A.; Witherspoon, J.P.; Till, J.E.

    1979-01-01

    Nuclear fuel cycles utilizing 233 U are currently the subject of considerable interest in the United States. This paper focuses on the identification of significant differences between the off-site radiological hazards posed by 232 Th/ 233 U (Th/U) and 238 U/ 239 Pu (U/Pu) fuel cycles, and represents a portion of our involvement in the Nonproliferation Alternative Systems Assessment Program (NASAP), to be used in support of the International Fuel Cycle Evaluation (INFCE). The major contributors to radiological dose are likely to be uranium mining and milling (58.5% of total fuel cycle dose), reprocessing (33.9%), and light-water reactor power generation (7.3%). The remainder of the cycle, including enrichment processes, fuel fabrication, transportation, and waste management, contributes only 0.3% to total estimated fuel cycle dose

  17. Bi-stability in accelerator driven 233U breeders

    International Nuclear Information System (INIS)

    Ghosh, Biplab; Degweker, S.B.

    2011-01-01

    Research on Accelerator Driven Systems (ADSs) is being carried out around the world primarily with the objective of waste transmutation. Presently, the volume of waste in India is small and therefore there is little incentive to develop ADS based waste transmutation technology immediately. On the other hand, the indigenous U availability is limited and hence there is a strong incentive for breeding. Moreover the large Th deposits in the country provide a clear incentive to develop Th related technologies. Th has the additional advantage that it produces very little trans-uranic waste. While Pu fuelled fast reactors using advanced metallic fuel can have high breeding ratios due to the hard spectrum in such reactors, Th fuelled critical reactors can at best be self sustaining or marginal breeders. A possible way to improve the breeding of Th fueled reactors is to use an external neutron source as is done in ADSs. ADSs can not only give improved breeding but also permit greater flexibility in type of fuel that may be used and have the potential to considerably simplify the Th fuel cycle as in the case of the Th burner. In this paper we study various issues associated with breeding in ADSs such as the energy economics of breeding in ADSs using various types of neutron sources and the effect of the reactor spectrum and the discharge fluence (or irradiation time) of the fuel on the breeding performance. We show that even with non-fissioning, non-power- producing targets such as Pb or LBE it is possible to choose the fuel irradiation time so that the breeder produces sufficient power to drive the accelerator and export the balance to the grid, without significantly diminishing the 233 U breeding rate. By increasing the discharge fluence (irradiation time) it is possible to increase the power. However, the 233 U production rate falls off rapidly to about half its maximum value. This is the Th burner region. As the equations governing the breeding process are non

  18. System Requirements Document for the Molten Salt Reactor Experiment 233U conversion system

    International Nuclear Information System (INIS)

    Aigner, R.D.

    2000-01-01

    The purpose of the conversion process is to convert the 233 U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019

  19. Comparison of Hansen--Roach and ENDF/B-IV cross sections for 233U criticality calculations

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1976-01-01

    A comparison is made between criticality calculations performed using ENDF/B-IV cross sections and the 16-group Hansen-- Roach library at ORNL. The area investigated is homogeneous systems of highly enriched 233 U in simple geometries. Calculations are compared with experimental data for a wide range of H/ 233 U ratios. Results show that calculations of k/sub eff/ made with the Hansen--Roach cross sections agree within 1.5 percent for the experiments considered. Results using ENDF/B-IV cross sections were in good agreement for well-thermalized systems, but discrepancies up to 7 percent in k/sub eff/ were observed in fast and epithermal systems

  20. Measurement of the fission cross section induced by fast neutrons of the {sup 232}Th/{sup 233}U nuclei within the innovating fuel cycles framework; Mesure de la section efficace de fission induite par neutrons rapides des noyaux {sup 232}Th/{sup 233}U dans le cadre des cycles de combustible innovants

    Energy Technology Data Exchange (ETDEWEB)

    Grosjean, C

    2005-03-15

    The thorium-U{sup 233} fuel cycle might provided safer and cleaner nuclear energy than the present Uranium/Pu fuelled reactors. Over the last 10 years, a vast campaign of measurements has been initiated to bring the precision of neutron data for the key nuclei (Th{sup 232}, Pa{sup 233} and U{sup 233}) at the level of those for the U-Pu cycle. This is the framework of these measurements, the energy dependent neutron induced fission cross section of Th{sup 232} and U{sup 233} has been measured from 1 to 7 MeV with a target accuracy lesser than 5 per cent. These measurements imply the accurate determination of the fission rate, the number of the target nuclei as well as the incident neutron flux impinging on the target, the latter has been obtained using the elastic scattering (n,p). The cross section of which is very well known in a large neutron energy domain ({approx} 0,5 % from 1 eV to 50 MeV) compared to the U{sup 235}(n,f) reaction. This technique has been applied for the first time to the Th{sup 232}(n,f) and U{sup 233}(n,f) cases. A Hauser-Feshbach statistical model has been developed. It consists of describing the different decay channels of the compound nucleus U{sup 234} from 0,01 to 10 MeV neutron energy. The parameters of this model were adjusted in order to reproduce the measured fission cross section of U{sup 233}. From these parameters, the cross sections from the following reactions could be extracted: inelastic scattering U{sup 233}(n,n'), radiative capture U{sup 233}(n,{gamma}) and U{sup 233}(n,2n). These cross sections are still difficult to measure by direct neutron reactions. The calculated values have allowed us to fill the lack of experimental data for the major fissile nucleus of the thorium cycle. (author)

  1. Study of the excited levels of 233Pa by the 237Np alpha decay

    International Nuclear Information System (INIS)

    Gonzalez, J.; Gaeta, R.; Vano, E.; Los Arcos, J. M.

    1978-01-01

    The excited levels in 233 P a following the 237 N p alpha decay have been studied, by performing different experiences to complete available data and supply new information. Thus, two direct alpha spectrum measurement, one alpha-gamma bidimensional coincidence experiment, three gamma-gamma and gamma-X ray coincidences and some other measurements of the gamma spectrum, direct and coincident with alpha-particles have been made. These last experiences have allowed to obviate usual radiochemical separation methods, the 233 P a radioactive descendent interferences being eliminated by means of the coincidence technic. As a result, a primary decay scheme has been elaborated, including 15 new gamma transitions and two new levels, not observed in the most recent works. (Author) 60 refs

  2. 29 CFR 779.233 - Independent contractors performing work “for” an enterprise.

    Science.gov (United States)

    2010-07-01

    ... Apply; Enterprise Coverage Leased Departments, Franchise and Other Business Arrangements § 779.233... section 3(r) has reference to an independent business which performs services for other businesses as an established part of its own business activities. The term “independent contractor” as used in 3(r) thus has...

  3. New calculation for the neutron-induced fission cross section of 233Pa between 1.0 and 3.0 MeV

    International Nuclear Information System (INIS)

    Mesa, J.; Deppman, A.; Likhachev, V.P.; Arruda-Neto, J.D.T.; Manso, M.V.; Garcia, C.E.; Rodriguez, O.; Guzman, F.; Garcia, F.

    2003-01-01

    The 233 Pa(n,f) cross section, a key ingredient for fast reactors and accelerators driven systems, was measured recently with relatively good accuracy [F. Tovesson et al., Phys. Rev. Lett. 88, 062502 (2002)]. The results are at strong variance with accepted evaluations and an existing indirect experiment. This circumstance led us to perform a quite detailed and complete evaluation of the 233 Pa(n,f) cross section between 1.0 and 3.0 MeV, where use of our newly developed routines for the parametrization of the nuclear surface and the calculation of deformation parameters and level densities (including low-energy discrete levels) were made. The results show good quantitative and excellent qualitative agreement with the experimental direct data obtained by Tovesson et al. [F. Tovesson et al., Phys. Rev. Lett. 88, 062502 (2002)]. Additionally, our methodology opens new possibilities for the analysis of subthreshold fission and above threshold second-chance fission for both 233 Pa and its decay product 233 U, as well as other strategically important fissionable nuclides

  4. Representations for the extreme zeros of orthogonal polynomials (vol 233, pg 847, 2009)

    NARCIS (Netherlands)

    van Doorn, Erik A.; van Foreest, Nicky D.; Zeifman, Alexander I.

    2013-01-01

    We correct representations for the endpoints of the true interval of orthogonality of a sequence of orthogonal polynomials that were stated by us in the Journal of Computational and Applied Mathematics 233 (2009) 847-851. (c) 2013 Elsevier B.V. All rights reserved.

  5. 48 CFR 52.233-4 - Applicable Law for Breach of Contract Claim.

    Science.gov (United States)

    2010-10-01

    ... Provisions and Clauses 52.233-4 Applicable Law for Breach of Contract Claim. As prescribed in 33.215(b), insert the following clause: Applicable Law for Breach of Contract Claim (OCT 2004) United States law... 48 Federal Acquisition Regulations System 2 2010-10-01 2010-10-01 false Applicable Law for Breach...

  6. Cost-based optimizations of power density and target-blanket modularity for 232Th/233U-based ADEP

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1995-01-01

    A cost-based parametric systems model is developed for an Accelerator-Driven Energy Production (ADEP) system based on a 232 Th/ 233 U fuel cycle and a molten-salt (LiF/BeF 2 /ThF 3 ) fluid-fuel primary system. Simplified neutron-balance, accelerator, reactor-core, chemical-processing, and balance-of-plant models are combined parametrically with a simplified costing model. The main focus of this model is to examine trade offs related to fission power density, reactor-core modularity, 233 U breeding rate, and fission product transmutation capacity

  7. ALARA Review of 233-S Process Hood D and D and Related Activities

    International Nuclear Information System (INIS)

    Landsman, S.D.

    1999-01-01

    This document is an as low as reasonably achievable (ALARA) review of design packages for planned work at the 233-S Facility. The ALARA review was initiated in accordance with the Bechtel Hanford, Inc. (BHI) Integrated Environmental, Safety, and Health Management System (ISMS) workflow process

  8. The status of 232Th and 233U for CENDL-3.0

    International Nuclear Information System (INIS)

    Liu Ping

    2003-01-01

    The new version CENDL-3.0 of China: Evaluated nuclear data library has been updated, and contains about 200 nuclides. Among them, the data of following nuclides have been newly evaluated or reevaluated: fissile nuclides 15, structure materials 18, light nuclides 3, fission products 116. The 232 Th and 233 U are newly evaluated

  9. Efficient Low-Voltage Operation of a CW Gyrotron Oscillator at 233 GHz.

    Science.gov (United States)

    Hornstein, Melissa K; Bajaj, Vikram S; Griffin, Robert G; Temkin, Richard J

    2007-02-01

    The gyrotron oscillator is a source of high average power millimeter-wave through terahertz radiation. In this paper, we report low beam power and high-efficiency operation of a tunable gyrotron oscillator at 233 GHz. The low-voltage operating mode provides a path to further miniaturization of the gyrotron through reduction in the size of the electron gun, power supply, collector, and cooling system, which will benefit industrial and scientific applications requiring portability. Detailed studies of low-voltage operation in the TE(2) (,) (3) (,) (1) mode reveal that the mode can be excited with less than 7 W of beam power at 3.5 kV. During CW operation with 3.5-kV beam voltage and 50-mA beam current, the gyrotron generates 12 W of RF power at 233.2 GHz. The EGUN electron optics code describes the low-voltage operation of the electron gun. Using gun-operating parameters derived from EGUN simulations, we show that a linear theory adequately predicts the low experimental starting currents.

  10. Three Small Planets Transiting the Bright Young Field Star K2-233

    Science.gov (United States)

    David, Trevor J.; Crossfield, Ian J. M.; Benneke, Björn; Petigura, Erik A.; Gonzales, Erica J.; Schlieder, Joshua E.; Yu, Liang; Isaacson, Howard T.; Howard, Andrew W.; Ciardi, David R.; Mamajek, Eric E.; Hillenbrand, Lynne A.; Cody, Ann Marie; Riedel, Adric; Schwengeler, Hans Martin; Tanner, Christopher; Ende, Martin

    2018-05-01

    We report the detection of three small transiting planets around the young K3 dwarf K2-233 (2MASS J15215519‑2013539) from observations during Campaign 15 of the K2 mission. The star is relatively nearby (d = 69 pc) and bright (V = 10.7 mag, K s = 8.4 mag), making the planetary system an attractive target for radial velocity follow-up and atmospheric characterization with the James Webb Space Telescope. The inner two planets are hot super-Earths (R b = 1.40 ± 0.06 {R}\\oplus , R c = 1.34 ± 0.08 {R}\\oplus ), while the outer planet is a warm sub-Neptune (R d = 2.6 ± 0.1 {R}\\oplus ). We estimate the stellar age to be {360}-140+490 Myr based on rotation, activity, and kinematic indicators. The K2-233 system is particularly interesting given recent evidence for inflated radii in planets around similarly aged stars, a trend potentially related to photo-evaporation, core cooling, or both mechanisms.

  11. Isolation and characterization of racemase from Ensifer sp. 23-3 that acts on α-aminolactams and α-amino acid amides.

    Science.gov (United States)

    Matsui, Daisuke; Fuhshuku, Ken-Ichi; Nagamori, Shingo; Takata, Momoko; Asano, Yasuhisa

    2017-11-01

    Limited information is available on α-amino-ε-caprolactam (ACL) racemase (ACLR), a pyridoxal 5'-phosphate-dependent enzyme that acts on ACL and α-amino acid amides. In the present study, eight bacterial strains with the ability to racemize α-amino-ε-caprolactam were isolated and one of them was identified as Ensifer sp. strain 23-3. The gene for ACLR from Ensifer sp. 23-3 was cloned and expressed in Escherichia coli. The recombinant ACLR was then purified to homogeneity from the E. coli transformant harboring the ACLR gene from Ensifer sp. 23-3, and its properties were characterized. This enzyme acted not only on ACL but also on α-amino-δ-valerolactam, α-amino-ω-octalactam, α-aminobutyric acid amide, and alanine amide.

  12. Core design options for high conversion BWRs operating in Th–233U fuel cycle

    International Nuclear Information System (INIS)

    Shaposhnik, Y.; Shwageraus, E.; Elias, E.

    2013-01-01

    Highlights: • BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. • Seed blanket optimization that includes assembly size array and axial dimensions. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal-hydraulic analysis includes MCPR observation. -- Abstract: Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone “sandwiched” between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233 U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design

  13. Alecto - results obtained with homogeneous critical experiments on plutonium 239, uranium 235 and uranium 233

    International Nuclear Information System (INIS)

    Bruna, J.G.; Brunet, J.P.; Caizegues, R.; Clouet d'Orval, Ch.; Kremser, J.; Tellier, H.; Verriere, Ph.

    1965-01-01

    In this report are given the results of the homogeneous critical experiments ALECTO, made on plutonium 239, uranium 235 and uranium 233. After a brief description of the equipment, the critical masses for cylinders of diameters varying from 25 to 42 cm, are given and compared with other values (foreign results, criticality guide). With respect to the specific conditions of neutron reflection in the ALECTO experiments the minimal values of critical masses are: Pu239 M c = 910 ± 10 g, U235 M c = 1180 ± 12 g and U233 M c = 960 ± 10 g. Experiments relating to cross sections and constants to be used on these materials are presented. Lastly, kinetic experiments allow to compare pulsed neutron methods to fluctuation methods [fr

  14. Criticality evaluation for the 233-S decontamination and decommissioning project

    International Nuclear Information System (INIS)

    1996-08-01

    This criticality evaluation document analyzes the potential of a criticality event as a result of decontaminating and decommissioning the 233-S Plutonium Concentration Facility. These calculations supplement the previous set of calculations performed under this same contract, which were performed on March 13, 1996. These calculations were performed using the same MCNP computer code as for the previous set; the validation calculations performed then are valid for this set as well. Hand calculations, using the method of Solid Angle, were also developed

  15. Allelic variation at the 8q23.3 colorectal cancer risk locus functions as a cis-acting regulator of EIF3H.

    Directory of Open Access Journals (Sweden)

    Alan M Pittman

    2010-09-01

    Full Text Available Common genetic variation at human 8q23.3 is significantly associated with colorectal cancer (CRC risk. To elucidate the basis of this association we compared the frequency of common variants at 8q23.3 in 1,964 CRC cases and 2,081 healthy controls. Reporter gene studies showed that the single nucleotide polymorphism rs16888589 acts as an allele-specific transcriptional repressor. Chromosome conformation capture (3C analysis demonstrated that the genomic region harboring rs16888589 interacts with the promoter of gene for eukaryotic translation initiation factor 3, subunit H (EIF3H. We show that increased expression of EIF3H gene increases CRC growth and invasiveness thereby providing a biological mechanism for the 8q23.3 association. These data provide evidence for a functional basis for the non-coding risk variant rs16888589 at 8q23.3 and provides novel insight into the etiological basis of CRC.

  16. 49 CFR 23.3 - What do the terms used in this part mean?

    Science.gov (United States)

    2010-10-01

    ... BUSINESS ENTERPRISE IN AIRPORT CONCESSIONS General § 23.3 What do the terms used in this part mean... participation of firms in the ACDBE program. Airport Concession Disadvantaged Business Enterprise (ACDBE) means... management and daily business operations are controlled by one or more of the socially and economically...

  17. Neutron inelastic-scattering cross sections of 232Th, 233U, 235U, 238U, 239Pu and 240Pu

    International Nuclear Information System (INIS)

    Smith, A.B.; Guenther, P.T.

    1982-01-01

    Differential-neutron-emission cross sections of 232 Th, 233 U, 235 U, 238 U, 239 Pu and 240 Pu are measured between approx. = 1.0 and 3.5 MeV with the angle and magnitude detail needed to provide angle-integrated emission cross sections to approx. 232 Th, 233 U, 235 U and 238 U inelastic-scattering values, poor agreement is observed for 240 Pu, and a serious discrepancy exists in the case of 239 Pu

  18. Neutronic studies of a 233U breeder

    International Nuclear Information System (INIS)

    Hansen, L.F.; Maniscalco, J.A.

    1978-09-01

    Neutronic calculations have been carried out to design a laser fusion driven hybrid blanket which maximizes 233 U production per unit of thermal energy (>1 kg/MW/sub T/-year) with acceptable fusion energy multiplication (M/sub F/ approx. 4). Two hybrid blankets, a thorium and a uranium--thorium blanket, are discussed in detail and their performance is evaluated by incorporating them into an existing hybrid design (the LLL/Bechtel design). The performance of these two blankets is discussed in terms of their energy multiplication, tritium breeding and fissile fuel production. The neutronic calculations have been done for two neutron libraries, the ENDF/B-IV and the ENDL with differences no larger than 10% in the results. An estimate is given of the number of equivalent thermal power fission reactors (LWR, HWR, SSCR, and HTGR) that these fusion breeders can fuel

  19. Passive neutron survey of the 233-S Plutonium Concentration Facility

    International Nuclear Information System (INIS)

    1996-08-01

    A passive neutron survey was performed at the 233-S Plutonium Concentration Facility (located at the Hanford Site in Richland, Washington) during late 1994 and early 1995. Four areas were surveyed: an abandoned filter box and pipe trench, column laydown trench, load-out hood, and process hood. The primary purpose of the survey was to identify locations that had plutonium to help direct decontamination and decommissioning activities. A secondary purpose of the survey was to determine the quantity of material when its presence was identified

  20. The use in nuclear reactors of plutonium and U233 produced in accelerators

    International Nuclear Information System (INIS)

    Gambier, G.

    1983-01-01

    After a review of the presently known energy production systems and the estimated world's energy cumulative consumption during the next century, the author considers the production of fertile isotopes Pu239 and U233 in proton accelerators and finally their different uses in conventional PWR or FBR and the thorium cycle. (A.F.)

  1. Development of Indian cross section data files for Th-232 and U-233 and integral validation studies

    International Nuclear Information System (INIS)

    Ganesan, S.

    1988-01-01

    This paper presents an overview of the tasks performed towards the development of Indian cross section data files for Th-232 and U-233. Discrepancies in various neutron induced reaction cross sections in various available evaluated data files have been obtained by processing the basic data into multigroup form and intercomparison of the latter. Interesting results of integral validation studies for capture, fission and (n,2n) cross sections for Th-232 by analyses of selected integral measurements are presented. In the resonance range, energy regions where significant differences in the calculated self-shielding factors for Th-232 occur have been identified by a comparison of self-shielded multigroup cross sections derived from two recent evaluated data files, viz., ENDF/B-V (Rev.2) and JENDL-2, for several dilutions and temperatures. For U-233, the three different basic data files ENDF/B-IV, JENDL-2 and ENDL-84 were intercompared. Interesting observations on the predictional capability of these files for the criticality of the spherical metal U-233 system are given. The current status of Indian data file is presented. (author) 62 ref

  2. Safety analysis for the 233-S decontamination and decommissioning project

    International Nuclear Information System (INIS)

    Thoren, S.

    1996-08-01

    Decommissioning of the 233-S Plutonium Concentration Facility (REDOX) is a proposed expedited response action that is regulated by the Comprehensive Environmental Response Compensation and Liability Act of 1980 and the Hanford Federal Facility Agreement and Consent Order. Due to progressive physical deterioration of this facility, a decontamination and decommissioning plan is being considered for the immediate future. This safety analysis describes the proposed actions involved in this D ampersand D effort; identifies the radioactive material inventories involved; reviews site specific environmental characteristics and postulates an accident scenario that is evaluated to identify resultant effects

  3. Mass determination of U-233 and Pu-239 by gamma spectrometry technique

    International Nuclear Information System (INIS)

    Moraes, M.A.P.V. de; Pugliesi, R.

    1988-09-01

    The gamma spectrometry technique has been used for masses determinations of uranium-233 and plutonium-239, granted by AERE-HARWELL. A high purity Ge semicondutor detector was used and the total efficiency curve was obtained for the counting system in the energy range 13 KeV to 135 KeV. The calculated values for the masses compared with that obtained by means of gravimetry technique. (author) [pt

  4. Metallurgical characteristics and fracture mechanical properties of unirradiated Kori-1 RPV weld: Linde 80, WF-233

    International Nuclear Information System (INIS)

    Hong, Jun Hwa; Lee, B. S.; Oh, Y. J.; Chi, S. H.; Kim, J. H.; Park, D. G.; Yoon, J. H.; Oh, J. M.

    2000-07-01

    The fracture toughness transition properties of the low upper shelf weld, Linde 80 WF-233, of Kori-1 RPV were evaluated by the master curve method, which is designated by ASTM E 1921, 'Standard test method for determination of reference temperature, T o , for ferritic steels in the transition range'. The reference temperature, T o =-83 deg C, was determined by PCVN specimens at -90 deg C. This value is similar to that of other high copper welds. The initial RT NDT was conservatively estimated as -26 deg F from the current fracture toughness results. From the studies on the chemistry and microstructure, the fracture mechanical properties of WF-233 weld is convincingly not worse than WF-70 and 72W welds

  5. Photonuclear reactions of U-233 and Pu-239 near threshold induced by thermal neutron capture gamma rays

    International Nuclear Information System (INIS)

    Moraes, M.A.P. de.

    1990-01-01

    The photonuclear cross sections of U-293 and Pu-239 have been studied by using monochromatic and discrete photons, in the energy interval from 5.49 to 9.72 MeV, produced by thermal neutron capture. The gamma fluxes incident on the samples were measured using a ( 3 x 3 )'' NaI (TI) crystal. The photofission fragments were detected in Makrofol-Kg (SSNTD). A possible structure was observed in the U-233 cross sections, near 7.23 MeV. The relative fissionability of the nuclides was determined at each excitation energy and shown to be energy independent: ( 2.12 ± 0.25) for U-233 and ( 3.32 ± 0.41 ) for Pu-239. The angular distribution of photofission fragments of Pu-239 were measured at two mean excitation energies of 5.43 and 7.35 MeV. An anisotropic distribution of ( 12.2 ± 3.6 ) % was observed at 5.43 MeV. The total neutron cross sections were measured by using a long counter detector. The photoneutron cross sections were calculated by using energy dependent neutron multiplicities values, γ(E), obtained in the literature. The competition Γn/γf was also determined at each excitation energy, and shown to be energy independent: ( 0.54 ± 0.05 ) for U-233 and ( 0.44 ± 0.05 ) for Pu-239, and were correlated to the parameters Z sup(2)/A, ( Ef'-Bn'), A. According to the FUJIMOTO-YAMAGUCHI and CONSTANT NUCLEAR TEMPERATURE models, the nuclear temperatures were calculated. The total photoabsorption cross sections were also calculated as a sum of the photofission and photoneutron cross sections at each energy excitation. From these results the competition Γf/ΓA, called fission probability Pf, were obtained: ( 0.66 ± 0.02) for U-233 and ( 0.70 ± 0.02 ) for Pu-239. (author)

  6. Use of sup(233)U for high flux reactors

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Liem, P.H.

    1991-01-01

    The feasibility design study on the graphite moderated gas cooled reactor as a high flux reactor has been performed. The core of the reactor is equipped with two graphite reflectors, i.e., the inner reflector and the outer reflector. The highest value of the thermal neutron flux and moderately high thermal neutron flux are expected to be achieved in the inner reflector region and in the outer reflector region respectively. This reactor has many merits comparing to the conventional high flux reactors. It has the inherent safety features associated with the modular high temperature reactors. Since the core is composed with pebble bed, the on-power refueling can be performed and the experiment time can be chosen as long as necessary. Since the thermal-to-fast flux ratio is large, the background neutron level is low and material damage induced by fast neutrons are small. The calculation was performed using a four groups diffusion approximation in a one-dimensional spherical geometry and a two-dimensional cylindrical geometry. By choosing the optimal values of the core-reflector geometrical parameters and moderator-to-fuel atomic density, high thermal neutron flux can be obtained. Because of the thermal neutron flux can be obtained. Because of the thermal design constraint, however, this design will produce a relatively large core volume (about 10 7 cc) and consequently a higher reactor power (100 MWth). Preliminary calculational results show that with an average power density of only 10 W/cc, maximum thermal neutron flux of 10 15 cm -2 s -1 can be achieved in the inner reflector. The eta value of 233 U is larger than 235 U. By introducing 233 U as the fissile material for this reactor, the thermal neutron flux level can be increased by about 15%. (author). 3 refs., 2 figs., 4 tabs

  7. Study on the adsorption of 233Pa in glass

    International Nuclear Information System (INIS)

    Natsumi, R.R.; Saiki, M.; Lima, F.W. de.

    1982-08-01

    It is intended to examine the adsorption of protactinium on glass in relation to pH, presence of complexing agents concentration and type of electrolytes. The study was made by using carrier-free 233 Pa solution and Pyrex glass tube was selected as adsorbent glass material surface. The adsorption curve of protactinium on glass surface as a function of the pH of the tracer solution showed the existence of two pronounced adsorption regions. It was found that this adsorption can be reduced by using electrolytes or complexing agents. Desorption of protactinium previously adsorbed on the Pyrex glass tube was also studied. Hidrochloric, oxalic and hydrofluoric acid solutions were used for the desorption experiments. (Author) [pt

  8. Study of the excited levels of 233{sup P}a by the 237{sup N}p alpha decay; Estudio de los niveles excitados en el 233{sup P}a por la desintegracion alfa del 237{sup N}p

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J; Gaeta, R; Vano, E; Los Arcos, J M

    1978-07-01

    The excited levels in 233{sup P}a following the 237{sup N}p alpha decay have been studied, by performing different experiences to complete available data and supply new information. Thus, two direct alpha spectrum measurement, one alpha-gamma bidimensional coincidence experiment, three gamma-gamma and gamma-X ray coincidences and some other measurements of the gamma spectrum, direct and coincident with alpha-particles have been made. These last experiences have allowed to obviate usual radiochemical separation methods, the 233{sup P}a radioactive descendent interferences being eliminated by means of the coincidence technic. As a result, a primary decay scheme has been elaborated, including 15 new gamma transitions and two new levels, not observed in the most recent works. (Author) 60 refs.

  9. Migration of uranium process wastes from the uranium-233--thorium-232 cycle

    International Nuclear Information System (INIS)

    Fried, S.; Sabau, C.; Hines, J.; Friedman, A.

    1978-03-01

    With the advent of fuel loadings of 233 U in the Shippingport Reactor, it has become important to understand the migratory behavior of uranium. The purpose of this study is the determination of the parameters influencing the migration of U(VI), the most likely chemical form of uranium to be mobilized from a repository. Samples of rhyolite tuff were used to measure the absorption coefficients of solutions of U(VI) in ground waters. In addition, columns of tuff were used to measure the elution behavior of U(VI) at various conditions of pH, U(VI) concentration, and flow saturation. These results indicate that there are several elution peaks with values of K/sub d/ between 35 and 120. This behavior is not the same as that of Pu(VI) on tuff; and the experimental results to date have not revealed the reason for this difference. Values of K/sub d/ in this range imply that geological containment would be difficult in strata of this type. It may be possible to find more retentive strata than tuff. Rocks containing reducing components are the most likely candidates and further investigation is urgently needed if the 233 U-Th cycle is to be widely used

  10. Storage and disposition of weapons usable fissile materials (FMD) PEIS: Blending of U-233 to <12% or <5% enrichment at the Idaho National Engineering Laboratory. Data report, Draft: Version 1

    International Nuclear Information System (INIS)

    Shaber, E.L.

    1995-08-01

    Uranium-233 (U-233), a uranium isotope, is a fissionable material capable of fueling nuclear reactors or being utilized in the manufacturing of nuclear weapons. As such, it is controlled as a special nuclear material. The Idaho National Engineering Laboratory (INEL) and Oak Ridge National Laboratory (ORNL) currently store the Department of Energy's (DOE's) supply of unirradiated U-233 fuel materials. Irradiated U-233 is covered by the national spent nuclear fuel (SNF) program and is not in the scope of this report. The U-233 stored at ORNL is relatively pure uranium oxide in the form of powder or monolithic solids. This material is currently stored in stainless steel canisters of variable lengths measuring about 3 inches in diameter. The ORNL material enrichment varies with some material containing considerable amounts of U-235. The INEL material is fuel from the Light Water Breeder Reactor (LWBR) Program and consists of enriched uranium and thorium oxides in zircaloy cladding. The DOE inventory of U-233 contains trace quantities of U-232, and daughter products from the decay of U-232 and U-233, resulting in increased radioactivity over time. These increased levels of radioactivity generally result in the need for special handling considerations

  11. A LMFBR for thorium utilization and for the U233/Th fuel rods specification

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Dias, A.F.

    1982-01-01

    The use of U 233 /Th as fuel in the middle part of LMFBR core and the Pu/U in the external part of the core, are proposed. The basic neutronic and safety characteristics and the specifications of fuel rods to be used in the internal core, are presented. (E.G.) [pt

  12. Thermoionic emission characteristics of uranium with application to its determination by MSID technique using 233U tracer

    International Nuclear Information System (INIS)

    Shihomatsu, H.M.; Iyer, S.S.

    1988-01-01

    Experimental details of the uranium determination in geological samples (50-1500 ppm range) by mass spectrometric isotope dilution technique (MSID) employing 233 U tracer are presented. For this purpose the thermoionic emission characteristics of uranium in various filament arrangements like simple plane, filament boat, double, are studied and the most efficient one selected for the isotope dilution analysis. The various experimental procedures involved in the MSID like sample dissolution, chemical separation and mass spectrometric analysis are developed and optimised. The experimental results on the uranium determination by MSID with 233 U tracer yielded precision and accuracy of 0,5% and 1% respectively. The importance of the sampling in the precise and accuracy determination of uranium in geological samples, where it is heterogeneously distributed, is discussed. (author) [pt

  13. ALARA Review of the Activation/Repair of Fire Detectors in Zone Three at the 233-S Facility

    International Nuclear Information System (INIS)

    Kornish, M.J.

    1998-07-01

    A formal as low as reasonably achievable (ALARA) review is required by BHI-SH-02, Vol. 1, Procedure 1.22, 'Planning Radiological Work', when radiological conditions exceed trigger levels. The level of contamination inside the viewing room meets this criterion. This ALARA review is for task instruction 1997-03-18-005-8.3.3 (mini task instruction to a living work package), 'Instructions for D ampersand D Support of Fire Detector Troubleshooting and Minor Maintenance Work at 233-S,' and DynCorp 2G-98-7207C, '233-S Reconnect Smoke Detectors Zone 3.' The Radiological Work Permit (RWP) request broke these two task instructions into four separate tasks. The four tasks identified in the RWP request were used to estimate airborne concentrations and the total exposure

  14. Study of the isotopic exchange associated with ionic exchange for the radiochemical separation of 233-Th

    International Nuclear Information System (INIS)

    Sepulveda Munita, C.J.A.

    1983-01-01

    The isotopic ion exchange procedure is applied in order to establish an analytical method for the determination of thorium by means of the 233 Th activity, when the presence of interfering elements does not allow a direct non-destructive activation analysis. The separation is based on the retention of 233 Th by a thorium saturated resin, due to the isotopic exchange effect, and subsequent elution of the interfering radioisotopes with a solution of thorium in diluted hydrochloric acid. The interfering elements were those which either present a great affinity for the resin or emit gamma rays with energies close to that of 233 Th (86.6 KeV), when a NaI(Tl) detector is used to obtain the gama-ray spectra of the irradiated samples. The equilibrium time for the thorium isotopic ion exchange and the distribution coefficients for the interfering elements were determined by using Bio-Rad AG 50W resins (100-200 mesh), with 4% to 8% of divinylbenzene. The best separation conditions were established in terms of the thorium and hydrochloric acid concentrations in the solution, the resin cross-linking degree, and the solution flow through the resin. The analytical method was applied to the determination of thorium in samples of ammonium diuranate as well in standard rock samples from the United States Geological Survey. The sensitivity, precision and accuracy of the method are also discussed. (Author) [pt

  15. Evaluation of temperature coefficients of reactivity for 233U--thorium fueled HTGR lattices. Final report

    International Nuclear Information System (INIS)

    Newman, D.F.; Leonard, B.R. Jr.; Trapp, T.J.; Gore, B.F.; Kottwitz, D.A.; Thompson, J.K.; Purcell, W.L.; Stewart, K.B.

    1977-05-01

    A comparison of calculated and measured neutron multiplication factors as a function of temperature was made for three graphite-moderated lattices in the High Temperature Lattice Test Reactor (HTLTR) using 233 UO 2 --ThO 2 fuels in varying amounts and configurations. Correlation of neutronic analysis methods and cross section data with the experimental measurements forms the basis for assessing the accuracy of the methods and data and developing confidence in the ability to predict the temperature coefficient of reactivity for various High Temperature Gas-Cooled Reactor (HTGR) conditions in which 233 U and thorium are present in the fuel. The calculated values of k/sub infinity/(T) were correlated with measured values using two least-squares-fitted correlation coefficients: (1) a normalization factor, and (2) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross section data

  16. 45 CFR 233.107 - Restriction in payment to households headed by a minor parent.

    Science.gov (United States)

    2010-10-01

    ... § 233.90(c)(1)(v) of this part provided that the residence is maintained as a home for the minor parent... the minor parent or dependent child would be jeopardized if they resided in the same residence with... residence of (i) a natural or adoptive parent or a stepparent, or (ii) a legal guardian as defined by the...

  17. Storage and disposition of weapons usable fissile materials (FMD) PEIS: Blending of U-233 to {lt}12% or {lt}5% enrichment at the Idaho National Engineering Laboratory. Data report, Draft: Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Shaber, E.L.

    1995-08-01

    Uranium-233 (U-233), a uranium isotope, is a fissionable material capable of fueling nuclear reactors or being utilized in the manufacturing of nuclear weapons. As such, it is controlled as a special nuclear material. The Idaho National Engineering Laboratory (INEL) and Oak Ridge National Laboratory (ORNL) currently store the Department of Energy`s (DOE`s) supply of unirradiated U-233 fuel materials. Irradiated U-233 is covered by the national spent nuclear fuel (SNF) program and is not in the scope of this report. The U-233 stored at ORNL is relatively pure uranium oxide in the form of powder or monolithic solids. This material is currently stored in stainless steel canisters of variable lengths measuring about 3 inches in diameter. The ORNL material enrichment varies with some material containing considerable amounts of U-235. The INEL material is fuel from the Light Water Breeder Reactor (LWBR) Program and consists of enriched uranium and thorium oxides in zircaloy cladding. The DOE inventory of U-233 contains trace quantities of U-232, and daughter products from the decay of U-232 and U-233, resulting in increased radioactivity over time. These increased levels of radioactivity generally result in the need for special handling considerations.

  18. Water Ingress Testing of the Turbula Jar and U-233 Lead Pig Containers

    Energy Technology Data Exchange (ETDEWEB)

    Reeves, Kirk Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Karns, Tristan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Smith, Paul Herrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-02

    Understanding the water ingress behavior of containers used at the TA-55 Plutonium Facility has significant implications for criticality safety. The purpose of this report is to document the water ingress behavior of the Turbula Jar with Bakelite lid and Viton gaskets (Turbula Jar) used in oxide blending operations and the U-233 lead pig container used to store and transport U-233 material. The technical basis for water resistant containers at TA-55 is described in LA-UR-15-22781, “Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program.” Testing of the water ingress behavior of various containers is described in LA-CP-13-00695, “Water Penetration Tests on the Filters of Hagan and SAVY Containers,” LA-UR-15-23121, “Water Ingress into Crimped Convenience Containers under Flooding Conditions,” and in LA-UR- 16-2411, “Water Ingress Testing for TA-55 Containers.” Water ingress criteria are defined in TA55-AP-522 “TA-55 Criticality Safety Program”, and in PA-RD-01009 “TA55 Criticality Safety Requirements.” The water ingress criteria for submersion is no more than 50 ml of water ingress at a 6” water column height for a period of 2 hours.

  19. Comparison of the U-233 dog data of Stevens et al. with uranium retention functions in ICRP Publication 30 and a 3-compartment mammillary model for uranium

    International Nuclear Information System (INIS)

    Bernard, S.R.

    1983-01-01

    Stevens measured the distribution, retention, and excretion of U-233 in seven beagles each given a single injection of U-233 citrate [2.8 μCi/kg U-233 (VI) (approx.3 mg/dog)]. These data, when plotted together with results obtained with the ICRP (Pub. 30) retention functions for purposes of comparison, are seen to differ only slightly from the ICRP-30 model. The number of transformations in the body, over a fifty-year period agree within a factor of 2. A three-compartment mammillary model has been parameterized from the data of Stevens by the method of Bernard. Retention in tissues of the body is represented by a linear combination of three compartments. The data plots for the dogs and ICRP-30 model will be presented and discussed together with the three compartment mammillary model for U-233 retention, distribution, and excretion. 3 figs., 2 tabs

  20. Analysis of Hydrogen Generation and Accumulation in U-233 Tube Vaults

    International Nuclear Information System (INIS)

    Ally, M.R.; Willis, K.J.

    1999-01-01

    The purpose of the 233 U Safe Storage Program is to enhance the safe storage of 233 U-bearing materials. This report describes the work done at the Oak Ridge National Laboratory's Radiochemical Development Facility (RDF) to address questions related to possible hydrogen generation and accumulation in 233 U tube vaults. The objective of this effort was to verify assumptions in the mathematical model used to estimate the hydrogen content of the gaseous atmosphere that possibly could occur inside the tube vaults in Building 3019 and to evaluate proposed measures for mitigating any hydrogen concerns. A mathematical model was developed using conservative assumptions to evaluate possible hydrogen generation and accumulation in the tube vaults. The model concluded that an equilibrium concentration would be established below the lower flammability limit (LFL) of 4.1% hydrogen. The major assumptions used in the model that were validated are as follows: (1) The shield plug does not form a seal with the tube vault wall, thus allowing the hydrogen gas to diffuse past the shield plug to the upper section of the tube vault. (2) The tube vault end-cap leaks sufficiently to allow air to be drawn into the tube vault by the off-gas system, thereby purging hydrogen from the upper section of the tube vault. (3) Any hydrogen gas generated completely mixes with the other gases present in the lower section of the tube vault and does not stratify beneath the shield plug. (4) The diffusion coefficient determined from the literature for constant diffusion of hydrogen in air is valid. The coefficient is corrected for temperatures from 0 to 25 C. Another assumption used in the model, that hydrogen generated by radiolytic decomposition of hydrogen-bearing materials (e.g., moisture and plastic) leaks from the cans under steady-state condition, as opposed to a sudden release resulting from rupture of the can(s), was beyond the scope of this investigation. Several parameters from the original

  1. Differences in the behavior of 233Pa, 237Np and 239 Pu in bentonite contaminated by sulfate-reducing bacteria

    International Nuclear Information System (INIS)

    Kudo, A.; Fujikawa, Y.; Takigami, H.; Zheng, J.; Asano, H.; Arai, K.; Yoshikawa, H.; Ito, M.

    1998-01-01

    The behaviors of 233 Pa, 237 Np and 239 Pu in high level radioactive wastes from nuclear fuel reprocessing were investigated by a laboratory experiment. Radioactive wastes are glassified and disposed of in geological repositories encased in bentonite as an additional artificial barrier to protect the environment. There is, however, the possibility that some anaerobic bacteria, especially sulfate-reducing bacteria, may flourish within the bentonite during the long disposal period (more than a century). The effects of sulfate-reducing bacteria on the behavior of the radionuclides within bentonite were investigated using the distribution coefficient (Kd) of 233 Pa, 237 Np and 239 Pu. The Kd was obtained with a 0.22 m membrane filter separating radionuclide contents in solid and liquid phases. The anaerobic bacteria, including sulfate-reducing bacteria, used for this investigation originated from the anaerobic treatment of pulp and paper waste and operated for more than one year at Eh around -85 mV. The bentonite used for this study was produced in Japan. The active anaerobic bacteria clearly accumulates considerable amounts of 233 Pa and 239 Pu by producing high Kd values of nearly 100,000, while Kds of 233 Pa and 239 Pu for the sterilized anaerobic bacteria were less than 10,000. In other words, live anaerobic bacteria can hold considerably higher amounts of the radionuclides compared to dead bacteria. Furthermore, high Kd values were obtained for anaerobic bacteria at pH 5-9. In contrast, Kd values for the radionuclide 237 Np were not influenced by the anaerobic bacteria but were controlled by chemical environmental conditions such as like pH. Another comparison was conducted for the radionuclides for mixtures of non-sterilized bacteria with bentonite. (author)

  2. On-site transportation and handling of uranium-233 special nuclear material: Preliminary hazards and accident analysis. Final

    International Nuclear Information System (INIS)

    Solack, T.; West, D.; Ullman, D.; Coppock, G.; Cox, C.

    1995-01-01

    U-233 Special Nuclear Material (SNM) currently stored at the T-Building Storage Areas A and B must be transported to the SW/R Tritium Complex for repackaging. This SNM is in the form of oxide powder contained in glass jars which in turn are contained in heat sealed double polyethylene bags. These doubled-bagged glass jars have been primarily stored in structural steel casks and birdcages for approximately 20 years. The three casks, eight birdcages, and one pail/pressure vessel will be loaded onto a transport truck and moved over an eight day period. The Preliminary Hazards and Accident Analysis for the on-site transportation and handling of Uranium-233 Special Nuclear Material, documented herein, was performed in accordance with the format and content guidance of DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports, dated July 1994, specifically Chapter Three, Hazard and Accident Analysis. The Preliminary Hazards Analysis involved detailed walkdowns of all areas of the U-233 SNM movement route, including the T-Building Storage Area A and B, T-Building truck tunnel, and the roadway route. Extensive discussions were held with operations personnel from the Nuclear Material Control Group, Nuclear Materials Accountability Group, EG and G Mound Security and the Material Handling Systems Transportation Group. Existing documentation related to the on-site transportation of hazardous materials, T-Building and SW/R Tritium Complex SARs, and emergency preparedness/response documentation were also reviewed and analyzed to identify and develop the complete spectrum of energy source hazards

  3. 45 CFR 233.35 - Computing the assistance payment under retrospective budgeting after the initial one or two...

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Computing the assistance payment under... FINANCIAL ASSISTANCE PROGRAMS § 233.35 Computing the assistance payment under retrospective budgeting after... shall be computed retrospectively, i.e., shall be based on income and other relevant circumstances in...

  4. 45 CFR 233.34 - Computing the assistance payment in the initial one or two months (AFDC).

    Science.gov (United States)

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Computing the assistance payment in the initial... § 233.34 Computing the assistance payment in the initial one or two months (AFDC). A State shall compute...) If the initial month is computed prospectively as in paragraph (a) of this section, the second month...

  5. Fiscal Year 2008 Phased Construction Completion Report for EU Z2-33 in Zone 2, East Tennessee Technology Park, Oak Ridge, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    Bechtel Jacobs

    2008-09-11

    The Record of Decision for Soil, Buried Waste, and Subsurface Structure Actions in Zone 2, East Tennessee Technology Park, Oak Ridge, Tennessee (DOE/OR/01-2161&D2) (Zone 2 ROD) acknowledged that most of the 800 acres in Zone 2 were contaminated, but that sufficient data to confirm the levels of contamination were lacking. The Zone 2 ROD further specified that a sampling strategy for filling the data gaps would be developed. The Remedial Design Report/Remedial Action Work Plan for Zone 2 Soils, Slabs, and Subsurface Structures, East Tennessee Technology Park, Oak Ridge, Tennessee (DOE/OR/01-2224&D3) (Zone 2 RDR/RAWP) defined the sampling strategy as the Dynamic Verification Strategy (DVS), generally following the approach used for characterization of the Zone 1 exposure units (EUs). The Zone 2 ROD divided the Zone 2 area into seven geographic areas and 44 EUs. To facilitate the data quality objectives (DQOs) of the DVS process, the Zone 2 RDR/RAWP regrouped the 44 EUs into 12 DQO scoping EU groups. These groups facilitated the DQO process by placing similar facilities and their support facilities together and allowing identification of data gaps. The EU groups were no longer pertinent after DQO planning was completed and characterization was conducted as areas became accessible. As the opportunity to complete characterization became available, the planned DVS program and remedial actions (RAs) were completed for EU Z2-33. Remedial action was also performed at two additional areas in adjacent EU Z2-42 because of their close proximity and similar nature to a small surface soil RA in EU Z2-33. Remedial actions for building slabs performed in EU Z2-33 during fiscal year (FY) 2007 were reported in the Fiscal Year 2007 Phased Construction Completion Report for the Zone 2 Soils, Slabs, and Subsurface Structures at East Tennessee Technology Park, Oak Ridge, Tennessee (DOE/OR/01-2723&D1). Recommended RAs for EU Z2-42 were described in the Fiscal Year 2006 Phased Construction

  6. Use of nuclear recoil for separating 228Ra, 224Ra, and 233Pa from colloidal thorium

    International Nuclear Information System (INIS)

    Beydon, J.; Gratot, I.

    1968-01-01

    By using α-recoil it is possible to separate by dialysis the α disintegration products (224 Ra; 228 Ra) of thorium from colloidal thorium hydroxide.The use of n, γ recoil allows the separation of 233 Pa produced by the neutron irradiation of thorium, on condition that the colloidal thorium hydroxide is irradiated in the presence of a dispersing. (author) [fr

  7. Using {sup 233}U-doped crystals to access the few-eV isomeric transition in {sup 229}Th

    Energy Technology Data Exchange (ETDEWEB)

    Stellmer, Simon; Schreitl, Matthias; Kazakov, Georgy A.; Sterba, Johannes H.; Schumm, Thorsten [Vienna Center for Quantum Science and Technology (VCQ) and Atominstitut, TU Wien, Vienna (Austria)

    2016-07-01

    The isotope {sup 229}Th possesses an exceptionally low-lying isomeric state at an energy of only a few eV. While direct laser excitation of the isomer is a tantalizing future prospect, the stage is not yet set for nuclear laser spectroscopy: too little is known about the energy, lifetime, and internal conversion pathways of the isomer. Alternative routes to populate the isomer are needed for further investigations. We use the alpha decay {sup 233}U →{sup 229g,m}Th to populate the isomer with a probability of 2%. The {sup 233}U is embedded into VUV-transparent crystals, as the isomer transition is expected around 160 nm. The wavelength of the gamma ray, emitted upon de-excitation of the isomer into the ground state, is measured with a spectrometer. Calculations show that the isomer emission is not obscured by radioluminescence of the crystal. We report on the current status of the experiment.

  8. Determination of the extraction efficiency for {sup 233}U source α-recoil ions from the MLL buffer-gas stopping cell

    Energy Technology Data Exchange (ETDEWEB)

    Wense, Lars v.d.; Seiferle, Benedict; Thirolf, Peter G. [Ludwig-Maximilians-Universitaet Muenchen, Garching (Germany); Laatiaoui, Mustapha [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); Helmholtz Institut Mainz, Mainz (Germany)

    2015-03-01

    Following the α decay of {sup 233}U, {sup 229}Th recoil ions are shown to be extracted in a significant amount from the MLL buffer-gas stopping cell. The produced recoil ions and subsequent daughter nuclei are mass purified with the help of a customized quadrupole mass spectrometer. The combined extraction and mass purification efficiency for {sup 229}Th{sup 3+} is determined via MCP-based measurements and via the direct detection of the {sup 229}Th α decay. A large value of (10±2)% for the combined extraction and mass purification efficiency of {sup 229}Th{sup 3+} is obtained at a mass resolution of about 1u/e. In addition to {sup 229}Th, also other α-recoil ions of the {sup 233,} {sup 232}U decay chains are addressed. (orig.)

  9. Development of automation and remotisation systems for fabrication of (Th-233U)O2 MOX fuel for AHWR

    International Nuclear Information System (INIS)

    Saraswat, Anupam; Danny, K.M.; Chakraborty, S.; Somayajulu, P.S.; Kumar, Arun; Mittal, R.; Prasad, R.S.; Mahule, K.N.; Panda, S.; Jayarajan, K.

    2011-01-01

    To meet the ever increasing power requirement of India, country is planning to utilize its large thorium reserves for the third stage of nuclear power program based on Thorium-Uranium 233 fuel in A.H.W.R. Although there are many advantages of (Th- 233 U)O 2 fuel cycle, presence of radiological hazards due to the presence of 1000-2000 ppm level of 232 U in the 233 U fuel and inertness of ThO 2 makes handling and fabrication of fuel difficult. The associated high alpha and gamma activity demands high level of automation and remote handling in alpha tight hot cells. To demonstrate automation and remotisation in (Th- 233 U)O 2 fuel fabrication, a mock up facility is being set up at BARC. This facility shall develop automation systems required for remote fuel fabrication in a simulated hot cell environment. There are many innovative schemes and systems being developed like integrated powder pellet system, remote viewing system for hot cell application etc. Low visibility inside the hot cell has always been a problem for the operator. To overcome this problem a remote viewing system has been developed by which entire hot cell area can be scanned with the use of a joystick and the display can be seen on a LCD monitor. The viewing system is made up of radiation resistant optics which can work even in high gamma fields. It consists of objective end assembly which is used to scan the hot cell area with the help of prism doublets and drive mechanism for capturing full 360 deg solid angle view. There is a Galilean telescope and focusing system used for focusing images of distant objects. Drive mechanism can be controlled by the joystick available to the operator. System has a high resolution CCD display and camera which gives a clear display of objects lying inside the hot cell area. Integrated powder pellet system is being developed for fabrication of MOX pellets from feed powder. This will be automated system which will take input in the form of MOX powder and convert it

  10. Resonance Region Covariance Analysis Method and New Covariance Data for Th-232, U-233, U-235, U-238, and Pu-239

    International Nuclear Information System (INIS)

    Leal, Luiz C.; Arbanas, Goran; Derrien, Herve; Wiarda, Dorothea

    2008-01-01

    Resonance-parameter covariance matrix (RPCM) evaluations in the resolved resonance region were done for 232Th, 233U, 235U, 238U, and 239Pu using the computer code SAMMY. The retroactive approach of the code SAMMY was used to generate the RPCMs for 233U, 235U. RPCMs for 232Th, 238U and 239Pu were generated together with the resonance parameter evaluations. The RPCMs were then converted in the ENDF format using the FILE32 representation. Alternatively, for computer storage reasons, the FILE32 was converted in the FILE33 cross section covariance matrix (CSCM). Both representations were processed using the computer code PUFF-IV. This paper describes the procedures used to generate the RPCM with SAMMY.

  11. Corrigendum to “Representations for the extreme zeros of orthogonal polynomials” [J. Comput. Appl. Math. 233 (2009) 847–851

    OpenAIRE

    van Doorn, Erik A.; van Foreest, N.D.; Zeifman, Alexander I.

    2013-01-01

    We correct representations for the endpoints of the true interval of orthogonality of a sequence of orthogonal polynomials that were stated by us in the Journal of Computational and Applied Mathematics 233 (2009) 847–851.

  12. An in-depth analysis identifies two new independent signals in 11q23.3 associated with vitiligo in the Chinese Han population.

    Science.gov (United States)

    Zhao, Suli; Fang, Fang; Tang, Xianfa; Dou, Jinfa; Wang, Wenjun; Zheng, Xiaodong; Sun, Liangdan; Zhang, Anping

    2017-10-01

    Vitiligo is an autoimmune disease, characterized by progressive loss of skin pigmentation, which is caused by the interactions of multiple factors, such as heredity, immunity and environment. Recently, a single nucleotide polymorphism (SNP) rs638893 at 11q23.3 region was identified as a risk factor for vitiligo in genome-wide association studies and multiple SNPs in this region have been associated with other autoimmune diseases. This study aims to identify additional susceptibility variants associated with vitiligo at 11q23.3 in the Chinese Han population. We selected and genotyped 26 SNPs at 11q23.3 in an independent cohort including 2924 cases and 4048 controls using the Sequenom MassArray iPLEX ® system. Bonferroni adjustment was used for multiple comparisons and P value vitiligo (OR=1.21, 95% CI: 1.11-1.31, P=1.20×10 -5 ; OR=1.14, 95% CI: 1.07-1.23, P=1.90×10 -4 , respectively). The C allele of rs638893 (a previously reported one) located upstream of DDX6 was also significantly associated with vitiligo (OR=1.25, 95% CI: 1.12-1.38, P=3.04×10 -5 ). The genotypes distribution of 3 SNPs also showed significant differences between case and control (rs613791: P=7.00×10 -6 , rs523604: P=4.00×10 -3 , rs638893: P=1.20×10 -5 , respectively). The two newly identified SNPs (rs613791 and rs523604) showed independent associations with vitiligo by linkage disequilibrium analysis and conditional logistic regression. The study identified two new independent signals in the associated locus 11q23.3 for vitiligo. The presence of multiple independent variants emphasizes an important role of this region in disease susceptibility. Copyright © 2017 Japanese Society for Investigative Dermatology. Published by Elsevier B.V. All rights reserved.

  13. Critical Experiments With Aqueous Solutions of 233UO2(NO3)2

    International Nuclear Information System (INIS)

    Thomas, J.T.

    2001-01-01

    This report provides the critical experimenter's interpretations and descriptions of informal critical experiment logbook notes and associated information (e.g., experimental equipment designs/sketches, chemical and isotopic analyses, etc.) for the purpose of formally documenting the results of critical experiments performed in the late 1960s at the Oak Ridge Critical Experiments Facility. The experiments were conducted with aqueous solutions of 97.6 wt % 233 U uranyl nitrate having uranium densities varying between about 346 g U/l and 45 g U/l. Criticality was achieved with single simple units (e.g., cylinders and spheres) and with spaced subcritical simple cylindrical units arranged in unreflected, water-reflected, and polyethylene reflected critical arrays

  14. Within the framework of the new fuel cycle 232Th/233U, determination of the 233Pa(n.γ) radiative capture cross section for neutron energies ranging between 0 and 1 MeV

    International Nuclear Information System (INIS)

    Boyer, S.

    2004-10-01

    The Thorium cycle Th 232 /U 233 may face brilliant perspectives through advanced concepts like molten salt reactors or accelerator driven systems but it lacks accurate nuclear data concerning some nuclei. Pa 233 is one of these nuclei, its high activity makes the direct measurement of its radiative neutron capture cross-section almost impossible. This difficulty has been evaded by considering the transfer reaction Th 232 (He 3 ,p)Pa 234 * in which the Pa 234 nucleus is produced in various excited states according to the amount of energy available in the reaction. The first chapter deals with the thorium cycle and its assets to contribute to the quenching of the fast growing world energy demand. The second chapter gives a detailed description of the experimental setting. A scintillation detector based on deuterated benzene (C 6 D 6 ) has been used to counter gamma ray cascades. The third chapter is dedicated to data analysis. In the last chapter we compare our experimental results with ENDF and JENDL data and with computed values from 2 statistical models in the 0-1 MeV neutron energy range. Our results disagree clearly with evaluated data: our values are always above ENDF and JENDL data but tend to near computed values. We have also perform the measurement of the radiative neutron cross-section of Pa 231 for a 110 keV neutron: σ(n,γ) 2.00 ± 0.14 barn. (A.C.)

  15. Fabrication of zero power reactor fuel elements containing 233U3O8 powder

    International Nuclear Information System (INIS)

    Nicol, R.G.; Parrott, J.R.; Krichinsky, A.M.; Box, W.D.; Martin, C.W.; Whitson, W.R.

    1982-05-01

    Oak Ridge National Laboratory, under contract with Argonne National Laboratory, completed the fabrication of 1743 fuel elements for use in their Zero Power Reactor. The contract also included recovery of 20 kg of 233 U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U 3 O 8 powder (suitable for fuel) and subsequent charging, sealing, decontamination, and testing of the fuel elements (packets) preparatory to shipment. The nuclear safety, radiation exposures, and quality assurance aspects of the program are discussed

  16. Retention and translocation of inhaled uranyl nitrate (233U and 232U) in rats

    International Nuclear Information System (INIS)

    Ballou, J.E.; Gies, R.A.; Wogman, N.A.

    1978-01-01

    The uranium-thorium breeder reactors proposed for nuclear power production, and other thorium fuel systems in conventional reactors, utilize fuels and fuel recycle process solutions that have not been evaluated for biological hazard. This project emphasizes studies of the metabolism of the oxide fuels and the nitrate solutions of the major radionuclides, following inhalation, ingestion, or cutaneous application in rodents. Preliminary data are reported for the clearance of inhaled 233 UO 2 (NO 3 ) 2 and 232 UO 2 (NO 3 ) 2 from the lung and their translocation to skeleton

  17. Investigation of the origin of elements of the uranium-235 family noticed in excess around the EL4 experimental nuclear reactor during its dismantling. Site of the Monts d'Arree - Brennilis (29) power plant. Years 2007-2008. Report and appendices with results

    International Nuclear Information System (INIS)

    2009-02-01

    The presence of actinium-227 has been noticed in the Mont d'Arree region (Finistere district) and such a presence in the environment had never been reported before. Thus, a study has been performed to investigate the origin of this element: about 300 samples have been analysed. After an indication of the investigation chronology, the report outlines that there is no relationship between the excess of actinium-227 and land amendment or embankments, that there is no obvious relationship between this presence and radioactive liquid effluents or atmospheric effluents from the Brennilis nuclear site. It shows that there is an obvious relationship with the radiological quality. It states that this excess of actinium 227 is related to the management of rain waters about the Brennilis site. An appendix specifies the location, nature and agenda of samplings (bio-indicators, samplings in sludge, soils, food and tap water, aquatic foams and sediments, ponds, wet lands, and at the vicinity of the power plant channel), presents detailed results obtained by gamma spectrometry, and measurement equipment and methods

  18. Young planets under extreme UV irradiation. I. Upper atmosphere modelling of the young exoplanet K2-33b

    Science.gov (United States)

    Kubyshkina, D.; Lendl, M.; Fossati, L.; Cubillos, P. E.; Lammer, H.; Erkaev, N. V.; Johnstone, C. P.

    2018-04-01

    The K2-33 planetary system hosts one transiting 5 R⊕ planet orbiting the young M-type host star. The planet's mass is still unknown, with an estimated upper limit of 5.4 MJ. The extreme youth of the system (age of the system indicates that the planet is more massive than 10 M⊕.

  19. Study of the mass, isotopic and kinetic energy distributions of the 233U(nth, f) and 241Pu(nth, f) fission products measured at the Lohengrin mass spectrometer (ILL)

    International Nuclear Information System (INIS)

    Martin, F.

    2013-01-01

    Fission product yields are significant nuclear data for neutronic simulations. The purpose of this work is to improve fission yield knowledge for two fissile nuclei: 241 Pu and 233 U. Those are respectively involved in the uranium and thorium nuclear fuel cycle. The measurements are performed at the Lohengrin mass spectrometer of the Institut Laue-Langevin (ILL) located in Grenoble. The spectrometer is combined with an ionization chamber to measure mass yields of 241 Pu and 233 U and with a gamma spectrometry set-up to determine isotopic yields of 233 U. A new analysis method of experimental data has been developed in order to control systematics and to reduce experimental biases. For the first time, the experimental variance-covariance matrix of our measured fission yields could be deduced. (author) [fr

  20. Mass dependence of azimuthal asymmetry in the fission of 232Th and 233,235,236,238U by polarized photons

    International Nuclear Information System (INIS)

    Denyak, V.V.; Khvastunov, V.M.; Paschuk, S.A.; Schelin, H.R.

    2013-01-01

    Fission of the even-even nuclei 232 Th, 236,238 U and even-odd nuclei 233,235 U by linearly polarized photons has been studied at excitation energies in the region of a giant dipole resonance. The performed investigations unambiguously showed the existence of the fragment mass dependence of the cross section azimuthal asymmetry in the photofission of 236 U and 238 U. In addition, the obtained results provided the first evidence for the possible difference between the asymmetry values in asymmetric and symmetric mass distribution regions in the case of 236 U. The measured cross section azimuthal asymmetry of the fission of 232 Th does not show any fragment mass dependence. In the even-odd nuclei 233 U and 235 U the difference between the far-asymmetric and other mass distribution regions was also observed but with the statistical uncertainty not small enough for definitive conclusion. (orig.)

  1. Amster: a molten-salt reactor concept generating its own 233U and incinerating transuranium elements

    International Nuclear Information System (INIS)

    Lecarpentier, D.; Garzenne, C.; Vergnes, J.; Mouney, H.; Delpech, M.

    2002-01-01

    In the coming century, sustainable development of atomic energy will require the development of new types of reactors able to exceed the limits of the existing reactor types, be it in terms of optimum use of natural fuel resources, reduction in the production of long-lived radioactive waste, or economic competitiveness. Of the various candidates with the potential to meet these needs, molten-salt reactors are particularly attractive, in the light of the benefits they offer, arising from two fundamental features: - A liquid fuel does away with the constraints inherent in solid fuel, leading to a drastic simplification of the fuel cycle, in particular making in possible to carry out on-line pyrochemical reprocessing; - Thorium cycle and thermal spectrum breeding. The MSBR concept proposed by ORNL in the 1970's thus gave a breeding factor of 1.06, with a doubling time of about 25 years. However, given the tight neutron balance of the thorium cycle (the η of 233 U is about 2.3), MSBR performance is only possible if there are strict constraints set on the in-line reprocessing unit: all the 233 Pa must be removed from the core so that it can decay on the 233 U in no more than about ten days (or at least 15 tonnes of salt to be extracted from the core daily), and the absorbing fission products, in particular the rare earths, must be extracted in about fifty days. With the AMSTER MSR concept, which we initially developed for incinerating transuranium elements, we looked to reduce the mass of salt to be reprocessed in order to minimise the size and complexity of the reprocessing unit coupled to the reactor, and the quantity of transuranium elements sent for disposal, as this is directly proportional to the mass of salt reprocessed for extraction of the fission products. Given that breeding was not an absolute necessity, because the reactor can be started by incinerating the transuranium elements from the spent fuel assemblies of current reactors, or if necessary by loading

  2. Cross sections and neutron yields for U233, U235 and Pu239 at 2200 m/sec

    International Nuclear Information System (INIS)

    Sjoestrand, N.G.; Story, J.S.

    1960-04-01

    The experimental information on the 2200 m/sec values for σ abs , σ f , α, ν and η for 233 U , 235 U and 23 been collected and discussed. The values will later be used in an evaluation of a 'best' set of data. In appendix the isotopic abundances of the uranium isotopes are discussed and also the alpha activities of the uranium isotopes and Pu-239

  3. Remaining Sites Verification Package for the 600-233 Waste Site, Vertical Pipe Near 100-B Electrical Laydown Area. Attachment to Waste Site Reclassification Form 2005-041

    International Nuclear Information System (INIS)

    Carlson, R.A.

    2005-01-01

    The 600-233 waste site consisted of three small-diameter pipelines within the 600-232 waste site, including previously unknown diesel fuel supply lines discovered during site remediation. The 600-233 waste site has been remediated to achieve the remedial action objectives specified in the Remaining Sites ROD. The results of verification sampling show that residual contaminant concentrations do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also demonstrate that residual contaminant concentrations are protective of groundwater and the Columbia River

  4. The European Expression Of Interest For High Purity U-233 Materials

    Energy Technology Data Exchange (ETDEWEB)

    Giaquinto, Joseph M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Younkin, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    The purpose of this letter report is to document the response for an Expression of Interest (EOI) sent to the European Safeguards and research and development (R&D) scientific communities for the distribution of small amounts of high purity 233U materials for use in safeguards, nonproliferation, and basic R&D in the nuclear disciplines. The intent for the EOI was to gauge the level of international interest for these materials from government and research institutions with programmatic missions in the nuclear security or nuclear R&D arena. The information contained herein is intended to provide information to assist key decision makers in DOE as to the ultimate disposition path for the high purity materials currently being recovered at Oak Ridge National Laboratory (ORNL) and only those items for which there is no United States (U.S.) sponsor identified.

  5. Fuel utilization improvement in PWRs using the denatured 233U-Th cycle

    International Nuclear Information System (INIS)

    Jones, H.M.; Schwenk, G.A.; Toops, E.C.; Yotinen, V.O.

    1980-06-01

    A number of changes in PWR core design and/or operating strategy were evaluated to assess the fuel utilization improvement achievable by their implementation in a PWR using thorium-based fuel and operating in a recycle mode. The reference PWR for this study was identical to the B and W Standard Plant except that the fuel pellets were of denatured ( 233 U/ 238 U-Th)O 2 . An initial scoping study identified the three most promising improvement concepts as (1) a very tight lattice, (2) thorium blankets, and (3) ThO 2 rods placed in available guide tubes. A conceptual core design incorporating these changes was then developed, and the fuel utilization of this modified design was compared with that of the reference case

  6. A study of sodium-cooled fast breeder reactor with thorium blanket for supply of U-233 to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Yoshida, H.; Nishimura, H.; Osugi, T.

    1978-08-01

    Symbiotic energy system between fast breeder reactor and thermal reactor would have a potential merit for nuclear proliferation problem. And when using HTGR as the thermal reactor in the system, the energy system appears to be promising as an energy system self-sufficient in fuels, which can generate both electricity and high temperature process heat. In the system the fast breeder reactor has to supply sufficient amount of fissile plutonium to keep the reactor going, and also produce U-233 necessary to the associated U-233 fuelled process heat production HTGR. Three types of LMFBR concepts with thorium blanket, conventional homogeneous core LMFBR, and axial and radial parfait heterogeneous core LMFBRs, have been investigated to find out suitable configurations of LMFBR for supply of U-233 to the HTGR with relatively high conversion ratio of 0.85, in the symbiotic energy system between LMFBR and HTGR. The investigation on LMFBR has been made on fuel sufficiency of the system, inherent safety such as sodium-void and Doppler coefficients, and fuel cycle cost. The followings were revealed; (1) Conventional homogeneous core LMFBR with thorium radial blanket well satisfies the condition of fuel sufficiency, if adequate radial blanket thickness is chosen. However, the sodium-void coefficient and fuel cycle cost are inferior to the other concepts. (2) Axial parfait heterogeneous core LMFBR can be regarded as one of the best LMFBR concepts installed in the symbiotic energy system, from the viewpoints of fuel sufficiency, inherent safety and fuel cycle cost. However, further investigations should be needed on reliability and operationability of the concept. (3) Radial parfait heterogeneous core LMFBR seems inadequate as the LMFBR in the system, because the configurations based on this concept does not satisfy plutonium and U-233 breedings, simultaneously. This LMFBR concept, however, has excellent breeding performance in the internal radial blanket. So further

  7. Evaluation of neutron nuclear data for 233U in thermal and resonance regions

    International Nuclear Information System (INIS)

    Kikuchi, Yasuyuki

    1981-02-01

    The thermal and resonance cross sections of 233 U were evaluated for JENDL-2. The cross sections below 1 eV are given as point-wise data and were evaluated by the use of the measured fission and capture cross sections. The resolved resonance parameters are derived up to 100 eV. The parameters were obtained by using NDES so as to reproduce the measured total and fission cross sections. The cross sections from 100 eV to 30 keV are represented by the unresolved resonance parameters. The fission and capture resonance integrals calculated from these parameters are 771 and 138 barns, respectively, which agree with the measured data within the quoted errors. (author)

  8. Thermal-Neutron-Induced Fission of U235, U233 and Pu239

    International Nuclear Information System (INIS)

    Thomas, T.D.; Gibson, W.M.; Safford, G.J.

    1965-01-01

    We have used solid-state detectors to measure the kinetic energies of the coincident fission fragments in the thermal-neutron-induced fission of U 235 , U 233 and Pu 239 . Special care has been taken to eliminate spurious-events near symmetry to give an accurate measure of such quantities as the average total kinetic energy at symmetry. For each fissioning system over 10 6 events were recorded. As a result the statistics are good enough to see definite evidence for fine structure over a wide range of masses and energies. The data have been analysed to give mass yield curves, average kinetic energies as a function of mass, and other quantities of interest. For each fissioning system the average total kinetic energy goes through a maximum for a heavy fragment mass of about 132 and for the corresponding light fragment mass. There is a pronounced minimum at symmetry, although not as deep as that found in time-of-flight experiments. The difference between the maximum average kinetic energy and that at symmetry is about 32 MeV for U 235 , 18 MeV for U 233 and 20 MeV for Pu 239 . The dispersion of kinetic energies at symmetry is also smaller than that found in time-of-flight experiments. Fine structure is apparent in two different representations of the data. The energy spectrum of heavy fragments in coincidence with light fragment energies is greater than the most probable value. This structure becomes more pronounced as the light fragment energy increases. The mass yield curves for a given total kinetic energy show a structure suggesting a preference for fission fragments with masses ∼134, ∼140 and ∼145 (and their light fragment partners). Much of the structure observed can be understood by considering a semi-empirical mass surface and a simple model for the nuclear configuration at the saddle point. (author) [fr

  9. Within the framework of the new fuel cycle {sup 232}Th/{sup 233}U, determination of the {sup 233}Pa(n.{gamma}) radiative capture cross section for neutron energies ranging between 0 and 1 MeV; Dans le cadre du nouveau cycle de combustible {sup 232}Th/{sup 233}U, determination de la section efficace de capture radiative {sup 233}Pa(n,{gamma}) pour des energies de neutrons comprises entre 0 et 1 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, S

    2004-10-15

    The Thorium cycle Th{sup 232}/U{sup 233} may face brilliant perspectives through advanced concepts like molten salt reactors or accelerator driven systems but it lacks accurate nuclear data concerning some nuclei. Pa{sup 233} is one of these nuclei, its high activity makes the direct measurement of its radiative neutron capture cross-section almost impossible. This difficulty has been evaded by considering the transfer reaction Th{sup 232}(He{sup 3},p)Pa{sup 234}* in which the Pa{sup 234} nucleus is produced in various excited states according to the amount of energy available in the reaction. The first chapter deals with the thorium cycle and its assets to contribute to the quenching of the fast growing world energy demand. The second chapter gives a detailed description of the experimental setting. A scintillation detector based on deuterated benzene (C{sub 6}D{sub 6}) has been used to counter gamma ray cascades. The third chapter is dedicated to data analysis. In the last chapter we compare our experimental results with ENDF and JENDL data and with computed values from 2 statistical models in the 0-1 MeV neutron energy range. Our results disagree clearly with evaluated data: our values are always above ENDF and JENDL data but tend to near computed values. We have also perform the measurement of the radiative neutron cross-section of Pa{sup 231} for a 110 keV neutron: {sigma}(n,{gamma}) 2.00 {+-} 0.14 barn. (A.C.)

  10. Cross sections and neutron yields for U-233, U-235 and Pu-239 at 2200 m/sec

    Energy Technology Data Exchange (ETDEWEB)

    Sjoestrand, N G; Story, J S

    1960-04-15

    The experimental information on the 2200 m/sec values for {sigma}{sub abs}, {sigma}{sub f}, {alpha}, {nu} and {eta} for {sup 233}U , {sup 235}U and {sup 23} been collected and discussed. The values will later be used in an evaluation of a 'best' set of data. In appendix the isotopic abundances of the uranium isotopes are discussed and also the alpha activities of the uranium isotopes and Pu-239.

  11. 6q16.3q23.3 duplication associated with Prader-Willi-like syndrome.

    Science.gov (United States)

    Desch, Laurent; Marle, Nathalie; Mosca-Boidron, Anne-Laure; Faivre, Laurence; Eliade, Marie; Payet, Muriel; Ragon, Clemence; Thevenon, Julien; Aral, Bernard; Ragot, Sylviane; Ardalan, Azarnouche; Dhouibi, Nabila; Bensignor, Candace; Thauvin-Robinet, Christel; El Chehadeh, Salima; Callier, Patrick

    2015-01-01

    Prader-Willi syndrome (PWS) is characterized by hypotonia, delayed neuropsychomotor development, overeating, obesity and mental deficiency. This phenotype is encountered in other conditions, defining Prader-Willi-like syndrome (PWLS). We report a 14-year-old boy with a complex small supernumerary marker chromosome (sSMC) associated with PWLS. The propositus presents clinical features commonly found in patients with PWLS, including growth hormone deficit. Banding karyotype analysis and fluorescence in situ hybridization (FISH) revealed a marker derived from chromosome 6 and a neocentromere as suspected, but array-CGH enabled us to characterize this marker as a der(10)t(6;10)(6qter → 6q23.3::10p11.1 → 10p11.21)dn. As far as we know, this is the first diagnosed case of PWLS associated with a complex sSMC, involving a 30.9 Mb gain in the 6q16.3q23.3 region and a 3.5 Mb gain in the 10p11.21p11.1 region. Several genes have been mapped to the 6q region including the TCBA1 gene, which is associated with developmental delay and recurrent infections, the ENPP1 gene, associated with insulin resistance and susceptibility to obesity and the BMIQ3 gene, associated with body mass index (BMI). No OMIM gene was found in the smallest 10p11.21p11.1 region. We suggest that the duplicated chromosome segment 6q16.3q23.3 may be responsible for the phenotype of our case and may also be a candidate locus of PWLS.

  12. Review on transactinium isotope build-up and decay in reactor fuel and related sensitivities to cross section changes and results and main conclusions of the IAEA-Advisory Group Meeting on Transactinium Nuclear Data, held at Karlsruhe, November 1975

    International Nuclear Information System (INIS)

    Kuesters, H.; Lalovic, M.

    1976-04-01

    In this report a review is given on the actinium isotope build-up and decay in LWRs, LMFBRs and HTRs. The dependence of the corresponding physical aspects on reactor type, fuel cycle strategy, calculational methods and cross section uncertainties is discussed. Results from postirradiation analyses and from integral experiments in fast zero power assemblies are compared with theoretical predictions. Some sensitivity studies about the influence of actinium nuclear data uncertainties on the isotopic concentration, decay heat, and the radiation out-put in fuel and waste are presented. In a second part, the main results of the IAEA-Advisory Group Meeting on Transactinium Nuclear Data are summarized and discussed. (orig.) [de

  13. Coulomb effects in isobaric cold fission from reactions 233U(nth,f), 235U(nth,f),239Pu(nth,f) and 252Cf(sf)

    International Nuclear Information System (INIS)

    Montoya, Modesto

    2013-01-01

    The Coulomb effect hypothesis, formerly used to interpret fluctuations in the curve of maximal total kinetic energy as a function of light fragment mass in reactions 233 U(n th ,f), 235 U(n th ,f) and 239 Pu(n th ,f), is confirmed in high kinetic energy as well as in low excitation energy windows, respectively. Data from reactions 233 U(n th ,f), 235 U(n th ,f), 239 Pu(n th ,f) and 252 Cf(sf) show that, between two isobaric fragmentations with similar Q-values, the more asymmetric charge split reaches the higher value of total kinetic energy. Moreover, in isobaric charge splits with different Q-values, similar preference for asymmetrical fragmentations is observed in low excitation energy windows. (author).

  14. Absolute M1 and E2 Transition Probabilities in 233U

    International Nuclear Information System (INIS)

    Malmskog, S.G.; Hoejeberg, M.

    1967-08-01

    Using the delayed coincidence technique, the following half lives have been determined for different excited states in 233 U: T 1/2 (311.9 keV level) = (1.20 ± 0.15) x 10 -10 sec, T 1/2 (340.5 keV level) = (5.2 ± 1.0) x 10 -11 sec, T 1/2 (398.6 keV level) = (5.5 ± 2.0) x 10 -11 sec and T 1/2 (415.8 keV level) -11 sec. From these half life determinations, together with earlier known electron intensities and conversion coefficients, 22 reduced B(Ml) and B(E2) transition probabilities (including 9 limits) have been deduced. The rotational transitions give information on the parameters δ and (g K - g R ) . The experimental M1 and E2 transition rates between members of different bands have been analysed in terms of the predictions of the Nilsson model, taking also pairing correlations and Coriolis coupling effects into account

  15. Site-Specific Health and Safety Plan, 233-S Decontamination and Decommissioning

    International Nuclear Information System (INIS)

    Hobbs, B.J.

    1998-01-01

    The 233-S Facility operated from January 1952 until July 1967, at which time the building entered the U.S. Department of Energy's Surplus Facility Management Program as a retired facility. The facility has since undergone severe degradation due to exposure to extreme weather conditions. A freeze and thaw cycle occurred at the Hanford Site during February 1996, which caused cracking failure of portions of the building roof. This resulted in significant infiltration of water into the facility, which creates a pathway for potential release of radioactive material into the environment (air and/or ground). Additionally, the weather caused existing cracks in concrete structures of the building to lengthen, thereby increasing the potential for failed confinement of the building's radioactive material. Differential settlement has also occurred, causing portions of the facility to separate from the main building structure, increasing the potential for release of radioactive material to the environment. An expedited response is proposed to remove this threat and ensure protection of human health and the environment

  16. Defense In-Depth Accident Analysis Evaluation of Tritium Facility Bldgs. 232-H, 233-H, and 234-H

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'The primary purpose of this report is to document a Defense-in-Depth (DID) accident analysis evaluation for Department of Energy (DOE) Savannah River Site (SRS) Tritium Facility Buildings 232-H, 233-H, and 234-H. The purpose of a DID evaluation is to provide a more realistic view of facility radiological risks to the offsite public than the bounding deterministic analysis documented in the Safety Analysis Report, which credits only Safety Class items in the offsite dose evaluation.'

  17. Aerosols generated by 239PU and 233U droplets burning in air

    International Nuclear Information System (INIS)

    Nelson, L.S.; Raabe, O.G.

    1978-01-01

    The inhalation hazards of radioactive aerosols produced by the explosive disruption and subsequent combustion of metallic plutonium in air are not adequately understood. Results of a study to determine whether uranium can be substituted for plutonium in such a situation in which experiments were performed under identical conditions with laser-ignited, single, freely falling droplets of 239 Pu and 233 U are reported. The total amounts of aerosol produced were studied quantitatively as a function of time during the combustion. Also, particle size distributions of selected aerosols were studied with aerodynamic particle separation techniques. Results showed that the ultimate quantity of aerosols, their final particle size distributions, and depositions as a function of time are not identical mainly because of the different vapor pressures of the metals, and the unlike degrees of violence of the explosions of the droplets

  18. Defense In-Depth Accident Analysis Evaluation of Tritium Facility Bldgs. 232-H, 233-H, and 234-H

    Energy Technology Data Exchange (ETDEWEB)

    Blanchard, A.

    1999-05-10

    'The primary purpose of this report is to document a Defense-in-Depth (DID) accident analysis evaluation for Department of Energy (DOE) Savannah River Site (SRS) Tritium Facility Buildings 232-H, 233-H, and 234-H. The purpose of a DID evaluation is to provide a more realistic view of facility radiological risks to the offsite public than the bounding deterministic analysis documented in the Safety Analysis Report, which credits only Safety Class items in the offsite dose evaluation.'

  19. Mass dependence of azimuthal asymmetry in the fission of {sup 232}Th and {sup 233,235,236,238}U by polarized photons

    Energy Technology Data Exchange (ETDEWEB)

    Denyak, V.V. [National Science Center ' ' Kharkov Institute of Physics and Technology' ' , Kharkiv (Ukraine); Pele Pequeno Principe Research Institute, Curitiba (Brazil); Khvastunov, V.M. [National Science Center ' ' Kharkov Institute of Physics and Technology' ' , Kharkiv (Ukraine); Paschuk, S.A. [Federal University of Technology - Parana, Curitiba (Brazil); Schelin, H.R. [Federal University of Technology - Parana, Curitiba (Brazil); Pele Pequeno Principe Research Institute, Curitiba (Brazil)

    2013-04-15

    Fission of the even-even nuclei {sup 232}Th, {sup 236,238}U and even-odd nuclei {sup 233,235}U by linearly polarized photons has been studied at excitation energies in the region of a giant dipole resonance. The performed investigations unambiguously showed the existence of the fragment mass dependence of the cross section azimuthal asymmetry in the photofission of {sup 236}U and {sup 238}U. In addition, the obtained results provided the first evidence for the possible difference between the asymmetry values in asymmetric and symmetric mass distribution regions in the case of {sup 236}U. The measured cross section azimuthal asymmetry of the fission of {sup 232}Th does not show any fragment mass dependence. In the even-odd nuclei {sup 233}U and {sup 235}U the difference between the far-asymmetric and other mass distribution regions was also observed but with the statistical uncertainty not small enough for definitive conclusion. (orig.)

  20. 233U breeding in accelerator-driven sub-critical fast reactor

    International Nuclear Information System (INIS)

    Yang Yongwei; An Yu

    1999-01-01

    Accelerator-driven Sub-critical Fast Reactor (ADFR) is chosen as fissile-material-breeding reactor. (U-Pu)O x is chosen as fuel in the core and ThO 2 as fertile material in the blanket zone to breed 233 U. Molten lead is chosen as coolant because of its better neutronic and chemical characteristics over sodium. The program system used for neutronics study consists of: LAHET, for the simulation of the interaction between the proton with medium energy and the nuclei of the target; MCNP4A, for the simulation of neutron transport with energy below 20 MeV in the sub-critical reactor; CONNECT1, for the processing of some tallies provided by the output of MCNP4A in order to prepare micro-cross sections for elements used for burnup calculation; ORIGEN2, used for multi-region burnup calculation; CONNECT2, for the processing of atom densities of some elements provided in the output of ORIGEN2 in order to prepare input to LAHET calculation for next time step. The calculated results show that the proposed case is feasible for breeding fissile material considering the criticality safety, power density, burnup, etc

  1. Chemical and spectrochemical production analysis of ThO2 and 233UO2-ThO2 pellets for the light water breeder reactor core for Shippingport (LWBR development program)

    International Nuclear Information System (INIS)

    Bukowski, J.F.; Hollis, E.D.

    1975-06-01

    The Bettis Atomic Power Laboratory has utilized wet chemical, emission spectrochemical, and mass spectrometric analytical techniques for the production analysis of the ThO 2 and 233 UO 2 -ThO 2 (1 to 6 wt percent 233 UO 2 ) pellets for the Light Water Breeder Reactor (LWBR) core for Shippingport. Proof of the fuel breeding concept necessitates measurement of precise and accurate chemical characterization of all fuel pellets before core life. Chemistry's efforts toward this goal are presented in three main sections: (1) general discussions relating the chemical requirements for ThO 2 and 233 UO 2 -ThO 2 core materials to the analytical capabilities, (2) technical discussions of the chemical and instrumental technology applied for the analysis of aluminum, boron, calcium, carbon, chloride plus bromide, chromium, cobalt, copper, dysprosium, europium, fluoride, gadolinium, iron, magnesium, manganese, mercury, molybdenum, nickel, nitrogen, samarium, silicon, titanium, vanadium, thorium, and uranium (total, trace, and uranium VI), and (3) a formal presentation of the analytical procedures as applied to the LWBR Development Program. (U.S.)

  2. Neutron multipilication factors as a function of temperature: a comparison of calculated and measured values for lattices using 233UO2-ThO2 fuel in graphite

    International Nuclear Information System (INIS)

    Newman, D.F.; Gore, B.F.

    1978-01-01

    Neutron multiplication factors calculated as a function of temperature for three graphite-moderated 233 UO 2 -ThO 2 -fueled lattices are correlated with the values measured for these lattices in the high-temperature lattice test reactor (HTLTR). The correlation analysis is accomplished by fitting calculated values of k/sub infinity/(T) to the measured values using two least-squares-fitted correlation coefficients: (a) a normalization factor and (b) a temperature coefficient bias factor. These correlations indicate the existence of a negative (nonconservative) bias in temperature coefficients of reactivity calculated using ENDF/B-IV cross-section data. Use of an alternate cross-section data set for thorium, which has a smaller resonance integral than ENDF/B-IV data, improved the agreement between calculated and measured temperature coefficients of reactivity for the three experimental lattices. The results of the correlations are used to estimate the bias in the temperature coefficient of reactivity calculated for a lattice typical of fresh 233 U recycle fuel for a high-temperature gas-cooled reactor (HTGR). This extrapolation to a lattice having a heavier fissile loading than the experimental lattices is accomplished using a sensitivity analysis of the estimated bias to alternate thorium cross-section data used in calculations of k/sub infinity/(T). The envelope of uncertainty expected to contain the actual values for the temperature coefficient of the reactivity for the 233 U-fueled HTGR lattice studied remains negative at 1600 K (1327 0 C). Although a broader base of experimental data with improved accuracy is always desirable, the existing data base provided by the HTLTR experiments is judged to be adequate for the verification of neutronic calculations for the HTGR containing 233 U fuel at its current state of development

  3. Fission Product Yields of 233U, 235U, 238U and 239Pu in Fields of Thermal Neutrons, Fission Neutrons and 14.7-MeV Neutrons

    Science.gov (United States)

    Laurec, J.; Adam, A.; de Bruyne, T.; Bauge, E.; Granier, T.; Aupiais, J.; Bersillon, O.; Le Petit, G.; Authier, N.; Casoli, P.

    2010-12-01

    The yields of more than fifteen fission products have been carefully measured using radiochemical techniques, for 235U(n,f), 239Pu(n,f) in a thermal spectrum, for 233U(n,f), 235U(n,f), and 239Pu(n,f) reactions in a fission neutron spectrum, and for 233U(n,f), 235U(n,f), 238U(n,f), and 239Pu(n,f) for 14.7 MeV monoenergetic neutrons. Irradiations were performed at the EL3 reactor, at the Caliban and Prospero critical assemblies, and at the Lancelot electrostatic accelerator in CEA-Valduc. Fissions were counted in thin deposits using fission ionization chambers. The number of fission products of each species were measured by gamma spectrometry of co-located thick deposits.

  4. Advantages and implications of U233 fueled thermionic space power energy conversion

    International Nuclear Information System (INIS)

    Terrell, C.W.

    1992-01-01

    In this paper two recent analyses are reported which demonstrate advantages of a U233 fueled thermionic fuel element (TFE) compared to 93 w/o U235, and that application (mission) has broad latitude in how space power reactor systems could or should be optimized. A reference thermionic reactor system was selected to provide the basis for the fuel comparisons. Both oxide and metal fuel forms were compared. Of special interest was to estimate the efficiencies of the four fuel forms to produce electrical power. A figure of merit (FOM) was defined which is directly proportional to the electrical average electrical power produced is proportional to the electrical power produced per unit uranium mass. In a TFE the average electrical power produced is proportional to the emitter surface area (Esa), hence the ratio Esa/Mu was selected as the FOM. Results indicate that the choice of fuel type and form leads to wide variations in critical and system masses FOM values, and system total power

  5. Investigation of the fission yields of the fast neutron-induced fission of {sup 233}U; Mesure de la distribution en masse et en charge des produits de la fission rapide de l'{sup 233}U

    Energy Technology Data Exchange (ETDEWEB)

    Galy, J

    1999-09-01

    As a stars, a survey of the different methods of investigations of the fission product yields and the experimental data status have been studied, showing advantages and shortcomings for the different approaches. An overview of the existing models for the fission product distributions has been as well intended. The main part of this thesis was the measurement of the independent yields of the fast neutron-induced fission of{sup 233}U, never investigated before this work. The experiment has been carried out using the mass separator OSIRIS (Isotope Separator On-Line). Its integrated ion-source and its specific properties required an analysis of the delay-parameter and ionisation efficiency for each chemical species. On the other hand, this technique allows measurement of independent yields and cumulative yields for elements from Cu to Ba, covering most of the fission yield distribution. Thus, we measured about 180 independent yields from Zn (Z=30) to Sr (Z=38) in the mass range A=74-99 and from Pd (Z=46) to Ba (Z=56) in the mass range A=113-147, including many isomeric states. An additional experiment using direct {gamma}-spectroscopy of aggregates of fission products was used to determine more than 50 cumulative yields of element with half-life from 15 min to a several days. All experimental data have been compared to estimates from a semi-empirical model, to calculated values and to evaluated values from the European library JEF 2.2. Furthermore, a study of both thermal and fast neutron-induced fission of {sup 233}U measured at Studsvik, the comparison of the OSIRIS and LOHENGRIN facilities and the trends in new data for the Reactors Physics have been discussed. (author)

  6. Termination of Safeguards on ULWBR Material

    International Nuclear Information System (INIS)

    Ivan R. Thomas; Ernest L. Laible

    2008-01-01

    The Department of Energy (DOE), Office of Environmental Management, has approved the disposition of 31 metric tons of Unirradiated Light Water Breeder Reactor (ULWBR) material in canisters stored within dry wells of the Underground Fuel Storage Facility at the Idaho Nuclear Technology and Engineering Center (INTEC). This unirradiated material consists primarily of ceramic pellets of thorium oxide in stainless steel cladding, but it also contains 300 kilograms of uranium that is 98 wt% U-233. The ULWBR material was not processed at the INTEC because it was incompatible with prior chemical separation schemes. Other economical recovery options have not been identified, and expressions of interest for consolidating the material with existing projects at other DOE sites have not been received. The U-233 could be used for producing the medical isotope Actinium-225, but the proof-of-principle demonstration and follow-on pilot program have not been developed to the point of requiring production quantities of U-233. Consequently, the selected disposition of the ULWBR material was burial as Low Level Waste at the Nevada Test Site (NTS), which required terminating safeguards controls for the contained Category II quantity of Attractiveness Level D special nuclear material (SNM). The requested termination followed the twelve point evaluation criteria of the Historical Defense Program Discard Guidance and included a security analysis for evaluating the risks of theft, diversion, and radiological sabotage associated with the material. Continuity of knowledge in the book inventory was assured by documenting that the original shipper's measurements accurately reflected the quantities of materials received and that the ULWBR materials had remained under adequate physical protection and had been subject to periodic physical inventories. The method selected for substantiating the book values as the basis for terminating safeguards was the nondestructive assay used during physical

  7. Monte Carlo analyses of simple U233 O2-ThO2 and U235 O2-ThO2 lattices with ENDF/B-IV data (AWBA development program)

    International Nuclear Information System (INIS)

    Hardy, J. Jr.; Ullo, J.J.

    1980-09-01

    A number of water-moderated Th-U235 and Th-U233 lattice integral experiments were analyzed in a consistent manner, with ENDF/B-IV data and detailed Monte Carlo methods. These experiments provide a consistent test of the nuclear data. The ENDF/B-IV data are found to perform reasonably well. Adequate agreement is found with integral measurements of thorium capture. Calculated K/sub eff/ values show a generally coherent pattern which is consistent with K/sub eff/ results obtained for homogeneous aqueous critical assemblies. Harder prompt fission spectra for U233 and U235 can correct the principal discrepancy observed with ENDF/B-IV, a bias trend in K/sub eff/ attributed to an underprediction of leakage

  8. Behaviour of uranium series radionuclides in surface water (Crouzille, Limousin). Geochemical implications

    International Nuclear Information System (INIS)

    Moulin, J.

    2008-06-01

    Understanding natural radionuclides behaviour in surface water is a required step to achieve uranium mine rehabilitation and preserve water quality. The first objective of this thesis is to determine which are the radionuclides sources in a drinking water reservoir. The second objective is to improve the knowledge about the behaviour of uranium series radionuclides, especially actinium. The investigated site is a brook (Sagnes, Limousin, France) which floods a peat bog contaminated by a former uranium mine and which empties into the Crouzille lake. It allows studying radionuclides transport in surface water and radionuclides retention through organic substance or water reservoir. Radionuclides distribution in particulate, colloidal and dissolved phases is determined thanks to ultra-filtrations. Gamma spectrometry allows measuring almost all natural radionuclides with only two counting stages. However, low activities of 235 U series radionuclides impose the use of very low background well-type Ge detectors, such as those of the Underground Laboratory of Modane (France). Firstly, this study shows that no or few radionuclides are released by the Sagnes peat bog, although its radioactivity is important. Secondly, it provides details on the behaviour of uranium series radionuclides in surface water. More specifically, it provides the first indications of actinium solubility in surface water. Actinium's behaviour is very close to uranium's even if it is a little less soluble. (author)

  9. Status of thorium cycle nuclear data evaluations: Comparison of cross-section line shapes of JENDL-3 and ENDF-B-VI files for 230Th, 232Th, 231Pa, 233Pa, 232U, 233U and 234U

    International Nuclear Information System (INIS)

    Ganesan, S.; McLaughlin, P.K.

    1992-02-01

    Since 1990, one of the most interesting developments in the field of nuclear data for nuclear technology applications is that several new evaluated data files have been finalized and made available to the International Atomic Energy Agency (IAEA) for distribution to its Member States. Improved evaluated nuclear data libraries such as ENDF/B-VI from the United States and JENDL-3 from Japan were developed over a period of 10-15 years. This report is not an evaluation of the evaluations. The report as presented here gives a first look at the cross section line shapes of the isotopes that are important to the thorium fuel cycle derived from the two recently evaluated data files: JENDL-3 and ENDF/B-VI. The basic evaluated data files JENDL-3 and ENDF/B-VI were point-processed successfully using the codes LINEAR and RECENT. The point data were multigrouped in three different group structures using the GROUPIE code. Graphs of intercomparisons of cross section line shapes of JENDL-3 and ENDF/B-VI are presented in this paper for the following isotopes of major interest to studies of the thorium fuel cycle: 230 Th, 232 Th, 231 Pa, 233 Pa, 232 U, 233 U and 234 U. Comparisons between JENDL-3 and ENDF/B-VI which were performed at the point and group levels show large discrepancies in various cross sections. We conclude this report with a general remark that it is necessary to perform sensitivity studies to assess the impacts of the discrepancies between the two different sets of data on calculated reactor design and safety parameters of specific reactor systems and, based on the results of such sensitivity studies, to undertake new tasks of evaluations. (author). 2 refs, 245 figs, 8 tabs

  10. Study of the origin of elements of the uranium-235 family observed in excess in the vicinity of the experimental nuclear EL4 reactor under dismantling. Lessons got at this day and conclusions

    International Nuclear Information System (INIS)

    2007-01-01

    This study resumes the discovery of an excess of actinium 227 found around by EL4 nuclear reactor actually in dismantling. The search for the origin of this excess revealed a real inquiry of investigation during three years. Because a nuclear reactor existed in this area a particular attention will have concerned this region. The doubt became the line of conduct to find the answer to the human or natural origin of this excess. Finally and against any evidence, it appears that the origin of this phenomenon was natural, consequence of the particular local geology. The detail of the different investigations is given: search of a possible correlation with the composition of elevations constituent of lanes, search (and underlining) of new sites in the surroundings of the Rusquec pond and the Plouenez station, study of the atmospheric deposits under winds of the nuclear power plant and in the east direction, search of a possible relationship with the gaseous effluents of the nuclear power plant in the past, historical study of radioactive effluents releases in the fifty last years by the analysis of the sedimentary deposits in the Saint-Herbiot reservoir, search of a possible correlation between the excess of actinium 227 and the nuclear power plant activity; search of a possible correlation with a human activity without any relationship with the nuclear activities, search of a correlation with the underground waters, search of a correlation with the geological context, collect of information on the possible transfers in direction of the food chain, determination of the radiological composition of the underground waters ( not perturbed by human activity), search of the cause of an excess of actinium 227 in the old channel of liquid effluents release of the nuclear power plant. The results are given and discussed. And contrary to all expectations the origin of the excess of actinium 227 is completely natural. (N.C.)

  11. Study of the origin of elements of the uranium-235 family observed in excess in the vicinity of the experimental nuclear EL4 reactor under dismantling. Lessons got at this day and conclusions; Etude de l'origine des elements de la famille de l'uranium-235 observes en exces dans les environs du reacteur nucleaire experimental EL4 en cours de demantelement. Enseignements retires a ce jour et conclusion

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    This study resumes the discovery of an excess of actinium 227 found around by EL4 nuclear reactor actually in dismantling. The search for the origin of this excess revealed a real inquiry of investigation during three years. Because a nuclear reactor existed in this area a particular attention will have concerned this region. The doubt became the line of conduct to find the answer to the human or natural origin of this excess. Finally and against any evidence, it appears that the origin of this phenomenon was natural, consequence of the particular local geology. The detail of the different investigations is given: search of a possible correlation with the composition of elevations constituent of lanes, search (and underlining) of new sites in the surroundings of the Rusquec pond and the Plouenez station, study of the atmospheric deposits under winds of the nuclear power plant and in the east direction, search of a possible relationship with the gaseous effluents of the nuclear power plant in the past, historical study of radioactive effluents releases in the fifty last years by the analysis of the sedimentary deposits in the Saint-Herbiot reservoir, search of a possible correlation between the excess of actinium 227 and the nuclear power plant activity; search of a possible correlation with a human activity without any relationship with the nuclear activities, search of a correlation with the underground waters, search of a correlation with the geological context, collect of information on the possible transfers in direction of the food chain, determination of the radiological composition of the underground waters ( not perturbed by human activity), search of the cause of an excess of actinium 227 in the old channel of liquid effluents release of the nuclear power plant. The results are given and discussed. And contrary to all expectations the origin of the excess of actinium 227 is completely natural. (N.C.)

  12. Sorption studies of radioelements on geological materials

    International Nuclear Information System (INIS)

    Berry, John A.; Yui, Mikazu; Kitamura, Akira

    2007-11-01

    Batch sorption experiments have been carried out to study the sorption of uranium, technetium, curium, neptunium, actinium, protactinium, polonium, americium and plutonium onto bentonite, granodiorite and tuff. Mathematical modelling using the HARPHRQ program and the HATCHES database was carried out to predict the speciation of uranium and technetium in the equilibrated seawater, and neptunium, americium and plutonium in the rock equilibrated water. Review of the literature for thermodynamic data for curium, actinium, protactinium and polonium was carried out. Where sufficient data were available, predictions of the speciation and solubility were made. This report is a summary report of the experimental work conducted by AEA Technology during April 1991-March 1998, and the main results have been presented at Material Research Society Symposium Proceedings and published as proceedings of them. (author)

  13. Final characterization report for the non-process areas of the 233-S Plutonium Concentration Facility

    International Nuclear Information System (INIS)

    Encke, D.B.; Harris, R.A.

    1997-04-01

    This report addresses the 233-S Plutonium Concentration Facility characterization survey data collected from January 21, 1997 through February 3, 1997. The characterization activities evaluated the radiological status and identified the hazardous materials locations. The scope of this report is limited to the nonprocess areas in the facility, which include the special work permit (SWP) change room, toilet, equipment room, electrical cubicle, control room, and pipe gallery. A portion of the roof (excluding the roof over the process hood and viewing room) was also included. Information in this report will be used to identify waste streams, provide specific chemical and radiological data to aid in planning decontamination and demolition activities, and allow proper disposal of the demolition debris, as required by the Comprehensive Environmental Response, Compensation, and Liability Act of 1980

  14. Fc Gamma Receptor 3B (FCGR3Bc.233C>A-rs5030738) Polymorphism Modifies the Protective Effect of Malaria Specific Antibodies in Ghanaian Children

    DEFF Research Database (Denmark)

    Adu, Bright; Jepsen, Micha Phill Grønholm; Gerds, Thomas A

    2014-01-01

    Immunoglobulin G (IgG) cross-linking with Fc gamma receptor IIIB (FcγRIIIB) triggers neutrophil degranulation, releasing reactive oxygen species with high levels associated with protection against malaria. The FCGR3B-c.233C>A polymorphism thought to influence the interaction between IgG and Fcγ...

  15. Importance of coccolithophore-associated organic biopolymers for fractionating particle-reactive radionuclides (234Th, 233Pa, 210Pb, 210Po, and 7Be) in the ocean

    Science.gov (United States)

    Lin, Peng; Xu, Chen; Zhang, Saijin; Sun, Luni; Schwehr, Kathleen A.; Bretherton, Laura; Quigg, Antonietta; Santschi, Peter H.

    2017-08-01

    Laboratory incubation experiments using the coccolithophore Emiliania huxleyi were conducted in the presence of 234Th, 233Pa, 210Pb, 210Po, and 7Be to differentiate radionuclide uptake to the CaCO3 coccosphere from coccolithophore-associated biopolymers. The coccosphere (biogenic calcite exterior and its associated biopolymers), extracellular (nonattached and attached exopolymeric substances), and intracellular (sodium-dodecyl-sulfate extractable and Fe-Mn-associated metabolites) fractions were obtained by sequentially extraction after E. huxleyi reached its stationary growth phase. Radionuclide partitioning and the composition of different organic compound classes, including proteins, total carbohydrates (TCHO), and uronic acids (URA), were assessed. 210Po was closely associated with the more hydrophobic biopolymers (high protein/TCHO ratio, e.g., in attached exopolymeric substances), while 234Th and 233Pa showed similar partitioning behavior with most activity being distributed in URA-enriched, nonattached exopolymeric substances and intracellular biopolymers. 234Th and 233Pa were nearly undetectable in the coccosphere, with a minor abundance of organic components in the associated biopolymers. These findings provide solid evidence that biogenic calcite is not the actual main carrier phase for Th and Pa isotopes in the ocean. In contrast, both 210Pb and 7Be were found to be mostly concentrated in the CaCO3 coccosphere, likely substituting for Ca2+ during coccolith formation. Our results demonstrate that even small cells (E. huxleyi) can play an important role in the scavenging and fractionation of radionuclides. Furthermore, the distinct partitioning behavior of radionuclides in diatoms (previous studies) and coccolithophores (present study) explains the difference in the scavenging of radionuclides between diatom- and coccolithophore-dominated marine environments.

  16. Synthesis of 8-phenyl-10H-pyrido[1,2-α]indole salts from 2,3,3-trimethyl-3H-indole chlorides with cinnamaldehyde

    International Nuclear Information System (INIS)

    Shachkus, A.A.; Degutis, Yu.A.

    1987-01-01

    Reaction of 2,3,3-trimethyl-3H-indole chloride with cinnamic and 4-dimethylaminocinnamic aldehydes led to salts of 8-phenyl and 8-(4-dimethylaminophenyl)-10,10-dimethyl-10H-pyrido[1,2-α]indole. PMR spectra were recorded on a Tesla BS-487C (80 MHz) instrument (internal standard HMDS) and IR spectra on a UR-20 spectrometer (KBr pellets)

  17. Objectieve en subjectieve verkeersveiligheid van het N233-kruispunt Rhenen-Achterberg : inventarisatie van zorgpunten bij bewoners, enquête onder (ouders van) scholieren en beoordeling van de huidige en toekomstige verkeerssituatie. Onderzoek in opdracht van de Provincie Utrecht.

    OpenAIRE

    Bax, C.A. Hoekstra, A.T.G. & Schermers, G.

    2017-01-01

    Objective and subjective road safety of the Rhenen-Achterberg intersection on the N233 provincial road : Inventory of concerns among residents, survey among (parents of) students and assessment of the current and future traffic situation. The province of Utrecht asked SWOV to investigate the objective and subjective traffic safety at the intersection of the Bergweg/Achterbergsestraatweg and provincial road N233. It is an intersection between a 80 km/h rural distributor road under the authorit...

  18. Material control and accountability aspects of safeguards for the USA 233U/Th fuel recycle plant

    International Nuclear Information System (INIS)

    Carpenter, J.A. Jr.; McNeany, S.R.; Angelini, P.; Holder, N.D.; Abraham, L.

    1978-01-01

    The materials control and accountability aspects of the reprocessing and refabrication of a conceptual large-scale HTGR fuel recycle plant have been discussed. Two fuel cycles were considered. The traditional highly enriched uranium cycle uses an initial or makeup fuel element with a fissile enrichment of 93% 235 U. The more recent medium enriched uranium cycle uses initial or makeup fuel elements with a fissile enrichment less than 20% 235 U. In both cases, 233 U bred from the fertile thorium is recycled. Materials control and accountability in the plant will be by means of a real-time accountability method. Accountability data will be derived from monitoring of total material mass through the processes and a system of numerous assays, both destructive and nondestructive

  19. A 233U/236U/242Pu/244Pu spike for isotopic and isotope dilution analysis by mass spectrometry with internal calibration

    International Nuclear Information System (INIS)

    Stepanov, A.; Belyaev, B.; Buljanitsa, L.

    1989-11-01

    The Khlopin Radium Institute prepared on behalf of the IAEA a synthetic mixture of 233 U, 236 U, 242 Pu and 244 Pu isotopes. The isotopic composition and elemental concentration of uranium and plutonium were certified on the basis of analyses done by four laboratories of the IAEA Network, using mass spectrometry with internal standardization. The certified values for 233 U/ 236 U ratio and the 236 U chemical concentration have a coefficient of variation of 0.05%. The latter is fixed by the uncertainty in the 235 U/ 238 U ratio of NBS500 used as internal standard. The coefficients of variation of the 244 Pu/ 242 Pu ratio and the 242 Pu chemical concentration are respectively 0.10% and 0.16% and limited by the uncertainty in the 240 Pu/ 239 Pu ratio of NBS947. This four isotope mixture was used as an internal standard as well as a spike, to analyze 30 batches of LWR spent fuel solutions. The repeatability of the mass spectrometric measurements have a coefficient of variation of 0.025% for the uranium concentration, and of 0.039% for the plutonium concentration. The spiking and treatment errors had a coefficient of variation of 0.048%. (author). Refs, figs and tabs

  20. Measurements of neutron induced capture and fission reactions on $^{233}$ U (EAR1)

    CERN Multimedia

    The $^{233}$U plays the essential role of ssile nucleus in the Th-U fuel cycle, which has been proposed as a safer and cleaner alternative to the U-Pu fuel cycle. Considered the scarce data available to assess the capture cross section, a measurement was proposed and successfully performed at the n_TOF facility at CERN using the 4$\\pi$ Total Absorp- tion Calorimeter (TAC). The measurement was extremely dicult due to the need to accurately distinguish between capture and fission $\\gamma$-rays without any additional discrim-ination tool and the measured capture cross section showed a signicant disagreement in magnitude when compared with the ENDF/B-VII.1 library despite the agreement in shape. We propose a new measurement that is aimed at providing a higher level of dis-crimination between competing nuclear reactions, to extend the neutron energy range and to obtain more precise and accurate data, thus fullling the demands of the "NEA High Priority Nuclear Data Request List". The setup is envisaged as a combin...

  1. De novo 12q22.q23.3 duplication associated with temporal lobe epilepsy.

    Science.gov (United States)

    Vari, Maria Stella; Traverso, Monica; Bellini, Tommaso; Madia, Francesca; Pinto, Francesca; Minetti, Carlo; Striano, Pasquale; Zara, Federico

    2017-08-01

    Temporal lobe epilepsy (TLE) is the most common form of focal epilepsy and may be associated with acquired central nervous system lesions or could be genetic. Various susceptibility genes and environmental factors are believed to be involved in the aetiology of TLE, which is considered to be a heterogeneous, polygenic, and complex disorder. Rare point mutations in LGI1, DEPDC5, and RELN as well as some copy number variations (CNVs) have been reported in families with TLE patients. We perform a genetic analysis by Array-CGH in a patient with dysmorphic features and temporal lobe epilepsy. We report a de novo duplication of the long arm of chromosome 12. We confirm that 12q22-q23.3 is a candidate locus for familial temporal lobe epilepsy with febrile seizures and highlight the role of chromosomal rearrangements in patients with epilepsy and intellectual disability. Copyright © 2017 British Epilepsy Association. Published by Elsevier Ltd. All rights reserved.

  2. Association of sequence variants on chromosomes 20, 11, and 5 (20q13.33, 11q23.3, and 5p15.33) with glioma susceptibility in a Chinese population.

    Science.gov (United States)

    Chen, Hongyan; Chen, Yuanyuan; Zhao, Yao; Fan, Weiwei; Zhou, Keke; Liu, Yanhong; Zhou, Liangfu; Mao, Ying; Wei, Qingyi; Xu, Jianfeng; Lu, Daru

    2011-04-15

    Two genome-wide association studies of glioma in European populations identified 14 genetic variants strongly associated with risk of glioma, but it is unknown whether these variants are associated with glioma risk in Asian populations. The authors genotyped these 14 variants in 976 glioma patients and 1,057 control subjects to evaluate their associations with risk of glioma, particularly high-grade glioma (glioblastoma; n = 312), in a Chinese population (2004-2009). Overall, the authors identified 3 susceptibility loci for glioma risk at 20q13.33 (RTEL1 rs6010620 (P = 2.79 × 10(-6))), 11q23.3 (PHLDB1 rs498872 (P = 3.8 × 10(-6))), and 5p15.33 (TERT rs2736100 (P = 3.69 × 10(-4))) in this study population; these loci were also associated with glioblastoma risk (20q13.33: RTEL1 rs6010620 (P = 3.57 × 10(-7)); 11q23.3: PHLDB1 rs498872 (P = 7.24 × 10(-3)); 5p15.33: TERT rs2736100 and TERT rs2736098 (P = 1.21 × 10(-4) and P = 2.84 × 10(-4), respectively)). This study provides further evidence for 3 glioma susceptibility regions at 20q13.33, 11q23.3, and 5p15.33 in Chinese populations.

  3. Amorphous silicon passivation for 23.3% laser processed back contact solar cells

    Science.gov (United States)

    Carstens, Kai; Dahlinger, Morris; Hoffmann, Erik; Zapf-Gottwick, Renate; Werner, Jürgen H.

    2017-08-01

    This paper presents amorphous silicon deposited at temperatures below 200 °C, leading to an excellent passivation layer for boron doped emitter and phosphorus doped back surface field areas in interdigitated back contact solar cells. A higher deposition temperature degrades the passivation of the boron emitter by an increased hydrogen effusion due to lower silicon hydrogen bond energy, proved by hydrogen effusion measurements. The high boron surface doping in crystalline silicon causes a band bending in the amorphous silicon. Under these conditions, at the interface, the intentionally undoped amorphous silicon becomes p-type conducting, with the consequence of an increased dangling bond defect density. For bulk amorphous silicon this effect is described by the defect pool model. We demonstrate, that the defect pool model is also applicable to the interface between amorphous and crystalline silicon. Our simulation shows the shift of the Fermi energy towards the valence band edge to be more pronounced for high temperature deposited amorphous silicon having a small bandgap. Application of optimized amorphous silicon as passivation layer for the boron doped emitter and phosphorus doped back surface field on the rear side of laser processed back contact solar cells, fabricated using four laser processing steps, yields an efficiency of 23.3%.

  4. Collective and single-particle excitations in the heavy deformable nuclei 234U, 233U, 231Th, 230Pa and 232Pa

    International Nuclear Information System (INIS)

    Kotthaus, Tanja

    2010-01-01

    In this thesis five heavy deformed isotopes from the mass region A≥230, namely 234 U, 233 U, 231 Th, 230 Pa and 232 Pa, were investigated by means of deuteron-induced neutron transfer reactions. The even-even isotope 234 U has been studied with the 4π-γ-spectrometer MINIBALL at the Cologne Tandem accelerator. Excited nuclei in the isotope 234 U were produced using the reaction 235 U(d,t) at a beam energy of 11 MeV. The target thickness was 3.5 mg/cm 2 . The analysis of the γγ-coincidence data yielded a reinterpretation of the level scheme in 12 cases. Considering its decay characteristics, the 4 + state at an excitation energy of 1886.7 keV is a potential candidate for a two-phonon vibrational state. The isotopes 233 U, 231 Th, 230 Pa and 232 Pa were investigated at the Munich Q3D spectrometer. For each isotope an angular distribution with angles between 5 and 45 were measured. In all four cases the energy of the polarized deuteron beam (vector polarization of 80%) was 22 MeV. As targets 234 U (160 μg/cm 2 ), 230 Th (140 μg/cm 2 ) and 231 Pa (140 μg/cm 2 ) were used. The experimental angular distributions were compared to results of DWBA calculations. For the odd isotope 233 U spin and parity for 33 states are assigned and in the other odd isotope 231 Th 22 assignments are made. The excitation spectra of the two odd-odd isotopes 230 Pa and 232 Pa were investigated for the first time. For the isotope 230 Pa 63 states below an excitation energy of 1.5 MeV are identified. Based on the new experimental data the Nilsson configuration of the ground state is either 1/2[530] p -5/2[633] n or 1/2[530] p +3/2[631] n . In addition 12 rotational bands are proposed and from this six values for the GM splitting energy are deduced as well as two new values for the Newby shift. In the other odd-odd isotope 232 Pa 40 states below an excitation energy of 850 keV are observed and suggestions for the groundstate band and its GM partner are made. From this one GM splitting

  5. Quantitative trait loci at the 11q23.3 chromosomal region related to dyslipidemia in the population of Andhra Pradesh, India.

    Science.gov (United States)

    Pranavchand, Rayabarapu; Reddy, Battini Mohan

    2017-06-13

    Given the characteristic atherogenic dyslipidemia of south Indian population and crucial role of APOA1, APOC3, APOA4 and APOA5 genes clustered in 11q23.3 chromosomal region in regulating lipoprotein metabolism and cholesterol homeostasis, a large number of recently identified variants are to be explored for their role in regulating the serum lipid parameters among south Indians. Using fluidigm SNP genotyping platform, a prioritized set of 96 SNPs of the 11q23.3 chromosomal region were genotyped on 516 individuals from Hyderabad, India, and its vicinity and aged >45 years. The linear regression analysis of the individual lipid traits viz., TC, LDLC, HDLC, VLDL and TG with each of the 78 SNPs that confirm to HWE and with minor allele frequency > 1%, suggests 23 of those to be significantly associated (p ≤ 0.05) with at least one of these quantitative traits. Most importantly, the variant rs632153 is involved in elevating TC, LDLC, TG and VLDLs and probably playing a crucial role in the manifestation of dyslipidemia. Additionally, another three SNPs rs633389, rs2187126 and rs1263163 are found risk conferring to dyslipidemia by elevating LDLC and TC levels in the present population. Further, the ROC (receiver operating curve) analysis for the risk scores and dyslipidemia status yielded a significant area under curve (AUC) = 0.675, suggesting high discriminative power of the risk variants towards the condition. The interaction analysis suggests rs10488699-rs2187126 pair of the BUD13 gene to confer significant risk (Interaction odds ratio = 14.38, P = 7.17 × 10 5 ) towards dyslipidemia by elevating the TC levels (β = 37.13, p = 6.614 × 10 5 ). On the other hand, the interaction between variants of APOA1 gene and BUD13 and/or ZPR1 regulatory genes at this region are associated with elevated TG and VLDL. The variants at 11q23.3 chromosomal region seem to determine the quantitative lipid traits and in turn dyslipidemia in the population of Hyderabad

  6. Fission cross-section measurements on 233U and minor actinides at the CERN n-TOF facility

    International Nuclear Information System (INIS)

    Calviani, M.; Cennini, P.; Chiaveri, E.; Dahlfors, M.; Ferrari, A.; Herrera-Martinez, A.; Kadi, Y.; Sarchiapone, L.; Vlachoudis, V.; Colonna, N.; Terlizzi, R.; Abbondanno, U.; Marrone, S.; Belloni, F.; Fujii, K.; Moreau, C.; Aerts, G.; Andriamonje, S.; Berthoumieux, E.; Dridi, W.; Gunsing, F.; Pancin, J.; Perrot, L.; Plukis, A.; Alvarez, H.; Duran, I.; Paradela, C.; Alvarez-Velarde, F.; Cano-Ott, D.; Embid-Sesura, M.; Gonzalez-Romero, E.; Guerrero, C.; Martinez, T.; Vincente, M. C.; Andrzejewski, J.; Assimakopoulos, P.; Audouin, L.; David, S.; Ferrant, L.; Stephan, C.; Tassan-Got, L.; Badurek, G.; Jericha, E.; Leeb, H.; Oberhummer, H.; Pigni, M. T.; Baumann, P.; Kerveno, M.; Lukic, S.; Rudolf, G.; Becvar, F.; Calvino, F.; Capote, R.; Carrapico, C.; Chepel, V.; Ferreira-Marques, R.; Goncalves, I.; Lindote, A.; Lopes, I.; Neves, F.; Cortes, G.; Poch, A.; Pretel, C.; Couture, A.; Cox, J.; O'Brien, S.; Wiescher, M.; Dillmann, I.; Heil, M.; Kaeppeler, F.; Mosconi, M.; Plag, R.; Walter, S.; Wisshak, K.; Domingo-Pardo, C.; Eleftheriadis, C.; Furman, W.; Goverdovski, A.; Gramegna, F.; Mastinu, P.; Praena, J.; Haas, B.; Haight, R.; Igashira, M.; Karadimos, D.; Karamanis, D.; Ketlerov, V.; Koehler, P.; Konovalov, V.; Kossionides, E.; Krticka, M.; Lampoudis, C.; Lozano, M.; Marganiec, J.; Massimi, C.; Mengoni, A.; Milazzo, P. M.; Papachristodoulou, C.; Papadopoulos, C.; Patronis, N.; Pavlik, A.; Pavlopoulos, P.; Plompen, A.; Quesada, J.; Rauscher, T.; Reifarth, R.; Rubbia, C.; Rullhusen, P.; Salgado, J.; Santos, C.; Savvidis, I.; Tagliente, G.; Tain, J. L.; Tavora, L.; Vannini, G.; Vaz, P.; Ventura, A.; Villamarin, D.; Vlastou, R.; Voss, F.

    2010-01-01

    Neutron-induced fission cross-sections of minor actinides have been measured at the white neutron source n-TOF at CERN, Geneva. The studied isotopes include 233 U, interesting for Th/U based nuclear fuel cycles, 241, 243 Am and 245 Cm, relevant for transmutation and waste reduction studies in new generation fast reactors (Gen-IV) or Accelerator Driven Systems. The measurements take advantage of the unique features of the n-TOF facility, namely the wide energy range, the high instantaneous neutron flux and the low background. Results for the involved isotopes are reported from ∼30 meV to around 1 MeV neutron energy. The measurements have been performed with a dedicated Fission Ionization Chamber (FIC), relative to the standard cross-section of the 235 U fission reaction, measured simultaneously with the same detector. Results are here reported. (authors)

  7. Remote fabrication of (Th, {sup 233}U)O{sub 2} pellet-type fuels for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A

    1981-05-15

    Thorium fuels enriched with {sup 233}U must be fabricated in shielded cells because of high gamma and alpha activity. A conceptual design of a remotely operated plant to produce gamma-active pellet fuels has been made. The plant consists of eight fabrication canyons, two repair canyons, and several miscellaneous cells. Process equipment is modular, easily disconnected, and mounted on plates for easy removal. Equipment consists of a combination of robotics, hard automation, and conventional process equipment. The plant is operated from a central control room with the assistance of a sophisticated computer-based control and information system. Many of the automated process steps are preprogrammed on the control computer and executed on demand by the supervising operator. The technology to build such a plant exists today but needs to be adapted to the needs of the recycle fuel industry. (author)

  8. Hydrogen storage properties of Mg-23.3wt.%Ni eutectic alloy prepared via hydriding combustion synthesis followed by mechanical milling

    International Nuclear Information System (INIS)

    Liquan Li; Yunfeng Zhu; Xiaofeng Liu

    2006-01-01

    A Mg-23.3wt.%Ni eutectic alloy was prepared by the process of hydriding combustion synthesis followed by mechanical milling (HCS+MM). The product showed a high hydriding rate at 373 K and the dehydrogenation started at temperature as low as 423 K. Several reasons contributing to the improvement in hydrogen storage properties were presented. The result of this study will provide attractive information for mobile applications of magnesium hydrogen storage materials, and the process of HCS+MM developed in this study showed its potential for synthesizing magnesium based hydrogen storage materials with novel hydriding/de-hydriding properties. (authors)

  9. Evaluation of fission cross sections and covariances for 233U, 235U, 238U, 239Pu, 240Pu, and 241Pu

    International Nuclear Information System (INIS)

    Kawano, Toshihiko; Matsunobu, Hiroyuki; Murata, Toru

    2000-02-01

    A simultaneous evaluation code SOK (Simultaneous evaluation on KALMAN) has been developed, which is a least-squares fitting program to absolute and relative measurements. The SOK code was employed to evaluate the fission cross sections of 233 U, 235 U, 238 U, 239 Pu, 240 Pu, and 241 Pu for the evaluated nuclear data library JENDL-3.3. Procedures of the simultaneous evaluation and the experimental database of the fission cross sections are described. The fission cross sections obtained were compared with evaluated values given in JENDL-3.2 and ENDF/B-VI. (author)

  10. Reforma tributária: os efeitos macroeconômicos e setoriais da PEC 233/2008

    Directory of Open Access Journals (Sweden)

    Nelson Leitão Paes

    2011-06-01

    Full Text Available A despeito de um histórico desalentador, o atual governo enviou nova proposta de reforma tributária ao Congresso Nacional, a PEC 233/2008. A proposta unifica alguns tributos federais do consumo no IVA-F, simplifica e diminui drasticamente a legislação do ICMS, alivia a tributação sobre a folha de pagamento e bens essenciais e desonera investimentos. Para a análise do impacto destas mudanças, foi construído um modelo de equilíbrio geral, que contempla 55 firmas no lado produtivo da economia. Os resultados sugerem que do lado agregado haverá modesta expansão do produto, consumo, emprego e investimento, com pequena perda de arrecadação e aumento de bem-estar. Houve substanciais alterações no produto setorial, com uma tendência de aumento dos setores industrial e agropecuário em detrimento dos serviços.

  11. Multiplicity and energy of neutrons from {sup 233}U(n{sub th},f) fission fragments

    Energy Technology Data Exchange (ETDEWEB)

    Nishio, Katsuhisa; Kimura, Itsuro; Nakagome, Yoshihiro [Kyoto Univ. (Japan)

    1998-03-01

    The correlation between fission fragments and prompt neutrons from the reaction {sup 233}U(n{sub th},f) was measured with improved accuracy. The results determined the neutron multiplicity and emission energy as a function of fragment mass and total kinetic energy. The average energy as a function of fragment mass followed a nearly symmetric distribution centered about the equal mass-split and formed a remarkable contrast with the saw-tooth distribution of the average neutron multiplicity. The neutron multiplicity from the specified fragment decreases linearly with total kinetic energy, and the slope of multiplicity with kinetic energy had the minimum value at about 130 u. The level density parameter versus mass determined from the neutron data showed a saw-tooth structure with the pronounced minimum at about 128 and generally followed the formula by Gilbert and Cameron, suggesting that the neutron emission process was very much affected by the shell-effect of the fission fragment. (author)

  12. Simultaneous measurement of neutrons and fission fragments of thermal neutron fission of U-233

    International Nuclear Information System (INIS)

    Itsuro Kimura; Katsuhisa Nishio; Yoshihiro Nakagome

    2000-01-01

    The multiplicity and the energy of prompt neutrons from the fragments for 233 U(n th , f) were measured as functions of fragment mass and total kinetic energy. Average neutron energy against the fragment mass showed a nearly symmetric distribution about the half mass division with two valleys at 98 and 145 u. The slope of the neutron multiplicity with total kinetic energy depended on the fragment mass and showed the minimum at about 130 u. The obtained neutron data were applied to determine the total excitation energy of the system, and the resulting value in the typical asymmetric fission lied between 22 and 25 MeV. The excitation energy agreed with that determined by subtracting the total kinetic energy from the Q-value within 1 MeV, thus satisfied the energy conservation. In the symmetric fission, where the mass yield was drastically suppresses, the total excitation energy is significantly large and reaches to about 40 MeV, suggesting that fragment pairs are preferentially formed in a compact configuration at the scission point [ru

  13. Fase aguda da doença de Chagas na Amazônia brasileira: estudo de 233 casos do Pará, Amapá e Maranhão observados entre 1988 e 2005 Acute phase of Chagas disease in the Brazilian Amazon region: study of 233 cases from Pará, Amapá and Maranhão observed between 1988 and 2005

    Directory of Open Access Journals (Sweden)

    Ana Yecê das Neves Pinto

    2008-12-01

    Full Text Available Foram estudados 233 casos de fase aguda da doença de Chagas, oriundos do Pará, Amapá e Maranhão, observados no período de 1988 a 2005, cento e sessenta deles retrospectivamente de 1988 a 2002 e setenta e três prospectivamente de 2003 a 2005. Entre os casos estudados 78,5% (183/233 faziam parte de surtos provavelmente por transmissão oral, acometendo em média 4 pessoas e 21,5% (50/233 eram casos isolados. Foram considerados casos agudos aqueles que apresentaram exames parasitológicos diretos (a fresco, gota espessa ou Quantitative Buffy Coat - QBC e/ou IgM anti-Trypanosoma cruzi positivos. Foram feitos ainda xenodiagnósticos em 224 pacientes e hemoculturas em 213. Todos foram avaliados clinica e epidemiologicamente. As manifestações clínicas mais freqüentes foram febre (100%, cefaléia (92,3%, mialgia (84,1%, palidez (67%, dispnéia (58,4%, edema de membros inferiores (57,9%, edema de face (57,5% dor abdominal (44,2%, miocardite (39,9% e exantema (27%. O eletrocardiograma mostrou alterações de repolarização ventricular em 38,5% dos casos, baixa voltagem de QRS em 15,4% e desvio de SAQRS em 11,5%, extra-sístoles ventriculares em 5,8%, bradicardia em 5,8% e taquicardia em 5,8%, bloqueio de ramo direito em 4,8% e fibrilação atrial em 4,8%. A alteração mais freqüente vista no ecocardiograma foi o derrame pericárdico em 46,2% dos casos. Treze (5,6% pacientes evoluíram para o óbito, 10 (76,9% dos quais por comprometimento cardiovascular, dois por complicações de origem digestiva e um de causa mal definida.Two hundred and thirty-three cases of the acute phase of Chagas disease, from Pará, Amapá and Maranhão, were observed between 1988 and 2005. One hundred and sixty were studied retrospectively from 1988 to 2002 and seventy-three were prospectively followed up from 2003 to 2005. Among the cases studied, 78.5% (183/233 formed part of outbreaks, probably due to oral transmission (affecting a mean of 4 individuals, and 21

  14. Dissolution rates of unirradiated UO2, UO2 doped with 233U, and spent fuel under normal atmospheric conditions and under reducing conditions using an isotope dilution method

    International Nuclear Information System (INIS)

    Ollila, Kaija; Albinsson, Yngve; Oversby, Virginia; Cowper, Mark

    2003-10-01

    The experimental results given in this report allow us to draw the following conclusions. 1) Tests using unirradiated fuel pellet materials from two different manufacturers gave very different dissolution rates under air atmosphere testing. Tests for fragments of pellets from different pellets made by the same manufacturer gave good agreement. This indicates that details of the manufacturing process have a large effect on the behavior of unirradiated UO 2 in dissolution experiments. Care must be taken in interpreting differences in results obtained in different laboratories because the results may be affected by manufacturing effects. 2) Long-term tests under air atmosphere have begun to show the effects of precipitation. Further testing will be needed before the samples reach steady state. 3) Testing of unirradiated UO 2 in systems containing an iron strip to produce reducing conditions gave [U] less than detection limits ( 235 U added as spike was recovered, indicating that 90% of the spike had precipitated onto the solid sample or the iron strip. 9) Tests of UO 2 pellet materials containing 233 U to provide an alpha decay activity similar to that expected for spent fuel 3000 and 10,000 years after disposal showed that the pellet materials behaved as expected under air atmosphere conditions, showing that the manufacturing method was successful. 10) Early testing of the 233 U-doped materials under reducing conditions showed relatively rapid (30 minute) dissolution of small amounts of U at the start of the puff test procedure. Results of analyses of an acidified fraction of the same solutions after 1 or 2 weeks holding indicate that the solutions were inhomogeneous, indicating the presence of colloidal material or small grains of solid. 11) Samples from the 233 U-doped tests initially indicated dissolution of solid during the first week of testing, with some indication of more rapid dissolution of the material with the higher doping. 12) The second cycle of testing

  15. Characterization of Extracellular Dextranase from a Novel Halophilic Bacillus subtilis NRC-B233b a Mutagenic Honey Isolate under Solid State Fermentation

    Directory of Open Access Journals (Sweden)

    Mona A. Esawy

    2012-01-01

    Full Text Available Bacillus subtilis NRC-B233b was isolated from Libyan honey sample proved to be a potent dextranase producer by applying solid state fermentation and utilizing corn flour as the sole carbon source. The optimized culture conditions for dextranase productions were 37°C, pH 10, 32 h, and 20% (v/w moisture content. A unique character of this isolate is its ability to produce steady dextranase irrespective to the presence of NaCl in the medium. The addition of 0.175 Mm CrCl3 increased the enzyme production by about 4.5 fold. Further improvement in enzyme production was achieved by simple UV mutation which increased the enzyme production up to about 2842 U/g. The crude extract has been partially purified about 112-fold from crude extract by only two purification steps involving ultra-filtration. The partially purified dextranase showed its maximum activity at pH 9.2 and 70°C. It retained full activity (100% at 75°C for one hour. Dextranase activity increased about 4 fold in the presence of 10% NaCl. This enzyme showed variable degradation effect on different types of dextran and its derivatives. The treatment of viscous sugar cane juice with the enzyme preparation resulted in clear visual dextran hydrolysis. These results suggest that the dextranase produced by Bacillus subtilis NRC-B233b is industrially applicable.

  16. Splenectomy as a curative treatment for immune thrombocytopenia: a retrospective analysis of 233 patients with a minimum follow up of 10 years

    Science.gov (United States)

    Vianelli, Nicola; Palandri, Francesca; Polverelli, Nicola; Stasi, Roberto; Joelsson, Joel; Johansson, Eva; Ruggeri, Marco; Zaja, Francesco; Cantoni, Silvia; Catucci, Angelo Emanuele; Candoni, Anna; Morra, Enrica; Björkholm, Magnus; Baccarani, Michele; Rodeghiero, Francesco

    2013-01-01

    The treatment of choice in steroid-resistant immune thrombocytopenia is still controversial due to the recent advent of new drugs (anti-CD20 antibodies and thrombopoietin mimetics) that have encouraged a generalized tendency to delay splenectomy. Consequently, it is extremely importance to define the efficacy and safety of splenectomy in the long term. We retrospectively analyzed the data of 233 patients affected by immune thrombocytopenia who underwent splenectomy between 1959 and 2001 in 6 European hematologic institutions and who have now a minimum follow up of ten years from surgery. Of the 233 patients, 180 (77%) achieved a complete response and 26 (11%) a response. Sixty-eight of 206 (33%) responsive patients relapsed, mostly (75%) within four years from first response. In 92 patients (39.5%), further treatment was required after splenectomy that was effective in 76 cases (83%). In 138 patients (59%), response was maintained free of any treatment at last contact. No significant association between baseline characteristics and likelihood of stable response was found. Overall, 73 (31%) and 58 (25%) patients experienced at least one infectious or hemorrhagic complication, which was fatal in 2 and 3 patients, respectively. A stable response to splenectomy was associated with a lower rate of infections (P=0.004) and hemorrhages (PSplenectomy achieved a long-term stable response in approximately 60% of cases. Complications mainly affected non-responding patients and were fatal in a minority. PMID:23144195

  17. Optimization of the binary breeder reactor. VIII annular core fueled with 233U - 238U and Pu-238U

    International Nuclear Information System (INIS)

    Nascimento, J.A. do; Ishiguro, Y.

    1988-04-01

    First cycle burnup characteristics of a 1200 MWe binary breeder reactor with annular core fueled with metallic 233 U- 238 U-Zr, Pu- 238 U-Zr and Th in the blankets have been analysed. The Doppler effect is small as expected in a metal fueled fast reactor. The sodium void reactivity is, in general, smaller than in metal fueled homogeneous fast reactors of 1 m core height. The estimates of the required and available control rod worths show a large shutdown margin throughout the operational cycle. There are flexibilities in the blanket fueling and well balanced breeding in the two cycles, uranium and thorium, with doubling times of about 20 years are possible. (author) [pt

  18. Overview of the recovery and processing of 233U from the Oak Ridge molten salt reactor experiment (MSRE) remediation activities

    International Nuclear Information System (INIS)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.; Trowbridge, L.D.; Williams, D.F.; Toth, L.M.; Dai, S.

    2001-01-01

    The Molten Salt Reactor Experiment (MSRE) was operated at Oak Ridge National Laboratory (ORNL) from 1965 to 1969 to test the concept of a high-temperature, homogeneous, fluid-fueled reactor. The discovery that UF 6 and F 2 migrated from the storage tanks into distant pipes and a charcoal bed resulted in significant activities to remove and recover the 233 U and to decommission the reactor. The recovered fissile uranium will be converted into uranium oxide (U 3 O 8 ), which is a suitable form for long-term storage. This publication reports the research and several new developments that were needed to carry out these unique activities. (author)

  19. Formation and evolution of ultrafine particles produced by radiolysis and photolysis

    International Nuclear Information System (INIS)

    Madelaine, G.J.; Perrin, M.L.; Renoux, A.

    1980-01-01

    Results are presented, concerning the formation, the size distribution, and the behavior of ultrafine particles produced by alpha disintegration of actinium and uv irradiation in filtered and natural atmospheric air. The characterization of these particles is obtained by electrical aerosol analyzer and diffusion battery method. Measurements are made in the range between 0.003 and 0.5 micrometer. Some qualitative indications are obtained on the different mechanisms which govern the evolution of ultrafine particles in the atmosphere (nucleation, coagulation, and condensation). It is now well established that the photo-oxydation of SO 2 in the atmosphere leads to the production of sulphuric acid and of sulphate, which are usually found in the form of submicronic particles. This paper concerns the evolution of ultrafine particles generated in the presence of a preexisting aerosol. They are either instantaneously produced by the alpha disintegrations of actinium 219 or continuously produced by the transformation of SO 2 under uv irradiation

  20. Actinide metals

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Paul L. [Geochem Australia, Kiama, NSW (Australia); Ekberg, Christian [Chalmers Univ. of Technology, Goeteborg (Sweden). Nuclear Chemistry/Industrial Materials Recycling

    2016-07-01

    All isotopes of actinium are radioactive and exist in aqueous solution only in the trivalent state. There have been very few studies on the hydrolytic reactions of actinium(III). The hydrolysis reactions for uranium would only be important in alkaline pH conditions. Thermodynamic parameters for the hydrolysis species of uranium(VI) and its oxide and hydroxide phases can be determined from the stability and solubility constants. The hydrolytic behaviour of neptunium(VI) is quite similar to that of uranium(VI). The solubility constant of NpO{sub 2}OH(am) has been reported a number of times for both zero ionic strength and in fixed ionic strength media. Americium can form four oxidation states in aqueous solution, namely trivalent, tetravalent, pentavalent and hexavalent. Desire, Hussonnois and Guillaumont determined stability constants for the species AmOH{sup 2+} for the actinides, plutonium(III), americium(III), curium(III), berkelium(III) and californium(III) using a solvent extraction technique.

  1. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, J.D.; Horton, K.E.

    1991-01-01

    An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Subsequent to the directive to shut down the Fast Flux Test Facility in early 1990, a variety of activities were undertaken to provide support for continued operation. The United States has made substantial progress in achieving ALMR program objectives. The metal fuel cycle is designed to recycle and burn its own actiniums, and has the potential to be a very effective burner of actiniums generated in the LWRs. The current emphasis in the IFR Program is on the comprehensive development of the IFR (Integral Fast Reactor) technology, to be followed by a period of technology demonstration which would verify the economic feasibility of the concept. The United States has been active in international cooperative activities in the fast reactor sector since 1969. (author). 11 figs, 1 tab

  2. Actinide metals

    International Nuclear Information System (INIS)

    Brown, Paul L.; Ekberg, Christian

    2016-01-01

    All isotopes of actinium are radioactive and exist in aqueous solution only in the trivalent state. There have been very few studies on the hydrolytic reactions of actinium(III). The hydrolysis reactions for uranium would only be important in alkaline pH conditions. Thermodynamic parameters for the hydrolysis species of uranium(VI) and its oxide and hydroxide phases can be determined from the stability and solubility constants. The hydrolytic behaviour of neptunium(VI) is quite similar to that of uranium(VI). The solubility constant of NpO 2 OH(am) has been reported a number of times for both zero ionic strength and in fixed ionic strength media. Americium can form four oxidation states in aqueous solution, namely trivalent, tetravalent, pentavalent and hexavalent. Desire, Hussonnois and Guillaumont determined stability constants for the species AmOH 2+ for the actinides, plutonium(III), americium(III), curium(III), berkelium(III) and californium(III) using a solvent extraction technique.

  3. On the Relative Signs of "ROT-Effects" in Ternary and Binary Fission of 233U and 235U Nuclei Induced by Polarized Cold Neutrons

    Science.gov (United States)

    Danilyan, G. V.

    2018-02-01

    Signs of the ROT-effects in ternary fission of 233U and 235U experimentally defined by PNPI group are the same, whereas in binary fission defined by ITEP group are opposite. This contradiction cannot be explained by the errors in the experiments of both groups, since such instrumental effects would be too large not to be noticed. Therefore, it is necessary to find the answer to this problem in the differences of the ternary and binary fission mechanisms.

  4. Study of the variation with the energy of the fission cross-sections of {sup 233}U, {sup 235}U, {sup 239}Pu for the fast neutrons; Etude de la variation avec l'energie des sections efficaces de fission de {sup 233}U, {sup 235}U, {sup 239}Pu pour les neutrons rapides

    Energy Technology Data Exchange (ETDEWEB)

    Szteinsznaider, D; Naggiar, V; Netter, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    This measurements have been done while taking the value of the fission cross-sections of {sup 238}U as reference. The neutrons are produced by the reaction {sup 7}Li(p,n) in the Van de Graaff generator of Saclay. The explored domain spreads from some tenths to 2000 keV. We find: for {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, cross-section constant between 150 and 2000 keV, for {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, cross-section constant between 700 and 1000 keV, for {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, for neutrons of 850 keV. (authors) [French] Ces mesures ont ete effectuees en prenant la valeur de la section efficace de fission de {sup 238}U comme reference. Les neutrons sont produits par la reaction {sup 7}Li(p,n) au generateur Van de Graaff de Saclay. Le domaine explore s'etend de quelques dizaines de kev a 2000 kev. Nous trouvons: pour {sup 239}Pu: {sigma}{sub f} = 2,04 {+-} 0,12 barns, section efficace constante entre 150 et 2000 kev. pour {sup 235}U: {sigma}{sub f} = 1,15 {+-} 0,15 barns, section efficace constante entre 700 et 1000 kev. pour {sup 233}U: {sigma}{sub f} = 1,92 {+-} 0,25 barns, pour des neutrons de 850 kev. (auteurs)

  5. cis- and trans-2,3,3a,4,5,9b-Hexahydro-1H-benz[e]indoles: synthesis and evaluation of dopamine D2, and D3 receptor binding affinity

    DEFF Research Database (Denmark)

    Song, Xiaodong; Crider, Michael A.; Cruse, Sharon F.

    1999-01-01

    cis- and trans-2,3,3a,4,5,9b-hexahydro-1H-benz [e]indoles were synthesized as conformationally rigid analogues of 3-phenylpyrrolidine and evaluated for dopamine (DA) D2S and D3 receptor binding affinity. The tricyclic benz[e]indole nucleus was constructed by a previously reported reductive...... configuration. These novel ligands may be useful tools for gaining additional information about the DA D3 receptor. Copyright Elsevier, Paris.dopamine / D2S receptor / D3 receptor / cis- and trans-2,3,3a,4,5,9b-hexahydro-1H-benz[e]indoles / receptor binding affinity....... receptors was shown by compounds substituted with N-n-propyl or N-allyl groups. The cis-(+-)-N-allyl derivative 21e demonstrated a D2S/D3 selectivity of 290. Resolution of cis-(+-)-5 and trans-(+-)- 21c into individual enantiomers showed that in both series the more active isomer had 3aR absolute...

  6. Measurement of the Neutron Capture Cross Sections of $^{233}$U, $^{237}$Np, $^{240,242}$Pu, $^{241,243}$Am and $^{245}$Cm with a Total Absorption Calorimeter at n_TOF

    CERN Multimedia

    Beer, H; Wiescher, M; Cox, J; Rapp, W; Embid, M; Dababneh, S

    2002-01-01

    Accurate and reliable neutron capture cross section data for actinides are necessary for the poper design, safety regulation and precise performance assessment of transmutation devices such as Fast Critical Reactors or Accelerator Driven Systems (ADS). The goal of this proposal is the measurement of the neutron capture cross sections of $^{233}$U, $^{237}$Np, $^{240,242}$Pu, $^{241,243}$Am and $^{245}$Cm at n_TOF with an accuracy of 5~\\%. $^{233}$U plays an essential role in the Th fuel cycle, which has been proposed as a safer and cleaner alternative to the U fuel cycle. The capture cross sections of $^{237}$Np,$^{240,242}$Pu, $^{241,243}$Am and $^{245}$Cm play a key role in the design and optimization of a strategy for the Nuclear Waste Transmutation. A high accuracy can be achieved at n_TOF in such measurements due to a combination of features unique in the world: high instantaneous neutron fluence and excellent energy resolution of the facility, innovative Data Acquisition System based on flash ADCs and t...

  7. Fission Cross-section Measurements of (233)U, (245)Cm and (241,243)Am at CERN n_TOF Facility

    CERN Document Server

    Calviani, M; Andriamonje, S; Chiaveri, E; Vlachoudis, V; Colonna, N; Meaze, M H; Marrone, S; Tagliente, G; Terlizzi, R; Belloni, F; Abbondanno, U; Fujii, K; Milazzo, P M; Moreau, C; Aerts, G; Berthoumieux, E; Dridi, W; Gunsing, F; Pancin, J; Perrot, L; Plukis, A; Alvarez, H; Duran, I; Paradela, C; Alvarez-Velarde, F; Cano-Ott, D; Gonzalez-Romero, E; Guerrero, C; Martinez, T; Villamarin, D; Vicente, M C; Andrzejewski, J; Marganiec, J; Assimakopoulos, P; Karadimos, D; Karamanis, D; Papachristodoulou, C; Patronis, N; Audouin, L; David, S; Ferrant, L; Isaev, S; Stephan, C; Tassan-Got, L; Badurek, G; Jericha, E; Leeb, H; Oberhummer, H; Pigni, M T; Baumann, P; Kerveno, M; Lukic, S; Rudolf, G; Becvar, F; Krticka, M; Calvino, F; Capote, R; Carrillo De Albornoz, A; Marques, L; Salgado, J; Tavora, L; Vaz, P; Cennini, P; Dahlfors, M; Ferrari, A; Gramegna, F; Herrera-Martinez, A; Kadi, Y; Mastinu, P; Praena, J; Sarchiapone, L; Wendler, H; Chepel, V; Ferreira-Marques, R; Goncalves, I; Lindote, A; Lopes, I; Neves, F; Cortes, G; Poch, A; Pretel, C; Couture, A; Cox, J; O'brien, S; Wiescher, M; Dillman, I; Heil, M; Kappeler, F; Mosconi, M; Plag, R; Voss, F; Walter, S; Wisshak, K; Dolfini, R; Rubbia, C; Domingo-Pardo, C; Tain, J L; Eleftheriadis, C; Savvidis, I; Frais-Koelbl, H; Griesmayer, E; Furman, W; Konovalov, V; Goverdovski, A; Ketlerov, V; Haas, B; Haight, R; Reifarth, R; Igashira, M; Koehler, P; Kossionides, E; Lampoudis, C; Lozano, M; Quesada, J; Massimi, C; Vannini, G; Mengoni, A; Oshima, M; Papadopoulos, C; Vlastou, R; Pavlik, A; Pavlopoulos, P; Plompen, A; Rullhusen, P; Rauscher, T; Rosetti, M; Ventura, A

    2011-01-01

    Neutron-induced fission cross-sections of minor actinides have been measured using the n_TOF white neutron source at CERN, Geneva, as part of a large experimental program aiming at collecting new data relevant for nuclear astrophysics and for the design of advanced reactor systems. The measurements at n_TOF take advantage of the innovative features of the n_TOF facility, namely the wide energy range, high instantaneous neutron flux and good energy resolution. Final results on the fission cross-section of 233U, 245Cm and 243Am from thermal to 20 MeV are here reported, together with preliminary results for 241Am. The measurement have been performed with a dedicated Fast Ionization Chamber (FIC), a fission fragment detector with a very high efficiency, relative to the very well known cross-section of 235U, measured simultaneously with the same detector.

  8. Measurement of the $^{233}$U neutron capture cross section at the n_TOF facility at CERN

    CERN Document Server

    Carrapiço, Carlos; Berthoumieux, Eric; Gonçalves, Isabel; Gunsing, Frank

    2012-12-12

    The Thorium-Uranium (Th-U) fuel cycle has been envisaged as an alternative to the Uranium-Plutonium (U-Pu) fuel cycle for electricity generation using nuclear power reactors. Indeed, thorium can be used as a nuclear fuel, and several studies and R&D programs seem to provide evidence on the sustainability of the Th-U fuel cycle, due to (i) the natural abundance of Thorium, (ii) the improved proliferation resistance offered by the Th-U fuel cycle relative to the U-Pu fuel cycle, (iii) the better neutronics performance of the Th-U fuel cycle throughout the whole neutron energy range compared to the U-Pu fuel cycle, (iv) the lower radiotoxicity of the generated spent fuel in reactors with Th-U fuel cycle and, consequently (v) better economics and public acceptance of the reactors operated using the Th-U fuel cycle compared to those using the U-Pu fuel cycle (prior to the Generation IV nuclear reactors). In a nuclear reactor operated using the Th-U fuel cycle, $^{233}$U is a key nuclide governing the neutr...

  9. DISSOLVED CONCENTRATION LIMITS OF RADIOACTIVE ELEMENTS

    Energy Technology Data Exchange (ETDEWEB)

    P. Bernot

    2005-07-13

    The purpose of this study is to evaluate dissolved concentration limits (also referred to as solubility limits) of elements with radioactive isotopes under probable repository conditions, based on geochemical modeling calculations using geochemical modeling tools, thermodynamic databases, field measurements, and laboratory experiments. The scope of this activity is to predict dissolved concentrations or solubility limits for elements with radioactive isotopes (actinium, americium, carbon, cesium, iodine, lead, neptunium, plutonium, protactinium, radium, strontium, technetium, thorium, and uranium) relevant to calculated dose. Model outputs for uranium, plutonium, neptunium, thorium, americium, and protactinium are provided in the form of tabulated functions with pH and log fCO{sub 2} as independent variables, plus one or more uncertainty terms. The solubility limits for the remaining elements are either in the form of distributions or single values. Even though selection of an appropriate set of radionuclides documented in Radionuclide Screening (BSC 2002 [DIRS 160059]) includes actinium, transport of Ac is not modeled in the total system performance assessment for the license application (TSPA-LA) model because of its extremely short half-life. Actinium dose is calculated in the TSPA-LA by assuming secular equilibrium with {sup 231}Pa (Section 6.10); therefore, Ac is not analyzed in this report. The output data from this report are fundamental inputs for TSPA-LA used to determine the estimated release of these elements from waste packages and the engineered barrier system. Consistent modeling approaches and environmental conditions were used to develop solubility models for the actinides discussed in this report. These models cover broad ranges of environmental conditions so they are applicable to both waste packages and the invert. Uncertainties from thermodynamic data, water chemistry, temperature variation, and activity coefficients have been quantified or

  10. DISSOLVED CONCENTRATION LIMITS OF RADIOACTIVE ELEMENTS

    International Nuclear Information System (INIS)

    P. Bernot

    2005-01-01

    The purpose of this study is to evaluate dissolved concentration limits (also referred to as solubility limits) of elements with radioactive isotopes under probable repository conditions, based on geochemical modeling calculations using geochemical modeling tools, thermodynamic databases, field measurements, and laboratory experiments. The scope of this activity is to predict dissolved concentrations or solubility limits for elements with radioactive isotopes (actinium, americium, carbon, cesium, iodine, lead, neptunium, plutonium, protactinium, radium, strontium, technetium, thorium, and uranium) relevant to calculated dose. Model outputs for uranium, plutonium, neptunium, thorium, americium, and protactinium are provided in the form of tabulated functions with pH and log fCO 2 as independent variables, plus one or more uncertainty terms. The solubility limits for the remaining elements are either in the form of distributions or single values. Even though selection of an appropriate set of radionuclides documented in Radionuclide Screening (BSC 2002 [DIRS 160059]) includes actinium, transport of Ac is not modeled in the total system performance assessment for the license application (TSPA-LA) model because of its extremely short half-life. Actinium dose is calculated in the TSPA-LA by assuming secular equilibrium with 231 Pa (Section 6.10); therefore, Ac is not analyzed in this report. The output data from this report are fundamental inputs for TSPA-LA used to determine the estimated release of these elements from waste packages and the engineered barrier system. Consistent modeling approaches and environmental conditions were used to develop solubility models for the actinides discussed in this report. These models cover broad ranges of environmental conditions so they are applicable to both waste packages and the invert. Uncertainties from thermodynamic data, water chemistry, temperature variation, and activity coefficients have been quantified or otherwise

  11. Kinetics of radioisotope exchange between brine and rock in a geothermal system

    International Nuclear Information System (INIS)

    Hammond, D.E.; Zukin, J.G.; Teh-Lung Ku

    1988-01-01

    A wide range of isotopes in the /sup 238/U, /sup 235/U, and /sup 232/Th decay chains was measured in geothermal brines collected from two production zones at 1898 and 3220 m in the Salton Sea Scientific Drilling Project well. High concentrations of radium, radon, and lead isotopes are generated and maintained by the input of these isotopes from solid phases into brine by both recoil and leaching processes, by the high chloride content of the brine which complexes radium and lead, and by the apparent absence of suitable unoccupied adsorption sites. In contrast, uranium, thorium, actinium, bismuth, and polonium isotopes all have low concentrations due to their efficient sorption from brine to rock. Measurements of short-lived isotopes in these decay series yield insights regarding the mechanisms controlling radioisotope exchange, and they permit estimation of rates of brine-rock interaction. For example, the /sup 228/Ac//sup 228/Ra activity ratio of 0.2 in brines indicates that the mean residence time of actinium in solution before sorption onto solid surfaces is less than 2.5 hours

  12. Study of the production of uranium-233 in natural thorium subjected to a neutron flux of 14 MeV

    International Nuclear Information System (INIS)

    Abel, G.; Martel, J.G.; St Germain, J.P.

    1979-01-01

    This is a study of neutron flux and reactivity in several simplified models of a fusion reactor blanket composed of thorium, carbon and a stainless steel structure. The objective is the comparative determination, theoretical and experimental, of values for the conversion of nuclei of Th-232 into fissile U-233 nuclei in such a blanket. Theoretical calculations are carried out using the ANISN (transport equation) and MORSE (Monte Carlo) codes. Experimental values are measured in a stainless steel rack structure allowing simplified blanket models to be mounted around a 14 MeV neutron source. In the first year of work theoretical flux values have been obtained, the necesssary detectors constructed, the rack structure set up, programs worked out for the reactivity calculations, and finally verification methods worked out in conditions similar to those that will prevail. (LL) [fr

  13. Study of the variation with the energy of the fission cross-sections of 233U, 235U, 239Pu for the fast neutrons

    International Nuclear Information System (INIS)

    Szteinsznaider, D.; Naggiar, V.; Netter, F.

    1955-01-01

    This measurements have been done while taking the value of the fission cross-sections of 238 U as reference. The neutrons are produced by the reaction 7 Li(p,n) in the Van de Graaff generator of Saclay. The explored domain spreads from some tenths to 2000 keV. We find: for 239 Pu: σ f = 2,04 ± 0,12 barns, cross-section constant between 150 and 2000 keV, for 235 U: σ f = 1,15 ± 0,15 barns, cross-section constant between 700 and 1000 keV, for 233 U: σ f = 1,92 ± 0,25 barns, for neutrons of 850 keV. (authors) [fr

  14. Fission cross section ratios for 233,234,236U relative to 235U from 0.5 to 400 MeV

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Gavron, A.; Parker, W.E.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

    1991-01-01

    Neutron-induced fission cross section ratios from 0.5 to 400 MeV for samples of 233, 234, 236 U relative to 235 U have been measured at the WNR neutron Source at Los Alamos. The fission reaction rate was determined using a fast parallel plate ionization chamber at a 20-m flight path. Cross sections over most the energy range were also extracted using the neutron fluence determined with three different proton telescope arrangements. Those data provided the shape of the 235 U(n,f) cross section relative to the hydrogen scattering cross section. That shape was then normalized to the very accurately known value for 235 U(n,f) at 14.1 MeV to allow us to obtain cross section section values from the ratio data and our values for 235 U(n,f). 6 refs., 1 fig

  15. Fission cross section ratios for 233,234,236U relative to 235U from 0.5 to 400 MeV

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Gavron, A.; Parker, W.E.; Balestrini, S.J.; Carlson, A.D.; Wasson, O.A.; Hill, N.W.

    1992-01-01

    Neutron-induced fission cross section ratios from 0.5 to 400 MeV for samples of 233,234,236 U relative to 235 U have been measured at the WNR neutron Source at Los Alamos. The fission reaction rate was determined using a fast parallel plate ionization chamber at a 20-m flight path. Cross sections over most of the energy range were also extracted using the neutron fluence determined with three different proton telescope arrangements. Those data provided the shape of the 235 U(n, f) cross section relative to the hydrogen scattering cross section. That shape was then normalized to the very accurately known value for 235 U(n, f) at 14.1 MeV which will allow us to obtain cross section values from the ratio data and our values for 235 U(n, f). (orig.)

  16. Calculated critical parameters in simple geometries for oxide and nitrate water mixtures of U-233, U-235 and Pu-239 with thorium. Final report

    International Nuclear Information System (INIS)

    Converse, W.E.; Bierman, S.R.

    1979-11-01

    Calculations have been performed on water mixtures of oxides and nitrates of 233 U, 235 U, and 239 Pu with chemically similar thorium compounds to determine critical dimensions for simple geometries (sphere, cylinder, and slab). Uranium enrichments calculated were 100%, 20%, 10%, and 5%; plutonium calculations assumed 100% 239 Pu. Thorium to uranium or plutonium weight ratios (Th: U or Pu) calculated were 0, 1, 4, and 8. Both bare and full water reflection conditions were calculated. The results of the calculations are plotted showing a critical dimension versus the uranium or plutonium concentration. Plots of K-infinity and material buckling for each material type are also shown

  17. Determination of 233U, 235U, 238U and 239Pu fission yields induced by fission and 14.7 MeV neutrons

    International Nuclear Information System (INIS)

    Laurec, Jean; Adam, Albert; Bruyne, Thierry de.

    1981-12-01

    The 233 U, 235 U, 238 U, 239 Pu fission yields have been determined by a radiochemical method. A target and a fission chamber made of same fissible material are irradied together. The total fission number is measured from the fission chamber. The fission product activities are directly measured on the target using calibrated Ge-Li detectors. The fissible material masses are determined by alpha and mass spectrometries. The irradiations were made on the critical assemblies PROSPERO and CALIBAN and on the 14 MeV neutron generator of C.E. VALDUC. 3 to 5% fission yield errors are got for the most measured nuclides: 95 Zr, 97 Zr, 99 Mo, 103 Ru, 131 I, 132 Te, 140 Ba, 141 Ce, 143 Ce, 144 Ce, 147 Nd [fr

  18. Objectieve en subjectieve verkeersveiligheid van het N233-kruispunt Rhenen-Achterberg : inventarisatie van zorgpunten bij bewoners, enquête onder (ouders van) scholieren en beoordeling van de huidige en toekomstige verkeerssituatie. Onderzoek in opdracht van de Provincie Utrecht.

    NARCIS (Netherlands)

    Bax, C.A. Hoekstra, A.T.G. & Schermers, G.

    2017-01-01

    Objective and subjective road safety of the Rhenen-Achterberg intersection on the N233 provincial road : Inventory of concerns among residents, survey among (parents of) students and assessment of the current and future traffic situation. The province of Utrecht asked SWOV to investigate the

  19. Measurement of neutron-induced fission cross-sections of Th232, U238, U233 and Np237 relative to U235 from 1 MeV to 200 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Shcherbakov, O.A.; Laptev, A.B.; Petrov, G.A. [Petersburg Nuclear Physics Inst., Gatchina, Leningrad district (Russian Federation); Fomichev, A.V.; Donets, A.Y.; Osetrov, O.I.

    1998-11-01

    The measurements of neutron-induced cross-section ratios for Th232, U238, U233 and Np237 relative to U235 have been carried out in the energy range from 1 MeV up to 200 MeV using the neutron time-of-flight spectrometer GNEIS based on 1 GeV proton synchrocyclotron. Below 20 MeV, the results of present measurements are roughly in agreement with evaluated data though there are some discrepances to be resolved. (author)

  20. LASL experience in decontamination of the environment

    International Nuclear Information System (INIS)

    Ahlquist, A.J.

    1981-01-01

    This discussion represents one part of a major effort in soil decontamination at the Los Alamos site. A contaminated industrial waste line in the Los Alamos townsite was removed, and a plutonium incineration facility, and a filter building contaminated with actinium-227 were dismantled. The former plutonium handling facility has been decontaminated, and canyons and an old firing site contaminated with strontium-90 have been surveyed

  1. Determination of radionuclides in discharged water from gold ...

    African Journals Online (AJOL)

    Long-lived radionuclides from the Uranium-, Thorium- and Actinium-decay chains in the discharged water into the environment were radiochemically separated and the activity concentrations determined for 238U-series ranged from 3.8 ± 1.5 to 178 ± 19 mBqL-1, 232Th-series ranged from < 2.0 to 47.8 ± 7.3 mBqL-1 and ...

  2. PROJECT EXPERIENCE REPORT DEMOLITION OF HANFORDS 233-S PLUTONIUM CONCENTRATION FACILITY

    International Nuclear Information System (INIS)

    BERLIN, G.T.; ORGILL, T.K.

    2004-01-01

    This report provides a summary of the preparation, operations, innovative work practices, and lessons learned associated with demolition of the 2334 Plutonium Concentration Facility. This project represented the first open-air demolition of a highly-contaminated plutonium facility at the Hanford Site. This project may also represent the first plutonium facility in the US. Department of Energy (DOE) complex to have been demolished without first decontaminating surfaces to near ''free release'' standards. Demolition of plutonium contaminated structures, if not properly managed, can subject cleanup personnel and the environment to significant risk. However, with proper sequencing and innovative use of commercially available equipment, materials, and services, this project demonstrated that a plutonium processing facility can be demolished while avoiding the need to perform extensive decontamination or to construct large enclosures. This project utilized an excavator with concrete shears, diamond circular saws, water misting and fogging equipment, commercially available fixatives and dust suppressants, conventional mobile crane and rigging services, and near real-time modeling of meteorological and radiological conditions. Following a significant amount of preparation, actual demolition of the 233-S Facility began in October 2003 and was completed in late April 2004. The knowledge and experience gained on this project are important to the Hanford Site as additional plutonium processing facilities are scheduled for demolition in the near future. Other sites throughout the DOE Complex may also be faced with similar challenges. Numerous innovations and effective work practices were implemented on this project. Accordingly, a series of ''Lessons Learned and Innovative Practices Fact Sheets'' were developed and are included as an appendix to this report. This collection of fact sheets is not intended to capture every innovative work practice and lesson learned, but rather

  3. Development and validation of an ICP-MS method for the determination of elemental impurities in TP-6076 active pharmaceutical ingredient (API) according to USP 〈232〉/〈233〉.

    Science.gov (United States)

    Chahrour, Osama; Malone, John; Collins, Mark; Salmon, Vrushali; Greenan, Catherine; Bombardier, Amy; Ma, Zhongze; Dunwoody, Nick

    2017-10-25

    The new guidelines of the United States pharmacopeia (USP), European pharmacopeia (EP) and international conference on harmonization (ICH) regulating elemental impurities limits in pharmaceuticals signify the end of unspecific analysis of metals as outlined in USP 〈231〉. The new guidelines specify both daily doses and concentration/limits of elemental impurities in pharmaceutical final products, active pharmaceutical ingredients (API) and excipients. In chapter USP 〈233〉 method implementation, validation and quality control during the analytical process are described. We herein report the use of a stabilising matrix that overcomes low spike recovery problem encountered with Os and allows the determination of all USP required elemental impurities (As, Cd, Hg, Pb, V, Cr, Ni, Mo, Cu, Pt, Pd, Ru, Rh, Os and Ir) in a single analysis. The matrix was used in the validation of a method to determine elemental impurities in TP-6076 active pharmaceutical ingredient (API) by ICP-MS according to the procedures defined in USP〈233〉 and to GMP requirements. This validation will support the regulatory submission of TP-6076 which is a novel tetracycline analogue effective against the most urgent multidrug-resistant gram-negative bacteria. Evaluation of TP-6076 in IND-enabling toxicology studies has led to the initiation of a phase 1 clinical trial. Copyright © 2017 Elsevier B.V. All rights reserved.

  4. Evaluation of fission cross sections and covariances for {sup 233}U, {sup 235}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu

    Energy Technology Data Exchange (ETDEWEB)

    Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan); Matsunobu, Hiroyuki [Data Engineering, Inc. (Japan); Murata, Toru [AITEL Corporation, Tokyo (JP)] [and others

    2000-02-01

    A simultaneous evaluation code SOK (Simultaneous evaluation on KALMAN) has been developed, which is a least-squares fitting program to absolute and relative measurements. The SOK code was employed to evaluate the fission cross sections of {sup 233}U, {sup 235}U, {sup 238}U, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu for the evaluated nuclear data library JENDL-3.3. Procedures of the simultaneous evaluation and the experimental database of the fission cross sections are described. The fission cross sections obtained were compared with evaluated values given in JENDL-3.2 and ENDF/B-VI. (author)

  5. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    International Nuclear Information System (INIS)

    Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

    1979-12-01

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

  6. Nylon and teflon scribe effect on NBR to Chemlok 233 and NBR to NBR bond interfaces

    Science.gov (United States)

    Jensen, S. K.

    1990-01-01

    A study was requested by Manufacturing Engineering to determine what effects marking with nylon (6/6) and Teflon scribes may have on subsequent bonding. Witness panel bond specimens were fabricated by the development lab to test both acrylonitrile butadiene rubber (NBR) to Chemlok and NBR to NBR after controlled exposure. The nylon rod used as a scribe tool demonstrates virtually no bond deterioration when used to scribe lines on either the Chemlok to NBR surfaces or the NBR to NBR interface. Lab test results indicate that the nylon rod-exposed samples produce tensile and peel values very similar to the control samples and the Teflon exposed samples produce tensile and peel values much lower than the control samples. Visual observation of the failure surfaces of the tested samples shows that Teflon scribing produces an obvious contamination to the surface and the nylon produces no effect. Photographs of test samples are provided. It is concluded that Teflon stock used as a scribe tool on a Chemlok 233 to NBR surface or an NBR to NBR surface has a detrimental effect on the bond integrity on either of these bond interfaces. Therefore, it is recommended that the nylon rod continue to be used where a scribe line is required in the redesigned solid rocket motor segment insulation layup operations. The use of Teflon scribes should not be considered.

  7. Dissolution of unirradiated UO{sub 2} and UO{sub 2} doped with {sup 233}U under reducing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ollila, K. [VTT Processes (Finland); Oversby, V.M. [VMO Konsult (Sweden)

    2005-01-01

    Experiments have been conducted to determine an upper limit to the dissolution rate of UO{sub 2} under reducing conditions appropriate to those in a geologic repository for spent fuel disposal in Finland and Sweden. Test duration ranged from 52 to 140 days. The total amount of U recovered in each test was converted into a dissolution rate per year for the sample. The dissolution rate was then used to calculate an expected lifetime for the samples under the test conditions. The dissolution rate did not depend on the length of the testing period. Rather, the dissolution rate appeared to decrease as the samples were exposed to sequential testing periods. This indicates that the results are still influenced by transient effects such as high-energy surface sites, which implies that the dissolution rates measured are upper limits. The sample lifetimes calculated from the last two testing periods, which had a total of 269 days, ranged from 7 to 10 million years. There was no indication of an effect of alpha radiolysis on the dissolution rate results for samples with doping levels of 0, 5, and 10% {sup 233}U.

  8. Impact of uranium-233/thorium cycle on advanced accountability concepts and fabrication facilities. Addendum 2 to application of advanced accountability concepts in mixed oxide fabrication

    International Nuclear Information System (INIS)

    Bastin, J.J.; Jump, M.J.; Lange, R.A.; Crandall, C.C.

    1977-11-01

    The Phase I study of the application of advanced accountability methods (DYMAC) in a uranium/plutonium mixed oxide facility was extended to cover the possible fabrication of uranium-233/thorium fuels. Revisions to Phase II of the DYMAC plan which would be necessitated by such a process are specified. These revisions include shielding requirements, measurement systems, licensing conditions, and safeguards considerations. The impact of the uranium/thorium cycle on a large-scale fuel fabrication facility was also reviewed; it was concluded that the essentially higher radioactivity of uranium/thorium feeds would lead to increased difficulties which tend to preclude early commercial application of the process. An amended schedule for Phase II is included

  9. Preliminary study of the {alpha} ratio measurement, ratio of the neutron capture cross section to the fission one for {sup 233}U, on the PEREN platform. Development and study of the experimental setup; Etude preliminaire de la mesure du rapport {alpha}, rapport de la section efficace moyenne de capture sur celle de fission de l'{sup 233}U, sur la plateforme PEREN. Developpement et etude du dispositif experimental

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, M A

    2007-12-15

    Producing nuclear energy in order to reduce anthropic CO{sub 2} emission and to meet high energy demand, implies three conditions to the nuclear plants of the IV. generation: safety improvements, radioactive waste minimization, and fuel breeding for a sustainable use of the resources. The Thorium fuel cycle used in Molten Salt Reactors seems promising. Many numerical studies based on probabilistic codes are carried out in order to analyse the behaviour of such reactors. Nevertheless, one of the most important parameters is badly known: the alpha ratio of {sup 233}U, ratio of the neutron capture cross section to fission one for {sup 233}U. This key-parameter is necessary to calculate the breeding ratio and thus, the deployment capacities of those reactors. This Ph-D thesis was intended to prepare a precise measurement of the alpha ratio of {sup 233}U between 1 eV and 10 keV. Preliminary measurements have been performed on the experimental platform PEREN. This experimental environment is composed of a lead slowing-down time spectrometer associated with an intense pulsed neutron generator. Capture and fission rates are measured thanks to eight scintillators with their photomultipliers, surrounding a fission chamber. A software analysis sets the coincidence rate between the scintillators. In order to understand perfectly the experimental setup, preliminary tests using a {sup 235}U fission chamber have been done. This experiment resulted in a very low signal to background ratio (1 %). The background coming from the scintillators themselves seriously handicapped the measurement. Another series of experiment has been done with scintillators 5 times thinner. Nevertheless, the signal to background ratio should still be increased to measure the capture of {sup 235}U. To make sure that the experimental setup has totally been understood, we made many comparisons between experimental results and simulations. Two simulation codes were mainly used: MCNP and GEANT4. We paid

  10. Preliminary study of the {alpha} ratio measurement, ratio of the neutron capture cross section to the fission one for {sup 233}U, on the PEREN platform. Development and study of the experimental setup; Etude preliminaire de la mesure du rapport {alpha}, rapport de la section efficace moyenne de capture sur celle de fission de l'{sup 233}U, sur la plateforme PEREN. Developpement et etude du dispositif experimental

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, M.A

    2007-12-15

    Producing nuclear energy in order to reduce anthropic CO{sub 2} emission and to meet high energy demand, implies three conditions to the nuclear plants of the IV. generation: safety improvements, radioactive waste minimization, and fuel breeding for a sustainable use of the resources. The Thorium fuel cycle used in Molten Salt Reactors seems promising. Many numerical studies based on probabilistic codes are carried out in order to analyse the behaviour of such reactors. Nevertheless, one of the most important parameters is badly known: the alpha ratio of {sup 233}U, ratio of the neutron capture cross section to fission one for {sup 233}U. This key-parameter is necessary to calculate the breeding ratio and thus, the deployment capacities of those reactors. This Ph-D thesis was intended to prepare a precise measurement of the alpha ratio of {sup 233}U between 1 eV and 10 keV. Preliminary measurements have been performed on the experimental platform PEREN. This experimental environment is composed of a lead slowing-down time spectrometer associated with an intense pulsed neutron generator. Capture and fission rates are measured thanks to eight scintillators with their photomultipliers, surrounding a fission chamber. A software analysis sets the coincidence rate between the scintillators. In order to understand perfectly the experimental setup, preliminary tests using a {sup 235}U fission chamber have been done. This experiment resulted in a very low signal to background ratio (1 %). The background coming from the scintillators themselves seriously handicapped the measurement. Another series of experiment has been done with scintillators 5 times thinner. Nevertheless, the signal to background ratio should still be increased to measure the capture of {sup 235}U. To make sure that the experimental setup has totally been understood, we made many comparisons between experimental results and simulations. Two simulation codes were mainly used: MCNP and GEANT4. We paid

  11. Short communication an interferon-γ ELISPOT assay with two cytotoxic T cell epitopes derived from HTLV-1 tax region 161-233 discriminates HTLV-1-associated myelopathy/tropical spastic paraparesis patients from asymptomatic HTLV-1 carriers in a Peruvian population.

    Science.gov (United States)

    Best, Ivan; López, Giovanni; Talledo, Michael; MacNamara, Aidan; Verdonck, Kristien; González, Elsa; Tipismana, Martín; Asquith, Becca; Gotuzzo, Eduardo; Vanham, Guido; Clark, Daniel

    2011-11-01

    HTLV-1-associated myelopathy/tropical spastic paraparesis (HAM/TSP) is a chronic and progressive disorder caused by the human T-lymphotropic virus type 1 (HTLV-1). In HTLV-1 infection, a strong cytotoxic T cell (CTL) response is mounted against the immunodominant protein Tax. Previous studies carried out by our group reported that increased IFN-γ enzyme-linked immunospot (ELISPOT) responses against the region spanning amino acids 161 to 233 of the Tax protein were associated with HAM/TSP and increased HTLV-1 proviral load (PVL). An exploratory study was conducted on 16 subjects with HAM/TSP, 13 asymptomatic carriers (AC), and 10 HTLV-1-seronegative controls (SC) to map the HAM/TSP-associated CTL epitopes within Tax region 161-233. The PVL of the infected subjects was determined and the specific CTL response was evaluated with a 6-h incubation IFN-γ ELISPOT assay using peripheral blood mononuclear cells (PBMCs) stimulated with 16 individual overlapping peptides covering the Tax region 161-233. Other proinflammatory and Th1/Th2 cytokines were also quantified in the supernatants by a flow cytometry multiplex assay. In addition, a set of human leukocyte antigen (HLA) class I alleles that bind with high affinity to the CTL epitopes of interest was determined using computational tools. Univariate analyses identified an association between ELISPOT responses to two new CTL epitopes, Tax 173-185 and Tax 181-193, and the presence of HAM/TSP as well as an increased PVL. The HLA-A*6801 allele, which is predicted to bind to the Tax 181-193 peptide, was overpresented in the HAM/TSP patients tested.

  12. Preliminary study of the α ratio measurement, ratio of the neutron capture cross section to the fission one for 233U, on the PEREN platform. Development and study of the experimental setup

    International Nuclear Information System (INIS)

    Cognet, M.A.

    2007-12-01

    Producing nuclear energy in order to reduce anthropic CO 2 emission and to meet high energy demand, implies three conditions to the nuclear plants of the IV. generation: safety improvements, radioactive waste minimization, and fuel breeding for a sustainable use of the resources. The Thorium fuel cycle used in Molten Salt Reactors seems promising. Many numerical studies based on probabilistic codes are carried out in order to analyse the behaviour of such reactors. Nevertheless, one of the most important parameters is badly known: the alpha ratio of 233 U, ratio of the neutron capture cross section to fission one for 233 U. This key-parameter is necessary to calculate the breeding ratio and thus, the deployment capacities of those reactors. This Ph-D thesis was intended to prepare a precise measurement of the alpha ratio of 233 U between 1 eV and 10 keV. Preliminary measurements have been performed on the experimental platform PEREN. This experimental environment is composed of a lead slowing-down time spectrometer associated with an intense pulsed neutron generator. Capture and fission rates are measured thanks to eight scintillators with their photomultipliers, surrounding a fission chamber. A software analysis sets the coincidence rate between the scintillators. In order to understand perfectly the experimental setup, preliminary tests using a 235 U fission chamber have been done. This experiment resulted in a very low signal to background ratio (1 %). The background coming from the scintillators themselves seriously handicapped the measurement. Another series of experiment has been done with scintillators 5 times thinner. Nevertheless, the signal to background ratio should still be increased to measure the capture of 235 U. To make sure that the experimental setup has totally been understood, we made many comparisons between experimental results and simulations. Two simulation codes were mainly used: MCNP and GEANT4. We paid special attention to quantify the

  13. Distinct Patterns of Association of Variants at 11q23.3 Chromosomal Region with Coronary Artery Disease and Dyslipidemia in the Population of Andhra Pradesh, India.

    Directory of Open Access Journals (Sweden)

    Rayabarapu Pranav Chand

    Full Text Available In our attempt to comprehensively understand the nature of association of variants at 11q23.3 apolipoprotein gene cluster region, we genotyped a prioritized set of 96 informative SNPs using Fluidigm customized SNP genotyping platform in a sample of 508 coronary artery disease (CAD cases and 516 controls. We found 12 SNPs as significantly associated with CAD at P <0.05, albeit only four (rs2849165, rs17440396, rs6589566 and rs633389 of these remained significant after Benjamin Hochberg correction. Of the four, while rs6589566 confers risk to CAD, the other three SNPs reduce risk for the disease. Interaction of variants that belong to regulatory genes BUD13 and ZPR1 with APOA5-APOA4 intergenic variants is also observed to significantly increase the risk towards CAD. Further, ROC analysis of the risk scores of the 12 significant SNPs suggests that our study has substantial power to confer these genetic variants as predictors of risk for CAD, as illustrated by AUC (0.763; 95% CI: 0.729-0.798, p = <0.0001. On the other hand, the protective SNPs of CAD are associated with elevated Low Density Lipoprotein Cholesterol and Total Cholesterol levels, hence with dyslipidemia, in our sample of controls, which may suggest distinct effects of the variants at 11q23.3 chromosomal region towards CAD and dyslipidemia. It may be necessary to replicate these findings in the independent and ethnically heterogeneous Indian samples in order to establish this as an Indian pattern. However, only functional analysis of the significant variants identified in our study can provide more precise understanding of the mechanisms involved in the contrasting nature of their effects in manifesting dyslipidemia and CAD.

  14. Fission cross section ratios for sup 233,234,236 U relative to sup 235 U from 0. 5 to 400 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Lisowski, P.W.; Gavron, A.; Parker, W.E.; Balestrini, S.J. (Los Alamos National Lab., NM (USA)); Carlson, A.D.; Wasson, O.A. (National Inst. of Standards and Technology, Gaithersburg, MD (USA)); Hill, N.W. (Oak Ridge National Lab., TN (USA))

    1991-01-01

    Neutron-induced fission cross section ratios from 0.5 to 400 MeV for samples of {sup 233, 234, 236}U relative to {sup 235}U have been measured at the WNR neutron Source at Los Alamos. The fission reaction rate was determined using a fast parallel plate ionization chamber at a 20-m flight path. Cross sections over most the energy range were also extracted using the neutron fluence determined with three different proton telescope arrangements. Those data provided the shape of the {sup 235}U(n,f) cross section relative to the hydrogen scattering cross section. That shape was then normalized to the very accurately known value for {sup 235}U(n,f) at 14.1 MeV to allow us to obtain cross section section values from the ratio data and our values for {sup 235}U(n,f). 6 refs., 1 fig.

  15. An aerial radiological survey of the Sandia National Laboratories and surrounding area

    International Nuclear Information System (INIS)

    Riedhauser, S.R.

    1994-06-01

    A team from the Remote Sensing Laboratory conducted an aerial radiological survey of the area surrounding the Sandia National Laboratories and Kirtland Air Force Base in Albuquerque, New Mexico, during March and April 1993. The survey team measured the terrestrial gamma radiation at the site to determine the levels of natural and man-made radiation. This survey includes the areas covered by a previous survey in 1981. The results of the aerial survey show a background exposure rate which varies between 5 and 18 μR/h plus an approximate 6 μR/h contribution from cosmic rays. The major radioactive isotopes found in this survey were: potassium-40, thallium-208, bismuth-214, and actinium-228, which are all naturally-occurring isotopes, and cobalt-60, cesium-137, and excess amounts of thallium-208 and actinium-228, which are due to human actions in the survey area. In regions away from man-made activity, the exposure rates inferred from this survey's gamma ray measurements agree almost exactly with the exposure rates inferred from the 1981 survey. In addition to the aerial measurements, another survey team conducted in situ and soil sample radiation measurements at three sites within the survey perimeter. These ground-based measurements agree with the aerial measurements within ± 5%

  16. Preliminary radiological safety assessment for decommissioning of thoria dissolver of the 233U pilot plant, Trombay

    International Nuclear Information System (INIS)

    Priya, S.; Srinivasan, P.; Gopalakrishnan, R. K.

    2012-01-01

    The thoria dissolver, used for separation of 233 U from reactor-irradiated thorium metal and thorium oxide rods, is no longer operational. It was decided to carry out assessment of the radiological status of the dissolver cell for planning of the future decommissioning/dismantling operations. The dissolver interiors are expected to be contaminated with the dissolution remains of irradiated thorium oxide rods in addition to some of the partially dissolved thoria pellets. Hence, 220 Rn, a daughter product of 228 Th is of major radiological concern. Airborne activity of thoron daughters 212 Pb (Th-B) and 212 Bi (Th-C) was estimated by air sampling followed by high-resolution gamma spectrometry of filter papers. By measuring the full-energy peaks counts in the energy windows of 212 Pb, 212 Bi and 208 Tl, concentrations of thoron progeny in the sampled air were estimated by applying the respective intrinsic peak efficiency factors and suitable correction factors for the equilibration effects of 212 Pb and 212 Bi in the filter paper during the delay between sampling and counting. Then the thoron working level (TWL) was evaluated using the International Commission on Radiological Protection (ICRP) methodology. Finally, the potential effective dose to the workers, due to inhalation of thoron and its progeny during dismantling operations was assessed by using dose conversion factors recommended by ICRP. Analysis of filter papers showed a maximum airborne thoron progeny concentration of 30 TWLs inside the dissolver. (authors)

  17. Separation of the rare earths by high pressure liquid chromatography and the fission yield on sup(148m)Pm and sup(148g)Pm using thermal neutron induced fission of 233U and 239Pu

    International Nuclear Information System (INIS)

    Zwicky, H.U.

    1979-03-01

    This report is in two parts: in the first part, the method of high pressure liquid chromatography is described with particular reference to rare earth nuclei produced in nuclear reactions; in the second part, the results of a study of the fission yield of sup(148m)Pm and sup(148g)Pm from the thermal fission of 233 U and 239 Pu are presented. (G.T.H.)

  18. Synthesis of methyl ((chloro-2 ethyl)-3 nitroso-3 Ureido)-3 Didesoxy-2,3. alpha. -D-Arabino-hexopyrannoside labelled with carbon-14 or carbon-13 (CY 233 - SR 90008). Synthese du methyl ((chloro-2 ethyl)-3 nitroso-3 Ureido)-3 Didesoxy-2,3. alpha. -D-Arabino-hexopyrannoside marque au carbone-14 ou carbone-13 (CY 233 - SR 90008)

    Energy Technology Data Exchange (ETDEWEB)

    Sion, R.; Schumer, A.; Durme, E. van (Sanofi Recherche, Brussels (Belgium)); Gouyette, A. (Centre de Lutte Contre le Cancer Gustave-Roussy, 94 - Villejuif (France)); Geslin, M.; Fournier, J.P.; Roger, P. (Sanofi Recherche, Montrouge (France). Inst. Choay); Berger, Y. (Sanofi Recherche, Montpellier (France))

    1990-06-01

    CY 233 (Ecomustine or SR 90098) is a new antitumour nitrosourea: it is characterized by a 2-chloroethylnitrosourea substituent on a dideoxycarbohydrate. It has been labelled with {sup 14}C on (a) the carbonyl group of the urea in four stages starting with {sup 14}COCl{sub 2}, (b) the second carbon of the chloroethyl group in four stages starting with ({sup 14}C) ethanolamine, and (c) on the methyl group on the anomeric centre of the carbohydrate in three stages starting with {sup 14}CH{sub 3}OH. The final position was also labelled with {sup 13}C starting with {sup 13}CH{sub 3}OH. These differently labelled compounds are suitable for mechanistic studies of antitumour activity. (author).

  19. Reflector drums as control mechanism for craft thermionic reactors with constant emitter heating containing U-233 as fuel and beryllium as moderator

    International Nuclear Information System (INIS)

    Sahin, S.; Selvi, S.

    1980-01-01

    The suitability of borated reflector drums has been investigated and shown as a control mechanism for space craft thermionic reactors with constant emitter heating using U-233 as fuel and beryllium to be moderator, mainly due to their extremce compactness and their very soft neutron sepctrum. The achievable change in ksub(eff) allows long-term control operation with success. The use of reflector drums keeps the cone diameter and the mass of the radiation shield on minimum. The distortion of the emitter heating field remains under acceptable tolerances, mainly due to the enhanced neutron production at the outer core region and the remaining reflector part between the boron layer and the core. All neutron physics calculations have been carried out using the multigroup Ssub(N) methods. Three data groups for r-theta-calculations in S 4 -P 1 approximation (16 space angles) have been evaluated from a 123-energy-groups data library using transport theoretical methods. (orig.) [de

  20. Decree No85-968 of 27 August 1986 amending Article R.233-83 of the Labour Code and defining the health and safety conditions to be met by gamma industrial radiography equipment

    International Nuclear Information System (INIS)

    1986-01-01

    This Decree, amending Article R 233-83 of the Labour Code, applies to gamma-ray industrial radiography equipment whether mobile or fixed. It contains specific technical conditions concerning the equipment, in particular, safety devices located at the entry and exit points of the source's shield. The equipment must carry a notice mentioning''Radioactive''prepared by the constructor or the importer of the source, which explains the conditions for handling, setting-up, use and maintenance of the source, including the frequency of revisions. The Decree entered into force on 1 June 1986. (NEA) [fr

  1. Synthesis of methyl [(chloro-2 ethyl)-3 nitroso-3 Ureido]-3 Didesoxy-2,3 α-D-Arabino-hexopyrannoside labelled with carbon-14 or carbon-13 (CY 233 - SR 90008)

    International Nuclear Information System (INIS)

    Sion, R.; Schumer, A.; Durme, E. van; Gouyette, A.; Geslin, M.; Fournier, J.P.; Roger, P.

    1990-01-01

    CY 233 (Ecomustine or SR 90098) is a new antitumour nitrosourea: it is characterized by a 2-chloroethylnitrosourea substituent on a dideoxycarbohydrate. It has been labelled with 14 C on a) the carbonyl group of the urea in four stages starting with 14 COCl 2 , b) the second carbon of the chloroethyl group in four stages starting with [ 14 C] ethanolamine, and c) on the methyl group on the anomeric centre of the carbohydrate in three stages starting with 14 CH 3 OH. The final position was also labelled with 13 C starting with 13 CH 3 OH. These differently labelled compounds are suitable for mechanistic studies of antitumour activity. (author)

  2. Procedure for preparation of 3-fluor-D-alanine, 2-deutero-3-fluor-D-alanine and 2,3,3-trideutero-3-fluor-D-alanine and their salts

    International Nuclear Information System (INIS)

    Kollonitsch, J.; Kahan, F.M.

    1971-01-01

    Procedures for the preparation of 3-fluor-D-alanine, 2-deutero-3-fluor-D-alanine and 2,3,3-trideutero-3-fluor-D-alanine, and salts of these compounds, are described. These new compounds are useful antibacterial substances not only applicable in the disinfection of pharmaceutical, dental and medical equipment, but also in the treatment of diseases caused by bacteria, and may be administered orally. While 3-fluor-L-alanine metabolises rapidly with toxic results, 3-fluor-D-alanine is much more slowly broken down in vivo and is not harmful in normal doses. Further it has been found that deuteration gives new deutero-analogues which are less subject to metabolic breaking down and still retain the antibacterial strength of the original compound. The in vivo activity is thereby increased and maintained. (JIW)

  3. Complexes of groups 3,4, the lanthanides and the actinides containing neutral phophorus donor ligands

    International Nuclear Information System (INIS)

    Fryzuk, M.D.; Haddad, T.S.; Berg, D.J.

    1990-01-01

    Of relevance to this review are complexes of the early transition elements, in particular groups 3 and 4 and the lanthanides and actinides. In this review the authors have attempted to collect all the data up to the end of 1988 for complexed of groups 3 and 4, the lanthanides and the actinides that contain phosphorus donor ligands. The 1989s have seen a renaissance of the use of phosphine donors for the early d elements (groups 3 and 4) and the f elements. Neutral phosphorus donors are defined as primary (PH 2 R), secondary (PH 2 ) or tertiary phosphines (PR 3 ), including complexes of phosphine, PH 3 . Also reviewed are complexes of PF 3 and phosphites, P(OR) 3 . Specifically excluded are phosphido derivates, PR 2 . The ability of a neutral phosphorus donor to bind the metals of groups 3 and 4, the lanthanides and the actinides is now well established. While there are still no examples of lanthanum or actinium phosphine complexes, such derivatives should be accessible at least for lanthanum. series. However, there is no obvious chemical reason to suggest that such derivatives cannot be generated. The phosphine ligands that appear to generate the most stable phosphine-metal interaction are chelating phosphines such as dmpe, trmpe and trimpsi. In addition, the use of the chelate effect in conjunction with a hard ligand such as the amide in - N(SiMe 2 CH 2 PMe 2 ) 2 , or an alkoxide as found in - OC(BU t ) 2 CH 2 PMe 2 , also appears to be effective in anchoring the phosphine donor to the metal. The majority of low oxidation state derivatives of the group 4 elements are stabilized by phosphine donors in contrast with other parts of the transition series where one finds that classic π-acceptor-type ligands such as CO or RNC are utilized. 233 refs

  4. Brad Patterson, Tom Brooking, and Jim McAloon, Unpacking the Kist: The Scots in New Zealand. McGill-Queen's Studies in Ethnic History Series, No. 2.33. Montreal: McGill-Queen’s University Press, 2013. Pp. 412. ISBN 978-0-7735-4190-0. CAD $100.00.

    Directory of Open Access Journals (Sweden)

    Seán Gerard Brosnahan

    2014-11-01

    Full Text Available Brad Patterson, Tom Brooking, and Jim McAloon, Unpacking the Kist: The Scots in New Zealand. McGill-Queen's Studies in Ethnic History Series, No. 2.33. Montreal: McGill-Queen’s University Press, 2013. Pp. 412. ISBN 978-0-7735-4190-0. CAD $100.00.

  5. Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for {sup 233}U fuel production: 1. Concept of a stationary Tokamak as a neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Azizov, E. A.; Gladush, G. G., E-mail: gladush@triniti.ru; Dokuka, V. N.; Khayrutdinov, R. R. [State Research Center of the Russian Federation, Troitsk Institute for Innovation and Fusion Research (Russian Federation)

    2015-12-15

    On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of {sup 233}U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

  6. Complications associated with transobturator sling procedures: analysis of 233 consecutive cases with a 27 months follow-up

    Directory of Open Access Journals (Sweden)

    Dubuisson Jean-Bernard

    2009-09-01

    Full Text Available Abstract Backround The transobturator tape procedure (TOT is an effective surgical treatment of female stress urinary incontinence. However data concerning safety are rare, follow-up is often less than two years, and complications are probably underreported. The aim of this study was to describe early and late complications associated with TOT procedures and identify risk factors for erosions. Methods It was a 27 months follow-up of a cohort of 233 women who underwent TOT with three different types of slings (Aris®, Obtape®, TVT-O®. Follow-up information was available for 225 (96.6% women. Results There were few per operative complications. Forty-eight women (21.3% reported late complications including de novo or worsening of preexisting urgencies (10.2%, perineal pain (2.2%, de novo dyspareunia (9%, and vaginal erosion (7.6%. The risk of erosion significantly differed between the three types of slings and was 4%, 17% and 0% for Aris®, Obtape® and TVT-O® respectively (P = 0.001. The overall proportion of women satisfied by the procedure was 72.1%. The percentage of women satisfied was significantly lower in women who experienced erosion (29.4% compared to women who did not (78.4% (RR 0.14, 95% CI 0.05-0.38, P Conclusion Late post operative complications are relatively frequent after TOT and can impair patient's satisfaction. Women should be informed of these potential complications preoperatively and require careful follow-up after the procedure. Choice of the safest sling material is crucial as it is a risk factor for erosion.

  7. Complex three-way translocation involving MLL, ELL, RREB1, and CMAHP genes in an infant with acute myeloid leukemia and t(6;19;11)(p22.2;p13.1;q23.3)

    DEFF Research Database (Denmark)

    Tuborgh, A; Meyer, C; Marschalek, R

    2013-01-01

    until progression to acute myeloid leukemia, AML-M5. The leukemic cells harbored a novel apparent 3-way translocation t(6;19;11)(p22.2;p13.1;q23.3). We utilized advanced molecular cytogenetic methods including 24-color karyotyping, high-resolution array comparative genomic hybridization (aCGH) and DNA...... in the initial stages of disease before clear morphological signs of bone marrow involvement. The patient responded well to therapy and remains in remission>6 years from diagnosis. This apparent 3-way translocation is remarkable because of its rarity and presentation with myeloid sarcoma, and may, as more cases...

  8. Measurements of the prompt neutron spectra in 233U, 235U, 239Pu thermal neutron fission in the energy range of 0.01-5 MeV and in 252Cf spontaneous fission in the energy range of 0.01-10 MeV

    International Nuclear Information System (INIS)

    Starostov, B.I.; Semenov, A.F.; Nefedov, V.N.

    1978-01-01

    The measurement results on the prompt neutron spectra in 233 U, 235 U, 239 Pu thermal neutron fission in the energy range of 0.01-5 MeV and in 252 Cf spontaneous fission in the energy range of 0.01-10 MeV are presented. The time-of-flight method was used. The exceeding of the spectra over the Maxwell distributions is observed at E 252 Cf neutron fission spectra. The spectra analysis was performed after normalization of the spectra and corresponding Maxwell distributions for one and the same area. In the range of 0.05-0.22 MeV the yield of 235 U + nsub(t) fission neutrons is approximately 8 and approximately 15 % greater than the yield of 252 Cf and 239 Pu + nsub(t) fission neutrons, respectively. In the range of 0.3-1.2 MeV the yield of 235 U + nsub(t) fission neutrons is 8 % greater than the fission neutron yield in case of 239 Pu + nsub(t) fission. The 235 U + nsub(t) and 233 U + nsub(t) fission neutron spectra do not differ from one another in the 0.05-0.6 MeV range

  9. Radiometric determination of {sup 226}, {sup 228}Ac and {sup 40}K in fly ashes and building materials

    Energy Technology Data Exchange (ETDEWEB)

    Harangozo, M; Toelgyessy, J; Lesny, J; Cik, G [Slovak Technical Univ., Bratislava (Slovakia). Fac. of Chemical Technology, Dept. of Environmental Science

    1996-12-31

    In this paper the activities of radium-226, actinium-228 and potassium-40 in fly ashes and building materials of Slovakia were determined. Different origin of coals combusted results in significant differences in specific activities of radium-226 and activities-228 of measured fly-ashes and building materials. The knowledge of the specific activity of selected nuclides contained in fly-ashes is, therefore, very important and in specific cases can indicate the possibilities of their further technological use. (J.K.) 1 tab., 3 refs.

  10. A Radium-223 microgenerator from cyclotron-produced trace Actinium-227

    International Nuclear Information System (INIS)

    Abou, Diane S.; Pickett, Juile; Mattson, John E.; Thorek, Daniel L.J.

    2017-01-01

    The alpha particle emitter Radium-223 dichloride ("2"2"3RaCl_2) has recently been approved for treatment of late-stage bone metastatic prostate cancer. There is considerable interest in studying this new agent outside of the clinical setting, however the supply of "2"2"3Ra is limited and expensive. We have engineered a "2"2"3Ra microgenerator using traces of "2"2"7Ac previously generated from cyclotron-produced "2"2"5Ac. Radiochemically pure "2"2"3RaCl_2 was made, characterized, evaluated in vivo, and the source was recovered in high yield for regeneration of the microgenerator. - Highlights: • A "2"2"3Ra microgenerator was built using residual "2"2"7Ac from cyclotron-produced "2"2"5Ac. • Following "2"2"5Ac decay, the residual "2"2"7Ac was processed into pure "2"2"3Ra. • "2"2"7Ac and "2"2"7Th were recovered in high yield for a permanent supply of "2"2"3Ra. • Clinically supplied and generator-produced "2"2"3Ra have equivalent in vivo distribution. • Microdose column provides sufficient material for research use.

  11. Dissolution rates of unirradiated UO{sub 2}, UO{sub 2} doped with {sup 233}U, and spent fuel under normal atmospheric conditions and under reducing conditions using an isotope dilution method

    Energy Technology Data Exchange (ETDEWEB)

    Ollila, Kaija [VTT Processes, Helsinki (Finland); Albinsson, Yngve [Chalmers Univ. of Technology, Goeteborg (Sweden); Oversby, Virginia [VMO Konsult, Stockholm (Sweden); Cowper, Mark [AEA Technology, Harwell (United Kingdom)

    2003-10-01

    additional meaningful data. 8) A test procedure that used several short exposures of the sample to solution - the puff test procedure - gave results that showed very little recovery of the spike solution at the end of the tests. Only 10% of the {sup 235}U added as spike was recovered, indicating that 90% of the spike had precipitated onto the solid sample or the iron strip. 9) Tests of UO{sub 2} pellet materials containing {sup 233}U to provide an alpha decay activity similar to that expected for spent fuel 3000 and 10,000 years after disposal showed that the pellet materials behaved as expected under air atmosphere conditions, showing that the manufacturing method was successful. 10) Early testing of the {sup 233}U-doped materials under reducing conditions showed relatively rapid (30 minute) dissolution of small amounts of U at the start of the puff test procedure. Results of analyses of an acidified fraction of the same solutions after 1 or 2 weeks holding indicate that the solutions were inhomogeneous, indicating the presence of colloidal material or small grains of solid. 11) Samples from the {sup 233}U-doped tests initially indicated dissolution of solid during the first week of testing, with some indication of more rapid dissolution of the material with the higher doping. 12) The second cycle of testing of the {sup 233}U-doped materials also showed dissolution occurring during the dilution stages of the puff test. The subsequent week of testing also showed small amounts of further dissolution, with hints that the doped samples were dissolving faster than the undoped samples. 13) At the end of 2 weeks of cycle 2 the remaining solution and solid was transferred to a new reaction vessel, the solution was made up to original volume, and a new dose of spike was added. The results of analyses of [U] and isotopic composition show that the measured U is that expected from dilution of the original solution plus adding the spike. 14) Samples taken during 2 weeks of testing of

  12. Purification of cerium, neodymium and gadolinium for low background experiments

    Directory of Open Access Journals (Sweden)

    Boiko R.S.

    2014-01-01

    Full Text Available Cerium, neodymium and gadolinium contain double beta active isotopes. The most interesting are 150Nd and 160Gd (promising for 0ν2β search, 136Ce (2β+ candidate with one of the highest Q2β. The main problem of compounds containing lanthanide elements is their high radioactive contamination by uranium, radium, actinium and thorium. The new generation 2β experiments require development of methods for a deep purification of lanthanides from the radioactive elements. A combination of physical and chemical methods was applied to purify cerium, neodymium and gadolinium. Liquid-liquid extraction technique was used to remove traces of Th and U from neodymium, gadolinium and for purification of cerium from Th, U, Ra and K. Co-precipitation and recrystallization methods were utilized for further reduction of the impurities. The radioactive contamination of the samples before and after the purification was tested by using ultra-low-background HPGe gamma spectrometry. As a result of the purification procedure the radioactive contamination of gadolinium oxide (a similar purification efficiency was reached also with cerium and neodymium oxides was decreased from 0.12 Bq/kg to 0.007 Bq/kg in 228Th, from 0.04 Bq/kg to <0.006 Bq/kg in 226Ra, and from 0.9 Bq/kg to 0.04 Bq/kg in 40K. The purification methods are much less efficient for chemically very similar radioactive elements like actinium, lanthanum and lutetium.

  13. 'Masurium' and the 'early transuranium elements' or how discovery of nuclear fission was not clearly seen

    International Nuclear Information System (INIS)

    Keller, C.

    1988-01-01

    Fifty years after the discovery of fission, the scientific community is aware that this type of nuclear reaction could have been discovered more than a decade earlier. Noddack, Tacke and Berg announced in 1925 the discovery of elements Z = 43 (masurium) and rhenium (Z = 75), the first one could be detected only in U-bearing minerals. A recent re-examination by P.H.M. von Assche of the published data clearly showed that the original claim for element Z = 43 of the authors in 1925 was correct and, therefore, they detected not only element Z = 43 but also the first fission product. Because this discovery of element Z = 43 could not be repeated by other authors as that time, the scientific credibility of Noddack-Tacke was very low in order to give credit to her proposal that the 'early' transuranium elements by Enrico Fermi might also be fragments of known (lighter) elements. Enrico Fermi in 1934 obtained these 'early' (and as we today know: wrong) transuranium isotopes by irradiation of uranium with neutrons. A 'wrong' periodic system in the thirties which placed Th, Pa and U as 6d-elements and not as 5f-actinides chemically helped to consider these fission products as transuranium elements Z = 93/94. In 1937/38 I. Curie and P. Savitch discovered an 'actinium-nuclide' with 3,5 h half-life which, however, had properties similar to lanthanium and not to actinium, as they stated. (orig.) [de

  14. Shape of intrinsic alpha pulse height spectra in lanthanide halide scintillators

    Science.gov (United States)

    Wolszczak, W.; Dorenbos, P.

    2017-06-01

    Internal contamination with actinium-227 and its daughters is a serious drawback in low-background applications of lanthanide-based scintillators. In this work we showed the important role of nuclear γ de-excitations on the shape of the internal alpha spectrum measured in scintillators. We calculated with Bateman equations the activities of contamination isotopes and the time evolution of actinium-227 and its progenies. Next, we measured the intrinsic background spectra of LaBr3(Ce), LaBr3(Ce,Sr) and CeBr3 with a digital spectroscopy technique, and we analyzed them with a pulse shape discrimination method (PSD) and a time-amplitude analysis. Finally, we simulated the α background spectrum with Geant4 tool-kit, consequently taking into account complex α-γ-electron events, the α / β ratio dependence on the α energy, and the electron/γ nonproportionality. We found that α-γ mixed events have higher light yield than expected for alpha particles alone, which leads to overestimation of the α / β ratio when it is measured with internal 227Th and 223Ra isotopes. The time-amplitude analysis showed that the α peaks of 219Rn and 215Po in LaBr3(Ce) and LaBr3(Ce,Sr) are not symmetric. We compared the simulation results with the measured data and provided further evidence of the important role of mixed α-γ-electron events for understanding the shape of the internal α spectrum in scintillators.

  15. The production of lymphokines by primary alloreactive T-cell clones: a co-ordinate analysis of 233 clones in seven lymphokine assays.

    Science.gov (United States)

    Sanderson, C J; Strath, M; Warren, D J; O'Garra, A; Kirkwood, T B

    1985-01-01

    A total of 233 primary alloreactive T-cell clones have been tested for the production of interleukin-2 (IL-2), interleukin-3 (IL-3), immune(gamma) interferon (IFN) and granulocyte-macrophage colony-stimulating factor (CSF-2), B-cell growth factor I and II (BCGFI, BCGFII), and eosinophil differentiation factor (EDF). EDF was assayed by means of the eosinophil differentiation assay (EDA). Two principal correlations were observed: IL-3 was shown to be the major lymphokine detected in the bone marrow proliferation assay (BMPA) used to detect CSF-2, and there was a high correlation between the EDA and BCGFII. Subsequent work has suggested that this latter correlation is because a single factor is responsible for both activities. Apart from these two exceptions, and low level correlations probably due to the fact that different assays detect more than one lymphokine, there was no evidence for co-ordinate expression of lymphokines. There was a large variation in amounts of individual lymphokines produced. More clones produced multiple lymphokines than would be expected from independent control. Taken together, this pattern of regulation is consistent with the hypothesis that antigen stimulation of T cells results in the activation of all the lymphokine genes, but the amount of each produced is determined by secondary controlling mechanisms. PMID:3935571

  16. Actinides

    International Nuclear Information System (INIS)

    Martinot, L.; Fuger, J.

    1985-01-01

    The oxidation behavior of the actinides is explained on the basis of their electronic structure. The actinide elements, actinium, thorium, protactinium, uranium, neptunium, plutonium, americium, curium, berkelium, californium, einsteinium, fermium, mendelevium, nobelium, and laurencium are included. For all except the last three elements, the points of discussion are oxidation states, Gibbs energies and potentials, and potential diagram for the element in acid solution; and thermodynamic properties of these same elements are tabulated. References are cited following discussion of each element with a total of 97 references being cited. 13 tables

  17. Calculation of entropy of liquid metals using acoustic measurements in the framework of the rigid sphere model

    International Nuclear Information System (INIS)

    Tekuchev, V.V.; Barashkov, B.I.; Rygalov, L.N.; Dolzhikov, Yu.S.

    2001-01-01

    For the first time one obtained the polytherms of ultrasound velocity for liquid high-melting metals within wide temperature range. In terms of the rigid sphere model on the basis of the acoustic data one calculated the entropy values for 34 liquid metals at the melting point. The average discrepancy of the calculated values of entropy with the published one constitutes 8.2%. With increase of metal valency the error increases from 2.8 up to 13%. In case of francium, radium, promethium, actinium, hafnium, polonium, rhenium one obtained data for the first time [ru

  18. Status of the lanthanides and actinides in the periodic table

    International Nuclear Information System (INIS)

    Holden, N.E.

    1985-01-01

    In extended discussions and correspondence with Ekkehard Fluck, the author was made aware of a problem with the Periodic Table, i.e., which element should be shown in the main table as the representative of the lanthanide series and the actinide series. In earlier discussion, he came to the conclusion that lanthanum and actinium are not the elements which should appear, but rather lutetium and lawrencium are more appropriate for inclusion in their place. This paper will attempt to justify the reasons for the above conclusions. 4 refs

  19. Delayed Fission Gamma-ray Characteristics of Th-232 U-233 U-235 U-238 and Pu-239

    Energy Technology Data Exchange (ETDEWEB)

    Lane, Taylor [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Parma, Edward J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-08-01

    Delayed fission gamma-rays play an important role in determining the time dependent ioniz- ing dose for experiments in the central irradiation cavity of the Annular Core Research Reactor (ACRR). Delayed gamma-rays are produced from both fission product decay and from acti- vation of materials in the core, such as cladding and support structures. Knowing both the delayed gamma-ray emission rate and the time-dependent gamma-ray energy spectrum is nec- essary in order to properly determine the dose contributions from delayed fission gamma-rays. This information is especially important when attempting to deconvolute the time-dependent neutron, prompt gamma-ray, and delayed gamma-ray contribution to the response of a diamond photo-conducting diode (PCD) or fission chamber in time frames of milliseconds to seconds following a reactor pulse. This work focused on investigating delayed gamma-ray character- istics produced from fission products from thermal, fast, and high energy fission of Th-232, U-233, U-235, U-238, and Pu-239. This work uses a modified version of CINDER2008, a transmutation code developed at Los Alamos National Laboratory, to model time and energy dependent photon characteristics due to fission. This modified code adds the capability to track photon-induced transmutations, photo-fission, and the subsequent radiation caused by fission products due to photo-fission. The data is compared against previous work done with SNL- modified CINDER2008 [ 1 ] and experimental data [ 2 , 3 ] and other published literature, includ- ing ENDF/B-VII.1 [ 4 ]. The ability to produce a high-fidelity (7,428 group) energy-dependent photon fluence at various times post-fission can improve the delayed photon characterization for radiation effects tests at research reactors, as well as other applications.

  20. On the Use of 233U-236U Double-Spike for TIMS Measurements of Uranium Isotopes: A Simulation Study

    International Nuclear Information System (INIS)

    Williams, R W

    2004-01-01

    Synthetic ion beams with instantaneous and temporal characteristics appropriate to thermal ionization mass spectrometry (TIMS) were mathematically generated and analyzed to determine the effects of using a mixed 233 U- 236 U spike (double-spike) in the analysis of uranium isotopes. The instantaneous beam characteristics are the intensities (e.g., counts per second) modeled with a Poisson distribution plus a component of random noise that simulates the detection processes. Several beam intensity and mass fractionation vs. time functions were modeled to simulate a range of sample sizes and the commonly employed methods of data collection. These beam profiles were also generated with different noise levels, and signal-to-noise vs. analytical precision diagrams are presented. Modeling focused on natural uranium, where 238 U/ 235 U = 137.88, and on the ability of a given method to determine precisely and accurately small variations in this ratio. Practical limits on precision were determined to be 20-30 ppm, which is consistent with precision seen for other elements by state-of-the-art TIMS. The TIMS total evaporation method was compared directly with the double-spike method. While similar analytical precisions are obtained with either method, the double-spike method of correcting for analytical bias gives more accurate results. The results of a total evaporation analysis will deviate from true by more than the analytical precision if as little as 0.05% of the signal is not integrated, whereas the accuracy and precision of the double-spiked analyses are always linked

  1. Neutron-induced fission cross-section of 233U, 241Am and 243Am in the energy range 0.5 MeV ≤ En ≤ 20 MeV

    International Nuclear Information System (INIS)

    Belloni, F.; Milazzo, P.M.; Calviani, M.

    2011-01-01

    Neutron-induced fission cross-sections of 233 U, 241 Am and 243 Am relative to 235 U have been measured in a wide energy range at the neutron time of flight facility n-TOF in Geneva to address the present discrepancies in evaluated and experimental databases for reactions and isotopes relevant for transmutation and new generation fast reactors. A dedicated fast ionization chamber was used. Each isotope was mounted in a different cell of the modular detector. The measurements took advantage of the characteristics of the n-TOF installation. Its intrinsically low background, coupled to its high instantaneous neutron flux, results in high accuracy data. Its wide energy neutron spectrum helps to reduce systematic uncertainties due to energy-domain matching problems while the 185 m flight path and a 6 ns pulse width assure an excellent energy resolution. This paper presents results obtained between 500 keV and 20 MeV neutron energy. (authors)

  2. A study of uranium and thorium migration at the Koongarra uranium deposit with application to actinide transport from nuclear waste repositories

    International Nuclear Information System (INIS)

    Payne, T.E.

    1991-01-01

    One way to gain confidence in modelling possible radionuclide releases is to study natural systems which are similar to components of the multibarrier waste repository. Several such analogues are currently under study and these provide useful data about radionuclide behaviour in the natural environment. One such system is the Koongarra uranium deposit in the Northern Territory. In this dissertation, the migration of actinides, primarily uranium and thorium, has been studied as an analogue for the behaviour of transuranics in the far-field of a waste repository. The major findings of this study are: 1. the main process retarding uranium migration in the dispersion fan at Koongarra is sorption, which suppresses dissolved uranium concentrations well below solubility limits, with ferrihydrite being a major sorbing phase; 2. thorium is extremely immobile, with very low dissolved concentrations and corresponding high distribution ratios for 230 Th. Overall, it is estimated that colloids are relatively unimportant in Koongarra groundwater. Uranium migrates mostly as dissolved species, whereas thorium and actinium are mostly adsorbed to larger, relatively immobile particles and the stationary phase. However, of the small amount of 230 Th that passes through a 1μm filter, a significant proportion is associated with colloidal particles. Actinium appears to be slightly more mobile than thorium and is associated with colloids to a greater extent, although generally present in low concentrations. These results support the possibility of colloidal transport of trivalent and tetravalent actinides in the vicinity of a nuclear waste repository. 112 refs., 23 tabs., 32 figs

  3. Flexible synthesis of poison-frog alkaloids of the 5,8-disubstituted indolizidine-class. II: Synthesis of (--209B, (--231C, (--233D, (--235B", (--221I, and an epimer of 193E and pharmacological effects at neuronal nicotinic acetylcholine receptors

    Directory of Open Access Journals (Sweden)

    Garraffo H Martin

    2007-09-01

    Full Text Available Abstract Background The 5,8-disubstituted indolizidines constitute the largest class of poison-frog alkaloids. Some alkaloids have been shown to act as noncompetitive blockers at nicotinic acetylcholine receptors but the proposed structures and the biological activities of most of the 5,8-disubstituted indolizidines have not been determined because of limited supplies of the natural products. We have therefore conducted experiments to confirm proposed structures and determine biological activities using synthetic compounds. Recently, we reported that one of this class of alkaloids, (--235B', acts as a noncompetitive antagonist for α4β2 nicotinic receptors, and its sensitivity is comparable to that of the classical competitive antagonist for this receptor, dihydro-β-erythroidine. Results The enantioselective syntheses of (--209B, (--231C, (--233D, (--235B", (--221I, and what proved to be an epimer of natural 193E, starting from common chiral lactams have been achieved. When we performed electrophysiological recordings to examine the effects of the synthetic alkaloids on two major subtypes of nicotinic receptors (α4β2 and α7 expressed in Xenopus laevis oocytes, (--231C effectively blocked α4β2 receptor responses (IC50 value, 1.5 μM with a 7.0-fold higher potency than for blockade of α7 receptor responses. In contrast, synthetic (--221I and (--epi-193E were more potent in blocking α7 receptor responses (IC50 value, 4.4 μM and 9.1 μM, respectively than α4β2 receptor responses (5.3-fold and 2.0-fold, respectively. Conclusion We achieved the total synthesis of (--209B, (--231C, (--233D, (--235B", (--221I, and an epimer of 193E starting from common chiral lactams, and the absolute stereochemistry of natural (--233D was determined. Furthermore, the relative stereochemistry of (--231C and (--221I was also determined. The present asymmetric synthesis of the proposed structure for 193E revealed that the C-8 configuration of natural 193E

  4. An eighteen-membered macrocyclic ligand for actinium-225 targeted alpha therapy

    International Nuclear Information System (INIS)

    Thiele, Nikki A.; MacMillan, Samantha N.; Wilson, Justin J.; Rodriguez-Rodriguez, Cristina

    2017-01-01

    The 18-membered macrocycle H 2 macropa was investigated for 225 Ac chelation in targeted alpha therapy (TAT). Radiolabeling studies showed that macropa, at submicromolar concentration, complexed all 225 Ac (26 kBq) in 5 min at RT. [ 225 Ac(macropa)] + remained intact over 7 to 8 days when challenged with either excess La 3+ ions or human serum, and did not accumulate in any organ after 5 h in healthy mice. A bifunctional analogue, macropa-NCS, was conjugated to trastuzumab as well as to the prostate-specific membrane antigen-targeting compound RPS-070. Both constructs rapidly radiolabeled 225 Ac in just minutes at RT, and macropa-Tmab retained >99 % of its 225 Ac in human serum after 7 days. In LNCaP xenograft mice, 225 Ac-macropa-RPS-070 was selectively targeted to tumors and did not release free 225 Ac over 96 h. These findings establish macropa to be a highly promising ligand for 225 Ac chelation that will facilitate the clinical development of 225 Ac TAT for the treatment of soft-tissue metastases. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  5. An eighteen-membered macrocyclic ligand for actinium-225 targeted alpha therapy

    Energy Technology Data Exchange (ETDEWEB)

    Thiele, Nikki A.; MacMillan, Samantha N.; Wilson, Justin J. [Cornell Univ., Ithaca, NY (United States). Chemistry and Chemical Biology; Brown, Victoria; Jermilova, Una; Ramogida, Caterina F.; Robertson, Andrew K.H.; Schaffer, Paul; Radchenko, Valery [TRIUMF, Vancouver, BC (Canada). Life Science Div.; Kelly, James M.; Amor-Coarasa, Alejandro; Nikolopoulou, Anastasia; Ponnala, Shashikanth; Williams, Clarence Jr.; Babich, John W. [Radiology, Weill Cornell Medicine, New York, NY (United States); Rodriguez-Rodriguez, Cristina [British Columbia Univ., Vancouver, BC (Canada). Dept. of Physics and Astronomy and Centre for Comparative Medicine

    2017-11-13

    The 18-membered macrocycle H{sub 2}macropa was investigated for {sup 225}Ac chelation in targeted alpha therapy (TAT). Radiolabeling studies showed that macropa, at submicromolar concentration, complexed all {sup 225}Ac (26 kBq) in 5 min at RT. [{sup 225}Ac(macropa)]{sup +} remained intact over 7 to 8 days when challenged with either excess La{sup 3+} ions or human serum, and did not accumulate in any organ after 5 h in healthy mice. A bifunctional analogue, macropa-NCS, was conjugated to trastuzumab as well as to the prostate-specific membrane antigen-targeting compound RPS-070. Both constructs rapidly radiolabeled {sup 225}Ac in just minutes at RT, and macropa-Tmab retained >99 % of its {sup 225}Ac in human serum after 7 days. In LNCaP xenograft mice, {sup 225}Ac-macropa-RPS-070 was selectively targeted to tumors and did not release free {sup 225}Ac over 96 h. These findings establish macropa to be a highly promising ligand for {sup 225}Ac chelation that will facilitate the clinical development of {sup 225}Ac TAT for the treatment of soft-tissue metastases. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  6. Journal of Naval Science. Volume 2. Number 3. July 1976

    Science.gov (United States)

    1976-07-01

    supplementary food to the beakers containing stage VI nauplii: cyprids do not feed. The larvae in each beaker were con- fined within a close-fitting plastic...99-274% of Natural U) Uranium-234 (234U) (0-006% of Natural U) 2-48 X 10"’yrs Thorium-230 (230Th) Polonium -218(2I8Po) /through short lived...Natural U) Actinium-227 (--7Ac) FIG. 1. Nuclide chart. 4-51 X 10’Yrs 7 hours P 8 X 10’ Yrs 4 days Lead- 210 (210Pb) (22 yrs,/3"). 1-39 X 10

  7. Study of the desintegration of short-life fission products. Application to the mass distribution in the fission of 238U and 233U induced by 14MeV neutrons

    International Nuclear Information System (INIS)

    Cavallini, Pierre.

    1975-01-01

    Nuclear spectrometry of short-life fission products was investigated, together with direct applications to the study of mass and charge distribution in fission reactions. It is shown that, by choosing judiciously the target in which the fission product is created and owing to the differences in stabilities and evaporation temperatures of the compounds obtained, it is possible to separate some elements. For example, niobium was separated by heating after irradiation of a mixture of UC and RuCl 3 , and sublimation in a tube with temperature gradient. It was thus possible to study the 99 Nb isotope. Other classical chemical separation processes were used for yttrium and strontium. The half-lifes beta and gamma spectra, decay schemes of 93 Sr, 94 Y and 95 Y were studied. It was shown how to obtain mass distribution in fission using a nondestructive gamma analysis method. As an application, results obtained in the fission of 233 U and 238 U at 14 MeV are given [fr

  8. The fission cross sections of 230Th, 232Th, 233U, 234U, 236U, 238U, 237Np, 239Pu and 242Pu relative 235U at 14.74 MeV neutron energy

    International Nuclear Information System (INIS)

    Meadows, J.W.

    1986-12-01

    The measurement of the fission cross section ratios of nine isotopes relative to 235 U at an average neutron energy of 14.74 MeV is described with particular attention to the determination of corrections and to sources of error. The results are compared to ENDF/B-V and to other measurements of the past decade. The ratio of the neutron induced fission cross section for these isotopes to the fission cross section for 235 U are: 230 Th - 0.290 +- 1.9%; 232 Th - 0.191 +- 1.9%; 233 U - 1.132 +- 0.7%; 234 U - 0.998 +- 1.0%; 236 U - 0.791 +- 1.1%; 238 U - 0.587 +- 1.1%; 237 Np - 1.060 +- 1.4%; 239 Pu - 1.152 +- 1.1%; 242 Pu - 0.967 +- 1.0%. 40 refs., 11 tabs., 9 figs

  9. Purification of radium-226 for the manufacturing of actinium-225 in a cyclotron for alpha-immunotherapy; Radium-Aufreinigung zur Herstellung von Actinium-225 am Zyklotron fuer die Alpha-Immuntherapie

    Energy Technology Data Exchange (ETDEWEB)

    Marx, Sebastian Markus

    2014-09-23

    The thesis describes the development of methods for the purification of Ra-226. The objective was to obtain the radionuclide in the quality that is needed to be used as starting material in the manufacturing process for Ac-225 via proton-irradiated Ra-226. The radionuclide has been gained efficiently out of huge excesses of impurities. The high purity of the obtained radium affords its use as staring material in a pharmaceutical manufacturing process.

  10. 78 FR 64175 - Hashemite Kingdom of Jordan Loan Guarantees Issued Under the Further Continuing Appropriations...

    Science.gov (United States)

    2013-10-28

    ... procedures and standard terms and conditions applicable to loan guarantees to be issued for the benefit of... Definitions. 233.03 The Guarantee. 233.04 Guarantee eligibility. 233.05 Non-impairment of the Guarantee. 233... Default; Application for Compensation; payment. 233.09 No acceleration of Eligible Notes. [[Page 64176...

  11. Origin of elements of the Uranium-235 family observed in the Ellez river near the EL-4 experimental nuclear reactor in dismantling (Monts d'Arree- Finistere department); Origine des elements de la famille de l'uranium-235 observes dans la riviere Ellez a proximite du reacteur nucleaire experimental EL4 en cours de demantelement (Mont d'Arree - departement du Finistere). Resultats et premiers constats annee 2006

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    In a previous study which concerned the catchment basin of the harbour of Brest, the A.C.R.O. put in evidence a marking by artificial radioelements around the power plant of Brennilis which can be imputed without ambiguities to the nuclear installation. It also put in evidence abnormalities concerning the natural radioactivity which justifies this new study. In the area of the Monts d'Arree, actinium 227 ({sup 227}Ac), non born by its ascendents which are {sup 235}U and {sup 231}Pa is observed. This phenomenon is characterized by mass activities superior to these ones of {sup 235}U and able to reach these ones of {sup 238}U. Its presence corresponds with the drainage of the Ellez river since the former channel of radioactive effluents releases from the nuclear power plant EL-4 up to the reservoir Saint-Herblot situated 6 km downstream. The strongest values of radioactivity are registered near the disused power plant, at this place a relationship exists between the level of actinium 227 and this one of the artificial radioactivity as it exists a relationship with the decay products of radon exhaled from the subsoil ({sup 210}Pb). But its presence is not limited to a part of the Ellez river, it is equally observed in terrestrial medium, in places in priori not influenced by the direct liquid effluents of the power plant. This place is situated at more than 4 km and without any connection with the Ellez waters. At this stage of the study, it is not possible to answer with certainty the question of the origin of this phenomenon. A new reorientation is considered indispensable to clarify definitively the origin of this unknown phenomenon in the scientific publications and the environmental monitoring. (N.C.)

  12. origin of elements of the Uranium-235 family observed in the Ellez river near the EL-4 experimental nuclear reactor in dismantling (Monts d'Arree- Finistere department)

    International Nuclear Information System (INIS)

    2006-01-01

    In a previous study which concerned the catchment basin of the harbour of Brest, the A.C.R.O. put in evidence a marking by artificial radioelements around the power plant of Brennilis which can be imputed without ambiguities to the nuclear installation. It also put in evidence abnormalities concerning the natural radioactivity which justifies this new study. In the area of the Monts d'Arree, actinium 227 ( 227 Ac), non born by its ascendents which are 235 U and 231 Pa is observed. This phenomenon is characterized by mass activities superior to these ones of 235 U and able to reach these ones of 238 U. Its presence corresponds with the drainage of the Ellez river since the former channel of radioactive effluents releases from the nuclear power plant EL-4 up to the reservoir Saint-Herblot situated 6 km downstream. The strongest values of radioactivity are registered near the disused power plant, at this place a relationship exists between the level of actinium 227 and this one of the artificial radioactivity as it exists a relationship with the decay products of radon exhaled from the subsoil ( 210 Pb). But its presence is not limited to a part of the Ellez river, it is equally observed in terrestrial medium, in places in priori not influenced by the direct liquid effluents of the power plant. This place is situated at more than 4 km and without any connection with the Ellez waters. At this stage of the study, it is not possible to answer with certainty the question of the origin of this phenomenon. A new reorientation is considered indispensable to clarify definitively the origin of this unknown phenomenon in the scientific publications and the environmental monitoring. (N.C.)

  13. 233 - 238 Musa Kifi

    African Journals Online (AJOL)

    User

    2015-12-02

    Dec 2, 2015 ... Dry pods of Piliostigma reticulatum were obtained from. Damfamin –Tofa ... stable flame gas cooker for the respective treatment periods that has to do with ..... composition of Horse Eye bean (Mucuna urens) Asian Journal of ...

  14. 227 - 233_Yahaya_Antimicrobial

    African Journals Online (AJOL)

    pc

    PREVALENCE OF PATHOGENIC MICR ... hygiene, it serves as an abrasive that aids in remov the dental plaque and food ... plaque, dental caries and periodontal disease (Clar. 1924). ...... dental caries in children and adolescents. (Review).

  15. Colorectal cancer risk variants at 8q23.3 and 11q23.1 are associated with disease phenotype in APC mutation carriers.

    Science.gov (United States)

    Ghorbanoghli, Z; Nieuwenhuis, M H; Houwing-Duistermaat, J J; Jagmohan-Changur, S; Hes, F J; Tops, C M; Wagner, A; Aalfs, C M; Verhoef, S; Gómez García, E B; Sijmons, R H; Menko, F H; Letteboer, T G; Hoogerbrugge, N; van Wezel, T; Vasen, H F A; Wijnen, J T

    2016-10-01

    Familial adenomatous polyposis (FAP) is a dominantly inherited syndrome caused by germline mutations in the APC gene and characterized by the development of multiple colorectal adenomas and a high risk of developing colorectal cancer (CRC). The severity of polyposis is correlated with the site of the APC mutation. However, there is also phenotypic variability within families with the same underlying APC mutation, suggesting that additional factors influence the severity of polyposis. Genome-wide association studies identified several single nucleotide polymorphisms (SNPs) that are associated with CRC. We assessed whether these SNPs are associated with polyp multiplicity in proven APC mutation carriers. Sixteen CRC-associated SNPs were analysed in a cohort of 419 APC germline mutation carriers from 182 families. Clinical data were retrieved from the Dutch Polyposis Registry. Allele frequencies of the SNPs were compared for patients with APC genotype as a covariate. We found a trend of association of two of the tested SNPs with the ≥100 adenoma phenotype: the C alleles of rs16892766 at 8q23.3 (OR 1.71, 95 % CI 1.05-2.76, p = 0.03, dominant model) and rs3802842 at 11q23.1 (OR 1.51, 95 % CI 1.03-2.22, p = 0.04, dominant model). We identified two risk variants that are associated with a more severe phenotype in APC mutation carriers. These risk variants may partly explain the phenotypic variability in families with the same APC gene defect. Further studies with a larger sample size are recommended to evaluate and confirm the phenotypic effect of these SNPs in FAP.

  16. Radionuclide interactions with marine sediments

    International Nuclear Information System (INIS)

    Higgo, J.J.W.

    1987-09-01

    A critical review of the literature on the subject of the interactions of radionuclides with marine sediments has been carried out. On the basis of the information available, an attempt has been made to give ranges and 'best estimates' for the distribution ratios between seawater and sediments. These estimates have been based on an understanding of the sediment seawater system and the porewater chemistry and mineralogy. Field measurements, laboratory measurements and estimates based on stable-element geochemical data are all taken into account. Laboratory measurements include distribution-ratio and diffusion-coefficient determinations. The elements reviewed are carbon, chlorine, calcium, nickel, selenium, strontium, zirconium, niobium, technetium, tin, iodine, caesium, lead, radium, actinium, thorium, protactinium, uranium, neptunium, plutonium, americium and curium. (author)

  17. Calculation technique of free and impurity ion electronic structures

    International Nuclear Information System (INIS)

    Kulagin, N.A.; Sviridov, D.T.

    1986-01-01

    The monograph deals with calculation technique of free and impurity ion spectra with completed nl N -shell. The principles of the theory of irreducible tensor operators, genealogical coefficients, calculation technique of angular and radial parts of matrix elements operators are stated. The correlation accounting methods in free ions are considered in detail. The principles of the theory of crystal field and ligand field, the method of self-consistent field for impurity ions are reported. The technique efficiency based on example of lanthanum and actinium group ions is shown. Experimental data by nf N -ion spectra are given. The tables of angular coefficients, energy values of X-ray lines of rare earth ions and genealogical coefficients are given in the appendix

  18. JAEA thermodynamic database for performance assessment of geological disposal of high-level and TRU wastes. Refinement of thermodynamic data for trivalent actinoids and samarium

    International Nuclear Information System (INIS)

    Kitamura, Akira; Fujiwara, Kenso; Yui, Mikazu

    2010-01-01

    Within the scope of the JAEA thermodynamic database project for performance assessment of geological disposal of high-level radioactive and TRU wastes, the refinement of the thermodynamic data for the inorganic compounds and complexes of trivalent actinoids (actinium(III), plutonium(III), americium(III) and curium(III)) and samarium(III) was carried out. Refinement of thermodynamic data for these elements was based on the thermodynamic database for americium published by the Nuclear Energy Agency in the Organisation for Economic Co-operation and Development (OECD/NEA). Based on the similarity of chemical properties among trivalent actinoids and samarium, complementary thermodynamic data for their species expected under the geological disposal conditions were selected to complete the thermodynamic data set for the performance assessment of geological disposal of radioactive wastes. (author)

  19. Formerly utilized MED/AEC sites Remedial Action Program. Radiological survey of the St. Louis Airport Storage Site, St. Louis, Missouri. Final report. [U, Ra-bearing wastes stored in 1940-60's

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-09-01

    Results of two radiological surveys of the St. Louis-Lambert Airport property, formerly known as the Airport Storage Site, St. Louis, Missouri, are presented. Uranium- and radium-bearing waste materials were stored from the 1940's to the late 1960's in this area. The surveys included direct measurements of beta-gamma radiation; determination of uranium, actinium, and radium concentrations in soil samples and from bore holes; determination of radionuclide concentrations in groundwater and surface water; measurement of radon flux from the ground surface; and measurements of /sup 222/Rn in air near the site. Results indicate that some offsite drainage pathways are becoming contaminated, probably by runoff from the site; no migration of /sup 222/Rn from the site was observed.

  20. The effect of dissolved hydrogen on the dissolution of 233U doped UO2(s) high burn-up spent fuel and MOX fuel

    International Nuclear Information System (INIS)

    Carbol, P.; Spahiu, K.

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of 233 U doped UO 2 (s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H 2 pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H 2 pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO 2 , high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10 -6 /yr - 10 -8 /yr with a recommended value of 4x10 -7 /yr for dissolved hydrogen concentrations above 10 -3 M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO 2 and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB

  1. Non-invasive ventilation (NIV) in the clinical management of acute COPD in 233 UK hospitals: results from the RCP/BTS 2003 National COPD Audit.

    Science.gov (United States)

    Kaul, Sundeep; Pearson, Michael; Coutts, Ian; Lowe, Derek; Roberts, Michael

    2009-06-01

    Non-invasive ventilation (NIV) is a clinically proven, cost-effective intervention for acidotic exacerbations of COPD that is recommended by UK national guidelines. This study examines the extent to which these recommendations are being followed in the UK. Between August and October 2003 a national audit of COPD exacerbations was conducted by the Royal College of Physicians and the British Thoracic Society. 233 (94%) UK hospitals submitted data for 7,529 prospectively recruited acute COPD admissions, documenting process of care and outcomes from a retrospective case note audit. They also completed a resources and organisation of care proforma. Nineteen hospitals (8%) reported they did not offer NIV. There was no access to NIV in 92 (39%) intensive care units in 88 (36%), high-dependency units or on general wards of 85 (34%) hospitals. In 74 (30%) NIV was available on all 3 sites. A low pH (hospital mortality (26% v 14%) and at 90 days (37% v 24%) and longer hospital stays (median 9 v 7 days) than those not receiving NIV. Hospitals with least usage of NIV had similar mortality rates to those using NIV more often. A comprehensive NIV service is not available in many hospitals admitting patients with acute respiratory failure secondary to COPD. Access to acute NIV is inadequate and does not conform with NICE and BTS guidelines. These observational audit data do not demonstrate benefits of NIV on survival when compared to conventional management, contrary to results from randomised trials. Reasons for this are unclear but unmeasured confounding factors and poor patient selection for NIV are likely explanations.

  2. Scoping Review of the Zika Virus Literature.

    Directory of Open Access Journals (Sweden)

    Lisa A Waddell

    Full Text Available The global primary literature on Zika virus (ZIKV (n = 233 studies and reports, up to March 1, 2016 has been compiled using a scoping review methodology to systematically identify and characterise the literature underpinning this broad topic using methods that are documented, updateable and reproducible. Our results indicate that more than half the primary literature on ZIKV has been published since 2011. The articles mainly covered three topic categories: epidemiology of ZIKV (surveillance and outbreak investigations 56.6% (132/233, pathogenesis of ZIKV (case symptoms/ outcomes and diagnosis 38.2% (89/233 and ZIKV studies (molecular characterisation and in vitro evaluation of the virus 18.5% (43/233. There has been little reported in the primary literature on ZIKV vectors (12/233, surveillance for ZIKV (13/233, diagnostic tests (12/233 and transmission (10/233. Three papers reported on ZIKV prevention/control strategies, one investigated knowledge and attitudes of health professionals and two vector mapping studies were reported. The majority of studies used observational study designs, 89.7% (209/233, of which 62/233 were case studies or case series, while fewer (24/233 used experimental study designs. Several knowledge gaps were identified by this review with respect to ZIKV epidemiology, the importance of potential non-human primates and other hosts in the transmission cycle, the burden of disease in humans, and complications related to human infection with ZIKV. Historically there has been little research on ZIKV; however, given its current spread through Australasia and the Americas, research resources are now being allocated to close many of the knowledge gaps identified in this scoping review. Future updates of this project will probably demonstrate enhanced evidence and understanding of ZIKV and its impact on public health.

  3. Scoping Review of the Zika Virus Literature

    Science.gov (United States)

    2016-01-01

    The global primary literature on Zika virus (ZIKV) (n = 233 studies and reports, up to March 1, 2016) has been compiled using a scoping review methodology to systematically identify and characterise the literature underpinning this broad topic using methods that are documented, updateable and reproducible. Our results indicate that more than half the primary literature on ZIKV has been published since 2011. The articles mainly covered three topic categories: epidemiology of ZIKV (surveillance and outbreak investigations) 56.6% (132/233), pathogenesis of ZIKV (case symptoms/ outcomes and diagnosis) 38.2% (89/233) and ZIKV studies (molecular characterisation and in vitro evaluation of the virus) 18.5% (43/233). There has been little reported in the primary literature on ZIKV vectors (12/233), surveillance for ZIKV (13/233), diagnostic tests (12/233) and transmission (10/233). Three papers reported on ZIKV prevention/control strategies, one investigated knowledge and attitudes of health professionals and two vector mapping studies were reported. The majority of studies used observational study designs, 89.7% (209/233), of which 62/233 were case studies or case series, while fewer (24/233) used experimental study designs. Several knowledge gaps were identified by this review with respect to ZIKV epidemiology, the importance of potential non-human primates and other hosts in the transmission cycle, the burden of disease in humans, and complications related to human infection with ZIKV. Historically there has been little research on ZIKV; however, given its current spread through Australasia and the Americas, research resources are now being allocated to close many of the knowledge gaps identified in this scoping review. Future updates of this project will probably demonstrate enhanced evidence and understanding of ZIKV and its impact on public health. PMID:27244249

  4. Dissolved Concentration Limits of Radioactive Elements

    Energy Technology Data Exchange (ETDEWEB)

    Y. Chen; E.R. Thomas; F.J. Pearson; P.L. Cloke; T.L. Steinborn; P.V. Brady

    2003-06-20

    The purpose of this study is to evaluate dissolved concentration limits (also referred to as solubility limits) of radioactive elements under possible repository conditions, based on geochemical modeling calculations using geochemical modeling tools, thermodynamic databases, and measurements made in laboratory experiments and field work. The scope of this modeling activity is to predict dissolved concentrations or solubility limits for 14 radioactive elements (actinium, americium, carbon, cesium, iodine, lead, neptunium, plutonium, protactinium, radium, strontium, technetium, thorium, and uranium), which are important to calculated dose. Model outputs are mainly in the form of look-up tables plus one or more uncertainty terms. The rest are either in the form of distributions or single values. The results of this analysis are fundamental inputs for total system performance assessment to constrain the release of these elements from waste packages and the engineered barrier system. Solubilities of plutonium, neptunium, uranium, americium, actinium, thorium, protactinium, lead, and radium have been re-evaluated using the newly updated thermodynamic database (Data0.ymp.R2). For all of the actinides, identical modeling approaches and consistent environmental conditions were used to develop solubility models in this revision. These models cover broad ranges of environmental conditions so that they are applicable to both waste packages and the invert. Uncertainties from thermodynamic data, water chemistry, temperature variation, activity coefficients, and selection of solubility controlling phase have been quantified or otherwise addressed. Moreover, a new blended plutonium solubility model has been developed in this revision, which gives a mean solubility that is three orders of magnitude lower than the plutonium solubility model used for the Total System Performance Assessment for the Site Recommendation. Two alternative neptunium solubility models have also been

  5. Dissolved Concentration Limits of Radioactive Elements

    International Nuclear Information System (INIS)

    Y. Chen; E.R. Thomas; F.J. Pearson; P.L. Cloke; T.L. Steinborn; P.V. Brady

    2003-01-01

    The purpose of this study is to evaluate dissolved concentration limits (also referred to as solubility limits) of radioactive elements under possible repository conditions, based on geochemical modeling calculations using geochemical modeling tools, thermodynamic databases, and measurements made in laboratory experiments and field work. The scope of this modeling activity is to predict dissolved concentrations or solubility limits for 14 radioactive elements (actinium, americium, carbon, cesium, iodine, lead, neptunium, plutonium, protactinium, radium, strontium, technetium, thorium, and uranium), which are important to calculated dose. Model outputs are mainly in the form of look-up tables plus one or more uncertainty terms. The rest are either in the form of distributions or single values. The results of this analysis are fundamental inputs for total system performance assessment to constrain the release of these elements from waste packages and the engineered barrier system. Solubilities of plutonium, neptunium, uranium, americium, actinium, thorium, protactinium, lead, and radium have been re-evaluated using the newly updated thermodynamic database (Data0.ymp.R2). For all of the actinides, identical modeling approaches and consistent environmental conditions were used to develop solubility models in this revision. These models cover broad ranges of environmental conditions so that they are applicable to both waste packages and the invert. Uncertainties from thermodynamic data, water chemistry, temperature variation, activity coefficients, and selection of solubility controlling phase have been quantified or otherwise addressed. Moreover, a new blended plutonium solubility model has been developed in this revision, which gives a mean solubility that is three orders of magnitude lower than the plutonium solubility model used for the Total System Performance Assessment for the Site Recommendation. Two alternative neptunium solubility models have also been

  6. TMFunction data: 233 [TMFunction[Archive

    Lifescience Database Archive (English)

    Full Text Available Biol Chem. 2000 Jul 14;275(28):21017-24 mutagenesis ... affinity chromatography 1AJJ ... LDLR_HUMAN (P01130) Helix ... ligand binding site; surface exposed; acidic residue; conserved

  7. Agro. no 2 december 233

    African Journals Online (AJOL)

    plant species in the forest are mostly secondary colonizers climbers, shrubs and trees such as Culcasia .... some tree saplings are now emerging through this undergrowth, there is so far little sign of ..... issotis rotundifolia. (S ... Ficus exasperate.

  8. 43 CFR 23.3 - Definitions.

    Science.gov (United States)

    2010-10-01

    ... Lands: Interior Office of the Secretary of the Interior SURFACE EXPLORATION, MINING AND RECLAMATION OF... Leasing Act for Acquired Lands (30 U.S.C. 351-359); (b) Mining Supervisor means the Area Mining Supervisor... administrative jurisdiction of and responsibility for the land covered by a permit, lease, contract, application...

  9. 45 CFR 233.39 - Age.

    Science.gov (United States)

    2010-10-01

    ...; or age 18 if a full-time student in a secondary school, or in the equivalent level of vocational or... AABD with respect to the blind, any age; (iv) In APTD or AABD with respect to the disabled, 18 years of...

  10. Agro. no 2 december 233

    African Journals Online (AJOL)

    results of the logistic regression model showed that farm size, contacts with .... For this study, farmers with application rate of less than 25 percent of the .... in dry season amaranthus vegetable production in the study area were high cost of.

  11. 45 CFR 233.70 - Blindness.

    Science.gov (United States)

    2010-10-01

    ...). Such physician is responsible for making the agency's decision that the applicant or recipient does or... XVI of the Social Security Act must: (1) Contain a definition of blindness in terms of ophthalmic measurement. The following definition is recommended: An individual is considered blind if he has central...

  12. Agro. no 2 december 233

    African Journals Online (AJOL)

    cemented floor, the substrate raw materials mixed thoroughly and water content ... determined by drying in an oven at 80 C for 2 days according to the standard and method .... Mushroom biology; concise basics and current development.

  13. Energy Magazine. V. 23(3)

    International Nuclear Information System (INIS)

    1999-01-01

    The permanent secretariat of OLADE, with financial support of the European Commission, will carry out the National Energetic Information System , through which the tools will be developed for the elaboration and administration of the energy statistics in the member states of the organization. It is also included a vision on the action program in energy for the Caribbean to unify efforts and to coordinate actions leading the energy development and the diversification of the energy supply in order to satisfy their requirements. A section of the magazine, it is dedicated to the transformation experienced by the energy sector in Peru

  14. 45 CFR 233.40 - Residence.

    Science.gov (United States)

    2010-10-01

    .... For purposes of this section: (1) A resident of a State is one: (i) Who is living in the State... resident of the State in which he or she is living other than on a temporary basis. Residence may not depend upon the reason for which the individual entered the State, except insofar as it may bear upon...

  15. 40_ _230 - 233__Hassan _Structural

    African Journals Online (AJOL)

    User

    structure for glasses and many ceramics. The structure of a ... Quartz crystals have piezoelectric properties, meaning that they are ... glass, ceramics, building materials and for research in ... were irregular in shape and having porous texture. In.

  16. 40_ _230 - 233__Hassan _Structural

    African Journals Online (AJOL)

    User

    Kimura, 2005). Pure quartz (SiO2) has three common polymorphs: cristobolite, tridymite (high-temperature), and quartz. (low-temperature). The silica structure is the basic structure for glasses and many ceramics. The structure of a material may be divided into four levels: atomic structure, atomic arrangement, microstructure,.

  17. 233 Haematological Characteristics and Blood

    African Journals Online (AJOL)

    2008-12-02

    Dec 2, 2008 ... nitrogen of Sokoto Red kids fed varying levels of Fore-stomach digesta (FSD) replacing cowpea husk at 0, 10, 20 and 30 ... Haematological Characteristics and Blood Urea Nitrogen of Sokoto Red Goat Kids Fed …… 228. Formulation of ..... significant rise in Hb level in treatment D. This is an indication that ...

  18. 45 CFR 233.80 - Disability.

    Science.gov (United States)

    2010-10-01

    ... competent persons—not less than a physician and a social worker qualified by professional training and... or XVI of the Social Security Act must: (1) Contain a definition of permanently and totally disabled..., skills, and work experience, and the probable functioning of the individual in his particular situation...

  19. 12 CFR 233.2 - Definitions.

    Science.gov (United States)

    2010-01-01

    ... actual team that is a member of an amateur or professional sports organization (as those terms are... users to a computer server, including specifically a service or system that provides access to the....C. 3001 et seq.); (ii) 28 U.S.C. chapter 178 (professional and amateur sports protection); (iii) The...

  20. Power density effect on feasibility of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    Sidik, Permana; Takaki, Naoyuki; Sekimoto, Hiroshi

    2008-01-01

    Breeding is made possible by the high value of neutron regeneration ratio η for 233 U in thermal energy region. The reactor is fueled by 233 U-Th oxide and it has used the light water as moderator. Some characteristics such as spectrum, η value, criticality, breeding performance and number density are evaluated. Several power densities are evaluated in order to analyze its effect to the breeding performance. The η value of fissile 233 U obtains higher value than 2 which may satisfy the breeding capability especially for thermal reactor for all investigated MFR. The increasing enrichment and decreasing conversion ratio are more significant for MFR 233 U enrichment. Number density of 233 Pa decreases significantly with decreasing power density which leads the reactor has better breeding performance because lower capture rate of 233 Pa. (author)

  1. How fission was discovered

    International Nuclear Information System (INIS)

    Fluegge, S.

    1989-01-01

    After the great survey of neutron induced radioactivity by Fermi and co-workers, the laboratories in Paris and Berlin-Dahlen tried to disentangle the complex results found in uranium. At that time neutron sources were small, activities low, and equipment very simple. Chemistry beyond uranium still was unknown. Hahn and Meitner believed to have observed three transuranic isomeric chains, a doubtful result even then. Early in 1938, Curie and Savic in Paris found an activity interpreted to be actinium, and Hahn and Meitner another to be radium. Both interpretations seemed impossible from energy considerations. Hahn and Strassmann, therefore, continued this work and succeeded to separate the new activity from radium. There remained no doubt that a barium isotope had been produced, the uranium nucleus splitting in the yet-unknown process we now call fission

  2. New method for large scale production of medically applicable Actinium-225 and Radium-223

    International Nuclear Information System (INIS)

    Aliev, R.A.; Vasilyev, A.N.; Ostapenko, V.; Kalmykov, S.N.; Zhuikov, B.L.; Ermolaev, S.V.; Lapshina, E.V.

    2014-01-01

    Alpha-emitters ( 211 At, 212 Bi, 213 Bi, 223 Ra, 225 Ac) are promising for targeted radiotherapy of cancer. Only two alpha decays near a cell membrane result in 50% death of cancer cell and only a single decay inside the cell is required for this. 225 Ac may be used either directly or as a mother radionuclide in 213 Bi isotope generator. Production of 225 Ac is provided by three main suppliers - Institute for Transuranium Elements in Germany, Oak Ridge National Laboratory in USA and Institute of Physics and Power Engineering in Obninsk, Russia. The current worldwide production of 225 Ac is approximately 1.7 Ci per year that corresponds to only 100-200 patients that could be treated annually. The common approach for 225 Ac production is separation from mother 229 Th or irradiation of 226 Ra with protons in a cyclotron. Both the methods have some practical limitations to be applied routinely. 225 Ac can be also produced by irradiation of natural thorium with medium energy protons . Cumulative cross sections of 225 Ac, 227 Ac, 227 Th, 228 Th formations have been obtained recently. Thorium targets (1-9 g) were irradiated by 114-91 MeV proton beam (1-50 μA) at INR linear accelerator. After dissolution in 8 M HNO 3 + 0.004 M HF thorium was removed by double LLX by HDEHP in toluene (1:1). Ac and REE were pre-concentrated and separated from Ra and most fission products by DGA-Resin (Triskem). After washing out by 0.01 M HNO 3 Ac was separated from REE by TRU Resin (Triskem) in 3 M HNO 3 media. About 6 mCi 225 Ac were separated in hot cell with chemical yield 85%. The method may be upscaled for production of Ci amounts of the radionuclide. The main impurity is 227 Ac (0.1% at the EOB) but it does not hinder 225 Ac from being used for medical 225 Ac/ 213 Bi generators. (author)

  3. The marine geochemistry of actinium-227: Evidence for its migration through sediment pore water

    International Nuclear Information System (INIS)

    Nozaki, Yoshiyuki; Yamada, Masatoshi; Nikaido, Hirofumi

    1990-01-01

    227 Ac with a half life of 21.8 years has a potential utility as a tracer of deep water circulation and mixing studies on time scales less than 100 years. Here the authors present the first measurement of 227 Ac profile in the pore water of Northwest Pacific deep-sea sediment and in the ∼10,000 m long water column of Izu-Ogasawara Trench. The results clearly show that 227 Ac is supplied from the sediment to the overlying water through migration in the pore water. The model calculation indicates that the molecular diffusion alone through sediment porewater can support only a half of the standing crop of excess 227 Ac in the water column and the enhanced supply of 227 Ac by particle mixing is necessary to account for the remainder. Thus, bioturbation in the deep sea plays an important role in controlling the flux of some short-lived radionuclides such as 227 Ac and 228 Ra across the sediment-water interface

  4. 78 FR 45557 - Gulf of Mexico, Outer Continental Shelf (OCS), Western Planning Area (WPA) Oil and Gas Lease Sale...

    Science.gov (United States)

    2013-07-29

    ... Supplemental Environmental Impact Statement (WPA 233/ CPA 231 Supplemental EIS). WPA Lease Sale 233, scheduled... EIS evaluated the environmental and socioeconomic impacts for WPA Lease Sale 233. SUPPLEMENTARY... DEPARTMENT OF THE INTERIOR Bureau of Ocean Energy Management [MMAA104000] Gulf of Mexico, Outer...

  5. 77 FR 11437 - Inspection Service Authority; Seizure and Forfeiture

    Science.gov (United States)

    2012-02-27

    ... POSTAL SERVICE 39 CFR Part 233 Inspection Service Authority; Seizure and Forfeiture AGENCY: Postal... Service's rules and regulations regarding the seizure and forfeiture of property into three sections, 39.... The proposed revision consolidates sections 233.8 and 233.9, and treats seizures involving personal...

  6. Journal of Business Research: Contact

    African Journals Online (AJOL)

    Principal Contact. Goski Alabi Mrs Institute of Professional Studies (IPS) P. 0 Box 149 Institute Of Professional Studies (IPS) Legon, Accra Ghana Phone: +233 24 64 52798. Fax: +233 21 513539. Email: goskia@yahoo.com. Support Contact. Anthony Afeadie. Phone: +233 21 500171. Email: ipsjournal@yahoo.com.

  7. Consultants' Meeting on Review Benchmarking of Nuclear Data for the Th/U Fuel Cycle. Summary Report

    International Nuclear Information System (INIS)

    Capote Noy, R.

    2011-02-01

    A summary is given of the Consultants' Meeting (CM) on Review and Benchmarking of Nuclear Data for the Th/U Fuel Cycle. An IAEA Coordinated Research Project (CRP) on 'Nuclear Data for Th/U Fuel Cycle' was concluded in 2005. The CRP activities resulted in new evaluated nuclear data files for 232 Th, 231 , 233 Pa (later adopted for the ENDF/B-VII.0 library) and improvements to existing evaluations for 232 , 233 , 234 , 236 U. Available nuclear data evaluations for 230 - 232 Th, 231,233 Pa and 232 , 233 , 234 U were reviewed including ROSFOND2010, CENDL-3.1, JENDL-4, JEFF-3.1.1, MINSKACT, and ENDF/B-VII.0 libraries. Benchmark results of available evaluations for 232 Th and 233 U were also discussed. (author)

  8. Process for radioisotope recovery and system for implementing same

    Science.gov (United States)

    Meikrantz, David H [Idaho Falls, ID; Todd, Terry A [Aberdeen, ID; Tranter, Troy J [Idaho Falls, ID; Horwitz, E Philip [Naperville, IL

    2009-10-06

    A method of recovering daughter isotopes from a radioisotope mixture. The method comprises providing a radioisotope mixture solution comprising at least one parent isotope. The at least one parent isotope is extracted into an organic phase, which comprises an extractant and a solvent. The organic phase is substantially continuously contacted with an aqueous phase to extract at least one daughter isotope into the aqueous phase. The aqueous phase is separated from the organic phase, such as by using an annular centrifugal contactor. The at least one daughter isotope is purified from the aqueous phase, such as by ion exchange chromatography or extraction chromatography. The at least one daughter isotope may include actinium-225, radium-225, bismuth-213, or mixtures thereof. A liquid-liquid extraction system for recovering at least one daughter isotope from a source material is also disclosed.

  9. The fission cross sections of /sup 230/Th, /sup 232/Th, /sup 233/U, /sup 234/U, /sup 236/U, /sup 238/U, /sup 237/Np, /sup 239/Pu and /sup 242/Pu relative /sup 235/U at 14. 74 MeV neutron energy

    Energy Technology Data Exchange (ETDEWEB)

    Meadows, J.W.

    1986-12-01

    The measurement of the fission cross section ratios of nine isotopes relative to /sup 235/U at an average neutron energy of 14.74 MeV is described with particular attention to the determination of corrections and to sources of error. The results are compared to ENDF/B-V and to other measurements of the past decade. The ratio of the neutron induced fission cross section for these isotopes to the fission cross section for /sup 235/U are: /sup 230/Th - 0.290 +- 1.9%; /sup 232/Th - 0.191 +- 1.9%; /sup 233/U - 1.132 +- 0.7%; /sup 234/U - 0.998 +- 1.0%; /sup 236/U - 0.791 +- 1.1%; /sup 238/U - 0.587 +- 1.1%; /sup 237/Np - 1.060 +- 1.4%; /sup 239/Pu - 1.152 +- 1.1%; /sup 242/Pu - 0.967 +- 1.0%. 40 refs., 11 tabs., 9 figs.

  10. ß-Lysine discrimination by lysyl-tRNA synthetase

    DEFF Research Database (Denmark)

    Gilreath, Marla S; Roy, Hervé; Bullwinkle, Tammy J

    2011-01-01

    guided by the PoxA structure. A233S LysRS behaved as wild type with a-lysine, while the G469A and A233S/G469A variants decreased stable a-lysyl-adenylate formation. A233S LysRS recognized ß-lysine better than wildtype, suggesting a role for this residue in discriminating a- and ß-amino acids. Both...

  11. Fast Thorium Molten Salt Reactors Started with Plutonium

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Heuer, D.; Le Brun, C.; Brissot, R.; Liatard, E.; Meplan, O.; Nuttin, A.; Mathieu, L.

    2006-01-01

    One of the pending questions concerning Molten Salt Reactors based on the 232 Th/ 233 U fuel cycle is the supply of the fissile matter, and as a consequence the deployment possibilities of a fleet of Molten Salt Reactors, since 233 U does not exist on earth and is not yet produced in the current operating reactors. A solution may consist in producing 233 U in special devices containing Thorium, in Pressurized Water or Fast Neutrons Reactors. Two alternatives to produce 233 U are examined here: directly in standard Molten Salt Reactors started with Plutonium as fissile matter and then operated in the Th/ 233 U cycle; or in dedicated Molten Salt Reactors started and fed with Plutonium as fissile matter and Thorium as fertile matter. The idea is to design a critical reactor able to burn the Plutonium and the minor actinides presently produced in PWRs, and consequently to convert this Plutonium into 233 U. A particular reactor configuration is used, called 'unique channel' configuration in which there is no moderator in the core, leading to a quasi fast neutron spectrum, allowing Plutonium to be used as fissile matter. The conversion capacities of such Molten Salt Reactors are excellent. For Molten Salt Reactors only started with Plutonium, the assets of the Thorium fuel cycle turn out to be quickly recovered and the reactor's characteristics turn out to be equivalent to Molten Salt Reactors operated with 233 U only. Using a combination of Molten Salt Reactors started or operated with Plutonium and of Molten Salt Reactors started with 233 U, the deployment capabilities of these reactors fully satisfy the condition of sustainability. (authors)

  12. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  13. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  14. Remarks on the thorium cycle

    International Nuclear Information System (INIS)

    Teller, E.

    1978-01-01

    The use of thorium and neutrons to make 233 U would provide energy for many thousands of years. Thorium is more abundant than uranium and 233 U is the best fissile material for thermal neutron reactors. Four approaches to the use of thorium are worth developing: heavy water moderated reactors with conversion ratios greater than 0.9, such as modified CANDU with lower cost of separating D 2 O and 235 U; molten salt breeder reactors, from which fission products and excess fuel may be continuously removed; fusion-fission hybrids that produce adequate tritium and excess neutrons for sustenance and 233 U production in a subcritical thorium 233 U blanket; and by fission-initiated thermo-nuclear explosions in cavities in salt beds one mile below the earth's surface, yielding 233 U from the excess neutrons and thorium and decontaminated steam for power production. (author)

  15. Publications | Page 233 | IDRC - International Development ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    ... most advanced Internet censorship and surveillance regime in cyberspace. ... The Evaluation Unit, in partnership with Universalia Management Group, has ... We have developed an organizational performance assessment framework, ...

  16. 34 CFR 668.233 - Student eligibility.

    Science.gov (United States)

    2010-07-01

    ... high school diploma, a recognized equivalent of a high school diploma, or have passed an ability to... evaluation and diagnosis of an intellectual disability by a psychologist or other qualified professional; or...

  17. 48 CFR 233.215 - Contract clause.

    Science.gov (United States)

    2010-10-01

    ... acquisition is for— (i) Aircraft (ii) Spacecraft and launch vehicles (iii) Naval vessels (iv) Missile systems (v) Tracked combat vehicles (vi) Related electronic systems; (2) The contracting officer determines...

  18. 46_231 - 233_BIO 054 Mohammed

    African Journals Online (AJOL)

    userpc

    Hepatotoxicity from these drugs have been linked to in part, the alterations in ... these drugs even in concurrent use. This study was ... being a clinical situation in cancer and malaria .... Effect of Oral Administration of Ethanolic Leaf. Extract of ...

  19. 40 CFR 233.21 - General permits.

    Science.gov (United States)

    2010-07-01

    ... ensure compliance with existing permit conditions an any reporting monitoring, or prenotification... apply for an individual permit. This discretionary authority will be based on concerns for the aquatic environment including compliance with paragraph (b) of this section and the 404(b)(1) Guidelines (40 CFR part...

  20. 40 CFR 233.23 - Permit conditions.

    Science.gov (United States)

    2010-07-01

    .... (7) Monitoring, reporting and recordkeeping requirements as needed to safeguard the aquatic environment. (Such requirements will be determined on a case-by-case basis, but at a minimum shall include monitoring and reporting of any expected leachates, reporting of noncompliance, planned changes or transfer...

  1. Publications | Page 233 | IDRC - International Development ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    It is the middle of one of the worst droughts to hit East Africa in. ... risk for obstructed labour and its adverse outcomes in south-western Uganda (open access) ... for state recognition of cultural difference and for the granting of collective rights.

  2. Potential for the near-term use of the thorium cycle in a sustainable way

    International Nuclear Information System (INIS)

    Wider, H.; Tucek, K.; Carlsson, J.

    2007-01-01

    Nuclear sustainability is generally believed to be only reachable through the building of many fast breeder reactors. This paper shows that there is another possibility by using existing reactors that are either thermal breeders or have at least a high conversion ratio and considerably smaller critical masses than fast systems. Earlier it was believed that thermal molten salt breeders could eventually use the thorium / 233U cycle, which doesn't generate minor actinides and is therefore a cleaner fuel cycle. In the meantime, it has become rather clear that CANDU reactors that use heavy water cooling can also be self-breeders. The CANDU reactors could generate themselves 233U in thorium targets and could become selfsustaining after 12 years. However, additional 233U could also be generated in LWRs and fast reactors. It is shown that this generation of 233U will allow a faster large-term nuclear expansion than fast reactors alone. There could actually be a synergy between thermal and fast breeders if the latter are run with Pu/Minor Actinides/Th fuel, which burns the minor actinides and generates sizeable amounts of 233U. The main problem is still the necessary reprocessing on which India is working and intends to have in 10 years a large scale reprocessing facility available. However, there is at least an existing method for removing the 233U by the fluoride volatility method and to further use it in CANDUs. For the preparation of the use of 233U, we should attempt to run thorium subassemblies in CANDUs, LWRs, and fast reactors. Besides breeding 233U or at least having a high conversion ratio, CANDUs have the further advantage that they don't need a pressure vessel and therefore could be built in large numbers faster than LWRs. (author)

  3. The chemistry of the actinide elements. Volume I

    International Nuclear Information System (INIS)

    Katz, J.J.; Seaborg, G.T.; Morss, L.R.

    1986-01-01

    The Chemistry of the Actinide Elements is a comprehensive, contemporary and authoritative exposition of the chemistry and related properties of the 5f series of elements: actinium, thorium, protactinium, uranium and the first eleven. This second edition has been completely restructured and rewritten to incorporate current research in all areas of actinide chemistry and chemical physics. The descriptions of each element include accounts of their history, separation, metallurgy, solid-state chemistry, solution chemistry, thermo-dynamics and kinetics. Additionally, separate chapters on spectroscopy, magnetochemistry, thermodynamics, solids, the metallic state, complex ions and organometallic compounds emphasize the comparative chemistry and unique properties of the actinide series of elements. Comprehensive lists of properties of all actinide compounds and ions in solution are given, and there are special sections on such topics as biochemistry, superconductivity, radioisotope safety, and waste management, as well as discussion of the transactinides and future elements

  4. A neutron source of variable fluence

    International Nuclear Information System (INIS)

    Brachet, Guy; Demichel, Pascal; Prigent, Yvon; Riche, J.C.

    1975-01-01

    The invention concerns a variable fluence neutron source, like those that use in the known way a reaction between a radioactive emitter and a target, particularly of type (α,n). The emitter being in powder form lies in a carrier fluid forming the target, inside a closed containment. Facilities are provided to cause the fluidisation of the emitter by the carrier fluid in the containment. The fluidisation of the emitting powder is carried out by a booster with blades, actuated from outside by a magnetic coupling. The powder emitter is a α emitter selected in the group of curium, plutonium, thorium, actinium and americium oxides and the target fluid is formed of compounds of light elements selected from the group of beryllium, boron, fluorine and oxygen 18. The target fluid is a gas used under pressure or H 2 O water highly enriched in oxygen 18 [fr

  5. Potential use of thorium through fusion breeders in the Indian context

    International Nuclear Information System (INIS)

    Srinivasan, M.; Basu, T.K.; Subba Rao, K.

    1991-01-01

    The Indian Nuclear Programme is based on a three stage strategy: the first stage of about 10 GWe comprises of natural uranium fuelled Pressurised Heavy Water Reactors (PHWRs); the second stage would consist of Liquid Metal Cooled Fast Breeder Reactors (LMFBRs) to be fuelled with plutonium generated in the first stage PHWRs and the third stage is envisaged to be based on advanced converters/breeders operating on the Th/U-233 cycle. It has generally been assumed that the initial inventory of U-233 for the third stage reactors would be generated in the blankets of LMFBRs containing thorium. But the success of this strategy depends crucially on the attainment of LMFBR doubling times as short as 14 years. The progress registered in recent years in the magnetic confinement of fusion plasmas has opened up the prospects of developing Fusion Breeders for the direct conversion of fertile 232 Th into fissile 233 U using the 14 MeV neutron released in the (D-T) fusion reaction. A detailed study of the dependence of the 233 U production characteristics as well as energy cost of fissile fuel production of such systems on parameters such as plasma energy gain Q, blanket neutron multiplication has been carried out. The growth rate dynamics of the symbiotic combination of 233 U generating fusion breeders with PHWRs operating on the Th/U-233 cycle in the so called near-breeder regime has been examined. 95% of the energy generated by PHWRs operating with Th/ 233 U fuel would arise from thorium consumption rather than fission of the initially loaded 233 U. A few sub-engineering breakeven fusion breeders producing U-233 at an energy cost well under 200 MeV per atom are adequate to give the requisite nuclear capacity growth rates in conjunction with such near breeder PHWRs. This corresponds to only a 5% diversion of the grid electrical power for the operation of such fusion breeders. In summary the symbiotic combination of a few fusion breeders with a number of PHWRs gives fresh hopes

  6. Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    Highlights: • This work analyzes important trends of the running-in phase of a thorium breeder PBR. • Depletion equations are solved for important actinides and a fission product pair. • Breeding U-233 is achieved in 7 years by cleverly adjusting the feed fuel enrichment. • A safety analysis shows the thorium PBR is passively safe during the running-in phase. - Abstract: The present work investigates the running-in phase of a 100 MW th Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated. For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme. Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300 days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7 years after start of reactor operation. The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the

  7. Purification of radium-226 for the manufacturing of actinium-225 in a cyclotron for alpha-immunotherapy

    International Nuclear Information System (INIS)

    Marx, Sebastian Markus

    2014-01-01

    The thesis describes the development of methods for the purification of Ra-226. The objective was to obtain the radionuclide in the quality that is needed to be used as starting material in the manufacturing process for Ac-225 via proton-irradiated Ra-226. The radionuclide has been gained efficiently out of huge excesses of impurities. The high purity of the obtained radium affords its use as staring material in a pharmaceutical manufacturing process.

  8. Linear free energy relationship applied to trivalent cations with lanthanum and actinium oxide and hydroxide structure

    International Nuclear Information System (INIS)

    Ragavan, Anpalaki J.

    2006-01-01

    Linear free energy relationships for trivalent cations with crystalline M 2 O 3 and, M(OH) 3 phases of lanthanides and actinides were developed from known thermodynamic properties of the aqueous trivalent cations, modifying the Sverjensky and Molling equation. The linear free energy relationship for trivalent cations is as ΔG f,MvX 0 =a MvX ΔG n,M 3+ 0 +b MvX +β MvX r M 3+ , where the coefficients a MvX , b MvX , and β MvX characterize a particular structural family of MvX, r M 3+ is the ionic radius of M 3+ cation, ΔG f,MvX 0 is the standard Gibbs free energy of formation of MvX and ΔG n,M 3+ 0 is the standard non-solvation free energy of the cation. The coefficients for the oxide family are: a MvX =0.2705, b MvX =-1984.75 (kJ/mol), and β MvX =197.24 (kJ/molnm). The coefficients for the hydroxide family are: a MvX =0.1587, b MvX =-1474.09 (kJ/mol), and β MvX =791.70 (kJ/molnm).

  9. Development of ion beam sputtering techniques for actinide target preparation

    International Nuclear Information System (INIS)

    Aaron, W.S.; Zevenbergen, L.A.; Adair, H.L.

    1985-01-01

    Ion beam sputtering is a routine method for the preparation of thin films used as targets because it allows the use of minimum quantity of starting material, and losses are much lower than most other vacuum deposition techniques. Work is underway in the Isotope Research Materials Laboratory (IRML) at ORNL to develop the techniques that will make the preparation of actinide targets up to 100 μg/cm 2 by ion beam sputtering a routinely available service from IRML. The preparation of the actinide material in a form suitable for sputtering is a key to this technique, as is designing a sputtering system that allows the flexibility required for custom-ordered target production. At present, development work is being conducted on low-activity in a bench-top system. The system will then be installed in a hood or glove box approved for radioactive materials handling where processing of radium, actinium, and plutonium isotopes among others will be performed. (orig.)

  10. Development of ion beam sputtering techniques for actinide target preparation

    Science.gov (United States)

    Aaron, W. S.; Zevenbergen, L. A.; Adair, H. L.

    1985-06-01

    Ion beam sputtering is a routine method for the preparation of thin films used as targets because it allows the use of a minimum quantity of starting material, and losses are much lower than most other vacuum deposition techniques. Work is underway in the Isotope Research Materials Laboratory (IRML) at ORNL to develop the techniques that will make the preparation of actinide targets up to 100 μg/cm 2 by ion beam sputtering a routinely available service from IRML. The preparation of the actinide material in a form suitable for sputtering is a key to this technique, as is designing a sputtering system that allows the flexibility required for custom-ordered target production. At present, development work is being conducted on low-activity actinides in a bench-top system. The system will then be installed in a hood or glove box approved for radioactive materials handling where processing of radium, actinium, and plutonium isotopes among others will be performed.

  11. Proposed training program for construction personnel involved in remedial action work at sites contaminated by naturally occurring radionuclides

    International Nuclear Information System (INIS)

    Berven, B.A.; Goldsmith, W.A.; Haywood, F.F.; Schiager, K.J.

    1979-01-01

    Many sites used during the early days of the US atomic energy program are contaminated with radionuclides of the primordial decay chains (uranium, thorium, and actinium series). This contamination consists of residues resulting from refining and processing uranium and thorium. Preparation of these sites for release to unrestricted private use will involve the assistance of construction workers, many of whom have limited knowledge of the hazards associated with radioactive materials. Therefore, there is a need to educate these workers in the fundamentals of radioactive material handling to minimize exposures and possible spread of contamination. This training should disseminate relevant information at an appropriate educational level and should instill a cautious, common-sense attitude toward the handling of radioactive materials. The training should emphasize basic information concerning environmental radiation within a context of relative risk. A multi-media format, including colorful visual aids, demonstration, and discussion, should be used to maximize motivation and retention. A detailed, proposed training program design is presented

  12. Production of Thorium-229 at the ORNL High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Boll, Rose Ann; Garland, Marc A.; Mirzadeh, Saed

    2008-01-01

    The investigation of targeted cancer therapy using -emitters has developed considerably in recent years and clinical trials have generated promising results. In particular, the initial clinical trials for treatment of acute myeloid leukemia have demonstrated the effectiveness of the -emitter 213Bi in killing cancer cells. Pre-clinical studies have also shown the potential application of both 213Bi and its 225Ac parent radionuclide in a variety of cancer systems and targeted radiotherapy. Bismuth-213 is obtained from a radionuclide generator system from decay of the 10-d 225Ac parent, a member of the 7340-y 229Th chain. Currently, 233U is the only viable source for high purity 229Th; however, due to increasing difficulties associated with 233U safeguards, processing additional 233U is presently unfeasible. The recent decision to downblend and dispose of enriched 233U further diminished the prospects for extracting 229Th from 233U stock. Nevertheless, the anticipated growth in demand for 225Ac may soon exceed the levels of 229Th (∼40 g or ∼8 Ci; ∼80 times the current ORNL 229Th stock) present in the aged 233U stockpile. The alternative routes for the production of 229Th, 225Ra and 225Ac include both reactor and accelerator approaches. Here, we describe production of 229Th via neutron transmutation of 226Ra targets in the ORNL High Flux Isotope Reactor (HFIR).

  13. An optimized symbiotic fusion and molten-salt fission reactor system

    International Nuclear Information System (INIS)

    Blinkin, V.L.; Novikov, V.M.

    A symbiotic fusion-fission reactor system which breeds nuclear fuel is discussed. In the blanket of the controlled thermonuclear reactor (CTR) uranium-233 is generated from thorium, which circulates in the form of ThF 4 mixed with molten sodium and beryllium fluorides. The molten-salt fission reactor (MSR) burns up the uranium-233 and generates tritium for the fusion reactor from lithium, which circulates in the form of LiF mixed with BeF 2 and 233 UF 4 through the MSR core. With a CTR-MSR thermal power ratio of 1:11 the system can produce electrical energy and breed fuel with a doubling time of 4-5 years. The system has the following special features: (1) Fuel reprocessing is much simpler and cheaper than for contemporary fission reactors; reprocessing consists simply in continuous removal of 233 U from the salt circulating in the CTR blanket by the fluorination method and removal of xenon from the MSR fuel salt by gas scavenging; the MSR fuel salt is periodically exchanged for fresh salt and the 233 U is then removed from it; (2) Tritium is produced in the fission reactor, which is a much simpler system than the fusion reactor; (3) The CTR blanket is almost ''clean''; no tritium is produced in it and fission fragment activity does not exceed the activity induced in the structural materials; (4) Almost all the thorium introduced into the CTR blanket can be used for producing 233 U

  14. Economics of fusion driven symbiotic energy systems

    International Nuclear Information System (INIS)

    Renier, J.P.; Hoffman, T.J.

    1979-01-01

    The economic analysis of symbiotic energy systems in which U233 (to fuel advanced converters burning U233 fuel) is generated in blankets surrounding fusioning D-T plasma's depends on factors such as the plasma performance parameters, ore costs, and the relative costs of Fusion Breeders (CTR) to Advanced Fission Converters. The analysis also depends on detailed information such as initial, final makeup fuel requirements, fuel isotopics, reprocessing and fabrication costs, reprocessing losses (1%) and delays (2 years), the cost of money, and the effect of the underutilization of the factory thermal installation at the beginning of cycle. In this paper we present the results of calculations of overall fuel cycle and power costs, ore requirements, proliferation resistance and possibilities for grid expansion, based on detailed mass and energy flow diagrams and standard US INFCE cost data and introduction constraints, for realistic symbiotic scenarios involving CTR's (used as drivers) and denatured CANDU's (used as U233 burners). We compare the results with those obtained for other strategies involving heterogeneous LMFBR's which burn Pu to produce U233 for U233-burners such as the advanced CANDU converters

  15. Tritium Facilities Modernization and Consolidation Project Process Waste Assessment (Project S-7726)

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H. [Westinghouse Savannah River Company, AIKEN, SC (United States); Oji, L.N.

    1997-11-14

    Under the Tritium Facility Modernization {ampersand} Consolidation (TFM{ampersand}C) Project (S-7726) at the Savannah River Site (SS), all tritium processing operations in Building 232-H, with the exception of extraction and obsolete/abandoned systems, will be reestablished in Building 233-H. These operations include hydrogen isotopic separation, loading and unloading of tritium shipping and storage containers, tritium recovery from zeolite beds, and stripping of nitrogen flush gas to remove tritium prior to stack discharge. The scope of the TFM{ampersand}C Project also provides for a new replacement R&D tritium test manifold in 233-H, upgrading of the 233- H Purge Stripper and 233-H/234-H building HVAC, a new 234-H motor control center equipment building and relocating 232-H Materials Test Facility metallurgical laboratories (met labs), flow tester and life storage program environment chambers to 234-H.

  16. Study on Utilization of Super Grade Plutonium in Molten Salt Reactor FUJI-U3 using CITATION Code

    Science.gov (United States)

    Wulandari, Cici; Waris, Abdul; Pramuditya, Syeilendra; Asril, Pramutadi AM; Novitrian

    2017-07-01

    FUJI-U3 type of Molten Salt Reactor (MSR) has a unique design since it consists of three core regions in order to avoid the replacement of graphite as moderator. MSR uses floride as a nuclear fuel salt with the most popular chemical composition is LiF-BeF2-ThF4-233UF4. ThF4 and 233UF4 are the fertile and fissile materials, respectively. On the other hand, LiF and BeF2 working as both fuel and heat transfer medium. In this study, the super grade plutonium will be utilized as substitution of 233U since plutonium is easier to be obtained compared to 233U as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2002 code with JENDL 3.2 as nuclear data library.

  17. Chain and independent fission product yields adjusted to conform with physical conservation laws. Part 2

    International Nuclear Information System (INIS)

    Crouch, E.A.C.

    1976-01-01

    Previously reported adjustments to the chain yields and independent yields for the thermal neutron induced fission of 233 U, 235 U, 239 Pu and 241 Pu, the fast neutron induced fission of 232 Th, 233 U, 235 U, 238 U, 239 Pu, 240 Pu and 241 Pu, and the 14 MeV neutron induced fission of 232 Th, 233 U, 235 U and 238 U, have been recalculated using the principle of least squares. The adjustments to the chain yields so found are much smaller than those previously reported. (author)

  18. Special Analysis for the Disposal of the Consolidated Edison Uranium Solidification Project Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Environmental Management

    2013-01-31

    The purpose of this Special Analysis (SA) is to determine if the Oak Ridge (OR) Consolidated Edison Uranium Solidification Project (CEUSP) uranium-233 (233U) waste stream (DRTK000000050, Revision 0) is acceptable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). The CEUSP 233U waste stream requires a special analysis because the concentrations of thorium-229 (229Th), 230Th, 232U, 233U, and 234U exceeded their NNSS Waste Acceptance Criteria action levels. The acceptability of the waste stream is evaluated by determining if performance assessment (PA) modeling provides a reasonable expectation that SLB disposal is protective of human health and the environment. The CEUSP 233U waste stream is a long-lived waste with unique radiological hazards. The SA evaluates the long-term acceptability of the CEUSP 233U waste stream for near-surface disposal as a two tier process. The first tier, which is the usual SA process, uses the approved probabilistic PA model to determine if there is a reasonable expectation that disposal of the CEUSP 233U waste stream can meet the performance objectives of U.S. Department of Energy Manual DOE M 435.1-1, “Radioactive Waste Management,” for a period of 1,000 years (y) after closure. The second tier addresses the acceptability of the OR CEUSP 233U waste stream for near-surface disposal by evaluating long-term site stability and security, by performing extended (i.e., 10,000 and 60,000 y) modeling analyses, and by evaluating the effect of containers and the depth of burial on performance. Tier I results indicate that there is a reasonable expectation of compliance with all performance objectives if the OR CEUSP 233U waste stream is disposed in the Area 5 RWMS SLB disposal units. The maximum mean and 95th percentile PA results are all less than the performance objective for 1,000 y. Monte Carlo uncertainty analysis indicates that there is a high likelihood of

  19. Centrally acting serotonergic and dopaminergic agents. 1. Synthesis and structure-activity relationships of 2,3,3a,4,5,9b-hexahydro-1H-benz[e]indole derivatives.

    Science.gov (United States)

    Lin, C H; Haadsma-Svensson, S R; Lahti, R A; McCall, R B; Piercey, M F; Schreur, P J; Von Voigtlander, P F; Smith, M W; Chidester, C G

    1993-04-16

    The synthesis and structure-activity relationships (SAR) of 2,3,3a,4,5,9b-hexahydro-1H-benz[e]indole derivatives (3) are described. These compounds are conformationally restricted, angular tricyclic analogs of 2-aminotetralin. The synthesis was achieved in several steps from the corresponding 2-tetralones. The enantiomers of the cis analogs were obtained from either fractional recrystallizations of the diastereomeric salts of di-p-toluoyl-L-(or D)-tartaric acid or an asymmetric synthesis using chiral (R)-alpha-methylbenzylamine. All analogs were evaluated in the in vitro 5-HT1A and D2 binding assays and selected analogs were investigated further in biochemical and behavioral tests. Analogs with 9-methoxy substitution (R1 in 3) showed mixed 5-HT1A agonist and dopamine antagonist activities whereas the corresponding 9-hydroxy analogs displayed selective 5-HT1A agonist activity. The cis analogs were found to be more potent than the corresponding trans analogs and in the cis series, the (3aR)-(-)-enantiomers displayed higher potency. Nitrogen substitution (R2 in 3) with either an n-propyl or an allyl group produced similar activities whereas replacement with a bulky alpha-methylbenzyl group resulted in loss of activity. Analogs without aromatic substitution (R1 = H in 3) still showed good 5-HT1A agonist activity, although less potent than the 9-methoxy series. In this case, the trans analogs possessed equal or higher in vitro 5-HT1A affinity than the corresponding cis analogs. Analogs with either 6-methoxy or 6-hydroxy substitution (R1 in 3) were found to display dopamine antagonist properties. However, only N-allyl analogs showed this activity. In the 6-methoxy-N-allyl series, the cis analog was found to be more potent than the trans analog. Again, between the pair of cis enantiomers, the (3aR)-(-)-enantiomer showed higher potency. Incorporation of an additional methyl group into 9-methoxy cis analogs at C-2 resulted in retention of potent 5-HT1A agonist activity.

  20. Gender | Page 233 | IDRC - International Development Research ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Gender. Sexospécificités. Il s'agissait d'une politique née d'une promesse mirifique ... from women's experiences of decentralization in Nepal, Pakistan, and India. ... last few decades, gender inequality and gender-based inequities continue to ...

  1. 39 CFR 233.2 - Circulars and rewards.

    Science.gov (United States)

    2010-07-01

    .... (viii) Murder or manslaughter of a postal employee. (ix) Mailing or receiving through the mail any... the following offenses: Murder or Manslaughter, $100,000. The unlawful killing of any officer or...

  2. 16 CFR 233.1 - Former price comparisons.

    Science.gov (United States)

    2010-01-01

    ... each. His usual markup is 50 percent over cost; that is, his regular retail price is $7.50. In order... that the former price is not a fictitious one. If the former price, or the amount or percentage of...

  3. 48 CFR 1352.233-70 - Agency protests.

    Science.gov (United States)

    2010-10-01

    ... served upon the Contract Law Division of the Office of the General Counsel within one day of filing a... Contract Law Division shall be made as follows: U.S. Department of Commerce, Office of the General Counsel, Chief, Contract Law Division, Room 5893, Herbert C. Hoover Building, 14th Street and Constitution Avenue...

  4. 12 CFR 233.3 - Designated payment systems.

    Science.gov (United States)

    2010-01-01

    ... following payment systems could be used by participants in connection with, or to facilitate, a restricted... include check cashing, currency exchange, or the issuance or redemption of money orders, travelers' checks...

  5. 45 CFR 233.50 - Citizenship and alienage.

    Science.gov (United States)

    2010-10-01

    ... and Nationality Act: (1) Section 207(c), in effect after March 31, 1980—Aliens Admitted as Refugees. (2) Section 203(a)(7), in effect prior to April 1, 1980—Individuals who were Granted Status as Conditional Entrant Refugees. (3) Section 208—Aliens Granted Political Asylum by the Attorney General. (4...

  6. 49 CFR 190.233 - Corrective action orders.

    Science.gov (United States)

    2010-10-01

    ... effective, the Associate Administrator, OPS may request the Attorney General to bring an action for appropriate relief in accordance with § 190.235. (i) Upon petition by the Attorney General, the District... must notify the Associate Administrator, OPS of that election in writing within 10 days of service of...

  7. Ecology of marine deposit feeders. Ed. by G. Lopez, C. Taghon and J. Levinton

    Digital Repository Service at National Institute of Oceanography (India)

    Royan, J

    stream_size 2 stream_content_type text/plain stream_name Indian_J_Mar_Sci_19_233.pdf.txt stream_source_info Indian_J_Mar_Sci_19_233.pdf.txt Content-Encoding ISO-8859-1 Content-Type text/plain; charset=ISO-8859-1 ...

  8. Ghana Mining Journal: Contact

    African Journals Online (AJOL)

    Principal Contact. Professor Daniel Mireku-Gyimah Editor-in-Chief University of Mines & Technology Ghana Mining Journal University of Mines & Technology P. O. BOX 237 Tarkwa Ghana Phone: +233 362 20280/20324. Fax: +233 362 20306. Email: dm.gyimah@umat.edu.gh ...

  9. Tritium Facilities Modernization and Consolidation Project Process Waste Assessment (Project S-7726)

    International Nuclear Information System (INIS)

    Hsu, R.H.; Oji, L.N.

    1997-01-01

    Under the Tritium Facility Modernization ampersand Consolidation (TFM ampersand C) Project (S-7726) at the Savannah River Site (SS), all tritium processing operations in Building 232-H, with the exception of extraction and obsolete/abandoned systems, will be reestablished in Building 233-H. These operations include hydrogen isotopic separation, loading and unloading of tritium shipping and storage containers, tritium recovery from zeolite beds, and stripping of nitrogen flush gas to remove tritium prior to stack discharge. The scope of the TFM ampersand C Project also provides for a new replacement R ampersand D tritium test manifold in 233-H, upgrading of the 233- H Purge Stripper and 233-H/234-H building HVAC, a new 234-H motor control center equipment building and relocating 232-H Materials Test Facility metallurgical laboratories (met labs), flow tester and life storage program environment chambers to 234-H

  10. Protein expression of saccharomyces cerevisiae in response to uranium exposure

    International Nuclear Information System (INIS)

    Sakamoto, Fuminori; Nankawa, Takuya; Kozai, Naofumi; Ohnuki, Toshihiko; Fujii, Tsutomu; Iefuji, Haruyuki; Francis, A.J.

    2007-01-01

    Protein expression of Saccharomyces cerevisiae grown in the medium containing 238 U (VI) and 233 U (VI) was examined by two-dimensional gel electrophoresis. Saccharomyces cerevisiae of BY4743 was grown in yeast nitrogen base medium containing glucose and glycerol 2-phosphate and 238 U of 0, 2.0, and 5.0 x 10 -4 M or 233 U of 2.5 x 10 -6 M (radioactivity was higher by 350 times than 2.0 x 10 -4 M 238 U) and 5.0 x 10 -6 M for 112 h at 30 degC. The growth of Saccharomyces cerevisiae was monitored by measuring OD 600 at 112 h after the inoculation. Uranium concentrations in the media also were measured by radiometry using a liquid scintillation counter. The growths of the yeast grown in the above media were in the following order: control>2.5 x 10 -6 M 233 U>2.0 x 10 -4 M 238 U>5.0 x 10 -6 M 233 U>5.0 x 10 -4 M 238 U. This result indicated that not only radiological but also chemical effect of U reduced the growth of the yeast. The concentrations of U in the medium containing 238 U or 233 U decreased, suggesting U accumulation by the yeast cells. The 2-D gel electrophoresis analysis showed the appearance of several spots after exposure to 238 U or to 233 U but not in the control containing no uranium. These results show that the yeast cells exposed to U express several specific proteins. (author)

  11. Determination of natural uranium in urine ({sup 233}U); Dosage de l'uranium dans l'urine ({sup 233}U)

    Energy Technology Data Exchange (ETDEWEB)

    Jeanmaire, L; Jammet, H [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    A procedure for the quantitative analysis of uranium in urine is described. The residue obtained by mineralization is dissolved in diluted hydrochloric acid. Uranium is separated by fixation on a permutit 50 column, elution with 0,2 M oxalic acid and electrodeposition on nickel. Uranium is then measured by {alpha} counting. It is thus possible to detect less than 1 pico-curie of uranium in the sample. (author) [French] Cet article decrit une technique de dosage de l'uranium dans l'urine. Apres mineralisation, le residu est dissous dans l'acide chlorhydrique dilue. L'uranium est separe par fixation, sur une colonne de permutite 50, elution au moyen d'acide oxalique 0,2 M et depot electrolytique sur nickel. La mesure faite par comptage {alpha} permet de detecter moins de 1 picocurie d'uranium dans l'echantillon. (auteur)

  12. Summary and conclusions

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, D.G; Murthy, K.S; Neprochnov, Y.P; Subrahmanyam, C.

    stream_size 4 stream_content_type text/plain stream_name Mem_Geol_Soc_India_1998_39_233.pdf.txt stream_source_info Mem_Geol_Soc_India_1998_39_233.pdf.txt Content-Encoding ISO-8859-1 Content-Type text/plain; charset=ISO-8859-1 ...

  13. 78 FR 72914 - Changes in Flood Hazard Determinations

    Science.gov (United States)

    2013-12-04

    ... INFORMATION CONTACT: Luis Rodriguez, Chief, Engineering Management Branch, Federal Insurance and Mitigation... County Department, 233 lomc. 1509P). Judge, Paul North Pecos-La Elizondo Tower, Trinidad Street, 101 West... Works www.msc.fema.gov/ County (13-06- Bexar County Department, 233 lomc. 2845P). Judge, Paul North...

  14. African Journal of Management Research: Contact

    African Journals Online (AJOL)

    Principal Contact. Prof. Anthony Q. Q. Aboagye Editor University of Ghana Business School. University of Ghana Business School P.O. Box LG 78. Legon Accra Ghana. Phone: +233-24-425-2596. Email: qaboagye@ug.edu.gh. Support Contact. Sylvia Ahudzo (Editorial Assistant) Phone: +233-24-318-7075

  15. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  16. Design Feasible Area on Water Cooled Thorium Breeder Reactor in Equilibrium States

    International Nuclear Information System (INIS)

    Sidik Permana; Naoyuki Takaki; Hiroshi Sekimoto

    2006-01-01

    Thorium as supplied fuel has good candidate for fuel material if it is converted into fissile material 233 U which shows superior characteristics in the thermal region. The Shippingport reactor used 233 U-Th fuel system, and the molten salt breeder reactor (MSBR) project showed that breeding is possible in a thermal spectrum. In the present study, feasibility of water cooled thorium breeder reactor is investigated. The key properties such as flux, η value, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The results show the feasibility of breeding for different MFR and burn-ups. The required 233 U enrichment is about 2% - 9% as charge fuel. The lower MFR and the higher enrichment of 233 U are preferable to improve the average burn-up; however the design feasible window is shrunk. This core shows the design feasible window especially in relation to MFR with negative void reactivity coefficient. (authors)

  17. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    International Nuclear Information System (INIS)

    Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

    2016-01-01

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  18. Research of natural resources saving by design studies of Pressurized Light Water Reactors and High Conversion PWR cores with mixed oxide fuels composed of thorium/uranium/plutonium

    International Nuclear Information System (INIS)

    Vallet, V.

    2012-01-01

    Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3. generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium/uranium/plutonium, and in the other hand, to elaborate multi-recycling strategies of both plutonium and 233 U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium/uranium/plutonium, and secondly the setting up of multi-recycling strategies of both plutonium and 233 U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO 2 fuel, and the second is with standard or high MR and ThUO 2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233 U production and low plutonium consumption. This second step came up with two specific concepts, from which multi-recycling strategies have been elaborated. The exclusive production and recycling of 233 U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233 U production is the limiting factor. That is why it was eventually proposed to study how the production of 233 U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233 U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%. (author) [fr

  19. 9th Annual Systems Engineering Conference: Volume 2 Tuesday

    Science.gov (United States)

    2006-10-26

    Enterprise modeling • MBSE standards emerging (SysML, AP233, BPMN , UPDM, ..) MBSE State of Practice - 2010 • Commonly practiced across broad range of...and practiced in an ad-hoc manner. MBSE standards are emerging (OMG SysML, XMI, AP233, BPMN , Architecture Frameworks). • The projected state of MBSE

  20. 78 FR 70065 - Agency Information Collection Activities: Customs Declaration

    Science.gov (United States)

    2013-11-22

    ... technology; and (e) the annual costs burden to respondents or record keepers from the collection of... Respondents: 105,606,000. Estimated Number of Total Annual Responses: 105,606,000. Estimated Time per Response... Respondents: 233,000,000. Estimated Number of Total Annual Responses: 233,000,000. Estimated Time per Response...

  1. 78 FR 45941 - Changes in Flood Hazard Determinations

    Science.gov (United States)

    2013-07-30

    ... Center at www.msc.fema.gov . FOR FURTHER INFORMATION CONTACT: Luis Rodriguez, Chief, Engineering... Department of Public 06-0093P). County Judge, Paul Works, 233 North Elizondo Tower, 101 Pecos-La Trinidad... County (12- W. Wolff, Bexar Department of Public 06-1791P). County Judge, Paul Works, 233 North Elizondo...

  2. Role of thorium in the industry advantage of atomic energy

    International Nuclear Information System (INIS)

    Souza Santos, M.D. de; Goldemberg, J.; Lopes, J.L.

    1985-01-01

    Based in the utilization of others fossil substances, such as plutonium and uranium 233, produzed through the thorium and natural uranium (238), it is discussed the relative merits of alternative processes: to produce U233 on Pu 239 to substitute the initial load of U235. (M.C.K.) [pt

  3. Overexpression of the catalytically impaired Taspase1 T234V or Taspase1 D233A variants does not have a dominant negative effect in T(4;11 leukemia cells.

    Directory of Open Access Journals (Sweden)

    Carolin Bier

    Full Text Available BACKGROUND: The chromosomal translocation t(4;11(q21;q23 is associated with high-risk acute lymphoblastic leukemia of infants. The resulting AF4•MLL oncoprotein becomes activated by Taspase1 hydrolysis and is considered to promote oncogenic transcriptional activation. Hence, Taspase1's proteolytic activity is a critical step in AF4•MLL pathophysiology. The Taspase1 proenzyme is autoproteolytically processed in its subunits and is assumed to assemble into an αββα-heterodimer, the active protease. Therefore, we investigated here whether overexpression of catalytically inactive Taspase1 variants are able to interfere with the proteolytic activity of the wild type enzyme in AF4•MLL model systems. METHODOLOGY/FINDINGS: The consequences of overexpressing the catalytically dead Taspase1 mutant, Taspase1(T234V, or the highly attenuated variant, Taspase1(D233A, on Taspase1's processing of AF4•MLL and of other Taspase1 targets was analyzed in living cancer cells employing an optimized cell-based assay. Notably, even a nine-fold overexpression of the respective Taspase1 mutants neither inhibited Taspase1's cis- nor trans-cleavage activity in vivo. Likewise, enforced expression of the α- or β-subunits showed no trans-dominant effect against the ectopically or endogenously expressed enzyme. Notably, co-expression of the individual α- and β-subunits did not result in their assembly into an enzymatically active protease complex. Probing Taspase1 multimerization in living cells by a translocation-based protein interaction assay as well as by biochemical methods indicated that the inactive Taspase1 failed to assemble into stable heterocomplexes with the wild type enzyme. CONCLUSIONS: Collectively, our results demonstrate that inefficient heterodimerization appears to be the mechanism by which inactive Taspase1 variants fail to inhibit wild type Taspase1's activity in trans. Our work favours strategies targeting Taspase1's catalytic activity

  4. Some physics problems in the design of thorium-fuelled CANDU reactors

    International Nuclear Information System (INIS)

    Milgram, M.S.; Walker, W.H.

    1976-08-01

    This paper is the text of a presentation to the American Nuclear Society Conference, Toronto, 14-1 8 June, 1976. The contents deal with the adequacy of fission product representations in U 233 -Th cycles, the sensitivity of burnup predictions to cross-section data, and the flux dependence phenomenon associated with Pa 233

  5. Program Management Educational Needs of Idaho Business and Marketing Teachers

    Science.gov (United States)

    Kitchel, Allen; Cannon, John; Duncan, Dennis

    2009-01-01

    The purpose of this study was to determine the perceived program management professional development needs of Idaho secondary business/marketing teachers (N = 233) in order to guide pre-service curriculum development and in-service training activities. Sixty-two percent (n = 146) of the 233 teachers completed a modified version of Joerger's (2002)…

  6. Fused salt power reactor study: Minutes of discussion meeting No. 2

    International Nuclear Information System (INIS)

    Alexander, L. G.

    1956-01-01

    Remarks made by participants in a 1956 meeting are sketched. Economics was a major concern. Significant topics included development of a new alloy for use in the heat exchanger, conversion ratios in a U-233 breeder, the effects of ThF 4 on corrosion, and means of producing various transmutation products other than U-233.

  7. Geological disposal: security and R and D. Security of 'second draft for R and D of geological disposal'

    International Nuclear Information System (INIS)

    Shiotsuki, Masao; Miyahara, Kaname

    2003-01-01

    The second draft for R and D of geological disposal (second draft) was arranged in 1999. The idea of security of geological disposal in the second draft is explained. The evaluation results of the uncertainty analysis and an example of evaluation of the effect of separation nuclear transmutation on the geological disposal are shown. The construction of strong engineered barrier is a basic idea of geological disposal system. Three processes such as isolation, engineering countermeasures and safety evaluation are carried out for the security of geological disposal. The security of geological environment for a long time of 12 sites in Japan was studied by data. Provability of production and enforcement of engineered barrier were confirmed by trial of over pack, tests and the present and future technologies developed. By using the conditions of reference case in the second draft, the evaluation results of dose effects in the two cases: 1) 90 to 99% Cs and Sr removed from HLW (High Level radioactive Waste) and 2) high stripping ratio of actinium series are explained. (S.Y.)

  8. Ac, La, and Ce radioimpurities in {sup 225}Ac produced in 40-200 MeV proton irradiations of thorium

    Energy Technology Data Exchange (ETDEWEB)

    Engle, Jonathan W.; Ballard, Beau D. [Los Alamos National Laboratory, NM (United States); Weidner, John W. [Air Force Institute of Technology, Wright Patterson Air Force Base, OH (United States); and others

    2014-10-01

    Accelerator production of {sup 225}Ac addresses the global supply deficiency currently inhibiting clinical trials from establishing {sup 225}Ac's therapeutic utility, provided that the accelerator product is of sufficient radionuclidic purity for patient use. Two proton activation experiments utilizing the stacked foil technique between 40 and 200 MeV were employed to study the likely co-formation of radionuclides expected to be especially challenging to separate from {sup 225}Ac. Foils were assayed by nondestructive γ-spectroscopy and by α-spectroscopy of chemically processed target material. Nuclear formation cross sections for the radionuclides {sup 226}Ac and {sup 227}Ac as well as lower lanthanide radioisotopes {sup 139}Ce, {sup 141}Ce, {sup 143}Ce, and {sup 140}La whose elemental ionic radii closely match that of actinium were measured and are reported. The predictions of the latest MCNP6 event generators are compared with measured data, as they permit estimation of the formation rates of other radionuclides whose decay emissions are not clearly discerned in the complex spectra collected from {sup 232}Th(p,x) fission product mixtures. (orig.)

  9. Background radiation and individual dosimetry in the coastal area of Tamil Nadu (India)

    International Nuclear Information System (INIS)

    Matsuda, N.; Brahmanandhan, G. M.; Yoshida, M.; Takamura, N.; Suyama, A.; Koguchi, Y.; Juto, N.; Raj, Y. L.; Winsley, G.; Selvasekarapandian, S.

    2011-01-01

    South coast of India is known as the high-level background radiation area (HBRA) mainly due to beach sands that contain natural radionuclides as components of the mineral monazite. The rich deposit of monazite is unevenly distributed along the coastal belt of Tamil Nadu and Kerala. An HBRA site that laid in 2x7 m along the sea was found in the beach of Chinnavillai, Tamil Nadu, where the maximum ambient dose equivalent reached as high as 162.7 mSv y -1 . From the sands collected at the HBRA spot, the high-purity germanium semi-conductor detector identified six nuclides of thorium series, four nuclides of uranium series and two nuclides belonging to actinium series. The highest radioactivity observed was 43.7 Bq g -1 of Th-228. The individual dose of five inhabitants in Chinnavillai, as measured by the radiophotoluminescence glass dosimetry system, demonstrated the average dose of 7.17 mSv y -1 ranging from 2.79 to 14.17 mSv y -1 . (authors)

  10. Soil nuclide distribution coefficients and their statistical distributions

    International Nuclear Information System (INIS)

    Sheppard, M.I.; Beals, D.I.; Thibault, D.H.; O'Connor, P.

    1984-12-01

    Environmental assessments of the disposal of nuclear fuel waste in plutonic rock formations require analysis of the migration of nuclides from the disposal vault to the biosphere. Analyses of nuclide migration via groundwater through the disposal vault, the buffer and backfill, the plutonic rock, and the consolidated and unconsolidated overburden use models requiring distribution coefficients (Ksub(d)) to describe the interaction of the nuclides with the geological and man-made materials. This report presents element-specific soil distribution coefficients and their statistical distributions, based on a detailed survey of the literature. Radioactive elements considered were actinium, americium, bismuth, calcium, carbon, cerium, cesium, iodine, lead, molybdenum, neptunium, nickel, niobium, palladium, plutonium, polonium, protactinium, radium, samarium, selenium, silver, strontium, technetium, terbium, thorium, tin, uranium and zirconium. Stable elements considered were antimony, boron, cadmium, tellurium and zinc. Where sufficient data were available, distribution coefficients and their distributions are given for sand, silt, clay and organic soils. Our values are recommended for use in assessments for the Canadian Nuclear Fuel Waste Management Program

  11. Performance of LMFBR fuel pins with (Pu,Th)O/sub 2-x/ and UO2

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1983-09-01

    The irradiation performance of (Pu,Th)O/sub 2-x/ and UO 2 fueled pins for breeder reactor application were compared to the extensive performance data base for the (U,Pu)O/sub 2-x/ fuel system. Th-Pu and 238 U- 233 U based fuel systems were candidate fuel fertile/fissile isotopic combinations for development of alternatives to the current LMFBR fuel cycle. Initial screening tests were conducted in the EBR-II to obtain comparative performance data because of the limited experience with these fuel systems. In some cases, 235 U was used as a substitute for 233 U because of the difficulties in fabrication of available 233 U due to its high gamma ray emission rate

  12. Lower bounds on the independence number of certain graphs of odd girth at least seven

    DEFF Research Database (Denmark)

    Pedersen, A. S.; Rautenbach, D.; Regen, F.

    2011-01-01

    Heckman and Thomas [C.C. Heckman, R. Thomas, A new proof of the independence ratio of triangle-free cubic graphs, Discrete Math. 233 (2001) 233-237] proved that every connected subcubic triangle-free graph G has an independent set of order at least (4n(G) - m(G) - 1)/7 where n(G) and m(G) denote...

  13. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

    2014-01-01

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

  14. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

    Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

  15. Study on the thorium-based breeder with molten fluoride salt blanket in the Nuclear Hot Spring - 5420

    International Nuclear Information System (INIS)

    Bing, X.; Yingzhong, L.

    2015-01-01

    Nuclear Hot Spring (NHS) is an innovative reactor type featured by pool-type molten-salt-cooled pebble-bed reactor core with the capability of natural circulation under full power operation. Except for the potential applications in power generation and high temperature process heat, thorium-based breeding is also a promising feature of the NHS. In order to take advantage of both the highly inherent safety and the on-line processing capability of fluid thorium-based fuels, a breeder design of NHS equipped with a blanket of molten salt with thorium fluoride outside the pebble-bed core is proposed in this work. For the purpose of keeping cleanness of the primary loop and blanket loop, both loops are isolated physically from each other, and the rapid on-line extraction of converted 233 Pa and 233 U is employed for the processing of blanket salt. The conversion ratio, defined as the ratio of converted 233 Pa and 233 U to the consumed fissile uranium in seed fuels, is investigated by varying the relevant parameters such as the circulation flux of blanket salt and the discharge burn-up of seed fuels. It is found that breeding can be achieved for the pure 233 U seed scheme with relatively low discharge burn-up and low blanket salt flux. However, the reprocessing for the HTGR fuels with TRISO particles has to be taken into account to ensure the breeding. (authors)

  16. Determination of natural uranium in urine (233U)

    International Nuclear Information System (INIS)

    Jeanmaire, L.; Jammet, H.

    1959-01-01

    A procedure for the quantitative analysis of uranium in urine is described. The residue obtained by mineralization is dissolved in diluted hydrochloric acid. Uranium is separated by fixation on a permutit 50 column, elution with 0,2 M oxalic acid and electrodeposition on nickel. Uranium is then measured by α counting. It is thus possible to detect less than 1 pico-curie of uranium in the sample. (author) [fr

  17. 45 CFR 233.90 - Factors specific to AFDC.

    Science.gov (United States)

    2010-10-01

    ... physical or mental injury, sexual abuse or exploitation, or negligent treatment or maltreatment of such...) payments with respect to a pregnant woman with no other children receiving assistance, and additionally, at... pregnancy both for the pregnant woman with no other children as well as for the pregnant woman receiving...

  18. Cover Image, Volume 233, Number 7, July 2018.

    Science.gov (United States)

    Yin, Chong; Zhang, Yan; Hu, Lifang; Tian, Ye; Chen, Zhihao; Li, Dijie; Zhao, Fan; Su, Peihong; Ma, Xiaoli; Zhang, Ge; Miao, Zhiping; Wang, Liping; Qian, Airong; Xian, Cory J

    2018-07-01

    Cover: The cover image, by Yin et al., is based on the Original Research Article, Mechanical unloading reduces microtubule actin crosslinking factor 1 expression to inhibit β-Catenin Signaling and osteoblast proliferation, DOI: 10.1002/jcp.26374. © 2018 Wiley Periodicals, Inc.

  19. 40 CFR 233.61 - Determination of Tribal eligibility.

    Science.gov (United States)

    2010-07-01

    ..., such as, but not limited to, the exercise of police powers affecting (or relating to) the health, safety, and welfare of the affected population; taxation; and the exercise of the power of eminent domain... narrative statement describing the capability of the Indian Tribe to administer an effective 404 permit...

  20. 39 CFR 233.7 - Forfeiture authority and procedures.

    Science.gov (United States)

    2010-07-01

    ...; advertisement; declaration of forfeiture. (1) The Postal Inspection Service must cause written notice of the... forfeiture unless the petitioner establishes that: (1) The petitioner has a valid, good faith and legally...

  1. Dicty_cDB: VHN233 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available ns full open reading fra... 76 7e-13 AF134593_1( AF134593 |pid:none) Homo sapiens L-pipecolic acid oxid... 7...7e-13 BC114006_1( BC114006 |pid:none) Bos taurus L-pipecolic acid oxidas... 76 9e-13 protein update 2009. 7.

  2. 40 CFR 233.53 - Withdrawal of program approval.

    Science.gov (United States)

    2010-07-01

    ... exercise control over activities required to be regulated under this part, including failure to issue... 22.02—(use of number/gender); (B) Section 22.04—(authorities of Presiding Officer); (C) Section 22.06... business day. (2) Extensions of time. The Administrator, Regional Administrator, or Presiding Officer, as...

  3. Dicty_cDB: VSK233 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available anslated Amino Acid sequence yk*ksyi*ylns*firessfgn*iyrf**nkfk*ink*iny*rfyffyvryt...iklw*lnisilik*i*ink*inkllkilffxc Frame B: yk*ksyi*ylns*firessfgn*iyrf**nkfk*ink*iny*rfyffyvrytypvgey*F MKIKKK--- ---yk*ksyi*ylns*fire

  4. Dicty_cDB: VSF233 [Dicty_cDB

    Lifescience Database Archive (English)

    Full Text Available lllks*vit*eimvikilisc*nqldlilrikknnnnnki*msl Frame B: ***sptrisctyc*kvr*llkr*wllry*ylvktn*i*f*elkkiiiiikfkch...y*nk*i kkk--- ---***sptrisctyc*kvr*llkr*wllry*ylvktn*i*f*elkkiiiiikfkch Frame C:

  5. Prevalence and incidence of Parkinson's disease in The Faroe Islands

    DEFF Research Database (Denmark)

    Wermuth, Lene; Bech, Sara Brynhild Winther; Petersen, Maria Skaalum

    2008-01-01

    A study in The Faroe Islands in 1995 suggested a high prevalence of idiopathic Parkinson's disease (IPD) and total parkinsonism of 187.6 and 233.4 per 100,000 inhabitants respectively.......A study in The Faroe Islands in 1995 suggested a high prevalence of idiopathic Parkinson's disease (IPD) and total parkinsonism of 187.6 and 233.4 per 100,000 inhabitants respectively....

  6. Generation of efficient mutants of endoglycosidase from Streptococcus pyogenes and their application in a novel one-pot transglycosylation reaction for antibody modification.

    Directory of Open Access Journals (Sweden)

    Mitsuhiro Iwamoto

    Full Text Available The fine structures of Fc N-glycan modulate the biological functions and physicochemical properties of antibodies. By remodeling N-glycan to obtain a homogeneous glycoform or chemically modified glycan, antibody characteristics can be controlled or modified. Such remodeling can be achieved by transglycosylation reactions using a mutant of endoglycosidase from Streptococcus pyogenes (Endo-S and glycan oxazoline. In this study, we generated improved mutants of Endo-S by introducing additional mutations to the D233Q mutant. Notably, Endo-S D233Q/Q303L, D233Q/E350Q, and several other mutations resulted in transglycosylation efficiencies exceeding 90%, with a single-digit donor-to-substrate ratio of five, and D233Q/Y402F/D405A and several other mutations resulted in slightly reduced transglycosylation efficiencies accompanied by no detectable hydrolysis activity for 48 h. We further demonstrated that the combined use of mutants of Endo-S with Endo-M or Endo-CC, endoglycosidases from Mucor hiemalis and Coprinopsis cinerea, enables one-pot transglycosylation from sialoglycopeptide to antibodies. This novel reaction enables glycosylation remodeling of antibodies, without the chemical synthesis of oxazoline in advance or in situ.

  7. Determination of the conjugated linoleic acid-containing triacylglycerols in New Zealand bovine milk fat.

    Science.gov (United States)

    Robinson, N P; MacGibbon, A K

    2000-07-01

    Reversed-phase high-performance liquid chromatography (HPLC) with ultraviolet (UV) detection at 233 nm was used to separate, quantify, and identify the triacylglycerols (TAG) of milk fat that contain conjugated linoleic acid (CLA). The absorbance at 233 nm was substantially due to CLA-TAG (chromatography of some representative TAG devoid of CLA, such as tripalmitin and triolein, showed poor responses at 233 nm, 1/800th that of CLA-TAG). A CLA molar extinction coefficient at 233 nm of 23,360 L mol(-1) cm(-1) and an HPLC UV response factor were obtained from a commercially available cis-9,trans-11-CLA standard. This molar extinction coefficient was only 86% of reported literature values. Summation of all chromatographic peaks absorbing at 233 nm using the corrected response factor gave good agreement with independent determinations of total CLA by gas chromatography and UV spectrophotometry. This agreement allowed quantification of individual CLA-TAG peaks in the HPLC separation of a typical New Zealand bovine milk fat. Three CLA-containing TAG, CLA-dipalmitin, CLA-oleoyl-palmitin and CLA-diolein, were prepared by interesterification of tripalmitin with the respective fatty acid methyl esters and used to assign individual peaks in the reversed-phase chromatography of total milk fat, of which CLA-oleoyl-palmitin was coincident with the largest UV peak. Band fractions from argentation thin-layer chromatography of total milk fat were similarly employed to identify five predominant CLA-TAG groups in total milk fat: CLA-disaturates, CLA-oleoyl-saturates, CLA-vaccenyl-saturates, CLA-vaccenyl-olein, and CLA-diolein.

  8. Nuclear performance optimization of the Be/Li/Th blanket for the fusion breeder

    International Nuclear Information System (INIS)

    Lee, J.D.; Bandini, B.R.

    1985-01-01

    More rigorous nuclear analysis, including treatment of resonance self-shielding effects coupled with an optimization procedure, has resulted in improved performance of the Be/Li/Th blanket. Net U-233 breeding ratio has increased 36% (to 0.84) while at an average U-233/Th ratio of 0.5 a/o average energy multiplication has increased only 12% (to 2.1) compared with earlier results

  9. Author Index

    Indian Academy of Sciences (India)

    Huang, Y., 53. Hudec, L., 121. Hudec, R., 91, 121. Hughes, P. A., 5. Iguchi, S., 61. Jia, L. W., 309. Jiang, D. R., 261. Jorstad, S. G., 233, 239. Kang, H., 301. Kardashev, N. S., 105. Kong, M. Z., 209. Kovacs, Z., 189. Krichbaum, T. P., 29, 57. Krishan, V., 265. Kurtanidze, O., 67. Lähteenmäki, A., 105, 233, 239. Larinov, M. G., 105,.

  10. Reklamen går STEALTH

    DEFF Research Database (Denmark)

    Andersen, Lars Pynt

    2010-01-01

    At snige sig uset gennem radarsystemerne er dog ikke kun noget der appellerer til luftvåbnet. Det er tilsyneladende også mere og mere væsentligt i markedskommunikation. Udgivelsesdato: 23.3.......At snige sig uset gennem radarsystemerne er dog ikke kun noget der appellerer til luftvåbnet. Det er tilsyneladende også mere og mere væsentligt i markedskommunikation. Udgivelsesdato: 23.3....

  11. HPMA copolymer-based polymer conjugates for the delivery and controlled release of retinoids

    Czech Academy of Sciences Publication Activity Database

    Lidický, Ondřej; Šírová, Milada; Etrych, Tomáš

    2016-01-01

    Roč. 65, Suppl. 2 (2016), S233-S241 ISSN 0862-8408 R&D Projects: GA MŠk(CZ) LQ1604 Institutional support: RVO:61389013 ; RVO:61388971 Keywords : polymer conjugate * retinoid * HPMA Subject RIV: EB - Genetics ; Molecular Biology; EA - Cell Biology (MBU-M) Impact factor: 1.461, year: 2016 http://www.biomed.cas.cz/physiolres/pdf/65%20Suppl%202/65_S233.pdf

  12. Nuclear energy from thorium

    International Nuclear Information System (INIS)

    Coote, G.E.

    1977-06-01

    Relevant topics in nuclear and reactor physics are outlined. These include: the thorium decay series; generation of fissile from fertile nuclides, in particular U-233 from Th-232; the princiiples underlying thermal breeder reactors; the production of U-232 in thorium fuel and its important influence on nuclear safeguards and the recycling of U-233. Development work is continuing on several types of reactor which could utilise thorium; each of these is briefly described and its possible role is assessed. Other tipics covered include safety aspects of thorium oxide fuel, reprocessing, fabrication of recycle fuel and the possibility of denaturing U-233 by adding natural uranium. It is concluded that previoue arguments for development of the thorium cycle are still valid but those relating to non-proliferation of weapons may become even more compelling. (auth.)

  13. Studies on use of reflector material and its position within FBR core for reducing U{sup 232} content of U produced in ThO{sub 2} radial blankets

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Sujoy, E-mail: sujoy@igcar.gov.in [Core Design Group, IGCAR, Kalpakkam (India); Prasad, Rajeev Ranjan; Bagchi, Subhrojit [Core Design Group, IGCAR, Kalpakkam (India); Mohanakrishnan, P. [MCNS, Manipal University, Manipal (India); Arul, A. John; Puthiyavinayagam, P. [Core Design Group, IGCAR, Kalpakkam (India)

    2015-11-15

    Highlights: • Nuclear data processing for multigroup neutron transport calculation. • Discrete ordinate and Monte Carlo neutron transport. • Breeding of Thorium in Fast Reactor. • Minimization of U{sup 232} in U{sup 233}. • Fuel burn up using Neutron Diffusion. - Abstract: Presence of U{sup 232} in U{sup 233} bred in thorium blanket of fast reactor is a major concern in fuel reprocessing. The former's daughter products being hard gamma emitter and the isotope itself having substantial half life, its presence beyond 10 ppm makes fuel recycle complicated and expensive. In this study possibility of decreasing U{sup 232} production in a typical FBR blanket by means of spectrum modification is examined. SS, depleted B{sub 4}C, SiC, Mo and W regions were introduced between core and radial blanket and evolution of isotopes were studied to arrive at an optimal configuration that satisfies requirements of breeding U{sup 233} and lowering U{sup 232}concentration. SS, B{sub 4}C, SiC, Mo and W are known to be high temperature material with appropriate stability in harsh fast reactor environment. Study has shown that introducing two SS reflector rows can achieve the required low value of U{sup 232}concentration without greatly compromising the U{sup 233}production.

  14. Evaluation of Accuracy of Calculational Prediction of Criticality Based on ICSBEP Handbook Experiments

    International Nuclear Information System (INIS)

    Golovko, Yury; Rozhikhin, Yevgeniy; Tsibulya, Anatoly; Koscheev, Vladimir

    2008-01-01

    Experiments with plutonium, low enriched uranium and uranium-233 from the ICSBEP Handbook are being considered in this paper. Among these experiments it was selected only those, which seem to be the most relevant to the evaluation of uncertainty of critical mass of mixtures of plutonium or low enriched uranium or uranium-233 with light water. All selected experiments were examined and covariance matrices of criticality uncertainties were developed along with some uncertainties were revised. Statistical analysis of these experiments was performed and some contradictions were discovered and eliminated. Evaluation of accuracy of prediction of criticality calculations was performed using the internally consistent set of experiments with plutonium, low enriched uranium and uranium-233 remained after the statistical analyses. The application objects for the evaluation of calculational prediction of criticality were water-reflected spherical systems of homogeneous aqueous mixtures of plutonium or low enriched uranium or uranium-233 of different concentrations which are simplified models of apparatus of external fuel cycle. It is shows that the procedure allows to considerably reduce uncertainty in k eff caused by the uncertainties in neutron cross-sections. Also it is shows that the results are practically independent of initial covariance matrices of nuclear data uncertainties. (authors)

  15. Consistent Set of Experiments from ICSBEP Handbook for Evaluation of Criticality Calculation Prediction of Apparatus of External Fuel Cycle with Different Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Golovko, Yury E. [FSUE ' SSC RF-IPPE' , 249033, Bondarenko Square 1, Obninsk (Russian Federation)

    2008-07-01

    Experiments with plutonium, low enriched uranium and uranium-233 from the ICSBEP1 Handbook are being considered in this paper. Among these experiments it was selected only those, which seem to be the most relevant to the evaluation of uncertainty of critical mass of mixtures of plutonium or low enriched uranium or uranium-233 with light water. All selected experiments were examined and covariance matrices of criticality uncertainties were developed along with some uncertainties were revised. Statistical analysis of these experiments was performed and some contradictions were discovered and eliminated. Evaluation of accuracy of prediction of criticality calculations was performed using the internally consistent set of experiments with plutonium, low enriched uranium and uranium-233 remained after the statistical analyses. The application objects for the evaluation of calculational prediction of criticality were water-reflected spherical systems of homogeneous aqueous mixtures of plutonium or low enriched uranium or uranium-233 of different concentrations which are simplified models of apparatus of external fuel cycle. It is shows that the procedure allows to considerably reduce uncertainty in k{sub eff} caused by the uncertainties in neutron cross-sections. Also it is shows that the results are practically independent of initial covariance matrices of nuclear data uncertainties. (authors)

  16. Thorex reprocessing characterization

    International Nuclear Information System (INIS)

    1978-11-01

    The purpose of this report is to bring together, in highly condensed form, information which would need to be considered in planning a commercial reprocessing plant for recovering 233 U-Th reactor fuel. This report does not include a discussion of process modifications which would be required for thorium-base fuels that contain plutonium (such as would be required for thorium fuels containing 235 U or 233 U denatured with 238 U). It is the intent of this paper to address only the basic Thorex process for treating 233 U-Th fuels. As will be pointed out, the degree of development of the various proposed operations varies widely, from preliminary laboratory experiments for the dissolution of Zircaloy-clad thoria to engineering scale demonstration of the recovery of moderately irradiated thorium by a solvent extraction process (Thorex)

  17. Influence of moderator to fuel ratio (MFR) on burning thorium in a subcritical assembly

    International Nuclear Information System (INIS)

    Wojciechowski, Andrzej

    2014-01-01

    The conversion ratio (CR) of Th-232 to U-233 calculation results for a subcritical reactor assembly is presented as a function of MFR, burnup, power density (PD) and fissile concentration. The calculated model is based on subcritical assembly which makes configuration of fuel rods and volumes of moderator and coolant changes possible. This comfortable assembly enables investigation of CR in a thorium cycle for different value of MFR. Additionally, the calculation results of U-233 saturation concentration are explained by mathematical model. The value of MFR main influences the saturation concentration of U-233 and fissile and the fissile concentration dependence of CR. The saturation value of CR is included in the range CR ∈ (0.911, 0.966) and is a slowly increasing function of MFR. The calculations were done with a MCNPX 2.7 code

  18. Minor Actinide Burning in Thermal Reactors. A Report by the Working Party on Scientific Issues of Reactor Systems

    International Nuclear Information System (INIS)

    Hesketh, K.; Porsch, D.; Rimpault, G.; Taiwo, T.; Worrall, A.

    2013-01-01

    The actinides (or actinoids) are those elements in the periodic table from actinium upwards. Uranium (U) and plutonium (Pu) are two of the principal elements in nuclear fuel that could be classed as major actinides. The minor actinides are normally taken to be the triad of neptunium (Np), americium (Am) and curium (Cm). The combined masses of the remaining actinides (i.e. actinium, thorium, protactinium, berkelium, californium, einsteinium and fermium) are small enough to be regarded as very minor trace contaminants in nuclear fuel. Those elements above uranium in the periodic table are known collectively as the transuranics (TRUs). The operation of a nuclear reactor produces large quantities of irradiated fuel (sometimes referred to as spent fuel), which is either stored prior to eventual deep geological disposal or reprocessed to enable actinide recycling. A modern light water reactor (LWR) of 1 GWe capacity will typically discharge about 20-25 tonnes of irradiated fuel per year of operation. About 93-94% of the mass of uranium oxide irradiated fuel is comprised of uranium (mostly 238 U), with about 4-5% fission products and ∼1% plutonium. About 0.1-0.2% of the mass is comprised of neptunium, americium and curium. These latter elements accumulate in nuclear fuel because of neutron captures, and they contribute significantly to decay heat loading and neutron output, as well as to the overall radio-toxic hazard of spent fuel. Although the total minor actinide mass is relatively small - approximately 20-25 kg per year from a 1 GWe LWR - it has a disproportionate impact on spent fuel disposal, and thus the longstanding interest in transmuting these actinides either by fission (to fission products) or neutron capture in order to reduce their impact on the back end of the fuel cycle. The combined masses of the trace actinides actinium, thorium, protactinium, berkelium and californium in irradiated LWR fuel are only about 2 parts per billion, which is far too low for

  19. Safety and Performance Achievement of Indian Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kumar, Randhir

    2011-01-01

    The Nuclear Power Programme in India is based on three stage. The first stage is based on setting up of Pressurized Heavy Water Reactors (PHWRs) using indigenously available natural uranium producing electricity and plutonium. This will be followed by the second stage by plutonium fuelled Fast Breeder Reactors (FBRs) producing electricity and additional quantity of plutonium and also uranium 233 from thorium. The third stage of reactors will be based on thorium uranium 233 cycle.

  20. The Ascoli property for function spaces and the weak topology of Banach and Fréchet spaces

    Czech Academy of Sciences Publication Activity Database

    Gabriyelyan, S.; Kąkol, Jerzy; Plebanek, G.

    2016-01-01

    Roč. 233, č. 2 (2016), s. 119-139 ISSN 0039-3223 R&D Projects: GA ČR GF16-34860L Institutional support: RVO:67985840 Keywords : locally convex-space Subject RIV: BA - General Mathematics Impact factor: 0.535, year: 2016 https://www.impan.pl/pl/wydawnictwa/czasopisma-i-serie-wydawnicze/studia-mathematica/all/233/2/91577/the-ascoli-property-for-function-spaces- and -the-weak-topology-of-banach- and -frechet-spaces

  1. Protactinium-231 found in natural thorium irradiated in JMTR

    International Nuclear Information System (INIS)

    Suzuki, Susumu; Mitsugashira, Toshiaki; Hara, Mitsuo; Satoh, Isamu; Shiokawa, Yoshinobu; Satoh, Michiko

    1987-01-01

    Natural thorium dioxides, which differed in the content of 230 Th, were irradiated in JMTR(Japan Material Testing Reactor). 232 U, 233 U, 231 Pa, 233 Pa, and remaining Th were measured radiometrically. High production of 231 Pa and high consumption of 230 Th were observed and it was necessary to assume large resonance capture of 230 Th in order to explain the production of 231 Pa and the consumption of 230 Th. (author)

  2. Electronic structure and dynamics of ordered clusters with ME or RE ions on oxide surface

    Energy Technology Data Exchange (ETDEWEB)

    Kulagin, N.A., E-mail: nkulagin@bestnet.kharkov.u [Kharkiv National University for Radio Electronics, Avenue Shakespeare 6-48, 61045 Kharkiv (Ukraine)

    2011-03-15

    Selected data of ab initio simulation of the electronic structure and spectral properties of either cluster with ions of iron, rare earth or actinium group elements have been presented here. Appearance of doped Cr{sup +4} ions in oxides, Cu{sup +2} in HTSC, Nd{sup +2} in solids has been discussed. Analysis of experimental data for plasma created ordered structures of crystallites with size of about 10{sup -9} m on surface of separate oxides are given, too. Change in the spectroscopic properties of clusters and nano-structures on surface of strontium titanate crystals discussed shortly using the X-ray line spectroscopy experimental results. - Research highlights: External influence and variation of technology induce changes in valence of nl ions in compounds. Wave function of cluster presented as anti-symmetrical set of ions wave functions. The main equation describes the self-consistent field depending on state of all electrons of cluster. Level scheme of Cr{sup 4+} ions in octo- and tetra-site corresponds to doped oxides spectra after treatment. Plasma treatment effects in appearance of systems of unit crystallites with size of about 10{sup -6}-10{sup -9} m.

  3. Electronic structure and dynamics of ordered clusters with ME or RE ions on oxide surface

    International Nuclear Information System (INIS)

    Kulagin, N.A.

    2011-01-01

    Selected data of ab initio simulation of the electronic structure and spectral properties of either cluster with ions of iron, rare earth or actinium group elements have been presented here. Appearance of doped Cr +4 ions in oxides, Cu +2 in HTSC, Nd +2 in solids has been discussed. Analysis of experimental data for plasma created ordered structures of crystallites with size of about 10 -9 m on surface of separate oxides are given, too. Change in the spectroscopic properties of clusters and nano-structures on surface of strontium titanate crystals discussed shortly using the X-ray line spectroscopy experimental results. - Research highlights: → External influence and variation of technology induce changes in valence of nl ions in compounds. → Wave function of cluster presented as anti-symmetrical set of ions wave functions. → The main equation describes the self-consistent field depending on state of all electrons of cluster. → Level scheme of Cr 4+ ions in octo- and tetra-site corresponds to doped oxides spectra after treatment. → Plasma treatment effects in appearance of systems of unit crystallites with size of about 10 -6 -10 -9 m.

  4. Review of radionuclide source terms used for performance-assessment analyses

    International Nuclear Information System (INIS)

    Barnard, R.W.

    1993-06-01

    Two aspects of the radionuclide source terms used for total-system performance assessment (TSPA) analyses have been reviewed. First, a detailed radionuclide inventory (i.e., one in which the reactor type, decay, and burnup are specified) is compared with the standard source-term inventory used in prior analyses. The latter assumes a fixed ratio of pressurized-water reactor (PWR) to boiling-water reactor (BWR) spent fuel, at specific amounts of burnup and at 10-year decay. TSPA analyses have been used to compare the simplified source term with the detailed one. The TSPA-91 analyses did not show a significant difference between the source terms. Second, the radionuclides used in source terms for TSPA aqueous-transport analyses have been reviewed to select ones that are representative of the entire inventory. It is recommended that two actinide decay chains be included (the 4n+2 ''uranium'' and 4n+3 ''actinium'' decay series), since these include several radionuclides that have potentially important release and dose characteristics. In addition, several fission products are recommended for the same reason. The choice of radionuclides should be influenced by other parameter assumptions, such as the solubility and retardation of the radionuclides

  5. Renal uptake of bismuth-213 and its contribution to kidney radiation dose following administration of actinium-225-labeled antibody

    Energy Technology Data Exchange (ETDEWEB)

    Schwartz, J; O' Donoghue, J A; Humm, J L [Department of Medical Physics, Memorial Sloan-Kettering Cancer Center, 1275 York Avenue, New York, NY 10065 (United States); Jaggi, J S [Bristol-Myers Squibb, Plainsboro, NJ (United States); Ruan, S; Larson, S M [Nuclear Medicine Service Department of Radiology, Memorial Sloan-Kettering Cancer Center, 1275 York Avenue, New York, NY 10065 (United States); McDevitt, M; Scheinberg, D A, E-mail: schwarj1@mskcc.org [Molecular Pharmacology and Chemistry, Sloan-Kettering Institute, 1275 York Avenue, New York, NY 10065 (United States)

    2011-02-07

    Clinical therapeutic studies using {sup 225}Ac-labeled antibodies have begun. Of major concern is renal toxicity that may result from the three alpha-emitting progeny generated following the decay of {sup 225}Ac. The purpose of this study was to determine the amount of {sup 225}Ac and non-equilibrium progeny in the mouse kidney after the injection of {sup 225}Ac-huM195 antibody and examine the dosimetric consequences. Groups of mice were sacrificed at 24, 96 and 144 h after injection with {sup 225}Ac-huM195 antibody and kidneys excised. One kidney was used for gamma ray spectroscopic measurements by a high-purity germanium (HPGe) detector. The second kidney was used to generate frozen tissue sections which were examined by digital autoradiography (DAR). Two measurements were performed on each kidney specimen: (1) immediately post-resection and (2) after sufficient time for any non-equilibrium excess {sup 213}Bi to decay completely. Comparison of these measurements enabled estimation of the amount of excess {sup 213}Bi reaching the kidney ({gamma}-ray spectroscopy) and its sub-regional distribution (DAR). The average absorbed dose to whole kidney, determined by spectroscopy, was 0.77 (SD 0.21) Gy kBq{sup -1}, of which 0.46 (SD 0.16) Gy kBq{sup -1} (i.e. 60%) was due to non-equilibrium excess {sup 213}Bi. The relative contributions to renal cortex and medulla were determined by DAR. The estimated dose to the cortex from non-equilibrium excess {sup 213}Bi (0.31 (SD 0.11) Gy kBq{sup -1}) represented {approx}46% of the total. For the medulla the dose contribution from excess {sup 213}Bi (0.81 (SD 0.28) Gy kBq{sup -1}) was {approx}80% of the total. Based on these estimates, for human patients we project a kidney-absorbed dose of 0.28 Gy MBq{sup -1} following administration of {sup 225}Ac-huM195 with non-equilibrium excess {sup 213}Bi responsible for approximately 60% of the total. Methods to reduce renal accumulation of radioactive progeny appear to be necessary for the success of {sup 225}Ac radioimmunotherapy.

  6. Distribution of trace elements in land plants and botanical taxonomy with special reference to rare earth elements and actinium

    International Nuclear Information System (INIS)

    Koyama, Mutsuo

    1989-01-01

    Distribution profiles of trace elements in land plants were studied by neutron activation analysis and radioactivity measurements without activation. Number of botanical samples analyzed were more than three thousand in which more than three hundred botanical species were included. New accumulator plants of Co, Cr, Zn, Cd, rare earth elements, Ac, U, etc., were found. Capabilities of accumulating trace elements can be related to the botanical taxonomy. Discussions are given from view points of inorganic chemistry as well as from botanical physiology

  7. An Investigation of the Combat Air Patrol Stationing in an Integrated Air Defense Scenario

    Science.gov (United States)

    1990-12-01

    guided missiles, drones , decoys and short range missiles and bombs. The defensive elements include a C2 netting of early warning (EW), ground...M, 69 andar 70045 Brasilia, DF Brasil 6. Departamento de Ensino Da Aerondutica Av. Mal CAmara , 233, 10 andar 20020 Rio de Janeiro, RJ Brasil 7...Diretoria de Informdtica e Estatfstica da Aeron~utica Av. Mal CAmara , 233, 8Q andar 20020 Rio de Janeiro, RJ Brasil 11. Centro de Computagdo da Aeronutica-BR

  8. W-007H B Plant Process Condensate Treatment Facility. Revision 3

    International Nuclear Information System (INIS)

    Rippy, G.L.

    1995-01-01

    B Plant Process Condensate (BCP) liquid effluent stream is the condensed vapors originating from the operation of the B Plant low-level liquid waste concentration system. In the past, the BCP stream was discharged into the soil column under a compliance plan which expired January 1, 1987. Currently, the BCP stream is inactive, awaiting restart of the E-23-3 Concentrator. B Plant Steam Condensate (BCS) liquid effluent stream is the spent steam condensate used to supply heat to the E-23-3 Concentrator. The tube bundles in the E-23-3 Concentrator discharge to the BCS. In the past, the BCS stream was discharged into the soil column. Currently, the BCS stream is inactive. This project shall provide liquid effluent systems (BCP/BCS/BCE) capable of operating for a minimum of 20 years, which does not include the anticipated decontamination and decommissioning (D and D) period

  9. Spin coating and plasma process for 2.5D integrated photonics on multilayer polymers

    International Nuclear Information System (INIS)

    Zebda, A.; Camberlein, L.; Beche, B.; Gaviot, E.; Beche, E.; Duval, D.; Zyss, J.; Jezequel, G.; Solal, F.; Godet, C.

    2008-01-01

    Polymer spin coating, surface plasma treatment and selective UV-lithography processes have been developed to realize 2.5D photonic micro-resonators, made of disk- or ring-shaped upper rib waveguides, using common polymers such as SU8 (biphenol A ether glycidyl), PS233 (polymeric silane) and SOG (siloxane Spin on Glass). Both oxygen and argon plasma treatments, applied to PS233 and SOG before spin-coating the SU8, improve substantially the grip of multilayer devices (SU8 / PS233 or SU8 / SOG). Surface energy components derived from contact angle measurements have been used to optimize the processing conditions. In such integrated photonic devices, the both single-electromagnetic-modes called transverse electric (TE 00 ) and transverse magnetic (TM 00 ) have been excited in a SU8 micro-disk, with a single mode propagation strongly localized near the edge of the disk (i.e. the so called whispering gallery modes)

  10. W-007H B Plant Process Condensate Treatment Facility. Revision 3

    Energy Technology Data Exchange (ETDEWEB)

    Rippy, G.L.

    1995-01-20

    B Plant Process Condensate (BCP) liquid effluent stream is the condensed vapors originating from the operation of the B Plant low-level liquid waste concentration system. In the past, the BCP stream was discharged into the soil column under a compliance plan which expired January 1, 1987. Currently, the BCP stream is inactive, awaiting restart of the E-23-3 Concentrator. B Plant Steam Condensate (BCS) liquid effluent stream is the spent steam condensate used to supply heat to the E-23-3 Concentrator. The tube bundles in the E-23-3 Concentrator discharge to the BCS. In the past, the BCS stream was discharged into the soil column. Currently, the BCS stream is inactive. This project shall provide liquid effluent systems (BCP/BCS/BCE) capable of operating for a minimum of 20 years, which does not include the anticipated decontamination and decommissioning (D and D) period.

  11. Publications | Page 233 | CRDI - Centre de recherches pour le ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Compostage artisanale d'ordure ménagère à Pouytenga : risques de contamination et stratégie de leur réduction (restricted access) · Sahel agroforesterie, no. 6, avril - juin 2006 (open access) · Numérisation et mise sur Internet des ressources documentaires de l'IFAN Cheikh Anta Diop : rapport technique final; 2002-2006 ...

  12. 45 CFR 233.20 - Need and amount of assistance.

    Science.gov (United States)

    2010-10-01

    ... section 236 of the National Housing Act” means Department of Housing and Urban Development assisted... debts); Fair market value means the price an item of a particular make, model, size, material or... agency establishes policy under which assistance from other agencies and organizations will not be...

  13. 233 DETERMINANTS OF CAPITAL FLIGH: THE CASE OF NIGERIA ...

    African Journals Online (AJOL)

    Through the least square regression analysis, this study constructs a model ..... β1……. β10 = (betas) are the regression coefficients or the slope parameters for the ... The term Ut, otherwise known as the stochastic term of the regression is ...

  14. 233 Meaning and the Second Language Learner Jane Nkechi ...

    African Journals Online (AJOL)

    Ike Odimegwu

    second language learner seeks to interpret word meaning without reference to the .... Meaning can be natural but language is conventional - that is there is no .... language learner is usually engaged in processing contextual information to ...

  15. All projects related to | Page 233 | IDRC - International Development ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Assessing the Impact of Current National Policies to Reduce Salt and Trans Fatty ... to the Center Stage in Mexico: Case Study for World Development Report 2013 ... and the impact of employment on well-being, social inclusion, and political ...

  16. 42 CFR 456.233 - Initial continued stay review date.

    Science.gov (United States)

    2010-10-01

    ... SERVICES (CONTINUED) MEDICAL ASSISTANCE PROGRAMS UTILIZATION CONTROL Utilization Control: Mental Hospitals... must provide that— (a) When a recipient is admitted to the mental hospital under admission review... will be reviewed; (b) If an individual applies for Medicaid while in the mental hospital, the committee...

  17. Comparison of the response (in terms of accumulation, cellular and genetic impacts) of the crayfish Procambarus clarkii after exposure to a metallic pollutant (cadmium) and to a radiological pollutant (uranium 238 and 233)

    International Nuclear Information System (INIS)

    Al Kaddissi, S.

    2012-01-01

    The study of the effects of radionuclides and metals on organisms is necessary for the evaluation of their toxicity and their ecological threats. We first aimed to study the impacts of cadmium (Cd) and Uranium (U) on different biological levels of the crayfish Procambarus clarkii after acute and chronic exposures. We evaluated their impacts on mitochondria, oxidative stress responses, on histological structures, and the survival rates. We tried to connect these effects between them and to the bioaccumulation in the gills and the hepato pancreas. We also tried to discriminate the chemo and the radiotoxicity of U by exposing crayfish to either depleted or enriched U ( 233 U: presenting a higher specific activity) using the same criteria of effects. We demonstrated that the gene mt encoding for the metallothionein was always over-expressed in the presence of Cd. Therefore, it seems to be a good bio-marker of Cd toxicity in P. clarkii. The follow up of mitochondrial genes expressions (12s, atp6 and cox1), showed that both metals affect mitochondria and that their mechanisms of action do not seem to be always the same. We also observed that U generates more oxidative stress than Cd when comparing the expression levels of genes encoding for antioxidants (sod (Mn) and mt) and the enzymatic activities of superoxide dismutase, the catalase, the glutathione peroxidase and the glutathione S transferase. However, the symptoms of histo-pathological damages after Cd and U contamination were similar in both conditions. After comparing the survival rates of the crayfish, we concluded that Cd was more toxic than the radioelement. Moreover, we demonstrated that the toxic effect of U on P. clarkii exposed to a low environmental concentration is mainly due to its chemo-toxicity rather than to its radiotoxicity. We established that, the molecular answers vary according to the intensity and the duration of the chemical stress applied to the organisms. We suggested the use of the

  18. The thorium fuel cycle

    International Nuclear Information System (INIS)

    Merz, E.R.

    1977-01-01

    The utilization of the thorium fuel cycle has long since been considered attractive owing to the excellent neutronic characteristics of 233 U, and the widespread and cheap thorium resources. Rapidly increasing uranium prices, public reluctance for widespread Pu recycling and expected delays for the market penetration of fast breeders have led to a reconsideration of the thorium fuel cycle merits. In addition, problems associated with reprocessing and waste handling, particularly with re-fabrication by remote handling of 233 U, are certainly not appreciably more difficult than for Pu recycling. To divert from uranium as a nuclear energy source it seems worth while intensifying future efforts for closing the Th/ 233 U fuel cycle. HTGRs are particularly promising for economic application. However, further research and development activities should not concentrate on this reactor type alone. Light- and heavy-water-moderated reactors, and even future fast breeders, may just as well take advantage of a demonstrated thorium fuel cycle. (author)

  19. Electron-capture delayed fission properties of neutron-deficient einsteinium nuclei

    International Nuclear Information System (INIS)

    Shaughnessy, Dawn A.

    2000-01-01

    Electron-capture delayed fission (ECDF) properties of neutron-deficient einsteinium isotopes were investigated using a combination of chemical separations and on-line radiation detection methods. 242 Es was produced via the 233 U( 14 N,5n) 242 Es reaction at a beam energy of 87 MeV (on target) in the lab system, and was found to decay with a half-life of 11 ± 3 seconds. The ECDF of 242 Es showed a highly asymmetric mass distribution with an average pre-neutron emission total kinetic energy (TKE) of 183 ± 18 MeV. The probability of delayed fission (P DF ) was measured to be 0.006 ± 0.002. In conjunction with this experiment, the excitation functions of the 233 U( 14 N,xn) 247-x Es and 233 U( 15 N,xn) 248-x Es reactions were measured for 243 Es, 244 Es and 245 Es at projectile energies between 80 MeV and 100 MeV

  20. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    Lowry, L.L.

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

  1. 800-MeV proton irradiation of thorium and depleted uranium targets

    Energy Technology Data Exchange (ETDEWEB)

    Russell, G.J.; Brun, T.O.; Pitcher, E.J. [Los Alamos National Laboratory, NM (United States)] [and others

    1995-10-01

    As part of the Los Alamos Fertile-to-Fissile-Conversion (FERFICON) program in the late 1980`s, thick targets of the fertile materials thorium and depleted uranium were bombarded by 800-MeV protons to produce the fissile materials {sup 233}U and {sup 239}Pu, respectively. The amount of {sup 233}U made was determined by measuring the {sup 233}Pa activity, and the yield of {sup 239}Pu was deduced by measuring the activity of {sup 239}Np. For the thorium target, 4 spallation products and 34 fission products were also measured. For the depleted uranium target, 3 spallation products and 16 fission products were also measured. The number of fissions in each target was deduced from fission product mass-yield curves. In actuality, axial distributions of the products were measured, and the distributions were then integrated over the target volume to obtain the total number of products for each reaction.

  2. Helicobacter Pylori Related Functional Dyspepsia in a Defined Malaysian Population

    OpenAIRE

    Nafeeza, M.I.; Isa, M.R.; Kudva, M.V.; Ishak, M.S.; Mazlam, M.Z.; Haron, A.; Najib, R.; Shahimi, M.M.

    2000-01-01

    The objective of the study was to determine the prevalence of H. pylori in functional dyspepsia among the three main races in Malaysia. Gastric antral biopsies from 233 (98 males, 135 females; age range: 17–75 years, mean age 39.5 years) patients attending the Universiti Kebangsaan Malaysia (UKM) gastroenterology clinic were assessed for the presence of H. pylori by culture and histology. About a third of the cases (79 of 233 (34%); 34 males, 45 females; mean age 42.6 yrs) were positive for H...

  3. Photoneutron and Photonuclear Cross Sections According to Packed cluster Model

    International Nuclear Information System (INIS)

    El-Mekkawi, L.S.; El-Bakty, O.M.

    1998-01-01

    Photonuclear gross sections have been estimated for 232 Th, 237 Np, 239 Pu, 233 U, 234 U, 235 U, 238 U in the energy range from threshold up to 20 MeV, by perturbation balance in Packed Cluster. The Packed Cluster (gamma, f) and (gamma, n) cross sections require complete absence of any (gamma,2n) or (gamma,nf) cross sections for 233 U and 234 U as in experiment. It also explains the early (gamma,n) and gamma,nf) reactions in 235 U

  4. Design of a thorium fuelled Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    2009-01-01

    Full text: The main objective for development of Advanced Heavy Water Reactor (AHWR) is to demonstrate thorium fuel cycle technologies, along with several other advanced technologies required for next generation reactors, so that these are readily available in time for launching the third stage. The AHWR under design is a 300 MWe vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The fuel consists of (Th-Pu)O 2 and ( 233 ThU)O 2 pins. The fuel cluster is designed to generate maximum energy out of 233 U, which is bred in-situ from thorium and has a slightly negative void coefficient of reactivity, negative fuel temperature coefficient and negative power coefficient. For the AHWR, the well -proven pressure tube technology and online fuelling have been adopted. Core heat removal is by natural circulation of coolant during normal operation and shutdown conditions. Thus, it combines the advantages of light water reactors and PHWRs and removes the disadvantages of PHWRs. It has several passive safety systems for reactor normal operation, decay heat removal, emergency core cooling, confinement of radioactivity etc. The fuel cycle is based on the in-situ conversion of naturally available thorium into fissile 233 U in self sustaining mode. The uranium in the spent fuel will be reprocessed and recycled back into the reactor. The plutonium inventory will be kept a minimum and will come from fuel irradiated in Indian PHWRs. The 233 U required initially can come from the fast reactor programme or it can be produced by specially designing the initial core of AHWR using (Th,Pu)MOX fuel. There will be gradual transition from the initial core which will not contain any 233 U to an equilibrium core, which will have ( 233 U, Th) MOX fuel pins also in a composite cluster. The self sustenance is being achieved by a differential fuel loading of low and a relatively higher Pu in the composite clusters. The AHWR burns the

  5. Methodology of simultaneous analysis of Uranium and Thorium by nuclear and atomic techniques. Application to the Uranium and Thorium dosing in mineralogic samples

    International Nuclear Information System (INIS)

    Fakhi, S.

    1988-01-01

    This work concerns essentially the potential applications of 100 kW nuclear reactor of Strasbourg Nuclear Research Centre to neutron activation analysis of Uranium and Thorium. The Uranium dosing has been made using: 239-U, 239-Np, fission products or delayed neutrons. Thorium has been showed up by means of 233-Th or 233-Pa. The 239-U and 233-Th detection leads to a rapid and non-destructive analysis of Uranium and Thorium. The maximum sensitivity is of 78 ng for Uranium and of 160 ng for Thorium. The Uranium and Thorium dosing based on 239-Np and 233-Pa detection needs chemical selective separations for each of these radionuclides. The liquid-liquid extraction has permitted to elaborate rapid and quantitative separation methods. The sensitivities of the analysis after extraction reach 30 ng for Uranium and 50 ng for Thorium. The fission products separation study has allowed to elaborate the La, Ce and Nd extractions and its application to the Uranium dosing gives satisfying results. A rapid dosing method with a sensitivity of 0.35 microgramme has been elaborated with the help of delayed neutrons measurement. These different methods have been applied to the Uranium and Thorium dosing in samples coming from Oklo mine in Gabon. The analyses of these samples by atomic absorption spectroscopy and by the proton induced X-ray emission (PIXE) method confirm that the neutron activation analysis methods are reliable. 37 figs., 14 tabs., 50 refs

  6. ORF Alignment: NC_003279 [GENIUS II[Archive

    Lifescience Database Archive (English)

    Full Text Available PTFIG 60 ... Query: 174 MQTSPTAPTTLRACAPMIYKQEGLTGFFKGLPPLWTRQIPYTMMKFTCFEKTVELLYQYV 233 ... MQTSPTAPTTLRACAPM...IYKQEGLTGFFKGLPPLWTRQIPYTMMKFTCFEKTVELLYQYV Sbjct: 121 MQTSPTAPTTLRACAPMIYK

  7. Diabetic peripheral angiopathy treatment using a multi-laser therapy device

    OpenAIRE

    Zabulonov, Y.; Chukhraiyeva, O.; Vladimirov, A.; Chukhraiyev, M.; Zukow, W.

    2015-01-01

    Zabulonov Y., Chukhraiyeva O., Vladimirov A., Chukhraiyev M., Zukow W. Diabetic peripheral angiopathy treatment using a multi-laser therapy device. Journal of Education, Health and Sport. 2015;5(10):227-233. ISSN 2391-8306. DOI http://dx.doi.org/10.5281/zenodo.32801 http://ojs.ukw.edu.pl/index.php/johs/article/view/2015%3B5%2810%29%3A227-233 https://pbn.nauka.gov.pl/works/662978 Formerly Journal of Health Sciences. ISSN 1429-9623 / 2300-665X. Archives 2011–2014 http://journal.rsw....

  8. Diabetic peripheral angiopathy treatment using a multi-laser therapy device

    OpenAIRE

    Y. Zabulonov; O. Chukhraiyeva; A. Vladimirov; M. Chukhraiyev; W. Zukow

    2015-01-01

    Zabulonov Y., Chukhraiyeva O., Vladimirov A., Chukhraiyev M., Zukow W. Diabetic peripheral angiopathy treatment using a multi-laser therapy device. Journal of Education, Health and Sport. 2015;5(10):227-233. ISSN 2391-8306. DOIhttp://dx.doi.org/10.5281/zenodo.32801 http://ojs.ukw.edu.pl/index.php/johs/article/view/2015%3B5%2810%29%3A227-233 https://pbn.nauka.gov.pl/works/662978 Formerly Journal of Health Sciences. ISSN 1429-9623 / 2300-665X. Archives 2011–2014http://journal.rsw.ed...

  9. Report to Congress on Sustainable Ranges, 2015

    Science.gov (United States)

    2015-03-01

    107th FS at Selfridge ANGB MI, F-16 at Toledo ANGB OH, A-10 at Fort Wayne ANGB IN, and all units deployed in training at Alpena CRTC. The range also...Sheppard AFB, TX 76311 DSN 736- 2675/4995, C817-676-2675/4995. Sunrise-Sunset Mon-Fri, OT by NOTAM 233 VR1624 ALPENA CRTC/OTM, 5884 A. Sreet, Alpena , MI...49707-8125 DSN 741-6509/6226. Same as Originating Activity Sunrise-Sunset 233 VR1625 ALPENA CRTC/OTM, 5884 A. Sreet, Alpena , MI 49707-8125 DSN 741

  10. Determination of thorium and uranium at the nanogram per gram level in semiconductor potting plastics by neutron activation analysis

    International Nuclear Information System (INIS)

    Dyer, F.F.; Emery, J.F.; Bate, L.C.

    1985-01-01

    A method was developed to determine thorium and uranium in semiconductor potting plastics. The method is based on neutron activation and subsequent radiochemical separation to isolate and permit measurement of the induced 233 Pa and 239 Np. These plastics typically contain macro amounts of silicon, bromine and antimony and nanogram per gram amounts of thorium and uranium. The radiochemical method provides the necessary sensitivity and makes it possible to easily attain adequate decontamination of the tiny amounts of 233 Pa and 239 Np from the high levels of radioactive bromine and antimony. 8 refs

  11. Burnup characteristics of binary breeder reactors

    International Nuclear Information System (INIS)

    Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.

    1983-01-01

    Burnup calculations of a binary breeder reactor have been done for two cases of fueling. In one case the U 233 /TH fueled inner core and the Pu/U-fueled outer core have the same number of fuel assemblies. In the other case two outermost rings in the inner core are Pu/U-fueled. The second case is considered for an initial phase of thorim cycle introduction when the supply of U 233 could be limited. Results show an efficient breeding on the thorium cycle in both cases. (Author) [pt

  12. ORF Alignment: NC_005363 [GENIUS II[Archive

    Lifescience Database Archive (English)

    Full Text Available ICQQLGHSVVLVNHRGAGEGAPFAKRPYHS 60 ... Query: 174 SGSLLLKSGFNRVYDMRFVLRLRKLVEEKHRLGLITEKYEIPKWATVWDMDQIYTAPASG 2...33 ... SGSLLLKSGFNRVYDMRFVLRLRKLVEEKHRLGLITEKYEIPKWATVWDMDQIYTAPASG Sbjct: 121 SGSLLLKSGFNRVYDMRFVLRLRKLVEEKHRLGLITEKYEIPKWATVWDMDQIYTAPASG 180 ...

  13. Insights into phosphate cooperativity and influence of substrate modifications on binding and catalysis of hexameric purine nucleoside phosphorylases.

    Directory of Open Access Journals (Sweden)

    Priscila O de Giuseppe

    Full Text Available The hexameric purine nucleoside phosphorylase from Bacillus subtilis (BsPNP233 displays great potential to produce nucleoside analogues in industry and can be exploited in the development of new anti-tumor gene therapies. In order to provide structural basis for enzyme and substrates rational optimization, aiming at those applications, the present work shows a thorough and detailed structural description of the binding mode of substrates and nucleoside analogues to the active site of the hexameric BsPNP233. Here we report the crystal structure of BsPNP233 in the apo form and in complex with 11 ligands, including clinically relevant compounds. The crystal structure of six ligands (adenine, 2'deoxyguanosine, aciclovir, ganciclovir, 8-bromoguanosine, 6-chloroguanosine in complex with a hexameric PNP are presented for the first time. Our data showed that free bases adopt alternative conformations in the BsPNP233 active site and indicated that binding of the co-substrate (2'deoxyribose 1-phosphate might contribute for stabilizing the bases in a favorable orientation for catalysis. The BsPNP233-adenosine complex revealed that a hydrogen bond between the 5' hydroxyl group of adenosine and Arg(43* side chain contributes for the ribosyl radical to adopt an unusual C3'-endo conformation. The structures with 6-chloroguanosine and 8-bromoguanosine pointed out that the Cl(6 and Br(8 substrate modifications seem to be detrimental for catalysis and can be explored in the design of inhibitors for hexameric PNPs from pathogens. Our data also corroborated the competitive inhibition mechanism of hexameric PNPs by tubercidin and suggested that the acyclic nucleoside ganciclovir is a better inhibitor for hexameric PNPs than aciclovir. Furthermore, comparative structural analyses indicated that the replacement of Ser(90 by a threonine in the B. cereus hexameric adenosine phosphorylase (Thr(91 is responsible for the lack of negative cooperativity of phosphate binding

  14. Assembly-level analysis of heterogeneous Th–Pu PWR fuel

    International Nuclear Information System (INIS)

    Zainuddin, Nurjuanis Zara; Parks, Geoffrey T.; Shwageraus, Eugene

    2017-01-01

    Highlights: • We directly compare homogeneous and heterogeneous Th–Pu fuel. • Examine whether there is an increase in Pu incineration in the latter. • Homogeneous fuel was able to achieve much higher Pu incineration. • In the heterogeneous case, U-233 breeding is faster (larger power fraction), thus decreasing incineration of Pu. - Abstract: This study compares homogeneous and heterogeneous thorium–plutonium (Th–Pu) fuel assemblies (with high Pu content – 20 wt%), and examines whether there is an increase in Pu incineration in the latter. A seed-blanket configuration based on the Radkowsky thorium reactor concept is used for the heterogeneous assembly. This separates the thorium blanket from the uranium seed, or in this case a plutonium seed. The seed supplies neutrons to the subcritical thorium blanket which encourages the in situ breeding and burning of "2"3"3U, allowing the fuel to stay critical for longer, extending burnup of the fuel. While past work on Th–Pu seed-blanket units shows superior Pu incineration compared to conventional U–Pu mixed oxide fuel, there is no literature to date that directly compares the performance of homogeneous and heterogeneous Th–Pu assembly configurations. Use of exactly the same fuel loading for both configurations allows the effects of spatial separation to be fully understood. It was found that the homogeneous fuel with and without burnable poisons was able to achieve much higher Pu incinerations than the heterogeneous fuel configurations, while still attaining a reasonably high discharge burnup. This is because in the heterogeneous cases, "2"3"3U breeding is faster, thereby contributing to a much larger fraction of total power produced by the assembly. In contrast, "2"3"3U build-up is slower in the homogeneous case and therefore Pu burning is greater. This "2"3"3U begins to contribute a significant fraction of power produced only towards the end of life, thus extending criticality, allowing more Pu to

  15. Monopole conversion hidden by penetration effect in magnetic dipole transitions

    International Nuclear Information System (INIS)

    Bikit, I.; Anichin, I.; Marinkov, L.

    1977-01-01

    The 191 keV 197 Au nad 340 keV 233 U transitions are investigated and the effect of penetration into the M1-component is accounted for. Theoretical internal conversion coefficients (ICC) and electron parameters to account for the penetration effect have been obtained by interpolating the data of the Hager and Zeltzer tables. The ICC values and ratios are analyzed under the assumption that the 191 keV 197 Au transition has multipolarities M1 + E2 and E 0 +M1. A common overlapping occurs when the nuclear penetration parameter lambda for magnetic dipole transition is lambda = 34.2+-2.2. For the 340 keV 233 U transition the ICC has been found to equal αk=0.69+-0.07, and the relative conversion-line intensities have been determined. It is concluded that the 191 keV 197 Au nad 340 keV 233 U transitions involve an electric monopole component concealed by the penetration effect in the M1-conversion. The matrix elements of the E0-transition have been evaluated

  16. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    International Nuclear Information System (INIS)

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  17. Radioactive materials production

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    The Radiochemical Processing Plant (RPP) at ORNL has served as the national repository and distribution center for 233 U for > 20 years. Several hundred kilograms of uranium, containing approximately 90 to 98% 233 U, are stored there in the form of metal, oxides, and nitrate solutions. All of these uranium materials contain small, but significant, concentrations of 232 U, ranging from 2 to 225 ppm. Most of the radioactivity associated with the 233 U comes from the decay daughters of 232 U (74-year half-life). The 252 Cf Industrial Sales/Loan Program involves loans of 252 Cf neutron sources to agencies of the US Government and sales of 252 Cf as the bulk oxide and as palladium-californium alloy pellets and wires. The program has been operated since 1968 in temporary facilities at the Savannah River Laboratory (SRL). The obsolete hot-cell facilities at SRL are now being decommissioned, and the program activities are being transferred to ORNL's Californium Facility in Bldg. 7930, which is managed by the staff of the Transuranium Processing Plant

  18. Emanations and 'induced' radioactivity: from mystery to (mis)use

    International Nuclear Information System (INIS)

    Kolar, Z.I.

    1999-01-01

    The natural Rn isotopes were discovered within the period 1899-1902 and at that time referred to as emanations because they came out (emanated) of sources/materials containing actinium, thorium and radium, respectively. The (somewhat mysterious) emanations appeared to disintegrate into radioactive decay products which by depositing at solid surfaces gave rise to 'induced' radioactivity i.e. radioactive substances with various half-lives. Following the discovery of the emanations the volume of the research involving them and their disintegration products grew steeply. The identity of a number of these radioactive products was soon established. Radium emanation was soon used as a source of RaD ( 210 Pb) to be applied as an 'indicator' (radiotracer) for lead in a study on the solubility of lead sulphide and lead chromate. Moreover, radium and its emanation were introduced into the medical practice. Inhaling radon and drinking radon-containing water became an accepted medicinal use (or misuse?) of that gas. Shortly after the turn of the century, the healing (?) action of natural springs (spas) was attributed to their radium emanation, i.e. radon. Bathing in radioactive spring water and drinking it became very popular. Even today, bathing in radon-containing water is still a common medical treatment in Jachymov, Czech Republic. (author)

  19. Study of Soil Decontamination Method Using Supercritical Carbon Dioxide and TBP

    International Nuclear Information System (INIS)

    Park, Jihye; Park, Kwangheon; Jung, Wonyoung

    2014-01-01

    The result of this study means that we have a possible new method for cheap and less wasteful nuclear waste decontamination. When severe accidents such as the incident at the Fukushima nuclear site occur, the soil near the power plant is contaminated with fission products or the activation metal structure of the power plant. The soil pollution form depends on the environment and soil characteristics of the contaminated areas. Thus, a- single-decontamination method is not effective for site cleanup. In addition, some soil decontamination methods are expensive and large amounts of secondary waste are generated. Therefore, we need new soil decontamination methods. In this study, instead of using a conventional solvent method that generates secondary waste, supercritical carbon dioxide was used to remove metal ions from the soil. Supercritical carbon dioxide is known for good permeation characteristics. We expect that we will reduce the cost of soil pollution management. Supercritical carbon dioxide can decontaminate soil easily, as it has the ability to penetrate even narrow gaps with very good moisture permeability. We used TBP, which is a known for extractant of actinium metal. TBP is usually used for uranium and strontium extraction. Using TBP-HNO 3 complex and supercritical carbon dioxide, we did extraction experiments for several heavy metals in contaminated soil

  20. Study of Soil Decontamination Method Using Supercritical Carbon Dioxide and TBP

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jihye; Park, Kwangheon; Jung, Wonyoung [Kyunghee Univ., Yongin (Korea, Republic of)

    2014-05-15

    The result of this study means that we have a possible new method for cheap and less wasteful nuclear waste decontamination. When severe accidents such as the incident at the Fukushima nuclear site occur, the soil near the power plant is contaminated with fission products or the activation metal structure of the power plant. The soil pollution form depends on the environment and soil characteristics of the contaminated areas. Thus, a- single-decontamination method is not effective for site cleanup. In addition, some soil decontamination methods are expensive and large amounts of secondary waste are generated. Therefore, we need new soil decontamination methods. In this study, instead of using a conventional solvent method that generates secondary waste, supercritical carbon dioxide was used to remove metal ions from the soil. Supercritical carbon dioxide is known for good permeation characteristics. We expect that we will reduce the cost of soil pollution management. Supercritical carbon dioxide can decontaminate soil easily, as it has the ability to penetrate even narrow gaps with very good moisture permeability. We used TBP, which is a known for extractant of actinium metal. TBP is usually used for uranium and strontium extraction. Using TBP-HNO{sub 3} complex and supercritical carbon dioxide, we did extraction experiments for several heavy metals in contaminated soil.