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Sample records for actinium 233

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

    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.

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

    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.

  3. Extraction of actinium with di-(2-ethylhexyl)phosphoric acid from hydrochloric and nitric acid solutions

    The extraction of actinium with HDEHP from Cl- and NO3- systems has been investigated. It was found that extraction of actinium from HCl solutions is much better than from HNO3 solutions. Stability constants of actinium complexes Ac(X-)+2 with Cl- and NO3- ligands were determined. Our results show that the actinium formed less stable complexes with Cl- than with NO3- ligands. 5 refs., 3 figs., 1 tab. (author)

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

    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

  5. Discovery of the actinium, thorium, protactinium, and uranium isotopes

    Fry, C; Thoennessen, M

    2012-01-01

    Currently, 31 actinium, 31 thorium, 28 protactinium, and 23 uranium isotopes have so far been observed; the discovery of these isotopes is discussed. For each isotope a brief summary of the first refereed publication, including the production and identification method, is presented.

  6. Separation of Actinium 227 from the uranium minerals

    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. Spectroscopic and computational investigation of actinium coordination chemistry.

    Ferrier, Maryline G; Batista, Enrique R; Berg, John M; Birnbaum, Eva R; Cross, Justin N; Engle, Jonathan W; La Pierre, Henry S; Kozimor, Stosh A; Lezama Pacheco, Juan S; Stein, Benjamin W; Stieber, S Chantal E; Wilson, Justin J

    2016-01-01

    Actinium-225 is a promising isotope for targeted-α therapy. Unfortunately, progress in developing chelators for medicinal applications has been hindered by a limited understanding of actinium chemistry. This knowledge gap is primarily associated with handling actinium, as it is highly radioactive and in short supply. Hence, Ac(III) reactivity is often inferred from the lanthanides and minor actinides (that is, Am, Cm), with limited success. Here we overcome these challenges and characterize actinium in HCl solutions using X-ray absorption spectroscopy and molecular dynamics density functional theory. The Ac-Cl and Ac-OH2O distances are measured to be 2.95(3) and 2.59(3) Å, respectively. The X-ray absorption spectroscopy comparisons between Ac(III) and Am(III) in HCl solutions indicate Ac(III) coordinates more inner-sphere Cl(1-) ligands (3.2±1.1) than Am(III) (0.8±0.3). These results imply diverse reactivity for the +3 actinides and highlight the unexpected and unique Ac(III) chemical behaviour. PMID:27531582

  8. Production of high-purity radium-223 from legacy actinium-beryllium neutron sources.

    Soderquist, Chuck Z; McNamara, Bruce K; Fisher, Darrell R

    2012-07-01

    Radium-223 is a short-lived alpha-particle-emitting radionuclide with potential applications in cancer treatment. Research to develop new radiopharmaceuticals employing (223)Ra has been hindered by poor availability due to the small quantities of parent actinium-227 available world-wide. The purpose of this study was to develop innovative and cost-effective methods to obtain high-purity (223)Ra from (227)Ac. We obtained (227)Ac from two surplus actinium-beryllium neutron generators. We retrieved the actinium/beryllium buttons from the sources and dissolved them in a sulfuric-nitric acid solution. A crude actinium solid was recovered from the solution by coprecipitation with thorium fluoride, leaving beryllium in solution. The crude actinium was purified to provide about 40 milligrams of actinium nitrate using anion exchange in methanol-water-nitric acid solution. The purified actinium was then used to generate high-purity (223)Ra. We extracted (223)Ra using anion exchange in a methanol-water-nitric acid solution. After the radium was separated, actinium and thorium were then eluted from the column and dried for interim storage. This single-pass separation produces high purity, carrier-free (223)Ra product, and does not disturb the (227)Ac/(227)Th equilibrium. A high purity, carrier-free (227)Th was also obtained from the actinium using a similar anion exchange in nitric acid. These methods enable efficient production of (223)Ra for research and new alpha-emitter radiopharmaceutical development. PMID:22697483

  9. Radium, thorium, and actinium extraction from seawater using an improved manganese-oxide-coated fiber

    Laboratory experiments were conducted to determine the efficiency with which improved manganese-oxide-coated acrylic fibers extract radium, thorium, and actinium from seawater. Tests were made using surface seawater spiked with 227Ac, 227Th and 223Ra. For sample volumes of approximately 30 liters and flow rates up to 0.5 liters per minute, radium and actinium are removed quantitatively. Approximately 80-95% of the thorium is removed under these same conditions. (Auth.)

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

    This report identifies an appropriate thermal stabilization temperature for 233U 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 233U (DOE 2000), written as part of the response to Recommendation 97-1 of the Defense Nuclear Facilities Safety Board (DNFSB), addressing safe storage of 233U

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

    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.

  12. Disposition Options for Uranium-233

    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

  13. Excited levels of Pa-233

    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(T1) detectors. (Author) 54 refs

  14. 8 CFR 233.1 - Contracts.

    2010-01-01

    ... 8 Aliens and Nationality 1 2010-01-01 2010-01-01 false Contracts. 233.1 Section 233.1 Aliens and Nationality DEPARTMENT OF HOMELAND SECURITY IMMIGRATION REGULATIONS CONTRACTS WITH TRANSPORTATION LINES § 233.1 Contracts. The contracts with transportation lines referred to in section 233(c) of the Act may...

  15. Neutron-Induced Fission of Actinium-227, Protactinium-231 and Neptunium-237: Mass Distribution

    Results of radiochemical studies on the mass distribution in the neutron-induced fission of actinium-227, protactinium-231 and neptunium-237 have been presented. This work has been carried out as part of a programme to determine the mass distribution in the fission of heavy elements as a function of Z and A. All irradiations have been carried out in the core of the swimming-pool type reactor APSARA with cadmium shielding wherever necessary. Relative yields of several fission product nuclides have been obtained by a method involving a comparison of the fission product activities from the respective targets with those formed from uranium-235 simultaneously irradiated. Thermal-neutron fission yields of uranium-235 have been assumed. These results indicate a predominantly asymmetric mass distribution in all the three cases, and also a distinct though small symmetric peak in the case of actinium-227. (author)

  16. A new method for the determination of low-level actinium-227 in geological samples

    We developed a new method for the determination of 227Ac in geological samples. The method uses extraction chromatographic techniques and alpha-spectrometry and is applicable for a range of natural matrices. Here we report on the procedure and results of the analysis of water (fresh and seawater) and rock samples. Water samples were acidified and rock samples underwent total dissolution via acid leaching. A DGA (N,N,N',N'-tetra-n-octyldiglycolamide) extraction chromatographic column was used for the separation of actinium. The actinium fraction was prepared for alpha spectrometric measurement via cerium fluoride micro-precipitation. Recoveries of actinium in water samples were 80 ± 8 % (number of analyses n = 14) and in rock samples 70 ± 12 % (n = 30). The minimum detectable activities (MDA) were 0.017-0.5 Bq kg-1 for both matrices. Rock sample 227Ac activities ranged from 0.17 to 8.3 Bq kg-1 and water sample activities ranged from below MDA values to 14 Bq kg-1of 227Ac. From the analysis of several standard rock and water samples with the method we found very good agreement between our results and certified values. (author)

  17. 49 CFR 233.9 - Reports.

    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...

  18. 14 CFR 23.3 - Airplane categories.

    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...

  19. 22 CFR 23.3 - Refunds.

    2010-04-01

    ... 22 Foreign Relations 1 2010-04-01 2010-04-01 false Refunds. 23.3 Section 23.3 Foreign Relations DEPARTMENT OF STATE FEES AND FUNDS FINANCE AND ACCOUNTING § 23.3 Refunds. (a) Rectifications and readjustments. See § 22.6 of this chapter for outline of circumstances under which fees which have...

  20. 7 CFR 58.233 - Skim milk.

    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...

  1. 45 CFR 233.70 - Blindness.

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Blindness. 233.70 Section 233.70 Public Welfare... FINANCIAL ASSISTANCE PROGRAMS § 233.70 Blindness. (a) State plan requirements. A State plan under title X or XVI of the Social Security Act must: (1) Contain a definition of blindness in terms of...

  2. 40 CFR 233.71 - New Jersey.

    2010-07-01

    ... Director of the Federal Register in accordance with 552(a) and 1 CFR part 51. Material is incorporated as... 40 Protection of Environment 24 2010-07-01 2010-07-01 false New Jersey. 233.71 Section 233.71... REGULATIONS Approved State Programs § 233.71 New Jersey. The applicable regulatory program for discharges...

  3. 49 CFR 233.13 - Criminal penalty.

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Criminal penalty. 233.13 Section 233.13 Transportation Other Regulations Relating to Transportation (Continued) FEDERAL RAILROAD ADMINISTRATION, DEPARTMENT OF TRANSPORTATION SIGNAL SYSTEMS REPORTING REQUIREMENTS § 233.13 Criminal penalty....

  4. 45 CFR 233.80 - Disability.

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Disability. 233.80 Section 233.80 Public Welfare... FINANCIAL ASSISTANCE PROGRAMS § 233.80 Disability. (a) State plan requirements. A State plan under title XIV...; and (ii) “Totally” is related to the degree of disability. The following definition is...

  5. 12 CFR 23.3 - Lease requirements.

    2010-01-01

    ... 12 Banks and Banking 1 2010-01-01 2010-01-01 false Lease requirements. 23.3 Section 23.3 Banks and Banking COMPTROLLER OF THE CURRENCY, DEPARTMENT OF THE TREASURY LEASING General Provisions § 23.3 Lease... connection with leasing that property, and may engage in activities incidental thereto, if the...

  6. 49 CFR 234.233 - Rail joints.

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Rail joints. 234.233 Section 234.233..., DEPARTMENT OF TRANSPORTATION GRADE CROSSING SIGNAL SYSTEM SAFETY AND STATE ACTION PLANS Maintenance, Inspection, and Testing Maintenance Standards § 234.233 Rail joints. Each non-insulated rail joint...

  7. Plutonium and U-233 mines

    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

  8. 12 CFR 233.4 - Exemptions.

    2010-01-01

    ... FUNDING OF UNLAWFUL INTERNET GAMBLING (REGULATION GG) § 233.4 Exemptions. (a) Automated clearing house... business are exempt from this regulation's requirements for establishing written policies and...

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

    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.

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

    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

  11. Dicty_cDB: SLB233 [Dicty_cDB

    Full Text Available SL (Link to library) SLB233 (Link to dictyBase) slb233 - - Contig-U16290-1 SLB233P ...(Link to Original site) SLB233F 697 SLB233Z 631 SLB233P 1328 - - Show SLB233 Library SL (Link to library) Clone ID SLB...233 (Link to dictyBase) Atlas ID slb233 NBRP ID - dictyBase ID - Link to Contig Contig-U16290-1 Or...iginal site URL http://dictycdb.biol.tsukuba.ac.jp/CSM/SL/SLB2-B/SLB233Q.Seq.d/ R...epresentative seq. ID SLB233P (Link to Original site) Representative DNA sequence >SLB233 (SLB233Q) /CSM/SL/SLB2-B/SLB

  12. Dicty_cDB: SLD233 [Dicty_cDB

    Full Text Available SL (Link to library) SLD233 (Link to dictyBase) sld233 - - Contig-U15622-1 SLD233P ...(Link to Original site) SLD233F 296 SLD233Z 383 SLD233P 679 - - Show SLD233 Library SL (Link to library) Clone ID SLD2...33 (Link to dictyBase) Atlas ID sld233 NBRP ID - dictyBase ID - Link to Contig Contig-U15622-1 Ori...ginal site URL http://dictycdb.biol.tsukuba.ac.jp/CSM/SL/SLD2-B/SLD233Q.Seq.d/ Re...presentative seq. ID SLD233P (Link to Original site) Representative DNA sequence >SLD233 (SLD233Q) /CSM/SL/SLD2-B/SLD2

  13. 48 CFR 233.215 - Contract clause.

    2010-10-01

    ... 48 Federal Acquisition Regulations System 3 2010-10-01 2010-10-01 false Contract clause. 233.215....215 Contract clause. Use Alternate I of the clause at FAR 52.233-1, Disputes, when— (1) The... (v) Tracked combat vehicles (vi) Related electronic systems; (2) The contracting officer...

  14. Preserving high-purity 233U

    The MARC X Conference hosted a workshop for the scientific community to communicate needs for high-purity 233U 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 233U are being preserved. However, many additional kilograms of 233U (>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)

  15. 49 CFR 233.3 - Application.

    2010-10-01

    ... TRANSPORTATION SIGNAL SYSTEMS REPORTING REQUIREMENTS § 233.3 Application. (a) Except as provided in paragraph (b... the general railroad system of transportation. (b) This part does not apply to rail rapid...

  16. The Pa-233 fission cross section

    The energy dependent neutron-induced fission cross section of 233Pa has for the first time been measured directly with mono-energetic neutrons. This isotope is produced in the thorium fuel cycle and serves as an intermediate step between the 232Th source material and the 233U fuel material. Four neutron energies between 1.0 and 3.0 MeV have been measured in a first campaign. Some preliminary results are presented and compared to literature. (author)

  17. In-source laser spectroscopy developments at TRILIS—towards spectroscopy on actinium and scandium

    Resonance Ionization Laser Ion Sources (RILIS) have become a versatile tool for production and study of exotic nuclides at Isotope Separator On-Line (ISOL) facilities such as ISAC at TRIUMF. The recent development and addition of a grating tuned spectroscopy laser to the TRIUMF RILIS solid state laser system allows for wide range spectral scans to investigate atomic structures on short lived isotopes, e.g., those from the element actinium, produced in uranium targets at ISAC. In addition, development of new and improved laser ionization schemes for rare isotope production at ISAC is ongoing. Here spectroscopic studies on bound states, Rydberg states and autoionizing (AI) resonances on scandium using the existing off-line capabilities are reported. These results allowed to identify a suitable ionization scheme for scandium via excitation into an autoionizing state at 58,104 cm − 1 which has subsequently been used for ionization of on-line produced exotic scandium isotopes.

  18. Dicty_cDB: SFL233 [Dicty_cDB

    Full Text Available MKIKIXXPPXRKXXVWIGGXILXSLSTFXQMWXSKX Translated Amino Acid sequence (All Frames) Frame A: iasdfg**rkrfspftki...SF (Link to library) SFL233 (Link to dictyBase) - - - Contig-U16382-1 SFL233P (Link... to Original site) SFL233F 480 SFL233Z 142 SFL233P 622 - - Show SFL233 Library SF (Link to library) Clone ID... SFL233 (Link to dictyBase) Atlas ID - NBRP ID - dictyBase ID - Link to Contig Contig-U16382-1 Original site...xptnvdxxx Frame B: slptsvnnekdfrlllkyfrepks*rqlslff*iglpipktnkln*iknkngw*rcssfs y**rfwyv*srfcw*rcstccfpincws

  19. Workshop on Preserving High Purity Uranium-233

    Krichinsky, Alan M [ORNL; Giaquinto, Joseph [ORNL; Canaan, R Douglas {Doug} [ORNL

    2016-01-01

    A workshop was held on at the MARC X conference to provide a forum for the scientific community to communicate needs for high-purity 233U and its by-products in order to preserve critical items otherwise slated for downblending and disposal. Currently, only a small portion of the U.S. holdings of separated 233U is being preserved. However, many additional kilograms of 233U (>97% pure) still are destined to be downblended which will permanently destroy their potential value for many other applications. It is not likely 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.

  20. Benchmark testing of 233U evaluations

    In this paper we investigate the adequacy of available 233U cross-section data (ENDF/B-VI and JENDL-3) for calculation of critical experiments. An ad hoc revised 233U evaluation is also tested and appears to give results which are improved relative to those obtained with either ENDF/B-VI or JENDL-3 cross sections. Calculations of keff were performed for ten fast benchmarks and six thermal benchmarks using the three cross-section sets. Central reaction-rate-ratio calculations were also performed

  1. 12 CFR 233.7 - Regulatory enforcement.

    2010-01-01

    ... PROHIBITION ON FUNDING OF UNLAWFUL INTERNET GAMBLING (REGULATION GG) § 233.7 Regulatory enforcement. The... regulators, with respect to the designated payment systems and participants therein that are subject to the... Commission, with respect to designated payment systems and participants therein not otherwise subject to...

  2. Preserving Ultra-Pure Uranium-233

    Krichinsky, Alan M [ORNL; Goldberg, Dr. Steven A. [DOE SC - Chicago Office; Hutcheon, Dr. Ian D. [Lawrence Livermore National Laboratory (LLNL)

    2011-10-01

    Uranium-233 ({sup 233}U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium ({sup 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 {sup 233}U is stored in vaults at Oak Ridge National Laboratory (ORNL). These materials represent a broad spectrum of {sup 233}U from the standpoint of isotopic purity - the purest being crucial for precise analyses in safeguarding uranium. All {sup 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 {sup 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 {sup 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 {sup 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 {sup 233}U were identified as candidates for establishing a lasting supply of

  3. 34 CFR 668.233 - Student eligibility.

    2010-07-01

    ... Intellectual Disabilities § 668.233 Student eligibility. A student with an intellectual disability is eligible... intellectual disability, as described in paragraph (1) of the definition of a student with an intellectual... intellectual disability, such as— (1) A documented comprehensive and individualized...

  4. Star formation in the S233 region

    Ladeyschikov, D A; Parfenov, S Yu; Alexeeva, S A; Bieging, J H

    2015-01-01

    The main objective of this paper is to study the possibility of triggered star formation on the border of the HII region S233, which is formed by a B-star. Using high-resolution spectra we determine the spectral class of the ionizing star as B0.5 V and the radial velocity of the star to be -17.5(1.4) km/s. This value is consistent with the velocity of gas in a wide field across the S233 region, suggesting that the ionizing star was formed from a parent cloud belonging to the S233 region. By studying spatial-kinematic structure of the molecular cloud in the S233 region, we detected an isolated clump of gas producing CO emission red-shifted relative to the parent cloud. In the UKIDSS and WISE images, the clump of gas coincides with the infrared source containing a compact object and bright-rimmed structure. The bright-rimmed structure is perpendicular to the direction of the ionizing star. The compact source coincides in position with IRAS source 05351+3549. All these features indicate a possibility of triggeri...

  5. 233-S plutonium concentration facility hazards assessment

    This document establishes the technical basis in support of Emergency Planning activities for the 233-S Plutonium Concentration Facility on the Hanford Site. The document represents an acceptable interpretation of the implementing guidance document for DOE ORDER 5500.3A. Through this document, the technical basis for the development of facility specific Emergency Action Levels and the Emergency Planning Zone is demonstrated

  6. 28 CFR 23.3 - Applicability.

    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...

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

    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.; Huyse, M.; Kron, T.; Kudryavtsev, Yu.; Laatiaoui, M.; Lecesne, N.; Luton, F.; Moore, I. D.; Martinez, Y.; Mogilevskiy, E.; Naubereit, P.; Piot, J.; Rothe, S.; Savajols, H.; Sels, S.; Sonnenschein, V.; Traykov, E.; Van Beveren, C.; Van den Bergh, P.; Van Duppen, P.; Wendt, K.; Zadvornaya, A.

    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.

  8. Dicty_cDB: CHR233 [Dicty_cDB

    Full Text Available CH (Link to library) CHR233 (Link to dictyBase) - - - Contig-U16471-1 - (Link to Original site) - - CHR...233Z 789 - - - - Show CHR233 Library CH (Link to library) Clone ID CHR233 (Link to dicty...iol.tsukuba.ac.jp/CSM/CH/CHR2-B/CHR233Q.Seq.d/ Representative seq. ID - (Link to ...Original site) Representative DNA sequence >CHR233 (CHR233Q) /CSM/CH/CHR2-B/CHR233Q.Seq.d/ XXXXXXXXXXTATATGA...DNA Score E Sequences producing significant alignments: (bits) Value SHB457 (SHB457Q) /CSM/SH/SHB4-C/SHB457Q.Seq.d/ 747 0.0 CHR

  9. Track 8: health and radiological applications. Isotopes and radiation: general. 3. Extraction of 229Th from 233U for Medical Research Applications

    the handling of 233U. After separation, the thorium is transferred to the ORNL Radiochemical Development Laboratory (RDL) for decay and periodic extraction of actinium. The safety basis for the RMAL has recently been upgraded to allow the processing of batches containing up to 500 g of 233U. The 233U from Mound Laboratories was delivered in 20 packages, each containing 232U contamination ranges from 2 to 16 ppm. Individual packages of uranium oxide are transferred from the RDF to the RMAL, the package is opened, and the quantity of uranium is verified. The uranium oxide is dissolved in nitric acid, and the solution is adjusted to 8 M. The solution is passed over a Bio-Rad 1x4 anion exchange column at a rate of ∼5 ml/ min. Thorium and plutonium load on the column; uranium, radium, and actinium pass through. The column is washed with 8 M nitric acid to remove residual uranium. Thorium is then eluted from the resin with 0.1 M nitric acid. This solution, typically containing ∼4 mCi of 229Th, is transferred to the RDL for further purification and is then added to the existing thorium inventory. The uranium and acid wash solutions are combined and evaporated to 3O8 by heating in air to 800 deg. C. The oxide is sampled, weighed, packaged, and returned to storage in the RDF. In fiscal year 2000, the process was set up and two small batches of uranium, with 2 ppm 232U, were processed. Because of the low dose associated with this material and to allow refinements to the process, these initial batches were processed in a glove box. The process performed well, with more than 90% recovery of thorium. Just over 5 mCi of 229Th was added to the inventory. Funding already provided for fiscal year 2001 will allow the extraction of another 30 mCi of 229Th; if additional funding can be identified, a total of nearly 70 mCi can be separated by the end of the fiscal year. With this additional thorium, and with other improvements to the actinium extraction facilities at the RDL, the

  10. 45 CFR 233.52 - Overpayment to aliens.

    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...

  11. 17 CFR 256.233 - Notes payable to associate companies.

    2010-04-01

    ... companies. 256.233 Section 256.233 Commodity and Securities Exchanges SECURITIES AND EXCHANGE COMMISSION (CONTINUED) UNIFORM SYSTEM OF ACCOUNTS FOR MUTUAL SERVICE COMPANIES AND SUBSIDIARY SERVICE COMPANIES, PUBLIC UTILITY HOLDING COMPANY ACT OF 1935 7. Current and Accrued Liabilities § 256.233 Notes payable...

  12. 40 CFR 233.4 - Conflict of interest.

    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...

  13. 45 CFR 233.36 - Monthly reporting (AFDC).

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Monthly reporting (AFDC). 233.36 Section 233.36... ELIGIBILITY IN FINANCIAL ASSISTANCE PROGRAMS § 233.36 Monthly reporting (AFDC). (a) Except as provided in paragraph (b) of this section, a State plan for AFDC shall require the caretaker relative, or another...

  14. 45 CFR 233.90 - Factors specific to AFDC.

    2010-10-01

    ... meet the needs of the unborn child. (Refer to Medicaid regulations at 42 CFR 435.115 for Medicaid... 45 Public Welfare 2 2010-10-01 2010-10-01 false Factors specific to AFDC. 233.90 Section 233.90... ELIGIBILITY IN FINANCIAL ASSISTANCE PROGRAMS § 233.90 Factors specific to AFDC. (a) State plan requirements....

  15. 45 CFR 233.31 - Budgeting methods for AFDC.

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Budgeting methods for AFDC. 233.31 Section 233.31... ELIGIBILITY IN FINANCIAL ASSISTANCE PROGRAMS § 233.31 Budgeting methods for AFDC. (a) Requirements for State plans. A State plan for AFDC shall specify that all factors of eligibility shall be...

  16. Performance analysis of 233U for fixed bed nuclear reactors

    Criticality and burn up behavior of the Fixed Bed Nuclear Reactor (FBNR) are investigated for the mixed fuel 233UO2/ThO2 as an alternative to low enriched 235UO2 fuel. CERMET fuel with a zirconium matrix and cladding has been used throughout the study. The main results of the study can be summarized as follows: Reactor criticality is already achieved by ∝2% 233UO2 with the mixed 233UO2/ThO2 fuel. At higher 233U fractions, reactor criticality rises rapidly and exceeds keff > 1.5 already by 9% 233UO2. With 100% 233UO2, start up criticality can reach keff = 2.0975. Time dependent reactor criticality keff and fuel burn up have been investigated for two different mixed fuel 233UO2/ThO2 compositions, namely: 4% 233UO2 + 96% ThO2 for a reactor power of 40 MWel (120 MWth) and 9% 233UO2 + 91% ThO2 for a reactor power of 70 MWel (210 MWth). Sufficient reactor criticality (keff > 1.06) for continuous operation without fuel change can be sustained during ∝ 5 and 12 years with 4% and 9% 233UO2 fractions in the mixed fuel, leading to burn ups of ∝ 36000 and > 105000 MWD/t, respectively. Thorium based fuel produces no prolific uranium. Plutonium production remains negligible. (orig.)

  17. Update of 233U, 229-232Th and 230-233Pa Fission Data

    The influence of the 235U(n,f) and 239Pu(n,f) prompt fission neutron spectra (PFNS) on modeling of integral benchmarks was estimated to be significant. For 233U(n,f) PFNS similar sensitivities could be envisaged. For the variety of Th/U fuels and systems, ranging from metal fast to deep thermal solutions, large positive/negative swings in calculated Keff can be expected. Th/U fuelled core criticality calculations would be sensitive to the modelled soft tail of fission neutrons or deficiency of hard tail fission neutrons, as revealed for U and Pu fuels. The deficiency of the 233U(nth,f) PFNS, adopted for the ENDF/B-VII.0. could be traced back to the 'propagation' of calculated 235U(nth,f) PFNS shape at En= 0.5 MeV. At higher energies, average energies of PFNS coincide only at ∼5 MeV, at other En the average energies and spectra shapes are drastically different. However, it might be argued that the response of the criticality benchmark calculations for the 233U thermal and fast systems would be similar to that observed for the PFNS of 239Pu. Our approach allowed to solve the longstanding problem of inconsistency of 235U integral data testing and differential prompt fission neutron spectra data, a similar approach may hold for 233U. Using modified PFNS, we may avoid arbitrary tweaking of neutron cross sections or neutron multiplicities for Th/U fuel-cycle related nuclides (233U, 229,230,231,232Th and 230,231,232,233Pa) to compensate the ill-defined shape of 233U PFNS. For metal fast benchmarks much would depend on the inelastic scattering cross section of 233U, which should be considered rather uncertain at the moment. The unrealistic evaluations of poorly investigated cross sections related to the Th/U fuel cycle could be excluded by consistent analysis of the available fission data base. The evaluation of 229,230,231,232Th(n,f) and 230,231,232,233Pa(n,f) cross sections could be supplemented by description of surrogate and ratio surrogate fission data, coming from

  18. Neutron data evaluation of 233Pa

    Consistent evaluation of 233 Pa 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 band levels is calculated within rigid rotator model. Average resonance parameters are provided, which reproduce evaluated cross sections in the range of 16.5-70.9 keV. This work is performed under the Project Agreement B-404 with the International Science and Technology Center (Moscow). The Financing Party for the Project is Japan. (author)

  19. Uranium-233 analysis of biological samples

    Two liquid scintillation techniques were compared for 233U 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

  20. Evaluation of resonance parameters of U-233

    Compilation of nuclear data is in progress in Japan, and the second edition of Japanese Evaluated Nuclear Data Library (JENDL-2) will be published. The evaluation of the resonance parameters of U-233, which will be included in JENDL-2, has been made. The measured values of the resonance parameters after the publication of BNL-325 (second edition) were collected, and searched by using CINDA-78. The data by Blons, Kolar, Ryabov, and Bergen were used for the present evaluation. Complete set of the data was made for each measurement. Fission and capture areas integrated over energy intervals were obtained. The total, fission and capture cross-sections of U-233 were calculated from the various sets of complete resonance parameters, and shown in figures. The calculated values of total, fission and capture cross-sections based on the parameters by Blons were compared with the measured values. Correction of the resonance parameters with poor reproducibility was able to be made with the NDES system by Nakagawa. The final parameters which will be included in JENDL are shown in tables. (Kato, T.)

  1. 27 CFR 24.233 - Addition of spirits to wine.

    2010-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Addition of spirits to 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...

  2. Nuclear data evaluation for Pa-233

    In this report the evaluation of main neutron nuclear data for 233Pa, namely neutron cross sections (total, elastic, inelastic, radiative capture, fission, (n,2n), (n,3n)), as well as the elastic and inelastic angular distributions, and energy distributions of secondary neutrons from inelastic scattering, (n,2n), (n,3n) and fission reactions, is described. In the same time, radioactive decay data and average number of neutrons per fission are given. For the resolved and unresolved resonance energy range, the Breit-Wigner single level parameters have been estimated. The data cover the energy range between 10-5eV and 20MeV. The final set of evaluated data is given in ENDF/B format and have been checked against physical consistency and format correctness. Many of the data have been calculated using theoretical models. (author)

  3. 45 CFR 233.51 - Eligibility of sponsored aliens.

    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...

  4. 17 CFR 201.233 - Depositions upon oral examination.

    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...

  5. 40 CFR 233.61 - Determination of Tribal eligibility.

    2010-07-01

    .... 233.61 Section 233.61 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) OCEAN..., such as, but not limited to, the exercise of police powers affecting (or relating to) the health... environmental or public health programs administered by the Tribal governing body, and a copy of related...

  6. 49 CFR 238.233 - Interior fittings and surfaces.

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Interior fittings and surfaces. 238.233 Section... I Passenger Equipment § 238.233 Interior fittings and surfaces. (a) Each seat in a passenger car... determined by the railroad: (1) Longitudinal: 8g; (2) Vertical: 4g; and (3) Lateral: 4g. (c) Other...

  7. 48 CFR 952.233-2 - Service of protest.

    2010-10-01

    ... SOLICITATION PROVISIONS AND CONTRACT CLAUSES Text of Provisions and Clauses 952.233-2 Service of protest. As prescribed in 933.106(a), add the following to the end of the provision at 48 CFR 52.233-2: (c) Another copy... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Service of protest....

  8. 45 CFR 233.32 - Payment and budget months (AFDC).

    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 shall specify in its plan for AFDC the time period covered by the payment (payment month) and the...

  9. 232 Th - 233 Pa separation by extraction chromatography

    Thorium and protactinium behavior in extraction chromatography systems is presented, aiming its separation by selective retention of the micro constituent on the column. TBP/alumina, TBP/voltalef UF 300, TOA/alumina and TOA/voltalef UF 300 systems were verified. Column preparation as well the 233 Pa removal conditions were settled. The best 232 Th separation from its irradiation product, 233 Pa, has been achieved by using TBP/voltalef UF 300 system. 233 Pa was selectively retained on column from 10 M HCl solutions and eluted with 3 M HCl. (author)

  10. 16 CFR 233.5 - Miscellaneous price comparisons.

    2010-01-01

    ... DECEPTIVE PRICING § 233.5 Miscellaneous price comparisons. The practices covered in the provisions set forth... principles. For example, retailers should not advertise a retail price as a “wholesale” price. They...

  11. Feasibility to produce uranium-233 from thorium in PHWR

    Uranium-233 is a fissile isotope of uranium that is bred from thorium-232 as part of the thorium fuel cycle. It is used as fuel in nuclear reactor. 233U is produced by irradiating thorium in fast reactor or thermal reactor. In this paper feasibility to produce 233U by irradiation of fuel bundles of thorium mixed with plutonium or irradiation of thorium bundles along with slightly enriched Uranium bundles in 220 MWe and 540 MWe PHWRs has been studied. Study shows that production of 233U is possible by irradiating few fuel bundles of PuTh or Thorium bundles along with SEU bundles without affecting the power operation and the safety related parameters. (author)

  12. 12 CFR 233.5 - Policies and procedures required.

    2010-01-01

    ... SYSTEM PROHIBITION ON FUNDING OF UNLAWFUL INTERNET GAMBLING (REGULATION GG) § 233.5 Policies and... otherwise refuses to honor a transaction, shall not be liable to any party for such action if— (1)...

  13. 12 CFR 233.3 - Designated payment systems.

    2010-01-01

    ... PROHIBITION ON FUNDING OF UNLAWFUL INTERNET GAMBLING (REGULATION GG) § 233.3 Designated payment systems. The... remotely from a location other than a physical office of the money transmitting business; and (e)...

  14. Dicty_cDB: VSJ233 [Dicty_cDB

    Full Text Available itfmvdmfanksq vadavakmydvkvkrvntlitprgekkafvtlspefeaadvankigli...fsli **kk Frame B: agkkvksntpkqdlsvskskltsikapaaaikakaaasavkkgvsnkstrkvrtsvifrr pvtlnnpkkpayprrsvnkitkmdqfrilkaplttesatqkiegsnt...e E Sequences producing significant alignments: (bits) Value VSJ233 (VSJ233Q) /CS...-B/SLC129Q.Seq.d/ 1021 0.0 own update 2001.11.29 Homology vs DNA Score E Sequences producing significant alignments: (bits....6 1 dna update 2003. 7.18 Homology vs Protein Score E Sequences producing significant alignments: (bits) Va

  15. Repository criticality control for 233U using depleted uranium

    The US is evaluating methods for the disposition of excess weapons-usable 233U, including disposal in a repository. Isotopic dilution studies were undertaken to determine how much depleted uranium (DU) would need to be added to the 233U to minimize the potential for nuclear criticality in a repository. Numerical evaluations were conducted to determine the nuclear equivalence of different 235U enrichments to 233U isotopically diluted with DU containing 0.2 wt% 235U. A homogeneous system of silicon dioxide, water, 233U, and DU, in which the ratio of each component was varied, was used to determine the conditions of maximum nuclear reactivity. In terms of preventing nuclear criticality in a repository, there are three important limits from these calculations. 1. Criticality safe in any nonnuclear system: The required isotopic dilution to ensure criticality under all conditions, except in the presence of man-made nuclear materials (beryllium, etc.), is ≅1.0% 235U in 238U. The equivalent 233U enrichment level is 0.53 wt% 233U in DU. 2. Critically safe in natural systems: The lowest 235U enrichment found in a natural reactor at shutdown was approximately1.3%. French studies, based on the characteristics of natural uranium ore bodies, indicate that a minimum enrichment of approximately1.28% 235U is required for criticality. These data suggest that nuclear criticality from migrating uranium is not realistic unless the 235U enrichments exceed approximately1.3%, which is a result that is equivalent to 0.72% 233U in DU. 3. Criticality safety equivalent to light water reactor (LWR) spent nuclear fuel (SNF): The 233U can be diluted with DU so that the uranium criticality characteristics match SNF uranium. Whatever repository criticality controls are used for SNF can then be used for 233U. The average LWR SNF assay (after decay of plutonium isotopes to uranium isotopes) is 1.5% 235U equivalent in 238U. This is equivalent to diluting 233U to 0.81% in DU

  16. Experimental 233U nondestructive assay with a random driver

    Nondestructive assay (NDA) of 233U in quantities up to 15 grams containing 7 ppM 232U 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 233U had the same 232U content at the same age. Where the 232U content varies, determination of 233U fissile content would require active NDA. Active coincidence counting utilizing a 238Pu, 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 232U decay chain

  17. Testing of 233U evaluations with criticality benchmarks

    To validate and improve the quality of the complete set of evaluated nuclear reaction data for 233U, criticality benchmarks with fast, epithermal and thermal spectra from ICSBEP handbook were selected to test 233U evaluations from CENDL-3.1, ENDF/B-Ⅶ.0, JENDL-3.3 and JENDL-4.0. The effective multiplication factors keff of selected benchmarks were calculated with the Monte Carlo code MCNP5 and compared with the benchmark values. The results were analyzed with trend against energy spectrum index and sensitivity analysis. In present validation, the underestimation of keff for benchmarks with thermal, epithermal or some of fast spectra is the main problem existed in the tested evaluations. From the view of thermal reactors design, the 233U evaluation from ENDF/B-Ⅶ.0 shows better performance than other file tested, but still overestimates the contribution of capture reaction in resonance region. (authors)

  18. Decontamination of the 233-S building loadout hood

    This paper concerns the decontamination experience gained during the decontamination and decommissioning (DandD) operations on the Loadout Hood within the 233-S Building. The retired 233-S Building (Plutonium Concentration Facility) is being decommissioned as a demonstration project to develop baseline cost, technology, and operational data for DandD of alpha contaminated facilities. The Loadout Hood within the facility is a plutonium nitrate loadout system highly contaminated with transuranics. The paper consists of a brief description of the facility and Loadout Hood, a summary of the engineering and field work performed, and an evaluation of the technology and methods used

  19. Groundwater seepage from the Ranger uranium mine tailings dam: radioisotopes of radium, thorium and actinium. Supervising Scientist report 106

    Monitoring of bores near the Ranger uranium mine tailings dam has revealed deterioration in water quality in several bores since 1983. In a group of bores to the north of the dam, increases have been observed of up to 500 times for sulphate concentrations and of up to 5 times for 226Ra concentrations. Results are presented here of measurements of members of the uranium, thorium and actinium decay series in borewater samples collected between 1985 and 1993. In particular, measurements of all four naturally-occurring radium isotopes have been used in an investigation of the mechanism of radium concentration changes. For the most seepage-affected bores the major findings of the study include: 228Ra/226Ra 223Ra /226Ra and 224Ra/228Ra ratios all increased over the course of the study; barium concentrations show high seasonal variability, being lower in November than May, but strontium concentrations show a steady increase with time. Calculations show that the groundwater is probably saturated with respect to barite but not with respect to celestite or anglesite; sulphide concentrations are low in comparison with sulphate, and are higher in November than in May; and 227Ac concentrations have increased with time, but do not account for the high 223Ra/226Ra ratios. It is concluded on the basis of these observations that increases in Ra isotope concentrations observed in a number of seepage-affected bores arise from increases in salinity leading to desorption of radium from adsorption sites in the vicinity of the bore rather by direct transport of radium from the tailings. Increased salinity is also causing the observed increases in 227Ac and strontium concentrations, while formation of a barite solid phase in the groundwater is causing the removal of some radium from solution. This is the cause of the increasing radium isotope ratios noted above

  20. Dicty_cDB: VHJ233 [Dicty_cDB

    Full Text Available VH (Link to library) VHJ233 (Link to dictyBase) - - - Contig-U16440-1 - (Link to Original site) ... camonas acidamino... 34 3.2 (Q8TS91) RecName: Full=Imidazole ... glycerol phosphate synthase sub... 33 4.2 AF191043 ...

  1. 46 CFR 108.233 - Location and size.

    2010-10-01

    ... COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) A-MOBILE OFFSHORE DRILLING UNITS DESIGN AND EQUIPMENT Construction and Arrangement Helicopter Facilities § 108.233 Location and size. (a) Each helicopter deck must be— (1) At least the size of the rotor diameter of the largest single main...

  2. 48 CFR 1852.233-70 - Protests to NASA.

    2010-10-01

    ... 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... 48 Federal Acquisition Regulations System 6 2010-10-01 2010-10-01 true Protests to NASA....

  3. 45 CFR 233.20 - Need and amount of assistance.

    2010-10-01

    ... eligible child by applying the stepparent deeming formula at 45 CFR 233.20(a)(3)(xiv). (vii) If the State... Orange Product liability litigation, M.D.L. No. 381 (E.D.N.Y.). (t) Student financial assistance made..., items such as depreciation, personal business and entertainment expenses, personal...

  4. 7 CFR 1209.233 - Regional caucus chairpersons.

    2010-01-01

    ... chairpersons will coordinate the entire nomination process. In conducting the nominations process, each... PROMOTION, RESEARCH, AND CONSUMER INFORMATION ORDER Rules and Regulations Nomination Procedures § 1209.233..., and voting for importer nominees is limited to importers; and (2) Producer candidates for...

  5. 48 CFR 1452.233-2 - Service of Protest.

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Service of Protest. 1452... Protest. As prescribed in 1433.106, the provision at FAR 52.233-2, Service of Protest, shall be modified...: “Service of Protest Department of the Interior (JUL 1996) (Deviation)”; and adding the following...

  6. 48 CFR 52.233-2 - Service of Protest.

    2010-10-01

    ... 48 Federal Acquisition Regulations System 2 2010-10-01 2010-10-01 false Service of Protest. 52.233... Service of Protest. As prescribed in 33.106(a), insert the following provision: Service of Protest (SEP 2006) (a) Protests, as defined in section 33.101 of the Federal Acquisition Regulation, that are...

  7. 48 CFR 52.233-3 - Protest After Award.

    2010-10-01

    ... 48 Federal Acquisition Regulations System 2 2010-10-01 2010-10-01 false Protest After Award. 52....233-3 Protest After Award. As prescribed in 33.106(b), insert the following clause: Protest After Award (AUG 1996) (a) Upon receipt of a notice of protest (as defined in FAR 33.101) or a...

  8. 48 CFR 852.233-71 - Alternate protest procedure.

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Alternate protest... § 852.233-71 Alternate protest procedure. As prescribed in 833.106, insert the following provision: Alternate Protest Procedure (JAN 1998) As an alternative to filing a protest with the contracting...

  9. 48 CFR 952.233-5 - Agency protest review.

    2010-10-01

    ... level. The Department of Energy's agency protest procedures, set forth in 48 CFR 933.103, elaborate on... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Agency protest review. 952... SOLICITATION PROVISIONS AND CONTRACT CLAUSES Text of Provisions and Clauses 952.233-5 Agency protest review....

  10. The 233Pa fission cross-section measurement and evaluation

    233Pa is a conspicuous example of strongly discrepant data in the accepted nuclear data evaluations. The precise knowledge of the neutron-induced reaction cross-section of this highly β-active nuclide (T1/2 = 27.0 d) is essential for the successful implementation of the thorium-based fuel cycle in advanced nuclear applications. The reactions involving 233Pa are responsible for the balance of nuclei as well as the average number of prompt fission neutrons in a contemplated reactor scenario. In an IAEA report, it is stated that there is a need to know the 233Pa(n, f) cross-section with an accuracy of 20%. The different evaluated neutron data libraries show, however, a difference of a factor of two for this cross-section. It has previously been deemed not feasible to measure this reaction directly due to its short half-life, high radioactivity and the in-growth of the daughter product 233U. Hence, the entries in the neutron libraries are based on theoretical predictions, which explains the large discrepancies. As reported recently the neutron-induced fission cross-section of 233Pa has been measured for the first time directly with mono-energetic neutrons from 1.0 to 3.0 MeV at the Van-de-Graaff facility of the IRMM. In the meantime, during two further measurement campaigns, the energy range has been extended up to 8.5 MeV. The experimental results will be presented together with recent model calculations of the fission cross-section applying the statistical model code STATIS, which improve the cross-section evaluation up to the second chance fission threshold. (authors)

  11. 48 CFR 52.233-4 - Applicable Law for Breach of Contract Claim.

    2010-10-01

    ... 48 Federal Acquisition Regulations System 2 2010-10-01 2010-10-01 false Applicable Law for Breach of Contract Claim. 52.233-4 Section 52.233-4 Federal Acquisition Regulations System FEDERAL ACQUISITION REGULATION (CONTINUED) CLAUSES AND FORMS SOLICITATION PROVISIONS AND CONTRACT CLAUSES Text of Provisions and Clauses 52.233-4 Applicable Law...

  12. 47 CFR 51.233 - Significant degradation of services caused by deployment of advanced services.

    2010-10-01

    ... presumption that it is acceptable for deployment under § 51.230, the degraded service shall not prevail... deployment of advanced services. 51.233 Section 51.233 Telecommunication FEDERAL COMMUNICATIONS COMMISSION... § 51.233 Significant degradation of services caused by deployment of advanced services. (a) Where...

  13. 39 CFR 233.5 - Requesting financial records from a financial institution.

    2010-07-01

    ... 39 Postal Service 1 2010-07-01 2010-07-01 false Requesting financial records from a financial institution. 233.5 Section 233.5 Postal Service UNITED STATES POSTAL SERVICE ORGANIZATION AND ADMINISTRATION INSPECTION SERVICE AUTHORITY § 233.5 Requesting financial records from a financial institution....

  14. 48 CFR 5452.233-9001 - Disputes: Agreement To Use Alternative Dispute Resolution (ADR).

    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...

  15. 48 CFR 1352.233-71 - GAO and Court of Federal Claims protests.

    2010-10-01

    ....233-71 GAO and Court of Federal Claims protests. As prescribed in 48 CFR 1333.104-70(a), insert the following provision: GAO and Court of Federal Claims Protests (APR 2010) (a) A protest may be filed with... Claims protests. 1352.233-71 Section 1352.233-71 Federal Acquisition Regulations System DEPARTMENT...

  16. 45 CFR 233.37 - How monthly reports are treated and what notices are required (AFDC).

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false How monthly reports are treated and what notices are required (AFDC). 233.37 Section 233.37 Public Welfare Regulations Relating to Public Welfare... § 233.37 How monthly reports are treated and what notices are required (AFDC). (a) What happens if...

  17. 45 CFR 233.38 - Waiver of monthly reporting and retrospective budgeting requirements; AFDC.

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Waiver of monthly reporting and retrospective budgeting requirements; AFDC. 233.38 Section 233.38 Public Welfare Regulations Relating to Public Welfare... § 233.38 Waiver of monthly reporting and retrospective budgeting requirements; AFDC. (a) States...

  18. 48 CFR 2852.233-70 - Protests filed directly with the Department of Justice.

    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...

  19. Dicty_cDB: VSA233 [Dicty_cDB

    Full Text Available 4231 |BF294231.1 001PbG02 Pb cDNA #17, Tommaso Pace, Marta Ponzi, and Clara Frontali Plasmodium berghei cDNA...t: 0.00 m3a: 0.00 m3b: 0.00 m_ : 1.00 52.0 %: cytoplasmic 24.0 %: nuclear 8.0 %: cytoskeletal 4.0 %: Golgi 4.0 %: mitochondrial...VS (Link to library) VSA233 (Link to dictyBase) - - - Contig-U15040-1 VSA233Z (Link to Original...tlfiyrnikqylml*yn snraklsp Homology vs CSM-cDNA Score E Sequences producing significant alignments: (bits) Val...1. 8 Homology vs DNA Score E Sequences producing significant alignments: (bits) Value N AZ522056 |AZ522056.1

  20. Fast and Thermal Data Testing of U-233 Critical Assemblies

    Data testing has been performed for U-233 fast and thermal benchmarks. Results are presented for both ENDF/B-VI and a modified JENDL-3.2 evaluation. The revised JENDL-3.2 evaluation is summarized and comparisons with ENDF/B-VI and measured values are discussed. Calculated results using both cross section sets are presented for 10 fast benchmarks (reflected and unreflected U-233 metal) and 38 thermal benchmarks (uranyl-nitrate solutions in spherical and cylindrical geometry). Using the revised JENDL-3.2 evaluation, very good results are obtained for the calculated k-effs for almost all of the 48 benchmarks considered in this study. Possible future work is discussed briefly

  1. Updated and revised neutron reaction data for 233U

    YU Bao-Sheng; CHEN Guo-Chang; ZHANG Hua; CAO Wen-Tian; TANG Guo-You; TAO Xi

    2013-01-01

    A complete set of n+233U neutron reaction data from 10-5 eV-20 MeV is updated and revised based on the evaluated experimental data and the feedback information of various benchmark tests.The main revised quantities are nubars,cross sections as well as angular distributions,etc.The benchmark tests indicate that the present evaluated data achieve very promising results.

  2. Nuclear fuel cycle based on thorium and uranium-233

    The analysis of activities carried out in this country and abroad on a complex solution of principal problems of nuclear power advance. Demonstration of the potentiality of the above problems solution on the basis of conventional reactor plant development (light water cooled reactors and BN-type fast reactors) within the framework of nuclear fuel cycle using uranium-235, plutonium and uranium-233. 28 refs.; 1 fig.; 8 tabs

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

    This document provides a summary of technical information on the synthetic radioisotope 233U. It is one of a series of four reports that map out a national strategy for the future use and disposition of 233U. The technical information on 233U in this document falls into two main areas. First, material characteristics are presented along with the contrasts of 233U to the more well known strategic fissile materials, 235U 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 233U with nonfissile, depleted uranium (DU) are examined

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

    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

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

    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

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

    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

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

    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

  8. Isotopic dilution of 233U with depleted uranium for criticality safety in processing and disposal

    The disposal of excess 233U as waste is being considered. Because 233U is a fissile material, a key requirement for processing 233U to a final waste form and disposing of it is the avoidance of 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 233U with nonfissile 238U. 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 233U. The amount of isotopic dilution required to reduce criticality concerns to reasonable levels was determined in this study to be approximately 0.53 wt % 233U. The numerical calculations used to define this limit consisted of a homogeneous system of silicon dioxide (SiO2), water (H2O), 233U and depleted uranium (DU) in which the ratio of each component was varied to learn the conditions of maximum nuclear reactivity. About 188 parts of DU (0.2 wt % 235U) are required to dilute 1 part of 233U to this limit in a water-moderated system with no SiO2 present. Thus for the U.S. inventory of 233U, several hundred metric tons of DU would be required for isotopic dilution

  9. Study on the adsorption of 233Pa in glass

    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 233Pa 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)

  10. 45 CFR 233.106 - Denial of AFDC benefits to strikers.

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Denial of AFDC benefits to strikers. 233.106... COVERAGE AND CONDITIONS OF ELIGIBILITY IN FINANCIAL ASSISTANCE PROGRAMS § 233.106 Denial of AFDC benefits... refuse to seek or accept, employment. (2)(i) Provide for the denial of AFDC benefits to any family...

  11. 45 CFR 233.33 - Determining eligibility prospectively for all payment months (AFDC).

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Determining eligibility prospectively for all payment months (AFDC). 233.33 Section 233.33 Public Welfare Regulations Relating to Public Welfare OFFICE....33 Determining eligibility prospectively for all payment months (AFDC). (a) The State plan for...

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

    This assessment characterizes the 233U 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

  13. 9 CFR 2.33 - Attending veterinarian and adequate veterinary care.

    2010-01-01

    ... animal health, behavior, and well-being is conveyed to the attending veterinarian; (4) Guidance to... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Attending veterinarian and adequate veterinary care. 2.33 Section 2.33 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION...

  14. 24 CFR 5.233 - Mandated use of HUD's Enterprise Income Verification (EIV) System.

    2010-04-01

    ...) Project-based Voucher program under 24 CFR part 983; (v) Project-based Section 8 programs under 24 CFR... noncompliance. Failure to use the EIV system in its entirety may result in the imposition of sanctions and/or... Income Verification (EIV) System. 5.233 Section 5.233 Housing and Urban Development Office of...

  15. 46 CFR 10.233 - Obligations of the holder of a merchant mariner credential.

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Obligations of the holder of a merchant mariner credential. 10.233 Section 10.233 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY MERCHANT MARINE OFFICERS AND SEAMEN MERCHANT MARINER CREDENTIAL General Requirements for All Merchant Mariner...

  16. Separation and estimation of 229Th and 233U by alpha and gamma ray spectrometric technique

    The individual estimation of 233U and 229Th in a plancheted source made out of liquid sample were done by using an alpha and γ-ray spectrometric techniques. Estimation of 229Th in the plancheted source was done by γ-ray spectrometry and 233U by alpha spectrometry after subtracting the estimated amount of 229Th. In spite of the same alpha energy of 4.8 MeV, the individual estimation of 233U and 229Th based on present technique is superior to conventional techniques and important in the 232Th-233U fuel reprocessing cycle of AHWR and ADSs. The activity of 229Th was also radiochemically separated from its parent activity of 233U by using an ion exchange resin and the purity was checked by the above mentioned technique. (author)

  17. A 233 km Tunnel for Lepton and Hadron Colliders

    Summers, D J; Datta, A; Duraisamy, M; Luo, T; Lyons, G T

    2012-01-01

    A decade ago, a cost analysis was conducted to bore a 233 km circumference Very Large Hadron Collider (VLHC) tunnel passing through Fermilab. Here we outline implementations of $e^+e^-$, $p \\bar{p}$, and $\\mu^+ \\mu^-$ collider rings in this tunnel using recent technological innovations. The 240 and 500 GeV $e^+e^-$ colliders employ Crab Waist Crossings, ultra low emittance damped bunches, short vertical IP focal lengths, superconducting RF, and low coercivity, grain oriented silicon steel/concrete dipoles. Some details are also provided for a high luminosity 240 GeV $e^+ e^-$ collider and 1.75 TeV muon accelerator in a Fermilab site filler tunnel. The 40 TeV $p \\bar{p}$ collider uses the high intensity Fermilab $\\bar{p}$ source, exploits high cross sections for $p \\bar{p}$ production of high mass states, and uses 2 Tesla ultra low carbon steel/YBCO superconducting magnets run with liquid neon. The 35 TeV muon ring ramps the 2 Tesla superconducting magnets at 9 Hz every 0.4 seconds, uses 250 GV of superconduct...

  18. A 233 km tunnel for lepton and hadron colliders

    Summers, D. J.; Cremaldi, L. M.; Datta, A.; Duraisamy, M.; Luo, T.; Lyons, G. T. [Dept. of Physics and Astronomy, University of Mississippi-Oxford, University, MS 38677 (United States)

    2012-12-21

    A decade ago, a cost analysis was conducted to bore a 233 km circumference Very Large Hadron Collider (VLHC) tunnel passing through Fermilab. Here we outline implementations of e{sup +}e{sup -}, pp-bar , and {mu}{sup +}{mu}{sup -} collider rings in this tunnel using recent technological innovations. The 240 and 500 GeV e{sup +}e{sup -} colliders employ Crab Waist Crossings, ultra low emittance damped bunches, short vertical IP focal lengths, superconducting RF, and low coercivity, grain oriented silicon steel/concrete dipoles. Some details are also provided for a high luminosity 240 GeV e{sup +}e{sup -} collider and 1.75 TeV muon accelerator in a Fermilab site filler tunnel. The 40 TeV pp-bar collider uses the high intensity Fermilab p-bar source, exploits high cross sections for pp-bar production of high mass states, and uses 2 Tesla ultra low carbon steel/YBCO superconducting magnets run with liquid neon. The 35 TeV muon ring ramps the 2 Tesla superconducting magnets at 9 Hz every 0.4 seconds, uses 250 GV of superconducting RF to accelerate muons from 1.75 to 17.5 TeV in 63 orbits with 71% survival, and mitigates neutrino radiation with phase shifting, roller coaster motion in a FODO lattice.

  19. Statistical analysis of 233 cases in cerebovascular Diseases

    Hu Hao-Yu; Zhejiang; Jlnhua

    2000-01-01

    Objection:Monitoring Clinical in-patient constituent ratio in cerebral vasscular diseases. Methods: we monitored 233 cases of CVD in clinical in--patient 13170 cases for three years with unique registry card accroding to ICD--9. These pepole′s data was putted in computer. Age, sex, constituent ratio of each disease in CVD was observed and number of death and rank order of cases of death as well. Results: In our data, number of male is more than female(1.75:I). Constituent ratio of age is increasing at 40 years old. Main part of CH is middle age and senility, but CI age is senility. Constituent ratio of each disease is following :CI 67.74%, CH 22.13%, CT 6.06%, SAH 3.79% respectlvly. The rank order of death in CVD is third later in cardiovascular diseases and tumors. Conclusions: Our materials revealsed that it is improtant that intensiv′e care and treatment group in pead population of CVD and prevention high risk facters of CVD and health care education

  20. 45 CFR 233.35 - Computing the assistance payment under retrospective budgeting after the initial one or two...

    2010-10-01

    ... retrospective budgeting after the initial one or two months (AFDC). 233.35 Section 233.35 Public Welfare... the initial one or two months (AFDC). The State plan for AFDC shall provide: (a) After the initial one or two payment months of assistance under § 233.34, the amount of each subsequent month's...

  1. 20 CFR 404.233 - Adjustment of your guaranteed alternative when you become entitled after age 62.

    2010-04-01

    ... when you become entitled after age 62. 404.233 Section 404.233 Employees' Benefits SOCIAL SECURITY... Guaranteed Alternative for People Reaching Age 62 After 1978 But Before 1984 § 404.233 Adjustment of your guaranteed alternative when you become entitled after age 62. (a) If you do not become entitled to...

  2. 231Pa and 233Pa neutron-induced fission up to 20 MeV

    Consistency of neutron-induced fission cross section data of 231Pa and 233Pa and data extracted from transfer reactions is investigated. Present estimate of 233Pa(n,f) fission cross section is supported by smooth level density parameter systematic, validated in case of 231Pa(n,f) data description up to En=20 MeV. The fission probabilities of Pa, fissioning in 231,233Pa(n,nf) reactions, are defined by fitting (3He,d) or (3He,t) transfer reaction data

  3. Calculation of the neutron-induced fission cross section of 233Pa

    Since very recently, experimental data for the energy dependence of the 233Pa(n,f) cross section are finally available. This has stimulated a new, self-consistent cross section evaluation for the system n+233Pa in the incident neutron energy range 0.01-6 MeV. The results are quite different compared to earlier evaluation attempts. Since 233Pa is an important intermediary in the thorium based fuel cycle, its neutron reaction cross sections are key parameters in the modeling of future advanced reactor concepts

  4. Comparative studies on plutonium and 233U utilization in miniFUJI MSR

    Molten salt reactor (MSR) has many merits such as safety enhancement and capability to be used for hydrogen production. A comparative evaluation of plutonium and 233U utilization in miniFUJI MSR has been performed. Reactor grade plutonium (RGPu), weapon grade plutonium (WGPu), and super grade plutonium (SGPu) have been utilized in the present study. The reactors can obtain their criticality condition with the 233U concentration in the Th-233U fuel, RGPu concentration in Th-RGPu fuel, WGPu concentration in Th-WGPu fuel, and SGPu concentration in Th-SGPu fuel of 0.52%, 5.76%, 2.16%, and 1.96%, respectively. The Th-233U fuel results in the soft neutron spectra of miniFUJI reactor. The neutron spectra turn into harder with the enlarging of plutonium concentration in loaded fuel where Th-RGPu fuel gives the hardest neutron spectra. (author)

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

    2010-01-01

    ... the prevailing rate law for labor and agency representatives to participate in the wage survey process... REGULATIONS PREVAILING RATE SYSTEMS Prevailing Rate Determinations § 532.233 Preparation for full-scale...

  6. Improved AHWR equilibrium core cluster for self sustenance in 233U

    Advance Heavy Water Reactor (AHWR) is being designed with many advance features like negative coolant void reactivity, heat removal through natural circulation and other passive safety features. The AHWR is a 920 MWth, vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water and designed to maximise power production from thorium. The equilibrium fuel cycle is based on the conversion of naturally available thorium into fissile 233U, driven by plutonium as external fissile feed. The basic fuel cycle is based on the fact that the AHWR core should be self-sustaining in 233U. The self sustenance in 233U can be achieved by using two types of equilibrium core clusters. In this paper we have done a study where self sustenance in 233U can be achieved by using only single type of cluster. (author)

  7. 233Pa(2nth, f) cross-section determination using a fission track technique

    The 233Pa(2nth, f) cross-section has been experimentally determined for the first time using a fission track technique. It was found to be 4834 ± 57 b, which is significantly high and thus is very important for 232Th-233U-based fuel in advanced heavy-water reactors (AHWR) and accelerator-driven sub-critical systems (ADSs). This is because the 233Pa is an important intermediary in the thorium-based fuel cycle and thus its fission cross-section is a key parameter in the modeling of AHWR and ADSs. The 233Pa(2nth, f) cross-section was calculated theoretically using the TALYS computer code and found to be in good agreement with the experimental value after normalization with respect to 241Am(2nth, f). (orig.)

  8. Comparison of 233Pa(2nth, f) cross-section determined by fission track technique with TALYS 1.2

    In the last decade, an appreciable amount of work has been done in the new concepts of advanced heavy-water reactors (AHWR) and accelerator-driven sub-critical systems (ADSs) in the field of nuclear energy. In AHWR, 232Th - 233U is the primary fuel for power generation in AHWR and ADS. In the 232Th - 233U cycle, the fissile nucleus 233U is generated by two successive β-decays after a neutron capture of the fertile nucleus 232Th. The isotope 233Pa (T1/2 = 26.9 days) governs the production of 233U. Therefore, the neutron induced fission reaction with 233Pa greatly influences the production of 233U. The present work is carried out with the objective to study the 233Pa(2nth, f) reaction cross-section using the well known fission track-technique. The fission cross section has been experimentally found to be 4834 ± 57 barns, which is significantly high and thus is very important for 232Th-233U based fuel in advanced heavy water reactors (AHWR) and accelerator driven sub-critical systems (ADSs). In the present work, the 233Pa(2nth, f) cross-section has also been calculated theoretically using nuclear reaction model based computer code TALYS 1.2. The theoretically calculated 233Pa(2nth, f) cross-section found to be in good agreement with the experimental value after normalization with respect to 241Am(2nth, f). (author)

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

    The 233U storage facility at ORNL is Building 3019. The inventory stored in Building 3019 consists of 426.5 kg of 233U 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 233U in ∼ 4,000 gal. of solution. The facility is receiving additional 233U 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 233U inventory. Additionally, small quantities (233U 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 233U. The final assessment report for ORNL storage of 233U is scheduled for June 1999. The report will document the facility assessments, the specific package inspection plan, and the results of initial package inspections

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

    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 (233U). 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 233U and (2) arms control considerations, the U.S. government must decide how much of the existing 233U 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

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

    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.

  12. L233P mutation of the Tax protein strongly correlated with leukemogenicity of bovine leukemia virus.

    Inoue, Emi; Matsumura, Keiko; Soma, Norihiko; Hirasawa, Shintaro; Wakimoto, Mayuko; Arakaki, Yoshihiro; Yoshida, Takashi; Osawa, Yoshiaki; Okazaki, Katsunori

    2013-12-27

    The bovine leukemia virus (BLV) Tax protein is believed to play a crucial role in leukemogenesis by the virus. BLV usually causes asymptomatic infections in cattle, but only one-third develop persistent lymphocytosis that rarely progress after a long incubation period to lymphoid tumors, namely enzootic bovine leucosis (EBL). In the present study, we demonstrated that the BLV tax genes could be divided into two alleles and developed multiplex PCR detecting an L233P mutation of the Tax protein. Then, in order to define the relationship between the Tax protein and leukemogenicity, we examined 360 tumor samples randomly collected from dairy or breeding cattle in Japan, of which Tax proteins were categorized, for age at the time of diagnosis of EBL. The ages of 288 animals (80.0%) associated with L233-Tax and those of 70 animals (19.4%) with P233-Tax individually followed log-normal distributions. Only the two earliest cases (0.6%) with L233-Tax disobeyed the log-normal distribution. These findings suggest that the animals affected by EBL were infected with the virus at a particular point in life, probably less than a few months after birth. Median age of those with P233-Tax was 22 months older than that with L233-Tax and geometric means exhibited a significant difference (P<0.01). It is also quite unlikely that viruses carrying the particular Tax protein infect older cattle. Here, we conclude that BLV could be divided into two categories on the basis of amino acid at position 233 of the Tax protein, which strongly correlated with leukemogenicity. PMID:24139177

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

    A fuel cycle that employs 233U denatured with 238U 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 233U 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 233U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured 233U 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 233U 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. Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics

    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.

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

    The low level waste solution, generated from Thorex process during the processing of U233, contains thorium along with traces of Th228 and U233. 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 U233 could be recovered by Chitosan between pH 4 and 5. Loaded thorium and uranium could be eluted from the Chitosan by 1M HNO3 quantitatively. (author)

  16. Characteristics of Modular Fast Reactor SVBR-100 Using Thorium-Uranium (233) Fuel

    Conclusions: • The performed computations for three different types of fuel (oxide , nitride and metallic), have revealed that maximum of uranium-233 breeding ratio, which equals to 0.9, is achieved when nitride type of fuel is used. • Adding breeding zones or increasing of the core dimensions result in increasing uranium-233 breeding ratio (up to BR = 0,97 or BR = 0,96 respectively). • There is opportunity of using plutonium as initial fissile isotope to implement U-Th-Pu fuel cycle. Breeding ratio is assessed by 0,98 if nitride fuel composition (Th+Pu)N with effective density of 12.5 is used. • The obtained data have demonstrated that both for U-Th FC and U-Th-Pu FC there is an opportunity to achieve a value of U-233 BR to be over unity when using the breeding zones and slightly increased the core dimensions

  17. 231Pa and 233Pa Neutron-Induced Fission Data Analysis

    The 231Pa and 233Pa neutron-induced fission cross-section database is analyzed within the Hauser-Feshbach approach. The consistency of neutron-induced fission cross-section data and data extracted from transfer reactions is investigated. The fission probabilities of Pa, fissioning in 231,233Pa(n,nf) reactions, are defined by fitting (3He,d) or (3He,t) transfer-reaction data. The present estimate of the 233Pa(n,f) fission cross section above the emissive fission threshold is supported by smooth level-density parameter systematics, validated in the case of the 231Pa(n,f) data description up to En =20 MeV

  18. Evaluation of the fission cross sections for U-233 and U-235

    Activities of evaluation of nuclear data was started in Japan, 1963, and the results were published as JENDL-1., JENDL-2 and JENDL-3.1. The revised edition, JENDL-3.2 is now under preparation. The evaluation works of the fission cross sections for U-233 and U-235 and the encountered problems are discussed. For the 1 to 7 MeV region of U-233 cross section, the data by Poenitz adopted in JENDL-2 were discarded and finally ratio data by the Tohoku University were adopted in JENDL-3. The resolved resonance parameters for heavy nuclides including U-233 and U-235 are treated with the Reich-Moore formula in JENDL-3.2 instead of the previously used Breit and Wigner formula. (T.H.)

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

    Forsberg, C.W.; Storch, S.N. [Oak Ridge National Lab., TN (United States); Lewis, L.C. [Lockheed Martin Idaho Technology Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

    1998-07-07

    The US investigated the use of {sup 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 {sup 233}U on a large scale. Most of the {sup 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 {sup 233}U-containing materials. Because of these changes, significant activities associated with {sup 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 {sup 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.

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

    The US investigated the use of 233U 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 233U on a large scale. Most of the 233U-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 233U-containing materials. Because of these changes, significant activities associated with 233U-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 233U-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

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

    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 90Sr plume in the aquifer and measured groundwater velocities. (author). 23 refs., 4 tabs., 31 figs

  2. Breeding of 233U in the thorium–uranium fuel cycle in VVER reactors using heavy water

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U–232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement

  3. 16 CFR 23.3 - Misuse of the terms “hand-made,” “hand-polished,” etc.

    2010-01-01

    ... 16 Commercial Practices 1 2010-01-01 2010-01-01 false Misuse of the terms âhand-made,â âhand-polished,â etc. 23.3 Section 23.3 Commercial Practices FEDERAL TRADE COMMISSION GUIDES AND TRADE PRACTICE...-made,” “hand-polished,” etc. (a) It is unfair or deceptive to represent, directly or by...

  4. Registration of the Soft Red Winter Wheat Germplasm MD01W233-06-1 Resistant to Fusarium Head Blight

    MD01W233-06-1 (Reg. No., PI ) is a soft red winter wheat (SRWW) (Triticum aestivum L.) germplasm line developed at the University of Maryland and released by the Maryland Agricultural Experiment Station in 2009. MD01W233-06-1 was selected from a cross of ‘McCormick’/ ‘Choptank’ made in 2001. McCorm...

  5. 77 FR 67394 - Gulf of Mexico (GOM), Outer Continental Shelf (OCS), Western Planning Area (WPA) Lease Sale 233...

    2012-11-09

    ... Area (WPA) Lease Sale 233 and Central Planning Area (CPA) Lease Sale 231, Oil and Gas Lease Sales... Supplemental EIS for oil and gas lease sales tentatively scheduled in 2013 and 2014 in the WPA and CPA offshore... environmental and socioeconomic analyses for proposed WPA Lease Sale 233 and proposed CPA Lease Sale 231,...

  6. Breeding of 233U in the thorium-uranium fuel cycle in VVER reactors using heavy water

    Marshalkin, V. E.; Povyshev, V. M.

    2015-12-01

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U-232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement.

  7. Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

    Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M. [Russian Federal Nuclear Center All-Russian Research Institute of Experimental Physics (VNIIEF) (Russian Federation)

    2015-12-15

    A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

  8. Comparison on decay process of explosive products for 233U and weapon-grade plutonium

    Comparison on the nuclear explosive products' radioactivity, biological hazard potential , energy deposition after nuclear explosion between the model of 233U and the model of weapon- grade plutonium was made. The detail analysis of the process of above physics quantities in the model of weapon-grade Plutonium was also given

  9. 78 FR 18232 - Amendment of VOR Federal Airway V-233, Springfield, IL

    2013-03-26

    ... depiction of the airway. When V- 233 was amended in the Federal Register of August 8, 2005 (70 FR 45527... Department of Transportation (DOT) Regulatory Policies and Procedures (44 FR 11034; February 26, 1979); and..., 40113, 40120; E.O. 10854, 24 FR 9565, 3 CFR, 1959-1963 Comp., p. 389. Sec. 71.1 0 2. The...

  10. 16 CFR 233.2 - Retail price comparisons; comparable value comparisons.

    2010-01-01

    ... GUIDES AGAINST DECEPTIVE PRICING § 233.2 Retail price comparisons; comparable value comparisons. (a... fountain pens at $10, it is not dishonest for retailer Doe to advertise: “Brand X Pens, Price Elsewhere $10... here would be deceptive, since the price charged by the small suburban outlets would have no...

  11. 45 CFR 233.53 - Support and maintenance assistance (including home energy assistance) in AFDC.

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Support and maintenance assistance (including home... § 233.53 Support and maintenance assistance (including home energy assistance) in AFDC. (a) General. At State option, certain support and maintenance assistance (including home energy assistance) may...

  12. D2.3.3 Evaluation results of the LinkedUp VICI competition

    Drachsler, Hendrik

    2014-01-01

    This document D2.3.3 is the final report of Task 2.4 – Evaluation of challenge submissions. Task 2.4 is about the actual assessment of the participating projects within the LinkedUp Veni, Vidi and Vici competition on the basis of the LinkedUp Evaluation Framework (D2.2.1). The main objective of Task

  13. 46 CFR 153.233 - Separation of tanks from machinery, service and other spaces.

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Separation of tanks from machinery, service and other... Design and Equipment Cargo Containment Systems § 153.233 Separation of tanks from machinery, service and... joint: (1) Machinery spaces. (2) Service spaces. (3) Accommodation spaces. (4) Spaces for...

  14. Investigations on production of 233U using few pin thoria in existing PHWRs

    Thorium is not a fissile material and cannot be used to either start or sustain the chain reaction. Therefore, a reactor using thorium would also need either enriched uranium or plutonium to sustain the chain reaction until enough of the thorium has converted to fissile 233U. In order to retrieve and reprocess the irradiated fuel, the bundle is designed with few thoria pins and rest SEU pins. In the present study, different pin configurations of thoria in 19 and 37 element fuel clusters of Indian PHWRs have been considered. The lattice calculations have been done using the multi-group transport theory code CLUB. The variations of k∞ versus burn up are depicted in the paper. The production of 233U (considering also the decay of 233Pa into 233U) is also shown. Average discharge burn ups of the order 20 and 17 GWd/Te can be achieved with the use of thoria pins in 19 and 37 element fuel clusters respectively with appropriate bundle shift scheme. Derating of power is required during operation because of bundle power restrictions. It is found that 1 pin thoria configuration is preferable from the point of view of fuel requirements and power reduction consideration. Since 37 element fuel cluster used in 540 MWe PHWR fuel has large margins in bundle power, the restriction in power operation is much less than 19 element fuel cluster used in 220 MWe PHWR

  15. 48 CFR 852.233-70 - Protest content/alternative dispute resolution.

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Protest content... 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...

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

    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 233Pa produced by the neutron irradiation of thorium, on condition that the colloidal thorium hydroxide is irradiated in the presence of a dispersing. (author)

  17. Measured and calculated fission-product poisoning in neutron-irradiated uranium-233

    Samples of 233U and of natural thorium have been irradiated in high neutron-flux facilities, in both soft and hard neutron spectra, and for both short and long exposure times. Included are exposures resulting in depletions of more than 90 percent of the 233U in the fissile material and burnups of more than 30,000 MWd/MT in the fertile material. Fission-product poison cross sections in two energy groups (thermal and epithermal) exhibit differences between measurement and calculation that are believed to be attributable to a lack of adequate information on important fission products in the literature. Experimental results for transient absorbers in irradiated 233U give at least 20,000 b for the neutron absorption resonance integral of 149Pm. This is a factor of 15 higher than that obtained by a 1/v extrapolation of the thermal cross sections. For transient 135Xe, the measured absorption is 7.5 percent higher than that calculated using ENDF/B-IV data. Information is also provided concerning such matters as fission yields and neutron absorption of neodymium isotopes, the existence of significant transient fission-product poisons other than 135Xe and 149Sm, and the shielding of 233U by 232Th. Such shielding suggests the need for a change in the energy dependence of the 232Th thermal-neutron cross section

  18. Radiological safety experience in the fabrication of alloy plate fuels bearing 233U/Pu

    The first incidence of 233U-bearing fuel fabrication in India was the production of aluminum-clad Al-233U alloy fuel for the Kamini research reactor. The reactor physics experiments for this fuel are now being carried out in the Purnima III critical assembly, where Al-Pu alloy plate fuels will also be used. Both types of fuels were fabricated in the radiometallurgy laboratories of Bhabha Atomic Research Centre. The hazard potential of each step, evaluated from the site-specific radiological field data, is summarized. The parameters analyzed for this purpose include external and internal radiation hazards, contamination hazards, age of fuel material (i.e., time after separation), and experimental thermoluminescent detector exposure data. Gamma spectrometric data of the finished fuel plates were also analyzed for their utility in checking the material inventory. The collective dose equivalent from the fabrication operations for 12 subassemblies (9 bearing 233U and 3 bearing plutonium) was 67 mSv, arising from external exposures only. The internal exposure was nil. Fabrication of fuel plates constituted >60% of the total exposure. Fabrication of fuel subassemblies and quality control inspection at all the stages accounted for the remaining radiation exposure. Handling of 233U/Pu-bearing fuels is likely to increase in the years ahead in India. In this context, analysis of radiological field data has yielded useful guidelines for future work

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

    The electrodeposition technique to prepare alpha-counting plates of 233U for its determination is presented. To determine the optimum conditions for plating 233U 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 233U 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 233U from oxalate and nitrate solutions coming from chemical processing irradiated thorium. (Author)

  20. Recovery of 233U from waste and minimization of modifier with supercritical fluid extraction

    The supercritical fluid extraction (SFE) method was used to remove 233U from a real time tissue paper waste generated in our laboratory and resulted in about 97% of extraction efficiency. Optimization of modifier flow rate was carried out to minimize the generation of secondary liquid waste, a significant endeavour in the context of large-scale waste treatment plant. (author)

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

    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

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

    This document provides background information on the man-made radioisotope 233U. It is one of a series of four reports that map out potential national strategies for the future use and disposition of 233U pending action under the National Environmental Policy Act (NEPA). The scope of this report is separated 233U, where separated refers to nonwaste 233U or 233U that has been separated from fission products. Information on other 233U, 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 233U, the uses of 233U, and a discussion of the available facilities for storing 233U. The considerations for what fraction of the current inventory should be preserved for future use depend on several issues. First, 233U always contains a small amount of the contaminant isotope 232U. The decay products of 232U are highly radioactive and require special handling. The current inventory has a variety of qualities with regard to 232U content, ranging from 1 to about 200 ppm (on a total uranium basis). It is preferable to use 233U with the minimum amount of 232U in all applications. The second issue pertains to other isotopes of uranium mixed in with the 233U, specifically 235U and 238U. A large portion of the inventory has a high quantity of 235U associated with it. The presence of bulk amounts of 235U 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 233U. 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 applications, the

  3. 45 CFR 233.145 - Expiration of medical assistance programs under titles I, IV-A, X, XIV and XVI of the Social...

    2010-10-01

    ... 45 Public Welfare 2 2010-10-01 2010-10-01 false Expiration of medical assistance programs under titles I, IV-A, X, XIV and XVI of the Social Security Act. 233.145 Section 233.145 Public Welfare... FINANCIAL ASSISTANCE PROGRAMS § 233.145 Expiration of medical assistance programs under titles I, IV-A,...

  4. ALARA Review for the Decontamination, Deactivation and Housekeeping of the 233-S Viewing Room

    A formal as low as reasonably achievable (ALARA) review is required by BHI-SH-02, Vol. 1, Safety and Health Procedures, Procedure 1.22, 'Planning Radiological Work', when radiological conditions exceed trigger level. The level of contamination inside the viewing room of the 233-S Facility meets this criterion. This ALARA review is for task instructions 1997-03-18-005-8.3.1, 'Instructions for Routine Entries and Minor Maintenance Work at 233-S,' and 8.3.2, 'Instructions for Deactivation, Decon, and Housekeeping in Viewing Room.' The radiological work permit (RWP) request broke the two task instructions into nine separate tasks. The nine tasks identified in the RWP request were used to estimate airborne concentrations and the total exposure

  5. Review of thorium-U233 cycle thermal reactor benchmark studies

    A survey is made of many 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 the ENDF/B-IV leads to an underprediction of leakage. Extensive results from testing alternate thorium data sets are presented. For one evaluation due to Leonard they depict a possibly growing discrepance between measured integral parameters such as rho02 and I'232 and the differential data, which underpredicts them. 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

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

    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 rho02 and I232 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

  7. Emerging new options for harnessing the Th-233U cycle in India

    The recent development of the concept of Fusion Breeders especially the invention of the Fission Suppressed Blanket, coupled with parallel rapid strides in fusion technology (particularly of Tokamaks) has given a welcome new boost to the prospects of harnessing the Th-233U cycle. Their studies show that even sub-Lawson Fusion Breeders which are net consumers of electrical energy will be good enough to give adequate growth rates of nuclear generating capacity, provided the fusion bred 233U is used in either fast or thermal breeders of thermal near-breeders having high conversion ratios (>0.98). The paper presents an overview of the main results of studies underway in these areas both at the Bhabha Atomic Research Centre (BARC), Trombay and the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam

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

    The excited levels in 233Pa following the 237Np 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 233Pa 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. 233Pa(n, f) cross section up to En=8.5 MeV

    The energy dependence of the neutron-induced fission cross section of 233Pa has been measured directly for the first time from the fission threshold up to En=8.5 MeV. This reaction plays an important role in the thorium-uranium fuel cycle, and is thus of interest for the design and modeling of advanced reactor and transmutation facilities. The existing information in the ENDF/B-VI and JENDL-3.3 evaluated nuclear data files differ by a factor of two for the 233Pa(n, f) cross section values and show different fission threshold energies. Our new experimental data give lower cross section values than both evaluations and resolves the question about the threshold energy. In addition to the experimental investigation, also a new theoretical calculation of the reaction cross section has been performed with the statistical model code STATIS, showing a good agreement with the experimental data

  10. Final purification of 233U from thorium by Dowex 50x4 in Thorex process

    The 233U product obtained after a single cycle 5% tributyl phosphate/shell sol T extraction, scrubbing and stripping often contains significant amount of thorium as impurity. Further purification is normally carried out using ion exchange. The present paper summarises the results of the studies carried out to find out the various factors responsible for the extraordinary binding of thorium ion at the exchanger site while exploring the possibility of eluting total thorium using HNO3 alone

  11. Analysis of the BNL ThO2-233U exponential experiments

    The BNL ThO2--3 w/o 233U light-water-moderated exponential experiments were analyzed to evaluate (1) cross section library sets for 233U and 232Th, and (2) correlations with measured ThO2 resonance integral data. A total of six cross section library sets were evaluated, including ENDF/B-2 and ENDF/B-3 libraries for 232Th, ENDF/B-2 library for 233U, and ThO2 resonance integral correlations based on data by Weitman and Pettus, Hardy and Palowitch, and corrections to the latter data by Steen. A modified version of the LEOPARD code was used throughout this analysis. The principle results of this work are as follows: (1) The library set containing ENDF/B-2 data for 233U and ENDF/B-3 data for 232Th, together with ThO2 resonance integral correlation based on Steen's corrections to the Hardy and Palowitch data, yields the best agreement with measurements, giving an average k/sub eff/ of 0.9975 with a standard deviation of 0.0067 for the 21 analyzed configurations. (2) With respect to this ''best'' set, the ENDF/B-2 232Th data is less reactive than the corresponding ENDF/B-3 data by approximately 0.1 percent Δk. (3) The ThO2 resonance integral correlation based on data by Weitman and Pettus yields resonance integrals that are consistently higher than those produced by the correlation with Steen's values, even though the latter is normalized to an infinitely dilute resonance integral of 85.9 barns (0.5 ev cutoff), while the former is normalized to a corresponding value of 80 barns. Thus, with respect to the ''best'' set, the ThO2 resonance integral correlation based on the Weitman and Pettus data is less reactive by approximately 0.7 percent Δk

  12. Measurement and analysis of anti ν energy dependence for 233U, 238U, 239Pu

    The energy dependence of the average yield of prompt neutrons per fission (anti ν) for U233, U238 and Pu239 in the incident neutron energy range (Esub(n)) up to 5 MeV has been studied using an electrostatic accelerator. The recommended curves for anti ν (Esub(n)) has been obtained with the help of the polynomial least square method fitting. The resons of deviations from the linear dependence have been analysed

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

    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

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

    Highlights: • BWR core operating in a closed self-sustainable Th–233U 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–233U 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 233U. 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

  15. Characteristics of Modular Fast Reactor SVBR-100 Using Thorium-Uranium (233) Fuel

    Natural reserves of thorium are three times as much as those of uranium. For that reason, thorium is a very promising raw material for manufacturing an artificial fissionable isotope of uranium-233 that is formed when neutrons are absorbed by thorium. Many countries are investigating characteristics of reactors using thorium-uranium (233) fuel. First, breeding ratio (BR) is of interest because only when BR = 1, the reactor can operate in a closed fuel cycle in a mode of fuel self-providing without makeup by other fissionable isotopes. The report presents the results of calculations of neutron-physical and thermal-hydraulic characteristics of SVBR-100 - lead-bismuth cooled small power modular fast reactor using thorium-uranium (233) fuel. Reactor SVBR-100 has specific properties of inherent self-protection and passive safety. The NPP modular power-units, which power equals to a value divisible by 100 MWe, can be constructed on the basis of reactor modules SVBR-100. (author)

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

    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.

  17. Measurement of 233U fission spectrum-averaged cross sections for some threshold reactions

    The 233U fission spectrum-averaged cross sections for twelve threshold reactions were measured relative to the average cross section of 0.688 ± 0.040 mb for the 27Al(n,α)24Na reaction. The reference value was obtained by calculation using the energy dependent cross section in the Japanese Evaluated Nuclear Data Library (JENDL) Dosimetry File and the Watt-type fission spectrum in ENDF/B-VI. General agreement was seen between the measured and the calculated fission-spectrum averaged cross sections. However, there exist discrepancies of more than 10% between the measured and the calculated average cross sections for the 24Mg(n,p)24Na, 47Ti(n,p)47Sc, and 64Zn(n,p)64Cu reactions. The tendencies in the calculated-to-measured ratios are similar to those for 235U fission spectrum-averaged cross sections the authors previously measured. The measured average cross sections were also applied for the spectrum adjustment of the 233U fission neutrons using the Neutron Unfolding Package Code (NEUPAC). The adjusted spectrum is close to the Watt-type fission spectrum of 233U within the uncertainties of the obtained spectrum, although there exist some fluctuations in the ratio spectrum of the adjusted to the Watt-type

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

    Differential-neutron-emission cross sections of 232Th, 233U, 235U, 238U, 239Pu and 240Pu 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. 232Th, 233U, 235U and 238U inelastic-scattering values, poor agreement is observed for 240Pu, and a serious discrepancy exists in the case of 239Pu

  19. The calculation of prompt fission neutron spectrum for 233U(n,f) reaction by the semi-empirical method

    Chen, Yong-Jing; Min, Jia; Liu, Ting-Jin; Shu, Neng-Chuan

    2013-01-01

    The prompt fission neutron spectra for neutron-induced fission of 233U for low energy neutrons (below 6 MeV) are calculated using the nuclear evaporation theory with a semi-empirical method, in which the partition of the total excitation energy between the fission fragments for the nth+233U fission reactions are determined with the available experimental and evaluation data. The calculated prompt fission neutron spectra agree well with the experimental data. The proportions of high- energy ou...

  20. The study of 233Pa having the effect on reactivity swings during reactors start up and shutdown

    233Pa is an important nuclide in the Th-U conversion chain with a long half-life (27.4 d) and has large neutron absorption cross section, which influence the Th-U conversion ratio fuel and the operation of a Th-U fuelled reactor. On the basis of transport diffusion module and burn up module of the Dragon package, which gets the direct ratio relation- ship between 233Th, 233Pa and 233U concentration and the neutron flux levels, and gets the 233Pa having the effect on re activity swings during reactors start up and shutdown. The higher the neutron flux levels, the greater the loss of reactivity after reactors start up and the more significant increase of reactivity after reactors shutdown. Through the study of the 233Pa having the effect on reactivity swings during reactors start up and shutdown, it provides reference data for reactors using Th-U fuel with safe operation when reactors start up and shutdown. (authors)

  1. M233I Mutation in the β-Tubulin of Botrytis cinerea Confers Resistance to Zoxamide.

    Cai, Meng; Lin, Dong; Chen, Lei; Bi, Yang; Xiao, Lu; Liu, Xi-li

    2015-01-01

    Three phenotypes were detected in 161 Botrytis cinerea field isolates, including Zox(S)Car(S) (sensitive to zoxamide and carbendazim), Zox(S)Car(R) (sensitive to zoxamide and resistant to carbendazim), and Zox(R)Car(R) (resistant to zoxamide and carbendazim), but not Zox(R)Car(S) (resistant to zoxamide and sensitive to carbendazim). The baseline sensitivity to zoxamide was determined with a mean EC50 of 0.76 μg/ml. Two stable Zox(R)Car(S) isolates were obtained with a resistance factor of 13.28 and 20.43; there was a fitness penalty in mycelial growth rate, sporulation, virulence and sclerotium production. The results suggest that the resistance risk of B. cinerea to zoxamide is low where benzimidazoles have not been used. E198V, E198K and M233I, were detected in the β-tubulin of Zox(S)Car(R), Zox(R)Car(R), Zox(R)Car(S), respectively. Molecular docking indicated that position 198 in β-tubulin were targets for both zoxamide and carbendazim. The mutations at 198 prevented formation of hydrogen bonds between β-tubulin and carbendazim (E198V/K), and changed the conformation of the binding pocket of zoxamide (E198K). M233I had no effect on the binding of carbendazim but resulted in loss of a hydrogen bond between zoxamide and F200. M233 is suggested to be a unique target site for zoxamide and be very important in the function of β tubulin. PMID:26596626

  2. Gastrointestinal absorption and retention of plutonium-238 and uranium-233 in neonatal swine

    Newborn swine absorbed as much as one-half of a gavaged dose of 238Pu nitrate within 6 to 12 hr, and another one-fourth of the dose was retained in the stomach for up to 24 hr after intragastric administration. The small intestine accumulated one-third of the dose within 36 hr. Animals gavaged between 5 and 21 days of age absorbed decreasing amounts, especially at 14 and 21 days. Absorption of 233Pu (gavaged as the nitrate) by day-old pigs was similar to that of 238Pu

  3. A programme of evaluation, processing and testing of nuclear data for Th-232 and U-233

    As part of the IAEA-NDS sponsored Co-ordinated Research Programme on the intercomparison of evaluations of Actinide Neutron Nucelar Data, a programme has been undertaken at RRC for Th-232 and U-233. This paper presents results highlighting the extent to which evaluations in JENDL-2, INDIAN, ENDF/B-V, ENDL-84, FRENCH SET(1969), INDL/A-83(RUMANIAN), ENDF/B-IV and JENDL-1 files are consistent with the measured value of σ/sub c//σ/sub eta/ for Th-232 at the center of THOR assembly which emphasizes transport of neutrons in the fission source energy range

  4. Simultaneous measurement of the neutron capture and fission yields of {sup 233}U

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

    2008-07-01

    We have measured the neutron capture and fission cross section of {sup 233}U at the neutron time-of-flight facility n-TOF at CERN in the energy range from 1 eV to 1 MeV with high accuracy by using a high performance 4{pi} BaF{sub 2} Total Absorption Calorimeter (TAC) as a detection device. The method, based on the shape analysis of the TAC energy response, allowing to disentangle between {gamma}'s originating from fission and capture will be presented as well as the first very preliminary results. (authors)

  5. Simultaneous measurement of the neutron capture and fission yields of 233U

    We have measured the neutron capture and fission cross section of 233U at the neutron time-of-flight facility n-TOF at CERN in the energy range from 1 eV to 1 MeV with high accuracy by using a high performance 4π BaF2 Total Absorption Calorimeter (TAC) as a detection device. The method, based on the shape analysis of the TAC energy response, allowing to disentangle between γ's originating from fission and capture will be presented as well as the first very preliminary results. (authors)

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

    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 233U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U3O8 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

  7. Shutdown margin for high conversion BWRs operating in Th-{sup 233}U fuel cycle

    Shaposhnik, Y., E-mail: shaposhy@bgu.ac.il [NRCN – Nuclear Research Center Negev, POB 9001, Beer Sheva 84190 (Israel); Department of Nuclear Engineering, Ben-Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Elias, E. [Faculty of Mechanical Engineering, Technion – Israel Institute of Technology, Technion City 32000, Haifa (Israel)

    2014-09-15

    Highlights: • BWR core operating in a closed self-sustainable Th-{sup 233}U fuel cycle. • Shutdown Margin in Th-RBWR design. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal–hydraulic analysis includes MCPR observation. - Abstract: Several reactivity control system design options are explored in order to satisfy shutdown margin (SDM) requirements in a high conversion BWRs operating in Th-{sup 233}U fuel cycle (Th-RBWR). The studied core has an axially heterogeneous fuel assembly structure with a single fissile zone “sandwiched” between two fertile blanket zones. The utilization of an originally suggested RBWR Y-shape control rod in Th-RBWR is shown to be insufficient for maintaining adequate SDM to balance the high negative reactivity feedbacks, while maintaining fuel breeding potential, core power rating, and minimum Critical Power Ratio (CPR). Implementation of alternative reactivity control materials, reducing axial leakage through non-uniform enrichment distribution, use of burnable poisons, reducing number of pins as well as increasing pin diameter are also shown to be incapable of meeting the SDM requirements. Instead, an alternative assembly design, based on Rod Cluster Control Assembly with absorber rods was investigated. This design matches the reference ABWR core power and has adequate shutdown margin. The new concept was modeled as a single three-dimensional fuel assembly having reflective radial boundaries, using the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules.

  8. 232Th, 233Pa, and 234U capture cross-section measurements in moderated neutron flux

    The Th-U cycle was studied through the evolution of a 100 μg 232Th sample irradiated in a moderated neutron flux of 8.01014 n/cm2/s, intensity close to that of a thermal molten salt reactor. After 43 days of irradiation and 6 months of cooling, a precise mass spectrometric analysis, using both TIMS and MC-ICP-MS techniques, was performed, according to a rigorous methodology. The measured thorium and uranium isotopic ratios in the final irradiated sample were then compared with integral simulations based on evaluated data; an overall good agreement was seen. Four important thermal neutron-capture cross-sections were also extracted from the measurements, 232Th (7.34±0.21 b), 233Pa (38.34±1.78 b), 234U (106.12±3.34 b), and 235U (98.15±11.24 b). Our 232Th and 235U results confirmed existing values whereas the cross-sections of 233Pa and 234U (both key parameters) have been redefined

  9. Assessment of the available 233U cross-section evaluations in the calculation of critical benchmark experiments

    In this report we investigate the adequacy of the available 233U cross-section data for calculation of experimental critical systems. The 233U 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 233U 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 233U 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 233U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems

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

    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.

  11. A novel GH secretagogue, A233, exhibits enhanced growth activity and innate immune system stimulation in teleosts fish.

    Martinez, Rebeca; Ubieta, Kenia; Herrera, Fidel; Forellat, Alina; Morales, Reynold; de la Nuez, Ania; Rodriguez, Rolando; Reyes, Osvaldo; Oliva, Ayme; Estrada, Mario P

    2012-09-01

    In teleosts fish, secretion of GH is regulated by several hypothalamic factors that are influenced by the physiological state of the animal. There is an interaction between immune and endocrine systems through hormones and cytokines. GH in fish is involved in many physiological processes that are not overtly growth related, such as saltwater osmoregulation, antifreeze synthesis, and the regulation of sexual maturation and immune functions. This study was conducted to characterize a decapeptide compound A233 (GKFDLSPEHQ) designed by molecular modeling to evaluate its function as a GH secretagogue (GHS). In pituitary cell culture, the peptide A233 induces GH secretion and it is also able to increase superoxide production in tilapia head-kidney leukocyte cultures. This effect is blocked by preincubation with the GHS receptor antagonist [d-Lys(3)]-GHRP6. Immunoneutralization of GH by addition of anti-tilapia GH monoclonal antibody blocked the stimulatory effect of A233 on superoxide production. These experiments propose a GH-mediated mechanism for the action of A233. The in vivo biological action of the decapeptide was also demonstrated for growth stimulation in goldfish and tilapia larvae (Ptilapia larvae treated with this novel molecule. The decapeptide A233 designed by molecular modeling is able to function as a GHS in teleosts and enhance parameters of the innate immune system in the fish larvae. PMID:22707376

  12. Contribution to the study of 233U production with MOX-ThPu fuel in PWR reactor. Transition scenarios towards Th/233U iso-generating concepts in thermal spectrum. Development of the MURE fuel evolution code

    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)

  13. Calculation of the neutron induced fission cross-section of 233Pa up to 20 MeV

    Since very recently, direct measurements of the 233Pa(n,f) cross-section are available in the energy range from 1.0 to 8.5 MeV. This has stimulated a new, self-consistent, neutron cross-section evaluation for the n+233Pa system, in the incident neutron energy range 0.01-20 MeV. Since higher fission chances are involved also the lighter Pa-isotopes had to be re-evaluated in a consistent manner. The results are quite different compared to earlier evaluation attempts. Since 233Pa is a key isotope in the thorium based fuel cycle the quality of its reaction cross-sections is important for the modeling of future advanced fuel and reactor concepts. The present status of the evaluated libraries is that they differ by a factor of two in the absolute fission cross-section and also in the threshold energy value

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

    The purpose of the 233U Safe Storage Program is to enhance the safe storage of 233U-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 233U 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 model

  15. High conversion Th–U233 fuel for current generation of PWRs: Part II – 3D full core analysis

    Highlights: • Three-dimensional full core analysis of high conversion Th–U233 PWR was performed. • Thermal–hydraulic safety margins were evaluated. • Feasibility of achieving conversion ratio close to unity was demonstrated. • The major tradeoffs are lower power output and shorter fuel cycle. - Abstract: This study explores a possibility of designing a high conversion (HC) Th–U233 core for current generation of pressurized water reactors (PWRs). Increasing the conversion ratio in existing PWRs can potentially improve the utilization of natural resources, through the exploitation of vast thorium reserves and reduction in natural uranium demand. HC can be achieved through the use of heterogeneous seed-blanket (SB) Th–U233 fuel assembly design, where the supercritical seed works as a neutron supplier, while the subcritical blanket acts as U233 breeder. One of the main challenges associated with the heterogeneous SB fuel assembly designs is significant power imbalance between the seed and blanket regions caused by the high concentration of fissile material in the seed region and consequently requiring a substantial reduction in the core average power density. The main objectives of the current work are: (1) to design a high conversion SB Th–U233 fuel assembly which is directly retrofittable into existing PWRs without introducing significant modifications into the core and plant design; (2) to estimate the reasonably achievable core power density level at which reactor safety is not compromised by performing 3D coupled neutronic and thermal–hydraulic (T–H) analysis of a typical PWR core fully loaded with HC Th–U233 SB fuel. Part II of the two-part paper reports on the steady-state whole core analysis of 100% Th–U233 fueled PWR. The results of this study demonstrate the feasibility in principle of achieving conversion ratio close to unity for a Th–U233 PWR core operating at power density of 60 W/cc, in three-batch annual fuel cycle and without

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

    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

  17. 233U mass yield measurements around and within the symmetry region with the ILL Lohengrin spectrometer

    Chebboubi, A.; Kessedjian, G.; Sage, C.; Bernard, D.; Blanc, A.; Faust, H.; Köster, U.; Litaize, O.; Mutti, P.; Serot, O.

    2016-03-01

    The study of fission yields has a major impact on the characterization and understanding of the fission process and is mandatory for reactor applications. The LPSC in collaboration with ILL and CEA has developed a measurement program on fission fragment distributions at the Lohengrin spectrometer of the ILL, with a special focus on the masses constituting the heavy peak. We will present in this paper our measurement of the very low fission yields in the symmetry mass region and the heavy mass wing of the distribution for 233U thermal neutron induced fission. The difficulty due to the strong contamination by other masses with much higher yields will be addressed in the form of a new analysis method featuring the required contaminant correction. The apparition of structures in the kinetic energy distributions and possible interpretations will be discussed, such as a possible evidence of fission modes.

  18. Shutdown Margin for High Conversion BWRs Operating in Th-233U Fuel Cycle

    Shaposhnik, Yaniv; Elias, Ezra

    2013-01-01

    Several reactivity control system design options are explored in order to satisfy shutdown margin (SDM) requirements in a high conversion BWRs operating in Th-233U fuel cycle (Th-RBWR). The studied has an axially heterogeneous fuel assembly structure with a single fissile zone sandwiched between two fertile blanket zones. The utilization of an originally suggested RBWR Y-shape control rod in Th-RBWR is shown to be insufficient for maintaining adequate SDM to balance the high negative reactivity feedbacks, while maintaining fuel breeding potential, core power rating, and minimum Critical Power Ratio (CPR). Instead, an alternative assembly design, also relying on heterogeneous fuel zoning, is proposed for achieving fissile inventory ratio (FIR) above unity, adequate SDM and meeting minimum CPR limit at thermal core output matching the ABWR power. The new concept was modeled as a single 3-dimensional fuel assembly having reflective radial boundaries, using the BGCore system, which consists of the MCNP code coupl...

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

    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 233U is about 2.3), MSBR performance is only possible if there are strict constraints set on the in-line reprocessing unit: all the 233Pa must be removed from the core so that it can decay on the 233U 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

  20. 233U mass yield measurements around and within the symmetry region with the ILL Lohengrin spectrometer

    Chebboubi A.

    2016-01-01

    Full Text Available The study of fission yields has a major impact on the characterization and understanding of the fission process and is mandatory for reactor applications. The LPSC in collaboration with ILL and CEA has developed a measurement program on fission fragment distributions at the Lohengrin spectrometer of the ILL, with a special focus on the masses constituting the heavy peak. We will present in this paper our measurement of the very low fission yields in the symmetry mass region and the heavy mass wing of the distribution for 233U thermal neutron induced fission. The difficulty due to the strong contamination by other masses with much higher yields will be addressed in the form of a new analysis method featuring the required contaminant correction. The apparition of structures in the kinetic energy distributions and possible interpretations will be discussed, such as a possible evidence of fission modes.

  1. Batch extraction studies for the recovery of 233U from thoria irradiated in PHWR

    Batch equilibrium studies were carried out to optimise the extraction parameters for the recovery of 233U from thoria irradiated in PHWR. The thorium concentration and the acidity of the feed was adjusted to ca. 100 g/l and 4 M nitric acid respectively. The concentration of uranium was in the range of 1.4 g/L and it contained long lived fission product like 144Ce-144Pr, 134Cs, 137Cs, 106Ru-106Rh, 105Eu, 154Eu, 90Sr-90Y and 125Sb. 3% TBP in dodecane was used as the solvent. Four stages of batch extraction was followed by a single scrub stage of 4 M nitric acid. The scrubbed organic was stripped with 0.01 M HNO3 thrice. The stripped product was concentrated by evaporation and passed through a cation exchanger to remove the residual thorium. The results of the studies are discussed in detail. (author)

  2. 233U fuel production and 30-year utilization without reprocessing and refuelling using heavy water coolant

    This study examines the physics of a thorium fuel cycle based on generating the initial fissile (233U) fuel inventory in a Deuterium-Tritium fusion device and on operating a 600 MWth fission reactor. For both phases of the fuel cycle, the fuel form is an aqueous slurry consisting of thorium oxide micro-particles dispersed into heavy water. The slurry is the fuel carrier and the coolant. After 180 full power days in the fusion driven device, the fuel enrichment is 1.4%. The enrichments is defined as the ratio between the fissile actinides mass and the total actinides mass. After the removal of fission products, the 1.4% enriched slurry thorium-uranium fuel can be used for longer than 30 full power years in a 600 MWth critical reactor core, without adding any fissile material. The critical reactor has three zones: inner fissile, central fertile, and outer reflector. (author)

  3. Measurement and calculation of the 233Pa fission cross-section for advanced fuel cycles

    The energy dependence of the neutron-induced fission cross-section of 233Pa has been measured directly for the first time from the fission threshold up to 8.5 MeV. This fission cross-section is a key ingredient in feasibility studies on fast reactors and accelerator driven systems based on the Th-U fuel cycle. The results are at strong variance with the existing evaluations. The new experimental data give lower cross-section values and resolve the question about the threshold energy. Additionally a new theoretical calculation of the reaction cross-section has been performed with the statistical model code STATIS, showing a very good agreement with the experimental data. (authors)

  4. Health physics experience during recovery of 233U from irradiated thorium rods

    Recovery of 233U from the irradiated thorium rods (46 numbers) received from Bhabha Atomic Research Centre (BARC) was carried out successfully at Reprocessing Development Laboratory (RDL), Indira Gandhi Centre for Atomic Research (IGCAR). The reprocessing was done in five stages viz., charging of fuel rods into charging flask, decladding, dissolution, solvent extraction and reconversion. The complete operation, being first of its kind, undertaken at the centre needed extensive health physics surveillance and supervision at each stage of the operation. The operational radiation protection methods followed and the experience gained during this initial campaign in area and personnel monitoring, air monitoring and contamination are discussed. The results of routine stack monitoring and analysis of waste generated in the process are given. Special operations like decommissioning of the glove box are highlighted. A brief description of unusual occurrences is also given. (author)

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

    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 239Pu and 233U 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

  6. Investigation of the fission yields of the fast neutron-induced fission of 233U

    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 of233U, 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 γ-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 233U 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)

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

    We have used solid-state detectors to measure the kinetic energies of the coincident fission fragments in the thermal-neutron-induced fission of U235, U233 and Pu239. 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 106 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 U235, 18 MeV for U233 and 20 MeV for Pu239. 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)

  8. Measurement of average cross section for Pa-233 (n, 2n) Pa-232 reaction to neutrons with fission-type reactor spectrum

    Among some nuclides concerning thorium fuel cycle, the reaction cross sections of Pa-233 should be thoroughly investigated because of its relatively long life of 27 days half life. In the present works, the average cross section for Pa-233(n,2n)Pa-232 reaction, which has been considered to contribute to the production of troublesome concomitant U-232, was initially measured using the Pa-233 specimen as pure as possible followed by the re-irradiation in a fission-type neutron spectrum. The purest Pa-233 was produced from the first thermal neutron irradiation of Th-oxide, which was selected from the viewpoint of low Th-230 content to avoid the production of bothering Pa-231 having a large cross section for thermal neutron capture reaction. The chemically isolated Pa-233 was immediately re-irradiated with reactor neutrons having fission-type reactor spectrum in KUR, along with some flux monitors for fast neutrons. After completely decaying out Pa-233 to U-233, the chemical purification of uranium was performed and the resultant uranium isotopes were analysed with an alpha-spectrometry. By using the activity ratios of U-232/U-233, the objective cross section was evaluated to be 52.1 mbarn with an estimated overall experimental error of 10 % after correcting the inevitable bypath reaction by small amount of Pa-231 content. (author)

  9. 16 CFR 233.3 - Advertising retail prices which have been established or suggested by manufacturers (or other...

    2010-01-01

    ... Practices FEDERAL TRADE COMMISSION GUIDES AND TRADE PRACTICE RULES GUIDES AGAINST DECEPTIVE PRICING § 233.3..., therefore, deceptive. Typically, a list price is a price at which articles are sold, if not everywhere, then... insubstantial volume of sales in the area, advertising of the list price would be deceptive. (g) On the...

  10. 29 CFR 2.33 - Responsibilities of DOL, DOL social service providers and State and local governments...

    2010-07-01

    ... 29 Labor 1 2010-07-01 2010-07-01 true Responsibilities of DOL, DOL social service providers and... Organizations; Protection of Religious Liberty of Department of Labor Social Service Providers and Beneficiaries § 2.33 Responsibilities of DOL, DOL social service providers and State and local...

  11. Heavy coolant fast neutron reactor BRUS-150 for minor actinides burning and U-233 build-up

    The present paper deals with the calculational research into the performance of fast reactor BRUS-150 cooled with liquid metal coolant eutectic lead-bismuth alloy with reference to minor actinides (Np, Am, Cm) transmutation and isotopic pure U 233 build up. (authors). 10 refs., 2 figs

  12. Health physics surveillance during recovery of 233U from irradiated thorium rods at reprocessing development lab, IGCAR

    Second campaign for the recovery of 233U from the irradiated rods from CIRUS and DHRUVA reactors at BARC, was carried out successfully at Reprocessing Development Laboratory (RDL) at Indira Gandhi Centre for Atomic Research (IGCAR). Health physics surveillance was provided all through the operation. The operational radiation protection methods followed and the experience gained during the campaign are discussed in this paper. (author)

  13. 45 CFR 233.34 - Computing the assistance payment in the initial one or two months (AFDC).

    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...

  14. Application of the SCALE TSUNAMI Tools for the Validation of Criticality Safety Calculations Involving 233U

    Mueller, Don [ORNL; Rearden, Bradley T [ORNL; Hollenbach, Daniel F [ORNL

    2009-02-01

    The Radiochemical Development Facility at Oak Ridge National Laboratory has been storing solid materials containing 233U for decades. Preparations are under way to process these materials into a form that is inherently safe from a nuclear criticality safety perspective. This will be accomplished by down-blending the {sup 233}U materials with depleted or natural uranium. At the request of the U.S. Department of Energy, a study has been performed using the SCALE sensitivity and uncertainty analysis tools to demonstrate how these tools could be used to validate nuclear criticality safety calculations of selected process and storage configurations. ISOTEK nuclear criticality safety staff provided four models that are representative of the criticality safety calculations for which validation will be needed. The SCALE TSUNAMI-1D and TSUNAMI-3D sequences were used to generate energy-dependent k{sub eff} sensitivity profiles for each nuclide and reaction present in the four safety analysis models, also referred to as the applications, and in a large set of critical experiments. The SCALE TSUNAMI-IP module was used together with the sensitivity profiles and the cross-section uncertainty data contained in the SCALE covariance data files to propagate the cross-section uncertainties ({Delta}{sigma}/{sigma}) to k{sub eff} uncertainties ({Delta}k/k) for each application model. The SCALE TSUNAMI-IP module was also used to evaluate the similarity of each of the 672 critical experiments with each application. Results of the uncertainty analysis and similarity assessment are presented in this report. A total of 142 experiments were judged to be similar to application 1, and 68 experiments were judged to be similar to application 2. None of the 672 experiments were judged to be adequately similar to applications 3 and 4. Discussion of the uncertainty analysis and similarity assessment is provided for each of the four applications. Example upper subcritical limits (USLs) were

  15. Numbers of prompt neutrons per fission for U233, U235, Pu239, and Cf252

    An absolute measurement of #-v#, the average number of prompt neutrons emitted per fission, is being made for the spontaneous fission of Cf262. The relative values of #-v# are being measured for neutron-induced fission of U233, U235, and Pu239, and are being compared with the spontaneous fission #-v# of Cf252. Neutrons with energies between thermal and 15 MeV are used. Particular emphasis is put on studying the dependence of #-v# on the incident neutron energy. A fission counter containing the appropriate isotope is placed in the centre of a large cadmium-loade d liquid scintillator. Through the fissionable isotope is passed a collimated beam of neutrons. Fission events, identified by pulses from the fission counter, open an electronic gate between the large liquid scintillator and a scaler. Scintillator pulses due to capture in the scintillating solution of thermalized fission neutrons are counted during the gate. The fission neutrons are detected almost independently of energy and with very high efficiency. With this technique values of #-v# to an accuracy of 1 % are expected. (author)

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

    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

  17. Measurement of neutron capture and fission cross sections of 233U in the resonance region

    Tsekhanovich I.

    2012-02-01

    Full Text Available In the framework of studies concerning new fuel cycles and nuclear wastes incineration experimental data of the α ratio between capture and fission cross sections of 233U reactions play an important role in the Th/U cycle. The safety evaluation and the detailed performance assessment for the generation IV nuclear-energy system based on 232Th cycle strongly depend on this ratio. Since the current data are scarce and sometimes contradictory, new experimental studies are required. The measurement will take place at the neutron time-of-flight facility GELINA at Geel, designed to perform neutron cross section measurements with high incident neutron-energy resolution. A dedicated high efficiency fission ionization chamber (IC as fission fragment detector and six C6D6 liquid scintilators sensitive to γ-rays and neutrons will be used. The method, based on the IC energy response study, allowing to distinguish between gammas originating from fission and capture, in the resonance region, will be presented.

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

    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...

  19. ALARA Review of the Activation/Repair of Fire Detectors in Zone Three at 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 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

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

    Experimental details of the uranium determination in geological samples (50-1500 ppm range) by mass spectrometric isotope dilution technique (MSID) employing 233U 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 233U 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)

  1. Qualification and initial characterization of a high-purity 233U spike for use in uranium analyses

    Several high-purity 233U items potentially useful as isotope dilution mass spectrometry standards for safeguards, non-proliferation, and nuclear forensics measurements are identified and rescued from downblending. By preserving the supply of 233U materials of different pedigree for use as source materials for certified reference materials (CRMs), it is ensured that the safeguards community has high quality uranium isotopic standards required for calibration of the analytical instruments. One of the items identified as a source material for a high-purity CRM is characterized for the uranium isotope-amount ratios using thermal ionization mass spectrometry (TIMS). Additional verification measurements on this material using quadrupole inductively coupled plasma mass spectrometry (ICPMS) are also performed. As a result, the comparison of the ICPMS uranium isotope-amount ratios with the TIMS data, with much smaller uncertainties, validated the ICPMS measurement practices. ICPMS is proposed for the initial screening of the purity of items in the rescue campaign

  2. Determination of the 233Pa(n,f) reaction cross-section for thorium-fuelled reactors

    A direct measurement of the energy-dependent neutron-induced fission cross-section of 233Pa has been performed for the first time. The 233Pa isotope plays a key role in the thorium fuel cycle, serving as an intermediate isotope in the formation of the uranium fuel material. Since fission is one of the reactions determining the balance of nuclei at a given time, the cross-section is of vital importance for any calculation of a thorium-fuel-based nuclear-power device. In a first measurement series, four energies between 1.0 and 3.0 MeV were measured. The resulting average above-threshold cross-section found is lower than all literature values. (authors)

  3. Photofission cross sections of U-233 and Pu-239 near threshold induced by gamma-rays from thermal neutron capture

    The photofission cross sections of U-233 and PU-239 have been studied using monochromatic photons produced by thermal neutron capture in several materials placed in a radial beam hole of the IEA-R1, 2 MW pool type research reactor, in the energy interval from 5.43 MeV to 9.72 MeV. The gamma flux incident on the samples were measured using a (3X3) inch. NaI(Tl) crystal. The photofission fragments were detected in MAKROFOL-KG (solid state nuclear track detector) etched 30 min. in a KOH (35%wt) solution at 600C. The efficiency of the detector was obtained using a Californium-252 calibrated source and its value was (0.4323 ± 3%). The tracks were counted by means of an automatic spark counting. Analyzing the photofission data we have observed similarities between the cross sections obtained for the two samples in comparison with other authors. A structure was also observed in the U-233 cross section near the energy of 7.23 MeW. Acoording to the liquid drop model the height of the simple fission barrier were determined: (5.6 ± 0.2) MeV and (5.7 ± 0.2) MeV for U-233 and Pu-239 respectively. The relative fissionability of the samples to U-238 were also determined in each excitation energy and showed to be energy independent: (2.12 +-0.25) for U-233, and (3.32+-0.41) for Pu-239. (author)

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

    The experimental information on the 2200 m/sec values for σabs, σf, α, ν and η for 233U , 235U 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

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

    '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.'

  6. Study of (n,p) and (n,α) cross-sections for 232Th, 231Pa, 233U isotopes

    The study of neutron induced reaction cross-sections in the charged particle emission in this energy region will help us to understand the energy dependence of activation cross-sections in detail, thereby providing a complete database that will lead to better understanding of mechanisms of the nuclear reactions. The present study describes nuclear model calculations of (n,p) and (n,α) reaction cross-sections for 232Th, 231Pa and 233U isotopes

  7. Neutronic optimization in high conversion Th-233U fuel assembly with simulated annealing

    This paper reports on fuel design optimization of a PWR operating in a self sustainable Th-233U fuel cycle. Monte Carlo simulated annealing method was used in order to identify the fuel assembly configuration with the most attractive breeding performance. In previous studies, it was shown that breeding may be achieved by employing heterogeneous Seed-Blanket fuel geometry. The arrangement of seed and blanket pins within the assemblies may be determined by varying the designed parameters based on basic reactor physics phenomena which affect breeding. However, the amount of free parameters may still prove to be prohibitively large in order to systematically explore the design space for optimal solution. Therefore, the Monte Carlo annealing algorithm for neutronic optimization is applied in order to identify the most favorable design. The objective of simulated annealing optimization is to find a set of design parameters, which maximizes some given performance function (such as relative period of net breeding) under specified constraints (such as fuel cycle length). The first objective of the study was to demonstrate that the simulated annealing optimization algorithm will lead to the same fuel pins arrangement as was obtained in the previous studies which used only basic physics phenomena as guidance for optimization. In the second part of this work, the simulated annealing method was used to optimize fuel pins arrangement in much larger fuel assembly, where the basic physics intuition does not yield clearly optimal configuration. The simulated annealing method was found to be very efficient in selecting the optimal design in both cases. In the future, this method will be used for optimization of fuel assembly design with larger number of free parameters in order to determine the most favorable trade-off between the breeding performance and core average power density. (authors)

  8. Molecular Gas and Dust in the Massive Star Forming Region S 233 IR

    Rui-Qing Mao; Qin Zeng

    2004-01-01

    The massive star forming region S 233 IR is observed in the molecular lines CO J = 2-1, 3-2, NH3 (1,1), (2,2) and the 870μm dust continuum. Four submillimeter continuum sources, labelled SMM 1-4, are revealed in the 870μm dust emission. The main core, SMM1, is found to be associated with a deeply embedded near infrared cluster in the northeast; while the weaker source SMM2 coincides with a more evolved cluster in the southwest. The best fit spectral energy distribution of SMM1 gives an emissivity ofβ = 1.6, and temperatures of 32 K and 92 K for the cold- and hot-dust components. An SMM1 core mass of 246 M⊙ and a total mass of 445 M⊙ are estimated from the 870 μm dust continuum emission.SMM1 is found to have a temperature gradient decreasing from inside out, indicative of the presence of interior heating sources. The total outflow gas mass as traced by the CO J - 3-2 emission is estimated to be 35 M⊙. Low velocity outflows are also found in the NH3 (1,1) emission. The non-thermal dominant NH3 line width as well as the substantial core mass suggest that the SMM1 core is a "turbulent,massive dense core", in the process of forming a group or a cluster of stars. The much higher star formation efficiency found in the southwest cluster supports the suggestion that this cluster is more evolved than the northeast one. Large near infrared photometric variations found in the source PCS-IR93, a previously found highly polarized nebulosity, indicate an underlying star showing the FU Orionis type of behavior.

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

    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 FcγRI....../AC individuals compared with 233CC children. This genotype related effect modification may significantly influence malaria sero-epidemiological and vaccine trial studies....

  10. Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

    In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 233U isotope which has very high quality fission cross-section with thermal neutrons. 233U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2D + 3T →? 4He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li2BeF4, LiF-NaF-BeF2, Li20Sn80, natural Lithium and Li17Pb83, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li3N, Li2O, Li2O2, Li2TiO3, Li4SiO3, Li2ZrO3, LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S8-P3 approximation with Gaussian quadratures using

  11. Neutron-induced fission cross section of 233Pa between 1.0 and 3.0 MeV.

    Tovesson, F; Hambsch, F J; Oberstedt, A; Fogelberg, B; Ramström, E; Oberstedt, S

    2002-02-11

    The energy dependent neutron-induced fission cross section of 233Pa has for the first time been measured directly with monoenergetic neutrons. This nuclide is an important intermediary in a thorium based fuel cycle, and its fission cross section is a key parameter in the modeling of future advanced fuel and reactor concepts. A first experiment resulted in four cross section values between 1.0 and 3.0 MeV, establishing a fission threshold in excess of 1 MeV. Significant discrepancies were found with a previous indirect experimental determination and with model estimates. PMID:11863801

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

    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)

  13. Fractional independent yields of 141La and 142La from thermal-neutron-induced fission of 233U

    The fractional independent yields of 141La and 142La from thermal-neutron-induced fission of 233U were found to be 0.026 +- 0.006 and 0.068 +- 0.010, respectively. These yields are consistent with charge distributions for which σ = 0.56 +- 0.02 and 0.52 +- 0.02, respectively. These results are in good agreement with similar yields measured for fission of 235U, but not with those from fission of 249Cf. (author)

  14. Fission, total and neutron capture cross section measurements at ORELA for {sup 233}U, {sup 27}Al and natural chlorine

    Guber, K.H.; Spencer, R.R.; Leal, L.C.; Larson, D.C.; Santos, G. Dos; Harvey, J.A.; Hill, N.W.

    1998-08-01

    The authors have made use of the Oak Ridge Electron Linear Accelerator (ORELA) to measure the fission cross section of {sup 233}U in the neutron energy range of 0.36 eV to {approximately} 700 keV. This paper reports integral data and average cross sections. In addition they measured the total neutron cross section of {sup 27}Al and natural chlorine, as well as the capture cross section of Al over an energy range from 100 eV up to about 400 keV.

  15. Neutron-induced fission cross section of 233Pa between 1.0 and 3.0 MeV

    The energy dependent neutron-induced fission cross section of P233a has for the first time been measured directly with monoenergetic neutrons. This nuclide is an important intermediary in a thorium based fuel cycle, and its fission cross section is a key parameter in the modeling of future advanced fuel and reactor concepts. A first experiment resulted in four cross section values between 1.0 and 3.0 MeV, establishing a fission threshold in excess of 1 MeV. Significant discrepancies were found with a previous indirect experimental determination and with model estimates

  16. 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

    The Thorium cycle Th232/U233 may face brilliant perspectives through advanced concepts like molten salt reactors or accelerator driven systems but it lacks accurate nuclear data concerning some nuclei. Pa233 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 Th232(He3,p)Pa234* in which the Pa234 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 (C6D6) 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 Pa231 for a 110 keV neutron: σ(n,γ) 2.00 ± 0.14 barn. (A.C.)

  17. Measurement of the ionization yield of nuclear recoils in liquid argon at 80 and 233 keV

    Bondar, A; Dolgov, A; Grishnyaev, E; Polosatkin, S; Shekhtman, L; Shemyakina, E; Sokolov, A

    2014-01-01

    The energy calibration of nuclear recoil detectors is of primary importance to rare-event experiments such as those of direct dark matter search and coherent neutrino-nucleus scattering. In particular, such a calibration is performed by measuring the ionization yield of nuclear recoils in liquid Ar and Xe detection media, using neutron elastic scattering off nuclei. In the present work, the ionization yield for nuclear recoils in liquid Ar has for the first time been measured in the higher energy range, at 80 and 233 keV, using a two-phase Cryogenic Avalanche Detector (CRAD) and DD neutron generator. The ionization yield in liquid Ar at an electric field of 2.3 kV/cm amounted to 7.8+/-1.1 and 9.7+/-1.3 e-/keV at 80 and 233 keV respectively. Neither Jaffe model for nuclear recoil-induced ionization nor that of Thomas-Imel can consistently describe the energy dependence of the ionization yield.

  18. The structure, phase transition and molecular dynamics of [C(NH2)3]3[Sb2Br9

    The crystal structures of [C(NH2)3]3[Sb2Br9] (Gu3Sb2Br9) at 300 K and of [C(NH2)3]3[Sb2Cl9] (Gu3Sb2Cl9) at 90 and 300 K are determined. The compounds crystallize in the monoclinic space group: C 2/c. The structure is composed of Sb2X93- (X = Cl, Br) ions, which form two-dimensional layers through the crystal, and guanidinium cations. In Gu3Sb2Br9 the structural phase transformation of the first-order type is detected at 435/450 K (on cooling/heating) by the DSC and dilatometric techniques. The dielectric relaxation process in the frequency range between 75 kHz and 5 MHz over the low temperature phase indicates reorientations of weakly distorted guanidinium cations. The proton 1H NMR second-moment and spin-lattice relaxation time, T1, temperature runs for the polycrystalline Gu3Sb2Br9 sample indicate a complex cation motion. A significant dynamical non-equivalence of two guanidinium cations was found. The possible mechanism of the phase transition in Gu3Sb2Br9 is discussed on the basis of the results presented

  19. The structure, phase transition and molecular dynamics of [C(NH2)3]3[Sb2Br9

    Szklarz, P.; Zaleski, J.; Jakubas, R.; Bator, G.; Medycki, W.; Falinska, K.

    2005-04-01

    The crystal structures of [C(NH2)3]3[Sb2Br9] (Gu3Sb2Br9) at 300 K and of [C(NH2)3]3[Sb2Cl9] (Gu3Sb2Cl9) at 90 and 300 K are determined. The compounds crystallize in the monoclinic space group: C 2/c. The structure is composed of Sb2X93- (X = Cl, Br) ions, which form two-dimensional layers through the crystal, and guanidinium cations. In Gu3Sb2Br9 the structural phase transformation of the first-order type is detected at 435/450 K (on cooling/heating) by the DSC and dilatometric techniques. The dielectric relaxation process in the frequency range between 75 kHz and 5 MHz over the low temperature phase indicates reorientations of weakly distorted guanidinium cations. The proton 1H NMR second-moment and spin-lattice relaxation time, T1, temperature runs for the polycrystalline Gu3Sb2Br9 sample indicate a complex cation motion. A significant dynamical non-equivalence of two guanidinium cations was found. The possible mechanism of the phase transition in Gu3Sb2Br9 is discussed on the basis of the results presented.

  20. Emission probabilities of the KX-rays following the decay of 237 Np in equilibrium with 233 Pa

    Following participation in the international EUROMET project No. 416 and after our recent paper, concerning the measurement of the emission probability values of the main gamma-rays of 237 Np in equilibrium with 233 Pa, a complementary work has been done in the frame of the collaboration LNHB-VNIIM-KRI-IFIN (with the support of 'Ministere des Affaires Etrangeres' of France). The purpose was to determine the photon emission probabilities for the KX-rays following the decay of these two nuclides. Two different analysis methods have been used. At first, the KX-rays region was analyzed by fitting Voigt functions according to a least squares procedure, included in 'COLEGRAM' deconvolution code. In the second case, the analysis was performed by using full response functions. Thus, the work allowed the determination of the photon emission probabilities with a relative uncertainty of about 2%. This accurate set of data is useful in calculations related to the atomic level scheme of 237 Np/233 Pa and in X-ray spectrometry based applications. (authors)

  1. Environmental Assessment for the U-233 Disposition, Medical Isotope Production, and Building 3019 Complex Shutdown at the Oak Ridge National Laboratory

    The purpose of the proposed action evaluated in this environmental assessment (EA) is the processing of uranium-233 (233U) stored at the Oak Ridge National Laboratory (ORNL) and other small quantities of similar material currently stored at other U. S. Department of Energy (DOE) sites in order to render it suitable for safe, long-term, economical storage. The 233U is stored within Bldg. 3019A, which is part of the Bldg. 3019 Complex. The location of the Bldg. 3019 Complex is shown on Fig. 1.1. Additionally, the proposed action would increase the availability of medical isotopes needed for research and treatment and place the Bldg. 3019 Complex in safe and stable shutdown for transfer to the DOE program for decontamination and decommissioning (D and D). DOE has determined that there is no programmatic use for the 233U currently in storage at ORNL other than as a possible source of medical isotopes. Since 233U is a special nuclear material, continued long-term storage of the ORNL inventory in its current configuration represents a significant financial liability for DOE. Continued long-term storage in Bldg. 3019A would require major capital upgrades and retrofits to critical facility systems that have deteriorated due to aging or that may not meet current standards. Storing the material in its current form requires significant annual operating expenses to meet the material-handling requirements and to provide protection against nuclear criticality accidents or theft of the material. The ORNL inventory of 233U represents most of the readily available source of thorium-229 (229Th) in the Western Hemisphere. Actinum-225 (225Ac) and its daughter product, bismuth-213 (213Bi), are isotopes in the decay chain of 233U/229Th that are showing significant promise for ongoing cancer research, including clinical trials for treatment of acute myelogenous leukemia. These isotopes are also being explored for treatment of other cancers of the lungs, pancreas, and kidneys. Figure 1

  2. 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

    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)

  3. 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

    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)

  4. Cost-based optimizations of power density and target-blanket modularity for {sup 232}Th/{sup 233}U-based ADEP

    Krakowski, R.A.

    1995-07-01

    A cost-based parametric systems model is developed for an Accelerator-Driven Energy Production (ADEP) system based on a {sup 232}Th/{sup 233}U fuel cycle and a molten-salt (LiF/BeF{sub 2}/ThF{sub 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, {sup 233}U breeding rate, and fission product transmutation capacity.

  5. Basic characterization of 233U: Determination of age and 232U content using sector field ICP-MS, gamma spectrometry and alpha spectrometry

    The possibility to determine the age, i.e. the time since the last chemical separation, of 233U was studied using two fundamentally different measurement techniques: inductively coupled plasma mass spectrometry (ICP-MS) and gamma spectrometry. Moreover, the isotope ratio 232U/233U was measured using both alpha spectrometry and gamma spectrometry. For the two materials analysed, all measurement results were in agreement, i.e. consistent within the combined uncertainties. One of the materials was also measured using gamma spectrometry under field conditions. This measurement was also in agreement with the other results on this material

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

    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)

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

    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 233U, interesting for Th/U based nuclear fuel cycles, 241,243Am and 245Cm, 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 235U fission reaction, measured simultaneously with the same detector. Results are here reported. (authors)

  8. Determination of the 233Pa(n, f) reaction cross section from 0.5 to 10 MeV neutron energy using the transfer reaction 232Th(3He, p)234Pa

    The fission probability distributions of 232,233,234Pa and 231Th have been measured up to an excitation energy of 15 MeV, using the transfer reactions 232Th(3He, t)232Pa, 232Th(3He, d)233Pa, 232Th(3He, p)234Pa and 232Th(3He, 4He)231Th. From these measurements, the neutron induced fission cross sections of 231Pa, 233Pa and 230Th have been determined from the product of the fission probabilities of 232Pa, 233Pa and 231Th respectively with the calculated compound nucleus formation cross sections in the 231Pa+n, 233Pa+n and 230Th+n reactions. The validity of the applied method has been successfully tested with the existing neutron induced fission cross sections of 230Th and 231Pa. Special emphasis is put on the 233Pa(n, f) reaction which is of importance for thorium fueled nuclear reactors. Based on a statistical model analysis of the neutron induced fission cross section as a function of neutron energy, it has been possible to determine the barrier parameters of the 234Pa fissioning nucleus. Cross sections for the compound nucleus inelastic scattering 233Pa(n, n') and radiative capture 233Pa(n, γ) reactions have also been calculated and compared with recent evaluations

  9. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement

  10. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

    Marshalkin, V. E.; Povyshev, V. M.

    2015-12-01

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium-uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement.