Science.gov

Sample records for 232u tsentrobezhnym metodom

  1. Predicting 232U Content in Uranium

    SciTech Connect

    AJ Peurrung

    1999-01-07

    The minor isotope 232U may ultimately be used for detection or confirmation of uranium in a variety of applications. The primary advantage of 232 U as an indicator of the presence of enriched uranium is the plentiful and penetrating nature of the radiation emitted by its daughter radionuclide 208Tl. A possible drawback to measuring uranium via 232U is the relatively high uncertainty in 232U abundance both within and between material populations. An important step in assessing this problem is to ascertain what determines the 232U concentration within any particular sample of uranium. To this end, we here analyze the production and eventual enrichment of 232 U during fuel-cycle operations. The goal of this analysis is to allow approximate prediction of 232 U quantities, or at least some interpretation of the results of 232U measurements. We have found that 232U is produced via a number of pathways during reactor irradiation of uranium and is subsequently concentrated during the later enrichment of the uranium' s 235U Content. While exact calculations are nearly impossible for both the reactor-production and cascade-enrichment parts of the prediction problem, estimates and physical bounds can be provided as listed below and detailed within the body of the report. Even if precise calculations for the irradiation and enrichment were possible, the ultimate 212U concentration would still depend upon the detailed fuel-cycle history. Assuming that a thennal-diffusion cascade is used to produce highly enriched uranium (HEU), dilution of reactor-processed fuel at the cascade input and the long-term holdup of 232U within the cascade both affect the 232U concentration in the product. Similar issues could be expected to apply for the other isotope-separation technologies that are used in other countries. Results of this analysis are listed below: 0 The 232U concentration depends strongly on the uranium enrichment, with depleted uranium (DU) containing between 1600 and 8000 times

  2. Gamma-spectrometric determination of 232U in uranium-bearing materials

    NASA Astrophysics Data System (ADS)

    Zsigrai, Jozsef; Nguyen, Tam Cong; Berlizov, Andrey

    2015-09-01

    The 232U content of various uranium-bearing items was measured using low-background gamma spectrometry. The method is independent of the measurement geometry, sample form and chemical composition. Since 232U is an artificially produced isotope, it carries information about previous irradiation of the material, which is relevant for nuclear forensics, nuclear safeguards and for nuclear reactor operations. A correlation between the 232U content and 235U enrichment of the investigated samples has been established, which is consistent with theoretical predictions. It is also shown how the correlation of the mass ratio 232U/235U vs. 235U content can be used to distinguish materials contaminated with reprocessed uranium from materials made of reprocessed uranium.

  3. Electric dipole moments in {sup 230,232}U and implications for tetrahedral shapes

    SciTech Connect

    Ntshangase, S. S.; Bark, R. A.; Datta, P.; Lawrie, E. A.; Lawrie, J. J.; Lieder, R. M.; Mullins, S. M.; Aschman, D. G.; Mohammed, H.; Stankiewicz, M. A.; Bvumbi, S.; Masiteng, P. L.; Shirinda, O.; Davidson, P. M.; Nieminen, P.; Wilson, A. N.; Dinoko, T. S.; Sharpey-Shafer, J. F.; Elbasher, M. E. A.; Juhasz, K.

    2010-10-15

    The nuclei {sup 230}U and {sup 232}U were populated in the compound nucleus reactions {sup 232}Th({alpha},6n) and {sup 232}Th({alpha},4n), respectively. Gamma rays from these nuclei were observed in coincidence with a recoil detector. A comprehensive set of in-band E2 transitions were observed in the lowest lying negative-parity band of {sup 232}U while one E2 transition was also observed for {sup 230}U. These allowed B(E1;I{sup -{yields}}I{sup +}-1)/B(E2;I{sup -{yields}}I{sup -}-2) ratios to be extracted and compared with systematics. The values are similar to those of their Th and Ra isotones. The possibility of a tetrahedral shape for the negative-parity U bands appears difficult to reconcile with the measured Q{sub 2} values for the isotone {sup 226}Ra.

  4. Hyperdeformed sub-barrier fission resonances observed in {sup 232}U

    SciTech Connect

    Csige, L.; Csatlos, M.; Gacsi, Z.; Gulyas, J.; Krasznahorkay, A.; Faestermann, T.; Wirth, H.-F.; Habs, D.; Hertenberger, R.; Lutter, R.; Maier, H. J.; Thirolf, P. G.

    2009-07-15

    The fission probability of {sup 232}U has been measured using the {sup 231}Pa({sup 3}He,df) reaction with an energy resolution of 11 keV in the excitation energy region of E*=4.0-6.4 MeV. A number of sub-barrier fission resonances have been observed for the first time in the excitation energy range below E*=4.8 MeV and interpreted as rotational bands with a rotational parameter characteristic to a hyperdeformed nuclear shape (({Dirac_h}/2{pi}){sup 2}/2{theta}=1.96{+-}0.11 keV). The angular distribution of the associated fission fragments was measured to deduce the K value of the rotational bands. The fission barrier parameters of {sup 232}U have been determined by analyzing the overall features of the fission probability. A deep third minimum with an excitation energy of E{sub III}=3.2(2) MeV and rather low inner barrier height of E{sub A}=4.0(3) MeV could be established.

  5. Spectroscopy of 232U in the (p , t ) reaction: More information on 0+ excitations

    NASA Astrophysics Data System (ADS)

    Levon, A. I.; Alexa, P.; Graw, G.; Hertenberger, R.; Pascu, S.; Thirolf, P. G.; Wirth, H.-F.

    2015-12-01

    The excitation spectra in the deformed nucleus 232U have been studied by means of the (p ,t ) reaction, using the Q3D spectrograph facility at the Munich Tandem accelerator. The angular distributions of tritons were measured for 162 excitations seen in the triton spectra up to 3.25 MeV. 0+ assignments are made for 13 excited states by comparison of experimental angular distributions with the calculated ones using the chuck3 code. Assignments up to spin 6+ are made for other states. Sequences of states are selected which can be treated as rotational bands. Moments of inertia have been derived from these sequences, whose values may be considered as evidence of the two- or one-phonon nature of these 0+ excitations. Experimental data are compared with interacting boson model and quasiparticle-phonon model calculations.

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

    SciTech Connect

    Lane, Taylor; Parma, Edward J.

    2015-08-01

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

  7. Isolation and Puification of Uranium Isotopes for Measurement by Mass-Spectrometry (233, 234, 235, 236, 238U) and Alpha Spectrometry (232U)

    SciTech Connect

    Marinelli, R; Hamilton, T; Brown, T; Marchetti, A; Williams, R; Tumey, S

    2006-05-30

    This report describes a standardized methodology used by researchers from the Center for Accelerator Mass Spectrometry (CAMS) (Energy and Environment Directorate) and the Environmental Radiochemistry Group (Chemistry and Materials Science Directorate) at the Lawrence Livermore National Laboratory (LLNL) for the full isotopic analysis of uranium from solution. The methodology has largely been developed for use in characterizing the uranium composition of selected nuclear materials but may also be applicable to environmental studies and assessments of public, military or occupational exposures to uranium using in-vitro bioassay monitoring techniques. Uranium isotope concentrations and isotopic ratios are measured using a combination of Multi Collector Inductively Coupled Plasma Mass Spectrometry (MC ICP-MS), Accelerator Mass Spectrometry (AMS) and Alpha Spectrometry.

  8. The Potential Role of the Thorium Fuel Cycle in Reducing the Radiotoxicity of Long-Lived Waste - 13477

    SciTech Connect

    Hesketh, Kevin; Thomas, Mike

    2013-07-01

    The thorium (or more accurately the Th-232/U-233) fuel cycle is attracting growing interest world wide and one reason for this is the reduced radiotoxicity of long-lived waste, with the Th- 232/U-233 fuel cycle often being justified partly on the grounds of low radiotoxicity for long cooling times. This paper considers the evolution of heavy metal radiotoxicity in a Molten Salt Fast Reactor (MSFR) operating a closed Th-232/U-233 cycle during different operational phases. The paper shows that even in the MSFR core, the equilibrium radiotoxicity of the thorium fuel cycle is only reached after almost 100 years of operation. MSFR was chosen because it has many theoretical advantages that favour the Th-232/U-233 fuel cycle. Conventional solid fuel systems would be expected to behave similarly, but with even longer timescales and therefore the MSFR cycle can be used to define the limits of what is practically achievable. The results are used to argue the case that a fair approach to justifying the Th-232/U-233 breeder cycle should not quote the long term equilibrium radiotoxicity, but rather the somewhat less favourable radiotoxicity that could be achieved within the operational lifetime of the first generation of Th-232/U-233 breeder reactors. (authors)

  9. Basic characterization of highly enriched uranium by gamma spectrometry

    NASA Astrophysics Data System (ADS)

    Nguyen, Cong Tam; Zsigrai, József

    2006-05-01

    Gamma-spectrometric methods suitable for the characterization of highly enriched uranium samples encountered in illicit trafficking of nuclear materials are presented. In particular, procedures for determining the 234U, 235U, 238U, 232U and 236U contents and the age of highly enriched uranium are described. Consequently, the total uranium content and isotopic composition can be calculated. For determining the 238U and 232U contents a low-background chamber was used. In addition, age dating of uranium was also performed using low-background spectrometry.

  10. Enrichment Monitor for 235U Fuel Tubes

    SciTech Connect

    Winn, W.G.

    2001-08-22

    This report describes the performance of this prototype y-monitor of 235 Uranium enrichment. In this proposed method y-rates associated with 235U and 232U are correlated with enrichment. Instrumentation for appraising fuel tubes with this method has been assembled and tested.

  11. Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

    2015-12-01

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  12. Method of searching for neutron clusters

    NASA Astrophysics Data System (ADS)

    Dudkin, G. N.; Garapatskii, A. A.; Padalko, V. N.

    2014-10-01

    A new method of searching for neutron clusters (multineutrons) composed of neutrons bound by nuclear forces has been introduced and implemented. The method is based on the search for daughter nuclei that emerge at the nuclei cluster decay of 252Cf to neutron clusters. The effect of long-time build-up of daughter nuclei with a high atomic number and long half-life was utilized. The results are interpreted as evidence of the cluster decay of 252Cf to daughter nucleus 232U (half-life of T1/2= 68.9 years). The emergence of 232U is attributed to emission of neutron clusters consisting of eight neutrons - octaneutrons. The emission probability of octaneutrons against α-decay probability of 252Cf is defined equal to λC/λα=1.74×10-6.

  13. Radioactive dating of the elements

    NASA Technical Reports Server (NTRS)

    Cowan, John J.; Thielemann, Friedrich-Karl; Truran, James W.

    1991-01-01

    The extent to which an accurate determination of the age of the Galaxy, and thus a lower bound on the age of the universe, can be obtained from radioactive dating is discussed. Emphasis is given to the use of the long-lived radioactive nuclei Re-187, Th-232, U-238, and U-235. The nature of the production sites of these and other potential Galactic chronometers is examined along with their production ratios. Age determinations from models of nucleocosmochronology are reviewed and compared with age determination from stellar sources and age constraints form cosmological considerations.

  14. Extraction of protactinium-233 and separation from thermal neutron-irradiated thorium-232 using crown ethers

    SciTech Connect

    Jalhoom, Moayyed G.; Mohammed, Dawood A.; Khalaf, Jumah S.

    2008-07-01

    A new method was developed for the extraction and separation of {sup 233}Pa from thermal neutron-irradiated {sup 232}Th. Solutions of Pa{sup 233} were prepared in LiCI-HCl solutions from which appreciable extraction was obtained using dibenzo-18-crown-6 in 1,2-dichloroethane. The effects of cavity size, substitutions on the crown ring, type of the organic solvent, and temperature on extraction are discussed. Very high separation factors were obtained for the pairs {sup 233}Pa/{sup 232}Th (>105), {sup 233}Pa/{sup 233}U (> 1000), and {sup 232}U/{sup 232}Th (>60). (authors)

  15. Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed

    SciTech Connect

    Smirnov, A. Yu. Sulaberidze, G. A.; Alekseev, P. N.; Dudnikov, A. A.; Nevinitsa, V. A. Proselkov, V. N.; Chibinyaev, A. V.

    2012-12-15

    A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the {sup 236}U isotope is compensated by re-enrichment in the {sup 235}U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the {sup 232}U isotope are obtained.

  16. Uranium half-lives: a critical review

    SciTech Connect

    Holden, N.E.

    1981-01-01

    The experimental data are evaluated and values for the spontaneous fission half-life of /sup 238/U and the total half-lives for /sup 232/U, /sup 233/U, /sup 234/U, /sup 235/U, /sup 236/U, and /sup 238/U are recommended. Also the variation of the isotopic abundance of /sup 234/U in nature and the error involved in the assumption of secular equilibrium between /sup 234/U and /sup 238/U in the determination of the specific activity of natural uranium samples are discussed. The recommended half-life values and 95% confidence limits are: /sup 238/U spontaneous fission: 8.09 +- 0.26 x 10/sup 15/ years; /sup 232/U total: 69.8 +- 1.0 years; /sup 233/U total: 1.592 +- 0.002 x 10/sup 5/ years; /sup 234/U total: 2.454 +- 0.006 x 10/sup 5/ years; /sup 235/U total: 7.037 +- 0.011 x 10/sup 8/ years; /sup 236/U total: 2.342 +- 0.003 x 10/sup 7/ years /sup 238/U total: 4.468 +- 0.005 x 10/sup 9/ years.

  17. The concept of the use of recycled uranium for increasing the degree of security of export deliveries of fuel for light-water reactors

    NASA Astrophysics Data System (ADS)

    Alekseev, P. N.; Ivanov, E. A.; Nevinitsa, V. A.; Ponomarev-Stepnoi, N. N.; Rumyantsev, A. N.; Shmelev, V. M.; Borisevich, V. D.; Smirnov, A. Yu.; Sulaberidze, G. A.

    2010-12-01

    The present paper deals with investigation of the possibilities for reducing the risk of proliferation of fissionable materials by means of increasing the degree of protection of fresh fuel intended for light-water reactors against unsanctioned use in the case of withdrawal of a recipient country of deliveries from IAEA safeguards. It is shown that the use of recycled uranium for manufacturing export nuclear fuel makes transfer of nuclear material removed from the fuel assemblies for weapons purposes difficult because of the presence of isotope 232U, whose content increases when one attempts to enrich uranium extracted from fresh fuel. In combination with restricted access to technologies for isotope separation by means of establishing international centers for uranium enrichment, this technical measure can significantly reduce the risk of proliferation associated with export deliveries of fuel made of low-enriched uranium. The assessment of a maximum level of contamination of nuclear material being transferred by isotope 232U for the given isotope composition of the initial fuel is obtained. The concept of further investigations of the degree of security of export deliveries of fuel assemblies with recycled uranium intended for light-water reactors is suggested.

  18. Expeditious method to determine uranium in the process control samples of chemical plant separating (233)U from thoria irradiated in power reactors.

    PubMed

    Kedari, C S; Kharwandikar, B K; Banerjee, K

    2016-11-01

    Analysis of U in the samples containing a significant proportion of (232)U and high concentration of Th is of great concern. Transmutation of Th in the nuclear power reactor produces a notable quantity of (232)U (half life 68.9 years) along with fissile isotope (233)U. The decay series of (232)U is initiated with (228)Th (half life 1.9 year) and it is followed by several short lived α emitting progenies, (224)Ra, (220)Rn, (216)Po, (212)Bi and (212)Po. Even at the smallest contamination of (228)Th in the sample, a very high pulse rate of α emission is obtained, which is to be counted for the radiometric determination of [U]. A commercially available anionic type of extractant Alamine®336 is used to obtain the selective extraction of U from other alpha active elements and fission products present in the sample. Experimental conditions of liquid-liquid extraction (LLE) are optimized for obtaining maximum decontamination and recovery of U in the organic phase. The effect of some interfering ionic impurities in the sample on the process of separation is investigated. Depending on the level of the concentration of U in the samples, spectrophotometry or radiometry methods are adopted for its determination after separation by LLE. Under optimized experimental conditions, i.e. 5.5M HCl in the aqueous phase and 0.27M Alamin®336 in the organic phase, the recovery of U is about 100%, the decontamination factor with respect to Th is >2000 and the extraction of fission products like (90)Sr, (144)Ce and (134,137)Cs is negligible. The detection limit for [U] using α radiometry is 10mg/L, even in presence of >100g/L of Th in the sample. Accuracy and precision for the determination of U is also assessed. Reproducibility of results is within 5%. This method shows very good agreement with the results obtained by mass spectrometry. PMID:27591623

  19. Calculation of the radiation doses occurring in the human body for inadvertent ingestion of soil and other soil exposure pathways

    NASA Astrophysics Data System (ADS)

    Oner, F.; Okumuolu, N.

    2003-11-01

    We estimate the radiation doses in the human body, in the Gudalore region in India, following the inadvertent ingestion of soil and exposure to other soil pathways by measuring Th-232, U-238, and K-40. We estimate the equivalent dose in eleven different organs and the absorbed dose calculations for the whole body. The annual effective doses are calculated, the lowest is in Kariyasolai at 7.8 x 10(-3) mSv whereas the highest is in Ponnur at 8.9 x 10(-2) mSv. In all regions, the lowest equivalent doses through inadvertent soil ingestion are calculated in the kidney and thyroid whereas the highest doses are in the red marrow and on the bone surface.

  20. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    SciTech Connect

    Shott, Gregory

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  1. Technical Competencies for the Safe Interim Storage and Management of 233U at U.S. Department of Energy Facilities

    SciTech Connect

    Campbell, D.O.; Krichinsky, A.M.; Laughlin, S.S.; Van Essen, D.C.; Yong, L.K.

    1999-02-17

    Uranium-233 (with concomitant {sup 232}U) is a man-made fissile isotope of uranium with unique nuclear characteristics which require high-integrity alpha containment biological shielding, and remote handling. The special handling considerations and the fact that much of the {sup 233}U processing and large-scale handling was performed over a decade ago underscore the importance of identifying the people within the DOE complex who are currently working with or have worked with {sup 233}U. The availability of these key personnel is important in ensuring safe interim storage, management and ultimate disposition of {sup 233}U at DOE facilities. Significant programs are ongoing at several DOE sites with actinides. The properties of these actinide materials require many of the same types of facilities and handling expertise as does {sup 233}U.

  2. Australian Nuclear Science & Technology Organization (ANSTO) Interdicted Samples 24-Hour Report

    SciTech Connect

    Kristo, M J; Hutcheon, I D; Grant, P M; Borg, L E; Sharp, M A; Moody, K J; Conrado, C L; Wooddy, P T

    2011-01-27

    Categorization is complete. Samples 11-3-1 (NSR-F-270409-01) and 11-3-2 (NSR-F-270409-02) are depleted uranium powders of moderate purity ({approx}65-80 % U). The uranium feed stocks for 11-3-1 and 11-3-2 have both experienced a neutron flux (as demonstrated by the presence of {sup 232}U). Sample 11-3-3 is indistinguishable from a natural uranium ore concentrate of moderate purity ({approx}70-80% U). Two anomalous objects (11-3-1-4 and 11-3-2-5) were found in the material during aliquoting. These objects might be valuable for route attribution.

  3. Updated ENDL99 Cross Sections for U(n.y) and U(n,f)

    SciTech Connect

    Brown, D; Dietrich, F; Hill, T; McNabb, D

    2002-05-07

    In this note, we describe the first of two updates to the uranium isotopes in Livermore's Evaluated Neutron Data Library, ENDL99. Here, we concentrate on improving the (n, f) and (n, {gamma}) evaluations for a limited set of uranium isotopes. The first improvement consisted of creating an evaluation for {sup 232}U using a combination of fission and capture cross sections from the JENDL-3.2 database and the outgoing particle distributions from the exiting ENDL99 {sup 234}U evaluation. The second improvement consisted of updating existing (n, f) and (n, {gamma}) evaluations for uranium isotopes with A=233-238. These improvements are particularly apparent in the neutron resonance region as ENDL99 often contains gross averages over the resonances. We have propagated these updates into various Livermore application libraries.

  4. Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

    SciTech Connect

    Shmelev, A. N. Kulikov, G. G. Kurnaev, V. A. Salahutdinov, G. H. Kulikov, E. G. Apse, V. A.

    2015-12-15

    Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

  5. Annual Report on the Activities and Publications of the DHS-DNDO-NTNFC-Sponsored Post-doctoral Fellow at Los Alamos National Laboratory

    SciTech Connect

    Coleman, Magen E.; Tandon, Lav

    2012-04-24

    This report is a summary update on the projects Magen Coleman is working on as a postdoctoral fellow funded by DHS. These research projects are designed to explore different radioanalytical facets of nuclear forensics, including destructive and non-destructive analysis, chronometry measurements and modeling, spectral analysis, and actinide separations. Method development and instrumental analysis are integral parts of these projects, along with strong radiochemical knowledge. The current projects discussed here include method development for {sup 232}U analysis, the assembly of a gamma mapping system, the testing of neutron bubble detectors for neutron spectrometry, and continued work on the chronometry of plutonium and uranium materials. This report documents the work that has been performed since April 2011.

  6. Radiological Control of Water in Reactor Pond of MR Reactor in NRC 'Kurchatov Institute', During Dismantling Work - 13462

    SciTech Connect

    Stepanov, Alexey; Simirsky, Yury; Semin, Ilya; Volkovich, Anatoly; Ivanov, Oleg

    2013-07-01

    The analysis of the activity and radionuclide composition of water from the MR reactor pond for α,β,γ-ray radionuclides was made. To solve this problem we use a wide range of laboratory equipment: gamma spectrometric complex, beta spectrometric complex, vacuum alpha spectrometer, and spectrometric complex with liquid scintillator. The water from MR reactor pond contains: Cs-137 (2,6*10{sup 2} Bq/g), Co-60(1,8 Bq/g), Sr-90 (1,0*10{sup 2} Bq/g), H-3 (7,0*10{sup 3} Bq/g), and components of nuclear fuel (U-232,U-234,U-235,U-236,U-238). Therefore the cleaning water from radioactivity waste occurs to be quite a complicated radiochemical task. (authors)

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

    NASA Astrophysics Data System (ADS)

    v. d. Wense, Lars; Seiferle, Benedict; Laatiaoui, Mustapha; Thirolf, Peter G.

    2015-03-01

    Following the α decay of 233U, 229Th recoil ions are shown to be extracted in a significant amount from the MLL buffer-gas stopping cell. The produced recoil ions and subsequent daughter nuclei are mass purified with the help of a customized quadrupole mass spectrometer. The combined extraction and mass purification efficiency for 229Th3+ is determined via MCP-based measurements and via the direct detection of the 229Th α decay. A large value of (10±2)% for the combined extraction and mass purification efficiency of 229Th3+ is obtained at a mass resolution of about 1u/e. In addition to 229Th, also other α-recoil ions of the 233, 232U decay chains are addressed.

  8. Sequential determination of 210Po and uranium radioisotopes in drinking water by alpha-particle spectrometry.

    PubMed

    Benedik, L; Vasile, M; Spasova, Y; Wätjen, U

    2009-05-01

    Procedures for the sequential determination of low level (210)Po and uranium radioisotopes in drinking water by alpha-particle spectrometry are presented. After addition of (208)Po and (232)U tracers, the radionuclides were preconcentrated from water samples by co-precipitation on Fe(OH)(3) or MnO(2) at pH 9 using ammonia solution. The (210)Po source was prepared by spontaneous deposition onto a copper disc either before or after uranium separation. The uranium source for alpha-particle counting was prepared by micro co-precipitation with CeF(3). The procedures were tested on mineral water and the results obtained are compared. PMID:19231220

  9. Detection of previous neutron irradiation and reprocessing of uranium materials for nuclear forensic purposes.

    PubMed

    Varga, Zsolt; Surányi, Gergely

    2009-04-01

    The paper describes novel analytical methods developed for the detection of previous neutron irradiation and reprocessing of illicit nuclear materials, which is an important characteristic of nuclear materials of unknown origin in nuclear forensics. Alpha spectrometry and inductively coupled plasma sector-field mass spectrometry (ICP-SFMS) using solution nebulization and direct, quasi-non-destructive laser ablation as sample introduction were applied for the measurement of trace-level (232)U, (236)U and plutonium isotopes deriving from previous neutron irradiation of uranium-containing nuclear materials. The measured radionuclides and isotope ratios give important information on the raw material used for fuel production and enable confirm the supposed provenance of illicit nuclear material. PMID:19179085

  10. Nuclear Data Sheets for A = 232

    SciTech Connect

    Browne, E.

    2006-10-15

    The evaluator presents in this publication spectroscopic data and level schemes from radioactive decay and nuclear reactions for all nuclei with mass number A = 232. Highlights from this evaluation include the discovery of a new isotope of francium, {sup 232}Fr, and the study of its {beta} {sup -} decay to levels in {sup 232}Ra (2004Pe17,1990Me13). Also, it includes the first observation of spontaneous fission in {sup 232}Th and {sup 232}U, as well as the measurement of their respective partial half-lives of 1.2 (4) x 10{sup 21}y (1995Bo18) and 2.6 (5) x 10{sup 15}y (2000Bo46) for this type of decay.

  11. General constraints on the age and chemical evolution of the Galaxy

    NASA Technical Reports Server (NTRS)

    Meyer, Bradley S.; Schramm, David N.

    1986-01-01

    The formalism of Schramm and Wasserburg (1970) for determining the mean age of the elements is extended to develop as model-independent a range for the Galaxy's age as possible that takes all nuclear and meteoritic data uncertainties into account. A model-independent upper limit to that age is derived from an expansion of the equation giving the mean age of the elements in moments of the normalized effective nucleosynthesis rate. This limit depends only on the ratio of the mean time of formation of the elements to the total duration of nucleosynthesis, and on model-independent data. It is shown tht the Th-232/U-238. U-235/U-238, and Pu-244/U-238 chronometric pairs can give constraints on the relative rate of nucleosynthesis over the history of the synthesis of the solar system material.

  12. Analysis of HEU samples from the ULBA Metallurgical Plant

    SciTech Connect

    Gift, E.H.

    1995-05-01

    In early March 1994, eight highly enriched uranium (HEU) samples were collected from materials stored at the Ulba Metallurgical Plant in Oskamen (Ust Kamenogorsk), Kazakhstan. While at the plant site, portions of four samples were dissolved and analyzed by mass spectrograph at the Ulba analytical laboratory by Ulba analysts. Three of these mass spectrograph solutions and the eight HEU samples were subsequently delivered to the Y-12 Plant for complete chemical and isotopic analyses. Chemical forms of the eight samples were uranium metal chips, U0{sub 2} powder, uranium/beryllium oxide powder, and uranium/beryllium alloy rods. All were declared by the Ulba plant to have a uranium assay of {approximately}90 wt % {sup 235}U. The uranium/beryllium powder and alloy samples were also declared to range from about 8 to 28 wt % uranium. The chemical and uranium isotopic analyses done at the Y-12 Plant confirm the Ulba plant declarations. All samples appear to have been enriched using some reprocessed uranium, probably from recovery of uranium from plutonium production reactors. As a result, all samples contain some {sup 236}U and {sup 232}U and have small but measurable quantities of plutonium. This plutonium could be the result of either contamination carried over from the enrichment process or cross-contamination from weapons material. It is not the result of direct reactor exposure. Neither the {sup 232}U nor the plutonium concentrations are sufficiently high to provide a significant industrial health hazard. Both are well within established or proposed acceptance criteria for storage at Y-12. The trace metal analyses showed that, with the exception of beryllium, there are no trace metals in any of these HEU samples that pose a significant health hazard.

  13. Fission-suppressed fusion breeder on the thorium cycle and nonproliferation

    SciTech Connect

    Moir, R. W.

    2012-06-19

    Fusion reactors could be designed to breed fissile material while suppressing fissioning thereby enhancing safety. The produced fuel could be used to startup and makeup fuel for fission reactors. Each fusion reaction can produce typically 0.6 fissile atoms and release about 1.6 times the 14 MeV neutron's energy in the blanket in the fission-suppressed design. This production rate is 2660 kg/1000 MW of fusion power for a year. The revenues would be doubled from such a plant by selling fuel at a price of 60/g and electricity at $0.05/kWh for Q=P{sub fusion}/P{sub input}=4. Fusion reactors could be designed to destroy fission wastes by transmutation and fissioning but this is not a natural use of fusion whereas it is a designed use of fission reactors. Fusion could supply makeup fuel to fission reactors that were dedicated to fissioning wastes with some of their neutrons. The design for safety and heat removal and other items is already accomplished with fission reactors. Whereas fusion reactors have geometry that compromises safety with a complex and thin wall separating the fusion zone from the blanket zone where wastes could be destroyed. Nonproliferation can be enhanced by mixing {sup 233}U with {sup 238}U. Also nonproliferation is enhanced in typical fission-suppressed designs by generating up to 0.05 {sup 232}U atoms for each {sup 233}U atom produced from thorium, about twice the IAEA standards of 'reduced protection' or 'self protection.' With 2.4%{sup 232}U, high explosive material is predicted to degrade owing to ionizing radiation after a little over 1/2 year and the heat rate is 77 W just after separation and climbs to over 600 W ten years later. The fissile material can be used to fuel most any fission reactor but is especially appropriate for molten salt reactors (MSR) also called liquid fluoride thorium reactors (LFTR) because of the molten fuel does not need hands on fabrication and handling.

  14. Fission-suppressed fusion breeder on the thorium cycle and nonproliferation

    NASA Astrophysics Data System (ADS)

    Moir, R. W.

    2012-06-01

    Fusion reactors could be designed to breed fissile material while suppressing fissioning thereby enhancing safety. The produced fuel could be used to startup and makeup fuel for fission reactors. Each fusion reaction can produce typically 0.6 fissile atoms and release about 1.6 times the 14 MeV neutron's energy in the blanket in the fission-suppressed design. This production rate is 2660 kg/1000 MW of fusion power for a year. The revenues would be doubled from such a plant by selling fuel at a price of 60/g and electricity at 0.05/kWh for Q=Pfusion/Pinput=4. Fusion reactors could be designed to destroy fission wastes by transmutation and fissioning but this is not a natural use of fusion whereas it is a designed use of fission reactors. Fusion could supply makeup fuel to fission reactors that were dedicated to fissioning wastes with some of their neutrons. The design for safety and heat removal and other items is already accomplished with fission reactors. Whereas fusion reactors have geometry that compromises safety with a complex and thin wall separating the fusion zone from the blanket zone where wastes could be destroyed. Nonproliferation can be enhanced by mixing 233U with 238U. Also nonproliferation is enhanced in typical fission-suppressed designs by generating up to 0.05 232U atoms for each 233U atom produced from thorium, about twice the IAEA standards of "reduced protection" or "self protection." With 2.4% 232U, high explosive material is predicted to degrade owing to ionizing radiation after a little over 1/2 year and the heat rate is 77 W just after separation and climbs to over 600 W ten years later. The fissile material can be used to fuel most any fission reactor but is especially appropriate for molten salt reactors (MSR) also called liquid fluoride thorium reactors (LFTR) because of the molten fuel does not need hands on fabrication and handling.

  15. Technical Review Report for the Justification for 233U Content Envelope Safety Analysis Report for Packaging Model 9975-85, Addendum 2

    SciTech Connect

    West, M

    2008-07-25

    This report documents the review of Addendum 2, Justification for {sup 233}U Content Envelope, Safety Analysis Report for Packaging, prepared by Savannah River Packaging Technology (SRPT) of Savannah River National Laboratory (SRNL),--the Submittal--at the request of the Department of Energy's (DOE) National Nuclear Security Agency's (NNSA) Albuquerque Operations Office, for the shipment of 233U-bearing material from Los Alamos National Laboratory (LANL), in support of the Technical Area 18 (TA-18) Materials Relocation Program. This Addendum supplements the Safety Analysis Report for Packaging (SARP), Revision 0, and Addendum 1 to Revision 0 of the 9975 SARP (called Revision 0 of the 9975 SARP in this Addendum). The {sup 233}U-bearing items are currently stored at TA-18, awaiting shipment in the Model 9975-85 Package as a new Content Envelope, C.9. Following acceptance of this Addendum by the DOE-Headquarters Certifying Official (EM-60), and subsequent revision to the current Certificate of Compliance (CoC), the new contents will be authorized for shipment in the Model 9975-85 Package. The new Content Configuration, C.9, along with the optional Shielded-Pig Convenience Container Configuration, will be incorporated into the next revision of the Model 9975-85 Package SARP. In addition to the {sup 233}U-bearing items stored at TA-18, kilogram quantities of {sup 233}U-bearing materials are stored at Oak Ridge National Laboratory (ORNL). About one quarter of the items is Highly Enriched Uranium (HEU) as U{sub 3}O{sub 8} with {sup 233}U and {sup 232}U. Highly Enriched Uranium implies a {sup 236}U enrichment of >93%. The remaining material located at ORNL is pure {sup 233}U (>90%) with varying amounts of {sup 232}U. The form of the material is U{sub 3}O{sub 8}, UO{sub 3}, UO{sub 2}, and U metal. Additional DOE Sites may also have {sup 233}U-bearing materials for shipment.

  16. Third Minima in Thorium and Uranium Isotopes in a Self-Consistent Theory

    SciTech Connect

    McDonnell, J. D.

    2013-01-01

    Background: Well-developed third minima, corresponding to strongly elongated and reflection-asymmetric shapes associated with dimolecular configurations, have been predicted in some non-self-consistent models to impact fission pathways of thorium and uranium isotopes. These predictions have guided the interpretation of resonances seen experimentally. On the other hand, self-consistent calculations consistently predict very shallow potential-energy surfaces in the third minimum region.

    Purpose: We investigate the interpretation of third-minimum configurations in terms of dimolecular (cluster) states. We study the isentropic potential-energy surfaces of selected even-even thorium and uranium isotopes at several excitation energies. In order to understand the driving effects behind the presence of third minima, we study the interplay between pairing and shell effects.

    Methods: We use the finite-temperature superfluid nuclear density functional theory. We consider two Skyrme energy density functionals: a traditional functional SkM and a recent functional UNEDF1 optimized for fission studies.

    Results: We predict very shallow or no third minima in the potential-energy surfaces of 232Th and 232U. In the lighter Th and U isotopes with N = 136 and 138, the third minima are better developed. We show that the reflection-asymmetric configurations around the third minimum can be associated with dimolecular states involving the spherical doubly magic 132Sn and a lighter deformed Zr or Mo fragment. The potential-energy surfaces for 228,232Th and 232U at several excitation energies are presented. We also study isotopic chains to demonstrate the evolution of the depth of the third minimum with neutron number.

    Conclusions: We show that the neutron shell effect that governs the existence of the dimolecular states around the third minimum is consistent with the spherical-to-deformed shape transition in the Zr andMo isotopes around N = 58.We demonstrate that the depth of

  17. Third minima in thorium and uranium isotopes in a self-consistent theory

    NASA Astrophysics Data System (ADS)

    McDonnell, J. D.; Nazarewicz, W.; Sheikh, J. A.

    2013-05-01

    Background: Well-developed third minima, corresponding to strongly elongated and reflection-asymmetric shapes associated with dimolecular configurations, have been predicted in some non-self-consistent models to impact fission pathways of thorium and uranium isotopes. These predictions have guided the interpretation of resonances seen experimentally. On the other hand, self-consistent calculations consistently predict very shallow potential-energy surfaces in the third minimum region.Purpose: We investigate the interpretation of third-minimum configurations in terms of dimolecular (cluster) states. We study the isentropic potential-energy surfaces of selected even-even thorium and uranium isotopes at several excitation energies. In order to understand the driving effects behind the presence of third minima, we study the interplay between pairing and shell effects.Methods: We use the finite-temperature superfluid nuclear density functional theory. We consider two Skyrme energy density functionals: a traditional functional SkM* and a recent functional UNEDF1 optimized for fission studies.Results: We predict very shallow or no third minima in the potential-energy surfaces of 232Th and 232U. In the lighter Th and U isotopes with N=136 and 138, the third minima are better developed. We show that the reflection-asymmetric configurations around the third minimum can be associated with dimolecular states involving the spherical doubly magic 132Sn and a lighter deformed Zr or Mo fragment. The potential-energy surfaces for 228,232Th and 232U at several excitation energies are presented. We also study isotopic chains to demonstrate the evolution of the depth of the third minimum with neutron number.Conclusions: We show that the neutron shell effect that governs the existence of the dimolecular states around the third minimum is consistent with the spherical-to-deformed shape transition in the Zr and Mo isotopes around N=58. We demonstrate that the depth of the third minimum

  18. Fission foil measurements of neutron and proton fluences in the A0015 experiment

    NASA Technical Reports Server (NTRS)

    Frank, A. L.; Benton, E. V.; Armstrong, T. W.; Colborn, B. L.

    1995-01-01

    Results are given from sets of fission foil detectors (FFD's) (Ta-181, Bi-209, Th-232, U-238) which were included in the A0015 experiment to measure combined proton/neutron fluences. Use has been made of recent FFD high energy proton calibrations for improved accuracy of response. Comparisons of track density measurements have been made with the predictions of environmental modeling based on simple 1-D (slab) geometry. At 1 g/cm(exp 2) (trailing edge) the calculations were approximately 25 percent lower than measurements; at 13 g/cm(exp 2) (Earthside) calculations were more than a factor of 2 lower. A future 3-D modeling of the experiment is needed for a more meaningful comparison. Approximate mission proton doses and neutron dose equivalents were found. At Earthside (13 g/cm(exp 2) the dose was 171 rad and dose equivalent was 82 rem. At the trailing edge (1 g/cm(exp 2) dose was 315 rad and dose equivalent was 33 rem. The proton doses are less than expected from TLD doses by 16 percent and 37 percent, respectively. These differences can be explained by uncertainties in the proton and neutron spectra and in the method used to separate proton and neutron contributions to the measurements.

  19. Laboratory column experiments for radionuclide adsorption studies of the Culebra dolomite member of the Rustler Formation

    SciTech Connect

    Lucero, D.A.; Heath, C.E.; Brown, G.O.

    1998-04-01

    Radionuclide transport experiments were carried out using intact cores obtained from the Culebra member of the Rustler Formation inside the Waste Isolation Pilot Plant, Air Intake Shaft. Twenty-seven separate tests are reported here and include experiments with {sup 3}H, {sup 22}Na, {sup 241}Am, {sup 239}Np, {sup 228}Th, {sup 232}U and {sup 241}Pu, and two brine types, AIS and ERDA 6. The {sup 3}H was bound as water and provides a measure of advection, dispersion, and water self-diffusion. The other tracers were injected as dissolved ions at concentrations below solubility limits, except for americium. The objective of the intact rock column flow experiments is to demonstrate and quantify transport retardation coefficients, (R) for the actinides Pu, Am, U, Th and Np, in intact core samples of the Culebra Dolomite. The measured R values are used to estimate partition coefficients, (kd) for the solute species. Those kd values may be compared to values obtained from empirical and mechanistic adsorption batch experiments, to provide predictions of actinide retardation in the Culebra. Three parameters that may influence actinide R values were varied in the experiments; core, brine and flow rate. Testing five separate core samples from four different core borings provided an indication of sample variability. While most testing was performed with Culebra brine, limited tests were carried out with a Salado brine to evaluate the effect of intrusion of those lower waters. Varying flow rate provided an indication of rate dependent solute interactions such as sorption kinetics.

  20. Comparison of the radiological hazard of thorium and uranium spent fuels from VVER-1000 reactor

    NASA Astrophysics Data System (ADS)

    Frybort, Jan

    2014-11-01

    Thorium fuel is considered as a viable alternative to the uranium fuel used in the current generation of nuclear power plants. Switch from uranium to thorium means a complete change of composition of the spent nuclear fuel produced as a result of the fuel depletion during operation of a reactor. If the Th-U fuel cycle is implemented, production of minor actinides in the spent fuel is negligible. This is favourable for the spent fuel disposal. On the other hand, thorium fuel utilisation is connected with production of 232U, which decays via several alpha decays into a strong gamma emitter 208Tl. Presence of this nuclide might complicate manipulations with the irradiated thorium fuel. Monte-Carlo computation code MCNPX can be used to simulate thorium fuel depletion in a VVER-1000 reactor. The calculated actinide composition will be analysed and dose rate from produced gamma radiation will be calculated. The results will be compared to the reference uranium fuel. Dependence of the dose rate on time of decay after the end of irradiation in the reactor will be analysed. This study will compare the radiological hazard of the spent thorium and uranium fuel handling.

  1. Radiogenic Heat Production of Rock from Three Rivers in Osun State of Nigeria

    NASA Astrophysics Data System (ADS)

    Alabi, O. O.; Akinluyi, F. O.; Ojo, M. O.; Adebo, B. A.

    Ten fresh rock samples were collected from three rivers in Osun State, namely Erin-Ijesha (EI), Osun-Osogbo river (OS) and Ishasha river in Edunabon near Ile-Ife (IS). The study area is underlain by the Precambrian Basement Complex of southwestern Nigeria. This is to determine their radioactive heat production and the contribution of each radionuclide content. The radiogenic heat production was determined by spectrometer which gives the area photopeak of the radionuclides contribution. These photo peaks were later converted to Bq Kg-1 and part per million (ppm) for radiogenic heat computation. The result shows that concentration and rate of heat production of 40K, 238U and 232Th in the samples varies significantly with geological location. The total heat production ranges from 8.21 to 235.82 pW kg-1. The highest concentration and heat production is recorded in Quatz of Osun-Osogbo rivers and the heat produced by 40k is highest in six samples. It is also noted that rock samples from Erin-Ijesha river are associated with high heat production of 232U.

  2. Elution behaviour of alpha-recoil atoms into etchant and ovservation of their tracks on the mica surface

    NASA Astrophysics Data System (ADS)

    Hashimoto, Tetsuo; Komatsu, Shigemi; Kido, Kazuo; Sotobayashi, Takeshi

    1980-12-01

    Muscovite samples, which were irradiated with alpha-recoil atoms emitted from a thinly electrodeposited 232U-source in a vacuum chamber of about 10 -2 Torr, were subjected to a chemical etching treatment with a hydrofluoric acid solution to develop alpha-recoil tracks. The transferred alpha-activities of 224Ra and 212Po, supported by 212Pb, on the mica surface were repeatedly measured after every etching treatment. The results showed that the 224Ra could be rapidly eluted out at earlier etching stages, in contrast to appreciably delayed elution of 212Po. These findings, along with annealing experiments on mica, imply that the recoil range of 224Ra, originated from the parent 228Th by a single decay process, is shorter than the total recoil range of 212Po, which can penetrate partially into inner mica layers through its preceding multiple alpha-decay processes after injection of its precursors into the mica. Scanning electron and phase-contrast microscopic observation of the etched mica surfaces indicated an apparent dependence of the recoil-track etch pit size on the number of succesive alpha-decays.

  3. Parametric Studies for 233U Gamma Spectrometry

    SciTech Connect

    Scheffing, C.C.; Krichinsky, A.

    2004-01-01

    Quantification of special nuclear material is a necessary aspect to assuring material accountability and is often accomplished using non-destructive gamma spectrometry. For 233U, gamma rays are affected by matrix and packaging attenuation and by a strong Compton continuum from decay products of 232U (inherently found in 233U) that obscure 233U gamma photopeaks. This project, based on current work at the national repository for separated 233U located at Oak Ridge National Laboratory (ORNL), explores the effects of various parameters on the quantification of 233U– including material form and geometry. Using an attenuation correction methodology for calculating the mass of 233U from NDA analysis, a bias of almost 75% less than the actual 233U mass was identified. The source of the bias needs to be understood at a more fundamental level for further use of this quantification method. Therefore, controlled experiments using well characterized packages of 233U were conducted at the repository and are presented in this paper.

  4. Nuclear forensic analysis of uranium oxide powders interdicted in Victoria, Australia

    DOE PAGESBeta

    Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael; Williams, Ross; Lindvall, Rachel; Eppich, Gary; Roberts, Sarah; Borg, Lars; Gaffney, Amy; Plaue, Jonathan; et al

    2015-04-13

    Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K2(UO2)3O4·4H2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarily UO3·2H2O, with minor phases ofmore » U3O8 and UO2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.« less

  5. In-beam studies of high-spin states of actinide nuclei

    SciTech Connect

    Stoyer, M.A. . Nuclear Science Div. California Univ., Berkeley, CA . Dept. of Chemistry)

    1990-11-15

    High-spin states in the actinides have been studied using Coulomb- excitation, inelastic excitation reactions, and one-neutron transfer reactions. Experimental data are presented for states in {sup 232}U, {sup 233}U, {sup 234}U, {sup 235}U, {sup 238}Pu and {sup 239}Pu from a variety of reactions. Energy levels, moments-of-inertia, aligned angular momentum, Routhians, gamma-ray intensities, and cross-sections are presented for most cases. Additional spectroscopic information (magnetic moments, M{sub 1}/E{sub 2} mixing ratios, and g-factors) is presented for {sup 233}U. One- and two-neutron transfer reaction mechanisms and the possibility of band crossings (backbending) are discussed. A discussion of odd-A band fitting and Cranking calculations is presented to aid in the interpretation of rotational energy levels and alignment. In addition, several theoretical calculations of rotational populations for inelastic excitation and neutron transfer are compared to the data. Intratheory comparisons between the Sudden Approximation, Semi-Classical, and Alder-Winther-DeBoer methods are made. In connection with the theory development, the possible signature for the nuclear SQUID effect is discussed. 98 refs., 61 figs., 21 tabs.

  6. Rapid determination of 226Ra and uranium isotopes in solid samples by fusion with lithium metaborate and alpha spectrometry.

    PubMed

    Bojanowski, R; Radecki, Z; Piekoś, R

    2002-07-01

    A simple and rapid method has been developed to determine 226Ra in rocks, soils, and sediments. Samples are decomposed by fusion with lithium metaborate and the melt is dissolved in a solution containing sulfates and citric acid. During the dissolution, a fine suspension of mixed barium and radium sulfates is formed. The microcrystals are collected on a membrane filter (pore size 0.1 microm) and analysed in an alpha spectrometer. Application of a 133Ba tracer enables us to assess the loss of the analyte, which only rarely exceeds 10%. All analytical operations, beginning from sample decomposition to source preparation for alpha spectrometry, can be accomplished within 1 or 2 h. With uranium determination, the filtrate is spiked with a 232U tracer and passed through a column loaded with a Dowex AG (1 x 4) anion-exchange resin in the sulfate form. Interfering elements are eluted with dilute sulfuric acid followed by concentrated hydrochloric acid. Uranium is eluted with water, electrodeposited on silver discs, and analysed in the alpha spectrometer. The method was tested on reference soil and sediment materials and was found to be accurate within the estimated uncertainties. PMID:12920318

  7. Shippingport LWBR (Th/U Oxide) Fuel Characteristics for Disposal Criticality Analysis

    SciTech Connect

    L. L. Taylor; H. H. Loo

    1999-09-01

    Department of Energy (DOE)-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into eight characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. The Shippingport Light Water Breeder Reactor (LWBR) fuels incorporate more of the conventional materials (zirconium cladding/heavy metal oxides) and fabrication details (rods and spacers) that make them comparable to a typical commercial fuel assembly. The LWBR seed/blanket configuration tested a light-water breeder concept with Th-232/U-233 binary fuel matrix. Reactor design used several assembly configurations at different locations within the same core . The seed assemblies contain the greatest fissile mass per (displaced) unit volume, but the blanket assemblies actually contain more fissile mass in a larger volume; the atom-densities are comparable.

  8. Tags to Track Illicit Uranium and Plutonium

    SciTech Connect

    Haire, M. Jonathan; Forsberg, Charles W.

    2007-07-01

    With the expansion of nuclear power, it is essential to avoid nuclear materials from falling into the hands of rogue nations, terrorists, and other opportunists. This paper examines the idea of detection and attribution tags for nuclear materials. For a detection tag, it is proposed to add small amounts [about one part per billion (ppb)] of {sup 232}U to enriched uranium to brighten its radioactive signature. Enriched uranium would then be as detectable as plutonium and thus increase the likelihood of intercepting illicit enriched uranium. The use of rare earth oxide elements is proposed as a new type of 'attribution' tag for uranium and thorium from mills, uranium and plutonium fuels, and other nuclear materials. Rare earth oxides are chosen because they are chemically compatible with the fuel cycle, can survive high-temperature processing operations in fuel fabrication, and can be chosen to have minimal neutronic impact within the nuclear reactor core. The mixture of rare earths and/or rare earth isotopes provides a unique 'bar code' for each tag. If illicit nuclear materials are recovered, the attribution tag can identify the source and lot of nuclear material, and thus help police reduce the possible number of suspects in the diversion of nuclear materials based on who had access. (authors)

  9. Heating and melting of small icy satellites by the decay of Al-26

    NASA Technical Reports Server (NTRS)

    Prialnik, Dina; Bar-Nun, Akiva

    1990-01-01

    The effect of radiogenic heating due to Al-26 on the thermal evolution of small icy satellites is studied. The object is to find the extent of internal melting as a function of the satellite radius and of the initial Al-26 abundance. The implicit assumption, based on observations of young stars, is that planet and satellite accretion occurred on a time scale of about 10 to the 6th yr (comparable with the lifetime of Al-26. The icy satellites are modeled as spheres of initially amorphous ice, with chondritic abundances of K-40, Th-232, U-235, and U-238, corresponding to an ice/dust mass ratio of 1. Evolutionary calculations are carried out, spanning 4.5 x 10 to the 9th yr, for different combinations of the two free parameters. Heat transfer by subsolidus convection is neglected for these small satellites. The main conclusion is that the initial Al-26 abundance capable of melting icy bodies of satellite size to a significant extent is more than 10 times lower than that prevailing in the interstellar medium (or that inferred from the Ca-Al rich inclusions of the Allende meteorite, about 7 x 10 to the -7th by mass).

  10. Natural radioactivity study in soil samples of South Konkan, Maharashtra, India.

    PubMed

    Dhawal, S J; Phadatare, M R; Thorat, N D; Kulkarni, G S; Pawar, S H

    2013-12-01

    This study assesses the level of natural radioactivity due to radionuclides, ²³⁸U, ²³²Th and ⁴⁰K, in 50 soil samples collected from South Konkan, Maharashtra, India. The mean activity concentrations of ²³⁸U, ²³²Th and ⁴⁰K are 44.97 ± 1.22 Bq kg⁻¹, 59.70 ± 2.17 Bq kg⁻¹ and 217.51 ± 8.75 Bq kg⁻¹, respectively, measured from all the soil samples studied. The good correlation between activity concentration of U-238 and Th-232; U-238 and K-40 as well as between activity concentration of Th-232 and K-40 was observed. The average calculated absorbed dose rate in air (68.08 nGy h⁻¹) was found to be higher than the world average of 57 nGy h⁻¹ (UNSCEAR 2000). Radium equivalent activity for all the villages was found to be lower than the worldwide value. The values of external hazard index and internal hazard index determined from all the soil samples were found to be within recommended limit. The calculated average annual effective dose was found to be 0.42 mSv y⁻¹, and it is lower than the worldwide value of 0.46 mSv y⁻¹.The annual effective dose values calculated from present study were comparable with previous studies carried out in other countries and in India. The data established from the study can be useful as baseline information on natural radioactivity in South Konkan, Maharashtra, India. PMID:23704360

  11. Identification of a Novel Di-D-Fructofuranose 1,2’:2,3’ Dianhydride (DFA III) Hydrolysis Enzyme from Arthrobacter aurescens SK8.001

    PubMed Central

    Yu, Shuhuai; Wang, Xiao; Zhang, Tao; Stressler, Timo; Fischer, Lutz; Jiang, Bo; Mu, Wanmeng

    2015-01-01

    Previously, a di-D-fructofuranose 1,2’:2,3’ dianhydride (DFA III)-producing strain, Arthrobacter aurescens SK8.001, was isolated from soil, and the gene cloning and characterization of the DFA III-forming enzyme was studied. In this study, a DFA III hydrolysis enzyme (DFA IIIase)-encoding gene was obtained from the same strain, and the DFA IIIase gene was cloned and expressed in Escherichia coli. The SDS-PAGE and gel filtration results indicated that the purified enzyme was a homotrimer holoenzyme of 145 kDa composed of subunits of 49 kDa. The enzyme displayed the highest catalytic activity for DFA III at pH 5.5 and 55°C, with specific activity of 232 U mg-1. Km and Vmax for DFA III were 30.7 ± 4.3 mM and 1.2 ± 0.1 mM min-1, respectively. Interestingly, DFA III-forming enzymes and DFA IIIases are highly homologous in amino acid sequence. The molecular modeling and docking of DFA IIIase were first studied, using DFA III-forming enzyme from Bacillus sp. snu-7 as a template. It was suggested that A. aurescens DFA IIIase shared a similar three-dimensional structure with the reported DFA III-forming enzyme from Bacillus sp. snu-7. Furthermore, their catalytic sites may occupy the same position on the proteins. Based on molecular docking analysis and site-directed mutagenesis, it was shown that D207 and E218 were two potential critical residues for the catalysis of A. aurescens DFA IIIase. PMID:26555784

  12. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels. PMID:23274823

  13. Characterization of the Oxidation State of 229 Th Recoils Implanted in MgF2 for the Search of the Low-lying 229 Th Isomeric State

    NASA Astrophysics Data System (ADS)

    Barker, Beau; Meyer, Edmund; Schacht, Mike; Collins, Lee; Wilkerson, Marianne; Zhao, Xinxin

    2016-05-01

    The low-lying (7.8 eV) isomeric state in 229 Th has the potential to become a nuclear frequency standard. 229 Th recoils from 233 U decays have been collected in MgF2 for use in the direct search of the transition. Of interest is the oxidation state of the implanted 229 Th atoms as this can have an influence on the decay mechanisms and photon emission rate. Too determine the oxidation state of the implanted 229 Th recoils we have employed laser induced florescence (LIF), and plan-wave pseudopotential DFT calculations to search for emission from thorium ions in oxidation states less than + 4. Our search focused on detecting emission from Th3+ ions. The DFT calculations predicted the Th3+ state to be the most likely to be present in the crystal after Th4+. We also calculated the band structure for the Th3+ doped MgF2 crystal. For LIF spectra a number of excitation wavelengths were employed, emission spectra in the visible to near-IR were recorded along with time-resolved emission spectra. We have found no evidence for Th3+ in the MgF2 plates. We also analyzed the detection limit of our apprentice and found that the minimum number of Th3+ atoms that we could detect is quite small compared to the number of implanted 229 Th recoils. The number of implanted 229 Th recoils was derived from a γ-ray spectrum by monitoring emission from the daughters of 228 Th. These were present in the MgF2 plates due to a 232 U impurity, which decays to 228 Th, in the source. LA-UR-16-20442.

  14. Sequential extraction procedure for determination of uranium, thorium, radium, lead and polonium radionuclides by alpha spectrometry in environmental samples

    NASA Astrophysics Data System (ADS)

    Oliveira, J. M.; Carvalho, F. P.

    2006-01-01

    A sequential extraction technique was developed and tested for common naturally-occurring radionuclides. This technique allows the extraction and purification of uranium, thorium, radium, lead, and polonium radionuclides from the same sample. Environmental materials such as water, soil, and biological samples can be analyzed for those radionuclides without matrix interferences in the quality of radioelement purification and in the radiochemical yield. The use of isotopic tracers (232U, 229Th, 224Ra, 209Po, and stable lead carrier) added to the sample in the beginning of the chemical procedure, enables an accurate control of the radiochemical yield for each radioelement. The ion extraction procedure, applied after either complete dissolution of the solid sample with mineral acids or co-precipitation of dissolved radionuclide with MnO2 for aqueous samples, includes the use of commercially available pre-packed columns from Eichrom® and ion exchange columns packed with Bio-Rad resins, in altogether three chromatography columns. All radioactive elements but one are purified and electroplated on stainless steel discs. Polonium is spontaneously plated on a silver disc. The discs are measured using high resolution silicon surface barrier detectors. 210Pb, a beta emitter, can be measured either through the beta emission of 210Bi, or stored for a few months and determined by alpha spectrometry through the in-growth of 210Po. This sequential extraction chromatography technique was tested and validated with the analysis of certified reference materials from the IAEA. Reproducibility was tested through repeated analysis of the same homogeneous material (water sample).

  15. Development of an Alternative Release Limit for a Former Uranium and Thorium Processing Plant in Cushing Oklahoma

    SciTech Connect

    Thatcher, A.H.

    2007-07-01

    The purpose of this presentation will be to describe how, through dose modeling and analysis, a complex site was able to obtain an Alternative Release Limit (ARL) that adequately protected the environment, met regulatory approval, and saved money in the process. The Kerr-McGee Refinery Site in Cushing, OK supported an experimental facility that processed nuclear fuel materials from 1963 to 1966. Radiological contaminants at the site as a result of operations consist of natural thorium and isotopes of uranium (Th-228, Th-232, U-234, U-235 and U-238). Site contamination existed in both surface and sub-surface soils and within a shallow aquifer. After the soil was remediated to acceptable regulatory limits, however, the potential existed for residual groundwater contamination to result in exposure to individuals following site closure. Traditional exposure pathway analysis for the resident farmer seemed to indicate that this exposure was excessive. A closer look at the exposure pathways present in this rural location showed that groundwater contamination existed in a shallow aquifer insufficient to support significant irrigation activities and was of sufficiently poor water quality that it could not be used for drinking water. Through the determination of aquifer yield pumping tests, agreement from the Oklahoma Department of Environmental Quality, and sensitivity and uncertainty analysis using Monte Carlo techniques, it was shown that the average member of the critical population was adequately protected in the current site configuration without further remediation. This paper describes the analytical methods and models used to apply the general dose limit of 0.25 mSv yr{sup -1} (25 mrem yr{sup -1}) to the particulars of the Cushing Site, and demonstrates how these methods achieved a much higher ARL for total uranium in groundwater that was accepted by the regulators and achieved significant savings for the Licensee. (authors)

  16. Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Novitrian, Waris, Abdul; Ismail, Suzuki, Mitsutoshi; Saito, Masaki

    2014-09-01

    Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by convertion rasio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loding scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

  17. Analysis on fuel breeding capability of FBR core region based on minor actinide recycling doping

    SciTech Connect

    Permana, Sidik; Novitrian,; Waris, Abdul; Ismail; Suzuki, Mitsutoshi; Saito, Masaki

    2014-09-30

    Nuclear fuel breeding based on the capability of fuel conversion capability can be achieved by conversion ratio of some fertile materials into fissile materials during nuclear reaction processes such as main fissile materials of U-233, U-235, Pu-239 and Pu-241 and for fertile materials of Th-232, U-238, and Pu-240 as well as Pu-238. Minor actinide (MA) loading option which consists of neptunium, americium and curium will gives some additional contribution from converted MA into plutonium such as conversion Np-237 into Pu-238 and it's produced Pu-238 converts to Pu-239 via neutron capture. Increasing composition of Pu-238 can be used to produce fissile material of Pu-239 as additional contribution. Trans-uranium (TRU) fuel (Mixed fuel loading of MOX (U-Pu) and MA composition) and mixed oxide (MOX) fuel compositions are analyzed for comparative analysis in order to show the effect of MA to the plutonium productions in core in term of reactor criticality condition and fuel breeding capability. In the present study, neptunium (Np) nuclide is used as a representative of MAin trans-uranium (TRU) fuel composition as Np-MOX fuel type. It was loaded into the core region gives significant contribution to reduce the excess reactivity in comparing to mixed oxide (MOX) fuel and in the same time it contributes to increase nuclear fuel breeding capability of the reactor. Neptunium fuel loading scheme in FBR core region gives significant production of Pu-238 as fertile material to absorp neutrons for reducing excess reactivity and additional contribution for fuel breeding.

  18. Sorption of radioactive contaminants by sediment from the Kara Sea

    SciTech Connect

    Fuhrmann, M.; Zhou, H.; Neiheisel, J.; Dyer, R.

    1995-02-01

    The purpose of this study is to quantify some of the parameters needed to perform near-field modeling of sites in the Kara Sea that were impacted by the disposal of radioactive waste. The parameters of interest are: the distribution coefficients (K{sub d}) for several important radionuclides, the mineralogy of the sediment, and the relationship of K{sub d} to liquid to solid ratio. Sediment from the Kara Sea (location: 73{degrees} 00` N, 58{degrees} 00` E) was sampled from a depth of 287 meters on August 23/24, 1992, during a joint Russian/Norwegian scientific cruise. Analysis of the material included mineralogy, grain size and total organic carbon. Uptake kinetics were determined for {sup 85}Sr, {sup 99}Tc, {sup 125}I, {sup 137}Cs, {sup 210}Pb, {sup 232}U, and {sup 241}Am and distribution coefficients (K{sub d}) were determined for these radionuclides using batch type experiments. Sorption isotherms were developed for {sup 85}Sr, {sup 99}Tc, and {sup 137}Cs to examine the effect that varying the concentration of a tracer has on the quantity of that tracer taken up by the solid. The effect of liquid to solid ratio on the uptake of contaminants was determined for {sup 99}Tc and {sup 137}Cs. In another set of experiments, the sediment was separated into four size fractions and uptake was determined for each fraction for {sup 85}Sr, {sup 99}Tc, and {sup 137}Cs. In addition, the sediment was analyzed to determine if it contains observable concentrations of anthropogenic radionuclides.

  19. A Gamma Radiolysis Study of UO{sub 2}F{sub 2} 0.4H{sub 2}O Using Spent Nuclear Fuel Elements from the High Flux Isotope Reactor

    SciTech Connect

    Icenhour, A.S.

    2002-01-24

    The development of a standard for the safe, long-term storage of {sup 233}U-containing materials resulted in the identification of several needed experimental studies. These studies were largely related to the potential for the generation of unacceptable pressures or the formation of deleterious products during storage of uranium oxides. The primary concern was that these conditions could occur as a result of the radiolysis of residual impurities--specifically fluorides and water-by the high radiation fields associated with {sup 233}U/{sup 232}U-containing materials. This report documents the results from a gamma radiolysis experiment in which UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O was loaded in helium. This experiment was performed using spent nuclear fuel elements from the High Flux Isotope Reactor as the gamma source and was a follow-on to experiments conducted previously. It was found that upon gamma irradiation, the UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O released 0{sub 2} with an initial G(O{sub 2}) = 0.01 molecule O{sub 2}/100 eV and that some of the uranium was reduced from U(VI) to U(IV). The high total dose achieved in the SNF elements was sufficient to reach a damage limit for the UO{sub 2}F{sub 2} {center_dot} 0.4H{sub 2}O. This damage limit, measured in terms of the amount of the U(IV) produced, was found to be about 9 wt%.

  20. ALARA Controls and the Radiological Lessons Learned During the Uranium Fuel Removal Projects at the Molten Salt Reactor Experiment

    SciTech Connect

    Gilliam, B. J.; Chapman, J. A.; Jugan, M. R.

    2002-02-26

    The removal of uranium-233 (233 U) from the auxiliary charcoal bed (ACB) of the Molten Salt Reactor Experiment (MSRE), performed from January through May 2001, created both unique radiological challenges and widely-applicable lessons learned. In addition to the criticality concerns and alpha contamination, 233U has an associated intense gamma photon from the cocontaminant uranium-232 (232U) decaying to thallium-208 (208Tl). Therefore, rigorous contamination controls and significant shielding were implemented. Extensive, timed mock-up training was also imperative to minimize individual and collective personnel exposures. Back-up shielding and containment techniques (that had been previously developed for defense in depth) were used successfully to control significant, changed conditions. Additional controls were placed on tests and on recovery designs to assure a higher level of safety throughout the removal operations. This paper delineates the manner in which each difficulty was solved, while relating the relevance of the results and the methodology to other projects with high dose-rate, highly-contaminated ionizing radiation hazards. Because of the distinctive features of and current interest in molten salt technology, a brief overview is provided. Also presented is the detailed, practical application of radiological controls integrated into, rather than added after, each evolution of the project--thus demonstrating the broad-based benefits of radiological engineering and ALARA reviews. The resolution of the serious contamination-control problems caused by unexpected uranium hexafluoride (UF6) gaseous diffusion is also explicated. Several tables and figures document the preparations, equipment and operations. A comparison of the pre-job dose calculations for the various functions of the uranium deposit removal (UDR) and the post-job dose-rate data are included in the conclusion.

  1. Observation of novel radioactive decay by spontaneous emission of complex nuclei

    SciTech Connect

    Barwick, S.W.

    1986-01-01

    Two years of experimental investigation on the subject of spontaneous emission of intermediate-mass fragments is described in this manuscript. A short introduction on this subject and a historical review are presented in chapter 1. In chapter 2, the author describe the experimental methods which led to the observation of /sup 14/C emission in polycarbonate etched-track detectors from the isotopes /sup 222/Ra, /sup 223/Ra, /sup 224/Ra and /sup 226/Ra at the branching ratios with respect to ..cap alpha..-decay of (3.7 +/- 0.6) x 10/sup -10/, (6.1 +/- 1.0) x 10/sup -10/, (4.3 +/- 1.2) x 10/sup -10/ and (2.9 +/- 1.0) x 10/sup -11/ respectively. Branching ratio limits for heavy-ion emission from /sup 221/Fr, /sup 221/Ra and /sup 225/Ac were determined to be at < 5.0 x 10/sup -14/, < 1.2 x 10/sup -13/ and < 4.0 x 10/sup -13/ respectively for the 90% C.L. The emission of /sup 24/Ne from /sup 232/U at a branching ratio of (2.0 +/- 0.5) x 10/sup -12/ has been discovered using polyethylene terephthalate etched-track plastics. A confirmation of /sup 24/Ne and/or /sup 25/Ne emission from /sup 233/U at a branching ratio of (5.3 +/- 2.3) x 10/sup -13/ is also reported. In chapter 3, three models of intermediate-mass decay are discussed-the analytic superasymmetric fission model, the model by Shi and Swiatecki, and a model based on a square-well + Coulomb potential.

  2. The Concentration of (236)Pu Daughters in Plutonium for Application to MOX Production from Plutonium from Dismantled US Nuclear Weapons

    SciTech Connect

    Sampson, T.E.; Cremers, T.L.

    2001-05-01

    The isotope {sup 236}Pu in the weapons-grade plutonium to be used in the US MOX (mixed-oxide) plant is of concern because the daughter products of {sup 236}Pu are sources of high-energy gamma rays. The {sup 208}Tl daughter of {sup 236}Pu emits intense, high-energy gamma rays that are important for radiation exposure calculations for plant design. It is generally thought that the concentrations of {sup 236}Pu and its daughters are well below 10{sup {minus}10}, but these concentrations are generally below the detection limits of most analytical techniques. One technique that can be used to determine the concentration {sup 208}Tl is the direct measurement of the intensity of the {sup 208}Tl gamma rays in the gamma-ray spectrum from plutonium. Thallium-208 will be in equilibrium with {sup 228}Th, and may very well be in equilibrium with {sup 232}U for most aged plutonium samples. We have used the FRAM isotopic analysis software to analyze dozens of archived high-resolution gamma ray spectra from various samples of US and foreign plutonium. We are able to quantify the ratio of minor isotopes with measurable gamma-ray emissions to the major isotope of plutonium and hence, through the measurement of the plutonium isotopic distribution of the sample, to elemental plutonium itself. Excluding items packaged in fluoropolymer vials, all samples analyzed with {sup 240}Pu < 9% gave {sup 228}Th/Pu ratios < 3.4 e-012 and all samples of US-produced plutonium, including {sup 240}Pu values up to 16.4%, gave {sup 228}Th/Pu ratios < 9.4 e-012. None of these values is significant from a radiation dose standpoint.

  3. ORNL production of the experimental alpha emitters bismuth-213 and actinium-225 for medical applications

    SciTech Connect

    Webb, O.F.; Krichinsky, A.M.; Yong, L.K.

    1998-09-01

    Due to their short range in tissue (50 to 80 {micro}m), alpha emitters are of considerable interest for certain radioimmunotherapy applications. These applications require the destruction of single cells or small clusters of cells. The radioisotope {sup 213}Bi, which is milked from {sup 225}Ac, is an alpha emitter that is currently being used in phase-1 human leukemia trials at the Memorial Sloan-Kettering Cancer Center. The most readily achievable route for producing {sup 225}Ac generators involves separating ingrown {sup 229}Th daughters from the {sup 233}U parent. Thorium-229 is then used as a parent generator for {sup 225}Ac. Thorium-229 is easier to handle than {sup 233}U, which is fissile and typically contains trace concentrations of {sup 232}U. Uranium-232 has a radioactive daughter, {sup 208}Tl, which emits a high-energy (2.6-MeV) photon when it decays. An alternative method for producing {sup 229}Th is through neutron irradiation of {sup 227}Ra. However, this method is less desirable due to the production of very high levels of {sup 228}Th. Thorium-229 accumulates in stored {sup 233}U oxides by natural decay. The current ORNL process for extracting {sup 229}Th from stored {sup 233}U oxides includes dissolution, strong-acid anion exchange, and calcination of the uranium. This ORNL process has provided high-purity {sup 225}Ac generators to medical researchers. Bismuth-213 has been extracted and used in initial human trials and already has demonstrated a potency and specificity for attacking cancerous cells.

  4. Why reconsider the thorium fuel cycle?

    SciTech Connect

    Krahn, S.; Croff, A.; Ault, T.; Wymer, R.

    2013-07-01

    In this paper we have endeavored to present the available technical information on the potential use of Th in nuclear fuel cycle (FC) applications as compared to U without subjective evaluations. Where helpful, we have compared the technical attributes of Th-232 as a fertile isotope and U-233 as a fissile isotope with other similar isotopes (i.e., U-238, and U-235 and Pu-239, respectively). In addition, we have summarized (a) experience gained to-date with fabricating and reprocessing of Th-232/U-233 fuels, (b) factors concerning Th fuel irradiation in both test reactors and power reactors, and (c) differences in the backend of the FC with emphasis on repository risks. As might be expected, many technical aspects of Th vs. U have not changed since the sixties. However, there are some factors elaborated in this paper that have changed. Changes potentially encouraging Th use are: (a) the ability to recover large amounts of Th as a byproduct with small attendant costs and environmental impacts, (b) the potential to produce fewer minor actinides (MA) and less Pu during power production, and (c) increased concerns about proliferation which might be somewhat mitigated by the high radioactivity and amenability to isotopic dilution of U-233. Changes challenging Th utilization are: (a) obtaining sufficient experience handling Th/U-233 fuels, (b) the existence of large inventories of depleted U and continuing discovery of large U resources, and (c) recognition that the extent to which U-233 might mitigate proliferation concerns is not as large as originally hoped.

  5. Quantitative determination of environmental levels of uranium, thorium and plutonium in bone by solvent extraction and alpha spectrometry

    NASA Astrophysics Data System (ADS)

    Singh, Narayani P.; Zimmerman, Carol J.; Lewis, Laura L.; Wrenn, McDonald E.

    1984-06-01

    Solvent extraction and alpha-spectrometry have been emplyed in the quantitative simultaneous determination of uranium. thorium and plutonium. The bone specimens, spiked with 232U, 229Th and 242Pu tracers, are wet ashed with HNO 3 followed by alternate additions of a new drops of HNO 3 and H 2O 2. Uranium is reduced to the tetravalent state with 200 mg SnCl 2 and 25 ml HI. Uranium, thorium and plutonium are then coprecipitated with calcium as oxalate, heated to 550°C, dissolved in 50 ml HCl, and the acidity adjusted to 10 M. Uranium and plutonium are extracted into a 20% tri-lauryl amine (TLA) solution in xylene, leaving thorium in the aqueous phase. Plutonium is first back-extracted from the TLA phase by shaking with a 1:1.5 volume of 0.05 M NH 4I in 8 M HCl, which reduces Pu(IV) to Pu(III). Uranium is then back-extracted with an equal volume of 0.1 M HCl. Thorium, which was left in the aqueous phase, is evaporated to dryness, dissolved in 4 M HNO 3, and the acidity adjusted to 4 M. Thorium is then extracted into 20% TLA solution in xylene pre-equilibrated with 4 M HNO 3, and back-extracted with 10 M HCl. Uranium, thorium, and plutonium are then electrodeposited separately onto platinum discs and counted by an alpha-spectrometer with a multi-channel analyzer and surface barrier silicon diodes. The mean recoveries of uranium, thorium, and plutonium in bovine, dog, and human bones were over 70%.

  6. Thorium, uranium and rare earth elements content in lanthanide concentrate (LC) and water leach purification (WLP) residue of Lynas advanced materials plant (LAMP)

    NASA Astrophysics Data System (ADS)

    AL-Areqi, Wadeeah M.; Majid, Amran Ab.; Sarmani, Sukiman

    2014-02-01

    Lynas Advanced Materials Plant (LAMP) has been licensed to produce the rare earths elements since early 2013 in Malaysia. LAMP processes lanthanide concentrate (LC) to extract rare earth elements and subsequently produce large volumes of water leach purification (WLP) residue containing naturally occurring radioactive material (NORM). This residue has been rising up the environmental issue because it was suspected to accumulate thorium with significant activity concentration and has been classified as radioactive residue. The aim of this study is to determine Th-232, U-238 and rare earth elements in lanthanide concentrate (LC) and water leach purification (WLP) residue collected from LAMP and to evaluate the potential radiological impacts of the WLP residue on the environment. Instrumental Neutron Activation Analysis and γ-spectrometry were used for determination of Th, U and rare earth elements concentrations. The results of this study found that the concentration of Th in LC was 1289.7 ± 129 ppm (5274.9 ± 527.6Bq/kg) whereas the Th and U concentrations in WLP were determined to be 1952.9±17.6 ppm (7987.4 ± 71.9 Bq/kg) and 17.2 ± 2.4 ppm respectively. The concentrations of Th and U in LC and WLP samples determined by γ- spectrometry were 1156 ppm (4728 ± 22 Bq/kg) & 18.8 ppm and 1763.2 ppm (7211.4 Bq/kg) &29.97 ppm respectively. This study showed that thorium concentrations were higher in WLP compare to LC. This study also indicate that WLP residue has high radioactivity of 232Th compared to Malaysian soil natural background (63 - 110 Bq/kg) and come under preview of Act 304 and regulations. In LC, the Ce and Nd concentrations determined by INAA were 13.2 ± 0.6% and 4.7 ± 0.1% respectively whereas the concentrations of La, Ce, Nd and Sm in WLP were 0.36 ± 0.04%, 1.6%, 0.22% and 0.06% respectively. This result showed that some amount of rare earth had not been extracted and remained in the WLP and may be considered to be reextracted.

  7. Determination the total neutron yields of several semiconductor compounds using various alpha emitters

    NASA Astrophysics Data System (ADS)

    Abdullah, Ramadhan Hayder; Sabr, Barzan Nehmat

    2016-03-01

    In the present work, the cross-sections of (α,n) reactions available in the literature as a function of α-particle energies for light and medium elements have been rearranged for α-particle energies from near threshold up to 10 MeV in steps of (0.050MeV) using the (Excel and Matlab) computer programs. The obtained data were used to calculate the neutron yields (n/106α) using the quick basic-computer program (Simpson Rules). The stopping powers of alpha particle energies from near threshold to 10 MeV for light and medium elements such as (nat.Be,10B,11B,13C,14N,nat.O,nat.F,nat.Mg,nat.Al,29Si,30Si, nat.P and 46.48Ti) have been calculated using the Zeigler formula. The kinetic energies (Tα) and the branching ratios of each α-emitters such as (211Bi, 210Po, 211Po, 215Po, 217At, 218Rn, 219Rn, 222Rn, 224Ra, 226Ra, 215Th, 228Th, 232U, 234U, 236U, 238U, 238Pu, 239Pu, 241Am, 245Es, 252Fm, 254Fm, 256Fm, 257Fm and 257Md) are taken into consideration to calculate the mean kinetic energy . The polynomial expressions were used to fitting the calculated weighted average of neutron yields (n/106α) for natural light and medium elements such as (Be, B, C, N, O, F, Mg, Al, Si, P and Ti) to determine the adopted neutron yields from the best fitting equation with minimum (CHISQ) at mean kinetic energies of various α-emitters. The total neutron yields (n/s/gx/ppmi) of the mentioned natural light and medium elements have been calculated using the adopted neutron yields (n/106α) from the fitting equations at mean kinetic energies of various α-emitters. The total neutron yields (n/s/gα-emitters/gcompounds) of semiconductor compounds such as (AlN, AlP, BN, BP, SiC, TiO2, BeSiN2, MgCN2, MgSiN2 and MgSiP2) have been calculated by mixing (1gram) of compounds with (1gram) of pure α-emitters using the quick basic computer program. The aim of the present work is to constructed and fabricate the neutron sources theoretically

  8. Thorium, uranium and rare earth elements content in lanthanide concentrate (LC) and water leach purification (WLP) residue of Lynas advanced materials plant (LAMP)

    SciTech Connect

    AL-Areqi, Wadeeah M. Majid, Amran Ab. Sarmani, Sukiman

    2014-02-12

    Lynas Advanced Materials Plant (LAMP) has been licensed to produce the rare earths elements since early 2013 in Malaysia. LAMP processes lanthanide concentrate (LC) to extract rare earth elements and subsequently produce large volumes of water leach purification (WLP) residue containing naturally occurring radioactive material (NORM). This residue has been rising up the environmental issue because it was suspected to accumulate thorium with significant activity concentration and has been classified as radioactive residue. The aim of this study is to determine Th-232, U-238 and rare earth elements in lanthanide concentrate (LC) and water leach purification (WLP) residue collected from LAMP and to evaluate the potential radiological impacts of the WLP residue on the environment. Instrumental Neutron Activation Analysis and γ-spectrometry were used for determination of Th, U and rare earth elements concentrations. The results of this study found that the concentration of Th in LC was 1289.7 ± 129 ppm (5274.9 ± 527.6Bq/kg) whereas the Th and U concentrations in WLP were determined to be 1952.9±17.6 ppm (7987.4 ± 71.9 Bq/kg) and 17.2 ± 2.4 ppm respectively. The concentrations of Th and U in LC and WLP samples determined by γ- spectrometry were 1156 ppm (4728 ± 22 Bq/kg) and 18.8 ppm and 1763.2 ppm (7211.4 Bq/kg) and 29.97 ppm respectively. This study showed that thorium concentrations were higher in WLP compare to LC. This study also indicate that WLP residue has high radioactivity of {sup 232}Th compared to Malaysian soil natural background (63 - 110 Bq/kg) and come under preview of Act 304 and regulations. In LC, the Ce and Nd concentrations determined by INAA were 13.2 ± 0.6% and 4.7 ± 0.1% respectively whereas the concentrations of La, Ce, Nd and Sm in WLP were 0.36 ± 0.04%, 1.6%, 0.22% and 0.06% respectively. This result showed that some amount of rare earth had not been extracted and remained in the WLP and may be considered to be reextracted.

  9. Fe-Mn substance in ocean as reason of regulation radionuclide pollution

    NASA Astrophysics Data System (ADS)

    Asavin, Alex; Martynov, Konstantin; Konstantinova, Lia

    2013-04-01

    Distribution of radionuclide in marine sediments as yet little studied [Choppin & Wong 1998]. The work mainly focused on effects of nuclear test fallout. In the works are examined isotopes of Pu - 238; Th - 232; U -234;238; Pu - 239,240,241; Am - 241; Np - 237; Cm -244 [Holm 1995]. It has been shown that seems to accumulate radionuclides in marine sediments. In particular, the importance attached to carbonate complexes (corals, etc.). But questions about the possibility of re-mobilization of radionuclide, forms their concentration, their participation in global geochemical cycles in the ocean, remain open. We believe a major factor controlling the distribution of heavy metals is the formation of ocean ferromanganese crusts and nodules hydrogenic at the bottom of the ocean and seamounts. It is likely that the process of formation of Fe-manganese hydrogenic can play a major role in the control of radioactive contamination in the oceanic sediment. At least for the U number of works on the subject [Sherman et al. 2008]. The high sensitivity of the Fe-manganese crust is known to the isotopic composition of lead [Loranger & Zayed 1994, Collen et al 2011]. Recent work [Wilkins etal 2006, Renshaw etal 2009] show a large role; Fe (III)-and Mn (IV)-reducing organisms that anaerobic bacteria in oxidation and therefore changes in mobility systems U and Pu. So much interest is data for sorption of radionuclide on hydroxides Fe and Mn. Unfortunately we are not aware of works on the subject. We have therefore taken their own experimental studies on sorption of radionuclide on natural Fe-Mn crusts (sample from Magellan seamount Pacific ocean) [Martynov et al 2012]. The results showed high sorption ability of material crusts for fixation of radionuclides: U-233, Np-237, Pu-238, Am-241. For all radionuclide experiment absorption has been reached already in the first hour it was 96.0% of total substance radionuclide absorbed from the solution, and after the first day it was reached

  10. Interpretation of actinide-distribution data obtained from non-destructive and destructive post-test analyses of an intact-core column of Culebra dolomite.

    PubMed

    Perkins, W G; Lucero, D A

    2001-02-01

    The US Department of Energy (DOE), with technical assistance from Sandia National Laboratories, has successfully received EPA certification and opened the Waste Isolation Pilot Plant (WIPP), a nuclear waste disposal facility located approximately 42 km east of Carlsbad, NM. Performance assessment (PA) analyses indicate that human intrusions by inadvertent, intermittent drilling for resources provide the only credible mechanisms for significant releases of radionuclides from the disposal system. For long-term brine releases, migration pathways through the permeable layers of rock above the Salado formation are important. Major emphasis is placed on the Culebra Member of the Rustler Formation because this is the most transmissive geologic layer overlying the WIPP site. In order to help quantify parameters for the calculated releases, radionuclide transport experiments have been carried out using intact-core columns obtained from the Culebra dolomite member of the Rustler Formation within the WIPP site. This paper deals primarily with results of analyses for 241Pu and 241Am distributions developed during transport experiments in one of these cores. Transport experiments were done using a synthetic brine that simulates Culebra brine at the core recovery location (the WIPP air-intake shaft (AIS)). Hydraulic characteristics (i.e., apparent porosity and apparent dispersion coefficient) for intact-core columns were obtained via experiments using the conservative tracer 22Na. Elution experiments carried out over periods of a few days with tracers 232U and 239Np indicated that these tracers were weakly retarded as indicated by delayed elution of the species. Elution experiments with tracers 241Pu and 241Am were attempted but no elution of either species has been observed to date, including experiments of many months' duration. In order to quantify retardation of the non-eluted species 241Pu and 241Am after a period of brine flow, non-destructive and destructive analyses of

  11. Interpretation of Actinide-Distribution Data Obtained from Non-Destructive and Destructive Post-Test Analyses of an Intact-Core Column of Culebra Dolomite

    SciTech Connect

    LUCERO, DANIEL A; PERKINS, W GEORGE

    1999-08-26

    The US DOE, with technical assistance from Sandia National Laboratories, has successfully received EPA certification and opened the Waste Isolation Pilot Plant (WIPP), a nuclear waste disposal facility located approximately 42 km east of Carlsbad, New Mexico. Performance assessment analyses indicate that human intrusions by inadvertent, intermittent drilling for resources provide the only credible mechanisms for releases of radionuclides from the disposal system. In modeling long-term brine releases, subsequent to a drilling event, potential migration pathways through the permeable layers of rock above the Salado formation were analyzed. Major emphasis is placed on the Culebra Member of the Rustler Formation because this is the most transmissive geologic layer overlying the WIPP site. In order to help quantify parameters for the calculated releases, radionuclide transport experiments have been earned out using intact-core columns obtained from the Culebra dolomite member of the Rustler Formation within the WIPP site. This paper deals primarily with results of analyses for {sup 241}Pu and {sup 241}Am distributions developed during transport experiments in one of these cores. Transport experiments were done using a synthetic brine that simulates Culebra brine at the core recovery location (the WIPP air-intake shaft--AIS). Hydraulic characteristics (i.e., apparent porosity and apparent dispersion coefficient) for intact-core columns were obtained via experiments using the conservative tracer {sup 22}Na. Elution experiments carried out over periods of a few days with tracers {sup 232}U and {sup 239}Np indicated that these tracers were weakly retarded as indicated by delayed elution of the species. Elution experiments with tracers {sup 241}Pu and {sup 241}Am were attempted, but no elution of either species has been observed to date, including experiments of many months' duration. In order to quantify retardation of the non-eluted species {sup 241}Pu and {sup 241}Am

  12. Supplemental Report on Nuclear Safeguards Considerations for the Pebble Bed Modular Reactor (PBMR)

    SciTech Connect

    Moses, David Lewis; Ehinger, Michael H

    2010-05-01

    Recent reports by Department of Energy National Laboratories have discussed safeguards considerations for the low enriched uranium (LEU) fueled Pebble Bed Modular Reactor (PBMR) and the need for bulk accountancy of the plutonium in used fuel. These reports fail to account effectively for the degree of plutonium dilution in the graphitized-carbon pebbles that is sufficient to meet the International Atomic Energy Agency's (IAEA's) 'provisional' guidelines for termination of safeguards on 'measured discards.' The thrust of this finding is not to terminate safeguards but to limit the need for specific accountancy of plutonium in stored used fuel. While the residual uranium in the used fuel may not be judged sufficiently diluted to meet the IAEA provisional guidelines for termination of safeguards, the estimated quantities of {sup 232}U and {sup 236}U in the used fuel at the target burn-up of {approx}91 GWD/MT exceed specification limits for reprocessed uranium (ASTM C787) and will require extensive blending with either natural uranium or uranium enrichment tails to dilute the {sup 236}U content to fall within specification thus making the PBMR used fuel less desirable for commercial reprocessing and reuse than that from light water reactors. Also the PBMR specific activity of reprocessed uranium isotopic mixture and its A{sub 2} values for effective dose limit if released in a dispersible form during a transportation accident are more limiting than the equivalent values for light water reactor spent fuel at 55 GWD/MT without accounting for the presence of the principal carry-over fission product ({sup 99}Tc) and any possible plutonium contamination that may be present from attempted covert reprocessing. Thus, the potentially recoverable uranium from PBMR used fuel carries reactivity penalties and radiological penalties likely greater than those for reprocessed uranium from light water reactors. These factors impact the economics of reprocessing, but a more significant

  13. Rapid analytical technique to identify alpha emitting isotopes in water, air-filters, urine, and solid matrices using a Frisch Grid detector.

    PubMed

    Scarpitta, Salvatore C; Miltenberger, Robert P; Gaschott, Robert; Carte, Nina

    2003-04-01

    A 5-inch-diameter Frisch Grid gas-proportional ionization chamber was utilized at Brookhaven National Laboratory (BNL) to rapidly characterize and quantify alpha-emitting actinides in unprocessed water, soil, air-filter, urine, and solid matrices. Instrument calibrations for the various matrices were performed by spiking representative samples with National Institute of Standards and Technology traceable isotopes of 230Th, 232U, 236Pu, and 243Am. Detection efficiencies were typically 15-20% for solid matrices (soil, concrete, filters, dry urine) and 45% for mass-less water samples. Instrument background over a 512-channel alpha-energy range of 3-8 MeV is very low at 0.01 cps. At optimum efficiency, minimum detectable levels of 0.56 mBq Kg(-1), 74 mBq L(-1) and 14.8 mBq filter(-1) were achievable for 40 x 10(-6) Kg soil, 1 x 10(-3) L tap water (or urine), and 4.5 cm diameter air-filter samples, respectively, each counted for 60 min. Data and spectra are presented showing the quality of results obtained using untreated samples obtained from the BNL Graphite Research Reactor Decommissioning Project. These samples contained Bq to MBq per gram amounts of (239,240)Pu, 241Am, and/or (234,235/238)U (as well as other beta/gamma emitters). Data and spectra are also presented for a very finely pulverized and homogeneous U.S. DOE/RESL soil reference standard (spiked with 239Pu, 241Am, and 233U) that was used to assess precision, accuracy, and reproducibility. Although this technique has its limitations, the advantages are (1) minimal sample preparation, (2) no separation chemistry required, (3) no chemical or hazardous waste generated, and (4) ability to immediately characterize and quantify alpha-emitting nuclides in most matrices. The benefits of this technique to the BNL/DOE Project Managers were rapid (1-2 d) turn-around times coupled with significant cost savings, as compared to commercial off-site analyses. PMID:12705448

  14. Critical Masses for Unreflected Metal Spheres

    SciTech Connect

    Westfall, Robert Michael; Wright, Richard Q

    2009-01-01

    Calculated critical masses of bare metal spheres for 28 actinide isotopes, using the SCALE/XSDRNPM one-dimensional, discrete-ordinates system, are presented. ENDF/B-VI, ENDF/B-VII, and JENDL-3.3 cross sections were used in the calculations. Results are given for isotopes of uranium, neptunium, plutonium, americium, curium, californium, and for one isotope of einsteinium. Calculated k values for these same nuclides are also given. We show that, for non-threshold or low-threshold fission nuclides, a good approximation for the nuclide k is the value of nubar at 1 MeV. A plot of the critical mass versus k values is given for 19 nuclides with A-numbers between 232 and 250. The peaks in the critical mass curve (for seven nuclides) correspond to dips in the k curve. For the seven cases with the largest critical mass, six are even-even nuclides. Neptunium-237, with a critical mass of about 62.7 kg (ENDF/B-VI calculation), has an odd number of protons and an even number of neutrons. However, two cases with quite small critical masses, 232U and 236Pu, are also even-even. These two nuclides do not exhibit threshold fission behavior like most other even-even nuclides. The largest critical mass is 208.8 kg for 243Am and the smallest is 2.44 kg for 251Cf. The calculated k values vary from 1.5022 for 234U to 4.4767 for 251Cf. A correlation between the calculated critical mass (kg) and the fission spectrum averaged value of is given for the elements U, Np, Pu, Am, Cm, and Cf. For each of the five elements, a fit to the data for that element is provided. In each case the fit employs a negative exponential of the form mass = exp(A + B ~ ln( ) The values of A and B are element dependent and vary slightly for each of the five elements. The method described here is mainly applicable for non-threshold fission nuclides (15 of the 28 nuclides considered in this paper). There are three exceptions, 238Pu, 244Cm, and 250Cf, which all exhibit threshold fission behavior.

  15. Hybrid Enrichment Assay Methods for a UF6 Cylinder Verification Station: FY10 Progress Report

    SciTech Connect

    Smith, Leon E.; Jordan, David V.; Orton, Christopher R.; Misner, Alex C.; Mace, Emily K.

    2010-08-01

    HPGe verification station at AREVA, and the IAEA’s uncertainty target values for feed, tail and product cylinders. A summary of the major findings from the field measurements and subsequent analysis follows: • Traditional enrichment-meter assay using specially collimated NaI spectrometers and a Square-Wave-Convolute algorithm can achieve uncertainties comparable to HPGe and LaBr for product, natural and depleted cylinders. • Non-traditional signatures measured using NaI spectrometers enable interrogation of the entire cylinder volume and accurate measurement of absolute 235U mass in product, natural and depleted cylinders. • A hybrid enrichment assay method can achieve lower uncertainties than either the traditional or non-traditional methods acting independently because there is a low degree of correlation in the systematic errors of the two individual methods (wall thickness variation and 234U/235U variation, respectively). This work has indicated that the hybrid NDA method has the potential to serve as the foundation for an unattended cylinder verification station. When compared to today’s handheld cylinder-verification approach, such a station would have the following advantages: 1) improved enrichment assay accuracy for product, tail and feed cylinders; 2) full-volume assay of absolute 235U mass; 3) assay of minor isotopes (234U and 232U) important to verification of feedstock origin; single instrumentation design for both Type 30B and Type 48 cylinders; and 4) substantial reduction in the inspector manpower associated with cylinder verification.

  16. Muscle Preactivity of Anterior Cruciate Ligament-Deficient and -Reconstructed Females During Functional Activities

    PubMed Central

    DeMont, Richard G.; Lephart, Scott M.; Giraldo, Jorge L.; Swanik, C. Buz; Fu, Freddie H.

    1999-01-01

    Objective: Underlying the ability of the hamstrings to decrease tibial anterior shear is the time of firing in comparison with the quadriceps. This timing may be aided by neural programming during a planned or expected activity. It is theorized that individuals who have better programming ability will suffer fewer anterior cruciate ligament (ACL) injuries due to joint protection through muscular stabilization. A component of this dynamic restraint is the development of muscular tension before the knee is loaded. The objective of our study was to compare the muscular activity before footstrike in ACL-deficient (ACL-D), ACL-reconstructed (ACL-R), and control (C) females during functional activities. Design and Setting: Active females were divided into groups based on their ACL status. The study was conducted in a neuromuscular research laboratory. Subjects: Twenty-four female subjects (ACL-D = 6, ACL-R = 12, C = 6). Measurements: Integrated electromyographic (IEMG) activity from the thigh (vastus medialis obliquus [VMO], vastus lateralis [VL], medial hamstring, and lateral hamstring) and leg (medial gastrocnemius and lateral gastrocnemius [LG]) and footswitch signals were recorded during downhill walking (15° at 0.92 m/s), running (2.08 m/s), hopping, and landing from a step (20.3 cm). IEMG activity was normalized to the mean amplitude of the sample and analyzed for area and mean amplitude for 150 milliseconds before heelstrike. Side-to-side differences were determined by t tests, and separate one-way analyses of variance (ANOVA) were used to detect differences among the 3 groups for each muscle of each activity. Results: IEMG area side-to-side differences for the ACL-D group appeared in the LG (involved [I] = 36.4 ± 19.7, uninvolved [U] = 60.1 ± 23.6) during landing, in the VMO (I = 11.4 ± 3.8, U = 7.2 ± 3.1) and VL (I = 13.3 ± 2.7, U = 8.9 ± 1.9) during running, and in the VMO (I = 9.2 ± 4.2, U = 19.5 ± 7.3) during downhill walking. IEMG mean amplitude side-to-side differences for the ACL-D group appeared in the LG (I = 79.7 ± 30.3, U = 122.3 ± 34.9) during downhill walking and in the VMO (I = 78.6 ± 23.2, U = 45.8 ± 18.9) during the run; IEMG mean amplitude side-to-side differences for the ACL-R group appeared in the LG (I = 74.7 ± 40.0, U = 52.8 ± 14.3) during the hop. The ACL-D group had higher IEMG means than control in the VL (ACL-D = 12.9 ± 5.8, C = 7.1 ± 3.9), but lower in the VMO (ACL-D = 9.2 ± 4.2, C = 15.7 ± 3.6). Conclusions: The side-to-side differences of the ACL-D and ACL-R groups, as well as the group differences between ACL-D and control, suggest that different muscle activation strategies are used by females when performing different dynamic activities. Therefore, muscle unit differentiation may be the cause of our results. These changes appear to be reversed through surgery or the associated postoperative rehabilitation. ImagesFigure 2.Figure 3.Figure 4. PMID:16558553

  17. Nuclear car wash status report, August 2005

    SciTech Connect

    Prussin, S; Slaughter, D; Pruet, J; Descalle, M; Bernstein, A; Hall, J; Accatino, M; Alford, O; Asztalos, S; Church, J; Loshak, A; Madden, N; Manatt, D; Moore, T; Norman, E; Petersen, D

    2005-07-29

    A large majority of US imports arrive at seaports in maritime cargo containers. The number of containers arriving is nearly 10 million per year, each with a cargo of up to 30 tons of various materials. This provides a vulnerable entry point for the importation of a nuclear weapon or its components by a terrorist group. Passive radiation sensors are being deployed at portals to detect radioactive material and portable instruments are carried by port personnel to augment detection. Those instruments can detect the neutrons and g-rays produced by {sup 240}Pu that is normally present in weapons grade plutonium in cases where cargo overburden is not too great. However, {sup 235}U produces almost no neutron output in its normal radioactive decay and its principal {gamma}-radiation is at 186 keV and is readily attenuated by small amounts of wood or packing materials. Impurities such as {sup 232}U, often present in reactor irradiated material at the 100-200 ppt level, can provide a detectable signal through significant cargo overburden but the wide variations among samples of HEU make this an unreliable means of detecting SNM. High quality radiography may be useful in determining that the majority of containers are clearly free of SNM. However, some containers will lead to ambiguous results from radiography and passive radiation sensing. For these reasons active neutron interrogation is proposed as a means to produce fission and thus greatly amplify the radiation output of fissionable material to facilitate its reliable detection even when well shielded by large cargo overburden. Historically, the fission signature utilized as the unique identifying feature of fissionable materials is the detection of delayed neutrons. However, these neutrons have very low yield {approx} 0.017 per fission in {sup 235}U, and their low energy results in very poor penetration of hydrogenous materials such as fuels, water, wood, or agricultural products. That signature alone does not provide

  18. Performance of Thorium-Based Mixed Oxide Fuels for the Consumption of Plutonium in Current and Advanced Reactors

    SciTech Connect

    Weaver, Kevan Dean; Herring, James Stephen

    2003-07-01

    -based fuels to achieve these burnups. Furthermore, thorium-based fuels could also be used as a strategy for reducing the amount of long-lived nuclides (including the minor actinides) and thus the radiotoxicity in spent nuclear fuel. Although the breeding of 233U is a concern, the presence of 232U and its daughter products (namely 208Tl) can aid in making this fuel self-protecting, and/or enough 238U can be added to denature the fissile uranium. From these calculations, it appears that thorium-based fuel for plutonium incineration is superior when compared to uranium-based fuel and should be considered as an alternative to traditional MOX in both current and future/advanced reactor designs.

  19. U(VI) behaviour in hyperalkaline calcite systems

    NASA Astrophysics Data System (ADS)

    Smith, Kurt F.; Bryan, Nicholas D.; Swinburne, Adam N.; Bots, Pieter; Shaw, Samuel; Natrajan, Louise S.; Mosselmans, J. Frederick W.; Livens, Francis R.; Morris, Katherine

    2015-01-01

    The behaviour of U(VI) in hyperalkaline fluid/calcite systems was studied over a range of U(VI) concentrations (5.27 × 10-5 μM to 42.0 μM) and in two high pH systems, young and old synthetic cement leachate in batch sorption experiments. These systems were selected to be representative of young- (pH 13.3) and old-stage (pH 10.5) leachate evolution within a cementitious geological disposal facility. Batch sorption experiments, modelling, extended X-ray absorption fine structure spectroscopy, electron microscopy, small angle X-ray scattering and luminescence spectroscopy were used to define the speciation of U(VI) across the systems of study. At the lowest concentrations (5.27 × 10-5 μM 232U(VI)) significant U removal was observed for both old and young cement leachates, and this was successfully modelled using a first order kinetic adsorption modelling approach. At higher concentrations (>4.20 μM) in the young cement leachate, U(VI) showed no interaction with the calcite surface over an 18 month period. Small angle X-ray scattering techniques indicated that at high U concentrations (42.0 μM) and after 18 months, the U(VI) was present in a colloidal form which had little interaction with the calcite surface and consisted of both primary and aggregated particles with a radius of 7.6 ± 1.1 and 217 ± 24 Å, respectively. In the old cement leachate, luminescence spectroscopy identified two surface binding sites for U(VI) on calcite: in the system with 0.21 μM U(VI), a liebigite-like Ca2UO2(CO3)3 surface complex was identified; at higher U(VI) concentrations (0.42 μM), a second binding site of undetermined coordination was identified. At elevated U(VI) concentrations (>2.10 μM) in old cement leachate, both geochemical data and luminescence spectroscopy suggested that surface mediated precipitation was controlling U(VI) behaviour. A focused ion beam mill was used to create a section across the U(VI) precipitate-calcite interface. Transmission electron

  20. Hydrofluoric Acid Corrosion Study of High-Alloy Materials

    SciTech Connect

    Osborne, P.E.

    2002-09-11

    A corrosion study involving high-alloy materials and concentrated hydrofluoric acid (HF) was conducted in support of the Molten Salt Reactor Experiment Conversion Project (CP). The purpose of the test was to obtain a greater understanding of the corrosion rates of materials of construction currently used in the CP vs those of proposed replacement parts. Results of the study will help formulate a change-out schedule for CP parts. The CP will convert slightly less than 40 kg of {sup 233}U from a gas (UF{sub 6}) sorbed on sodium fluoride pellets to a more stable oxide (U{sub 3}O{sub 8}). One by-product of the conversion is the formation of concentrated HF. Six moles of highly corrosive HF are produced for each mole of UF{sub 6} converted. This acid is particularly corrosive to most metals, elastomers, and silica-containing materials. A common impurity found in {sup 233}U is {sup 232}U. This impurity isotope has several daughters that make the handling of the {sup 233}U difficult. Traps of {sup 233}U may have radiation fields of up to 400 R at contact, a situation that makes the process of changing valves or working on the CP more challenging. It is also for this reason that a comprehensive part change-out schedule must be established. Laboratory experiments involving the repeated transfer of HF through 1/2-in. metal tubing and valves have proven difficult due to the corrosivity of the HF upon contact with all wetted parts. Each batch of HF is approximately 1.5 L of 33 wt% HF and is transferred most often as a vapor under vacuum and at temperatures of up to 250 C. Materials used in the HF side of the CP include Hastelloy C-276 and Monel 400 tubing, Haynes 230 and alloy C-276 vessels, and alloy 400 valve bodies with Inconel (alloy 600) bellows. The chemical compositions of the metals discussed in this report are displayed in Table 1. Of particular concern are the almost 30 vendor-supplied UG valves that have the potential for exposure to HF. These valves have been

  1. Special Analysis for the Disposal of the Consolidated Edison Uranium Solidification Project Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    SciTech Connect

    NSTec Environmental Management

    2013-01-31

    The purpose of this Special Analysis (SA) is to determine if the Oak Ridge (OR) Consolidated Edison Uranium Solidification Project (CEUSP) uranium-233 (233U) waste stream (DRTK000000050, Revision 0) is acceptable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). The CEUSP 233U waste stream requires a special analysis because the concentrations of thorium-229 (229Th), 230Th, 232U, 233U, and 234U exceeded their NNSS Waste Acceptance Criteria action levels. The acceptability of the waste stream is evaluated by determining if performance assessment (PA) modeling provides a reasonable expectation that SLB disposal is protective of human health and the environment. The CEUSP 233U waste stream is a long-lived waste with unique radiological hazards. The SA evaluates the long-term acceptability of the CEUSP 233U waste stream for near-surface disposal as a two tier process. The first tier, which is the usual SA process, uses the approved probabilistic PA model to determine if there is a reasonable expectation that disposal of the CEUSP 233U waste stream can meet the performance objectives of U.S. Department of Energy Manual DOE M 435.1-1, “Radioactive Waste Management,” for a period of 1,000 years (y) after closure. The second tier addresses the acceptability of the OR CEUSP 233U waste stream for near-surface disposal by evaluating long-term site stability and security, by performing extended (i.e., 10,000 and 60,000 y) modeling analyses, and by evaluating the effect of containers and the depth of burial on performance. Tier I results indicate that there is a reasonable expectation of compliance with all performance objectives if the OR CEUSP 233U waste stream is disposed in the Area 5 RWMS SLB disposal units. The maximum mean and 95th percentile PA results are all less than the performance objective for 1,000 y. Monte Carlo uncertainty analysis indicates that there is a high likelihood of

  2. Studies of flerovium and element 115 homologs with macrocyclic extractants

    NASA Astrophysics Data System (ADS)

    Despotopulos, John Dustin

    Study of the chemistry of the heaviest elements, Z ? 104, poses a unique challenge due to their low production cross-sections and short half-lives. Chemistry also must be studied on the one-atom-at-a-time scale, requiring automated, fast, and very efficient chemical schemes. Recent studies of the chemical behavior of copernicium (Cn, element 112) and flerovium (Fl, element 114) together with the discovery of isotopes of these elements with half-lives suitable for chemical studies have spurred a renewed interest in the development of rapid systems designed to study the chemical properties of elements with Z ≥ 114. This dissertation explores both extraction chromatography and solvent extraction as methods for development of a rapid chemical separation scheme for the homologs of flerovium (Pb, Sn, Hg) and element 115 (Bi, Sb), with the goal of developing a chemical scheme that, in the future, can be applied to on-line chemistry of both Fl and element 115. Macrocyclic extractants, specifically crown ethers and their derivatives, were chosen for these studies. Carrier-free radionuclides, used in these studies, of the homologs of Fl and element 115 were obtained by proton activation of high purity metal foils at the Lawrence Livermore National Laboratory (LLNL) Center for Accelerator Mass Spectrometry (CAMS): natIn(p,n)113Sn, natSn(p,n)124Sb, and Au(p,n)197m,gHg. The carrier-free activity was separated from the foils by novel separation schemes based on ion exchange and extraction chromatography techniques. Carrier-free Pb and Bi isotopes were obtained from development of a novel generator based on cation exchange chromatography using the 232U parent to generate 212Pb and 212Bi. Crown ethers show high selectivity for metal ions based on their size compared to the negatively charged cavity of the ether. Extraction by crown ethers occur based on electrostatic ion-dipole interactions between the negatively charged ring atoms (oxygen, sulfur, etc.) and the positively

  3. Studies of flerovium and element 115 homologs with macrocyclic extractants

    NASA Astrophysics Data System (ADS)

    Despotopulos, John Dustin

    Study of the chemistry of the heaviest elements, Z ? 104, poses a unique challenge due to their low production cross-sections and short half-lives. Chemistry also must be studied on the one-atom-at-a-time scale, requiring automated, fast, and very efficient chemical schemes. Recent studies of the chemical behavior of copernicium (Cn, element 112) and flerovium (Fl, element 114) together with the discovery of isotopes of these elements with half-lives suitable for chemical studies have spurred a renewed interest in the development of rapid systems designed to study the chemical properties of elements with Z ≥ 114. This dissertation explores both extraction chromatography and solvent extraction as methods for development of a rapid chemical separation scheme for the homologs of flerovium (Pb, Sn, Hg) and element 115 (Bi, Sb), with the goal of developing a chemical scheme that, in the future, can be applied to on-line chemistry of both Fl and element 115. Macrocyclic extractants, specifically crown ethers and their derivatives, were chosen for these studies. Carrier-free radionuclides, used in these studies, of the homologs of Fl and element 115 were obtained by proton activation of high purity metal foils at the Lawrence Livermore National Laboratory (LLNL) Center for Accelerator Mass Spectrometry (CAMS): natIn(p,n)113Sn, natSn(p,n)124Sb, and Au(p,n)197m,gHg. The carrier-free activity was separated from the foils by novel separation schemes based on ion exchange and extraction chromatography techniques. Carrier-free Pb and Bi isotopes were obtained from development of a novel generator based on cation exchange chromatography using the 232U parent to generate 212Pb and 212Bi. Crown ethers show high selectivity for metal ions based on their size compared to the negatively charged cavity of the ether. Extraction by crown ethers occur based on electrostatic ion-dipole interactions between the negatively charged ring atoms (oxygen, sulfur, etc.) and the positively