WorldWideScience

Sample records for bw standard reactor

  1. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  2. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Science.gov (United States)

    2013-12-09

    ... COMMISSION Operator Licensing Examination Standards for Power Reactors AGENCY: Nuclear Regulatory Commission... available for public comment a draft NUREG, NUREG-1021, Revision 10, ``Operator Licensing Examination Standards for Power Reactors.'' DATES: Submit comments by February 7, 2014. Comments received after...

  3. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  4. A proposed standard on medical isotope production in fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schenter, R. E. [Smart Bullets Inc., 2521 SW Luradel Street, Portland, OR 97219 (United States); Brown, G. J. [Ozarks Medical Center, Cancer Treatment Center, Shaw Medical Building, 1111 Kentucky Avenue, West Plains, MO 65775 (United States); Holden, C. S. [Thorenco LLC, 369 Pine Street, San Francisco, CA 94104 (United States)

    2006-07-01

    Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

  5. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  6. Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706(0)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel's service life (Fig. 1). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6 . More detailed writers' and users' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced documents, Section 2 and references (1-106) for use by individual writers and users. 1...

  7. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  8. Bgb expression in relation to the HLA-B17 antigen splits Bw57 and Bw58 and the cross-reactions of anti-Bgb antibodies.

    Science.gov (United States)

    Pollack, M S; Crawford, M N; Robinson, H M; Berger, R; Sabo, B; O'Neill, G J

    1982-07-01

    A substantial number of individuals typed as HLA-B17 do not have Bgb. To determine whether or not this discrepancy reflects genetic or ethnic group differences or the influence of B17 antigen subtypes, a large number of unrelated and related B17 individuals from the Caucasian, Hispanic, Black and Chinese ethnic groups were tested in parallel for Bgb and Bw57- or Bw58- B17 subtype. Higher Bgb expression was found in association with Bw57, but differences in expression of Bgb were also seen in different ethnic groups and in different related individuals carrying the same Bw57 or Bw58 haplotype. Genetic factors other than HLA thus influence the expression of HLa Antigens on red cells. Standard and antiglobulin lymphocytotoxicity tests indicate that all Bgb typing sera contain anti-Bw57 antilymphocyte antibodies and most also contain anti-Bw58 and cross-reacting anti-B12, B15 and/or Bw49 antibodies.

  9. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  10. Compiled reports on the applicability of selected codes and standards to advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, E.L.; Hoopingarner, K.R.; Markowski, F.J.; Mitts, T.M.; Nickolaus, J.R.; Vo, T.V.

    1994-08-01

    The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC.

  11. BW Trained HMM based Aerial Image Segmentation

    Directory of Open Access Journals (Sweden)

    R Rajasree

    2011-03-01

    Full Text Available Image segmentation is an essential preprocessing tread in a complicated and composite image dealing out algorithm. In segmenting arial image the expenditure of misclassification could depend on the factual group of pupils. In this paper, aggravated by modern advances in contraption erudition conjecture, I introduce a modus operandi to make light of the misclassification expenditure with class-dependent expenditure. The procedure assumes the hidden Markov model (HMM which has been popularly used for image segmentation in recent years. We represent all feasible HMM based segmenters (or classifiers as a set of points in the beneficiary operating characteristic (ROC space. optimizing HMM parameters is still an important and challenging work in automatic image segmentation research area. Usually the Baum-Welch (B-W Algorithm is used to calculate the HMM model parameters. However, the B-W algorithm uses an initial random guess of the parameters, therefore after convergence the output tends to be close to this initial value of the algorithm, which is not necessarily the global optimum of the model parameters. In this project, a Adaptive Baum-Welch (GA-BW is proposed.

  12. High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

    Energy Technology Data Exchange (ETDEWEB)

    McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.; Swearingen, Gary L.; Pugh, Ray

    2011-10-31

    Current interest expressed by industry in HTGR plants, particularly modular plants with power up to about 600 MW(e) per unit, has prompted NRC to task PNNL with assessing the currently available literature related to codes and standards applicable to HTGR plants, the operating history of past and present HTGR plants, and with evaluating the proposed designs of RPV and associated piping for future plants. Considering these topics in the order they are arranged in the text, first the operational histories of five shut-down and two currently operating HTGR plants are reviewed, leading the authors to conclude that while small, simple prototype HTGR plants operated reliably, some of the larger plants, particularly Fort St. Vrain, had poor availability. Safety and radiological performance of these plants has been considerably better than LWR plants. Petroleum processing plants provide some applicable experience with materials similar to those proposed for HTGR piping and vessels. At least one currently operating plant - HTR-10 - has performed and documented a leak before break analysis that appears to be applicable to proposed future US HTGR designs. Current codes and standards cover some HTGR materials, but not all materials are covered to the high temperatures envisioned for HTGR use. Codes and standards, particularly ASME Codes, are under development for proposed future US HTGR designs. A 'roadmap' document has been prepared for ASME Code development; a new subsection to section III of the ASME Code, ASME BPVC III-5, is scheduled to be published in October 2011. The question of terminology for the cross-duct structure between the RPV and power conversion vessel is discussed, considering the differences in regulatory requirements that apply depending on whether this structure is designated as a 'vessel' or as a 'pipe'. We conclude that designing this component as a 'pipe' is the more appropriate choice, but that the ASME BPVC

  13. The neutron standard fields at the BR1 reactor at SCK.CEN

    Energy Technology Data Exchange (ETDEWEB)

    Wagemans, J.; Malambu, E.; Borms, L. [SCKCEN, Boeretang 200, 2400 Mol (Belgium)

    2011-07-01

    The BR1 research reactor at SCK-CEN is characterized by a wide variety of irradiation possibilities, a large reactor core, and strong flexibility in its operation. A full MCNP model of BR1 has been recently developed in order to complement the results that can be obtained from activation dosimetry. After a general presentation of the reactor, this paper pays particular attention to its standard {sup 235}U(n,f) fast neutron field MARK III. This irradiation field is a useful tool for integral measurements and for detector calibrations. With the support of MCNP calculations, irradiations in MARK III can be directly referred to the pure {sup 235}U(n,f) fast neutron spectrum. (authors)

  14. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

  15. Shifts of neutrino oscillation parameters in reactor antineutrino experiments with non-standard interactions

    Directory of Open Access Journals (Sweden)

    Yu-Feng Li

    2014-11-01

    Full Text Available We discuss reactor antineutrino oscillations with non-standard interactions (NSIs at the neutrino production and detection processes. The neutrino oscillation probability is calculated with a parametrization of the NSI parameters by splitting them into the averages and differences of the production and detection processes respectively. The average parts induce constant shifts of the neutrino mixing angles from their true values, and the difference parts can generate the energy (and baseline dependent corrections to the initial mass-squared differences. We stress that only the shifts of mass-squared differences are measurable in reactor antineutrino experiments. Taking Jiangmen Underground Neutrino Observatory (JUNO as an example, we analyze how NSIs influence the standard neutrino measurements and to what extent we can constrain the NSI parameters.

  16. 76 FR 31381 - Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

    Science.gov (United States)

    2011-05-31

    ... Analysis Reports for Nuclear Power Plants,'' on a proposed Revision 4 to Standard Review Plan (SRP... Position (BTP) 8-8 on ``Onsite (Emergency Diesel Generators) and Offsite Power Sources Allowed Outage Time... COMMISSION Office Of New Reactors; Proposed Revision 4 to Standard Review Plan; Section 8.1 on Electric...

  17. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  18. EST Table: BW999205 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999205 SES2622 11/12/09 n.h 10/09/28 31 %/252 aa ref|XP_967989.1| PREDICTED: similar to lipoma...9|ref|XP_967989.1| PREDICTED: similar to lipoma preferred partner/lpp [Tribolium castaneum] BW999205 L8 ...

  19. The development and application of k -standardization method of neutron activation analysis at Es-Salam research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Alghem, L. [Departement d' Analyse par Activation Neutronique, CRNB, BP 180, Ain Oussera 17200, W Djelfa (Algeria)]. E-mail: lylia_25@hotmail.com; Ramdhane, M. [Departement de physique, Universite Mentouri de Constantine (Algeria); Khaled, S. [Departement d' Analyse par Activation Neutronique, CRNB, BP 180, Ain Oussera 17200, W Djelfa (Algeria); Akhal, T. [Departement d' Analyse par Activation Neutronique, CRNB, BP 180, Ain Oussera 17200, W Djelfa (Algeria)

    2006-01-01

    In recent years the k -NAA method has been applied and developed at the 15 MW Es-Salam research reactor, which includes: (1) the detection efficiency calibration of {gamma}-spectrometer used in k -NAA (2) the determination of reactor neutron spectrum parameters such as {alpha} and f factors in the irradiation channel, and (3) the validation of the developed k -NAA procedure by analysing SRM, namely AIEA-Soil7 and CRM, namely IGGE-GSV4. The analysis results obtained by k -NAA with 27 elements of Soil-7 standard and 14 elements of GSV-4 standard were compared with certified values. The analysis results showed that the deviations between experimental and certified values were mostly less than 10%. The k -NAA procedure established at Es-Salam research reactor has been regarded as a reliable standardization method of NAA and as available for practical applications.

  20. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  1. Mechanical Proofs about BW Multi-Party Contract Signing Protocol

    Institute of Scientific and Technical Information of China (English)

    ZHANG Ningrong; ZHANG Xingyuan; WANG Yuanyuan

    2006-01-01

    We report on the verification of a multi-party contract signing protocol described by Baum-Waidner and Waidner (BW). Based on Paulson' inductive approach, we give the protocol model that includes infinitely many signatories and contract texts signing simultaneously. We consider composite attacks of the dishonest signatory and the external intruder, formalize cryptographic primitives and protocol arithmetic including attack model, show formal description of key distribution, and prove signature key secrecy theorems and fairness property theorems of the BW protocol using the interactive theorem prover Isabelle/HOL.

  2. Evaluation of B&W UO2/ThO2 VIII experimental core: criticality and thermal disadvantage factor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Carlo Parisi; Emanuele Negrenti

    2017-02-01

    In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock & Wilcox (B&W) Spectral Shift Control Reactor (SSCR) critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO2-ThO2 fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated.

  3. EST Table: BW998133 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998133 ES3572 10/09/28 95 %/149 aa ref|NP_001040343.1| peripheral-type benzodiazepine... receptor [Bombyx mori] gb|ABF51223.1| peripheral-type benzodiazepine receptor [Bombyx mori] 10/08/29 52

  4. EST Table: BW999096 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999096 SES2266 10/09/28 85 %/193 aa ref|NP_001040402.1| preimplantation protein [...Bombyx mori] gb|ABF51322.1| preimplantation protein [Bombyx mori] 10/08/29 73 %/195 aa FBpp0252287|DwilGK231

  5. EST Table: BW997292 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available /09/10 n.h 10/09/10 33 %/121 aa gi|91077398|ref|XP_975310.1| PREDICTED: similar to Mdm2, transformed 3T3 cell double minute 2, p53 binding protein [Tribolium castaneum] BW997292 L8 ...

  6. EST Table: BW998485 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998485 ES4984 10/09/28 100 %/149 aa ref|NP_001116821.1| 18 wheeler [Bombyx mori] dbj|BAB85498.1| 18 wheele...-PA 10/09/10 58 %/154 aa gi|91076478|ref|XP_972409.1| PREDICTED: similar to 18 wheeler [Tribolium castaneum] FS922922 L8 ...

  7. EST Table: BW998803 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998803 SES1220 10/09/28 50 %/231 aa ref|XP_975194.1| PREDICTED: similar to something about...Amel|GB19080-PA 10/09/10 50 %/231 aa gi|91081571|ref|XP_975194.1| PREDICTED: similar to something about silencing protein 10 [Tribolium castaneum] FS937283 L8 ...

  8. EST Table: BW999179 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999179 SES2517 10/09/28 84 %/219 aa ref|NP_001040215.1| stathmin [Bombyx mori] gb|ABD36259.1| stathmin...7-PA 10/09/10 51 %/223 aa gi|91083957|ref|XP_975021.1| PREDICTED: similar to stathmin [Tribolium castaneum] FS915193 L8 ...

  9. EST Table: BW997643 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997643 ES1616 10/09/28 45 %/133 aa ref|XP_969699.1| PREDICTED: similar to fuzzy C...10/09/10 45 %/133 aa gi|91081315|ref|XP_969699.1| PREDICTED: similar to fuzzy CG13396-PA [Tribolium castaneum] CK514734 L8 ...

  10. EST Table: BW997930 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997930 ES2720 10/09/28 30 %/110 aa ref|XP_971125.1| PREDICTED: similar to werner .../10 30 %/110 aa gi|91094825|ref|XP_971125.1| PREDICTED: similar to werner helicase interacting protein [Tribolium castaneum] FS764071 L8 ...

  11. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  12. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  13. The cloud storage service bwSync&Share at KIT

    CERN Document Server

    CERN. Geneva

    2014-01-01

    The Karlsruhe Institute of Technology introduced the bwSync&Share collaboration service in January 2014. The service is an on-premise alternative to existing public cloud storage solutions for students and scientists in the German state of Baden-Württemberg, which allows the synchronization and sharing of documents between multiple devices and users. The service is based on the commercial software PowerFolder and is deployed on a virtual environment to support high reliability and scalability for potential 450,000 users. The integration of the state-wide federated identity management system (bwIDM) and a centralized helpdesk portal allows the service to be used by all academic institutions in the state of Baden-Württemberg. Since starting, approximately 15 organizations and 8,000 users joined the service. The talk gives an overview of related challenges, technical and organizational requirements, current architecture and future development plans.

  14. EST Table: BW999347 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999347 SES3121 10/09/28 43 %/121 aa gb|EFA09250.1| thickveins [Tribolium castaneu...| PREDICTED: similar to thickveins CG14026-PA [Tribolium castaneum] FS936628 L8 ... ...R:31432586:31434358:1|gene:AGAP009329 10/09/10 low homology 10/09/10 43 %/121 aa gi|91088837|ref|XP_970678.1

  15. EST Table: BW998785 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998785 SES1140 10/09/28 56 %/232 aa ref|XP_970433.1| PREDICTED: similar to Protein SDA1 homolog (Mystery... %/234 aa gnl|Amel|GB18390-PA 10/09/10 56 %/232 aa gi|91078808|ref|XP_970433.1| PREDICTED: similar to Protein SDA1 homolog (Mystery protein 45A) [Tribolium castaneum] FS731272 L8 ...

  16. EST Table: BW998041 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998041 ES3138 10/09/28 54 %/127 aa ref|NP_001165388.1| aliphatic nitrilase [Bomby...x mori] gb|ADB27116.1| aliphatic nitrilase [Bombyx mori] 10/08/29 n.h 10/08/28 n.h 10/09/10 n.h 10/09/10 n.h 10/09/10 n.h FS789730 L8 ...

  17. InterProScan Result: BW998866 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998866 BW998866_3_ORF2 19DAFFCDCA0766F1 PANTHER PTHR11875:SF7 NUCLEOSOME ASSEMBLY... PROTEIN 1 NA ? IPR002164 unintegrated Cellular Component: nucleus (GO:0005634)|Biological Process: nucleosome assembly (GO:0006334) ...

  18. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999574 BW999574_3_ORF1 398579F389EB61FC PROSITE PS00884 OSTEOPONTIN NA ? IPR019841 unintegrated Biological... Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ...

  19. InterProScan Result: BW998840 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998840 BW998840_2_ORF1 348217EBA47525C4 PANTHER PTHR11783:SF3 FLAVONOL SULFOTRANS...FERASE-LIKE NA ? IPR000863 unintegrated Molecular Function: sulfotransferase activity (GO:0008146) ...

  20. InterProScan Result: BW998840 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998840 BW998840_2_ORF1 348217EBA47525C4 PANTHER PTHR11783 SULFOTRANSFERASE (SULT)... 2.9e-59 T IPR000863 unintegrated Molecular Function: sulfotransferase activity (GO:0008146) ...

  1. InterProScan Result: BW998999 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998999 BW998999_1_ORF2 1916267D55526814 PANTHER PTHR11629:SF25 VACUOLAR PROTON TR...ANSLOCATING ATPASE 116 KDA SUBUNIT A ISOFORM 1 (VACUOLAR PROTON PUMP SUBUNIT 1) NA ? IPR002490 unintegrated

  2. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999574 BW999574_3_ORF1 398579F389EB61FC PRINTS PR00216 OSTEOPONTIN 2.2e-32 T IPR002038 Osteopontin Biolog...ical Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ...

  3. InterProScan Result: BW998866 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998866 BW998866_3_ORF2 19DAFFCDCA0766F1 PANTHER PTHR11875:SF7 NUCLEOSOME ASSEMBLY... PROTEIN 1 1.2e-30 T IPR002164 unintegrated Cellular Component: nucleus (GO:0005634)|Biological Process: nucleosome assembly (GO:0006334) ...

  4. InterProScan Result: BW997815 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997815 BW997815_3_ORF1 3ADEB1CFA0910F16 PANTHER PTHR23077:SF16 NUCLEAR VALOSIN-CO...NTAINING PROTEIN-LIKE (NUCLEAR VCP-LIKE PROTEIN) 7.2e-91 T IPR003960 unintegrated Molecular Function: ATP binding (GO:0005524) ...

  5. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available cal Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ... ...BW999574 BW999574_3_ORF1 398579F389EB61FC PRINTS PR00216 OSTEOPONTIN 2.2e-32 T IPR002038 Osteopontin Biologi

  6. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999574 BW999574_3_ORF1 398579F389EB61FC PFAM PF00865 Osteopontin 1.9e-62 T IPR002038 Osteopontin Biologica...l Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ...

  7. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available ical Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ... ...BW999574 BW999574_3_ORF1 398579F389EB61FC SMART SM00017 no description 0.0087 T IPR002038 Osteopontin Biolog

  8. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available ogical Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ... ...BW999574 BW999574_3_ORF1 398579F389EB61FC PANTHER PTHR10607 OSTEOPONTIN 6.2e-32 T IPR002038 Osteopontin Biol

  9. InterProScan Result: BW998452 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998452 BW998452_1_ORF1 60C99737A60E9D72 PANTHER PTHR11817 PYRUVATE KINASE 1.6e-32 T IPR001697 Pyruvat...e kinase Molecular Function: magnesium ion binding (GO:0000287)|Molecular Function: pyruvat

  10. InterProScan Result: BW998718 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998718 BW998718_3_ORF2 57ED3F4B93650AA3 PANTHER PTHR11523 SODIUM/POTASSIUM-DEPEND...ENT ATPASE BETA SUBUNIT 3.7e-91 T IPR000402 ATPase, P-type cation exchange, beta subunit Molecular Function: sodiu...006754)|Biological Process: potassium ion transport (GO:0006813)|Biological Process: sodiu

  11. InterProScan Result: BW998718 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998718 BW998718_3_ORF2 57ED3F4B93650AA3 PANTHER PTHR11523:SF15 SODIUM/POTASSIUM-D...EPENDENT ATPASE BETA SUBUNIT, INSECT NA ? IPR000402 unintegrated Molecular Function: sodium:potassium-exchan...ss: potassium ion transport (GO:0006813)|Biological Process: sodium ion transport (GO:0006814)|Cellular Component: membrane (GO:0016020) ...

  12. InterProScan Result: BW998840 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998840 BW998840_2_ORF1 348217EBA47525C4 PANTHER PTHR11783:SF3 FLAVONOL SULFOTRANS...FERASE-LIKE 2.9e-59 T IPR000863 unintegrated Molecular Function: sulfotransferase activity (GO:0008146) ...

  13. InterProScan Result: BW998999 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998999 BW998999_6_ORF2 A5018D40308EB7B7 PANTHER PTHR11629:SF25 VACUOLAR PROTON TR...ANSLOCATING ATPASE 116 KDA SUBUNIT A ISOFORM 1 (VACUOLAR PROTON PUMP SUBUNIT 1) NA ? IPR002490 unintegrated ...

  14. InterProScan Result: BW998999 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998999 BW998999_2_ORF2 03D2F37480EB153B PANTHER PTHR11629:SF25 VACUOLAR PROTON TR...ANSLOCATING ATPASE 116 KDA SUBUNIT A ISOFORM 1 (VACUOLAR PROTON PUMP SUBUNIT 1) NA ? IPR002490 unintegrated ...

  15. InterProScan Result: BW998999 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998999 BW998999_1_ORF2 1916267D55526814 PANTHER PTHR11629:SF25 VACUOLAR PROTON TR...ANSLOCATING ATPASE 116 KDA SUBUNIT A ISOFORM 1 (VACUOLAR PROTON PUMP SUBUNIT 1) 2.8e-154 T IPR002490 uninteg

  16. InterProScan Result: BW998999 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998999 BW998999_4_ORF2 243D69A457DBBB23 PANTHER PTHR11629:SF25 VACUOLAR PROTON TR...ANSLOCATING ATPASE 116 KDA SUBUNIT A ISOFORM 1 (VACUOLAR PROTON PUMP SUBUNIT 1) NA ? IPR002490 unintegrated ...

  17. InterProScan Result: BW998999 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998999 BW998999_3_ORF2 B9E0E84D15FB436E PANTHER PTHR11629:SF25 VACUOLAR PROTON TR...ANSLOCATING ATPASE 116 KDA SUBUNIT A ISOFORM 1 (VACUOLAR PROTON PUMP SUBUNIT 1) NA ? IPR002490 unintegrated ...

  18. InterProScan Result: BW997326 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997326 BW997326_1_ORF2 E9EB0AFC00B4092E PANTHER PTHR10119 GLUTAMYL/GLUTAMINYL-TRNA SYNTHETAS...E 2.9e-82 T IPR000924 Glutamyl/glutaminyl-tRNA synthetase, class Ic Molecular Function: nucleotid

  19. InterProScan Result: BW999159 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available luronic acid binding (GO:0005540)|Biological Process: cell adhesion (GO:0007155) ... ...BW999159 BW999159_2_ORF1 DA92A674762AF58C PROSITE PS01241 LINK_1 NA T IPR000538 Link Molecular Function: hya

  20. InterProScan Result: BW999159 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available nk Molecular Function: hyaluronic acid binding (GO:0005540)|Biological Process: cell adhesion (GO:0007155) ... ...BW999159 BW999159_2_ORF1 DA92A674762AF58C PFAM PF00193 Xlink 1.8e-27 T IPR000538 Li

  1. InterProScan Result: BW999292 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999292 BW999292_1_ORF2 3C8DCC56B3217B76 PANTHER PTHR11662:SF30 SIALIN (SOLUTE CAR...RIER FAMILY 17 MEMBER 5) (SODIUM/SIALIC ACID COTRANSPORTER) (AST) (MEMBRANE GLYCOPROTEIN HP59) 1.6e-86 T IPR020846 unintegrated ...

  2. InterProScan Result: BW999292 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999292 BW999292_3_ORF4 8B5BF80CEF36A721 PANTHER PTHR11662:SF30 SIALIN (SOLUTE CAR...RIER FAMILY 17 MEMBER 5) (SODIUM/SIALIC ACID COTRANSPORTER) (AST) (MEMBRANE GLYCOPROTEIN HP59) NA ? IPR020846 unintegrated ...

  3. InterProScan Result: BW999292 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999292 BW999292_5_ORF2 D6FAE59B632AC02A PANTHER PTHR11662:SF30 SIALIN (SOLUTE CAR...RIER FAMILY 17 MEMBER 5) (SODIUM/SIALIC ACID COTRANSPORTER) (AST) (MEMBRANE GLYCOPROTEIN HP59) NA ? IPR020846 unintegrated ...

  4. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  5. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Science.gov (United States)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  6. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  7. The Ultimate Light Curve of SN 1998bw/GRB 980425

    CERN Document Server

    Clocchiatti, Alejandro; Covarrubias, Ricardo; Candia, Pablo

    2011-01-01

    We present multicolor light curves of SN 1998bw which appeared in ESO184-G82 in close temporal and spacial association with GRB 980425. They are based on observations done at Cerro Tololo Inter-American Observatory and data from the literature. The CTIO photometry reaches ~86 days after the GRB in $U$ and ~160 days after the GRB in BV(RI)_C. The observations in U extend by about 30 days the previously known coverage, and determine the slope of the early exponential tail. We calibrate a large set of local standards in common with those of previous studies and use them to transform published observations of the SN to our realization of the standard photometric system. We show that the photometry from different sources merges smoothly and provide a unified set of 300 observations of the SN in five bands. Using the extensive set of spectra in public domain we compute extinction and K corrections, and build quasi-bolometric unreddened rest frame light curves. We provide low degree piecewise spline fits to these li...

  8. H-1 Upgrades (4BW/4BN) (H-1 Upgrades)

    Science.gov (United States)

    2015-12-01

    Nautical Miles R&M - Reliability and Maintainability RM - Reference Model TV-1 - Technical Standards Profile Univ . - Universal H-1 Upgrades December 2015...Speed (kts) 165 165 135 139 139 Payload (Hot Day) (lbs) 3500 lbs 3500 lbs 2500 lbs 6 Wing Stations 4 Universal Under Wing Stations 3429 3429 Weapon...Stations Universal Mounts 6 6 4 4 4 Precision Guided Munitions 16 16 12 16 16 Maneuverability/Agility (G’s) -0.5 to +2.5 -0.5 to +2.5 -0.5 to +2.5 -0.5 to

  9. 75 FR 6413 - Office of New Reactors; Proposed Revision to Standard Review Plan, Section 14.3.12 on Physical...

    Science.gov (United States)

    2010-02-09

    ...-3668; e-mail at Carol.Gallagher@nrc.gov . Mail comments to: Michael T. Lesar, Chief, Rulemaking and... Regulations, Part 73, Power Reactor Security Rule (published in the Federal Register (FR) on March 27, 2009 (74 FR 13926)). The previous version of this SRP section was published in March 2007 as an...

  10. InterProScan Result: BW998718 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW998718 BW998718_3_ORF2 57ED3F4B93650AA3 PANTHER PTHR11523:SF15 SODIUM/POTASSIUM-D...EPENDENT ATPASE BETA SUBUNIT, INSECT 3.7e-91 T IPR000402 unintegrated Molecular Function: sodium:potassium-e...Process: potassium ion transport (GO:0006813)|Biological Process: sodium ion transport (GO:0006814)|Cellular Component: membrane (GO:0016020) ...

  11. The equivalence between B-W and K-G and R-S equations

    Institute of Scientific and Technical Information of China (English)

    黄时中; 阮图南; 吴宁; 郑志鹏

    2003-01-01

    The equivalence between the Bargmann-Wigner (B-W) equations and the Klein-Gordon (K-G) equations for integral spin, and the Rarita-Schwinger (R-S) equations for half integral spin is established by explicit derivation, starting from the lowest spin cases. It is demonstrated that all the constraints or subsidiary conditions imposed on the K-G or R-S equations are included in the B-W equations.

  12. Application of UPTF data for modeling liquid draindown in the downcomer region of a PWR using RELAP5/MOD2-B&W

    Energy Technology Data Exchange (ETDEWEB)

    Wissinger, G.; Klingenfus, J. [B & W Nuclear Technologies, Lynchburg, VA (United States)

    1995-09-01

    B&W Nuclear Technologies (BWNT) currently uses an evaluation model that analyzes large break loss-of-coolant accidents in pressurized water reactors using several computer codes. These codes separately calculate the system performance during the blowdown, refill, and reflooding phases of the transient. Multiple codes are used, in part, because a single code has been unable to effectively model the transition from blowdown to reflood, particularly in the downcomer region where high steam velocities do not allow the injected emergency core cooling (ECC) liquid to penetrate and begin to refill the vessel lower plenum until after the end of blowdown. BWNT is developing a method using the RELAP5/MOD2-B&W computer code that can correctly predict the liquid draindown behavior in the downcomer during the late blowdown and refill phases. Benchmarks of this method have been performed against Upper Plenum Test Facility (UPTF) data for ECC liquid penetration and valves using both cold leg and downcomer ECC injection. The use of this new method in plant applications should result in the calculation of a shorter refill period, leading to lower peak clad temperature predictions and increased core peaking. This paper identifies changes made to the RELAP/MOD2-B&W code to improve its predictive capabilities with respect to the data obtained in the UPTF tests.

  13. Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1) This test method replaces Method E 418. This test method is more detailed and special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized. 1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically good results are obtained in the range of 5 to 800 000 tracks/cm2 and accurate results at higher track densities have been demonstrated for some cases. (2) Trac...

  14. 75 FR 29588 - Office of New Reactors: Proposed NUREG-0800; Standard Review Plan Section 13.6.6, Draft Revision...

    Science.gov (United States)

    2010-05-26

    ... NUREG-0800; Standard Review Plan Section 13.6.6, Draft Revision 0 on Cyber Security Plan AGENCY: Nuclear... public comment on NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for... revision of NUREG-0800, SRP Section 13.6.6 and Regulatory Guide 1.206, ``Combined License Applications...

  15. Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database: 1.1.1 MaterialsA 533 Type B Class 1 and 2, A302 Grade B, A302 G...

  16. Eigen rechtsingang voor minderjarigen : Ervaringen met artikel 1:162a BW

    NARCIS (Netherlands)

    Doornhein, L.

    1992-01-01

    In december 1990 is de nieuwe wet houdende een nadere regeling van de omgang in verband met scheiding in werking getreden. De Tweede Kamer is een evaluatie toegezegd van de ervaringen met het nieuwe artikel 1:162a BW, waarmee minderjarigen een (informele) eigen rechtsingang hebben gekregen. Jongeren

  17. HLA Bw35 antigen and mesangial IgA glomerulo-nephritis: a poor prognosis marker?

    Science.gov (United States)

    Berthoux, F C; Genin, C; Gagne, A; Le Petit, J C; Sabatier, J C

    1979-01-01

    Familial cases of mesangial IgA glomerulonephritis (MGN) have raised the possibility of a genetic control in this disease. In 50 patients with MGN, diagnosed on renal biopsy, and in 105 controls, we have compared the distribution of HLA antigens (A and B loci). We found a significant increase in the frequency of HLA Bw35 antigen in the patient group compared with controls (36% versus 13%: p less than 0.02). There was no significant difference between the Bw35 positive and negative MGN subgroups, in clinical, serological, and pathological data. Both subgroups had elevated mean serum IgA levels (154% of normal), and also mean serum IgM levels (146%). However, the follow-up data exhibited a significantly worse prognosis (p less than 0.01) in the Bw35 positive subgroup: 9 out of 18 patients versus 4 out of 32 progressed to chronic renal failure (serum creatinine greater than 1.5 mg/dl). We have established a genetic linkage between the HLA complex and the occurrence of MGN. The Bw35 antigen may serve as a marker (risk of disease = 4), in particular for poor prognosis cases.

  18. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  19. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  20. The regulated distribution transformer. Experiences gathered in the grid of EnBW Regional AG; Erfahrungen mit dem regelbaren Ortsnetztransformator im Netz der EnBW Regional AG

    Energy Technology Data Exchange (ETDEWEB)

    Hennig, Matthias; Koerner, Christian [EnBW Regional AG, Stuttgart (Germany); Schmid, Ronald [Siemens AG, Nuernberg (Germany); Handt, Karsten [Siemens AG, Erlangen (Germany)

    2012-07-01

    Regulated distribution transformers will play an important role in achieving voltage control for distributions grids with a high amount of volatile decentralized power generation. As a part of their research project named ''NetLab'', EnBW Regional AG tests a 400-kVA-substation-transformer provided by Siemens AG, which is fully integrated in the operating distribution grid. The transformer prototype with its thyristor-based on-load tap-changer can switch between three different transmission ratios. Its independent control unit is located in the secondary substation and supports the testing of different control algorithms. Extensive measurements allow a detailed monitoring of the voltage and load characteristics in the network and examining the transformer's switching behaviour as well as its effects on power quality. (orig.)

  1. Is there a 1998bw-like supernova in the afterglow of gamma ray burst 010921?

    CERN Document Server

    Dado, S; De Rújula, Alvaro; Dado, Shlomo; Dar, Arnon; Rujula, Alvaro De

    2002-01-01

    We use the very simple and successful Cannonball Model (CB) of gamma ray bursts (GRBs) and their afterglows (AGs) to analyze the observations of the strongly extinct optical AG of GRB 010921 with ground-based telescopes at early times, and with the HST at later time. We show that GRB 010921 was indeed associated with a 1998bw-like supernova at the GRB's redshift.

  2. Solving the generalized Sylvester matrix equation AV+BW=VF via Kronecker map

    Institute of Scientific and Technical Information of China (English)

    Aiguo WU; Siming ZHAO; Guangren DUAN

    2008-01-01

    This note considers the solution to the generalized Sylvester matrix equation AV+BW=VF with F being an arbitrary matrix,where V and W are the matrices to be determined.With the help of the Kronecker map,an explicit parametric solution to this matrix equation is established.The proposed solution possesses a very simple and neat form,and allows the matrix F to be undetermined.

  3. A summary of the quench behavior of B&W 1 m collider quadrupole model magnets

    Energy Technology Data Exchange (ETDEWEB)

    Rey, C.M.; Xu, M.F.; Hlasnicek, P.; Kelley, J.P.; Dixon, K.; Savignano, J.; Letterman, S.; Craig, P.; Maloney, J.; Boyes, D. [Babcock & Wilcox, Lynchburg, VA (United States)] [and others

    1994-12-31

    In order to evaluate the quench performance of a B&W-Siemens designed quadrupole magnet at the earliest possible stage, a model magnet program was developed at B&W for the support of the Superconducting Super Collider. The authors report the quench performance, training behavior, and the ramp rate dependence for the QSH-801 through QSH-804 series of short (1.2 meter) quadrupole model magnets.

  4. BW 3200 CMTS有线调制解调器头端系统

    Institute of Scientific and Technical Information of China (English)

    2004-01-01

    太力阳BW 3200 CMTS适用天高密度的MDU以及那些为客户提供宽带接入酒店。“Pizza盒”式的CMTS是业内唯一可扩展的解决方案,它由一个单独1RU交换控制单元组成。该交换控制单元支持

  5. Fasting heat production and metabolic BW in group-housed broilers.

    Science.gov (United States)

    Noblet, J; Dubois, S; Lasnier, J; Warpechowski, M; Dimon, P; Carré, B; van Milgen, J; Labussière, E

    2015-07-01

    Fasting heat production (FHP) is used for characterizing the basal metabolic rate of animals and the corresponding maintenance energy requirements and in the calculation of net energy value of feeds. In broilers, the most recent FHP estimates were obtained in the 1980s in slow-growing and fatter birds than nowadays. The FHP values (n=73; six experiments) measured in 3 to 6-week-old modern lines of broilers weighing 0.6 to 2.8 kg and growing at 80 to 100 g/day were used to update these literature values. Each measurement was obtained in a group of fasting broilers (5 to 14 birds) kept in a respiration chamber for at least 24 h. The FHP estimate corresponds to the asymptotic heat production corrected for zero physical activity obtained by modeling the decrease in heat production during the fasting day. The compilation of these data indicates that FHP was linearly related to the BW(0.70) (in kg), which can be considered as the metabolic BW of modern broilers. The 0.70 exponent differs from the conventional value of 0.75 used for mature animals. The FHP per kg of BW(0.70) ranged between 410 and 460 kJ/day according to the experiment (Pfasting (16 h) indicated that FHP values are higher than those obtained over at least 24 h of fasting. Our values are similar to those obtained previously on fatter and slow-growing birds, even though the comparison is difficult since measurement conditions and methodologies have changed during the last 30 years. The FHP values obtained in our trials represent a basis for energy nutrition of modern broilers.

  6. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  7. Is there a 1998bw-like supernova in the afterglow of gamma ray burst 011121?

    CERN Document Server

    Dado, S; De Rújula, Alvaro; Dado, Shlomo; Dar, Arnon; Rujula, Alvaro De

    2002-01-01

    We use the very simple and successful Cannonball Model (CB) of gamma ray bursts (GRBs) and their afterglows (AGs) to analyze the observations of the strongly extinct optical AG of the relatively nearby GRB 011121, which were made with ground-based telescopes at early times, and with the HST at later time. We show that GRB 011121 was indeed associated with a 1998bw-like supernova at the GRB's redshift, as we had specifically predicted for this GRB before the supernova could be observed.

  8. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  9. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  10. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  11. Screening a hybridoma producing a specific monoclonal antibody to HLA-A24+Bw4 antigen by cytotoxicity inhibition assay.

    Science.gov (United States)

    Hiroishi, S; Kaneko, T; Arita, J

    1987-02-01

    A hybridoma secreting a monoclonal antibody (Tsa-1, IgG3) reacting specifically to HLA-A24+Bw4 was screened by cytotoxicity inhibition assay and micrototoxicity test. The R value of the antibody was 0.843.

  12. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  13. The effect of colostrum period management on BW and immune system in lambs: from birth to weaning.

    Science.gov (United States)

    Hernández-Castellano, L E; Suárez-Trujillo, A; Martell-Jaizme, D; Cugno, G; Argüello, A; Castro, N

    2015-10-01

    The aim of this study was to investigate the BW and immune status of lambs reared under natural conditions or under artificial conditions fed two different colostrum amounts. In this study, 60 lambs were randomly divided into groups according to treatment. Twenty lambs remained with their dams (natural rearing (NR) group). Forty lambs were removed from their dams at birth. Lambs were bottle-fed with a pool of sheep colostrum, receiving either 4 g of IgG/kg of BW at birth (C4 group) or 8 g of IgG/kg of BW at birth (C8 group). The total colostrum amount was equally divided into three meals at 2, 14 and 24 h after birth. After this period, lambs were bottle-fed a commercial milk replacer. Blood plasma sample analysis and BW recordings were carried out before feeding at birth and then at 1, 2, 3, 4, 5 and 20 days after birth. Another blood sample analysis and BW recording was carried out when animals reached 10 kg of BW. During weaning (30 days), sampling was carried out every 5 days. Blood plasma was used to determine the concentrations of IgG and IgM and the complement system activity - total and alternative pathways. The NR group showed greater BW than the C4 and C8 groups during milk feeding period, whereas the C4 and C8 groups had greater BW than the NR group at the end of weaning period. The C8 and NR groups had greater plasma IgG and IgM concentrations than the C4 group during milk feeding period. In addition, C4 and C8 groups showed similar IgG concentrations and greater IgM concentrations than the NR group at the end of the weaning period. Complement system activity was greater in the NR group than in the C4 and C8 groups during the first 3 days after birth. In conclusion, lambs fed amounts of colostrum equivalent to 8 g of IgG/kg of BW showed similar immune variables compared to lambs reared under natural conditions, obtaining a greater BW at the end of the weaning period. Nevertheless, this study shows that not only the colostrum amount but also the

  14. New beamline optics of the x-ray undulator BW1 at DORIS

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, U.; Frahm, R.; Guertler, P.; Schulte-Schrepping, H. [Deutsches Elektronen-Synchrotron, Hamburg (Germany). Hamburger Synchrotronstrahlungslabor

    1996-12-31

    The X-ray undulator BW1 at the storage ring DORIS is a high brightness source for the spectral range from 2 to 20 keV. The undulator beam is used by three experiments with different distances to the source. The new optical elements allow the adaptation of the focal lengths to the needs of the experimental set-ups. The optical concept consists of a premirror with different optical surfaces, a double crystal monochromator and a focusing second mirror. Sagittal focusing is achieved either by using the cylindrical part of the premirror or by a bend crystal for a monochromatic beam, meridional focusing is done with a pneumatic driven mirror bender for the second mirror.

  15. An explicit solution to the matrix equation AV+BW=EV J

    Institute of Scientific and Technical Information of China (English)

    Aiguo WU; Guangren DUAN; Bin ZHOU

    2007-01-01

    In this note,the matrix equation AV+BW=EV J is considered,where E,A and B are given matrices of appropriate dimensions,J is an arbitrarily given Jordan matrix,V and W are the matrices to be determined.Firstly,a right factorization of(sE-A)-1 B is given based on the Leverriver algorithm for descriptor systems.Then based on this factorization and a proposed parametric solution,an alternative parametric solution to this matrix equation is established in terms of the R-controllability matrix of(E,A,B),the generalized symmetric operator and the observability matrix associated with the Jordan matrix J and a free parameter matrix.The proposed results provide great convenience for many analysis and design problems.Moreover,some equivalent forms are proposed.A numerical example is employed to illustrate the effect of the proposed approach.

  16. Biosimmer: A Virtual Reality Simulator for Training First Responders in a BW Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Shawver, D.M.; Sobel, A.L.; Stansfield, S.A.

    1998-11-11

    BioSimMER (Bioterrorism Simulated Medical Emergency Response) is a Virtual Reality-based mission rehearsal and training environment. BioSimMER employs contingency-oriented, multiple-path algorithms and MOESINIOPS focused on real-world operations. BioSimMER is network-based and immerses multiple trainees in a high resolution synthetic environment, including virtual casualties and instruments that they may interact with and manipulate. Trainees are represented as individuals by virtual human Avatars. The simulation consists of several components: virtual casualties dynamically manifest the symptoms of their injuries and respond to the intervention of the trainees. Agent transport analysis is used to simulate casualty exposures and to drive the responses of simulated sensors/detectors. The selected prototype scenario is representative of combined injuries anticipated in BW operations.

  17. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, F-13108 Saint Paul lez Durance (France); Vacelet, H. [CERCA, Romans (France); Dornbusch, D. [Technicatome, Aix en Provence (France)

    2000-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel are discussed. (author)

  18. HLA-Bw4及其KIR与宫颈癌的相关性研究%Correlation of cervical cancer with different combinations of HLA-Bw4 and its corresponding KIR

    Institute of Scientific and Technical Information of China (English)

    许联红; 蒋立新; 王永仿; 戚传平

    2013-01-01

    目的 了解江苏汉族人群HLA-Bw4及其相应的抑制性杀伤细胞免疫球蛋白样受体(KIR)和激活性KIR与宫颈癌是否存在关联性.方法 收集117例宫颈内皮样瘤变3级(CIN3)/宫颈癌患者和124例相匹配的无血缘关系健康女性外周血标本,采用PCR-SSP方法进行HLA-Bw4检测和SYBR Green Ⅰ real-time PCR方法进行KIR3DL1和3DS1分型,分析HLA-Bw4及其KIR在2组人群间的差异.结果 宫颈癌组与对照组相比,HLA-Bw4、KIR3DL1和3DS1的频率分布无显著性差异(P>0.05).HLA-Bw4与KIR3DL1和3DS1形成的不同配对组合在病例组与对照组中的频率分布,差异也无统计学意义(P>0.05).结论 江苏汉族人群HLA-Bw4及其KIR3DL1和KIR3DS1可能与宫颈癌的发生无关.%To explore the correlation of cervical cancer with HLA-Bw4 and its corresponding KIR in Han population of Jiangsu province,we collected peripheral blood samples from 117 patients with cervical intraepithelial neoplasia 3 (CIN3)/cervical cancer and 124 randomly ethnically matched healthy controls for typing KIR3DL1 and 3DS1 using SYBR Green Ⅰ real-time PCR genotyping assay and detecting HLA-Bw4 using PCR-SSP method.Different combinations of HLA-Bw4 and its corresponding KIR were compared between CIN3/cervical cancer patients and healthy controls.The results showed that the phenotype frequencies of HLA-Bw4,KIR3DL1 and 3DS1 showed no significant statistical differences between CIN3/cervical cancer group and the control group (P> 0.05).No evidence of association was found between any combination of HLA-Bw4/KIR and the risk of CIN3/cervical cancer.Thus we concluded that HLA-Bw4 and its corresponding KIR may be not associated with cervical cancer.

  19. Granulocyte-specific monoclonal antibody technetium-99m-BW 250/183 and indium-111 oxine-labelled leukocyte scintigraphy in inflammatory bowel disease

    Energy Technology Data Exchange (ETDEWEB)

    Segarra, I.; Roca, M.; Ricart, Y.; Mora, J.; Puchal, R.; Martin-Comin, J. (Hospital de Bella Vista, Barcelona (Spain). Servicio de Medicina Nuclear); Baliellas, C.; Vilar, L. (Hospital de Bella Vista, Barcelona (Spain). Servicio de Gastroenterologia)

    1991-09-01

    Thirty-three patients suspected of suffering from inflammatory bowel disease were studied. Autologous leukocytes were labelled with indium 111 oxine and re-injected simultaneously with 0.3-0.5 mg of {sup 99m}Tc-labelled granulocyte-specific monoclonal antibody BW 250/183. Two scans were obtained, the early scan 3-4 h post-injection (p.i.) and the late scan 18-24 h p.i. Using the endoscopy study as standard, the diagnostic accuracy of both agents was determined. Sensitivity, specificity and accuracy of {sup 111}In-scans was 88.8%, 100.0% and 93.7% at 4 h and 94.7%, 100.0% and 96.9% and 24 h, respectively. Concerning the results using antibodies, the values were 61.1%, 100.0% and 78.1% at 4 h and 78.9%, 92.8% and 84.8% at 24 h, respectively. Segmental analysis showed concordance in 89.3% and 93.3% of the cases at 4 and 24 h, respectively. Though less sensitive and less accurate than scanning employing indium 111 leukocytes, the BW 250/183 granulocyte-specific scintigraphy can be used for inflammatory bowel disease diagnosis and localization. (orig.).

  20. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

  1. Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Standard review plan and acceptance criteria. NUREG - 1537, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    NUREG - 1537, Part 2 gives guidance on the conduct of licensing action reviews to NRC staff who review non-power reactor licensing applications. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination.

  2. Licensed reactor nuclear safety criteria applicable to DOE reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  3. Evidence for Magnetar Formation in Broad-lined Type Ic Supernovae 1998bw and 2002ap

    Science.gov (United States)

    Wang, L. J.; Yu, H.; Liu, L. D.; Wang, S. Q.; Han, Y. H.; Xu, D.; Dai, Z. G.; Qiu, Y. L.; Wei, J. Y.

    2017-03-01

    Broad-lined type Ic supernovae (SNe Ic-BL) are peculiar stellar explosions that are distinct from ordinary SNe. Some SNe Ic-BL are associated with long-duration (≳ 2 {{s}}) gamma-ray bursts (GRBs). Black holes and magnetars are two types of compact objects that are hypothesized to be central engines of GRBs. In spite of decades of investigations, no direct evidence for the formation of black holes or magnetars has yet been found for GRBs. Here we report the finding that the early peak (t≲ 50 {days}) and late-time (t≳ 300 {days}) slow decay displayed in the light curves of SNe 1998bw (associated with GRB 980425) and 2002ap (not GRB-associated) can be attributed to magnetar spin-down with an initial rotation period {P}0∼ 20 {ms}, while the intermediate-time (50≲ t≲ 300 {days}) exponential decline is caused by the radioactive decay of 56Ni. The connection between the early peak and late-time slow decline in the light curves is unexpected in alternative models. We thus suggest that GRB 980425 and SN 2002ap were powered by magnetars.

  4. Evidence for magnetar formation in broad-lined type Ic supernovae 1998bw and 2002ap

    CERN Document Server

    Wang, L J; Liu, L D; Wang, S Q; Han, Y H; Xu, D; Dai, Z G; Qiu, Y L; Wei, J Y

    2016-01-01

    Broad-lined type Ic supernovae (SNe Ic-BL) are peculiar stellar explosions that distinguish themselves from ordinary SNe. Some SNe Ic-BL are associated with long-duration (\\gtrsim 2 s) gamma-ray bursts (GRBs). Black holes and magnetars are two types of compact objects that are hypothesized to be central engines of GRBs. In spite of decades of investigations, no direct evidence for the formation of black holes or magnetars has been found for GRBs so far. Here we report the discovery that the early peak (t \\lesssim 50 days) and late-time (t \\gtrsim 300 days) slow decay displayed in the light curves of both SNe 1998bw (associated with GRB 980425) and 2002ap (not GRB-associated) can be attributed to magnetar spin-down with initial rotation period P0 \\sim 20 ms, while the intermediate-time (50 \\lesssim t \\lesssim 300 days) linear decline is caused by radioactive decay of 56Ni. The connection between the early peak and late-time slow decline in the light curves is unexpected in alternative models. We thus suggest t...

  5. B.W. Vilakazi and the birth of the Zulu novel

    Directory of Open Access Journals (Sweden)

    N.N. Canonici

    2010-07-01

    Full Text Available B.W. Vilakazi is rightly famous for his Zulu poems that integrate the Zulu creative genius with established European poetic trends. He was also the creator of the Zulu romantic novel, having written the first three examples of the genre dealing with both personal and national romantic ideals. These are, however, seldom analysed. This article reflects on the emerging literatures in African languages, their aims, contents and forms. After a general introduction on Vilakazi’s life and innovative approach to creative writing within the context of the African mini-renaissance period of the 1930s, there is a brief exposition of Vilakazi’s vision of an African literature, rooted in the need for self-identification, and recognition of perceived historical greatness. Then each novel is contextualised and analysed, through a description of the characters that exert the greatest influence on the events, since plot and character are also the highest achievement of the folktale, when told by expert performers. An attempt is also made to identify Afro-centric narrative elements and to justify perceived shortcomings in plot construction.

  6. Nebular Spectra of SN 1998bw Revisited: Detailed Study by One and Two Dimensional Models

    CERN Document Server

    Maeda, K; Mazzali, P A; Deng, J

    2006-01-01

    Refined one- and two-dimensional models for the nebular spectra of the hyper-energetic Type Ic supernova (SN) 1998bw, associated with the gamma-ray burst GRB980425, from 125 to 376 days after B-band maximum are presented. One dimensional, spherically symmetric spectrum synthesis calculations show that reproducing features in the observed spectra, i.e., the sharply peaked [OI] 6300\\AA doublet and MgI] 4570\\AA emission, and the broad [FeII] blend around 5200\\AA, requires the existence of a high-density O-rich core expanding at low velocities ($\\lsim 8,000$ km s$^{-1}$) and of Fe-rich material moving faster than the O-rich material. Synthetic spectra at late phases from aspherical (bipolar) explosion models are also computed with a two-dimensional spectrum synthesis code. The above features are naturally explained by the aspherical model if the explosion is viewed from a direction close to the axis of symmetry ($\\sim 30^{\\rm o}$), since the aspherical model yields a high-density O-rich region confined along the ...

  7. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  8. Reactor vessel

    OpenAIRE

    Makkee, M.; Kapteijn, F.; Moulijn, J.A

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and comprising connections for supplying (9, 10, 11) and discharging (13, 14, 15) via the channels (3) gases and/or liquids entering into a reaction with each other and substances formed upon this reactio...

  9. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  10. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  11. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  12. Reactor Neutrinos

    OpenAIRE

    Lasserre, T.; Sobel, H.W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrin...

  13. Human platelet specical antigen-17bw (HPA-17bw) genotyping and genetic frequency study%血小板特异性抗原HPA-17bw基因分型及基因频率调查

    Institute of Scientific and Technical Information of China (English)

    刘衍春; 薛敏; 吴敏慧; 许剑锋; 唐荣才

    2009-01-01

    目的 对HPA-17bw基因分型并调查其在部分汉族人群中的分布.方法 采用PCR-SSP方法 对无亲缘关系的健康无偿献血者HPA-17bw进行基因分型,采用基因计数法进行基因频率、基因型频率的计算.结果 259名正常人的HPA-17bw基因频率为0.000; HPA-17 bw/bw的基因型频率为0.000.结论 HPA-17bw为低频抗原,259份汉族样本中尚检测不出HPA-17bw.

  14. Towards standardization of the dissemination measures and tritium solubility in materials of fusion reactors; Hacia la estandarizacion de las medidas de difusion y solubilidad de tritio en materiales de reactores de fusion

    Energy Technology Data Exchange (ETDEWEB)

    Alberto, G.; Penalva, I.; Aranburu, I.; Sarrionandia-Ibarra, A.; Legarda, F.; Martinez, P. M.; Sedano, L.; Moral, N.

    2011-07-01

    The standardization of the measurements of hydrogen isotope interaction with different materials is a challenge and goal of fusion technology programs worldwide. For decades the programs have promoted the need for a reference laboratory for measurements of hydrogen transport to the evolution of fusion technology, but that goal is still pending, in contrast to the situation in other goals I+D.

  15. Activating KIR and HLA Bw4 ligands are associated to decreased susceptibility to pemphigus foliaceus, an autoimmune blistering skin disease.

    Directory of Open Access Journals (Sweden)

    Danillo G Augusto

    Full Text Available The KIR genes and their HLA class I ligands have thus far not been investigated in pemphigus foliaceus (PF and related autoimmune diseases, such as pemphigus vulgaris. We genotyped 233 patients and 204 controls for KIR by PCR-SSP. HLA typing was performed by LABType SSO reagent kits. We estimated the odds ratio, 95% confidence interval and performed logistic regression analyses to test the hypothesis that KIR genes and their known ligands influence susceptibility to PF. We found significant negative association between activating genes and PF. The activating KIR genes may have an overlapping effect in the PF susceptibility and the presence of more than three activating genes was protective (OR=0.49, p=0.003. A strong protective association was found for higher ratios activating/inhibitory KIR (OR=0.44, p=0.001. KIR3DS1 and HLA-Bw4 were negatively associated to PF either isolated or combined, but higher significance was found for the presence of both together (OR=0.34, p<10(-3 suggesting that the activating function is the major factor to interfere in the PF pathogenesis. HLA-Bw4 (80I and 80T was decreased in patients. There is evidence that HLA-Bw4(80T may also be important as KIR3DS1 ligand, being the association of this pair (OR=0.07, p=0.001 stronger than KIR3DS1-Bw4(80I (OR=0.31, p=0.002. Higher levels of activating KIR signals appeared protective to PF. The activating KIR genes have been commonly reported to increase the risk for autoimmunity, but particularities of endemic PF, like the well documented influence the environmental exposure in the pathogenesis of this disease, may be the reason why activated NK cells probably protect against pemphigus foliaceus.

  16. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  17. On the Parametric Solution to the Second-Order Sylvester Matrix Equation EVF2−AVF−CV=BW

    Directory of Open Access Journals (Sweden)

    Wang Guo-Sheng

    2007-01-01

    Full Text Available This paper considers the solution to a class of the second-order Sylvester matrix equation EVF2−AVF−CV=BW. Under the controllability of the matrix triple (E,A,B, a complete, general, and explicit parametric solution to the second-order Sylvester matrix equation, with the matrix F in a diagonal form, is proposed. The results provide great convenience to the analysis of the solution to the second-order Sylvester matrix equation, and can perform important functions in many analysis and design problems in control systems theory. As a demonstration, an illustrative example is given to show the effectiveness of the proposed solution.

  18. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

  19. Radiation safety assessment and development of environmental radiation monitoring technology; standardization of input parameters for the calculation of annual dose from routine releases from commercial reactor effluents

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I. H.; Cho, D.; Youn, S. H.; Kim, H. S.; Lee, S. J.; Ahn, H. K. [Soonchunhyang University, Ahsan (Korea)

    2002-04-01

    This research is to develop a standard methodology for determining the input parameters that impose a substantial impact on radiation doses of residential individuals in the vicinity of four nuclear power plants in Korea. We have selected critical nuclei, pathways and organs related to the human exposure via simulated estimation with K-DOSE 60 based on the updated ICRP-60 and sensitivity analyses. From the results, we found that 1) the critical nuclides were found to be {sup 3}H, {sup 133}Xe, {sup 60}Co for Kori plants and {sup 14}C, {sup 41}Ar for Wolsong plants. The most critical pathway was 'vegetable intake' for adults and 'milk intake' for infants. However, there was no preference in the effective organs, and 2) sensitivity analyses showed that the chemical composition in a nuclide much more influenced upon the radiation dose than any other input parameters such as food intake, radiation discharge, and transfer/concentration coefficients by more than 102 factor. The effect of transfer/concentration coefficients on the radiation dose was negligible. All input parameters showed highly estimated correlation with the radiation dose, approximated to 1.0, except for food intake in Wolsong power plant (partial correlation coefficient (PCC)=0.877). Consequently, we suggest that a prediction model or scenarios for food intake reflecting the current living trend and a formal publications including details of chemical components in the critical nuclei from each plant are needed. Also, standardized domestic values of the parameters used in the calculation must replace the values of the existed or default-set imported factors via properly designed experiments and/or modelling such as transport of liquid discharge in waters nearby the plants, exposure tests on corps and plants so on. 4 figs., 576 tabs. (Author)

  20. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  1. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  2. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  3. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  4. 核测量系统采用标准输出信号的接口问题处理%New Interface Issues of Nuclear Reactor Neutron Flux Measurement System Using Unified Standard Signal

    Institute of Scientific and Technical Information of China (English)

    王学杰; 唐凤平; 朱世雷; 黄文; 钟定永

    2012-01-01

    核测量系统采用标准输出信号后,采用自动换档电流放大器的功率测量装置在档位切换时,输出信号有效值与量程的同步性可能出现问题,使保护系统、功率控制系统和报警系统信号接收端产生误动作.本工作对分析接口问题产生的原因进行分析,并提出解决办法.%When nuclear reactor control and protection system uses a unified standard electrical analog signal(4~20 mA)as an exchange signal, there exist problems of synchronization between the virtual value and range of output signals during switching of Neutron Flux Measurement System (NFMS) , and thus false actions occur in the protection system, the alarm system and the power control system. The cause of these new interface issues were described in this paper, and one solution for this problem was given in detail.

  5. Regional Fluctuation in the Functional Consequence of LINE-1 Insertion in the Mitf Gene: The Black Spotting Phenotype Arisen from the Mitfmi-bw Mouse Lacking Melanocytes.

    Science.gov (United States)

    Takeda, Kazuhisa; Hozumi, Hiroki; Ohba, Koji; Yamamoto, Hiroaki; Shibahara, Shigeki

    2016-01-01

    Microphthalmia-associated transcription factor (Mitf) is a key regulator for differentiation of melanoblasts, precursors to melanocytes. The mouse homozygous for the black-eyed white (Mitfmi-bw) allele is characterized by the white-coat color and deafness with black eyes due to the lack of melanocytes. The Mitfmi-bw allele carries LINE-1, a retrotransposable element, which results in the Mitf deficiency. Here, we have established the black spotting mouse that was spontaneously arisen from the homozygous Mitfmi-bw mouse lacking melanocytes. The black spotting mouse shows multiple black patches on the white coat, with age-related graying. Importantly, each black patch also contains hair follicles lacking melanocytes, whereas the white-coat area completely lacks melanocytes. RT-PCR analyses of the pigmented patches confirmed that the LINE-1 insertion is retained in the Mitf gene of the black spotting mouse, thereby excluding the possibility of the somatic reversion of the Mitfmi-bw allele. The immunohistochemical analysis revealed that the staining intensity for beta-catenin was noticeably lower in hair follicles lacking melanocytes of the homozygous Mitfmi-bw mouse and the black spotting mouse, compared to the control mouse. In contrast, the staining intensity for beta-catenin and cyclin D1 was higher in keratinocytes of the black spotting mouse, compared to keratinocytes of the control mouse and the Mitfmi-bw mouse. Moreover, the keratinocyte layer appears thicker in the Mitfmi-bw mouse, with the overexpression of Ki-67, a marker for cell proliferation. We also show that the presumptive black spots are formed by embryonic day 15.5. Thus, the black spotting mouse provides the unique model to explore the molecular basis for the survival and death of developing melanoblasts and melanocyte stem cells in the epidermis. These results indicate that follicular melanocytes are responsible for maintaining the epidermal homeostasis; namely, the present study has provided

  6. Reactor and method of operation

    Science.gov (United States)

    Wheeler, John A.

    1976-08-10

    A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

  7. Isolation and characterization of a virus (CvV-BW1 that infects symbiotic algae of Paramecium bursaria in Lake Biwa, Japan

    Directory of Open Access Journals (Sweden)

    Kasahara Masahiro

    2010-09-01

    Full Text Available Abstract Background We performed an environmental study of viruses infecting the symbiotic single-celled algae of Paramecium bursaria (Paramecium bursaria Chlorella virus, PBCV in Lake Biwa, the largest lake in Japan. The viruses detected were all Chlorella variabilis virus (CvV = NC64A virus. One of them, designated CvV-BW1, was subjected to further characterization. Results CvV-BW1 formed small plaques and had a linear DNA genome of 370 kb, as judged by pulsed-field gel electrophoresis. Restriction analysis indicated that CvV-BW1 DNA belongs to group H, one of the most resistant groups among CvV DNAs. Based on a phylogenetic tree constructed using the dnapol gene, CvV was classified into two clades, A and B. CvV-BW1 belonged to clade B, in contrast to all previously identified virus strains of group H that belonged to clade A. Conclusions We conclude that CvV-BW1 composes a distinct species within C. variabilis virus.

  8. System assessment of helical reactors in comparison with tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-10-01

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-{beta}{sub N} tokamak reactors. (author)

  9. Dopamine-dependent behavioural stimulation by non-peptide delta opioids BW373U86 and SNC 80: 2. Place-preference and brain microdialysis studies in rats.

    Science.gov (United States)

    Longoni, R; Cadoni, C; Mulas, A; Di Chiara, G; Spina, L

    1998-02-01

    The motivational properties of the non-peptide delta-opioid receptor agonists BW373U86 and SNC 80 were investigated using the place-conditioning paradigm. BW373U86 (0.5-1.0 mg/kg s.c.) and SNC 80 (1.25-5.0 mg/kg s.c.) elicited significant preference for the drug-paired compartment, in a dose-related fashion. Naltrindole (5.0 mg/kg s.c.) pretreatment, while failing to modify preference when given alone, completely prevented place-preference induced by BW373U86 (1.0 mg/kg s.c.) and SNC 80 (1.25 mg/kg s.c.). The dopamine D1 receptor antagonist SCH23390, given at doses that do not affect place-preference (0.012 mg/kg s.c.), completely prevented the place-preference induced by BW373U86 and SNC 80. At the doses effective in eliciting place-preference, BW373U86 and SNC 80 failed to modify extracellular dopamine in the medial nucleus accumbens, while in the dorso-lateral caudate-putamen BW373U86 (1.0 and 2.5 mg/kg s.c.) reduced extracellular dopamine, and this effect was prevented by naltrindole (5.0 mg/kg s.c.). SNC 80, only at the dose of 5 mg/kg s.c., significantly reduced extracellular DA in the dorso-lateral caudate-putamen. The results indicate that stimulation of delta-opioid receptors has incentive properties that might be related to an indirect amplification of post-synaptic dopamine transmission.

  10. Three protagonists in B.W. Vilakazi’s “Ezinkomponi” (“On the mine compounds”

    Directory of Open Access Journals (Sweden)

    N. Zondi

    2011-06-01

    Full Text Available In this poem the great Zulu poet B.W. Vilakazi is preoccupied with the surreal scene of a gold mine compound in the 1940s Johannesburg, and reflects on the three protagonists of the drama that plays out in front of him: the miners, mine magnates and the heavy machinery, all things that drive the entire enterprise of enslaving the workers. Feelings flood his imagination: about the terrible status of the miners (with whom he identifies; what they have left behind, their dreams and the reality they battle with; the unfeeling and overwhelming spectre of industrialisation, and distant capitalist interests; and the instruments of oppression: the deafening mine machines. These three protagonists(especially the first and the third, assume human characteristics and fight to justify their respective roles in the conflict. Vilakazi’s famous protest poem becomes a cry for help in the face of destructive industrial advancement as everpresent human drama, which pits values of gold/ and money against what is more fully human and worth living for; possibly unachievable present prosperity against a vision of future happiness and fulfilment.

  11. A single nucleotide polymorphism of the neuropeptide B/W receptor-1 gene influences the evaluation of facial expressions.

    Directory of Open Access Journals (Sweden)

    Noriya Watanabe

    Full Text Available Neuropeptide B/W receptor-1 (NPBWR1 is expressed in discrete brain regions in rodents and humans, with particularly strong expression in the limbic system, including the central nucleus of the amygdala. Recently, Nagata-Kuroiwa et al. reported that Npbwr1(-/- mice showed changes in social behavior, suggesting that NPBWR1 plays important roles in the emotional responses of social interactions.The human NPBWR1 gene has a single nucleotide polymorphism at nucleotide 404 (404A>T; SNP rs33977775. This polymorphism results in an amino acid change, Y135F. The results of an in vitro experiment demonstrated that this change alters receptor function. We investigated the effect of this variation on emotional responses to stimuli of showing human faces with four categories of emotional expressions (anger, fear, happiness, and neutral. Subjects' emotional levels on seeing these faces were rated on scales of hedonic valence, emotional arousal, and dominance (V-A-D. A significant genotype difference was observed in valence evaluation; the 404AT group perceived facial expressions more pleasantly than did the 404AA group, regardless of the category of facial expression. Statistical analysis of each combination of [V-A-D and facial expression] also showed that the 404AT group tended to feel less submissive to an angry face than did the 404AA group. Thus, a single nucleotide polymorphism of NPBWR1 seems to affect human behavior in a social context.

  12. Multi-Dimensional Simulations for Early Phase Spectra of Aspherical Hypernovae: SN 1998bw and Off-Axis Hypernovae

    CERN Document Server

    Tanaka, Masaomi; Mazzali, Paolo A; Nomoto, Ken'ichi

    2007-01-01

    Early phase optical spectra of aspherical jet-like supernovae (SNe) are presented. We focus on energetic core-collapse SNe, or hypernovae. Based on hydrodynamic and nucleosynthetic models, radiative transfer in SN atmosphere is solved with a multi-dimensional Monte-Carlo radiative transfer code, SAMURAI. Since the luminosity is boosted in the jet direction, the temperature there is higher than in the equatorial plane by ~ 2,000 K. This causes anisotropic ionization in the ejecta. Emergent spectra are different depending on viewing angle, reflecting both aspherical abundance distribution and anisotropic ionization. Spectra computed with an aspherical explosion model with kinetic energy 20 x 10^{51} ergs are compatible with those of the Type Ic SN 1998bw if ~ 10-20% of the synthesized metals are mixed out to higher velocities. The simulations enable us to predict the properties of off-axis hypernovae. Even if an aspherical hypernova explosion is observed from the side, it should show hypernova-like spectra but ...

  13. Protest against social inequalities in B.W. Vilakazi’s poem 'Ngoba ... sewuthi' ('Because ... you now say'

    Directory of Open Access Journals (Sweden)

    N. Zondi

    2005-07-01

    Full Text Available Long before the National Party institutionalised apartheid in 1948, individuals and organisations tried to highlight the injustices of the colonial capitalist system in South Africa, but, as Lodge (1983:6 puts it, “it all ended in speeches”. This article seeks to demonstrate how Benedict Wallet Vilakazi effectively broke the silence by bringing the plight of the black masses to the attention of the world. He strongly protested against the enslavement of black labourers, especially in the gold and diamond mines, that he depicts as responsible for the human, psychological and physical destruction of the black working classes. As a self-appointed spokesperson of the oppressed, he protested against the injustices through the medium of his poetry. One of his grave concerns was the fact that black workers had been reduced to a class with no name, no rights, practically with no life and no soul. The chosen poem “Ngoba … sewuthi” (Because … you now say is thus representative of the poems in which B.W Vilakazi externalised his commitment to the well-being of the black workers, and his protest against the insensitivity of white employers.

  14. Grinding behavior and surface appearance of (TiCp+TiBw)/Ti-6Al-4V titanium matrix composites

    Institute of Scientific and Technical Information of China (English)

    Ding Wenfeng; Zhao Biao; Xu Jiuhua; Yang Changyong; Fu Yucan; Su Honghua

    2014-01-01

    (TiCp+TiBw)/Ti-6Al-4V titanium matrix composites (PTMCs) have broad application prospects in the aviation and nuclear field. However, it is a typical difficult-to-cut material due to high hardness of the reinforcements, high strength and low thermal conductivity of Ti-6Al-4V alloy matrix. Grinding experiments with vitrified CBN wheels were conducted to analyze comparatively the grinding performance of PTMCs and Ti-6Al-4V alloy. Grinding force and force ratios, specific grinding energy, grinding temperature, surface roughness, ground surface appearance were dis-cussed. The results show that the normal grinding force and the force ratios of PTMCs are much larger than that of Ti-6Al-4V alloy. Low depth of cut and high workpiece speed are generally ben-eficial to achieve the precision ground surface for PTMCs. The hard reinforcements of PTMCs are mainly removed in the ductile mode during grinding. However, the removal phenomenon of the reinforcements due to brittle fracture still exists, which contributes to the lower specific grinding energy and grinding temperature of PTMCs than Ti-6Al-4V alloy.

  15. Role of oxidative stress in inactivation of Escherichia coli BW25113 by nanoscale zero-valent iron.

    Science.gov (United States)

    Chaithawiwat, Krittanut; Vangnai, Alisa; McEvoy, John M; Pruess, Birgit; Krajangpan, Sita; Khan, Eakalak

    2016-09-15

    An Escherichia coli BW25113 wildtype strain and mutant strains lacking genes that protect against oxidative stress were examined at different growth phases for susceptibility to zero-valent iron (nZVI). Viability of cells was determined by the plate count method. All mutant strains were more susceptible than the wild type strain to nZVI; however, susceptibility differed among the mutant strains. Consistent with the role of rpoS as a global stress regulator, an rpoS gene knockout mutant exhibited the greatest susceptibility to nZVI under the majority of conditions tested (except exponential and declining phases at longer exposure time). Mutants lacking genes encoding the inducible and constitutively expressed cytosolic superoxide dismutases, sodA and sodB, respectively, were more susceptible to nZVI than a mutant lacking the gene encoding sodC, a periplasmic superoxide dismutase. This suggests that nZVI induces oxidative stress inside the cells via superoxide generation. Quantitative polymerase chain reaction was used to examine the expression of katG, a gene encoding the catalase-peroxidase enzyme, in nZVI-treated E. coli at different growth phases. Results showed that nZVI repressed the expression of katG in all but lag phases.

  16. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  17. Reactor Simulation for Antineutrino Experiments using DRAGON and MURE

    CERN Document Server

    Jones, C L; Conrad, J M; Djurcic, Z; Fallot, M; Giot, L; Keefer, G; Onillon, A; Winslow, L

    2011-01-01

    Rising interest in nuclear reactors as a source of antineutrinos for experiments motivates validated, fast, and accessible simulations to predict reactor fission rates. Here we present results from the DRAGON and MURE simulation codes and compare them to other industry standards for reactor core modeling. We use published data from the Takahama-3 reactor to evaluate the quality of these simulations against the independently measured fuel isotopic composition. The propagation of the uncertainty in the reactor operating parameters to the resulting antineutrino flux predictions is also discussed.

  18. Dopamine-dependent behavioural stimulation by non-peptide delta opioids BW373U86 and SNC 80: 1. Locomotion, rearing and stereotypies in intact rats.

    Science.gov (United States)

    Spina, L; Longoni, R; Mulas, A; Chang, K J; Di Chiara, G

    1998-02-01

    The unconditioned behavioural effects of two non-peptide delta-opioid receptor agonists, BW 373U86 and SNC 80, were studied in the intact rat. BW 373U86 (0.1-2.5 mg/kg s.c.) and SNC 80 (2.5-10 mg/kg s.c.) dose-dependently elicited locomotion, rearing, stereotyped sniffing, licking and gnawing. These effects were abolished by pretreatment with the delta-opioid receptor antagonist naltrindole (5.0 mg/kg s.c.). In view of the phenomenological similarities between this syndrome and that elicited by dopamine-receptor agonists, the role played by dopamine receptors was investigated. The specific dopamine D1 receptor antagonist SCH 23390 and the specific dopamine D2/D3 receptor antagonist raclopride reduced or even abolished the behavioural stimulation induced by lower doses of BW 373U86 and SNC 80. When higher doses of BW 373U86 were used (2.5 mg/kg), however, raclopride, even at high cataleptic doses (6.0 mg/kg), only partly prevented the behavioural stimulation induced by the delta-opioid receptor agonist. The behavioural stimulation remaining after high doses of raclopride was abolished by the administration of SCH 23390. These results show that delta-opioid receptor stimulation elicits dopamine-dependent behavioural activation in the rat that depends on dopamine receptors, particularly of the D1 subtype.

  19. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  20. KIR polymorphisms modulate peptide-dependent binding to an MHC class I ligand with a Bw6 motif.

    Directory of Open Access Journals (Sweden)

    Arnaud D Colantonio

    2011-03-01

    Full Text Available Molecular interactions between killer immunoglobulin-like receptors (KIRs and their MHC class I ligands play a central role in the regulation of natural killer (NK cell responses to viral pathogens and tumors. Here we identify Mamu-A1*00201 (Mamu-A*02, a common MHC class I molecule in the rhesus macaque with a canonical Bw6 motif, as a ligand for Mamu-KIR3DL05. Mamu-A1*00201 tetramers folded with certain SIV peptides, but not others, directly stained primary NK cells and Jurkat cells expressing multiple allotypes of Mamu-KIR3DL05. Differences in binding avidity were associated with polymorphisms in the D0 and D1 domains of Mamu-KIR3DL05, whereas differences in peptide-selectivity mapped to the D1 domain. The reciprocal exchange of the third predicted MHC class I-contact loop of the D1 domain switched the specificity of two Mamu-KIR3DL05 allotypes for different Mamu-A1*00201-peptide complexes. Consistent with the function of an inhibitory KIR, incubation of lymphocytes from Mamu-KIR3DL05(+ macaques with target cells expressing Mamu-A1*00201 suppressed the degranulation of tetramer-positive NK cells. These observations reveal a previously unappreciated role for D1 polymorphisms in determining the selectivity of KIRs for MHC class I-bound peptides, and identify the first functional KIR-MHC class I interaction in the rhesus macaque. The modulation of KIR-MHC class I interactions by viral peptides has important implications to pathogenesis, since it suggests that the immunodeficiency viruses, and potentially other types of viruses and tumors, may acquire changes in epitopes that increase the affinity of certain MHC class I ligands for inhibitory KIRs to prevent the activation of specific NK cell subsets.

  1. Regional Fluctuation in the Functional Consequence of LINE-1 Insertion in the Mitf Gene: The Black Spotting Phenotype Arisen from the Mitfmi-bw Mouse Lacking Melanocytes.

    Directory of Open Access Journals (Sweden)

    Kazuhisa Takeda

    Full Text Available Microphthalmia-associated transcription factor (Mitf is a key regulator for differentiation of melanoblasts, precursors to melanocytes. The mouse homozygous for the black-eyed white (Mitfmi-bw allele is characterized by the white-coat color and deafness with black eyes due to the lack of melanocytes. The Mitfmi-bw allele carries LINE-1, a retrotransposable element, which results in the Mitf deficiency. Here, we have established the black spotting mouse that was spontaneously arisen from the homozygous Mitfmi-bw mouse lacking melanocytes. The black spotting mouse shows multiple black patches on the white coat, with age-related graying. Importantly, each black patch also contains hair follicles lacking melanocytes, whereas the white-coat area completely lacks melanocytes. RT-PCR analyses of the pigmented patches confirmed that the LINE-1 insertion is retained in the Mitf gene of the black spotting mouse, thereby excluding the possibility of the somatic reversion of the Mitfmi-bw allele. The immunohistochemical analysis revealed that the staining intensity for beta-catenin was noticeably lower in hair follicles lacking melanocytes of the homozygous Mitfmi-bw mouse and the black spotting mouse, compared to the control mouse. In contrast, the staining intensity for beta-catenin and cyclin D1 was higher in keratinocytes of the black spotting mouse, compared to keratinocytes of the control mouse and the Mitfmi-bw mouse. Moreover, the keratinocyte layer appears thicker in the Mitfmi-bw mouse, with the overexpression of Ki-67, a marker for cell proliferation. We also show that the presumptive black spots are formed by embryonic day 15.5. Thus, the black spotting mouse provides the unique model to explore the molecular basis for the survival and death of developing melanoblasts and melanocyte stem cells in the epidermis. These results indicate that follicular melanocytes are responsible for maintaining the epidermal homeostasis; namely, the present study

  2. Basis for Interim Operation for the K-Reactor in Cold Standby

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, B.

    1998-10-19

    The Basis for Interim Operation (BIO) document for K Reactor in Cold Standby and the L- and P-Reactor Disassembly Basins was prepared in accordance with the draft DOE standard for BIO preparation (dated October 26, 1993).

  3. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Hiroto

    1995-02-07

    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  4. Percutaneous ultrasound-guided ablation of BW7756-hepatoma using ethanol or acetic acid in a rat model

    Directory of Open Access Journals (Sweden)

    Galeotti Tommaso

    2007-12-01

    Full Text Available Abstract Background To compare tumor necrosis in hepatoma induced in rats by a single percutaneous injection of ethanol (PEI or acetic acid (PAI. Methods BW7756 hepatomas of 1 mm3 were implanted in the liver of 40 male healthy rats. After 14 days, the 36 surviving rats were treated, in a single session, by ultrasound-guided injection of 300 μl of 95% ethanol (n = 17 or 100 μl of 50% acetic acid (n = 19. They were sacrificed 14 days after treatment and explanted tumoral livers were examined. The same PAI procedure was repeated on 13 additional rats to exclude a suspected occurrence of technical failures during the experiment, due to a surprisingly high rate of deaths within 30 minutes after PAI. Results Four rats died within four days after tumor implantation; after PEI, 1/17 (6% died, whereas after PAI 9/19 (47% died. The remaining 26 rats, after 14 days post-percutaneous ablation, were sacrificed. Gross and microscopic examinations showed that the hepatoma's nodules treated with PEI had 45.3 ± 19.4% tumor necrosis compared to 49 ± 23.3% (P = NS for those treated with PAI. Complete tumor necrosis was not found in any animal. Peritoneal invasion was present in 4/16 (25% and 2/10 (20% rats treated with PEI or PAI, respectively (P = NS. Autopsy was performed in the 5 additional rats that died within 30 minutes after PAI. Conclusion Our results show that there is no significant difference in the percentage of tumor necrosis between two local ablation methods in spite of the different dosages used. However, mortality in the PAI-treated group was greater than in PEI-treated group, presumably due to greater acetic acid systemic diffusion and its metabolic side effects. In human subjects, HCC occurs in the setting of cirrhosis, where the non-tumoral tissue is firmer than the tumor structure, with consequent reduction of drug diffusion. This could be the reason why some human studies have concluded similar or even better safety and efficacy with PAI

  5. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  6. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  7. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  8. Design guide for Category III reactors: pool type reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-11-01

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems.

  9. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  10. Risk management activities at the DOE Class A reactor facilities

    Energy Technology Data Exchange (ETDEWEB)

    Sharp, D.A. [Westinghouse Savannah River Co., Aiken, SC (United States); Hill, D.J. [Argonne National Lab., IL (United States); Linn, M.A. [Oak Ridge National Lab., TN (United States); Atkinson, S.A. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Hu, J.P. [Brookhaven National Lab., Upton, NY (United States)

    1993-12-31

    The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented.

  11. Risk management activities at the DOE Class A reactor facilities

    Energy Technology Data Exchange (ETDEWEB)

    Sharp, D.A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Hill, D.J. (Argonne National Lab., IL (United States)); Linn, M.A. (Oak Ridge National Lab., TN (United States)); Atkinson, S.A. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Hu, J.P. (Brookhaven National Lab., Upton, NY (United States))

    1993-01-01

    The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented.

  12. Study on the Interaction between Rare Earth Salts of Heteropoly K15[Ce(BW11O39)2]·17H2O and Bovine Serum Albumin%稀土杂多配合物K15[Ce (BW11O39)2]·17H2O与BSA的相互作用

    Institute of Scientific and Technical Information of China (English)

    曾芳; 杨昌英; 潘家荣; 赵儒铭; 周百斌

    2006-01-01

    利用紫外光谱,以曙红Y(Eosin Y)为探针,考察了稀土杂多配合物K15[Ce(BW11O39)2]·17H2O与牛血清白蛋白(BSA)间的作用,K15[Ce(BW11O39)2]·17H2O与Eosin Y在BSA上有竞争结合,表面活性剂对两者间的竞争结合有增敏作用;并利用荧光光谱直接考察了K15[Ce(BW11O39)2]·17H2O与BSA的相互作用及其对BSA构象的影响.

  13. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  14. Standard PWR for Italy

    Energy Technology Data Exchange (ETDEWEB)

    Negroni, A.; Velona, F. (Ente Nazionale per l' Energia Elettrica, Rome (Italy))

    1983-03-01

    A description is given of the general design for the standard PWR which will be used in the seven to eight nuclear power stations provided for in the Italian national energy plan. Special features to meet Italian conditions include double containment and a common foundation mat for the reactor, auxiliary and fuel buildings.

  15. Sjak, mestre og skibsbyggeri. Arbejdsliv på B&W 1945-1996, [sammen med Torkil Adsersen

    DEFF Research Database (Denmark)

    Nielsen, Niels Jul

    2005-01-01

    Lukningen af B&W Skibsværft i 1996 markerede afslutningen på en epoke i dansk industrihistorie. Værtet på Refshaleøen var kendt som en af de toneangivende virksomheder i Danmark over en periode på næsten 150 år. Både med hensyn til størrelse, teknisk innovation og fagpoiltisk engagement. Mindre f...

  16. Experimental devices in the osiris reactor to study effects of radiations on fusion reactor materials

    Science.gov (United States)

    Lefevre, F.; Thevenot, G.

    1986-11-01

    Within the framework of the Technology Research Program on controlled fusion initiated by the European Communities, the Services des Piles de Saclay (SPS) of Commissariat à l'Energie Atomique (CEA) have been requested to perform some necessary experiments to study the irradiation behaviour of materials which are possible candidates for controlled fusion reactors. This paper describes the devices, generally adapted from a standard model "The COLIBRI", which allow one to carry out, in the OSIRIS reactor, irradiations on the three great families of fusion reactor materials: - lithium containing materials of breeding blanket for in-situ tritium production, - protection materials, and - structural materials.

  17. Experimental devices in the OSIRIS reactor to study effects of radiations on fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Lefevre, F.; Thevenot, G.

    Within the framework of the Technology Research Program on controlled fusion initiated by the European Communities, the Services des Piles de Saclay (SPS) of Commissariat a l'Energie Atomique (CEA) have been requested to perform some necessary experiments to study the irradiation behaviour of materials which are possible candidates for controlled fusion reactors. This paper describes the devices, generally adapted from a standard model The COLIBRI, which allow one to carry out, in the OSIRIS reactor, irradiations on the three great families of fusion reactor materials: Lithium containing materials of breeding blanket for in-situ tritium production, protection materials, and structural materials.

  18. Controllable two-scale network architecture and enhanced mechanical properties of (Ti5Si3+TiBw)/Ti6Al4V composites

    Science.gov (United States)

    Jiao, Y.; Huang, L. J.; Duan, T. B.; Wei, S. L.; Kaveendran, B.; Geng, L.

    2016-09-01

    Novel Ti6Al4V alloy matrix composites with a controllable two-scale network architecture were successfully fabricated by reaction hot pressing (RHP). TiB whiskers (TiBw) were in-situ synthesized around the Ti6Al4V matrix particles, and formed the first-scale network structure (FSNS). Ti5Si3 needles (Ti5Si3) precipitated in the β phase around the equiaxed α phase, and formed the secondary-scale network structure (SSNS). This resulted in increased deformation compatibility accompanied with enhanced mechanical properties. Apart from the reinforcement distribution and the volume fraction, the ratio between Ti5Si3 and TiBw fraction were controlled. The prepared (Ti5Si3 + TiBw)/Ti6Al4V composites showed higher tensile strength and ductility than the composites with a one-scale microstructure, and superior wear resistance over the Ti6Al4V alloy under dry sliding wear conditions at room temperature.

  19. SNTP program reactor design

    Science.gov (United States)

    Walton, Lewis A.; Sapyta, Joseph J.

    1993-06-01

    The Space Nuclear Thermal Propulsion (SNTP) program is evaluating the feasibility of a particle bed reactor for a high-performance nuclear thermal rocket engine. Reactors operating between 500 MW and 2,000 MW will produce engine thrusts ranging from 20,000 pounds to 80,000 pounds. The optimum reactor arrangement depends on the power level desired and the intended application. The key components of the reactor have been developed and are being tested. Flow-to-power matching considerations dominate the thermal-hydraulic design of the reactor. Optimal propellant management during decay heat cooling requires a three-pronged approach. Adequate computational methods exist to perform the neutronics analysis of the reactor core. These methods have been benchmarked to critical experiment data.

  20. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  1. Hybrid reactors. [Fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  2. Reactor pulse repeatability studies at the annular core research reactor

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

  3. Multi purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: vkrain@magnum.barc.ernet.in; Sasidharan, K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, Samiran [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, Tej [Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-04-15

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.

  4. INVAP's Research Reactor Designs

    Directory of Open Access Journals (Sweden)

    Eduardo Villarino

    2011-01-01

    Full Text Available INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors.

  5. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1997-04-04

    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  6. Assessing Pretreatment Reactor Scaling Through Empirical Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lischeske, James J.; Crawford, Nathan C.; Kuhn, Erik; Nagle, Nicholas J.; Schell, Daniel J.; Tucker, Melvin P.; McMillan, James D.; Wolfrum, Edward J.

    2016-12-01

    Pretreatment is a critical step in the biochemical conversion of lignocellulosic biomass to fuels and chemicals. Due to the complexity of the physicochemical transformations involved, predictively scaling up technology from bench- to pilot-scale is difficult. This study examines how pretreatment effectiveness under nominally similar reaction conditions is influenced by pretreatment reactor design and scale using four different pretreatment reaction systems ranging from a 3 g batch reactor to a 10 dry-ton/d continuous reactor. The reactor systems examined were an Automated Solvent Extractor (ASE), Steam Explosion Reactor (SER), ZipperClave(R) reactor (ZCR), and Large Continuous Horizontal-Screw Reactor (LHR). To our knowledge, this is the first such study performed on pretreatment reactors across a range of reaction conditions (time and temperature) and at different reactor scales. The comparative pretreatment performance results obtained for each reactor system were used to develop response surface models for total xylose yield after pretreatment and total sugar yield after pretreatment followed by enzymatic hydrolysis. Near- and very-near-optimal regions were defined as the set of conditions that the model identified as producing yields within one and two standard deviations of the optimum yield. Optimal conditions identified in the smallest-scale system (the ASE) were within the near-optimal region of the largest scale reactor system evaluated. A reaction severity factor modeling approach was shown to inadequately describe the optimal conditions in the ASE, incorrectly identifying a large set of sub-optimal conditions (as defined by the RSM) as optimal. The maximum total sugar yields for the ASE and LHR were 95%, while 89% was the optimum observed in the ZipperClave. The optimum condition identified using the automated and less costly to operate ASE system was within the very-near-optimal space for the total xylose yield of both the ZCR and the LHR, and was

  7. MOX in reactors: present and future

    Energy Technology Data Exchange (ETDEWEB)

    Arslan, Marc; Gros, Jean Pierre [AREVA NC - 33 rue La Fayette, 75009 Paris (France); Niquille, Aurelie; Marincic, Alexis [AREVA NP - Tour AREVA, 1 Place Jean Millier 92084 Paris La Defense (France)

    2010-07-01

    In Europe, MOX fuel has been supplied by AREVA for more than 30 years, to 36 reactors: 21 in France, 10 in Germany, 3 in Switzerland, 2 in Belgium. For the present and future, recycling is compulsory in the frame of sustainable development of nuclear energy. By 2030 the overall volume of used fuel will reach about 400 000 t worldwide. Their plutonium and uranium content represents a huge resource of energy to recycle. That is the reason why, the European Utilities issued an EUR (European Utilities Requirement) demanding new builds reactors to be able of using MOX Fuel Assemblies in up to 50 % of the core. AREVA GEN3+ reactors, like EPR{sup TM} or ATMEA{sup TM} designed with MHI partnership, are designed to answer any utility need of MOX recycling. The example of the EPR{sup TM} reactor operated with 100 % MOX core optimized for MOX recycling will be presented. A standard EPR{sup TM} can be operated with 100 % MOX core using an advanced homogeneous MOX (single Pu content) with highly improved performances (burn-up and Cycle length). The adaptations needed and the main operating and safety reactor features will be presented. AREVA offers the utilities throughout the world, fuel supply (UO{sub 2}, ERU, MOX), and reactors designed with all the needed capability for recycling. For each country and each utility, an adapted global solution, competitive and non proliferant can be proposed. (authors)

  8. Simulation of Reactors for Antineutrino Experiments Using DRAGON

    CERN Document Server

    Winslow, L

    2011-01-01

    From the discovery of the neutrino to the precision neutrino oscillation measurements in KamLAND, nuclear reactors have proven to be an important source of antineutrinos. As their power and our knowledge of neutrino physics has increased, more sensitive measurements have become possible. The next generation of reactor antineutrino experiments require more detailed simulations of the reactor core. Many of the reactor simulation codes are proprietary which makes detailed studies difficult. Here we present the results of the open source DRAGON code and compare it to other industry standards for reactor modeling. We use published data from the Takahama reactor to determine the quality of the simulations. The propagation of the uncertainty to the antineutrino flux is also discussed.

  9. Space Nuclear Reactor Engineering

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-06

    We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

  10. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  11. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  12. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  13. Preparation Before Signature of Upgrade of Algeria Heavy Water Research Reactor Contract

    Institute of Scientific and Technical Information of China (English)

    LI; Song; ZAN; Huai-qi; XU; Qi-guo; JIA; Yu-wen

    2012-01-01

    <正>Algeria heavy water research reactor (Birine) is a multiple-purpose research reactor, which was constructed with the help of China more than 20 years ago. By request of Algeria, China will upgrade the research reactor; so as to improve the status of current reactor such as equipment ageing, shortage of spare parts, several systems do not meet requirements of current standards and criteria etc.

  14. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  15. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G [Champaign, IL; Mitrovski, Svetlana M [Urbana, IL

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  16. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  17. Search for pair-produced vector-like top quark partners decaying to bW in the fully hadronic channel using jet substructure at 8 TeV

    CERN Document Server

    CMS Collaboration

    1900-01-01

    We present a search for pair-produced vector-like top quark partners in 19.7~fb$^{-1}$ of $\\rm pp$ collisions at $\\sqrt{s} = 8$ TeV collected with the CMS detector at the LHC in 2012. The search is optimized for the decay mode $\\rm T \\to bW$ in the fully hadronic final state. Due to the large mass of the $\\rm T$ quark, its decay products are significantly boosted, requiring the use of specialized jet substructure techniques. We select events with four jets, of which two are assigned to the hadronic W boson decay and the others are consistent with the hadronization of a b quark. We see no excess of events above the expected standard model backgrounds and set limits on the $\\rm T$ pair-production cross section. $\\rm T$ quarks with masses below 705 GeV/$c^2$ are excluded at the 95\\% confidence level. We also set limits on the $\\rm T \\to tZ$ and $\\rm T \\to tH$ decays.

  18. Iodine chemistry in a reactor regulation

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A. [Nuclear Regulatory Commission, Washington, DC (United States). Advisory Committee on Reactor Safeguards

    1996-12-01

    Radioactive iodine has always been an important consideration in the regulation of nuclear power reactors to assure the health and safety of the public. Regulators adopted conservatively bounding predictions of iodine behavior in the earliest days of the development of nuclear power because there was so little known about either accidents or the chemistry of iodine. Today there is a flood of new information and understanding of the chemistry of iodine under reactor accident conditions. This paper offers some thoughts on how the community of scientists engaged in the study of iodine chemistry can present the results of their work so that it is more immediately adopted by the regulator. It is suggested that the scientific community consider the concept of consensus standards so effectively used within the engineering community to define the status of the study of radioactive iodine chemistry for reactor safety. (author) 9 refs.

  19. Development of mechanical design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Keun Bae; Choi, Suhn; Kim, Kang Soo; Kim, Tae Wan; Jeong, Kyeong Hoon; Lee, Gyu Mahn

    1999-03-01

    While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose application such as small capacity power generation, co-generation and sea water desalination. With this in mind, an integral reactor SMART is under development. Design concepts, system layout and types of equipment of integral reactor are significantly different from those of loop type reactor. Conceptual design development of mechanical structures of integral reactor SMART is completed through the first stage of the project. Efforts were endeavored for the establishment of design basis and evaluation of applicable codes and standards. Design and functional requirements of major structural components were setup, and three dimensional structural modelling of SMART reactor vessel assembly was prepared. Also, maintenance and repair scheme as well as preliminary fabricability evaluation were carried out. Since small integral reactor technology includes sensitive technologies and know-how's, it is hard to achieve systematic and comprehensive technology transfer from nuclear-advanced countries. Thus, it is necessary to develop the related design technology and to verify the adopted methodologies through test and experiments in order to assure the structural integrity of reactor system. (author)

  20. Technology Selection for Offshore Underwater Small Modular Reactors

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1 a lead–bismuth fast reactor based on the Russian SVBR-100; (2 a novel organic cooled reactor; (3 an innovative superheated water reactor; (4 a boiling water reactor based on Toshiba's LSBWR; and (5 an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80% with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

  1. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  2. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  3. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  4. Helias reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Beidler, C.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Grieger, G. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Harmeyer, E. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Kisslinger, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Karulin, N. [Nuclear Fusion Institute, Moscow (Russian Federation); Maurer, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany); Nuehrenberg, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Rau, F. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Sapper, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Wobig, H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1995-10-01

    The present status of Helias reactor studies is characterised by the identification and investigation of specific issues which result from the particular properties of this type of stellarator. On the technical side these are issues related to the coil system, while physics studies have concentrated on confinement, alpha-particle behaviour and ignition conditions. The usual assumptions have been made in those fields which are common to all toroidal fusion reactors: blanket and shield, refuelling and exhaust, safety and economic aspects. For blanket and shield sufficient space has been provided, a detailed concept will be developed in future. To date more emphasis has been placed on scoping and parameter studies as opposed to fixing a specific set of parameters and providing a detailed point study. One result of the Helias reactor studies is that physical dimensions are on the same order as those of tokamak reactors. However, it should be noticed that this comparison is difficult in view of the large spectrum of tokamak reactors ranging from a small reactor like Aries, to a large device such as SEAFP. The notion that the large aspect ratio of 10 or more in Helias configurations also leads to large reactors is misleading, since the large major radius of 22 m is compensated by the average plasma radius of 1.8 m and the average coil radius of 5 m. The plasma volume of 1400 m{sup 3} is about the same as the ITER reactor and the magnetic energy of the coil system is about the same or even slightly smaller than envisaged in ITER. (orig.)

  5. Future Reactor Experiments

    OpenAIRE

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measure...

  6. Reactor Neutrino Experiments

    OpenAIRE

    Cao, Jun

    2007-01-01

    Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measu...

  7. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  8. Standard interface file handbook

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, A.; Huria, H.C. (Cincinnati Univ., OH (United States))

    1992-10-01

    This handbook documents many of the standard interface file formats that have been adopted by the US Department of Energy to facilitate communications between and portability of, various large reactor physics and radiation transport software packages. The emphasis is on those files needed for use of the VENTURE/PC diffusion-depletion code system. File structures, contents and some practical advice on use of the various files are provided.

  9. Moon base reactor system

    Science.gov (United States)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  10. 统一螺纹量规标准研究及其在AP1000反应堆压力容器主螺栓制造中的应用%Study on Standard of Gages for Unified Screw Threads and Application on AP1000 Reactor Pressure Vessel Main Stud Manufacturing

    Institute of Scientific and Technical Information of China (English)

    易飞; 杨杰

    2015-01-01

    研究了按ASME B1. 2标准规定的统一螺纹量规(通、 止规)的设计原则, 介绍了统一螺纹量规功能、 公差设置、 量规尺寸确定等, 并针对AP1000反应堆压力容器主螺栓螺纹制造和检测提出建议, 为业内及其他统一螺纹制造和检测提供参考.%This paper studies the design principles, function and determination of tolerances and dimensions of gages for unified inch screw thread according to ASME B1. 2 standard, based on which some suggestions are given for AP1000 reactor pressure vessel main studs manufacturing and testing, as reference for in industry sectors.

  11. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  12. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  13. The incumbent German power companies in a changing environment. A comparison of E.ON, RWE, EnBW and Vattenfall from 1998 to 2013

    Energy Technology Data Exchange (ETDEWEB)

    Kungl, Gregor

    2014-07-01

    This paper examines the actions and strategies of Germany's leading energy companies - E.ON, RWE, EnBW and Vattenfall - in the light of a changing regulatory framework and other circumstances. The liberalization of the German electricity market, measures to promote renewable energies, market developments as well as exogenous shocks such as the Fukushima nuclear disaster and the fiscal crisis all had far-reaching consequences for these companies. A comparative analysis of these companies from 1998 to 2013 shows their development from thriving growth at the start of liberalization up to the current state of crisis. Conducted with a focus on the context of the Energiewende - Germany's commitment to shift towards sustainable energy production - this article contributes to the current debate on the sustainable transformation of energy supply. The theory of strategic action fields by Fligstein and McAdam serves as a theoretical framework.

  14. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  15. Association of KIR3DL1/S1 and HLA-Bw4 with CD4 T cell counts in HIV-infected Mexican mestizos.

    Science.gov (United States)

    Hernández-Ramírez, Daniel; Esparza-Pérez, Mario A; Ramirez-Garcialuna, José L; Arguello, J Rafael; Mandeville, Peter B; Noyola, Daniel E; García-Sepúlveda, Christian A

    2015-08-01

    Certain genotypic combinations of killer-cell immunoglobulin-like receptors (KIR) and human leukocyte antigens (HLA) have been associated with favourable outcomes after exposure to human immunodeficiency virus in Caucasoid and African populations. Human immunodeficiency virus (HIV) infection is characterized by a rapid exhaustion of CD4 cells, which results in impaired cellular immunity. During this early phase of infection, it is thought that the natural killer (NK) cells represent the main effector arm of the host immune response to HIV. This study investigates whether KIR and HLA factors are associated to CD4 T cell numbers after HIV infection in Mexican mestizos as assessed at the time of initial medical evaluation and subsequent clinical follow-up. KIR and HLA-B gene carrier frequency differences were compared between groups of patients stratified by CD4 T cell numbers as assessed during their first medical evaluation (a point in time at which all patients were anti-retroviral therapy naïve). In addition, the influence that these genetic factors have on averaged historical CD4 cell counts in patients subjected to follow-up (mostly therapy-experienced) was also evaluated. Our results suggest a protective role for the HLA-Bw4 and KIR3D + Bw4 combination in both therapy-naïve and therapy-experienced patients. This report furthers our understanding on the way that immune genes modulate HIV disease progression in less-studied human populations such as the Mexican mestizos with a special focus on CD4 T cell number and behaviour.

  16. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system; Transport des radionucleides dans le circuit primaire d`un REP: comparaison des codes MAAP et ECART

    Energy Technology Data Exchange (ETDEWEB)

    Hervouet, C.; Ranval, W. [Electricite de France (EDF), 92 - Clamart (France); Parozzi, F.; Eusebi, M. [Ente Nazionale per l`Energia Elettrica, Rome (Italy)

    1996-04-01

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO{sub 2} and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs.

  17. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi; Iida, Masaaki; Moriki, Yasuyuki

    1994-10-18

    A reactor core is divided into a plurality of coolants flowrate regions, and electromagnetic pumps exclusively used for each of the flowrate regions are disposed to distribute coolants flowrates in the reactor core. Further, the flowrate of each of the electromagnetic pumps is automatically controlled depending on signals from a temperature detector disposed at the exit of the reactor core, so that the flowrate of the region can be controlled optimally depending on the burning of reactor core fuels. Then, the electromagnetic pumps disposed for every divided region are controlled respectively, so that the coolants flowrate distribution suitable to each of the regions can be attained. Margin for fuel design is decreased, fuels are used effectively, as well as an operation efficiency can be improved. Moreover, since the electromagnetic pump has less flow resistance compared with a mechanical type pump, and flow resistance of the reactor core flowrate control mechanism is eliminated, greater circulating flowrate can be ensured after occurrence of accident in a natural convection using a buoyancy of coolants utilizable for after-heat removal as a driving force. (N.H.).

  18. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported.

  19. Researches on a reactor core in heavy ion inertial fusion

    CERN Document Server

    Kondo, S; Iinuma, T; Kubo, K; Kato, H; Kawata, S; Ogoyski, A I

    2016-01-01

    In this paper a study on a fusion reactor core is presented in heavy ion inertial fusion (HIF), including the heavy ion beam (HIB) transport in a fusion reactor, a HIB interaction with a background gas, reactor cavity gas dynamics, the reactor gas backflow to the beam lines, and a HIB fusion reactor design. The HIB has remarkable preferable features to release the fusion energy in inertial fusion: in particle accelerators HIBs are generated with a high driver efficiency of ~30-40%, and the HIB ions deposit their energy inside of materials. Therefore, a requirement for the fusion target energy gain is relatively low, that would be ~50 to operate a HIF fusion reactor with a standard energy output of 1GW of electricity. In a fusion reactor the HIB charge neutralization is needed for a ballistic HIB transport. Multiple mechanical shutters would be installed at each HIB port at the reactor wall to stop the blast waves and the chamber gas backflow, so that the accelerator final elements would be protected from the ...

  20. PCCF flow analysis -- DR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Calkin, J.F.

    1961-04-26

    This report contains an analysis of PCCF tube flow and Panellit pressure relations at DR reactor. Supply curves are presented at front header pressures from 480 to 600 psig using cold water and the standard 0.236 inch orifice with taper down stream and the pigtail valve (plug or ball) open. Demand curves are presented for slug column lengths of 200 inches to 400 inches using 1.44 inch O.D. solid poison pieces (either Al or Pb-Cd) and cold water with a rear header pressure of 50 psig. Figure 1 is a graph of Panellit pressure vs. flow with the above supply and demand curves and clearly shows the effect of front header pressure and charge length on flow.

  1. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  2. Technology Selection for Offshore Underwater Small Modular Reactors

    OpenAIRE

    Koroush Shirvan; Ronald Ballinger; Jacopo Buongiorno; Charles Forsberg; Mujid Kazimi; Neil Todreas

    2016-01-01

    This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the ...

  3. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  4. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  5. Operation of Reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    3.1 Annual Report of SPR Operation Chu Shaochu Having overseen by National Nuclear Safety Administration and specialists, the reactor restarted up successfully after Safety renovation on April 16, 1996. In August 1996 the normal operation of SPR was approved by the authorities of Naitonal Nuclear Safety Administration. 1 Operation status In 1996, the reactor operated safely for 40 d and the energy released was about 137.3 MW·d. The operation status of SPR is shown in table 1. The reactor started up to higher power (power more than 1 MW) and lower power (for physics experiments) 4 times and 14 times respectively. Measurement of control rod efficiency and other measurement tasks were 2 times and 5 times respectively.

  6. An Overview of Reactor Concepts, a Survey of Reactor Designs.

    Science.gov (United States)

    1985-02-01

    Public Affairs Office and is releasaole to the National Technical Information Services (NTIS). At NTIS, it will be available to the general public...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept...reactor core at the top and discharged at the Dotton while the reactor is in operation. The discharged fuel can then b inspected to see if it can De used

  7. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  8. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  9. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  10. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  11. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  12. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  13. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  14. Chromatographic and Related Reactors.

    Science.gov (United States)

    1988-01-07

    special information about effects of surface heteroge- neity in the methanation reaction. Studies of an efficient multicolumn assembly for measuring...of organic basic catalysts such as pyridine and 4-methylpicoline. It was demonstrated that the chromatographic reactor gave special information about...Programmed Reaction to obtain special information about surface heterogeneity in the methanation reaction. Advantages of stopped flow over steady state

  15. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  16. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  17. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  18. The First Reactor.

    Science.gov (United States)

    Department of Energy, Washington, DC.

    On December 2, 1942, in a racquet court underneath the West Stands of Stagg Field at the University of Chicago, a team of scientists led by Enrico Fermi created the first controlled, self-sustaining nuclear chain reaction. This updated and revised story of the first reactor (or "pile") is based on postwar interviews (as told to Corbin…

  19. MULTISTAGE FLUIDIZED BED REACTOR

    Science.gov (United States)

    Jonke, A.A.; Graae, J.E.A.; Levitz, N.M.

    1959-11-01

    A multistage fluidized bed reactor is described in which each of a number of stages is arranged with respect to an associated baffle so that a fluidizing gas flows upward and a granular solid downward through the stages and baffles, whereas the granular solid stopsflowing downward when the flow of fluidizing gas is shut off.

  20. Brazilian multipurpose reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U{sub 3}Si{sub 2} dispersion-type Al having a density of up to 3.5 gU/cm{sup 3} and enrichment of 19.75% by weight of {sup 23{sup 5}}U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive

  1. Detection of anomalous reactor activity using antineutrino count evolution over the course of a reactor cycle

    Science.gov (United States)

    Bulaevskaya, Vera; Bernstein, Adam

    2011-06-01

    This paper analyzes the sensitivity of antineutrino count rate measurements to changes in the fissile content of civil power reactors. Such measurements may be useful in IAEA reactor safeguards applications. We introduce a hypothesis testing procedure to identify statistically significant differences between the antineutrino count rate evolution of a standard "baseline" fuel cycle and that of an anomalous cycle, in which plutonium is removed and replaced with an equivalent fissile worth of uranium. The test would allow an inspector to detect anomalous reactor activity, or to positively confirm that the reactor is operating in a manner consistent with its declared fuel inventory and power level. We show that with a reasonable choice of detector parameters, the test can detect replacement of 82 kg of plutonium in 90 days with 95% probability, while controlling the false positive rate at 5%. We show that some improvement on this level of sensitivity may be obtained by various means, including use of the method in conjunction with existing reactor safeguards methods. We also identify a necessary and sufficient minimum daily antineutrino count rate and a maximum tolerable background rate to achieve the quoted sensitivity, and list examples of detectors in which such rates have been attained.

  2. Modeling Chemical Reactors I: Quiescent Reactors

    CERN Document Server

    Michoski, C E; Schmitz, P G

    2010-01-01

    We introduce a fully generalized quiescent chemical reactor system in arbitrary space $\\vdim =1,2$ or 3, with $n\\in\\mathbb{N}$ chemical constituents $\\alpha_{i}$, where the character of the numerical solution is strongly determined by the relative scaling between the local reactivity of species $\\alpha_{i}$ and the local functional diffusivity $\\mathscr{D}_{ij}(\\alpha)$ of the reaction mixture. We develop an operator time-splitting predictor multi-corrector RK--LDG scheme, and utilize $hp$-adaptivity relying only on the entropy $\\mathscr{S}_{\\mathfrak{R}}$ of the reactive system $\\mathfrak{R}$. This condition preserves these bounded nonlinear entropy functionals as a necessarily enforced stability condition on the coupled system. We apply this scheme to a number of application problems in chemical kinetics; including a difficult classical problem arising in nonequilibrium thermodynamics known as the Belousov-Zhabotinskii reaction where we utilize a concentration-dependent diffusivity tensor $\\mathscr{D}_{ij}(...

  3. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  4. Antineutrino emission and gamma background characteristics from a thermal research reactor

    CERN Document Server

    Bui, V M; Fallot, M; Communeau, V; Cormon, S; Estienne, M; Lenoir, M; Peuvrel, N; Shiba, T; Cucoanes, A S; Elnimr, M; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Thiolliere, N; Yermia, F; Zakari-Issoufou, A -A

    2016-01-01

    The detailed understanding of the antineutrino emission from research reactors is mandatory for any high sensitivity experiments either for fundamental or applied neutrino physics, as well as a good control of the gamma and neutron backgrounds induced by the reactor operation. In this article, the antineutrino emission associated to a thermal research reactor: the OSIRIS reactor located in Saclay, France, is computed in a first part. The calculation is performed with the summation method, which sums all the contributions of the beta decay branches of the fission products, coupled for the first time with a complete core model of the OSIRIS reactor core. The MCNP Utility for Reactor Evolution code was used, allowing to take into account the contributions of all beta decayers in-core. This calculation is representative of the isotopic contributions to the antineutrino flux which can be found at research reactors with a standard 19.75\\% enrichment in $^{235}$U. In addition, the required off-equilibrium correction...

  5. Stress corrosion (Astm G30-90 standard) in 08x18H10T stainless steel of nuclear fuel storage pool in WWER reactors; Corrosion bajo esfuerzo (Norma ASTM G30-90) en acero inoxidable 08x18H10T de piscinas de almacenamiento de combustible nuclear en reactores V.V.E.R

    Energy Technology Data Exchange (ETDEWEB)

    Herrera, V.; Zamora R, L. [Centro de Estudios Aplicados al Desarrollo Nuclear (Cuba)

    1997-07-01

    At the water storage of the irradiated nuclear fuel has been an important factor in its management. The actual pools have its walls covered with inoxidable steel and heat exchangers to dissipate the residual heat from fuel. It is essential to control the water purity to eliminate those conditions which aid to the corrosion process in fuel and at related components. The steel used in this research was obtained from an austenitic inoxidizable steel standardized with titanium 08x18H10T (Type 321) similar to one of the two steel coatings used to cover walls and the pools floor. the test consisted in the specimen deformation through an U ply according to the Astm G30-90 standard. The exposition of the deformed specimen it was realized in simulated conditions to the chemical regime used in pools. (Author)

  6. Reactor monitoring using antineutrino detectors

    Science.gov (United States)

    Bowden, N. S.

    2011-08-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

  7. Reactor vessel support system. [LMFBR

    Science.gov (United States)

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  8. Importance of dam BW change and calf birth weight in double-muscled Belgian Blue cattle and its relationship with parity and calving interval.

    Science.gov (United States)

    Fiems, L O; Ampe, B

    2015-01-01

    Factors affecting calving interval (CI) in double-muscled Belgian Blue (DMBB) beef cows were investigated with regard to the BW yield (BWY) of the cow-calf pair, using 834 CI records from 386 females with parities 1 to 6. The effect of parity and CI on BWY was also studied. Cow-calf pair BWY was defined as calf birth weight plus dam BWY per CI. CI (mean±s.e.: 404±1.9 days) was affected by parity, calving season, suckling and calf birth weight/dam weight. Primiparous cows had a shorter CI than cows with three or more calvings (Pdams had longer CIs than non-suckling dams. There were interactions (Pdam weight of 6% to 10% resulted in shorter CI than a ratio of Dam contribution to cow-calf pair BWY was larger than calf birth weight in first- and second-calf cows, and increased with increasing CI. Dam contribution to cow-calf pair BWY was smaller than calf birth weight in older cows, varying from 0.2 to 1.0 depending on CI. A short CI is advised for DMBB cows because of a larger BWY and more efficient nutrient utilisation.

  9. Tool wear during high speed turning in situ TiCp/TiBw hybrid reinforced Ti-6Al-4V matrix composite

    Directory of Open Access Journals (Sweden)

    Ge Yingfei

    2016-10-01

    Full Text Available Chipping, adhesive wear, abrasive wear and crater wear are prevalent for both the polycrystalline diamond (PCD and the carbide tools during high speed turning of TiCp/TiBw hybrid reinforced Ti-6Al-4V (TC4 matrix composite (TMCs. The combined effects of abrasive wear and diffusion wear caused the big crater on PCD and carbide tool rake face. Compared to the PCD, bigger size of crater was found on the carbide tool due to much higher cutting temperature and the violent chemical reaction between the Ti element in the workpiece and the WC in the tool. However, the marks of the abrasive wear looked much slighter or even could not be observed on the carbide tool especially when low levels of cutting parameters were used, which attributes to much lower hardness and smaller size of WC combined with more significant chemical degradation of carbide. When cutting TC4 using PCD tool, notch wear was the most significant wear pattern which was not found when cutting the TMCs. However, chipping, adhesive wear and crater wear were much milder when compared to the cutting of titanium matrix composite. Due to the absence of abrasive wear when cutting TC4, the generated titanium carbide on the PCD protected the tool from fast wear, which caused that the tool life for TC4 was 6–10 times longer than that for TMCs.

  10. Tool wear during high speed turning in situ TiCp/TiBw hybrid reinforced Ti-6Al-4V matrix composite

    Institute of Scientific and Technical Information of China (English)

    Ge Yingfei; Xu Jiuhua; Huan Haixiang

    2016-01-01

    Chipping, adhesive wear, abrasive wear and crater wear are prevalent for both the poly-crystalline diamond (PCD) and the carbide tools during high speed turning of TiCp/TiBw hybrid reinforced Ti-6Al-4V (TC4) matrix composite (TMCs). The combined effects of abrasive wear and diffusion wear caused the big crater on PCD and carbide tool rake face. Compared to the PCD, bigger size of crater was found on the carbide tool due to much higher cutting temperature and the violent chemical reaction between the Ti element in the workpiece and the WC in the tool. However, the marks of the abrasive wear looked much slighter or even could not be observed on the carbide tool especially when low levels of cutting parameters were used, which attributes to much lower hardness and smaller size of WC combined with more significant chemical degradation of car-bide. When cutting TC4 using PCD tool, notch wear was the most significant wear pattern which was not found when cutting the TMCs. However, chipping, adhesive wear and crater wear were much milder when compared to the cutting of titanium matrix composite. Due to the absence of abrasive wear when cutting TC4, the generated titanium carbide on the PCD protected the tool from fast wear, which caused that the tool life for TC4 was 6–10 times longer than that for TMCs.

  11. The properties of the host galaxy and the immediate environment of GRB 980425 / SN 1998bw from the multi-wavelength spectral energy distribution

    CERN Document Server

    Michałowski, Michał J; Malesani, Daniele; Michałowski, Tadeusz; Cerón, José María Castro; Reinfrank, Robert F; Garrett, Michael A; Fynbo, Johan P U; Watson, Darach J; Jørgensen, Uffe G

    2008-01-01

    We present an analysis of the spectral energy distribution (SED) of the galaxy ESO 184-G82, the host of the closest known long gamma-ray burst (GRB) 980425 and its associated supernova SN 1998bw. We use our observations obtained at the Australia Telescope Compact Array (the third >3 sigma radio detection of a GRB host) as well as archival infrared (IR) and ultraviolet (UV) observations to estimate its star formation state. We find that ESO 184-G82 has a UV star formation rate (SFR) and stellar mass consistent with the population of cosmological GRB hosts and of local dwarf galaxies. It has however a higher specific SFR (per unit stellar mass) and lower molecular gas-to-dust ratio than luminous spiral galaxies. The mass of ESO 184-G82 is dominated by an older stellar population in contrast to the majority of GRB hosts. The Wolf-Rayet region ~800 pc from the supernova site experienced a starburst episode during which the majority of its stellar population was built up. Unlike that of the entire galaxy, its SED ...

  12. 葡甘聚糖/蜂蜡复合膜工艺优化研究%Study on optimization of konjac glucomannan (KGM)-beeswax (BW) composite membrane

    Institute of Scientific and Technical Information of China (English)

    赖明耀; 林好; 汪秀妹; 冯瑞; 兰润; 庞杰

    2013-01-01

    以魔芋葡甘聚糖(KGM)和蜂蜡(BW)为原料制备复合膜,以其拉伸强度为考察指标,通过单因素试验,分别研究KGM/BW配比、甘油添加量、烘干温度对复合膜力学性能的影响,从而得出其最适合的理论工艺,再采用响应面法优化其工艺条件.研究结果表明,KGM/BW配比、甘油添加量、烘干温度对复合膜的拉伸强度均有显著影响,且最优条件为KGM/BW配比6.14∶1、甘油添加量0.42%、烘干温度70.71℃,复合膜的拉伸强度理论预测值为42.156 0 MPa,验证值为41.78 MPa.

  13. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  14. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  15. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  16. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    OpenAIRE

    Sungjoo Lee; Byungun Yoon; Juneseuk Shin

    2016-01-01

    We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indic...

  17. LASIP-III, a generalized processor for standard interface files. [For creating binary files from BCD input data and printing binary file data in BCD format (devised for fast reactor physics codes)

    Energy Technology Data Exchange (ETDEWEB)

    Bosler, G.E.; O' Dell, R.D.; Resnik, W.M.

    1976-03-01

    The LASIP-III code was developed for processing Version III standard interface data files which have been specified by the Committee on Computer Code Coordination. This processor performs two distinct tasks, namely, transforming free-field format, BCD data into well-defined binary files and providing for printing and punching data in the binary files. While LASIP-III is exported as a complete free-standing code package, techniques are described for easily separating the processor into two modules, viz., one for creating the binary files and one for printing the files. The two modules can be separated into free-standing codes or they can be incorporated into other codes. Also, the LASIP-III code can be easily expanded for processing additional files, and procedures are described for such an expansion. 2 figures, 8 tables.

  18. ABO血型系统B101-O02等位基因杂交导致的Bw亚型%Hybridization of B101 and O02 alleles responsible for Bw subgroup in ABO blood group

    Institute of Scientific and Technical Information of China (English)

    李育; 金红; 陈秉宇

    2010-01-01

    Objective Bw subgroup of ABO blood group system was investigated to reveal its molecular genetic basis in Chinese Han population.Methods The Abw subgroup of three relatives were identified by standard blood group serological techniques.The enhancer,promoter and exon 1 to exon 7including flanking intron sequences of ABO gene were amplified by polymerase chain reaction.Direct sequencing was then performed on the gel-purified PCR products.Afterwards the exon 7 with polymorphic sites were cloned into pcDNA3.1(-)vector and transformed into DH5α to carry out a haploid analysis.Results According to the direct sequencing analysis,sequence characteristics of the 3 tested subjects were found specific to partial sequence of A102,B101 and O02 alleles,thought the genotypes could not be confirmed directly.The haploid analysis affirmed that one allele is A102 while the other is a B101-O02 hybrid,which harbored 646T>A,657T>C and 681G>A variants compared with B101 allele.This hybrid allele was not detected in a 110 randomly selected samples.Conclusion The B101-O02 hybrid allele may account for the Bw phenotype and its molecular genetic basis.%目的 研究中国汉族个体红细胞ABO血型Bw亚型的分子遗传背景.方法 通过标准血型血清学方法 鉴定了1个家庭3例ABw亚型,PCR扩增样本基因组DNA的ABO基因增强子、启动子和外显子1~7及侧翼内含子序列,PCR产物经割胶纯化后直接测序,并将含有多态性位点的外显子7克隆到pcDNA3.1(-)质粒,转化DH5α后进行单倍体序列分析.结果 3例ABw亚型的直接序列分析发现其ABO基因均有A102、B101和002等3种等位基因部分序列特征,但无法直接确定其基因型.单倍体序列分析表明,其中1个等位基因为A102,另1个为B101-O02杂交等位基因,表现为B101等位基因基础上的646T>A,657T>C和681G>A变异.在110份随机样本中未发现该杂交等位基因.结论 B101和O02等位基因杂交可能是导致该Bw亚型的分子机制.

  19. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

  20. Reactor Dosimetry State of the Art 2008

    Science.gov (United States)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    nuclides - 2008 / T. Golashvili -- Oral session 6: Test reactors, accelerators and advanced systems. Neutronic analyses in support of the HFIR beamline modifications and lifetime extension / I. Remec and E. D. Blakeman. Characterization of neutron test facilities at Sandia National Laboratories / D. W. Vehar ... [et al.]. LYRA irradiation experiments: neutron metrology and dosimetry / B. Acosta and L. Debarberis. Calculated neutron and gamma-ray spectra across the prismatic very high temperature reactor core / J. W. Sterbentz. Enhancement of irradiation capability of the experimental fast reactor joyo / S. Maeda ... [et al.]. Neutron spectrum analyses by foil activation method for high-energy proton beams / C. H. Pyeon ... [et al.] -- Oral session 7: Cross sections, nuclear data, damage correlations. Investigation of new reaction cross-section evaluations in order to update and extend the IRDF-2002 reactor dosimetry library / É. M. Zsolnay, H. J. Nolthenius and A. L. Nichols. A novel approach towards DPA calculations / A. Hogenbirk and D. F. Da Cruz. A new ENDFIB-VII.O based multigroup cross-section library for reactor dosimetry / F. A. Alpan and S. L. Anderson. Activities at the NEA for dosimetry applications / H. Henriksson and I. Kodeli. Validation and verification of covariance data from dosimetry reaction cross-section evaluations / S. Badikov. Status of the neutron cross section standards / A. D. Carlson -- Oral session 8: transport calculations. A dosimetry assessment for the core restraint of an advanced gas cooled reactor / D. A. Thornton ... [et al.]. Neutron dosimetry study in the region of the support structure of a VVER-1000 type reactor / G. Borodkin ... [et al.]. SNS moderator poison design and experiment validation of the moderator performance / W. Lu ... [et al.]. Analysis of OSIRIS in-core surveillance dosimetry for GONDOLE steel irradiation program by using TRIPOLI-4 Monte Carlo code / Y. K. Lee and F. Malouch.Reactor dosimetry applications using RAPTOR

  1. Some Movement Mechanisms and Characteristics in Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    Xingtuan Yang

    2014-01-01

    Full Text Available The pebblebed-type high temperature gas-cooled reactor is considered to be one of the promising solutions for generation IV advanced reactors, and the two-region arranged reactor core can enhance its advantages by flattening neutron flux. However, this application is held back by the existence of mixing zone between central and peripheral regions, which results from pebbles’ dispersion motions. In this study, experiments have been carried out to study the dispersion phenomenon, and the variation of dispersion region and radial distribution of pebbles in the specifically shaped flow field are shown. Most importantly, the standard deviation of pebbles’ radial positions in dispersion region, as a quantitative index to describe the size of dispersion region, is gotten through statistical analysis. Besides, discrete element method has been utilized to analyze the parameter influence on dispersion region, and this practice offers some strategies to eliminate or reduce mixing zone in practical reactors.

  2. MEANS FOR COOLING REACTORS

    Science.gov (United States)

    Wheeler, J.A.

    1957-11-01

    A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

  3. Compact fusion reactors

    CERN Document Server

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  4. Reactor Neutrino Spectra

    OpenAIRE

    Hayes, A. C.; Vogel, Petr

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these spectra and their associated uncertainties is crucial for neutrino oscillation studies. The spectra used to date have been determined either by converting measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that make up the spectra, using modern databases as input. The uncertainties in the subdominant corrections to β-decay plague both methods, and we ...

  5. REACTOR MODERATOR STRUCTURE

    Science.gov (United States)

    Greenstreet, B.L.

    1963-12-31

    A system for maintaining the alignment of moderator block structures in reactors is presented. Integral restraining grids are placed between each layer of blocks in the moderator structure, at the top of the uppermost layer, and at the bottom of the lowermost layer. Slots are provided in the top and bottom surfaces of the moderator blocks so as to provide a keying action with the grids. The grids are maintained in alignment by vertical guiding members disposed about their peripheries. (AEC)

  6. Future Scenarios for Fission Based Reactors

    Science.gov (United States)

    David, S.

    2005-04-01

    The coming century will see the exhaustion of standard fossil fuels, coal, gas and oil, which today represent 75% of the world energy production. Moreover, their use will have caused large-scale emission of greenhouse gases (GEG), and induced global climate change. This problem is exacerbated by a growing world energy demand. In this context, nuclear power is the only GEG-free energy source available today capable of responding significantly to this demand. Some scenarios consider a nuclear energy production of around 5 Gtoe in 2050, wich would represent a 20% share of the world energy supply. Present reactors generate energy from the fission of U-235 and require around 200 tons of natural Uranium to produce 1GWe.y of energy, equivalent to the fission of one ton of fissile material. In a scenario of a significant increase in nuclear energy generation, these standard reactors will consume the whole of the world's estimated Uranium reserves in a few decades. However, natural Uranium or Thorium ore, wich are not themselves fissile, can produce a fissile material after a neutron capture ( 239Pu and 233U respectively). In a breeder reactor, the mass of fissile material remains constant, and the fertile ore is the only material to be consumed. In this case, only 1 ton of natural ore is needed to produce 1GWe.y. Thus, the breeding concept allows optimal use of fertile ore and development of sustainable nuclear energy production for several thousand years into the future. Different sustainable nuclear reactor concepts are studied in the international forum "generation IV". Different types of coolant (Na, Pb and He) are studied for fast breeder reactors based on the Uranium cycle. The thermal Thorium cycle requires the use of a liquid fuel, which can be reprocessed online in order to extract the neutron poisons. This paper presents these different sustainable reactors, based on the Uranium or Thorium fuel cycles and will compare the different options in term of fissile

  7. BOILER-SUPERHEATED REACTOR

    Science.gov (United States)

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  8. Detection of Anomalous Reactor Activity Using Antineutrino Count Rate Evolution Over the Course of a Reactor Cycle

    CERN Document Server

    Bulaevskaya, Vera

    2010-01-01

    This paper analyzes the sensitivity of antineutrino count rate measurements to changes in the fissile content of civil power reactors. Such measurements may be useful in IAEA reactor safeguards applications. We introduce a hypothesis testing procedure to identify statistically significant differences between the antineutrino count rate evolution of a standard 'baseline' fuel cycle and that of an anomalous cycle, in which plutonium is removed and replaced with an equivalent fissile worth of uranium. The test would allow an inspector to detect anomalous reactor activity, or to positively confirm that the reactor is operating in a manner consistent with its declared fuel inventory and power level. We show that with a reasonable choice of detector parameters, the test can detect replacement of 73 kg of plutonium in 90 days with 95% probability, while controlling the false positive rate at 5%. We show that some improvement on this level of sensitivity may be expected by various means, including use of the method i...

  9. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  10. System 80+{trademark} Standard Design: CESSAR design certification. Volume 4: Amendment I

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-21

    This report, entitled Combustion Engineering Standard Safety Analysis Report - Design Certification (CESSAR-DC), has been prepared in support of the industry effort to standardize nuclear plant designs. These documents describe the Combustion Engineering, Inc. System 80+{sup TM} Standard Design. This report, Volume 4, provides a description of the reactor, reactor internals, fuel assemblies, and associated design requirements.

  11. 77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-10-09

    ... COMMISSION Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors... comment on NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power..., ``Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors.'' DATES: Submit comments...

  12. 77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-11-06

    ... COMMISSION Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors... the Commission), issued a NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports...), Section 19.0 ``Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors.'' The NRC...

  13. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere...

  14. 75 FR 1831 - Seeks Qualified Candidates for the Advisory Committee on Reactor Safeguards

    Science.gov (United States)

    2010-01-13

    ... used to evaluate candidates include education and experience, demonstrated skills in nuclear reactor safety matters, the ability to solve complex technical problems, and the ability to work collegially on a... on the adequacy of proposed reactor safety standards. Of primary importance are the safety...

  15. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic

  16. Turning points in reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  17. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  18. Acceptability of reactors in space

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1981-04-01

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

  19. Spiral-shaped disinfection reactors

    KAUST Repository

    Ghaffour, Noreddine

    2015-08-20

    This disclosure includes disinfection reactors and processes for the disinfection of water. Some disinfection reactors include a body that defines an inlet, an outlet, and a spiral flow path between the inlet and the outlet, in which the body is configured to receive water and a disinfectant at the inlet such that the water is exposed to the disinfectant as the water flows through the spiral flow path. Also disclosed are processes for disinfecting water in such disinfection reactors.

  20. Hydrogen Production in Fusion Reactors

    OpenAIRE

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M; Uenosono, C.

    1993-01-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated.

  1. Neutrino Oscillation Studies with Reactors

    CERN Document Server

    Vogel, Petr; Zhang, Chao

    2015-01-01

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  2. FAST NEUTRONIC REACTOR

    Science.gov (United States)

    Snell, A.H.

    1957-12-01

    This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

  3. Biparticle fluidized bed reactor

    Science.gov (United States)

    Scott, C.D.

    1993-12-14

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

  4. Small Punch Test on Before and Post Irradiated Domestic Reactor Pressure Steel

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Problems may be caused when applying the standard specimen to study the properties of irradiated reactor materials, because of its big dimension, e.g.: The inner temperature gradient of the specimen is high when irradiated, the radiation

  5. Prompt Neutron Lifetime for the NBSR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, A.L.; Diamond, D.

    2012-06-24

    In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

  6. Reactor technology assessment and selection utilizing systems engineering approach

    Science.gov (United States)

    Zolkaffly, Muhammed Zulfakar; Han, Ki-In

    2014-02-01

    The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology.

  7. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  8. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  9. Supernova 1998bw In ESO 184-G82. Repr. from Central Bureau for Astronomical Telegrams, no. 6969,10 Jul. 1998

    Science.gov (United States)

    Kay, L. E.; Halpern, Jules P.; Leighly, K. M.; Heathcote, S.; Magalhaes, A. M.; Oliversen, Ronald (Technical Monitor)

    2001-01-01

    Over the wavelength range 390-750 nm, we measure intrinsic linear polarization of 0.53 +/- 0.08 percent at position angle 49 +/- 3 deg, after correcting for Galactic interstellar polarization using the star HD 184100, which has polarization of 0.75 +/- 0.01 percent at p.a. 176.5 +/- 2.5 deg. This measured interstellar polarization is consistent with the Galactic extinction in this direction, estimated to be E(B-V) = 0.059 from IRAS maps, or E(B-V) = 0.079 from 21-cm H I. Interstellar polarization in the host galaxy ESO 184-G82 is expected to be negligible based on the relative absence of Na I D absorption at z = 0.00841 +/- 0.00005, the redshift of the environment of the supernova from narrow H II region emission lines in its spectrum. Polarization appears highest in between emission features in the total flux spectrum, which strengthens the interpretation of the polarization as intrinsic to the supernova. This modest polarization is less than that of some type-II supernovae, but greater than that of type-Ia supernovae, which are generally unpolarized. This supports the interpretation of SN 1998bw, a peculiar type-Ic supernova, as a core-collapse event in which the observed polarization is due to moderate asymmetry in either the photosphere of the ejecta or an overlying scattering envelope. However, this result does not strongly constrain arguments about whether some supernovae emit gamma-ray bursts, since such emission may come from a mildly relativistic shock associated with the radio emission and above the optical photosphere, without any requirements on beaming or orientation." A. V. Filippenko, University of California at Berkeley, comments on the total flux spectrum obtained above: "The spectrum most closely resembles those of the peculiar SN 1997ef, but perhaps evolving more slowly. It is not typical of type-Ic supernovae; indeed, the spectrum does not match any of the known spectral classes, but perhaps 'peculiar type Ic' is the best choice at this time

  10. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    Energy Technology Data Exchange (ETDEWEB)

    Cagle, C.D. (comp.)

    1982-10-01

    The Oak Ridge National Laboratory has three multipurpose research reactors which accommodate testing loops, target irradiations, and beam-type experiments. Since the experiments must share common or similar facilities and utilities, be designed and fabricated by the same groups, and meet the same safety criteria, certain standards for these have been developed. These standards deal only with those properties from which safety and economy of time and money can be maximized and do not relate to the intent of the experiment or quality of the data obtained. The necessity for, and the limitations of, the standards are discussed; and a compilation of general standards is included.

  11. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  12. The rehabilitation/upgrading of Philippine Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Renato, T. Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

    1998-10-01

    The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

  13. 标准型定位格架对超临界水在堆芯传热和流动影响的数值研究%Numerical research on heat and mass transfer of supercritical water in reactor core with standard grid spacer

    Institute of Scientific and Technical Information of China (English)

    朱晓静; 周盛; 邱庆刚; 沈胜强

    2016-01-01

    Numerical analysis is carried out for the heat and mass transfer ofsupercritical water in the sub-channels with standard grid spacer in supercritical reactor core with commercial software STARCCM+ 6.04 using unstructured polygonal mesh.The results show that the circumferential distribution of cladding temperatureis not uniform due to flow choking effects.The highest value and the lowest value appear at the narrow gap region the sub-channel center respectively.The local heat transferinside the grid spacer strapis greatly enhanceddue to the increased flow velocity.However,the standard grid spacer is negative for decreasing cladding temperature because it increases circumferential difference of cladding temperature in its downstream region and an obvious decreased heat transfer is caused by the aggravated flow choking effects. The decreased heat transfer downstream of a standard grid spacer maybe a unique feature of the tight rod bundle.%采用非结构化多面体网格,利用商业软件 STARCCM+6.04对标准型定位格架对超临界水在反应堆堆芯子通道内流动及传热特性的影响进行了数值研究。研究结果表明,由于流动阻塞效应的影响,燃料棒覆层表面温度周向分布不均,窄缝区最高、中心区最低。标准定位格架能够强化定位格架内部传热,但会造成其下游局部传热弱化,周向温度分布差异增大,不利于降低覆层温度。标准型定位格架下游的传热弱化现象或许是超临界反应堆紧凑型堆芯的特有现象,须在堆芯设计中加以重视。

  14. Brookhaven leak reactor to close

    CERN Multimedia

    MacIlwain, C

    1999-01-01

    The DOE has announced that the High Flux Beam Reactor at Brookhaven is to close for good. Though the news was not unexpected researchers were angry the decision had been taken before the review to assess the impact of reopening the reactor had been concluded (1 page).

  15. Thermochemical reactor systems and methods

    Energy Technology Data Exchange (ETDEWEB)

    Lipinski, Wojciech; Davidson, Jane Holloway; Chase, Thomas Richard

    2016-11-29

    Thermochemical reactor systems that may be used to produce a fuel, and methods of using the thermochemical reactor systems, utilizing a reactive cylindrical element, an optional energy transfer cylindrical element, an inlet gas management system, and an outlet gas management system.

  16. Chemical-vapor-deposition reactor

    Science.gov (United States)

    Chern, S.

    1979-01-01

    Reactor utilizes multiple stacked trays compactly arranged in paths of horizontally channeled reactant gas streams. Design allows faster and more efficient deposits of film on substrates, and reduces gas and energy consumption. Lack of dead spots that trap reactive gases reduces reactor purge time.

  17. Antineutrino Monitoring of Thorium Reactors

    CERN Document Server

    Akindele, Oluwatomi A; Norman, Eric B

    2015-01-01

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuels types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring.Thorium molten salt reactors (MSR) breed U-233, that if diverted constitute an IAEA direct use material. The antineutrino spectrum from the fission of U-233 has been determined, the feasibility of detecting the diversion of a significant quantity, 8 kg of U-233, within the IAEA timeliness goal of 30 days has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario at a 25 meter standoff by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos. It was found that the diversion of a signifi...

  18. Engineering reactors for catalytic reactions

    Indian Academy of Sciences (India)

    Vivek V Ranade

    2014-03-01

    Catalytic reactions are ubiquitous in chemical and allied industries. A homogeneous or heterogeneous catalyst which provides an alternative route of reaction with lower activation energy and better control on selectivity can make substantial impact on process viability and economics. Extensive studies have been conducted to establish sound basis for design and engineering of reactors for practising such catalytic reactions and for realizing improvements in reactor performance. In this article, application of recent (and not so recent) developments in engineering reactors for catalytic reactions is discussed. Some examples where performance enhancement was realized by catalyst design, appropriate choice of reactor, better injection and dispersion strategies and recent advances in process intensification/ multifunctional reactors are discussed to illustrate the approach.

  19. Unsteady processes in catalytic reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matros, Yu.Sh.

    1985-01-01

    In recent years a realization has occurred that reaction and reactor dynamics must be considered when designing and operating catalytic reactors. In this book, the author has focussed on both the processes occurring on individual porous-catalyst particles as well as the phenomena displayed by collections of these particles in fixed-bed reactors. The major topics discussed include the effects of unsteady-state heat and mass transfer, the influence of inhomogeneities and stagnant regions in fixed beds, and reactor operation during forced cycling of operating conditions. Despite the title of the book, attention is also paid to the determination of the number and stability of fixed-bed steady states, with the aim of describing the possibility of controlling reactors at unstable steady states. However, this development is somewhat dated, given the recent literature on multiplicity phenomena and process control.

  20. A model of reactor kinetics

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, A.S.; Thompson, B.R.

    1988-09-01

    The analytical model of nuclear reactor transients, incorporating both mechanical and nuclear effects, simulates reactor kinetics. Linear analysis shows the stability borderline for small power perturbations. In a stable system, initial power disturbances die out with time. With an unstable combination of nuclear and mechanical characteristics, initial disturbances persist and may increase with time. With large instability, oscillations of great magnitude occur. Stability requirements set limits on the power density at which particular reactors can operate. The limiting power density depends largely on the product of two terms: the fraction of delayed neutrons and the frictional damping of vibratory motion in reactor core components. As the fraction of delayed neutrons is essentially fixed, mechanical damping largely determines the maximum power density. A computer program, based on the analytical model, calculates and plots reactor power as a nonlinear function of time in response to assigned values of mechanical and nuclear characteristics.

  1. Metallic fuels for advanced reactors

    Science.gov (United States)

    Carmack, W. J.; Porter, D. L.; Chang, Y. I.; Hayes, S. L.; Meyer, M. K.; Burkes, D. E.; Lee, C. B.; Mizuno, T.; Delage, F.; Somers, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.

  2. Seismic base isolation technologies for Korea advanced liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, B.; Lee, J.-H.; Koo, G.-H.; Lee, H.-Y.; Kim, J.-B. [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    2000-06-01

    This paper describes the status and prospects of the seismic base isolation technologies for Korea Advanced Liquid Metal Reactor (KALIMER). The research and development program on the seismic base isolation for KALIMER began in 1993 by KAERI under the national long-term R and D program. The objective of this program is to enhance the seismic safety, to accomplish the economic design, and to standardize the plant design through the establishment of technologies on seismic base isolation for liquid metal reactors. In this paper, tests and analyses performed in the program are presented. (orig.)

  3. Neutrino Experiments at Reactors

    Science.gov (United States)

    Reines, F.; Gurr, H. S.; Jenkins, T. L.; Munsee, J. H.

    1968-09-09

    A description is given of the electron-antineutrino program using a large fission reactor. A search has been made for a neutral weak interaction via the reaction (electron antineutrino + d .> p + n + electron antineutrino), the reaction (electron antineutrino + d .> n + n + e{sup +}) has now been detected, and an effort is underway to observe the elastic scattering reaction (electron antineutrino + e{sup -} .> electron antineutrino + e{sup -}) as well as to measure more precisely the reaction (electron antineutrino + p .> n + e{sup+}). The upper limit on the elastic scattering reaction which we have obtained with our large composite NaI, plastic, liquid scintillation detector is now about 50 times the predicted value.

  4. Neutronic Reactor Shield

    Science.gov (United States)

    Fermi, Enrico; Zinn, Walter H.

    The argument of the present Patent is a radiation shield suitable for protection of personnel from both gamma rays and neutrons. Such a shield from dangerous radiations is achieved to the best by the combined action of a neutron slowing material (a moderator) and a neutron absorbing material. Hydrogen is particularly effective for this shield since it is a good absorber of slow neutrons and a good moderator of fast neutrons. The neutrons slowed down by hydrogen may, then, be absorbed by other materials such as boron, cadmium, gadolinium, samarium or steel. Steel is particularly convenient for the purpose, given its effectiveness in absorbing also the gamma rays from the reactor (both primary gamma rays and secondary ones produced by the moderation of neutrons). In particular, in the present Patent a shield is described, made of alternate layers of steel and Masonite (an hydrolized ligno-cellulose material). The object of the present Patent is not discussed in any other published paper.

  5. Accounting standards

    NARCIS (Netherlands)

    Stellinga, B.; Mügge, D.

    2014-01-01

    The European and global regulation of accounting standards have witnessed remarkable changes over the past twenty years. In the early 1990s, EU accounting practices were fragmented along national lines and US accounting standards were the de facto global standards. Since 2005, all EU listed companie

  6. Genetic associations between daily BW gain and live fleshiness of station-tested young bulls and carcass and meat quality traits of commercial intact males in Piemontese cattle.

    Science.gov (United States)

    Bonfatti, V; Albera, A; Carnier, P

    2013-05-01

    The aim of this study was to investigate genetic relationships between beef traits of station-tested young bulls and carcass and meat quality traits (MQ) of commercial intact males in Piemontese cattle. Phenotypes for daily gain (DG) and live fleshiness traits (width at withers: WW; shoulder muscularity: SM; loin width: LW; loin thickness: LT; thigh muscularity: TM; thigh profile: TP) and thinness of the shin bone (BT) were available for 3,109 and 2,183 performance-tested young bulls, respectively. Carcass daily gain (CDG), carcass conformation (SEUS), pH at 24 h (pH24h) and 8 d after slaughter (pH8d), lightness (L*), redness (a*), yellowness (b*), hue angle (HA), saturation index (SI), drip loss (DL), cooking loss (CL), and shear force (SF) were assessed for 1,208 commercial intact males. (Co) variance components were estimated in a set of twelve 9-traits analyses using REML and linear animal models including all performance-test traits and 1 carcass or MQ trait at a time. Heritabilities ± SE of beef traits ranged from 0.26 ± 0.03 (LW) to 0.47 ± 0.01 (DG), whereas those of carcass traits and MQ from 0.06 ± 0.03 (CL) to 0.63 ± 0.04 (HA). The genetic correlation (rg) between DG and CDG was 0.75 ± 0.10, indicating that DG, as measured at the test station, is a good indicator of the carcass gain achieved by commercial animals under farms conditions. Daily BW gain of station-tested bulls correlated positively with color traits (from 0.11 ± 0.12 to 0.54 ± 0.09), ph8d (rg ± SE = 0.31 ± 0.11), DL (rg ± SE = 0.29 ± 0.17), and CL (rg ± SE = 0.27 ± 0.18). Live fleshiness of station-tested bulls exhibited genetic correlations with MQ of commercial animals that were positive for L* and b* (from 0.13 ± 0.08 to 0.65 ± 0.14) and negative for pH (from -0.27 ± 0.15 to -0.57 ± 0.11), CL (from -0.16 ± 0.23 to -0.43 ± 0.22), and SF (TM: rg ± SE = -0.31 ± 0.15; TP: rg ± SE = -0.41 ± 0.17). The thinness of the shin bone correlated unfavorably with CDG (rg ± SE

  7. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-12-31

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

  8. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-01-01

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

  9. Clinical relevance of anti-CEA-radioimmunoscintigraphy with the [sup 99m]Tc-monoclonal antibody BW 431/26. Critical evaluation after 119 examinations. Klinische Relevanz der Anti-CEA-Immunszintigraphie mit dem [sup 99m]Tc-markierten monoklonalen Antikoerper BW 431/26. Kritische Bestandsaufnahme nach 119 Untersuchungen

    Energy Technology Data Exchange (ETDEWEB)

    Hach, A.; Piepenburg, R.; Steinert, H.; Lahmann, C.; Hahn, K. (Univ. Mainz, Nuklearmedizinische Klinik (Germany)); Wittig, B.; Dippold, W. (Univ. Mainz, 1. Medizinische Klinik (Germany))

    1992-07-01

    The results of 119 radioimmunoscintigraphies (RIS) in 113 patients with the [sup 99m]Tc-labeled monclonal anti-CEA-antibody BW 431/26 (Behring) have been analysed. The aim of our study was the estimation of the method's sensitivity and specificity under different aspects to find out for which indications and questions the [sup 99m]Tc-RIS is useful. Colorectal primary tumours in 19 patients were scintigraphically detected with a sensitivity of 83% and a specificity of 100%; 3 out of 7 other tumour sites were localised correctly. 55 patients were examined during the follow-up of colorectal cancer. There were 17 out of 22 true positive findings of local recurrences (sensitivity 77%, specificity 88%). Liver metastases were imaged as hot lesions with only 41% sensitivity and 86% specificity. In 14 patients with other non-colorectal carcinomas, RIS was successful in single cases. It is not helpful, however, when searching for tumours of unknown origin or for the screening of patients with elevated CEA levels without tumour history. The high technical, methodological and time effort required by RIS is justified in the follow-up of cancer patients when conventional diagnostic procedures are inconclusive or the status of morphological findings remains unclear. The use of RIS as an unspecific screening tool in tumour diagnosis must be rejected because of the not completely explored risks of the examination. Repeated applications of monoclonal antibodies require controls of the patients' HAMA titers before performing RIS. (orig.).

  10. Assessment of torsatrons as reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lyon, J.F. (Oak Ridge National Lab., TN (United States)); Painter, S.L. (Australian National Univ., Canberra, ACT (Australia))

    1992-12-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors because stellarators have no dangerous disruptions and no need for continuous current drive or power recirculated to the plasma, both easing the first wall, blanket, and shield design; less severe constraints on the plasma parameters and profiles; and better access for maintenance. This study shows that a reactor based on the torsatron configuration (a stellarator variant) could also have up to double the mass utilization efficiency (MUE) and a significantly lower cost of electricity (COE) than a conventional tokamak reactor (ARIES-I) for a range of assumptions. Torsatron reactors can have much smaller coil systems than tokamak reactors because the coils are closer to the plasma and they have a smaller cross section (higher average current density because of the lower magnetic field). The reactor optimization approach and the costing and component models are those used in the current stage of the ARIES-I tokamak reactor study. Typical reactor parameters for a 1-GW(e) Compact Torsatron reactor example are major radius R[sub 0] = 6.6-8.8 m, on-axis magnetic field B[sup 0] = 4.8-7.5 T, B[sub max] (on coils) = 16 T, MUE 140-210 kW(e)/tonne, and COE (in constant 1990 dollars) = 67-79 mill/kW(e)h. The results are relatively sensitive to assumptions on the level of confinement improvement and the blanket thickness under the inboard half of the helical windings but relatively insensitive to other assumptions.

  11. Concept for LEU Burst Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kimpland, Robert Herbert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-07

    Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

  12. Nuclear reactor downcomer flow deflector

    Energy Technology Data Exchange (ETDEWEB)

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    2011-02-15

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  13. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  14. Random processes in nuclear reactors

    CERN Document Server

    Williams, M M R

    1974-01-01

    Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

  15. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  16. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rohanda, Anis [Center for Reactor Technology and Nuclear Safety – BATAN Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 (Indonesia); Waris, Abdul [Physics Department of ITB, Indonesia anis-rohanda@yahoo.com (Indonesia)

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  17. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  18. Background Studies for the MINER Coherent Neutrino Scattering Reactor Experiment

    CERN Document Server

    Agnolet, G; Barker, D; Beck, R; Carroll, T J; Cesar, J; Cushman, P; Dent, J B; De Rijck, S; Dutta, B; Flanagan, W; Fritts, M; Gao, Y; Harris, H R; Hays, C C; Iyer, V; Jastram, A; Kadribasic, F; Kennedy, A; Kubik, A; Ogawa, I; Lang, K; Mahapatra, R; Mandic, V; Martin, R D; Mast, N; McDeavitt, S; Mirabolfathi, N; Mohanty, B; Nakajima, K; Newhouse, J; Newstead, J L; Phan, D; Proga, M; Roberts, A; Rogachev, G; Salazar, R; Sander, J; Senapati, K; Shimada, M; Strigari, L; Tamagawa, Y; Teizer, W; Vermaak, J I C; Villano, A N; Walker, J; Webb, B; Wetzel, Z; Yadavalli, S A

    2016-01-01

    The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 meters) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5 to 20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

  19. Reactor pressure vessels as type B transport containment boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E. [Applied Science and Technology, Inc., Poway, CA (United States); Griesbach, T.J. [ATI Consulting, Danville, CA (United States)

    1998-07-01

    Transportation risk and personnel exposure, as well as the cost of decommissioning nuclear power plants, can all be reduced significantly through the one-time use of the reactor pressure vessel as a containment boundary for shipping the activated internal components from the reactor site to a burial site. In order to help provide the technical basis for this end-use application, the ASME Board on Nuclear Codes and Standards, through its Subcommittee XI, has prepared a draft nuclear code case that contains requirements for any modifications to the vessel, including materials, design, fabrication, and examination. In particular, the requirements for evaluation of potential brittle fracture as the result of potentially low ambient shipping temperatures combined with hypothetical transportation accident loading are addressed. Existing ASME Code Section XI rules for linear elastic fracture mechanics evaluation of irradiated reactor pressure vessels have been adapted and included in the code case. (authors)

  20. Wastewater Treatment in a Hybrid Biological Reactor (HBR) :Nitrification Characteristics

    Institute of Scientific and Technical Information of China (English)

    JIAN-LONG WANG; LI-BO WU

    2004-01-01

    To investigate the nitrifying characteristics of both suspended- and attached- biomass in a hybrid bioreactor. Methods The hybrid biological reactor was developed by introducing porous ceramic particles into the reactor to provide the surface for biomass attachment. Microorganisms immobilized on the ceramics were observed using scanning electron microscopy (SEM). All chemical analyses were performed in accordance with standard methods. Results The suspended- and attached-biomass had approximately the same nitrification activity. The nitrifying kinetic was independent of the initial biomass concentration, and the attached-biomass had a stronger ability to resist the nitrification inhibitor. Conclusion The attached biomass is superior to suspended-biomass for nitrifying wastewater, especially that containing toxic organic compounds. The hybrid biological reactor consisting of suspended- and attached-biomass is advantageous in such cases.

  1. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  2. System Requirements Analysis for a Computer-based Procedure in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaek Wan; Jang, Gwi Sook; Seo, Sang Moon; Shin, Sung Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    This can address many of the routine problems related to human error in the use of conventional, hard-copy operating procedures. An operating supporting system is also required in a research reactor. A well-made CBP can address the staffing issues of a research reactor and reduce the human errors by minimizing the operator's routine tasks. A CBP for a research reactor has not been proposed yet. Also, CBPs developed for nuclear power plants have powerful and various technical functions to cover complicated plant operation situations. However, many of the functions may not be required for a research reactor. Thus, it is not reasonable to apply the CBP to a research reactor directly. Also, customizing of the CBP is not cost-effective. Therefore, a compact CBP should be developed for a research reactor. This paper introduces high level requirements derived by the system requirements analysis activity as the first stage of system implementation. Operation support tools are under consideration for application to research reactors. In particular, as a full digitalization of the main control room, application of a computer-based procedure system has been required as a part of man-machine interface system because it makes an impact on the operating staffing and human errors of a research reactor. To establish computer-based system requirements for a research reactor, this paper addressed international standards and previous practices on nuclear plants.

  3. Two cases report of rare platelet antigen HPA-21bw allele%两例罕见的血小板抗原HPA-21bw等位基因报告

    Institute of Scientific and Technical Information of China (English)

    周豪杰; 聂咏梅; 李晓帆; 汪传喜

    2013-01-01

    目的:调查中国人群人类血小板抗原HPA-21w多态性.方法:采用聚合酶链反应-序列特异性引物法(PCR-SSP)基因分型技术,对广州汉族血小板献血者进行HPA-21w基因分型,并使用PCR扩增HPA-21w基因片段,进行DNA测序验证.结果:在200例受检者中发现2例HPA-21 a/b杂合子个体,DNA测序表明GPIIIa编码基因1960位G→A,导致GPIIIa膜糖蛋白第628位谷氨酸被赖氨酸所取代,产生HPA-21 bw抗原特异性.中国人群HPA-21 bw的等位基因频率为0.50%.结论:在中国人群首次检出HPA-21 bw等位基因,提示在血小板同种免疫症的诊断和血小板输注治疗中,该抗原具有免疫学意义.%Objective:To study the polymorphism of human platelet antigen HPA-21 w in the Chinese population.Methods:A total of 200 Han voluntary platelet donors in Guangzhou were genotyped for HPA-21w by using a polymerase chain reaction-sequence specific primers (PCR-SSP) method.In order to confirm the genotyping result,a fragment of HPA-21w gene was amplified and then sequenced directly.Results:Two individuals with a rare HPA-21a/b genotype were found by using PCR-SSP method.DNA-sequencing showed that a single G to A substitution at nucleotide 1960 occurred,resulting in the replacement of a glutamic acid with a lysine acid at position 628 of GPIIIa protein.This substitution generated antigenic specificity HPA-21bw.The allele frequency of HPA-21bw was 0.50% in the Chinese population.Conclusion:This is the first study reported the rare HPA-21bw allele in the Chinese population.The detection of two rare HPA-21 bw alleles suggests that HPA-21 bw should be considered in platelet alloimmunization syndromes and platelet transfusion therapy.

  4. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  5. ADAPTIVE CONTROL SYSTEM OF INDUSTRIAL REACTORS

    Directory of Open Access Journals (Sweden)

    Vyacheslav K. Mayevski

    2014-01-01

    Full Text Available This paper describes a mathematical model of an industrial chemical reactor for production of synthetic rubber. During reactor operation the model parameters vary considerably. To create a control algorithm performed transformation of mathematical model of the reactor in order to obtain a dependency that can be used to determine the model parameters are changing during reactor operation.

  6. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  7. Breeder Reactors, Understanding the Atom Series.

    Science.gov (United States)

    Mitchell, Walter, III; Turner, Stanley E.

    The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

  8. Evolution of the tandem mirror reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, G.A.; Logan, B.G.

    1982-03-09

    We discuss the evolution of the tandem mirror reactor concept from the original conceptual reactor design (1977) through the first application of the thermal barrier concept to a reactor design (1979) to the beginning of the Mirror Advanced Reactor Study (1982).

  9. ANALISIS TRANSIEN PADA FIXED BED NUCLEAR REACTOR

    Directory of Open Access Journals (Sweden)

    M. Rizaal

    2015-03-01

    Full Text Available Desain teras Fixed Bed Nuclear Reactor (FBNR yang modular memungkinkan pengendalian daya dapat dilakukan dengan mengatur ketinggian suspended core dan laju aliran massa pendingin. Tujuan penelitian ini adalah mempelajari perubahan daya termal teras sebagai akibat perubahan laju aliran massa pendingin yang masuk ke teras reaktor dan perubahan ketinggian suspended core serta mempelajari karakteristik keselamatan melekat yang dimiliki FBNR saat terjadi kegagalan pelepasan kalor (loss of heat sink. Keadaan neutronik teras dimodelkan pada kondisi tunak dengan menggunakan paket program Standard Reactor Analysis Code (SRAC untuk memperoleh data fluks neutron, konstanta grup, fraksi neutron kasip, konstanta peluruhan prekursor neutron kasip, dan beberapa parameter teras penting lainnya. Selanjutnya data tersebut digunakan pada perhitungan transien sebagai syarat awal. Analisis transien dilakukan pada tiga kondisi, yaitu saat terjadi penurunan laju aliran massa pendingin, saat terjadi penurunan ketinggian suspended core, dan saat terjadi kegagalan sistem pelepasan kalor. Hasil yang diperoleh dari penelitian ini menunjukkan bahwa penurunan laju aliran massa pendingin sebesar 50%, dari kondisi normal, menyebabkan daya termal teras turun 28% dibanding daya sebelumnya. Penurunan ketinggian suspended core sebesar 30% dari ketinggian normal menyebabkan daya termal teras turun 17% dibanding daya sebelumnya. Sementara untuk kondisi kegagalan sistem pelepasan kalor, daya termal teras mengalami penurunan sebesar 76%. Dengan demikian, pengendalian daya pada FBNR dapat dilakukan dengan mengatur laju aliran massa pendingin dan ketinggian suspended core, serta keselamatan melekat yang handal pada kondisi kegagalan sistem pelepasan kalor. Kata kunci: FBNR, transien, daya, laju aliran massa, suspended core Modular in design enables Fixed Bed Nuclear Reactor (FBNR power controlled by the adjustment of suspended core and coolant flow rate. The main purposes of this paper

  10. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  11. Jules Horowitz Reactor, basic design

    Energy Technology Data Exchange (ETDEWEB)

    Bergamaschi, Y.; Bouilloux, Y.; Chantoin, P.; Guigon, B.; Bravo, X.; Germain, C.; Rommens, M.; Tremodeux, P

    2003-07-01

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it. It has the ambition to provide the necessary nuclear data and maintain a fission research capacity in Europe after 2010. This capacity should be service-oriented. It will be established in Cadarache. The Jules Horowitz reactor will also: - represent a significant step in term of performances and experimental capabilities, - be designed with a high flexibility, in order to satisfy the current demand from European industry, research and be able to accommodate future requirements, - reach a high level of safety, according to the best current practice. This paper will present the main functionalities and the design options resulting from the 'preliminary design' studies. (authors)

  12. Advanced Catalytic Hydrogenation Retrofit Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reinaldo M. Machado

    2002-08-15

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  13. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The overall objective of the Phase 1 effort was to demonstrate the technical feasibility of the Advanced Carbothermal Electric (ACE) Reactor concept. Unlike...

  14. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

    1989-12-01

    This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

  15. Unique features of space reactors

    Science.gov (United States)

    Buden, David

    Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K.

  16. Thermal Analysis for Mobile Reactor

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>Mobile reactor design in the paper is consisted of two grades of thermal electric conversion. The first grade is the thermionic conversion inside the core and the second grade is thermocouple conversion

  17. Teaching About Nature's Nuclear Reactors

    CERN Document Server

    Herndon, J M

    2005-01-01

    Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

  18. Advanced Carbothermal Electric Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop the Advanced Carbothermal Electric (ACE) reactor to efficiently extract oxygen from lunar regolith. Unlike state-of-the-art carbothermal...

  19. Determination of buckling in the IPEN/MB-01 Reactor in cylindrical configuration

    Energy Technology Data Exchange (ETDEWEB)

    Purgato, Rafael Turrini; Bitelli, Ulysses d' Utra; Aredes, Vitor Ottoni; Silva, Alexandre F. Povoa da; Santos, Diogo Feliciano dos; Mura, Luis Felipe L.; Jerez, Rogerio, E-mail: rtpurgato@ipen.br, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    One of the key parameters in reactor physics is the buckling of a reactor core. It is related to important parameters such as reaction rates, nuclear power operation, fuel burning, among others. In a critical reactor, the buckling depends on the geometric and material characteristics of the reactor core. This paper presents the results of experimental buckling in the reactor IPEN/MB-01 nuclear reactor in its cylindrical configuration with 28 fuel rods along its diameter. The IPEN/MB-01 is a zero power reactor designed to operate at a maximum power of 100 watts, it is a versatile nuclear facility which allows the simulation of all the characteristics of a large nuclear power reactor and ideal for this type of measurement. We conducted a mapping of neutron flux inside the reactor and thereby determined the total buckling of the cylindrical configuration. The reactor was operated for an hour. Then, the activation of the fuel rods was measured by gamma spectrometry on a rod scanner HPGe detector. We analyzed the gamma photons of the {sup 239}Np (276,6 keV) for neutron capture and the {sup 143}Ce (293,3 keV) for fission on both {sup 238}U and {sup 235}U, respectively. We analyzed the axial and radial directions. Other measurements were performed using wires and gold foils in the radial and axial directions of the reactor core. The results showed that the cylindrical configuration compared to standard rectangular configuration of the IPEN/MB-01 reactor has a higher neutron economy, since in this configuration there is less leakage of neutrons. The Buckling Total obtained from the three methods was 95.84 ± 2.67 m{sup -2}. (author)

  20. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

  1. China experimental fast reactor; Le reacteur rapide experimental chinois

    Energy Technology Data Exchange (ETDEWEB)

    Tianmin, X. [Institut d' Ingenierie Nucleaire de Pekin (China); Cunren, L. [Centre d' Etude de Surete de Pekin (China)

    2007-07-15

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*10{sup 15} n/cm{sup 2}/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  2. Effect of DUPIC cycle on CANDU reactor safety parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M. A. [Atomic Energy Authority, ETRR-2, Cairo (Egypt); Badawi, Alya [Dept. of Nuclear and Radiation Engineering, Alexandria University, Alexandria (Egypt)

    2016-10-15

    Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by UO{sub 2} enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

  3. Microchannel Reactors for ISRU Applications

    Science.gov (United States)

    Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

    2005-02-01

    Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

  4. Reactor antineutrinos and nuclear physics

    Science.gov (United States)

    Balantekin, A. B.

    2016-11-01

    Short-baseline reactor neutrino experiments successfully measured the neutrino parameters they set out to measure, but they also identified a shape distortion in the 5-7 MeV range as well as a reduction from the predicted value of the flux. Nuclear physics input into the calculations of reactor antineutrino spectra needs to be better refined if this anomaly is to be interpreted as due to sterile neutrino states.

  5. Communications standards

    CERN Document Server

    Stokes, A V

    1986-01-01

    Communications Standards deals with the standardization of computer communication networks. This book examines the types of local area networks (LANs) that have been developed and looks at some of the relevant protocols in more detail. The work of Project 802 is briefly discussed, along with a protocol which has developed from one of the LAN standards and is now a de facto standard in one particular area, namely the Manufacturing Automation Protocol (MAP). Factors that affect the usage of networks, such as network management and security, are also considered. This book is divided into three se

  6. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  7. Novel Catalytic Membrane Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stuart Nemser, PhD

    2010-10-01

    There are many industrial catalytic organic reversible reactions with amines or alcohols that have water as one of the products. Many of these reactions are homogeneously catalyzed. In all cases removal of water facilitates the reaction and produces more of the desired chemical product. By shifting the reaction to right we produce more chemical product with little or no additional capital investment. Many of these reactions can also relate to bioprocesses. Given the large number of water-organic compound separations achievable and the ability of the Compact Membrane Systems, Inc. (CMS) perfluoro membranes to withstand these harsh operating conditions, this is an ideal demonstration system for the water-of-reaction removal using a membrane reactor. Enhanced reaction synthesis is consistent with the DOE objective to lower the energy intensity of U.S. industry 25% by 2017 in accord with the Energy Policy Act of 2005 and to improve the United States manufacturing competitiveness. The objective of this program is to develop the platform technology for enhancing homogeneous catalytic chemical syntheses.

  8. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iwashige, Kengo

    1996-06-21

    In an LMFBR type reactor, partitions are disposed to a coolant channel at positions lower than the free liquid level, and the width of the partitions is adapted to have a predetermined condition. Namely, when low temperature fluid overflowing the wall of the coolant channel, flows down and collided against the free liquid surface in the coolant channel, since the dropping speed thereof is reduced abruptly, large pressure waves are caused by kinetic force of the low temperature fluid. However, if appropriate numbers of partitions having an appropriate shape are formed, the dropping speed of the low temperature fluid is moderated to reduce the pressure waves. In addition, since the pressure waves are dispersed to the circumferential and lateral directions of the coolant flow channel respectively, the propagation of the pressure waves can be prevented effectively. Further, when the flow of the low temperature fluid is changed to the circumferential direction, for example, by earthquakes, since the partitions act as members resisting against the circumferential change of the low temperature fluid, the change of the direction can be suppressed. (N.H.)

  9. Calculation of reactor antineutrino spectra in TEXONO

    CERN Document Server

    Chen Dong Liang; Mao Ze Pu; Wong, T H

    2002-01-01

    In the low energy reactor antineutrino physics experiments, either for the researches of antineutrino oscillation and antineutrino reactions, or for the measurement of abnormal magnetic moment of antineutrino, the flux and the spectra of reactor antineutrino must be described accurately. The method of calculation of reactor antineutrino spectra was discussed in detail. Furthermore, based on the actual circumstances of NP2 reactors and the arrangement of detectors, the flux and the spectra of reactor antineutrino in TEXONO were worked out

  10. Tritium management in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II.

  11. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  12. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications

    Science.gov (United States)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.

    2012-05-01

    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  13. Unification of reactor elastomeric sealing based on material

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India); Raj, Baldev [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India)

    2012-02-15

    The unification of elastomeric sealing applications of Indian nuclear reactors based on a few qualified fluoroelastomer/perfluoroelastomer compounds and standardized approaches for finite element analysis (FEA) based design, manufacturing process and antifriction coatings is discussed. It is shown that the advance polymer architecture based Viton{sup Registered-Sign} formulation developed for inflatable seals of 500 MWe Prototype Fast Breeder Reactor (PFBR) and its four basic variations can encompass other sealing applications of PFBR with minimum additional efforts on development and validation. Changing the blend ratio of Viton{sup Registered-Sign} GBL 200S and 600S in inflatable seal formulation could extend its use to Pressurized Heavy Water Reactors (PHWRs). The higher operating temperature of Advanced Heavy Water Reactor (AHWR) seals expands the choice to perfluoroelastomers. FEA based on plane-strain/axisymmetric modeling (with Mooney-Rivlin as the basic constitutive model), seal manufacture by cold feed extrusion and injection molding as well as plasma Teflon-like coating belonging to two variations obtained from the development of inflatable seals provide the necessary standardization for unification. The gains in simplification of design, development and operation of seals along with the enhancements of safety and reliability are expected to be substantial.

  14. Achieving Standardization

    DEFF Research Database (Denmark)

    Henningsson, Stefan

    2016-01-01

    competitive, national customs and regional economic organizations are seeking to establish a standardized solution for digital reporting of customs data. However, standardization has proven hard to achieve in the socio-technical e-Customs solution. In this chapter, the authors identify and describe what has...

  15. Achieving Standardization

    DEFF Research Database (Denmark)

    Henningsson, Stefan

    2014-01-01

    competitive, national customs and regional economic organizations are seeking to establish a standardized solution for digital reporting of customs data. However, standardization has proven hard to achieve in the socio-technical e-Customs solution. In this chapter, the authors identify and describe what has...

  16. Innovative hybrid biological reactors using membranes; Reactores biologico hibrido innovadores utilizando membranas

    Energy Technology Data Exchange (ETDEWEB)

    Diez, R.; Esteban-Garcia, A. L.; Florio, L. de; Rodriguez-Hernandez, L.; Tejero, I.

    2011-07-01

    In this paper we present two lines of research on hybrid reactors including the use of membranes, although with different functions: RBPM, biofilm reactors and membranes filtration RBSOM, supported biofilm reactors and oxygen membranes. (Author) 14 refs.

  17. Deterministic and risk-informed approaches for safety analysis of advanced reactors: Part I, deterministic approaches

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang Kyu [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Kim, Inn Seock, E-mail: innseockkim@gmail.co [ISSA Technology, 21318 Seneca Crossing Drive, Germantown, MD 20876 (United States); Oh, Kyu Myung [Korea Institute of Nuclear Safety, 19 Kusong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)

    2010-05-15

    The objective of this paper and a companion paper in this issue (part II, risk-informed approaches) is to derive technical insights from a critical review of deterministic and risk-informed safety analysis approaches that have been applied to develop licensing requirements for water-cooled reactors, or proposed for safety verification of the advanced reactor design. To this end, a review was made of a number of safety analysis approaches including those specified in regulatory guides and industry standards, as well as novel methodologies proposed for licensing of advanced reactors. This paper and the companion paper present the review insights on the deterministic and risk-informed safety analysis approaches, respectively. These insights could be used in making a safety case or developing a new licensing review infrastructure for advanced reactors including Generation IV reactors.

  18. Establishment of licensing process for development reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Yune, Young Gill; Kim, Woong Sik (and others)

    2006-02-15

    A study on licensing processes for development reactors has been performed to prepare the licensing of development reactors developed in Korea. The contents and results of the study are summarized as follows. The licensing processes for nuclear reactors in Korea, U.S.A., Japan, France, U.K., Canada, and IAEA were surveyed and analyzed to obtain technical bases necessary for establishing licensing processes applicable to development reactors in Korea. Based on the technical bases obtained the above analysis, the purpose, power output, and design characteristics of development reactors were analyzed in detail. The analysis results suggested that development reactors should be classified as a new reactor category (called as 'development reactor') separated from the current reactor categories such as the research reactor and the power reactor. Therefore, it is proposed to establish a new reactor category classified as 'development reactor' for the development reactors. And licensing processes, including licensing technical requirements, licensing document requirements, and other regulatory requirements, were also proposed for the development reactors. In order to institutionalize the licensing processes developed in this study, it is necessary to revise the current laws. Therefore, draft provisions of Atomic Energy Act, Enforcement Decree of the Atomic Energy Act, and Enforcement Regulation of the Atomic Energy Act have been developed for the preparation of the future legalization of the licensing processes proposed for the development reactors. Conclusively, a proposal of licensing processes and draft provisions of laws have been developed for the development reactors. The results proposed in this study can be applied directly to the licensing of the future development reactors. Furthermore, they will also contribute to establishing successfully the licensing processes of the development reactors.

  19. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-07-15

    Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

  20. Reactor Simulator Testing

    Science.gov (United States)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise Jon

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz. Keywords: fission, space power, nuclear, liquid metal, NaK.

  1. CANDU in-reactor quantitative visual-based inspection techniques

    Science.gov (United States)

    Rochefort, P. A.

    2009-02-01

    sealed by rolling its ends into the rolled joint area. During reactor refurbishment, the original FC calandria tubes are removed, potentially scratching the rolled joint area and, thereby, compromising the seal with the new FC calandria tube. The procedure involves delivering an inspection module having a radiation-resistant camera, standard lighting, and a structured lighting projector. The surface is inspected by rotating the module within the rolled joint area. If a flaw is detected, its depth and width are gauged from the profile variation of the structured lighting in a captured image. As well, the diameter profile of the area is measured from the analysis of a series of captured circumferential images of the structured lighting profiles on the surface.

  2. Reactivity determination in accelerator driven reactors using reactor noise analysis

    Directory of Open Access Journals (Sweden)

    Kostić Ljiljana 1

    2002-01-01

    Full Text Available Feynman-alpha and Rossi-alpha methods are used in traditional nuclear reactors to determine the subcritical reactivity of a system. The methods are based on the measurement of the mean value, variance and the covariance of detector counts for different measurement times. Such methods attracted renewed attention recently with the advent of the so-called accelerator driven reactors (ADS proposed some time ago. The ADS systems, intended to be used either in energy generation or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those traditionally used by radioactive sources. In such reactors the monitoring of the subcritical reactivity is very important, and a statistical method, such as the Feynman-alpha method, is capable of resolving this problem.

  3. Reactor antineutrino experiments

    OpenAIRE

    Lu, Haoqi

    2014-01-01

    Neutrinos are elementary particles in the standard model of particle physics. There are 3 flavors of neutrinos that oscillate among themselves. Their oscillation can be described by a 3$\\times$3 unitary matrix, containing three mixing angles $\\theta_{12}$, $\\theta_{23}$, $\\theta_{13}$, and one CP phase. Both $\\theta_{12}$ and $\\theta_{23}$ are known from previous experiments. $\\theta_{13}$ was unknown just two years ago. The Daya Bay experiment gave the first definitive non-zero value in 2012...

  4. Modeling and Control of a Large Nuclear Reactor A Three-Time-Scale Approach

    CERN Document Server

    Shimjith, S R; Bandyopadhyay, B

    2013-01-01

    Control analysis and design of large nuclear reactors requires a suitable mathematical model representing the steady state and dynamic behavior of the reactor with reasonable accuracy. This task is, however, quite challenging because of several complex dynamic phenomena existing in a reactor. Quite often, the models developed would be of prohibitively large order, non-linear and of complex structure not readily amenable for control studies. Moreover, the existence of simultaneously occurring dynamic variations at different speeds makes the mathematical model susceptible to numerical ill-conditioning, inhibiting direct application of standard control techniques. This monograph introduces a technique for mathematical modeling of large nuclear reactors in the framework of multi-point kinetics, to obtain a comparatively smaller order model in standard state space form thus overcoming these difficulties. It further brings in innovative methods for controller design for systems exhibiting multi-time-scale property,...

  5. Heterogeneous Transmutation Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. E. Bays

    2007-09-01

    The threshold-fission (fertile) nature of Am-241 is used to destroy this minor actinide by capitalizing upon neutron capture instead of fission within a sodium fast reactor. This neutron-capture and its subsequent decay chain leads to the breeding of even neutron number plutonium isotopes. A slightly moderated target design is proposed for breeding plutonium in an axial blanket located above the active “fast reactor” driver fuel region. A parametric study on the core height and fuel pin diameter-to-pitch ratio is used to explore the reactor and fuel cycle aspects of this design. This study resulted in both non-flattened and flattened core geometries. Both of these designs demonstrated a high capacity for removing americium from the fuel cycle. A reactivity coefficient analysis revealed that this heterogeneous design will have comparable safety aspects to a homogeneous reactor of comparable size. A mass balance analysis revealed that the heterogeneous design may reduce the number of fast reactors needed to close the current once-through light water reactor fuel cycle.

  6. Thermonuclear Reflect AB-Reactor

    CERN Document Server

    Bolonkin, Alexander

    2008-01-01

    The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical pr...

  7. MAN B&W智能气缸润滑技术在新公约下的优势%Superiority of MAN B&W smart cylinder lubrication system tinder the conditions of the new Convention

    Institute of Scientific and Technical Information of China (English)

    俞文胜; 蔡振雄; 刘建华

    2010-01-01

    介绍MARPOL73/78公约附则Ⅵ对船用燃油硫含量的强制要求及与之相适应的气缸润滑的调整,介绍MAN B&W Alpha ACC气缸润滑系统的组成和工作原理,对新公约要求的适应性,以及该系统在气缸润滑过程中所体现的优越性能和经济性.

  8. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    Energy Technology Data Exchange (ETDEWEB)

    Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rishel, Jeremy P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  9. Imaging Fukushima Daiichi reactors with muons

    Directory of Open Access Journals (Sweden)

    Haruo Miyadera

    2013-05-01

    Full Text Available A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  10. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  11. 75 FR 13610 - Office of New Reactors; Interim Staff Guidance on Implementation of a Seismic Margin Analysis for...

    Science.gov (United States)

    2010-03-22

    ... New Reactors Based on Probabilistic Risk Assessment AGENCY: Nuclear Regulatory Commission (NRC... Probabilistic Risk Assessment,'' (Agencywide Documents Access and Management System (ADAMS) Accession No..., ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,'' issued...

  12. (Terminology standardization)

    Energy Technology Data Exchange (ETDEWEB)

    Strehlow, R.A.

    1990-10-19

    Terminological requirements in information management was but one of the principal themes of the 2nd Congress on Terminology and Knowledge Engineering. The traveler represented the American Society for Testing and Materials' Committee on Terminology, of which he is the Chair. The traveler's invited workshop emphasized terminology standardization requirements in databases of material properties as well as practical terminology standardizing methods. The congress included six workshops in addition to approximately 82 lectures and papers from terminologists, artificial intelligence practitioners, and subject specialists from 18 countries. There were approximately 292 registrants from 33 countries who participated in the congress. The congress topics were broad. Examples were the increasing use of International Standards Organization (ISO) Standards in legislated systems such as the USSR Automated Data Bank of Standardized Terminology, the enhanced Physics Training Program based on terminology standardization in Physics in the Chinese province of Inner Mongolia, and the technical concept dictionary being developed at the Japan Electronic Dictionary Research Institute, which is considered to be the key to advanced artificial intelligence applications. The more usual roles of terminology work in the areas of machine translation. indexing protocols, knowledge theory, and data transfer in several subject specialties were also addressed, along with numerous special language terminology areas.

  13. The generation of denatured reactor plutonium by different options of the fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Broeders, C.H.M.; Kessler, G. [Inst. for Neutron Physics and Reactor Technology, Research Center Karlsruhe (Germany)

    2006-11-15

    Denatured (proliferation resistant) reactor plutonium can be generated in a number of different fuel cycle options. First denatured reactor plutonium can be obtained if, instead of low enriched U-235 PWR fuel, re-enriched U-235/U-236 from reprocessed uranium is used (fuel type A). Also the envisaged existing 2,500 t of reactor plutonium (being generated world wide up to the year 2010), mostly stored in intermediate fuel storage facilities at present, could be converted during a transition phase into denatured reactor plutonium by the options fuel type B and D. Denatured reactor plutonium could have the same safeguards standard as present low enriched (<20% U-235) LWR fuel. It could be incinerated by recycling once or twice in PWRs and subsequently by multi-recycling in FRs (CAPRA type or IFRs). Once denatured, such reactor plutonium could remain denatured during multiple recycling. In a PWR, e.g., denatured reactor plutonium could be destroyed at a rate of about 250 kg/GWey. While denatured reactor plutonium could be recycled and incinerated under relieved IAEA safeguards, neptunium would still have to be monitored by the IAEA in future for all cases in which considerable amounts of neptunium are produced. (orig.)

  14. Search for New Physics in reactor and accelerator experiments

    Science.gov (United States)

    Di Iura, A.; Girardi, I.; Meloni, D.

    2016-01-01

    We consider two scenarios of New Physics: the Large Extra Dimensions (LED), where sterile neutrinos can propagate in a (4+d) -dimensional space-time, and the Non Standard Interactions (NSI), where the neutrino interactions with ordinary matter are parametrized at low energy in terms of effective flavour-dependent complex couplings \\varepsilon_{αβ} . We study how these models have an impact on oscillation parameters in reactor and accelerator experiments.

  15. Dynamic Model of an Ammonia Synthesis Reactor Based on Open Information

    OpenAIRE

    Jinasena, Asanthi; Lie, Bernt; Glemmestad, Bjørn

    2016-01-01

    Ammonia is a widely used chemical, hence the ammonia manufacturing process has become a standard case study in the scientific community. In the field of mathematical modeling of the dynamics of ammonia synthesis reactors, there is a lack of complete and well documented models. Therefore, the main aim of this work is to develop a complete and well documented mathematical model for observing the dynamic behavior of an industrial ammonia synthesis reactor system. The mode...

  16. Synthesis of palladium nanoparticles using a continuous flow polymeric micro reactor.

    Science.gov (United States)

    Song, Yujun; Kumar, Challa S S R; Hormes, Josef

    2004-09-01

    A continuous flow polymeric micro reactor, fabricated using a negative photo resist SU-8 on a 10 x 10 cm PEEK (polyetheretherketone) substrate by standard UV lithography, was utilized to synthesize palladium nanoparticles. The nanoparticles were characterized by Transmission Electron Microscopy, Selected Area Electron Diffraction and X-ray Diffraction. The Pd nanoparticles synthesized in the micro reactor were found to have a narrower size distribution when compared with those obtained by the conventional batch process.

  17. Nuclear reactor decommissioning. (Latest citations from the NTIS bibliographic database). Published Search

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    The bibliography contains citations concerning nuclear power and research reactor decommissioning and decontamination plans, costs, and safety standards. References discuss the design and evaluation of protective confinement, entombment, and dismantling systems. Topics include decommissioning regulations and rules, public and occupational radiation exposure estimates, comparative evaluation, and reactor performance under high neutron flux conditions. Waste packaging and disposal, environmental compliance, and public opinion are examined. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  18. A review of qualitative inspection aspects of end fittings in an Indian pressurized heavy water reactor

    OpenAIRE

    Urva Pancholi; Dhaval Dave; Ajay Patel

    2016-01-01

    The paper provides a summarized description of the current state of knowledge and practices used in India, in the qualitative inspection of end fittings – a key component of the fuel channel assembly of a pressurized heavy water reactor (PHWR), generally of a Canadian Deuterium Uranium (CANDU) type. Further it discusses various quality inspection techniques; and the high standards and mechanical precision of the job required, to be accepted as viable nuclear reactor component. The techniqu...

  19. Plasma reactor waste management systems

    Science.gov (United States)

    Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

    1992-01-01

    The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

  20. Nuclear Reactor Engineering Analysis Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-12-31

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels.

  1. Utilisation of thorium in reactors

    Science.gov (United States)

    Anantharaman, K.; Shivakumar, V.; Saha, D.

    2008-12-01

    India's nuclear programme envisages a large-scale utilisation of thorium, as it has limited deposits of uranium but vast deposits of thorium. The large-scale utilisation of thorium requires the adoption of closed fuel cycle. The stable nature of thoria and the radiological issues associated with thoria poses challenges in the adoption of a closed fuel cycle. A thorium fuel based Advanced Heavy Water Reactor (AHWR) is being planned to provide impetus to development of technologies for the closed thorium fuel cycle. Thoria fuel has been loaded in Indian reactors and test irradiations have been carried out with (Th-Pu) MOX fuel. Irradiated thorium assemblies have been reprocessed and the separated 233U fuel has been used for test reactor KAMINI. The paper highlights the Indian experience with the use of thorium and brings out various issues associated with the thorium cycle.

  2. A tubular focused sonochemistry reactor

    Institute of Scientific and Technical Information of China (English)

    ZHOU GuangPing; LIANG ZhaoFeng; LI ZhengZhong; ZHANG YiHui

    2007-01-01

    This paper presents a new sonochemistry reactor, which consists of a cylindrical tube with a certain length and piezoelectric transducers at tube's end with the longitudinal vibration. The tube can effectively transform the longitudinal vibration into the radial vibration and thereby generates ultrasound. Furthermore, ultrasound can be focused to form high-intensity ultrasonic field inside tube. The reactor boasts of simple structure and its whole vessel wall can radiate ultrasound so that the electroacoustic transfer efficiency is high. The focused ultrasonic field provides good condition for sonochemical reaction. The length of the reactor can be up to 2 meters, and liquids can pass through it continuously, so it can be widely applied in liquid processing such as sonochemistry.

  3. A compact Tokamak transmutation reactor

    Institute of Scientific and Technical Information of China (English)

    QiuLi-Jian; XiaoBing-Jia

    1997-01-01

    The low aspect ration tokamak is proposed for the driver of a transmutation reactor.The main parameters of the reactor core,neutronic analysis of the blanket are given>the neutron wall loading can be lowered from the magnitude order of 1 MW/m2 to 0.5MW/m2 which is much easier to reach in the near future,and the transmutation efficiency (fission/absorption ratio)is raised further.The blanket power density is about 200MW/m3 which is not difficult to deal with.The key components such as diverter and center conductor post are also designed and compared with conventional TOkamak,Finally,by comparison with the other drivers such as FBR,PWR and accelerator,it can be anticipated that the low aspect ratio transmutation reactor would be one way of fusion energy applications in the near future.

  4. Investigation of KW reactor incident

    Energy Technology Data Exchange (ETDEWEB)

    Sturges, D G [USAEC Hanford Operations Office, Richland, WA (United States); Hauff, T W; Greager, O H [General Electric Co., Richland, WA (United States). Hanford Atomic Products Operation

    1955-02-11

    The new KW reactor was placed in operation on January 4, 1955, and had been running at relatively low power levels for only 17 hours when it was shut down because of a process tube water leak which appeared to be associated with a slug rupture. After several days of unrewarding effort to remove the slugs and tube by customary methods, it developed that considerable melting of the tube and slugs had taken place. It was then evident that removal of the stuck mass and repairs to the damaged tube channel would require unusual measures that were certain to extend the reactor outage for several weeks. This report documents the work and findings of the Committee which investigated the KW reactor incident. Its content represents unanimous agreement among the three Committee members.

  5. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  6. PITR: Princeton Ignition Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection.

  7. Analysis of Adiabatic Batch Reactor

    Directory of Open Access Journals (Sweden)

    Erald Gjonaj

    2016-05-01

    Full Text Available A mixture of acetic anhydride is reacted with excess water in an adiabatic batch reactor to form an exothermic reaction. The concentration of acetic anhydride and the temperature inside the adiabatic batch reactor are calculated with an initial temperature of 20°C, an initial temperature of 30°C, and with a cooling jacket maintaining the temperature at a constant of 20°C. The graphs of the three different scenarios show that the highest temperatures will cause the reaction to occur faster.

  8. External fuel thermionic reactor system.

    Science.gov (United States)

    Mondt, J. F.; Peelgren, M. L.

    1971-01-01

    Thermionic reactors are prime candidates for nuclear electric propulsion. The national thermionic reactor effort is concentrated on the flashlight concept with the external-fuel concept as the backup. The external-fuel concept is very adaptable to a completely modular power subsystem which is attractive for highly reliable long-life applications. The 20- to 25-cm long, externally-fueled converters have been designed, fabricated, and successfully tested with many thermal cycles by electrical heating. However, difficulties have been encountered during encapsulation for nuclear heated tests and none have been started to date. These nuclear tests are required to demonstrate the concept feasibility.

  9. Reactor shutdown delays medical procedures

    Science.gov (United States)

    Gwynne, Peter

    2008-01-01

    A longer-than-expected maintenance shutdown of the Canadian nuclear reactor that produces North America's entire supply of molybdenum-99 - from which the radioactive isotopes technetium-99 and iodine-131 are made - caused delays to the diagnosis and treatment of thousands of seriously ill patients last month. Technetium-99 is a key component of nuclear-medicine scans, while iodine-131 is used to treat cancer and other diseases of the thyroid. Production eventually resumed, but only after the Canadian government had overruled the Canadian Nuclear Safety Commission (CNSC), which was still concerned about the reactor's safety.

  10. Reactor Technology Options Study for Near-Term Deployment of GNEP Grid-Appropriate Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, Daniel T [ORNL; Poore III, Willis P [ORNL

    2007-09-01

    World energy demand is projected to significantly increase over the coming decades. The International Energy Agency projects that electricity demand will increase 50% by 2015 and double by 2030, with most of the increase coming in developing countries as they experience double-digit rates of economic growth and seek to improve their standards of living. Energy is the necessary driver for human development, and the demand for energy in these countries will be met using whatever production technologies are available. Recognizing this inevitable energy demand and its implications for the United States, the U.S. National Security Strategy has proposed the Global Nuclear Energy Partnership (GNEP) to work with other nations to develop and deploy advanced nuclear recycling and reactor technologies. This initiative will help provide reliable, emission-free energy with less of the waste burden of older technologies and without making available separated plutonium that could be used by rogue states or terrorists for nuclear weapons. These new technologies will make possible a dramatic expansion of safe, clean nuclear energy to help meet the growing global energy demand. In other words, GNEP seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy without increasing the risk of nuclear weapon proliferation. This global expansion of nuclear power is strategically important to the United States for several reasons, including the following: (1) National security, by reducing the competition and potential for conflict over increasingly scarce fossil energy resources; (2) Economic security, by helping maintain stable prices for nonrenewable resources such as oil, gas, and coal; (3) Environmental security, by replacing or off-setting large-scale burning of greenhouse gas-emitting fuels for electricity production; and (4) Regaining technical leadership, through deployment of innovative U.S. technology-based reactors. Fully meeting

  11. Designing a mini subcritical nuclear reactor; Diseno de un mini reactor nuclear subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M., E-mail: rafelaescobedo@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)

    2015-10-15

    In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of {sup 239}PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of {sup 239}PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k{sub e}-f{sub f}, the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k{sub eff}, the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons {sup 239}PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k{sub eff} was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k{sub eff} of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five

  12. 葡甘聚糖/蜂蜡复合液膜对鲜切琯溪蜜柚保鲜研究%Study on fresh-cut pomelo fresh-keeping by konjac glucomannan(KGM)-beeswax(BW)composite membrane

    Institute of Scientific and Technical Information of China (English)

    赖明耀; 兰润; 李雪晖; 庞杰

    2014-01-01

    该文研究室温(20℃)条件下葡甘聚糖/蜂蜡(KGM/BW)涂膜、葡甘聚糖(KGM)涂膜及对照组对鲜切琯溪蜜柚贮藏期间主要感观品质和理化性质影响。结果表明:KGM涂膜,尤其是KGM/BW涂膜能降低贮藏期间鲜切柚子的水分损失、抑制呼吸作用、减缓营养物质流失,具有良好的保鲜效果。%The effects of KGM/BW coating,KGM coating and the control group under the room temperature(20 ℃)on the sensory quality and physicochemical property of the fresh-cut honey pomelo(Guanxi)during postharvest storage were studied.The results showed that KGM/BW coating, compared to KGM coating group and the control,could protect the flesh-cut fruit best during storage from water loss and inhibit respiration as well as preserve the nutritious quality of the honey pomelo fruit.

  13. Using BW Data Warehouse Technology to Construct Enterprise Decision Analysis System%运用BW数据仓库技术构建企业决策分析系统

    Institute of Scientific and Technical Information of China (English)

    胡贵平

    2015-01-01

    数据仓库能够把来自企业内部和外部的大量异构数据按辅助决策主题的要求进行加工、集成,为高层管理人员提供各种类型的、有效的数据分析,成为目前信息处理技术中的研究热点。分析了BW(Busi原ness Information Warehouse,商务信息仓库)的特点,并概要地介绍了马钢整体ERP项目决策分析系统采用该技术实施过程,并对大型企业决策分析系统建立提出了建议和要求。%The data warehouse can process and integrate huge amounts of heterogeneous data coming from the enterprise or outside according to the requirement of auxiliary decision topic, to provide senior management with various kinds of effective data analysis, so it has become a hotspot of research in information processing technology at present. The characteris-tics of the Business Information Warehouse (BW) are analyzed, the implementation process of the ERP project decision analysis system adopting the data warehouse technology at MaSteel is briefly introduced and proposals as well as requirements on establishment of decision anal-ysis system at large enterprises are put forward.

  14. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    Energy Technology Data Exchange (ETDEWEB)

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

  15. Oregon State University TRIGA Reactor annual report

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, T.V.; Johnson, A.G.; Bennett, S.L.; Ringle, J.C.

    1979-08-31

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included.

  16. Reactor Antineutrino Signals at Morton and Boulby

    CERN Document Server

    Dye, Steve

    2016-01-01

    Increasing the distance from which an antineutrino detector is capable of monitoring the operation of a registered reactor, or discovering a clandestine reactor, strengthens the Non-Proliferation of Nuclear Weapons Treaty. This report presents calculations of reactor antineutrino interactions, from quasi-elastic neutrino-proton scattering and elastic neutrino-electron scattering, in a water-based detector operated >10 km from a commercial power reactor. It separately calculates signal from the proximal reactor and background from all other registered reactors. The main results are interaction rates and kinetic energy distributions of charged leptons scattered from quasi-elastic and elastic processes. Comparing signal and background distributions evaluates reactor monitoring capability. Scaling the results to detectors of different sizes, target media, and standoff distances is straightforward. Calculations are for two examples of a commercial reactor (P_th~3 GW) operating nearby (L~20 km) an underground facil...

  17. Frequency standards

    CERN Document Server

    Riehle, Fritz

    2006-01-01

    Of all measurement units, frequency is the one that may be determined with the highest degree of accuracy. It equally allows precise measurements of other physical and technical quantities, whenever they can be measured in terms of frequency.This volume covers the central methods and techniques relevant for frequency standards developed in physics, electronics, quantum electronics, and statistics. After a review of the basic principles, the book looks at the realisation of commonly used components. It then continues with the description and characterisation of important frequency standards

  18. Transmutation of actinides in power reactors.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  19. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG

  20. Current drive for stability of thermonuclear plasma reactor

    Science.gov (United States)

    Amicucci, L.; Cardinali, A.; Castaldo, C.; Cesario, R.; Galli, A.; Panaccione, L.; Paoletti, F.; Schettini, G.; Spigler, R.; Tuccillo, A.

    2016-01-01

    To produce in a thermonuclear fusion reactor based on the tokamak concept a sufficiently high fusion gain together stability necessary for operations represent a major challenge, which depends on the capability of driving non-inductive current in the hydrogen plasma. This request should be satisfied by radio-frequency (RF) power suitable for producing the lower hybrid current drive (LHCD) effect, recently demonstrated successfully occurring also at reactor-graded high plasma densities. An LHCD-based tool should be in principle capable of tailoring the plasma current density in the outer radial half of plasma column, where other methods are much less effective, in order to ensure operations in the presence of unpredictably changes of the plasma pressure profiles. In the presence of too high electron temperatures even at the periphery of the plasma column, as envisaged in DEMO reactor, the penetration of the coupled RF power into the plasma core was believed for long time problematic and, only recently, numerical modelling results based on standard plasma wave theory, have shown that this problem should be solved by using suitable parameter of the antenna power spectrum. We show here further information on the new understanding of the RF power deposition profile dependence on antenna parameters, which supports the conclusion that current can be actively driven over a broad layer of the outer radial half of plasma column, thus enabling current profile control necessary for the stability of a reactor.

  1. Instrumentation, Monitoring and NDE for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Leonard J.; Doctor, Steven R.; Bunch, Kyle J.; Good, Morris S.; Waltar, Alan E.

    2007-07-28

    The Global Nuclear Energy Partnership (GNEP) has been proposed as a viable system in which to close the fuel cycle in a manner consistent with markedly expanding the global role of nuclear power while significantly reducing proliferation risks. A key part of this system relies on the development of actinide transmutation, which can only be effectively accomplished in a fast-spectrum reactor. The fundamental physics for fast reactors is well established. However, to achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required--during both fabrication and operation. Since the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor – II (EBR-II) reactors were operational in the USA, there have been major advances in instrumentation, not the least being the move to digital systems. Some specific capabilities have been developed outside the USA, but new or at least re-established capabilities will be required. In many cases the only available information is in reports and papers. New and improved sensors and instrumentation will be required. Advanced instrumentation has been developed for high-temperature/high-flux conditions in some cases, but most of the original researchers and manufacturers are retired or no longer in business.

  2. Small modular reactor (SMR) development plan in Korea

    Science.gov (United States)

    Shin, Yong-Hoon; Park, Sangrok; Kim, Byong Sup; Choi, Swongho; Hwang, Il Soon

    2015-04-01

    Since the first nuclear power was engaged in Korean electricity grid in 1978, intensive research and development has been focused on localization and standardization of large pressurized water reactors (PWRs) aiming at providing Korean peninsula and beyond with economical and safe power source. With increased priority placed on the safety since Chernobyl accident, Korean nuclear power R&D activity has been diversified into advanced PWR, small modular PWR and generation IV reactors. After the outbreak of Fukushima accident, inherently safe small modular reactor (SMR) receives growing interest in Korea and Europe. In this paper, we will describe recent status of evolving designs of SMR, their advantages and challenges. In particular, the conceptual design of lead-bismuth cooled SMR in Korea, URANUS with 40˜70 MWe is examined in detail. This paper will cover a framework of the program and a strategy for the successful deployment of small modular reactor how the goals would entail and the approach to collaboration with other entities.

  3. Laminar Entrained Flow Reactor (Fact Sheet)

    Energy Technology Data Exchange (ETDEWEB)

    2014-02-01

    The Laminar Entrained Flow Reactor (LEFR) is a modular, lab scale, single-user reactor for the study of catalytic fast pyrolysis (CFP). This system can be employed to study a variety of reactor conditions for both in situ and ex situ CFP.

  4. Some new viewpoints in reactor noise analysis

    Institute of Scientific and Technical Information of China (English)

    罗征培; 李富; 等

    1996-01-01

    It is propsed that the linearity criterion and order criterion via frequency spectrum features without any limitation of the model's phase can be used in reactor noise analysis.The time constant,natural frequency as well as the recovered transfer function of reactors can bhe obtained via the analyzable model based on reactor noise.

  5. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

  6. Heavy Water Reactor; Reacteurs a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Yu, St.; HOpwood, J.; Meneley, D. [Energie Atomique du Canada (Canada)

    2000-04-01

    This document deals with the Heavy Water Reactor (HWR) technology and especially the Candu (Canada Deuterium Uranium) reactor. This reactors type offers many advantages that promote them for the future. General concepts, a description of the Candu nuclear power plants, the safety systems, the fuel cycle and economical and environmental aspects are included. (A.L.B.)

  7. Operating Modes Of Chemical Reactors Of Polymerization

    Directory of Open Access Journals (Sweden)

    Meruyert Berdieva

    2012-05-01

    Full Text Available In the work the issues of stable technological modes of operation of main devices of producing polysterol reactors have been researched as well as modes of stable operation of a chemical reactor have been presented, which enables to create optimum mode parameters of polymerization process, to prevent emergency situations of chemical reactor operation in industrial conditions.

  8. Nuclear Reactors and Technology; (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  9. A Simple Tubular Reactor Experiment.

    Science.gov (United States)

    Hudgins, Robert R.; Cayrol, Bertrand

    1981-01-01

    Using the hydrolysis of crystal violet dye by sodium hydroxide as an example, the theory, apparatus, and procedure for a laboratory demonstration of tubular reactor behavior are described. The reaction presented can occur at room temperature and features a color change to reinforce measured results. (WB)

  10. Silica-Immobilized Enzyme Reactors

    Science.gov (United States)

    2007-08-01

    immobilized artificial membrane chromatography and lysophospholipid micellar electrokinetic chromatography . J. Chromatogr. A 1998, 810, 95-103. 50...Journal of Liquid Chromatography and Related Technologies. Air Force Research Laboratory Materials and Manufacturing Directorate Airbase...immobilized enzyme reactors (IMERs) can also be integrated directly to further analytical methods such as liquid chromatography or mass spectrometry.[6] In

  11. British high flux beam reactor.

    Science.gov (United States)

    Egelstaff, P A

    1970-10-24

    The neutron scattering technique has become an accepted method for the study of condensed matter. Because of the great scientific and technical value of neutron experiments and the growing body of users, several proposals have been made during the past decade for a nuclear reactor devoted primarily to this technique. This article reviews the reasons for and history behind these proposals.

  12. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

  13. Heterogeneous Recycling in Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forget, Benoit; Pope, Michael; Piet, Steven J.; Driscoll, Michael

    2012-07-30

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  14. Neutrino Mixing Discriminates Geo-reactor Models

    CERN Document Server

    Dye, S T

    2009-01-01

    Geo-reactor models suggest the existence of natural nuclear reactors at different deep-earth locations with loosely defined output power. Reactor fission products undergo beta decay with the emission of electron antineutrinos, which routinely escape the earth. Neutrino mixing distorts the energy spectrum of the electron antineutrinos. Characteristics of the distorted spectrum observed at the earth's surface could specify the location of a geo-reactor, discriminating the models and facilitating more precise power measurement. The existence of a geo-reactor with known position could enable a precision measurement of the neutrino oscillation parameter delta-mass-squared.

  15. Reactor monitoring and safeguards using antineutrino detectors

    CERN Document Server

    Bowden, N S

    2008-01-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

  16. Reactor assessments of advanced bumpy torus configurations

    Energy Technology Data Exchange (ETDEWEB)

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1983-01-01

    Recently, several configurational approaches and concept improvement schemes were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These configurations include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator-snakey torus). Preliminary evaluations of reactor implications of each of these configurations have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties. Results indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past.

  17. Detection of antineutrinos for reactor monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeong Duk [Center for Underground Physics, Institute of Basic Science, Daejeon (Korea, Republic of)

    2016-04-15

    Reactor neutrinos have been detected in the past 50 years by various detectors for different purposes. Beginning in the 1980s, neutrino physicists have tried to use neutrinos to monitor reactors and develop an optimized detector for nuclear safeguards. Recently, motivated by neutrino oscillation physics, the technology and scale of reactor neutrino detection have progressed considerably. In this review, I will give an overview of the detection technology for reactor neutrinos, and describe the issues related to further improvements in optimized detectors for reactor monitoring.

  18. Use of standard spectra for the short life radionuclides and ratios for long life radionuclides in the wastes of EDF PWR type reactors; Utilisation de spectres types pour les radionucleides a vie courte et de ratios pour les radionucleides a vie longue dans les dechets de REP EDF

    Energy Technology Data Exchange (ETDEWEB)

    Lantes, B. [Electricite de France (EDF-DPN/Groupe Environnement), 31 - Toulouse (France); Bienvenu, Ph. [CEA Cadarache, Dept. d' Etudes des Dechets, DED, 13 - Saint-Paul-lez-Durance (France)

    2001-07-01

    This paper presents the type of declaration of radioactivity in the wastes of PWR type reactors park. Particularly, it insists on the justification of use of spectra for the declaration of short live radionuclides. It tackles the important developments of methods and measures of radiochemical analysis made by the Cea in order to determine the ratios to declare the long life radioisotopes. (N.C.)

  19. Strategies for reactor safety. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, K

    1997-12-01

    The NKS/RAK-1 project formed part of a four-year nuclear research program (1994-1997) in the Nordic countries, the NKS Programme. The project aims were to investigate and evaluate the safety work, to increase realism and reliability of the safety analysis, and to give ideas for how safety can be improved in selected areas. An evaluation of the safety work in nuclear installations in Finland and Sweden was made, and a special effort was devoted to plant modernisation and to see how modern safety standards can be met up with. A combination of more resources and higher efficiency is recommended to meet requirements from plant modernisation and plant renovations. Both the utilities and the safety authorities are recommended to actively follow the evolving safety standards for new reactors. Various approaches to estimating LOCA frequencies have been explored. In particular, a probabilistic model for pipe ruptures due to intergranular stress corrosion has been developed. A survey has been done over methodologies for integrated sequence analysis (ISA), and different approaches have been developed and tested on four sequences. Structured frameworks for integration between PSA and behavioural sciences have been developed, which e.g. have improved PSA. The status of maintenance strategies in Finland and Sweden has been studied and a new maintenance data information system has been developed. (au) 41 refs.

  20. CFD Simulation on Ethylene Furnace Reactor Tubes

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Different mathematical models for ethylene furnace reactor tubes were reviewed. On the basis of these models a new mathematical simulation approach for reactor tubes based on computational fluid dynamics (CFD) technique was presented. This approach took the flow, heat transfer, mass transfer and thermal cracking reactions in the reactor tubes into consideration. The coupled reactor model was solved with the SIMPLE algorithm. Some detailed information about the flow field, temperature field and concentration distribution in the reactor tubes was obtained, revealing the basic characteristics of the hydrodynamic phenomena and reaction behavior in the reactor tubes. The CFD approach provides the necessary information for conclusive decisions regarding the production optimization, the design and improvement of reactor tubes, and the new techniques implementation.

  1. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  2. In-reactor performance of pressure tubes in CANDU reactors

    Science.gov (United States)

    Rodgers, D. K.; Coleman, C. E.; Griffiths, M.; Bickel, G. A.; Theaker, J. R.; Muir, I.; Bahurmuz, A. A.; Lawrence, S. St.; Resta Levi, M.

    2008-12-01

    The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate Vc, and threshold stress intensity factor, KIH) change with irradiation; the former reach a limiting value at a fluence of Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

  3. Advanced CANDU reactor pre-licensing progress

    Energy Technology Data Exchange (ETDEWEB)

    Popov, N.K.; West, J.; Snell, V.G.; Ion, R.; Archinoff, G.; Xu, C. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada)]. E-mail: popovn@aecl.ca

    2005-07-01

    The Advanced CANDU Reactor (ACR) is an evolutionary advancement of the current CANDU 6 reactor, aimed at producing electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The Canadian Nuclear Safety Commission (CNSC) staff are currently reviewing the ACR design to determine whether, in their opinion, there are any fundamental barriers that would prevent the licensing of the design in Canada. This CNSC licensability review will not constitute a licence, but is expected to reduce regulatory risk. The CNSC pre-licensing review started in September 2003, and was focused on identifying topics and issues for ACR-700 that will require a more detailed review. CNSC staff reviewed about 120 reports, and issued to AECL 65 packages of questions and comments. Currently CNSC staff is reviewing AECL responses to all packages of comments. AECL has recently refocused the design efforts to the ACR-1000, which is a larger version of the ACR design. During the remainder of the pre-licensing review, the CNSC review will be focused on the ACR-1000. AECL Technologies Inc. (AECLT), a wholly-owned US subsidiary of AECL, is engaged in a pre-application process for the ACR-700 with the US Nuclear Regulatory Commission (USNRC) to identify and resolve major issues prior to entering a formal process to obtain standard design certification. To date, the USNRC has produced a Pre-Application Safety Assessment Report (PASAR), which contains their reviews of key focus topics. During the remainder of the pre-application phase, AECLT will address the issues identified in the PASAR. Pursuant to the bilateral agreement between AECL and the Chinese nuclear regulator, the National Nuclear Safety Administration (NNSA) and its Nuclear Safety Center (NSC), NNSA/NSC are reviewing the ACR in seven focus areas. The review started in September 2004, and will take three years. The main objective of the review is to determine how the ACR complies

  4. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  5. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  6. Performance of down-flow hanging sponge (DHS) reactor coupled with up-flow anaerobic sludge blanket (UASB) reactor for treatment of onion dehydration wastewater.

    Science.gov (United States)

    El-Kamah, Hala; Mahmoud, Mohamed; Tawfik, Ahmed

    2011-07-01

    In this study, a promising system consisting of up-flow anaerobic sludge blanket (UASB) reactor followed by down-flow hanging sponge (DHS) reactor was investigated for onion dehydration wastewater treatment. Laboratory experiments were conducted at two different phases, i.e., phase (1) at overall hydraulic retention time (HRT) of 11h (UASB reactor: 6h and DHS reactor: 5h) and phase (2) at overall HRT of 9.4h (UASB reactor: 5.2h and DHS reactor: 4.2h). Long-term operation results of the proposed system showed that its overall TCOD, TBOD, TSS, TKN and NH(4)-N removal efficiencies were 92 ± 5, 95 ± 2, 95 ± 2, 72 ± 6 and 99 ± 1.3%, respectively (phase 1). Corresponding values for the 2nd phase were 85.4 ± 5, 86 ± 3, 87 ± 6, 65 ± 8 and 95 ± 2.8%. Based on the available results, the proposed system could be more viable option for treatment of wastewater generated from onion dehydration industry in regions with tropical or sub-tropical climates and with stringent discharge standards.

  7. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    Science.gov (United States)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

  8. Nuclear reactor pulse calibration using a CdZnTe electro-optic radiation detector

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Kyle A., E-mail: knelson1@ksu.edu [S.M.A.R.T. Laboratory, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States); Geuther, Jeffrey A. [TRIGA Mark II Nuclear Reactor, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States); Neihart, James L.; Riedel, Todd A. [S.M.A.R.T. Laboratory, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States); Rojeski, Ronald A. [Nanometrics, Inc., 1550 Buckeye Drive, Milpitas, CA 95035 (United States); Saddler, Jeffrey L. [TRIGA Mark II Nuclear Reactor, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States); Schmidt, Aaron J.; McGregor, Douglas S. [S.M.A.R.T. Laboratory, Mechanical and Nuclear Engineering, Kansas State University, Manhattan, KS 66506 (United States)

    2012-07-15

    A CdZnTe electro-optic radiation detector was used to calibrate nuclear reactor pulses. The standard configuration of the Pockels cell has collimated light passing through an optically transparent CdZnTe crystal located between crossed polarizers. The transmitted light was focused onto an IR sensitive photodiode. Calibrations of reactor pulses were performed using the CdZnTe Pockels cell by measuring the change in the photodiode current, repeated 10 times for each set of reactor pulses, set between 1.00 and 2.50 dollars in 0.50 increments of reactivity. - Highlights: Black-Right-Pointing-Pointer We demonstrated the first use of an electro-optic device to trace reactor pulses in real-time. Black-Right-Pointing-Pointer We examined the changes in photodiode current for different reactivity insertions. Black-Right-Pointing-Pointer Created a linear best fit line from the data set to predict peak pulse powers.

  9. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

    Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

  10. CFD Simulation of a Hydrogen/Argon Plasma Jet Reactor for Coal Pyrolysis

    Institute of Scientific and Technical Information of China (English)

    CHEN H. G.; XIE K. C.

    2004-01-01

    A Computational Fluid Dynamics (CFD) model was formulated for DC arc hydrogen/argon plasma jet reactors used in the process of the thermal H2/Ar plasma pyrolysis of coal to acetylene. In this model, fluid flow, convective heat transfer and conjugate heat conductivity are considered simultaneously. The error caused by estimating the inner-wall temperature of a reactor is avoided. The thermodynamic and transport properties of the hydrogen/argon mixture plasma system, which are usually expressed by a set of discrete dats, are fitted into expressions that can be easily implemented in the program. The effects of the turbulence are modeled by two standard k-s equations. The temperature field and velocity field in the plasma jet reactor were calculated by employing SIMPLEST algorithm. The knowledge and insight obtained are useful for the design improvement and scale-up of plasma reactors.

  11. Results from the Daya Bay Reactor Neutrino Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Tsang, K.V. [Lawrence Berkeley National Laboratory, Berkeley, CA (United States); An, F.P. [Institute of High Energy Physics, Beijing (China); An, Q. [University of Science and Technology of China, Hefei (China); Bai, J.Z. [Institute of High Energy Physics, Beijing (China); Balantekin, A.B.; Band, H.R. [University of Wisconsin, Madison, WI (United States); Beriguete, W.; Bishai, M. [Brookhaven National Laboratory, Upton, NY (United States); Blyth, S. [National United University, Miao-Li (China); Brown, R.L. [Brookhaven National Laboratory, Upton, NY (United States); Cao, G.F.; Cao, J. [Institute of High Energy Physics, Beijing (China); Carr, R. [California Institute of Technology, Pasadena, CA (United States); Chan, W.T. [Brookhaven National Laboratory, Upton, NY (United States); Chang, J.F. [Institute of High Energy Physics, Beijing (China); Chang, Y. [National United University, Miao-Li (China); Chasman, C. [Brookhaven National Laboratory, Upton, NY (United States); Chen, H.S. [Institute of High Energy Physics, Beijing (China); Chen, H.Y. [Institute of Physics, National Chiao-Tung University, Hsinchu (China); Chen, S.J. [Nanjing University, Nanjing (China); and others

    2014-01-15

    The Daya Bay Reactor Neutrino Experiment was designed to achieve a sensitivity on the value of sin{sup 2}2θ{sub 13} to better than 0.01 at 90% CL. The experiment consists of eight antineutrino detectors installed underground at different baselines from six nuclear reactors. With data collected with six antineutrino detectors for 55 days, Daya Bay announced the discovery of a non-zero value for sin{sup 2}2θ{sub 13} with a significance of 5.2 standard deviations in March 2012. The most recent analysis with 139 days of data acquired in a six-detector configuration yields sin{sup 2}2θ{sub 13}=0.089±0.010(stat.)±0.005(syst.), which is the most precise measurement of sin{sup 2}2θ{sub 13} to date.

  12. Fluidized bed coal combustion reactor

    Science.gov (United States)

    Moynihan, P. I.; Young, D. L. (Inventor)

    1981-01-01

    A fluidized bed coal reactor includes a combination nozzle-injector ash-removal unit formed by a grid of closely spaced open channels, each containing a worm screw conveyor, which function as continuous ash removal troughs. A pressurized air-coal mixture is introduced below the unit and is injected through the elongated nozzles formed by the spaces between the channels. The ash build-up in the troughs protects the worm screw conveyors as does the cooling action of the injected mixture. The ash layer and the pressure from the injectors support a fluidized flame combustion zone above the grid which heats water in boiler tubes disposed within and/or above the combustion zone and/or within the walls of the reactor.

  13. Nuclear reactor alignment plate configuration

    Energy Technology Data Exchange (ETDEWEB)

    Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

    2014-01-28

    An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

  14. Transport simulation for EBT reactor

    Energy Technology Data Exchange (ETDEWEB)

    Uckan, T.; Uckan, N.A.; Jaeger, E.F.

    1983-08-01

    Transport simulation and modeling studies for the ELMO Bumpy Torus (EBT) reactor are carried out by using zero-dimensional (0-D) and one-and-one-half-dimensional (1 1/2-D) transport calculations. The time-dependent 0-D model is used for global analysis, whereas the 1 1/2-D radial transport code is used for accurate determination of density, temperature, and ambipolar potential profiles and of the role of these profiles in reactor plasma performance. Analysis with the 1 1/2-D transport code shows that profile effects near the outer edge of the hot electron ring lead to enhanced confinement by at least a factor of 2 to 5 beyond the simple scaling that is obtained from the global analysis. The radial profiles of core plasma density and temperatures (or core pressure) obtained from 1 1/2-D transport calculations are found to be similar to those theoretically required for stability.

  15. Gas-liquid autoxidation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morbidelli, M.; Paludetto, R.; Carra, S.

    1986-01-01

    A procedure for the simulation of autoxidation gas-liquid reactors has been developed based both on mathematical models and laboratory experiments. It has been shown that the complex radical chain mechanism of the autoxidation process can be simulated through two global parallel reactions, whose rates are obtained by assuming pseudo-steady-state concentration values for all the radical species involved. Using ethylbenzene autoxidation as a model reaction, an experimental analysis has been performed in order to estimate all the kinetic parameters of the model. The effect of the interaction between gas-liquid mass-transfer phenomena and the complex kinetic mechanism on the overall performance of an autoxidation reactor has been examined in detail within the framework of the liquid film model.

  16. Fast breeder reactor protection system

    Science.gov (United States)

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  17. Standard deviations

    CERN Document Server

    Smith, Gary

    2015-01-01

    Did you know that having a messy room will make you racist? Or that human beings possess the ability to postpone death until after important ceremonial occasions? Or that people live three to five years longer if they have positive initials, like ACE? All of these ‘facts' have been argued with a straight face by researchers and backed up with reams of data and convincing statistics.As Nobel Prize-winning economist Ronald Coase once cynically observed, ‘If you torture data long enough, it will confess.' Lying with statistics is a time-honoured con. In Standard Deviations, ec

  18. Reactor vessel lower head integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  19. Department of Energy's High Flux Beam Reactor (HFBR), September 15--19, 1980: An independent on-site safety review

    Energy Technology Data Exchange (ETDEWEB)

    1981-02-01

    The intent of this on-site safety review was to make a broad management assessment of HFBR operations, rather than conduct a detailed in-depth audit. The result of the review should only be considered as having identified trends or indications. The Team's observations and recommendations for the most part are based upon licensed reactor facility practices used to meet industry standards. These standards form the basis for many of the comments in this report. The Team believes that a uniform minimum standard of performance should be achieved in the operation of DOE reactors. In order to assure that this is accomplished, clear standards are necessary. Consistent with the past AEC and ERDA policy, the team has used the standards of the commercial nuclear power industry. It is recognized that this approach is conservative in that the HFBR reactor has a significantly greater degree of inherent safety (low pressure, temperature, power, etc.) than a licensed reactor.

  20. PR-EDB: Power Reactor Embrittlement Database - Version 3

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

    2008-03-01

    The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

  1. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  2. Actinide transmutation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bultman, J.H.

    1995-01-17

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP).

  3. Characterization of radioactive aerosols in Tehran research reactor containment

    Directory of Open Access Journals (Sweden)

    Moradi Gholamreza

    2015-01-01

    Full Text Available The objectives of this research were to determine the levels of radioactivity in the Tehran research reactor containment and to investigate the mass-size distribution, composition, and concentration of radionuclides during operation of the reactor. A cascade impactor sampler was used to determine the size-activity distributions of radioactive aerosols in each of the sampling stations. Levels of a and b activities were determined based on a counting method using a liquid scintillation counter and smear tests. The total average mass fractions of fine particles (particle diameter dp < 1 mm in all of the sampling stations were approximately 26.75 %, with the mean and standard deviation of 52.15 ± 19.75 mg/m3. The total average mass fractions of coarse particles were approximately 73.2%, with the mean and standard deviation of 71.34 ± 24.57 mg/m3. In addition to natural radionuclides, artificial radionuclides, such as 24Na, 91Sr, 131I, 133I, 103Ru, 82Br, and 140La, may be released into the reactor containment structure. Maximum activity was associated with accumulation-mode particles with diameters less than 400 nm. The results obtained from liquid scintillation counting suggested that the mean specific activity of alpha particles in fine and coarse-modes were 89.7 % and 10.26 %, respectively. The mean specific activity of beta particles in fine and coarse-modes were 81.15 % and 18.51 %, respectively. A large fraction of the radionuclides' mass concentration in the Tehran research reactor containment was associated with coarse-mode particles, in addition, a large fraction of the activity in the aerosol particles was associated with accumulation-mode particles.

  4. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  5. Investigation of molten salt fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kubota, Kenichi; Konomura, Mamoru [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2002-05-01

    On survey research for practicability strategy of fast reactor (FR) (phase 1), to extract future practicability image candidates of FR from wide options, in addition to their survey and investigation objects of not only solid fuel reactors of conventional research object but also molten salt reactor as a flowing fuel reactor, investigation on concept of molten salt FR plant was carried out. As a part of the first step of the survey research for practicability strategy, a basic concept on plant centered at nuclear reactor facility using chloride molten salt reactor capable of carrying out U-Pu cycle was examined, to perform a base construction to evaluate economical potential for a practical FBR. As a result, a result could be obtained that because of inferior fuel inventory and heat transmission to those in Na cooling reactor in present knowledge, mass of reactor vessel and intermediate heat exchanger were to widely increased to expect reduction of power generation unit price even on considering cheapness of its fuel cycle cost. Therefore, at present step further investigation on concept design of the chloride molten salt reactor plant system is too early in time, and it is at a condition where basic and elementary researches aiming at upgrading of economical efficiency such as wide reduction of fuel inventory, a measure expectable for remarkable rationalization effect of reprocessing system integrating a reactor to a processing facility, and so on. (G.K.)

  6. Event management in research reactors; Gestion de eventos en reactores de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Perrin, C.D. [Coordinador Reactores de Investigacion y Conjuntos Criticos, Autoridad Regulatoria Nuclear (Argentina)]. e-mail: cperrin@sede.arn.gov.ar

    2006-07-01

    In the Radiological and Nuclear Safety field, the Nuclear Regulatory Authority of Argentina controls the activities of three investigation reactors and three critical groups, by means of evaluations, audits and inspections, in order to assure the execution of the requirements settled down in the Licenses of the facilities, in the regulatory standards and in the documentation of mandatory character in general. In this work one of the key strategies developed by the ARN to promote an appropriate level of radiological and nuclear safety, based on the control of the administration of the abnormal events that its could happen in the facilities is described. The established specific regulatory requirements in this respect and the activities developed in the entities operators are presented. (Author)

  7. A toxicological safety assessment of a standardized extract of Sceletium tortuosum (Zembrin®) in rats.

    Science.gov (United States)

    Murbach, Timothy S; Hirka, Gábor; Szakonyiné, Ilona Pasics; Gericke, Nigel; Endres, John R

    2014-12-01

    A well-characterized standardized hydroethanolic extract of a traditionally recognized mak (mild) variety of Sceletium tortuosum, a South African plant with a long history of traditional ingestion, is marketed under the trade name Zembrin(®) as an ingredient for use in functional foods and dietary supplements. It is standardized to contain 0.35-0.45% total alkaloids (mesembrenone and mesembrenol ≥60%, and mesembrine <20%). A 14-day repeated oral toxicity study was conducted at 0, 250, 750, 2500, and 5000 mg/kg bw/day. A 90-day subchronic repeated oral toxicity study was conducted at 0, 100, 300, 450, and 600 mg/kg bw/day. Because S. tortuosum has a long history of human use for relieving stress and calming, a functional observation battery, including spontaneous locomotor activity measured using LabMaster ActiMot light-beam frames system, was employed. Several parameters, such as locomotion, rearing behavior, spatial parameters, and turning behavior were investigated in the final week of the study. No mortality or treatment-related adverse effects were observed in male or female Crl:(WI)BR Wistar rats in the 14- or 90-day studies. In the 14- and 90-day studies, the NOAELs were concluded as 5000 and 600 mg/kg bw/d, respectively, the highest dose groups tested.

  8. An approach to model reactor core nodalization for deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  9. An approach to model reactor core nodalization for deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

    2016-01-22

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  10. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  11. Applying moving bed biofilm reactor for removing linear alkylbenzene sulfonate using synthetic media

    Directory of Open Access Journals (Sweden)

    Jalaleddin Mollaei

    2015-01-01

    Full Text Available Detergents and problems of their attendance into water and wastewater cause varied difficulties such as producing foam, abnormality in the growth of algae, accumulation and dispersion in aqueous environments. One of the reactors was designated with 30% of the media with the similar conditions exactly same as the other which had filling rate about 10 %, in order to compare both of them together. A standard method methylene blue active substance was used to measure anionic surfactant. The concentrations of linear alkylbenzene sulfonate which examined were 50, 100, 200, 300 and 400 mg/l in HRT 72, 24 and 8 hrs. The removal percentage for both of reactors at the beginning of operating at50 mg/l concentration of pollutant had a bit difference and with gradually increasing the pollutant concentration and decreasing Hydraulic retention time, the variation between the removal percentage of both reactors became significant as the reactor that had the filling rate about 30 %, showed better condition than the other reactor with 10 % filling rate. Ideal condition in this experiment was caught at hydraulic retention time about 72 hrs and 200 mg/l pollutants concentration with 99.2% removal by the reactor with 30% filling rate. While the ideal condition for the reactor with 10% filling rate with the same hydraulic retention time and 100 mg/l pollutants concentrations was obtained about 99.4% removal. Regarding anionic surfactant standard in Iran which is 1.5 mg/l for surface water discharge, using this process is suitable for treating municipal wastewater and industrial wastewater which has a range of the pollutant between 100-200 mg/l. but for the industries that produce detergents products which make wastewater containing more than 200 mg/l surfactants, using secondary treatment process for achieving discharge standard is required.

  12. Molten-Salt Depleted-Uranium Reactor

    CERN Document Server

    Dong, Bao-Guo; Gu, Ji-Yuan

    2015-01-01

    The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results demonstrate how these reactors can possess and realize extraordinary excellent characteristics, no prompt critical, long-term safe and stable operation with negative feedback, closed uranium-plutonium cycle chain within the vessel, normal operation only with depleted-uranium, and depleted-uranium high burnup in reality, to realize with fission nuclear energy sufficiently satisfying humanity long-term energy resource needs, as well as thoroughly solve the challenges of nuclear criticality safety, uranium resource insuffic...

  13. Performance of a multipurpose research electrochemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Henquin, E.R. [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina); Bisang, J.M., E-mail: jbisang@fiq.unl.edu.ar [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina)

    2011-07-01

    Highlights: > For this reactor configuration the current distribution is uniform. > For this reactor configuration with bipolar connection the leakage current is small. > The mass-transfer conditions are closely uniform along the electrode. > The fluidodynamic behaviour can be represented by the dispersion model. > This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of {+-}10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  14. Sulfide toxicity kinetics of a uasb reactor

    Directory of Open Access Journals (Sweden)

    D. R. Paula Jr.

    2009-12-01

    Full Text Available The effect of sulfide toxicity on kinetic parameters of anaerobic organic matter removal in a UASB (up-flow anaerobic sludge blanket reactor is presented. Two lab-scale UASB reactors (10.5 L were operated continuously during 12 months. The reactors were fed with synthetic wastes prepared daily using glucose, ammonium acetate, methanol and nutrient solution. One of the reactors also received increasing concentrations of sodium sulfide. For both reactors, the flow rate of 16 L.d-1 was held constant throughout the experiment, corresponding to a hydraulic retention time of 15.6 hours. The classic model for non-competitive sulfide inhibition was applied to the experimental data for determining the overall kinetic parameter of specific substrate utilization (q and the sulfide inhibition coefficient (Ki. The application of the kinetic parameters determined allows prediction of methanogenesis inhibition and thus the adoption of operating parameters to minimize sulfide toxicity in UASB reactors.

  15. Development of a Monolithic Research Reactor Fuel Type at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Clark, C.R.; Briggs, R.J.

    2004-10-06

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been tasked with the conversion of research reactors from highly enriched to low-enriched uranium (LEU). To convert several high power reactors, monolithic fuel, a new fuel type, is being developed. This fuel type replaces the standard fuel dispersion with a fuel alloy foil, which allows for fuel densities far in excess of that found in dispersion fuel. The single-piece fuel foil also contains a significantly lower interface area between the fuel and the aluminum in the plate than the standard fuel type, limiting the amount of detrimental fuel-aluminum interaction that can occur. Implementation of monolithic fuel is dependant on the development of a suitable fabrication method as traditional roll-bonding techniques are inadequate.

  16. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  17. Microstructured reactors for hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Aartun, Ingrid

    2005-07-01

    Small scale hydrogen production by partial oxidation (POX) and oxidative steam reforming (OSR) have been studied over Rh-impregnated microchannel Fecralloy reactors and alumina foams. Trying to establish whether metallic microchannel reactors have special advantages for hydrogen production via catalytic POX or OSR with respect to activity, selectivity and stability was of special interest. The microchannel Fecralloy reactors were oxidised at 1000 deg C to form a {alpha}-Al2O3 layer in the channels in order to enhance the surface area prior to impregnation. Kr-BET measurements showed that the specific surface area after oxidation was approximately 10 times higher than the calculated geometric surface area. Approximately 1 mg Rh was deposited in the channels by impregnation with an aqueous solution of RhCl3. Annular pieces (15 mm o.d.,4 mm i.d., 14 mm length) of extruded {alpha}-Al2O3 foams were impregnated with aqueous solutions of Rh(NO3)3 to obtain 0.01, 0.05 and 0.1 wt.% loadings, as predicted by solution uptake. ICP-AES analyses showed that the actual Rh loadings probably were higher, 0.025, 0.077 and 0.169 wt.% respectively. One of the microchannel Fecralloy reactors and all Al2O3 foams were equipped with a channel to allow for temperature measurement inside the catalytic system. Temperature profiles obtained along the reactor axes show that the metallic microchannel reactor is able to minimize temperature gradients as compared to the alumina foams. At sufficiently high furnace temperature, the gas phase in front of the Rh/Al2O3/Frecralloy microchannel reactor and the 0.025 wt.% Rh/Al2O3 foams ignites. Gas phase ignition leads to lower syngas selectivity and higher selectivity to total oxidation products and hydrocarbon by-products. Before ignition of the gas phase the hydrogen selectivity is increased in OSR as compared to POX, the main contribution being the water-gas shift reaction. After gas phase ignition, increased formation of hydrocarbon by

  18. Plasma spark discharge reactor and durable electrode

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young I.; Cho, Daniel J.; Fridman, Alexander; Kim, Hyoungsup

    2017-01-10

    A plasma spark discharge reactor for treating water. The plasma spark discharge reactor comprises a HV electrode with a head and ground electrode that surrounds at least a portion of the HV electrode. A passage for gas may pass through the reactor to a location proximate to the head to provide controlled formation of gas bubbles in order to facilitate the plasma spark discharge in a liquid environment.

  19. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  20. Reactor Bolshoi Moshchnosti Kalani; Reacteurs RBMK

    Energy Technology Data Exchange (ETDEWEB)

    Bastien, D. [Conservatoire National des Arts et Metiers (CNAM), 75 - Paris (France)

    2000-01-01

    The Reactor Bolshoi Molshchnosti Kalani (RBMK) are pressure tubes reactor, boiling light water cooled. Exported since 1990 from the ex-USSR, they are today in three independent countries: Russian, Ukraine and Lithuania. Since this date, data exchange with the occident allowed the better knowledge of this reactor type. The design, the technical description (core, fuel, primary system), the safety and the improvement since Chernobyl are detailed. (A.L.B.)

  1. NASA Reactor Facility Hazards Summary. Volume 1

    Science.gov (United States)

    1959-01-01

    The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

  2. Heat for industry from nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kikoin, I.K.; Novikov, V.M.

    Two factors which incline nations toward the use of heat from nuclear reactors for industrial use are: 1) exhaustion of cheap fossil fuel resources, and 2) ecological problems associated both with extraction of fossil fuel from the earth and with its combustion. In addition to the usual problems that beset nuclear reactors, special problems associated with using heat from nuclear reactors in various industries are explored.

  3. D-D tokamak reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Evans, K.E. Jr.; Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Finn, P.A.; Jung, J.; Mattas, R.F.; Misra, B.; Smith, D.L.; Stevens, H.C.

    1980-11-01

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated.

  4. Final Report BW Sample Collection& Preparation Device

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, R P; Belgrader, P; Meyer, G; Benett, W J; Richards, J B; Hadley, D R; Stratton, P L; Milanovich, F P

    2002-01-31

    The objective of this project was to develop the technique needed to prepare a field collected sample for laboratory analysis and build a portable integrated biological detection instrument with new miniaturized and automated sample purification capabilities. The device will prepare bacterial spores, bacterial vegetative cells, and viral particles for PCR amplification.

  5. BW Trained HMM based Aerial Image Segmentation

    OpenAIRE

    R Rajasree; J. Nalini; S C Ramesh

    2011-01-01

    Image segmentation is an essential preprocessing tread in a complicated and composite image dealing out algorithm. In segmenting arial image the expenditure of misclassification could depend on the factual group of pupils. In this paper, aggravated by modern advances in contraption erudition conjecture, I introduce a modus operandi to make light of the misclassification expenditure with class-dependent expenditure. The procedure assumes the hidden Markov model (HMM) which has been popularly u...

  6. Initiating Events for Multi-Reactor Plant Sites

    Energy Technology Data Exchange (ETDEWEB)

    Muhlheim, Michael David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-09-01

    Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

  7. High Performance Photocatalytic Oxidation Reactor System Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Pioneer Astronautics proposes a technology program for the development of an innovative photocatalytic oxidation reactor for the removal and mineralization of...

  8. Savannah River Site reactor safety assessment. Draft

    Energy Technology Data Exchange (ETDEWEB)

    Woody, N.D.; Brandyberry, M.D. [eds.] [Westinghouse Savannah River Co., Aiken, SC (United States); Baker, W.H.; Brandyberry, M.D.; Kearnaghan, D.P.; O`Kula, K.R.; Woody, N.D. [Westinghouse Savannah River Co., Aiken, SC (United States); Amos, C.N.; Weingardt, J.J. [Science Applications International Corp., San Diego, CA (United States)

    1991-02-28

    This report gives the results of a Savannah River Site (SRS) Production Reactor risk assessment. Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide timely information to the US Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other Site programs in Heavy Water Reactor safety.

  9. History of fast reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kittel, J.H.; Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France))

    1992-01-01

    Most of the first generation of fast reactors that were operated at significant power levels employed solid metal fuels. They were constructed in the United States and United Kingdom in the 1950s and included Experimental Breeder Reactor (EBR)-I and -II operated by Argonne National Laboratory, United States, the Enrico Fermi Reactor operated by the Atomic Power Development Associates, United States and DFR operated by the U.K. Atomic Energy Authority (UKAEA). Their paper tracer pre-development of fast reactor fuel from these early days through the 1980s including ceramic fuels.

  10. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  11. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  12. Sandia National Laboratories Medical Isotope Reactor concept.

    Energy Technology Data Exchange (ETDEWEB)

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-04-01

    This report describes the Sandia National Laboratories Medical Isotope Reactor and hot cell facility concepts. The reactor proposed is designed to be capable of producing 100% of the U.S. demand for the medical isotope {sup 99}Mo. The concept is novel in that the fuel for the reactor and the targets for the {sup 99}Mo production are the same. There is no driver core required. The fuel pins that are in the reactor core are processed on a 7 to 21 day irradiation cycle. The fuel is low enriched uranium oxide enriched to less than 20% {sup 235}U. The fuel pins are approximately 1 cm in diameter and 30 to 40 cm in height, clad with Zircaloy (zirconium alloy). Approximately 90 to 150 fuel pins are arranged in the core in a water pool {approx}30 ft deep. The reactor power level is 1 to 2 MW. The reactor concept is a simple design that is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days. The fuel fabrication, reactor design and operation, and {sup 99}Mo production processing use well-developed technologies that minimize the technological and licensing risks. There are no impediments that prevent this type of reactor, along with its collocated hot cell facility, from being designed, fabricated, and licensed today.

  13. NCSU reactor sharing program. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    Perez, P.B.

    1997-01-10

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities. This report is the Final Technical Report for the DOE award reference number DE-FG05-95NE38136 which covers the period September 30, 1995 through September 30, 1996.

  14. Molecular ecology of anaerobic reactor systems

    DEFF Research Database (Denmark)

    Hofman-Bang, H. Jacob Peider; Zheng, D.; Westermann, Peter;

    2003-01-01

    Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible...... to the abundance of each microbe in anaerobic reactor systems by rRNA probing. This chapter focuses on various molecular techniques employed and problems encountered when elucidating the microbial ecology of anaerobic reactor systems. Methods such as quantitative dot blot/fluorescence in-situ probing using various...

  15. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  16. Nanostructured Catalytic Reactors for Air Purification Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR Phase I project proposes the development of lightweight compact nanostructured catalytic reactors for air purification from toxic gaseous organic...

  17. Phosphorus removal in aerated stirred tank reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ghigliazza, R.; Lodi, A.; Rovatti, M. [Inst. of Chemical and Process Engineering ``G.B. Bonino``, Univ. of Genoa (Italy)

    1999-03-01

    The possibility to obtain biological phosphorus removal in strictly aerobic conditions has been investigated. Experiments, carried out in a continuous stirred tank reactor (CSTR), show the feasibility to obtain phosphorus removal without the anaerobic phase. Reactor performance in terms of phosphorus abatement kept always higher then 65% depending on adopted sludge retention time (SRT). In fact increasing SRT from 5 days to 8 days phosphorus removal and reactor performance increase but overcoming this SRT value a decreasing in reactor efficiency was recorded. (orig.) With 6 figs., 3 tabs., 18 refs.

  18. Sodium fast reactors with closed fuel cycle

    CERN Document Server

    Raj, Baldev; Vasudeva Rao, PR 0

    2015-01-01

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

  19. Autonomous Control of Space Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation....

  20. Neutron imaging on the VR-1 reactor

    Science.gov (United States)

    Crha, J.; Sklenka, L.; Soltes, J.

    2016-09-01

    Training reactor VR-1 is a low power research reactor with maximal thermal power of 1 kW. The reactor is operated by the Faculty of Nuclear Science and Physical Engineering of the Czech Technical University in Prague. Due to its low power it suits as a tool for education of university students and training of professionals. In 2015, as part of student research project, neutron imaging was introduced as another type of reactor utilization. The low available neutron flux and the limiting spatial and construction capabilities of the reactor's radial channel led to the development of a special filter/collimator insertion inside the channel and choosing a nonstandard approach by placing a neutron imaging plate inside the channel. The paper describes preliminary experiments carried out on the VR-1 reactor which led to first radiographic images. It seems, that due to the reactor construction and low reactor power, the neutron imaging technique on the VR-1 reactor is feasible mainly for demonstration or educational and training purposes.

  1. Microchannel Methanation Reactors Using Nanofabricated Catalysts Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Makel Engineering, Inc. (MEI) and the Pennsylvania State University (Penn State) propose to develop and demonstrate a microchannel methanation reactor based on...

  2. Nanostructured Catalytic Reactors for Air Purification Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR Phase II project proposes the development of lightweight compact nanostructured catalytic reactors for air purification from toxic gaseous organic...

  3. Continuous steroid biotransformations in microchannel reactors.

    Science.gov (United States)

    Marques, Marco P C; Fernandes, Pedro; Cabral, Joaquim M S; Znidaršič-Plazl, Polona; Plazl, Igor

    2012-01-15

    The use of microchannel reactor based technologies within the scope of bioprocesses as process intensification and production platforms is gaining momentum. Such trend can be ascribed a particular set of characteristics of microchannel reactors, namely the enhanced mass and heat transfer, combined with easier handling and smaller volumes required, as compared to traditional reactors. In the present work, a continuous production process of 4-cholesten-3-one by the enzymatic oxidation of cholesterol without the formation of any by-product was assessed. The production was carried out within Y-shaped microchannel reactors in an aqueous-organic two-phase system. Substrate was delivered from the organic phase to aqueous phase containing cholesterol oxidase and the product formed partitions back to the organic phase. The aqueous phase was then forced through a plug-flow reactor, containing immobilized catalase. This step aimed at the reduction of hydrogen peroxide formed as a by-product during cholesterol oxidation, to avoid cholesterol oxidase deactivation due to said by-product. This setup was compared with traditional reactors and modes of operation. The results showed that microchannel reactor geometry outperformed traditional stirred tank and plug-flow reactors reaching similar conversion yields at reduced residence time. Coupling the plug-flow reactor containing catalase enabled aqueous phase reuse with maintenance of 30% catalytic activity of cholesterol oxidase while eliminating hydrogen peroxide. A final production of 36 m of cholestenone was reached after 300 hours of operation.

  4. Reactors for nuclear electric propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Angelo, J.A. Jr.

    1981-01-01

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades.

  5. Biodegradation of MTBE in reactors

    DEFF Research Database (Denmark)

    Waul, Christopher Kevin

    2007-01-01

    The fuel oxygenate methyl tert-butyl ether (MTBE) was first introduced in the 1970’s to improve gasoline combustion efficiency and reduce emission of harmful gases. However, it has caused groundwater contamination in Denmark and in many locations worldwide through accidental releases from leaking...... such as ammonium or benzene, toluene, ethyl benzene and xylene (BTEX) oxidizers, which can be present together in a single system. The competition resulted in reduced and/or delayed degradation of MTBE when there were limitations of oxygen or space in the reactor. The fraction of biologically active (BA) MTBE...

  6. 一个Bw亚型家系的血型分子机制及临床输血分析%Molecular basis and clinical transfusion of a family with Bw subtype of ABO blood group system

    Institute of Scientific and Technical Information of China (English)

    邓刚; 黄丹丹; 郭雯玉; 许德义; 杜勇; 马幼丽; 张哲

    2013-01-01

    Objective To study a family with Bw subtype of ABO blood group system,and to review safety issues in relation with clinical transfusion.Methods The molecular basis for the blood type was studied with serological assay,polymerase chain reaction-sequence specific primer (PCR-SSP) and DNA sequencing,TA clone and haplotype analysis in one blood donor whose ABO blood group were difficultly typed and her family.The bioinformatics analysis was carried out by biological analysis software to investigate the change of structure and function of enzymes influenced by the change amino acid.A retrospective survey was carried out to investigate what is the actual position that the donor blood was used in the clinical transfusion.Results Three members from the family were found to have a Bw subtype.A substitution of nucleotide C by T at position 721 in exon 7 was discovered,which resulted in replacement of amino acid Arg to Trp.Review of clinical record suggested that there has been no significant abnormality association with past three blood transfusions.Conclusion A 721C>T mutation of the ABO gene probably underlies the Bw subtype.Further research is needed for understanding the clinical significance of this subtype in the blood transfusion.%目的 对1个Bw亚型血型家系的分子机制进行研究,并探讨该亚型的临床输血情况.方法 对1名ABO血型正反定型不符的无偿献血者及其家人的血型进行血清学鉴定,并应用聚合酶链反应-序列特异性引物法、ABO基因直接测序、TA克隆单倍型分析及相关软件对该突变引起的酶结构及功能变化进行分析等方法,同时对该献血者以往3次所献的血样的临床输血情况进行回顾.结果 在该家系中发现了3例较为罕见的Bw亚型.该亚型是由于ABO基因第7外显子存在721C/T杂合,导致R241W氨基酸改变所致.临床回顾提示3次输血均未发现明显异常.结论 ABO基因721C>T突变是导致Bw亚型的分子遗传基础之一,

  7. Ge-on-Si: Single-Crystal Selective Epitaxial Growth in a CVD Reactor

    NARCIS (Netherlands)

    Sammak, A.; De Boer, W.B.; Nanver, L.K.

    2012-01-01

    A standard Si/SiGe ASM CVD reactor that was recently modified for merging GaAs and Si epitaxial growth in one system is utilized to achieve intrinsic and doped epitaxial Ge-on-Si with low threading dislocation and defect densities. For this purpose, the system is equipped with 2% diluted GeH4 as the

  8. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion.

    Science.gov (United States)

    Marques, J G; Sousa, M; Santos, J P; Fernandes, A C

    2011-08-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  9. Decision-support tool for assessing future nuclear reactor generation portfolios

    NARCIS (Netherlands)

    Jain, S.; Roelofs, F; Oosterlee, C.W.

    2014-01-01

    Capital costs, fuel, operation and maintenance (O&M) costs, and electricity prices play a key role in the economics of nuclear power plants. Often standardized reactor designs are required to be locally adapted, which often impacts the project plans and the supply chain. It then becomes difficult to

  10. Reactor scram experience for shutdown system reliability analysis. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Edison, G.E.; Pugliese, S.L.; Sacramo, R.F.

    1976-06-01

    Scram experience in a number of operating light water reactors has been reviewed. The date and reactor power of each scram was compiled from monthly operating reports and personal communications with the operating plant personnel. The average scram frequency from ''significant'' power (defined as P/sub trip//P/sub max greater than/ approximately 20 percent) was determined as a function of operating life. This relationship was then used to estimate the total number of reactor trips from above approximately 20 percent of full power expected to occur during the life of a nuclear power plant. The shape of the scram frequency vs. operating life curve resembles a typical reliability bathtub curve (failure rate vs. time), but without a rising ''wearout'' phase due to the lack of operating data near the end of plant design life. For this case the failures are represented by ''bugs'' in the plant system design, construction, and operation which lead to scram. The number of scrams would appear to level out at an average of around three per year; the standard deviations from the mean value indicate an uncertainty of about 50 percent. The total number of scrams from significant power that could be expected in a plant designed for a 40-year life would be about 130 if no wearout phase develops near the end of life.

  11. 3-flavor oscillations with current and future reactor experiments

    Science.gov (United States)

    Dwyer, Dan

    2017-01-01

    Nuclear reactors have been a crucial tool for our understanding of neutrinos. The disappearance of electron antineutrinos emitted by nuclear reactors has firmly established that neutrino flavor oscillates, and that neutrinos consequently have mass. The current generation of precision measurements rely on some of the world's most intense reactor facilities to demonstrate that the electron antineutrino mixes with the third antineutrino mass eigenstate (v3-). Accurate measurements of antineutrino energies robustly determine the tiny difference between the masses-squared of the v3- state and the two more closely-spaced v1- and v2- states. These results have given us a much clearer picture of neutrino mass and mixing, yet at the same time open major questions about how to account for these small but non-zero masses in or beyond the Standard Model. These observations have also opened the door for a new generation of experiments which aim to measure the ordering of neutrino masses and search for potential violation of CP symmetry by neutrinos. I will provide a brief overview of this exciting field. Work supported under DOE OHEP DE-AC02-05CH11231.

  12. An innovative reactor-type biosensor for BOD rapid measurement.

    Science.gov (United States)

    Wang, Jianlong; Zhang, Yixin; Wang, Yeyao; Xu, Runhua; Sun, Zhonghua; Jie, Zhou

    2010-03-15

    Biochemical oxygen demand (BOD) is one of the most important and widely used parameters for characterizing the organic pollution of water and wastewater. In this paper, a novel reactor-type biosensor for rapid measurement of BOD was developed, based on using immobilized microbial cell (IMC) beads as recognition bio-element in a completely mixed reactor which was used as determining chamber, replacing the traditionally used membrane as recognition bio-element. The IMC beads were freely suspended in the aqueous solution, so the mass transfer resistance for dissolved oxygen and organic compounds significantly reduced, and the quantity of the microbial cells used as recognition element can be easily adjusted, in comparison with the traditional membrane-type BOD biosensor, in which exists a unadjustable contradiction between the quantity of biomass and the thickness of the bio-membrane, thus limiting the stability and the detection limit. This novel kind of BOD biosensor significantly increased the sensitivity of the response, the detecting precision and prolonged the life time of the recognition element. The experimental data showed that the most appropriate temperature for biochemical reaction in the reactor was 30 degrees C, and the IMC beads could keep the bioactivity for about 70d at the detecting frequency of 8 times every day. The standard deviation of repeatability and the reproducibility of responses were within +/-6.4% and +/-5.0%, respectively, which are within acceptable bias limits, and meet the requirement of BOD rapid measurement.

  13. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  14. BNCT Technology Development on HANARO Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ki Jung; Park, Kyung Bae; Whang, Seung Ryul; Kim, Myong Seop

    2007-06-15

    So as to establish the biological effects of BNCT in the HANARO Reactor, biological damages in cells and animals with treatment of boron/neutron were investigated. And 124I-BPA animal PET image, analysis technology of the boron contents in the mouse tissues by ICP-AES was established. A Standard clinical protocol, a toxicity evaluation report and an efficacy investigation report of BNCT has been developed. Based on these data, the primary permission of clinical application was acquired through IRB of our hospital. Three cases of pre-clinical experiment for boron distribution and two cases of medium-sized animal simulation experiment using cat with verifying for 2 months after BNCT was performed and so the clinical demonstration with a patient was prepared. Also neutron flux, fast neutron flux and gamma ray dose of BNCT facility were calculated and these data will be utilized good informations for clinical trials and further BNCT research. For the new synthesis of a boron compound, o-carboranyl ethylamine, o-carboranylenepiperidine, o-carboranyl-THIQ and o-carboranyl-s-triazine derivatives were synthesized. Among them, boron uptake in the cancer cell of the triazine derivative was about 25 times than that of BPA and so these three synthesized methods of new boron compounds were patented.

  15. Constraining Sterile Neutrinos Using Reactor Neutrino Experiments

    CERN Document Server

    Girardi, Ivan; Ohlsson, Tommy; Zhang, He; Zhou, Shun

    2014-01-01

    Models of neutrino mixing involving one or more sterile neutrinos have resurrected their importance in the light of recent cosmological data. In this case, reactor antineutrino experiments offer an ideal place to look for signatures of sterile neutrinos due to their impact on neutrino flavor transitions. In this work, we show that the high-precision data of the Daya Bay experi\\-ment constrain the 3+1 neutrino scenario imposing upper bounds on the relevant active-sterile mixing angle $\\sin^2 2 \\theta_{14} \\lesssim 0.06$ at 3$\\sigma$ confidence level for the mass-squared difference $\\Delta m^2_{41}$ in the range $(10^{-3},10^{-1}) \\, {\\rm eV^2}$. The latter bound can be improved by six years of running of the JUNO experiment, $\\sin^22\\theta_{14} \\lesssim 0.016$, although in the smaller mass range $ \\Delta m^2_{41} \\in (10^{-4} ,10^{-3}) \\, {\\rm eV}^2$. We have also investigated the impact of sterile neutrinos on precision measurements of the standard neutrino oscillation parameters $\\theta_{13}$ and $\\Delta m^2...

  16. Immobilization of Fast Reactor First Cycle Raffinate

    Energy Technology Data Exchange (ETDEWEB)

    Langley, K. F.; Partridge, B. A.; Wise, M.

    2003-02-26

    This paper describes the results of work to bring forward the timing for the immobilization of first cycle raffinate from reprocessing fuel from the Dounreay Prototype Fast Reactor (PFR). First cycle raffinate is the liquor which contains > 99% of the fission products separated from spent fuel during reprocessing. Approximately 203 m3 of raffinate from the reprocessing of PFR fuel is held in four tanks at the UKAEA's site at Dounreay, Scotland. Two methods of immobilization of this high level waste (HLW) have been considered: vitrification and cementation. Vitrification is the standard industry practice for the immobilization of first cycle raffinate, and many papers have been presented on this technique elsewhere. However, cementation is potentially feasible for immobilizing first cycle raffinate because the heat output is an order of magnitude lower than typical HLW from commercial reprocessing operations such as that at the Sellafield site in Cumbria, England. In fact, it falls within the upper end of the UK definition of intermediate level waste (ILW). Although the decision on which immobilization technique will be employed has yet to be made, initial development work has been undertaken to identify a suitable cementation formulation using inactive simulant of the raffinate. An approach has been made to the waste disposal company Nirex to consider the disposability of the cemented product material. The paper concentrates on the process development work that is being undertaken on cementation to inform the decision making process for selection of the immobilization method.

  17. Polymerization in emulsion microdroplet reactors

    Science.gov (United States)

    Carroll, Nick J.

    The goal of this research project is to utilize emulsion droplets as chemical reactors for execution of complex polymerization chemistries to develop unique and functional particle materials. Emulsions are dispersions of immiscible fluids where one fluid usually exists in the form of drops. Not surprisingly, if a liquid-to-solid chemical reaction proceeds to completion within these drops, the resultant solid particles will possess the shape and relative size distribution of the drops. The two immiscible liquid phases required for emulsion polymerization provide unique and complex chemical and physical environments suitable for the engineering of novel materials. The development of novel non-ionic fluorosurfactants allows fluorocarbon oils to be used as the continuous phase in a water-free emulsion. Such emulsions enable the encapsulation of almost any hydrocarbon compound in droplets that may be used as separate compartments for water-sensitive syntheses. Here, we exemplify the promise of this approach by suspension polymerization of polyurethanes (PU), in which the liquid precursor is emulsified into droplets that are then converted 1:1 into polymer particles. The stability of the droplets against coalescence upon removal of the continuous phase by evaporation confirms the formation of solid PU particles. These results prove that the water-free environment of fluorocarbon based emulsions enables high conversion. We produce monodisperse, cross-linked, and fluorescently labeled PU-latexes with controllable mesh size through microfluidic emulsification in a simple one-step process. A novel method for the fabrication of monodisperse mesoporous silica particles is presented. It is based on the formation of well-defined equally sized emulsion droplets using a microfluidic approach. The droplets contain the silica precursor/surfactant solution and are suspended in hexadecane as the continuous oil phase. The solvent is then expelled from the droplets, leading to

  18. Dynamic neutronic and stability analysis of a burst mode, single cavity gas core reactor Brayton cycle space power system

    Science.gov (United States)

    Dugan, Edward T.; Kutikkad, Kiratadas

    The conceptual, burst-mode gaseous-core reactor (GCR) space nuclear power system presently subjected to reactor-dynamics and system stability studies operates on a closed Brayton cycle, via disk MHD generator for energy conversion. While the gaseous fuel density power coefficient of reactivity is found to be capable of rapidly stabilizing the GCR system, the power of this feedback renders standard external reactivity insertions inadequate for significant power-level changes during normal operation.

  19. Development of Guide System for a Reactor Head Maintenance Robot

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ho Cheol; Seo, Yong Chil; Jung, Kyung Min; Lee, Sung Uk; Kim, Seung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Park, Kwang Su [Doosan Heavy Industries and Construction Co., Ltd., Changwon (Korea, Republic of)

    2005-07-01

    The Control Rod Drive(CRD) nozzles for PWR nuclear power plants(NPP) house the control rod drives. The number of nozzle penetrations range from the mid-30's to over 100 in each reactor head. The integrity of CRD nozzles is very important, because the primary pressure boundary is established with the J-groove weld joining the nozzle to the head clad surface. The Alloy 600 PWSC CRD nozzle leaks discovered in the fall of 2000 and spring of 2001 in several US plants. Therefore the NRC has recommended a more proactive effort by US utilities to inspect similarly susceptible nozzles in all US plants. The primary safety concern is circumferential cracks that can permit the nozzles to separate from the head at high velocity and produce a large-break leak in the reactor vessel. A secondary concern is head leakage from any through-wall cracks in the nozzle or J-groove weld area. Numerous inspection and repair tools have been developed to address CRD nozzle inspection and repair issues. For example, Framatome-ANP has been developed several inspection and repair tools: bare-head visual inspection crawler, blade eddy current probes and rotating eddy current proves, ultrasonic volumetric test(UT) blade proves, rotating UT prove, remote dye-penetrant test(PT) tool and remote weld tool. And they developed tool delivering systems such as ARAMIS, ROCKY and SUMO ROCKY. KPS and Westing House also developed inspection tool and delivering system. In this paper, a guide system delivering a welding repair tool and robot was developed. The welding repair tool and robot is being developed by Doosan heavy industry. The guide system was designed to apply for the reactor head of Korean standard type NPP. The reactor head is placed on the laydown support during overhaul period. The maintenance of reactor head is carried out in the laydown support. First, work conditions of the job site were investigated to consider the entering and leaving convenience of the reactor head repair robot. The

  20. Report of the APS Neutrino Study Reactor Working Group

    Energy Technology Data Exchange (ETDEWEB)

    Abouzaid, E.; Anderson, K.; Barenboim, G.; Berger, B.; Blucher, E.; Bolton, T.; Choubey, S.; Conrad, J.; Formaggio, J.; Freedman, S.; Finely, D.; Fisher, P.; Fujikawa, B.; Gai, M.; Goodman, M.; de Goueva, A.; Hadley, N.; Hahn, R.; Horton-Smith, G.; Kadel, R.; Kayser, B.; Heeger, K.; Klein, J.; Learned, J.; Lindner, M.; Link, J.; Luk, K.-B.; McKeown, R.; Mocioiu, I.; Mohapatra, R.; Naples, D.; Peng, J.; Petcov, S.; Pilcher, J.; Rapidis, P.; Reyna, D.; Shaevitz, M.; Shrock, R.; Stanton, N.; Stefanski, R.; Yamamoto, R.; Worcester, M.

    2004-10-28

    The worldwide program to understand neutrino oscillations and determine the neutrino mixing parameters, CP violating effects, and mass hierarchy will require a broad combination of measurements. The group believes that a key element of this future neutrino program is a multi-detector neutrino experiment (with baselines of {approx} 200 m and {approx} 1.5 km) with a sensitivity of sin{sup 2} 2{theta}{sub 13} = 0.01. In addition to oscillation physics, the reactor experiment may provide interesting measurements of sin{sup 2} {theta}{sub W} at Q{sup 2} = 0, neutrino couplings, magnetic moments, and mixing with sterile neutrino states. {theta}{sub 13} is one of the twenty-six parameters of the standard model, the best model of electroweak interactions for energies below 100 GeV and, as such, is worthy of a precision measurement independent of other considerations. A reactor experiment of the proposed sensitivity will allow a measurement of {theta}{sub 13} with no ambiguities and significantly better precision than any other proposed experiment, or will set limits indicating the scale of future experiments required to make progress. Figure 1 shows a comparison of the sensitivity of reactor experiments of different scales with accelerator experiments for setting limits on sin{sup 2} 2{theta}{sub 13} if the mixing angle is very small, or for making a measurement of sin{sup 2} 2{theta}{sub 13} if the angle is observable. A reactor experiment with a 1% precision may also resolve the degeneracy in the {theta}{sub 23} parameter when combined with long-baseline accelerator experiments. In combination with long-baseline measurements, a reactor experiment may give early indications of CP violation and the mass hierarchy. The combination of the T2K and Nova long-baseline experiments will be able to make significant measurements of these effects if sin{sup 2} 2{theta}{sub 13} > 0.05 and with enhanced beam rates can improve their reach to the sin{sup 2} 2{theta}{sub 13} > 0.02 level

  1. Removal of anaerobic soluble microbial products in a biological activated carbon reactor.

    Science.gov (United States)

    Dong, Xiaojing; Zhou, Weili; He, Shengbing

    2013-09-01

    The soluble microbial products (SMP) in the biological treatment effluent are generally of great amount and are poorly biodegradable. Focusing on the biodegradation of anaerobic SMP, the biological activated carbon (BAC) was introduced into the anaerobic system. The experiments were conducted in two identical lab-scale up-flow anaerobic sludge blanket (UASB) reactors. The high strength organics were degraded in the first UASB reactor (UASB1) and the second UASB (UASB2, i.e., BAC) functioned as a polishing step to remove SMP produced in UASB1. The results showed that 90% of the SMP could be removed before granular activated carbon was saturated. After the saturation, the SMP removal decreased to 60% on the average. Analysis of granular activated carbon adsorption revealed that the main role of SMP removal in BAC reactor was biodegradation. A strain of SMP-degrading bacteria, which was found highly similar to Klebsiella sp., was isolated, enriched and inoculated back to the BAC reactor. When the influent chemical oxygen demand (COD) was 10,000 mg/L and the organic loading rate achieved 10 kg COD/(m3 x day), the effluent from the BAC reactor could meet the discharge standard without further treatment. Anaerobic BAC reactor inoculated with the isolated Klebsiella was proved to be an effective, cheap and easy technical treatment approach for the removal of SMP in the treatment of easily-degradable wastewater with COD lower than 10,000 mg/L.

  2. Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

    Science.gov (United States)

    Wright, Steven A.; Houts, Michael

    2001-02-01

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

  3. Journal standards.

    Science.gov (United States)

    Jackson, R

    2003-08-01

    Despite its many imperfections, the peer review process is a firmly established quality control system for scientific literature. It gives readers some assurance that the work and views that are reported meet standards that are acceptable to a journal. Maureen Revington's editorial in a recent issue of the Australian Veterinary Journal (Revington2002) gives a good concise warts and all overview of the process and is well worth reading. I have some concerns about several articles in the December 2002 issue of the New Zealand Veterinary Journal (Volume 50, Number 6), devoted to the health and welfare of farmed deer, that relate to extensive citing of non-peer reviewed papers. I can understand the need for information to flow from researchers to the wider community but that need is already satisfied by publications such as the proceedings of the Deer Branch of the New Zealand Veterinary Association and Proceedings of the New Zealand Society of Animal Production. Non-peer reviewed papers have been cited in the Journal in the past but never to the extent displayed in this particular issue. It degrades the peer-review process and creates an added burden for reviewers who are forced to grapple with the uncertainties of the science in non-peer reviewed citations. One of my fears is that this process allows science from non peer reviewed articles to be legitimised by its inclusion in a peer reviewed journal and perhaps go on to be accepted as dogma. This is a real danger given the difficulties associated with tracing back to original citations and the increasing volume of scientific literature. It also affords opportunities for agencies to pick up questionable and doubtful science and tout it as support for their products or particular points of view. If deer researchers choose to publish most of their work in proceedings then so be it. However this approach, which seems to becoming increasingly prevalent in the deer sector, is questionable from an established science point

  4. Compound cryopump for fusion reactors

    CERN Document Server

    Kovari, M; Shephard, T

    2013-01-01

    We reconsider an old idea: a three-stage compound cryopump for use in fusion reactors such as DEMO. The helium "ash" is adsorbed on a 4.5 K charcoal-coated surface, while deuterium and tritium are adsorbed at 15-22 K on a second charcoal-coated surface. The helium is released by raising the first surface to ~30 K. In a separate regeneration step, deuterium and tritium are released at ~110 K. In this way, the helium can be pre-separated from other species. In the simplest design, all three stages are in the same vessel, with a single valve to close the pump off from the tokamak during regeneration. In an alternative design, the three stages are in separate vessels, connected by valves, allowing the stages to regenerate without interfering with each other. The inclusion of the intermediate stage would not affect the overall pumping speed significantly. The downstream exhaust processing system could be scaled down, as much of the deuterium and tritium could be returned directly to the reactor. This could reduce ...

  5. K-East and K-West Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — Hanford's "sister reactors", the K-East and the K-West Reactors, were built side-by-side in the early 1950's. The two reactors went operational within four months of...

  6. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  7. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  8. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  9. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis

    Science.gov (United States)

    Turek, Thomas

    2009-01-01

    Summary Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid. PMID:20300506

  10. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis

    Directory of Open Access Journals (Sweden)

    Ulrich Kunz

    2009-11-01

    Full Text Available Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid.

  11. Flow through reactors for organic chemistry: directly electrically heated tubular mini reactors as an enabling technology for organic synthesis.

    Science.gov (United States)

    Kunz, Ulrich; Turek, Thomas

    2009-11-30

    Until recently traditional heating in organic chemistry has been done with oil heating baths or using electric heat exchangers. With the advent of microwave equipment, heating by microwaves was rapidly introduced as standard method in organic chemistry laboratories, mainly because of the convenient possibility to operate at high temperature accompanied by accelerated reaction rates. In the present contribution we discuss the method of heating small, continuously operated reactors by passing electric current directly through the reactor wall as an enabling technology in organic chemistry. The benefit of this method is that the heat is generated directly inside the reactor wall. By this means high heating rates comparable to microwave ovens can be reached but at much lower cost for the equipment. A tool for the comparison of microwave heating and traditional heating is provided. As an example kinetic data for the acid catalyzed hydrolysis of methyl formate were measured using this heating concept. The reaction is not only a suitable model but also one of industrial importance since this is the main production process for formic acid.

  12. Human reliability analysis of the Tehran research reactor using the SPAR-H method

    Directory of Open Access Journals (Sweden)

    Barati Ramin

    2012-01-01

    Full Text Available The purpose of this paper is to cover human reliability analysis of the Tehran research reactor using an appropriate method for the representation of human failure probabilities. In the present work, the technique for human error rate prediction and standardized plant analysis risk-human reliability methods have been utilized to quantify different categories of human errors, applied extensively to nuclear power plants. Human reliability analysis is, indeed, an integral and significant part of probabilistic safety analysis studies, without it probabilistic safety analysis would not be a systematic and complete representation of actual plant risks. In addition, possible human errors in research reactors constitute a significant part of the associated risk of such installations and including them in a probabilistic safety analysis for such facilities is a complicated issue. Standardized plant analysis risk-human can be used to address these concerns; it is a well-documented and systematic human reliability analysis system with tables for human performance choices prepared in consultation with experts in the domain. In this method, performance shaping factors are selected via tables, human action dependencies are accounted for, and the method is well designed for the intended use. In this study, in consultations with reactor operators, human errors are identified and adequate performance shaping factors are assigned to produce proper human failure probabilities. Our importance analysis has revealed that human action contained in the possibility of an external object falling on the reactor core are the most significant human errors concerning the Tehran research reactor to be considered in reactor emergency operating procedures and operator training programs aimed at improving reactor safety.

  13. Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.

    1972-07-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  14. Automated scoping methodology for liquid metal natural circulation small reactor

    Energy Technology Data Exchange (ETDEWEB)

    Son, Hyung M.; Suh, Kune Y., E-mail: kysuh@snu.ac.kr

    2014-07-01

    Highlights: • Automated scoping methodology for natural circulation small modular reactor is developed. • In-house code is developed to carry out system analysis and core geometry generation during scoping. • Adjustment relations are obtained to correct the critical core geometry out of diffusion theory. • Optimized design specification is found using objective function value. • Convex hull volume is utilized to quantify the impact of different constraints on the scope range. - Abstract: A novel scoping method is proposed that can automatically generate design variable range of the natural circulation driven liquid metal cooled small reactor. From performance requirements based upon Generation IV system roadmap, appropriate structure materials are selected and engineering constraints are compiled based upon literature. Utilizing ASME codes and standards, appropriate geometric sizing criteria on constituting components are developed to ensure integrity of the system during its lifetime. In-house one dimensional thermo-hydraulic system analysis code is developed based upon momentum integral model and finite element methods to deal with non-uniform descritization of temperature nodes for convection and thermal diffusion equation of liquid metal coolant. In order to quickly generate critical core dimensions out of given unit cell information, an adjustment relation that relates the critical geometry estimated from one-group diffusion and that from MCNP code is constructed and utilized throughout the process. For the selected unit cell dimension ranges, burnup calculations are carried out to check the cores can generate energy over the reactor lifetime. Utilizing random method, sizing criteria, and in-house analysis codes, an automated scoping methodology is developed. The methodology is applied to nitride fueled integral type lead cooled natural circulation reactor concept to generate design scopes which satisfies given constraints. Three dimensional convex

  15. Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

    Energy Technology Data Exchange (ETDEWEB)

    Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.

  16. Advances toward a transportable antineutrino detector system for reactor monitoring and safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Reyna, D. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bernstein, A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Lund, J.; Kiff, S.; Cabrera-Palmer, B. [Sandia National Laboratories, Livermore, CA 94550 (United States); Bowden, N. S.; Dazeley, S.; Keefer, G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

    2011-07-01

    Nuclear reactors have served as the neutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Our SNL/LLNL collaboration has demonstrated that such antineutrino based monitoring is feasible using a relatively small cubic meter scale liquid scintillator detector at tens of meters standoff from a commercial Pressurized Water Reactor (PWR). With little or no burden on the plant operator we have been able to remotely and automatically monitor the reactor operational status (on/off), power level, and fuel burnup. The initial detector was deployed in an underground gallery that lies directly under the containment dome of an operating PWR. The gallery is 25 meters from the reactor core center, is rarely accessed by plant personnel, and provides a muon-screening effect of some 20-30 meters of water equivalent earth and concrete overburden. Unfortunately, many reactor facilities do not contain an equivalent underground location. We have therefore attempted to construct a complete detector system which would be capable of operating in an aboveground location and could be transported to a reactor facility with relative ease. A standard 6-meter shipping container was used as our transportable laboratory - containing active and passive shielding components, the antineutrino detector and all electronics, as well as climate control systems. This aboveground system was deployed and tested at the San Onofre Nuclear Generating Station (SONGS) in southern California in 2010 and early 2011. We will first present an overview of the initial demonstrations of our below ground detector. Then we will describe the aboveground system and the technological developments of the two antineutrino

  17. Selective purge for hydrogenation reactor recycle loop

    Science.gov (United States)

    Baker, Richard W.; Lokhandwala, Kaaeid A.

    2001-01-01

    Processes and apparatus for providing improved contaminant removal and hydrogen recovery in hydrogenation reactors, particularly in refineries and petrochemical plants. The improved contaminant removal is achieved by selective purging, by passing gases in the hydrogenation reactor recycle loop or purge stream across membranes selective in favor of the contaminant over hydrogen.

  18. Radiochemical problems of fusion reactors. 1. Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Crespi, M.B.A.

    1984-02-01

    A list of fusion reactor candidate materials is given, for use in connection with blanket structure, breeding, moderation, neutron multiplication, cooling, magnetic field generation, electrical insulation and radiation shielding. The phenomena being studied for each group of materials are indicated. Suitable irradiation test facilities are discussed under the headings (1) accelerator-based neutron sources, (2) fission reactors, and (3) ion accelerators.

  19. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  20. Startup of an industrial adiabatic tubular reactor

    NARCIS (Netherlands)

    Verwijs, J.W.; Berg, van den H.; Westerterp, K.R.

    1992-01-01

    The dynamic behaviour of an adiabatic tubular plant reactor during the startup is demonstrated, together with the impact of a feed-pump failure of one of the reactants. A dynamic model of the reactor system is presented, and the system response is calculated as a function of experimentally-determine