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Sample records for bw standard reactor

  1. 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... Standards for Power Reactors.'' DATES: Submit comments by February 7, 2014. Comments received after this... of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory...

  2. Reactor Section standard analytical methods. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Sowden, D.

    1954-07-01

    the Standard Analytical Methods manual was prepared for the purpose of consolidating and standardizing all current analytical methods and procedures used in the Reactor Section for routine chemical analyses. All procedures are established in accordance with accepted practice and the general analytical methods specified by the Engineering Department. These procedures are specifically adapted to the requirements of the water treatment process and related operations. The methods included in this manual are organized alphabetically within the following five sections which correspond to the various phases of the analytical control program in which these analyses are to be used: water analyses, essential material analyses, cotton plug analyses boiler water analyses, and miscellaneous control analyses.

  3. Reference design for the standard mirror hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-05-22

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (/sup 239/Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket.

  4. Fusion reactor design studies: standard accounts for cost estimates

    Energy Technology Data Exchange (ETDEWEB)

    Schulte, S.C.; Willke, T.L.; Young, J.R.

    1978-05-01

    The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed.

  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) standard...

  6. 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.

  7. 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...

  8. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  9. 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 ...

  10. 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.

  11. Standardizing electrocoagulation reactor design: iron electrodes for NOM removal.

    Science.gov (United States)

    Dubrawski, Kristian L; Mohseni, Madjid

    2013-03-01

    A novel systematic approach for reactor design was described for iron electrocoagulation (EC) and applied to drinking water treatment. Suwannee NOM was used as a model compound; performance was quantified by UV-abs-254 and DOC removal. Significant EC design variables were identified and examined: current density (i) (2.43-26.8 mA cm(-2)), coagulant or charge loading rate (CLR) (100-1000 CL(-1) min(-1)), and flocculation methodology ("fast" and "slow"). A correlation was found between increased i and decreased current efficiency (φ), optimum NOM removal was found at i ~10 mA cm(-2). A lower CLR showed greater total DOC removal, while a higher CLR led to less reactor residence time and required either longer flocculation times or greater coagulant dose for similar NOM removal. This paper defines and describes the four general EC "classes" of operation that have implications on several important measures of success: coagulant dose, electrical consumption, process speed, volumetric footprint, and post-EC flocculation requirements. Two classes were further examined with or without pH adjustment for DOC removal, showing that a "fast" EC mode without flocculation is more appropriate for smaller applications, while a "slow" EC mode is more effective for large permanent applications, where flocculation and settling can reduce coagulant and electrical consumption. The effect of pH adjustment showed greater impact with the "fast" dosing mode than with the "slow" mode, adjustment to pH 6 with the "fast" mode gave 13.8% and 29.1% greater DOC and UV-abs-254 removal, respectively, compared to the baseline without pH adjustment. Copyright © 2012 Elsevier Ltd. All rights reserved.

  12. 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.

  13. 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

  14. Standard technical specifications for combustion engineering pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    1979-08-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Combustion Engineering plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident.

  15. Standard technical specifications for Westinghouse pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, P.C.

    1979-07-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Westinghouse plants currently being reviewed for an Operating License. Accordingly, the document contains specifications applicable to plants with (1) either 3 or 4 loops and (2) with and without loop stop valves. In addition, four separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, Ice Condenser, Sub-Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of the STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident.

  16. 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

    ... COMMISSION Office of New Reactors; Proposed Revision to Standard Review Plan, Section 14.3.12 on Physical... NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,'' on a proposed Revision 1 to Standard Review Plan (SRP), Section 14.3.12 on ``Physical Security...

  17. 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 preferred...9|ref|XP_967989.1| PREDICTED: similar to lipoma preferred partner/lpp [Tribolium castaneum] BW999205 L8 ...

  18. Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Results, E706(IA)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2001-01-01

    1.1 This practice covers the methodology, summarized in Annex A1, to be used in the analysis and interpretation of neutron exposure data obtained from LWR pressure vessel surveillance programs; and, based on the results of that analysis, establishes a formalism to be used to evaluate present and future condition of the pressure vessel and its support structures (1-70). 1.2 This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods (see Master Matrix E 706) (1, 5, 13, 48, 49). In order to make this practice at least partially self-contained, a moderate amount of discussion is provided in areas relating to ASTM and other documents. Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, and exposure units. Note 1—(Figure 1 is deleted in the latest update. The user is refered to Master Matrix E 706 for the latest figure of the standards interconnectivity). 1.3 This practice is restri...

  19. 76 FR 17159 - Office of New Reactors; Final Interim Staff Guidance on Standard Review Plan, Section 17.4...

    Science.gov (United States)

    2011-03-28

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Office of New Reactors; Final Interim Staff Guidance on Standard Review Plan, Section 17.4.... SUMMARY: The NRC staff is issuing its Final Interim Staff Guidance (ISG) DC/COL-ISG-018 (Agencywide...

  20. 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

  1. EST Table: BW999103 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999103 SES2281 10/09/28 50 %/165 aa ref|XP_001603591.1| PREDICTED: similar to Sorting and assembly machine...9238341|ref|XP_966817.2| PREDICTED: similar to sorting and assembly machinery component 50 homolog [Tribolium castaneum] FS904485 L8 ...

  2. 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...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 ...

  3. 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 wheel...-PA 10/09/10 58 %/154 aa gi|91076478|ref|XP_972409.1| PREDICTED: similar to 18 wheeler [Tribolium castaneum] FS922922 L8 ...

  4. EST Table: BW997911 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997911 ES2654 11/12/09 n.h 10/09/28 80 %/111 aa ref|XP_001850447.1| rabconnectin ...[Culex quinquefasciatus] gb|EDS32040.1| rabconnectin [Culex quinquefasciatus] 10/08/29 77 %/111 aa FBpp02746

  5. 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

  6. 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 ...

  7. EST Table: BW997372 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997372 ES0567 10/09/28 67 %/143 aa ref|XP_395977.1| PREDICTED: similar to mRNA dec...8.1| PREDICTED: similar to Decapping protein 2 CG6169-PA [Tribolium castaneum] FS769359 L8 ...

  8. 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.

  9. The cloud storage service bwSync&Share at KIT

    CERN Multimedia

    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.

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

    Science.gov (United States)

    2015-12-01

    Selected Acquisition Report (SAR) RCS: DD-A&T(Q&A)823-101 H-1 Upgrades (4BW/4BN) (H-1 Upgrades) As of FY 2017 President’s Budget Defense...5 Mission and Description 6 Executive Summary 7 Threshold Breaches 8 Schedule 9 Performance 11 Track to Budget 16 Cost and...Funding 17 Low Rate Initial Production 26 Foreign Military Sales 27 Nuclear Costs 27 Unit Cost 28 Cost Variance 31 Contracts

  11. EST Table: BW997512 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997512 ES1005 10/09/28 84 %/121 aa ref|XP_966342.1| PREDICTED: similar to bicaudal... CG3644-PA isoform 1 [Tribolium castaneum] ref|XP_975853.1| PREDICTED: similar to bicaudal CG3644-PA isofor.../10 84 %/121 aa gi|91089799|ref|XP_966342.1| PREDICTED: similar to bicaudal CG3644-PA isoform 1 [Tribolium castaneum] FS917098 L8 ...

  12. 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 ...

  13. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  14. 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...

  15. Application of coupled code technique to a safety analysis of a standard MTR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hamidouche, Tewfik [Division de l' Environnement, de la Surete et des Dechets Radioactifs, Centre de Recherche Nucleaire d' Alger (CRNA), Alger (Algeria); Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: t.hamidouche@crna.dz; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)], E-mail: b.salah@ing.unipi.it; Si-Ahmed, El Khider [Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: esi-ahmed@usthb.dz; Mokeddem, Mohamed Yazid [Division de la Physique et des Applications Nucleaires, Centre de Recherche Nucleaire de Draria (CRND) (Algeria); D' Auria, Franscesco [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)

    2009-10-15

    Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal-hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal-Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics-thermal-hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  16. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

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

  17. Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Lin; Gigax, Jonathan; Chen, Di; Kim, Hyosim; Garner, Frank A.; Wang, Jing; Toloczko, Mychailo B.

    2017-10-01

    Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and in some cases introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.

  18. Design of a weapons-grade plutonium assembly for optimal burnup in a standard pressurized water reactor

    Science.gov (United States)

    Alonso-Vargas, Gustavo

    We created a new MOX fuel assembly design that can be used in standard Westinghouse pressurized water reactors (PWR) to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which is called MIX-33, appears to be a good option for the disposition of weapons-grade plutonium (WG-Pu), increasing the plutonium disposition rate by 8% compared to a previous Westinghouse design. It is based in two novel ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the increase of the moderation ratio of the assembly. We replaced 8 fuel pins by water holes to increase the moderation ratio. We can transition smoothly from a full LEU core to a full MIX-33 core meeting the operational and safety regulations of a standard PWR. Given a MOX supply interruption scenario we can transition smoothly to full LEU meeting the safety regulations and using standard LEU assemblies with uniform enriched pin-wise distribution. If the MOX supply is interrupted for only one cycle, we are able to transition back to full MIX-33 core. However, in this case we probably need to de-rate the power by a few percent for a few weeks at the beginning of the cycle (BOC) to accommodate high peaking. For comparison we created another assembly design without extra water holes, which we called "MIX-25". It behaves in all the conditions analyzed in a similar way to the MIX-33 but it does present minor control problems. These can be solved by making small modifications to the control and safety systems, namely by enriching the boron-10 content of some boron absorbers. Thus, the addition of water holes replacing fuel pins helps to improve the MIX-33 performance and eliminate the difficulties seen in the MIX-25 design. We also performed a benchmarking analysis to test the code CASMO-3 to analyze WG-Pu assemblies, using the code MCNP-4A to compare. We found good agreement between CASMO-3 and

  19. 76 FR 39922 - Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power...

    Science.gov (United States)

    2011-07-07

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Office of New Reactors; Proposed Revision 4 to Standard Review Plan Section 8.1 on Electric Power... 8.1 on ``Electric Power-- Introduction.'' The notice period for this notice closes on June 30, 2011...

  20. InterProScan Result: BW999378 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999378 BW999378_3_ORF1 4DC407A334EB8726 SUPERFAMILY SSF103473 MFS general substra...te transporter 1.1e-08 T IPR016196 Major facilitator superfamily, general substrate transporter ...

  1. 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

  2. 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) ...

  3. InterProScan Result: BW999292 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW999292 BW999292_1_ORF2 3C8DCC56B3217B76 SUPERFAMILY SSF103473 MFS general substra...te transporter 8.9e-36 T IPR016196 Major facilitator superfamily, general substrate transporter ...

  4. InterProScan Result: BW997633 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

    Full Text Available BW997633 BW997633_3_ORF1 B98174A3666EA946 SUPERFAMILY SSF50677 ValRS/IleRS/LeuRS editing... domain 1.7e-19 T IPR009008 Valyl/Leucyl/Isoleucyl-tRNA synthetase, class Ia, editing Molecular Functio

  5. 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) ...

  6. InterProScan Result: BW999574 [KAIKOcDNA[Archive

    Lifescience Database Archive (English)

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

  7. 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 Biolog...ical Process: ossification (GO:0001503)|Biological Process: cell adhesion (GO:0007155) ...

  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. The ΣΣ secondary intermediate-energy standard neutron field development at the Romania TRIGA Reactor

    Science.gov (United States)

    Roth, C.; Gârlea, I.; Dobrea, D.; Aioanei, L.; Kelerman, C.; Gârlea, C.; Gugiu, D.; Datcu, A.; Preda, M.; Pavelescu, M.

    2004-04-01

    The ΣΣ intermediate-energy reference spectrum irradiation facility, operated until 1998 at the VVR-S Reactor from Bucharest, was put into operation at the Romania TRIGA Reactor. This paper presents the experimental devices developed for the ΣΣ system operating at the Steady State Reactor (SSR)-TRIGA core, including the monitoring system, and preliminary neutron characterization results. These results show that the spectrum characteristics are slowly deviating from those for the recommended ΣΣ spectrum. The high-energy neutron tail is slowly increasing, compared with the recommended ΣΣ spectrum and other similar facilities.

  10. 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.

  11. PWR Fuel Deposit Analysis at a B&W Plant with a 24 Month Fuel Cycle

    Science.gov (United States)

    Pop, Mike G.; Lamanna, Larry S.; Harne, Richard; Riddle, John

    The paper presents the crud analysis of a twice-burned fuel deposit using the AREVA crud sampling method in a domestic B&W PWR. The patented AREVA sampling method allowed the separation of deposit flakes that retain all the characteristics of unperturbed crud deposition.

  12. The protein crystallography beamline BW6 at DORIS - automatic operation and high-throughput data collection

    CERN Document Server

    Blume, H; Bourenkov, G P; Kosciesza, D; Bartunik, H D

    2001-01-01

    The wiggler beamline BW6 at DORIS has been optimized for de-novo solution of protein structures on the basis of MAD phasing. Facilities for automatic data collection, rapid data transfer and storage, and online processing have been developed which provide adequate conditions for high-throughput applications, e.g., in structural genomics.

  13. 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...

  14. 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...

  15. Short Baseline Reactor Antineutrino-Electron Scattering Experiments and Non-Standard Neutrino Interactions at Source and Detector

    OpenAIRE

    Khan, Amir N.; McKay, Douglas W.; Tahir, F

    2014-01-01

    We investigate non-standard interaction effects in antineutrino-electron scattering experiments with baselines short enough to ignore standard oscillation phenomena. The setup is free of ambiguities from the interference between new physics and oscillation effects and is sensitive to both semileptonic new physics at the source and purely leptonic new physics in the weak interaction scattering at the detector. We draw on the TEXONO experiment as the model system, extending its analysis of non-...

  16. Report detailing comparative analysis of results from high flux isotope reactor and national institute of standards technology small-angle neutron scattering experiments

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Littrell, Ken [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wells, Peter [Univ. of California, Santa Barbara, CA (United States); Cunningham, Nicholas J. [Univ. of California, Santa Barbara, CA (United States)

    2015-09-01

    discussed above, see Ref. [5] and [6] for details. UCSB has performed a large number of SANS experiments in the past at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR). These data are taken from RPV steels irradiated in a wide range of flux-fluence space and will be very useful in comparing to the upcoming UCSB ATR-2 irradiation characterization since most of the SANS experiments with ATR-2 materials will be performed at ORNL High Flux Isotope Reactor (HFIR). However in the previous report [7], some discrepancies were observed between HFIR and NCNR generated data. One of the hypotheses was that there was some kind of extra scattering occurring off the sample holders that results in the HFIR curves falling above the NCNR curves. To test this hypothesis, UCSB provided thermally aged samples that have been previously run at NCNR to ORNL for testing at HFIR while ORNL performed some improvements to experimental set up at HFIR. This report provides the status for the Level 3 Milestone (M3LW-15OR0402013), Complete report detailing comparative analysis of results from High Flux Isotope Reactor and National Institute of Standards and Technology small-angle neutron scattering experiments. This milestone is associated with small-angle neutron scattering characterization at the High Flux Isotope Reactor of various model alloys that had been previously characterized at NCNR by UCSB.

  17. 33 CFR 147.847 - Safety Zone; BW PIONEER Floating Production, Storage, and Offloading System Safety Zone.

    Science.gov (United States)

    2010-07-01

    ... Production, Storage, and Offloading System Safety Zone. 147.847 Section 147.847 Navigation and Navigable... ZONES § 147.847 Safety Zone; BW PIONEER Floating Production, Storage, and Offloading System Safety Zone. (a) Description. The BW PIONEER, a Floating Production, Storage and Offloading (FPSO) system, is in...

  18. 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.

  19. 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

  20. 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

  1. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  2. SU-E-I-62: Reduction of Susceptibility Artifacts by Increasing the Bandwidth (BW) and Echo Train Length (ETL)

    Energy Technology Data Exchange (ETDEWEB)

    Mavroidis, P [University of North Carolina, Chapel Hill, NC (United States); Boci, N; Kostopoulos, S; Ninos, C; Glotsos, D; Oikonomou, G; Bakas, A; Roka, V; Cavouras, D; Lavdas, E [Technological Education Institute of Athens, Athens, Attika (Greece); Sakkas, G; Tsagkalis, A [Animus Kyanoys Larissas Hospital, Larissa, Thessaly (Greece); Chatzivasileiou, V; Batsikas, G [IASO Thessalias Hospital, Larissa, Thessaly (Greece); Papanikolaou, N [University of Texas HSC SA, San Antonio, TX (United States); Stathakis, S [Cancer Therapy and Research Center, San Antonio, TX (United States)

    2015-06-15

    Purpose: The aim of this present study is to increase bandwidth (BW) and echo train length (ETL) in Proton Density Turbo Spin Echo (PD TSE) sequences with and without fat saturation (FS) as well as in Turbo Inversion Recovery Magnitude sequences (TIRM) in order to assess whether these sequences are capable of reducing susceptibility artifacts. Methods: We compared 1) TIRM coronal (COR) with the same sequence with increased both BW and ETL 2) Conventional PD TSE sagittal (SAG) with FS with an increased BW 3) Conventional PD TSE SAG without FS with an increased BW 4) Conventional PD TSE SAG without FS with increased both BW and ETL. A quantitative analysis was performed to measure the extent of the susceptibility artifacts. Furthermore, a qualitative analysis was performed by two radiologists in order to evaluate the susceptibility artifacts, image distortion and fat suppression. The depiction of cartilage, menisci, muscles, tendons and bone marrow were also qualitatively analyzed. Results: The quantitative analysis found that the modified TIRM sequence is significantly superior to the conventional one regarding the extent of the susceptibility artifacts. In the qualitative analysis, the modified TIRM sequence was superior to the corresponding conventional one in eight characteristics out of ten that were analyzed. The modified PD TSE with FS was superior to the corresponding conventional one regarding the susceptibility artifacts, image distortion and depiction of bone marrow and cartilage while achieving effective fat saturation. The modified PD TSE sequence without FS with a high (H) BW was found to be superior corresponding to the conventional one in the case of cartilage. Conclusion: Consequently, TIRM sequence with an increased BW and ETL is proposed for producing images of high quality and modified PD TSE with H BW for smaller metals, especially when FS is used.

  3. CONVECTION REACTOR

    Science.gov (United States)

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  4. CER. Research reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Estrade, Jerome [CEA, DEN, DER, Saint-Paul-lez-Durance (France). Jules Horowitz Reactor (JHR)

    2012-10-15

    Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)

  5. 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.

  6. Surface grain refinement mechanism of SMA490BW steel cross joints by ultrasonic impact treatment

    Science.gov (United States)

    He, Bo-lin; Xiong, Lei; Jiang, Ming-ming; Yu, Ying-xia; Li, Li

    2017-04-01

    Ultrasonic impact treatment (UIT) is a postweld technique for improving the fatigue strength of welded joints. This technique makes use of ultrasonic vibration to impact and plastically deform a weld toe and can achieve surface grain refinement of the weld toe, which is considered as the main reason for the improvement of fatigue strength. In this paper, the microstructure of the surface of a treated weld toe was observed by metallographic microscopy and transmission electron microscopy (TEM). The results show that UIT could produce severe plastic deformation on the surface layer of the weld toe and the maximum depth of plastic deformation extended to approximately 260 μm beneath the treated surface. Repeated processing could exacerbate the plastic deformation on the surface layer, resulting in finer grains. We can conclude that the surface grain refinement mechanism of SMA490BW welded joints is related to the high density of dislocation tangles and dislocation walls.

  7. REACTOR COOLING

    Science.gov (United States)

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  8. 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.

  9. 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

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

    Science.gov (United States)

    Segarra, I; Roca, M; Baliellas, C; Vilar, L; Ricart, Y; Mora, J; Puchal, R; Martin-Comin, J

    1991-01-01

    Thirty-three patients suspected of suffering from inflammatory bowel disease were studied. Autologous leucocytes were labelled with indium 111 oxine and re-injected simultaneously with 0.3-0.5 mg of technetium 99m granulocyte-specific monoclonal antibody BW 250/183. Two scans were obtained, the early scan 3-4 h postinjection (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 111In scans was 88.8%, 100.0% and 93.7% at 4 h and 94.7%, 100.0% and 96.9% at 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 leucocytes, BW 250/183 granulocyte-specific scintigraphy can be used for inflammatory bowel disease diagnosis and localization.

  11. 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.).

  12. Effects of the 5-HT2B receptor agonist, BW 723C86, on three rat models of anxiety.

    OpenAIRE

    Kennett, G. A.; Bright, F.; Trail, B.; Baxter, G. S.; Blackburn, T. P.

    1996-01-01

    1. BW 723C86 (3 and 10 mg kg-1, s.c. 30 min pretest), a 5-HT2B receptor agonist, increased total interaction, but not locomotion in a rat social interaction test, a profile consistent with anxiolysis. 2. The effect of BW 723C86 in the social interaction test is likely to be 5-HT2B receptor-mediated as it was prevented by pretreatment with the 5-HT2C/2B receptor antagonist, SB 200646A, (1 and 2 mg kg-1, p.o., 1 h pretest) which did not affect basal levels of social interaction at the doses use...

  13. 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.

  14. 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.

  15. 75 FR 19880 - Safety Zone; BW PIONEER at Walker Ridge 249, Outer Continental Shelf FPSO, Gulf of Mexico

    Science.gov (United States)

    2010-04-16

    ... shuttle tanker will connect its bow to the FPSO BW PIONEER and its stern to an attendant tug that will.... Basis and Purpose The Coast Guard is establishing a safety zone of 500 meters around the stern of the...) around the stern of the FPSO when it is moored to the turret buoy is a safety zone. If the FPSO detaches...

  16. Evidence of a Black Hole in the X-Ray Transient GS 1354-64 (=BW Circini)

    NARCIS (Netherlands)

    Casares, J.; Zurita, C.; Shahbaz, T.; Charles, P.A.; Fender, R.P.

    2004-01-01

    We present the first radial velocity curve of the companion star to BW Cir that demonstrates the presence of a black hole in this X-ray transient that recorded outbursts in 1987 and 1997 (and possibly 1971-1972). We identify weak absorption features corresponding to a G0-5 III donor star, strongly

  17. 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.

  18. NUCLEAR REACTOR

    Science.gov (United States)

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  19. 40 CFR 63.1406 - Reactor batch process vent provisions.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

  20. Heating device for nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shiratori, Yoshitake; Ijima, Takashi; Katano, Yoshiaki; Saito, Masaki

    1996-05-31

    The present invention provides a control system of a heating device which elevates the temperature of a reactor from a normal temperature to an operation temperature by using a nuclear heating. Namely, the device of the present invention comprises (1) means for detecting reactor temperature, (2) means for detecting reactor power, (3) means for memorizing the corresponding relation of each value of the means (1) and means (2) as standard data when temperature is elevated at a predetermined temperature elevation rate, (4) means for calculating the power corresponding to the current temperature based on the standard data upon elevation of the reactor temperature, and (5) means for controlling the progress or retraction of the power control material of the reactor core based on the power calculated by the means (4). With such a constitution, since the current reactor power elevation rate corresponding to the coolants is controlled based on the standard data upon actual start-up of the reactor, the control for the temperature of coolants can be facilitated. (I.S.)

  1. Studying at the Universities of the Armed Forces: Comparative Results from the 1999 Student Questionnaire (Studieren an den UniBw: Vergleichende Ergebnisse aus der Studentenbefragung 1999)

    Science.gov (United States)

    2002-12-01

    Phase des Studiums 10 6. Das kommunikative Miteinander 13 7. Die Studienanforderungen und das Erleben der Studiensituation an den UniBw 16 7.1. Das...studenti- schen Eigenkultur an der UniBw indifferent bleibt oder ffir AuBenstehende jedenfalls nicht erkenn- bar ist. 6. Das kommunikative Miteinander Der...gerweise war schon im Grandungskonzept der UniBw diesem Moment einer m~glichst nahen, en- gen und transparenten Kommunikation besondere Beachtung

  2. Catalytic combustion of propane in a membrane reactor with separate feed of reactants - III. Role of catalyst load on reactor performance

    NARCIS (Netherlands)

    Saracco, G; Veldsink, JW; Versteeg, GF; van Swaaij, WPM

    1996-01-01

    This paper deals with a pilot plant study on the catalytic combustion of propane in a membrane reactor with separate feed of reactants. The importance of the amount of catalyst (1% b.w. Pt on gamma-Al2O3), deposited on the pore walls of the membrane, is investigated. Two membranes were prepared by

  3. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry 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 neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  4. NEUTRONIC REACTORS

    Science.gov (United States)

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  5. Association of Two Polymorphisms in H2B.W Gene with Azoospermia and Severe Oligozoospermia in An Iranian Population

    Directory of Open Access Journals (Sweden)

    Haleh Haji Ebrahim Zargar

    2015-07-01

    Full Text Available Background: During spermatogenesis, the H2B family, member W (H2B.W gene, encodes a testis specific histone that is co-localized with telomeric sequences and has the potential role to mediate the sperm-specific chromatin remodeling. Previously H2B.W genetic variants were reported to be involved in susceptibility to spermatogenesis impairment. In the present study, two single nucleotide polymorphisms (SNPs in 5΄UTR and exon 1 of H2B.W gene were examined to investigate possible association of these polymorphisms with male infertility in Iranian population. Materials and Methods: This case control study was conducted in Royan institute during four-year period (2010–2013. Genetic alteration of two SNPs loci, −9C>T and 368A>G, in H2B.W gene were indicated in 92 infertile men who were divided into two main groups including azoospermia (n=46 and sever oligozoospermia (n=46, while there was 60 fertile men as control group. Azoosperima was also divided into three sub-groups including sertoli cell only syndrome (SCOS, n=21, complete maturation arrest (CMA, n=17 and hypo spermatogenesis (n=8 according to testicular biopsy. For analysis, polymerase chain reaction-restriction fragment length polymorphism (PCR-RFLP technique was applied. Results: The frequency of allele −9T was significantly higher in CMA group than in patients with SCOS (PT and 368A>G had no significant difference between the infertile groups and control (P=0.859 and P=0.812, respectively. Conclusion: This investigation showed that SNP −9C>T might be contribute to CMA in azoospermic patients and SNP 368A>G had no correlation with male infertility in Iranian population.

  6. Optimal dietary lysine to threonine ratio in pigs (30-110 kg BW) derived from observed dietary amino acid efficiency.

    Science.gov (United States)

    Wecke, C; Liebert, F

    2010-12-01

    Optimal ratio of lysine (Lys) to threonine (Thr) in diets for growing barrows [genotype: Piétrain × (Large White × German Landrace)] was established at 30, 50, 70, 90 and 110 kg body weight (BW). N balance studies were conducted with diets based on constant mixture of wheat, barley, soybean meal and field peas. Supplementation of crystalline amino acids (AA) provided diets with Lys (diet A) and Thr (diet B) in first limiting position. For each BW range, totally 28 individual N balance data were utilized for assessing model parameters (exponential N utilization model) of dietary efficiency of Lys and Thr, respectively. Observed maximal dietary efficiency of Lys and Thr provided the database for conclusion of optimal dietary Lys to Thr ratio. Based on applied experimental conditions, BW dependent Lys to Thr ratio could not be concluded. On average, the dietary ratio Lys to Thr = 1:0.61 ± 0.02 was obtained as optimal for growing pigs. © 2010 The Authors. Journal of Animal Physiology and Animal Nutrition © 2010 Blackwell Verlag GmbH.

  7. Neutronic reactor

    Science.gov (United States)

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  8. Neutronic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Menegus, R.L.; Wende, C.W.

    1983-01-04

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  9. 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.

  10. 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)

  11. Immunoscintigraphy of septic loosening of knee endoprosthesis: a retrospective evaluation of the antigranulocyte antibody BW 250/183

    Energy Technology Data Exchange (ETDEWEB)

    Klett, Rigobert; Khalisi, Alexander; Puille, Maximillian; Steiner, Dagmar; Bauer, Richard [Clinic of Nuclear Medicine, University Hospital Giessen, Friedrichstrasse 25, 35385, Giessen (Germany); Kordelle, Jens [Orthopedic Clinic, University Hospital Giessen (Germany); Stahl, Ulrich [Department of Pathology, University Hospital Giessen (Germany)

    2003-11-01

    Immunoscintigraphy with the use of the antigranulocyte antibody BW 250/183 is a reliable method for detecting infection, especially in septic loosening of hip prostheses, for which purpose quantitative interpretation of the time-activity course is employed. Therefore, we retrospectively studied whether similar results could be achieved in knee prostheses. We verified 28 scintigraphic examinations in 26 patients by histology. Scintigraphy was performed during an early (4-6 h post injection) and a late phase (23-25 h post injection). Infection was diagnosed when activity around the knee prosthesis increased by more than 10% compared with bone marrow. We found a sensitivity and a negative predictive value of 100%, a specificity of 80%, a positive predictive value of 81% and an accuracy of 89%. Specificity and accuracy are lower than in the evaluation of hip prostheses; however, in comparison to other scintigraphic modalities, scintigraphy with the antigranulocyte antibody BW 250/183 is superior in excluding infection and has better specificity and accuracy. Therefore, as in the case of hip prostheses, the described methodology appears to be the scintigraphic modality of choice for knee prostheses. (orig.)

  12. Nuclear Reactors. Revised.

    Science.gov (United States)

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  13. Photocatalytic reactor

    Science.gov (United States)

    Bischoff, Brian L.; Fain, Douglas E.; Stockdale, John A. D.

    1999-01-01

    A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

  14. Hybrid adsorptive membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    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.

  15. 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...

  16. Reactor transient

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.L.

    1956-05-31

    The authors are planning a calculation to be done on the Univac at the Louviers Building to estimate the effect of xenon transients, a high reactor power. This memorandum outlines the reasons why they prefer to do the work at Louviers rather than at another location, such as N.Y.U. They are to calculate the response of the reactor to a sudden change in position of the half rods. Qualitatively, the response will be a change in the rooftop ratio of the neutron flux. The rooftop ratio may oscillate with high damping, or, instead, it may oscillate for many cycles. It has not been possible for them to determine this response by hand calculation because of the complexity of the problem, and yet it is important for them to be certain that high power operation will not lead us to inherently unstable operation. Therefore they have resorted to machine computation. The system of differential equations that describes the response has seven dependent variables; therefore there are seven equations, each coupled with one or more of the others. The authors have discussed the problem with R.R. Haefner at the plant, and it is his opinion that the IBM 650 cannot adequately handle the system of seven equations because the characteristic time constants vary over a range of about 10{sup 8}. The Univac located at the Louviers Building is said to be satisfactory for this computation.

  17. Impaired development of melanoblasts in the black-eyed white Mitf(mi-bw) mouse, a model for auditory-pigmentary disorders.

    Science.gov (United States)

    Hozumi, Hiroki; Takeda, Kazuhisa; Yoshida-Amano, Yasuko; Takemoto, Yuji; Kusumi, Ryota; Fukuzaki-Dohi, Urara; Higashitani, Atsushi; Yamamoto, Hiroaki; Shibahara, Shigeki

    2012-06-01

    Microphthalmia-associated transcription factor (Mitf) is a regulator for differentiation of melanoblasts that are derived from the neural crest. The mouse homozygous for the black-eyed white (Mitf(mi-bw)) allele is characterized by the white coat color and deafness, with black eye that is associated with the lack of melanocytes in skin and inner ear. The Mitf(mi-bw) mutation is an insertion of the LINE1 retrotransposable element into intron 3 of the Mitf gene that causes the selective deficiency of the melanocyte-specific Mitf isoform, Mitf-M. Here, we show the expression of Mitf-M mRNA in the trunk region of the homozygous Mitf(mi-bw)(bw) mouse at embryonic days (E) 11.5 and E12.5, but Mitf-M mRNA is undetectable at E13.5. In addition, using bw mouse that carries the lacZ transgene under the control of a melanoblast-specific promoter, we show that the number of migrating melanoblasts in bw embryos was less than 10% of that in control embryos at E11.5 and E12.5, and melanoblasts disappear by E13.5. The loss of melanoblasts in bw embryos was probably caused by apoptosis. Finally, forced expression of Mitf-M in the cultured neural tube of bw embryos ensured the differentiation of melanoblasts. Therefore, the correct dose of Mitf-M is required for the normal development of melanoblasts. © 2012 The Authors. Journal compilation © 2012 by the Molecular Biology Society of Japan/Blackwell Publishing Ltd.

  18. Final Report on Isotope Ratio Techniques for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Mitchell, Mark R.; Meriwether, George H.; Reid, Bruce D.

    2009-07-01

    The Isotope Ratio Method (IRM) is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods.

  19. 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.

  20. Research on Hot Compressive Deformation of TiBw/TA15 Composite Cylindrical Billet with Network-Distributed Reinforcements

    Science.gov (United States)

    Zhang, Rui; Wang, DongJun; Yuan, ShiJian

    2017-10-01

    A TA15 matrix composite reinforced by 3.5 vol.% TiB whiskers with a network microstructure was fabricated by the reaction hot pressing method. After sintering, the composite developed a net-like microstructure. To improve the mechanical properties, a cylindrical billet of the as-sintered TiBw/TA15 composite was subjected to upsetting and swaging operations with intermediate heating at 1000°C. The original equiaxed network units were stretched slightly, and the matrix had a microstructure evolution from a basket structure to a bi-modal structure after hot forging. The results of tensile tests at room temperature showed that the ultimate tensile strength and elongation of the composite increased by 8.2% and 130%, respectively, while the yield strength appeared to remain constant. Furthermore, the mechanical properties of the as-forged composite also improved at an elevated temperature.

  1. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  2. Maintenance of fission and fusion reactors. 10. workshop on fusion reactor engineering

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    This report contains copies of OHP presented at the title meeting. The presented topics are as follows, maintenance of nuclear power plants and ITER, exchange of shroud in BWR type reactors, deterioration of fission and fusion reactor materials, standards of pressure vessels, malfunction diagnosis method with neural network. (J.P.N.)

  3. 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.

  4. 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.

  5. Significantly enhanced creep resistance of low volume fraction in-situ TiBw/Ti6Al4V composites by architectured network reinforcements.

    Science.gov (United States)

    Wang, S; Huang, L J; Geng, L; Scarpa, F; Jiao, Y; Peng, H X

    2017-01-17

    We present a new class of TiBw/Ti6Al4V composites with a network reinforcement architecture that exhibits a significant creep resistance compared to monolithic Ti6Al4V alloys. Creep tests performed at temperatures between 773 K and 923 K and stress range of 100 MPa-300 MPa indicate both a significant improvement of the composites creep resistance due to the network architecture made by the TiB whiskers (TiBw), and a decrease of the steady-state creep rates by augmenting the local volume fractions of TiBw in the network region. The deformation behavior is driven by a diffusion-controlled dislocation climb process. Moreover, the activation energies of these composites are significantly higher than that of Ti6Al4V alloys, indicating a higher creep resistance. The increase of the activation energy can be attributed to the TiBw architecture that severely impedes the movements of dislocation and grain boundary sliding and provides a tailoring of the stress transfer. These micromechanical mechanisms lead to a remarkable improvement of the creep resistance of these networked TiBw/Ti6Al4V composites featuring the special networked architecture.

  6. Significantly enhanced creep resistance of low volume fraction in-situ TiBw/Ti6Al4V composites by architectured network reinforcements

    Science.gov (United States)

    Wang, S.; Huang, L. J.; Geng, L.; Scarpa, F.; Jiao, Y.; Peng, H. X.

    2017-01-01

    We present a new class of TiBw/Ti6Al4V composites with a network reinforcement architecture that exhibits a significant creep resistance compared to monolithic Ti6Al4V alloys. Creep tests performed at temperatures between 773 K and 923 K and stress range of 100 MPa-300 MPa indicate both a significant improvement of the composites creep resistance due to the network architecture made by the TiB whiskers (TiBw), and a decrease of the steady-state creep rates by augmenting the local volume fractions of TiBw in the network region. The deformation behavior is driven by a diffusion-controlled dislocation climb process. Moreover, the activation energies of these composites are significantly higher than that of Ti6Al4V alloys, indicating a higher creep resistance. The increase of the activation energy can be attributed to the TiBw architecture that severely impedes the movements of dislocation and grain boundary sliding and provides a tailoring of the stress transfer. These micromechanical mechanisms lead to a remarkable improvement of the creep resistance of these networked TiBw/Ti6Al4V composites featuring the special networked architecture.

  7. 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.

  8. Examination of fast reactor fuels, FBR analytical quality assurance standards and methods, and analytical methods development: irradiation tests. Progress report, April 1--June 30, 1976, and FY 1976. [UO/sub 2/; PuO/sub 2/

    Energy Technology Data Exchange (ETDEWEB)

    Baker, R.D. (comp.)

    1976-08-01

    Characterization of unirradiated and irradiated LMFBR fuels by analytical chemistry methods will continue, and additional methods will be modified and mechanized for hot cell application. Macro- and microexaminations will be made on fuel and cladding using the shielded electron microprobe, emission spectrograph, radiochemistry, gamma scanner, mass spectrometers, and other analytical facilities. New capabilities will be developed in gamma scanning, analyses to assess spatial distributions of fuel and fission products, mass spectrometric measurements of burnup and fission gas constituents and other chemical analyses. Microstructural analyses of unirradiated and irradiated materials will continue using optical and electron microscopy and autoradiographic and x-ray techniques. Analytical quality assurance standards tasks are designed to assure the quality of the chemical characterizations necessary to evaluate reactor components relative to specifications. Tasks include: (1) the preparation and distribution of calibration materials and quality control samples for use in quality assurance surveillance programs, (2) the development of and the guidance in the use of quality assurance programs for sampling and analysis, (3) the development of improved methods of analysis, and (4) the preparation of continuously updated analytical method manuals. Reliable analytical methods development for the measurement of burnup, oxygen-to-metal (O/M) ratio, and various gases in irradiated fuels is described.

  9. Distribution of [sup 65]Zn labeled alpha-fetoprotein during proliferation of the BW7756 murine hepatoma

    Energy Technology Data Exchange (ETDEWEB)

    Keenan, J.F.

    1990-01-01

    The radiolabeling of alpha-fetoprotein (AFP) with [sup 65]Zn for the determination of its biodistribution was studied in mice bearing the BW7756 murine hepatoma as compared to that found with normal mice. AFP is an oncofetal protein of about 70,000 daltons associated with pregnancy and certain cancers (e.g., hepatoma). The AFP was purified from mouse amniotic fluid (MAF) using polyacrylamide gel electrophoresis (PAGE) and higher performance liquid chromatography (HPLC). The biological activity of AFP was maintained through the separation procedures and the purity was determined using double immunodiffusion (DID), immunoelectrophoresis (IEP) and sodium dodecyl sulfate electrophoresis (SDS). The labeling procedures included removal of intrinsic metal with EDTA, incubation with radiotracer ([sup 65]Zn) and buffer, followed by removal of unbound [sup 65]Zn using gel filtration chromatography. The results correlated well with Zn fluctuations recorded by other techniques (RIXRF, radiotracer [sup 65]Zn). Large amounts of [sup 65]Zn-AFP were localized in the liver, spleen and tumor with significant elevations above normal in the log growth phase (day 14-18). [sup 65]Zn-AFP levels in the skin, pancreas, brain and thyroid decreased as the tumor mass increased. Tumor [sup 65]Zn-AFP uptake increased with time but leveled off in the late log phase (day 21) due to tumor necrosis. In light of the results of this investigation, and previous work stating that AFP binds Zn with a higher affinity than does albumin, it is suggestive that the Zn fluctuations observed in the earlier hepatoma studies were due to the in vivo binding of Zn to AFP. These results confirm the thesis that intrinsic labeling (replacement of naturally bound ligands with radioactive analogs) does not alter the biochemical integrity as non-intrinsic labeling (e.g., Iodine) may.

  10. Hybrid plasmachemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lelevkin, V. M., E-mail: lelevkin44@mail.ru; Smirnova, Yu. G.; Tokarev, A. V. [Kyrgyz-Russian Slavic University (Kyrgyzstan)

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  11. Attrition reactor system

    Science.gov (United States)

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  12. Guidebook to nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen.

  13. Postmortem lung volume/body weight standards for term and preterm infants.

    Science.gov (United States)

    De Paepe, Monique E; Shapiro, Svetlana; Hansen, Katrine; Gündoğan, Füsun

    2014-01-01

    Assessment of lung growth is a critical component of the perinatal autopsy. Increased lung liquid content may lead to overestimation of lung growth based on (wet) lung weight. In contrast, lung volume is not influenced by intraalveolar lung liquid. Our aim was to establish age-specific reference values for postmortem lung volume/BW in preterm and term infants. We performed a retrospective analysis of fetuses/infants (16-41 weeks' gestation) without (N = 134) or with (N = 79) risk factors for pulmonary hypoplasia. Lungs were inflated at standardized pressure and volumes determined by water immersion method. Lung volume increased 11-fold between 16 and 41 weeks' gestation, concomitant with a 16-fold increase in BW. Mean lung volume/BW remained constant at 33-34 ml/kg between 16 and 31 weeks' gestation and decreased to 23.4 ml/kg at term. Lung volume/BW of infants with severe risk factors (renal anomalies, diaphragmatic hernia) was significantly lower than age-matched standards. In this group, all fetuses/infants diagnosed as having lung hypoplasia by lung volume/BW also had lung hypoplasia LW/BW standards. However, in infants with "softer" risk factors (rupture of membranes, chromosomal anomalies), 5/26 cases diagnosed with lung hypoplasia based on lung volume/BW had normal LW/BW ratios. In these discrepant cases, lung sections showed significant inflammation and edema, likely accounting for increased wet lung weight. In conclusion, we determined age-specific lung volume/BW reference values for preterm and term infants. In selected situations assessment of lung volume/BW may represent a useful complementary tool to LW/BW for postmortem evaluation of lung size. Copyright © 2013 Wiley Periodicals, Inc.

  14. Schadevergoeding voor derden in personenschadezaken. Een rechtsvergelijkende studie naar de artikelen 6:107 en 6:108 BW in relatie tot het onrechtmatige daadsrecht en schadevergoedingsrecht

    NARCIS (Netherlands)

    Rijnhout, R.

    2012-01-01

    The aim of this research is to consider whether the articles 6:107 and 6:108 BW should be reconsidered. These articles grant third parties an action to bring an independent claim against the tortfeasor in personal injury cases. The characteristics of these articles are threefold. First, these claims

  15. 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

  16. NUCLEAR REACTOR CONTROL SYSTEM

    Science.gov (United States)

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  17. 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).

  18. Nuclear reactor overflow line

    Science.gov (United States)

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. Neutron fluxes in test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Youinou, Gilles Jean-Michel [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

  5. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    OpenAIRE

    Shirvan, Koroush; Kazimi, Mujid

    2016-01-01

    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 ...

  6. Note: Comparison of grazing incidence small angle x-ray scattering of a titania sponge structure at the beamlines BW4 (DORIS III) and P03 (PETRA III)

    Energy Technology Data Exchange (ETDEWEB)

    Rawolle, M.; Koerstgens, V.; Ruderer, M. A.; Metwalli, E.; Guo, S.; Mueller-Buschbaum, P. [Lehrstuhl fuer Funktionelle Materialien, Physik-Department, Technische Universitaet Muenchen, James-Franck-Str. 1, 85748 Garching (Germany); Herzog, G.; Benecke, G.; Schwartzkopf, M.; Buffet, A.; Perlich, J.; Roth, S. V. [HASYLAB at DESY, Notkestr. 85, 22607 Hamburg (Germany)

    2012-10-15

    Grazing incidence small angle x-ray scattering (GISAXS) is a powerful technique for morphology investigation of nanostructured thin films. GISAXS measurements at the newly installed P03 beamline at the storage ring PETRA III in Hamburg, Germany, are compared to the GISAXS data from the beamline BW4 at the storage ring DORIS III, which had been used extensively for GISAXS investigations in the past. As an example, a titania thin film sponge structure is investigated. Compared to BW4, at beamline P03 the resolution of larger structures is slightly improved and a higher incident flux leads to a factor of 750 in scattered intensity. Therefore, the acquisition time in GISAXS geometry is reduced significantly at beamline P03.

  7. HORIZONTAL BOILING REACTOR SYSTEM

    Science.gov (United States)

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  8. 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...

  9. THERMAL NEUTRONIC REACTOR

    Science.gov (United States)

    Spinrad, B.I.

    1960-01-12

    A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

  10. 49 CFR 178.61 - Specification 4BW welded steel cylinders with electric-arc welded longitudinal seam.

    Science.gov (United States)

    2010-10-01

    ... exceed 1/8 inch per minute during yield strength determination. (k) Elongation. Physical test specimens... and must meet the requirements of CGA Pamphlet C-3. (2) Guided bend test. A root test specimen must be... specified for American Standard Taper Pipe threads. (iii) Straight threads, having at least 4 engaged...

  11. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  12. Development of reactor graphite

    Science.gov (United States)

    Haag, G.; Mindermann, D.; Wilhelmi, G.; Persicke, H.; Ulsamer, W.

    1990-04-01

    The German graphite development programme for High Temperature Reactors has been based on the assumption that reactor graphite for core components with lifetime fluences of up to 4 × 10 22 neutrons per cm 2 (EDN) at 400°C can be manufactured from regular pitch coke. The use of secondary coke and vibrational moulding techniques have allowed production of materials with very small anisotropy, high strength, and high purity which are the most important properties of reactor graphite. A variety of graphite grades has been tested in fast neutron irradiation experiments. The results show that suitable graphites for modern High Temperature Reactors with spherical fuel elements are available.

  13. Oscillatory flow chemical reactors

    National Research Council Canada - National Science Library

    Slavnić Danijela S; Bugarski Branko M; Nikačević Nikola M

    2014-01-01

    .... 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...

  14. Membrane reactors at Degussa.

    Science.gov (United States)

    Wöltinger, Jens; Karau, Andreas; Leuchtenberger, Wolfgang; Drauz, Karlheinz

    2005-01-01

    The review covers the development of membrane reactor technologies at Degussa for the synthesis of fine chemicals. The operation of fed-batch or continuous biocatalytic processes in the enzyme membrane reactor (EMR) is well established at Degussa. Degussa has experience of running EMRs from laboratory gram scale up to a production scale of several hundreds of tons per year. The transfer of the enzyme membrane reactor from biocatalysis to chemical catalysis in the chemzyme membrane reactor (CMR) is discussed. Various homogeneous catalysts have been investigated in the CMR, and the scope and limitation of this new technique is discussed.

  15. Pressurizing new reactors

    Energy Technology Data Exchange (ETDEWEB)

    Neill, J.S.

    1956-01-30

    The Technical Division was asked recently to consider designs for new reactors that would add 8000 MW capacity to the Savannah River Plant. One modification of the existing SRP design that would enable a higher power rating, and therefore require fewer new reactors, is an increase in the maximum pressure in the D{sub 2}O system. The existing reactors at SRP are designed for a maximum pressure in the gas plenum of only 5 psig. Higher pressures enable higher D{sub 2} temperatures and higher sheath temperatures without local boiling or burnout. The requirements in reactor cooling facilities at any given power level would therefore be reduced by pressurizing.

  16. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  17. Results of the Nucifer reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

    2016-07-01

    Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

  18. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.

    1962-06-25

    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  19. 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)

  20. 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.

  1. REFLECTOR FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  2. 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.

  3. 40 CFR 63.1407 - Non-reactor batch process vent provisions.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Non-reactor batch process vent... § 63.1407 Non-reactor batch process vent provisions. (a) Emission standards. (1) Owners or operators of non-reactor batch process vents located at new or existing affected sources with 0.25 tons per year (0...

  4. Mirror reactor surface study

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, A. L.; Damm, C. C.; Futch, A. H.; Hiskes, J. R.; Meisenheimer, R. G.; Moir, R. W.; Simonen, T. C.; Stallard, B. W.; Taylor, C. E.

    1976-09-01

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included.

  5. 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.

  6. 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.

  7. Nuclear reactor control column

    Science.gov (United States)

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  8. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  9. Reactor power monitoring device

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Naotaka; Igawa, Shinji; Kitazono, Hideaki

    1998-02-13

    The present invention provides a reactor power monitoring device capable of ensuring circumstance resistance, high reliability and high speed transmission even if an APRM is disposed in a reactor building (R/B). Namely, signal processing sections (APRM) for transmitting data to a central control chamber are distributed in the reactor building at an area at the lowest temperature among areas where the temperature control in an emergency state is regulated, and a transmission processing section (APRM-I/F) for transmitting data to the other systems is disposed to the central control chamber. An LPRM signal transmission processing section is constituted such that LPRM signals can be transmitted at a high speed by DMA. Set values relevant to reactor tripping (neutron flux high, thermal output high and sudden reduction of a reactor core flow rate) are stored in the APRM-I/F, and reactor tripping calculation is conducted in the APRM-I/F. With such procedure, a reactor power monitoring device having enhanced control function can be attained. (N.H.)

  10. Nuclear reactor reflector

    Science.gov (United States)

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  11. 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.

  12. Reactor flux calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lhuillier, D. [Commissariat à l' Énergie Atomique et aux Énergies Alternatives, Centre de Saclay, IRFU/SPhN, 91191 Gif-sur-Yvette (France)

    2013-02-15

    The status of the prediction of reactor anti-neutrino spectra is presented. The most accurate method is still the conversion of total β spectra of fissionning isotopes as measured at research reactors. Recent re-evaluations of the conversion process led to an increased predicted flux by few percent and were at the origin of the so-called reactor anomaly. The up to date predictions are presented with their main sources of error. Perspectives are given on the complementary ab-initio predictions and upcoming experimental cross-checks of the predicted spectrum shape.

  13. NEUTRONIC REACTOR CONTROL

    Science.gov (United States)

    Metcalf, H.E.

    1958-10-14

    Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

  14. Nuclear reactor reflector

    Science.gov (United States)

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  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. Effects of Hot-Hydrostatic Canned Extrusion on the Stock Utilization, Microstructure and Mechanical Properties of TiBw/TC4 Composites with Quasi-Continuous Network.

    Science.gov (United States)

    Feng, Yangju; Li, Bing; Cui, Guorong; Zhang, Wencong

    2017-10-25

    In-situ TiB whisker-reinforced Ti-6Al-4V (TC4) titanium matrix composites (TiBw/TC4) with quasi-continuous networks were successfully fabricated by vacuum hot-pressing sintering. The effects of the hot-hydrostatic canned extrusion on stock utilization, microstructure and mechanical properties of the TiBw/TC4 composites were investigated. It was satisfactory that the utilization of composites could be obviously improved by canned extrusion compared to that extruded without canned extrusion. The microstructure results showed that after canned extrusion the grain was refined and the TiB whiskers were distributed from a random array state to a state in which the whiskers were distributed along the extrusion direction. The properties testing results revealed that the tensile strength, the hardness and the ductility of the composites all significantly improved after extrusion due to the grain refinement and orientation of the TiB whisker caused by extrusion. Tensile fracture results showed that when the TiB whiskers were randomly distributed only part of them played a role in strengthening the matrix during the deformation process (as-sintered composites), while when the TiB whiskers were oriented all whiskers could strengthen the matrix during the tensile testing process (as-extruded composites).

  17. 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).

  18. 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)

  19. Technology selection for offshore underwater small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shivan, Koroush; Ballinger, Ronald; Buongiorno, Jacopo; Forsberg, Charles; Kazimi, Mujid; Todreas, Neil [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States)

    2016-12-15

    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 time frame. 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 CO{sub 2} 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.

  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. 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.

  2. 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....

  3. Natural convection reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Bernath, L.; Menegus, R.L.; Ring, H.F.

    1956-05-01

    A previous report described the conceptual design of a plutonium producing reactor that may be characterized as follows: Power output (2000 MW); cooling - (natural convection of light water through the reactor, up through a draft tube to an evaporative cooling pond, then back to the reactor, and fuel (400 to 500 tons of uranium enriched to 1.2% U-235). Because this reactor would be cooled by the natural convection of light water, it is believed that the construction costs would be significantly less than for a Savannah or Hanford type reactor. Such expensive items as water treatment and water pumping facilities would be eliminated entirely. The inventory of 500 tons of slightly enriched uranium, however, is an unattractive feature. It represents not only a large dollar investment but also makes the reactor less attractive for construction during periods of national emergency because of the almost certain scarcity of even slightly enriched uranium at that time. The Atomic Energy Commission asked that the design be reviewed with the objective of reducing the inventory of uranium, The results of this review are given in this report.

  4. Tax exemption for nuclear power plants. Eon, RWE and EnBW hope for billions Euro fuel tax exemptions - this is the only reason why so many reactors are still operated; Steuer-Aus fuer AKW. Eon, RWE und EnBW hoffen auf eine Steuerbefreiung der AKW-Brennelemente in Milliardenhoehe - nur deshalb sind viele Reaktoren noch am Netz

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2016-10-15

    Even after amortizations nuclear power plants cannot be operated with sufficient profit, this is the consequence of the increasing electricity production from renewable energies. The authors discuss the argument that the perspective of fuel tax exemptions is the economic factor for ongoing operation.

  5. 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.

  6. SPRAY CALCINATION REACTOR

    Science.gov (United States)

    Johnson, B.M.

    1963-08-20

    A spray calcination reactor for calcining reprocessin- g waste solutions is described. Coaxial within the outer shell of the reactor is a shorter inner shell having heated walls and with open regions above and below. When the solution is sprayed into the irner shell droplets are entrained by a current of gas that moves downwardly within the inner shell and upwardly between it and the outer shell, and while thus being circulated the droplets are calcined to solids, whlch drop to the bottom without being deposited on the walls. (AEC) H03 H0233412 The average molecular weights of four diallyl phthalate polymer samples extruded from the experimental rheometer were redetermined using the vapor phase osmometer. An amine curing agent is required for obtaining suitable silver- filled epoxy-bonded conductive adhesives. When the curing agent was modified with a 47% polyurethane resin, its effectiveness was hampered. Neither silver nor nickel filler impart a high electrical conductivity to Adiprenebased adhesives. Silver filler was found to perform well in Dow-Corning A-4000 adhesive. Two cascaded hot-wire columns are being used to remove heavy gaseous impurities from methane. This purified gas is being enriched in the concentric tube unit to approximately 20% carbon-13. Studies to count low-level krypton-85 in xenon are continuing. The parameters of the counting technique are being determined. The bismuth isotopes produced in bismuth irradiated for polonium production are being determined. Preliminary data indicate the presence of bismuth207 and bismuth-210m. The light bismuth isotopes are probably produced by (n,xn) reactions bismuth-209. The separation of uranium-234 from plutonium-238 solutions was demonstrated. The bulk of the plutonium is removed by anion exchange, and the remainder is extracted from the uranium by solvent extraction techniques. About 99% of the plutonium can be removed in each thenoyltrifluoroacetone extraction. The viscosity, liquid density, and

  7. 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...

  8. 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.

  9. 10 CFR Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements

    Science.gov (United States)

    2010-01-01

    ... vessel. These data will be used as described in section IV of appendix G to part 50. ASTM E 185-73, “Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels”; ASTM E 185-79, “Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels”; and ASTM...

  10. MERCHANT MARINE SHIP REACTOR

    Science.gov (United States)

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  11. 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.

  12. Nuclear reactor safety device

    Science.gov (United States)

    Hutter, Ernest

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  13. Heat dissipating nuclear reactor

    Science.gov (United States)

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  14. Reactor for exothermic reactions

    Science.gov (United States)

    Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

    1993-01-01

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

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

    Science.gov (United States)

    1985-02-01

    implies, this reactor uses gas as the primary coolant . The coolant has a higher exit temperature when leaving the core than the PWR water 6 AFWL-TN-84...nuclear reactors, coolants must be used to ensure material components are not subject to failure due to the temperature exceeding melting points...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept

  16. 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.

  17. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)

    2008-07-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  18. 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.

  19. Nuclear reactor apparatus

    Science.gov (United States)

    Wade, Elman E.

    1978-01-01

    A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

  20. 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.

  1. Safe reactor power

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.L.; Ring, H.F.; Bernath, L.

    1956-05-15

    The upper limit on reactor operating power is established not only by safety considerations during steady-state operation but also by the requirement that during an accident no permanent damage be inflicted upon the reactor or the fuel charge. Two general categories of accidents are recognized; they are the ``nuclear runaway`` and the ``loss of coolant flow`` incidents. In this memorandum an incident of the latter type is analyzed. It is assumed that the safety rods function normally, and a method is defined for establishing the highest operating power that may be permitted if the postulated accident is to do no damage.

  2. 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...

  3. 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.

  4. Technology Selection for Offshore Underwater Small Modular Reactors

    OpenAIRE

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

    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 ...

  5. 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.

  6. 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)

  7. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    Sep 4, 2015 ... criteria; passive shutdown and decay heat removal systems; fast breeder reactors in India. PACS No. 28.41.−i. 1. ... water reactors, mainly pressurized heavy water reactors (PHWRs) to extract ∼10 GWe capacity for ..... commissioning phase and most of the supporting systems have been commissioned and.

  8. Engineering reactors for catalytic reactions

    Indian Academy of Sciences (India)

    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 ...

  9. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  10. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    Sep 4, 2015 ... Home; Journals; Pramana – Journal of Physics; Volume 85; Issue 3. Fast reactor programme in India. P Chellapandi P R ... Keywords. Sodium fast reactor; design challenges; construction challenges; emerging safety criteria; passive shutdown and decay heat removal systems; fast breeder reactors in India.

  11. Coefficient of standardized total tract digestibility of phosphorus in ...

    African Journals Online (AJOL)

    정용대

    2016-12-14

    Dec 14, 2016 ... barrows (initial bodyweight (BW) ± standard deviation, 52.25 ± 2.57 kg) were allocated individually to metabolism cages. The experimental design was ... roles in metabolism and bone formation (NRC, 2012). ..... Degestibility of animal feed is based on the rate of intake and faecal output of dietary nutrients.

  12. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  13. 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.

  14. Nuclear reactor building

    Science.gov (United States)

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  15. JACKETED REACTOR FUEL ELEMENT

    Science.gov (United States)

    Smith, K.F.; Van Thyne, R.J.

    1958-12-01

    A fuel element is described for fast reactors comprised of a core of uranium metal containing material and a jacket around the core, the jacket consisting of from 2.5 to 15 percent of titanium, from 1 to 5 percent of niobium, and from 80 to 96.5 percent of vanadium.

  16. Nuclear rocket engine reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lanin, Anatoly

    2013-07-01

    Covers a new technology of nuclear reactors and the related materials aspects. Integrates physics, materials science and engineering Serves as a basic book for nuclear engineers and nuclear physicists. 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.

  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. Fossil fuel furnace reactor

    Science.gov (United States)

    Parkinson, William J.

    1987-01-01

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  19. NEUTRONIC REACTOR SHIELD

    Science.gov (United States)

    Fermi, E.; Zinn, W.H.

    1957-09-24

    The reactor radiation shield material is comprised of alternate layers of iron-containing material and compressed cellulosic material, such as masonite. The shielding material may be prefabricated in the form of blocks, which can be stacked together in ary desired fashion to form an effective shield.

  20. Neutronic Reactor Structure

    Science.gov (United States)

    Vernon, H. C.; Weinberg, A. M.

    1961-05-30

    The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

  1. NEUTRONIC REACTOR CONTROL ELEMENT

    Science.gov (United States)

    Beaver, R.J.; Leitten, C.F. Jr.

    1962-04-17

    A boron-10 containing reactor control element wherein the boron-10 is dispersed in a matrix material is describeri. The concentration of boron-10 in the matrix varies transversely across the element from a minimum at the surface to a maximum at the center of the element, prior to exposure to neutrons. (AEC)

  2. 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.

  3. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  4. Sequencing Bacth Reactors; Reactores biologicos secuenciados

    Energy Technology Data Exchange (ETDEWEB)

    Nolasco, D.; Manoharan, M.

    1999-06-01

    The application of sequencing batch reactors (SBR) for wastewater treatment is becoming increasingly popular. However, published information on process performance and construction costs for SBRs is scarce. For this reason. Environment Canada, the Ontario Ministry of the Environment (MOE), and the Water Environment Association of Ontario (WEAO) decided to sponsor a program to evaluate the performance of 75 municipal SBRs in Canada and the United States. Effluent quality, construction costs, and design and operating problems were investigated. Areas for optimization found as a result of this investigation were classified an prioritized based on their impact on operational costs, treatment capacity, effluent quality, and frequency of occurrence. A list of recommendations for process optimization was prepared. A construction cost comparison between activated sludge systems of continuous flow and SBRs was prepared. (Author) 12 refs.

  5. 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.

  6. 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...

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

    Directory of Open Access Journals (Sweden)

    Ding Wenfeng

    2014-10-01

    Full Text Available (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 discussed. 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 beneficial 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.

  8. A decision-tree model to detect post-calving diseases based on rumination, activity, milk yield, BW and voluntary visits to the milking robot.

    Science.gov (United States)

    Steensels, M; Antler, A; Bahr, C; Berckmans, D; Maltz, E; Halachmi, I

    2016-09-01

    Early detection of post-calving health problems is critical for dairy operations. Separating sick cows from the herd is important, especially in robotic-milking dairy farms, where searching for a sick cow can disturb the other cows' routine. The objectives of this study were to develop and apply a behaviour- and performance-based health-detection model to post-calving cows in a robotic-milking dairy farm, with the aim of detecting sick cows based on available commercial sensors. The study was conducted in an Israeli robotic-milking dairy farm with 250 Israeli-Holstein cows. All cows were equipped with rumination- and neck-activity sensors. Milk yield, visits to the milking robot and BW were recorded in the milking robot. A decision-tree model was developed on a calibration data set (historical data of the 10 months before the study) and was validated on the new data set. The decision model generated a probability of being sick for each cow. The model was applied once a week just before the veterinarian performed the weekly routine post-calving health check. The veterinarian's diagnosis served as a binary reference for the model (healthy-sick). The overall accuracy of the model was 78%, with a specificity of 87% and a sensitivity of 69%, suggesting its practical value.

  9. Evidence of mercury trapping in biofilm-EPS and mer operon-based volatilization of inorganic mercury in a marine bacterium Bacillus cereus BW-201B.

    Science.gov (United States)

    Dash, Hirak R; Basu, Subham; Das, Surajit

    2017-04-01

    Biofilm-forming mercury-resistant marine bacterium Bacillus cereus BW-201B has been explored to evident that the bacterial biofilm-EPS (exopolymers) trap inorganic mercury but subsequently release EPS-bound mercury for induction of mer operon-mediated volatilization of inorganic mercury. The isolate was able to tolerate 50 ppm of mercury and forms biofilm in presence of mercury. mer operon-mediated volatilization was confirmed, and -SH was found to be the key functional group of bacterial EPS responsible for mercury binding. Biofilm-EPS-bound mercury was found to be internalized to the bacterial system as confirmed by reversible conformational change of -SH group and increased expression level of merA gene in a timescale experiment. Biofilm-EPS trapped Hg after 24 h of incubation, and by 96 h, the volatilization process reaches to its optimum confirming the internalization of EPS-bound mercury to the bacterial cells. Biofilm disintegration at the same time corroborates the results.

  10. 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...

  11. Nuclear reactor construction with bottom supported reactor vessel

    Science.gov (United States)

    Sharbaugh, John E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment

  12. Reactor core stability monitoring method for BWR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanemoto, Shigeru; Ebata, Shigeo.

    1992-09-01

    In an operation for a BWR type reactor, reactor power is usually increased or decreased by controlling both of control rods and reactor core flow rate. Under a certain condition, the reactor core is made unstable by the coupling of nuclear and thermohydrodynamic characteristics in the reactor. Therefore, the reactor power and the reactor core flow rate are changed within a range predetermined by a design calculation. However, if reactor core stability can be always measured and monitored, it is useful for safe operation, as well as an existent operation range can be extended to enable more effective operation. That is, autoregressive a coefficient is determined successively on real time based on fluctuation components of neutron flux signals. Based on the result, an amplification ratio, as a typical measure of the reactor core stability, is determined on a real time. A time constant of the successive calculation for the autoregressive coefficient can be made variable by the amplification ratio. Then, the amplification ratio is estimated at a constant accuracy. With such procedures, the reactor core stability can be monitored successively in an ON-line manner at a high accuracy, thereby enabling to improve the operation performance. (I.S.).

  13. BEACON TSM application system to the operation of PWR reactors; Aplicacion del Sistema BEACON TSM a la operacion de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-07-01

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  14. 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

  15. Compact fusion reactors

    CERN Multimedia

    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.

  16. HOMOGENEOUS NUCLEAR REACTOR

    Science.gov (United States)

    Hammond, R.P.; Busey, H.M.

    1959-02-17

    Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.

  17. AIR COOLED NEUTRONIC REACTOR

    Science.gov (United States)

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  18. NEUTRONIC REACTOR CORE INSTRUMENT

    Science.gov (United States)

    Mims, L.S.

    1961-08-22

    A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

  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. A COOLED NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Creutz, E.C.

    1960-03-15

    A nuclear reactor comprising a pair of graphite blocks separated by an air gap is described. Each of the blocks contains a plurality of channels extending from the gap through the block with a plurality of fuel elements being located in the channels. Means are provided for introducing air into the gap between the graphite blocks and for exhausting the air from the ends of the channels opposite the gap.

  1. Gaseous fuel reactor research

    Science.gov (United States)

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  2. 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.)

  3. NUCLEAR REACTOR CORE DESIGN

    Science.gov (United States)

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  4. Nuclear reactor sealing system

    Science.gov (United States)

    McEdwards, James A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  5. 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.

  6. 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.

  7. 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)

  8. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

    2013-06-01

    On Friday, March 11, 2011, at 2:46pm (Japan Standard Trme), the Tohoku region on the east coast of northern Japan experi­enced what would become known as the largest earthquake in the country's history at magnitude 9.0 on the Richter scale. The Fukushima Daiichi nuclear power plant suffered exten­sive and irreversible damage. Six operating units were at the site, each with a boiling water reactor. When the earthquake struck, three of the six reactors were operating and the others were in a periodic inspection outage phase. In one reactor, all of the fuel had been relocated to a spent fuel pool in the reactor building. The seismic acceleration caused by the earthquake brought the three operating units to an automatic shutdown. Since there was damage to the power transmission lines, the emergency diesel generators (EDG) were automat­ically started to ensure continued cooling of the reactors and spent fuel pools. The situation was under control until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 meters, which was three times taller than the sea wall of 5m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to five of the six reactors. The flooding also resulted in the loss of instrumentation that would have other­ wise been used to monitor and control the emergency. The ugly aftermath included high radiation exposure to operators at the nuclear power plants and early contamina­tion of food supplies and water within several restricted areas in Japan, where high radiation levels have rendered them un­safe for human habitation. While the rest of the story will remain a tragic history, it is this part of the series of unfortunate events that has inspired our research. It has indubitably highlighted the need for a novel sensor and instrumentation system that can withstand similar or worse conditions to avoid future catastrophe and assume damage

  9. 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.

  10. TMI-2 reactor vessel head removal

    Energy Technology Data Exchange (ETDEWEB)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

  11. TMI-2 reactor vessel head removal

    Energy Technology Data Exchange (ETDEWEB)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities.

  12. Simultaneous optical/X-ray study of GS 1354-64 (=BW Cir) during hard outburst: evidence for optical cyclo-synchrotron emission from the hot accretion flow

    Science.gov (United States)

    Pahari, Mayukh; Gandhi, Poshak; Charles, Philip A.; Kotze, Marissa M.; Altamirano, Diego; Misra, Ranjeev

    2017-07-01

    We present results from simultaneous optical [South African Large Telescope (SALT)] and X-ray (Swift and INTEGRAL) observations of GS 1354-64/BW Cir during the 2015 hard state outburst. During the rising phase, optical/X-ray time series shows a strong anti-correlation with X-ray photons lagging optical. Optical and X-ray power spectra show quasi-periodic oscillations (QPOs) at a frequency of ˜18 mHz with a confidence level of at least 99 per cent. Simultaneous fitting of Swift/XRT and INTEGRAL spectra in the range 0.5-1000.0 keV shows non-thermal, power-law-dominated (>90 per cent) spectra with a hard power-law index of 1.48 ± 0.03, inner disc temperature of 0.12 ± 0.01 keV and an inner disc radius of ˜3000 km. All evidence is consistent with cyclo-synchrotron radiation in a non-thermal, hot electron cloud extending to ˜100 Schwarzschild radii being a major physical process for the origin of optical photons. At outburst peak about one month later, when the X-ray flux rises and the optical drops, the apparent features in the optical/X-ray correlation vanish and the optical auto correlation widens. Although ˜0.19 Hz QPO is observed from the X-ray power spectra, the optical variability is dominated by the broad-band noise, and the inner disc temperature increases. These results support a change in the dominant optical emission source between outburst rise and peak, consistent with a weakening of hot flow as the disc moves in.

  13. TRATAMIENTO DE AGUAS RESIDUALES DE UNA INDUSTRIA PROCESADORA DE PESCADO EN REACTORES ANAERÓBICOS DISCONTINUOS/WASTEWATER TREATMENT OF A FISH PROCESSING INDUSTRY IN BATCH ANAEROBIC REACTORS

    National Research Council Canada - National Science Library

    Julio César Marín Leal; Carlos Aníbal Chinga Panta; Abrahan Isaac Velásquez Ferrín; Pierre Andrés González Cabo; Luz María Zambrano Rodríguez

    2015-01-01

      In this paper, the treatment of wastewater from a fish processing industry (Manta, Ecuador) in batch anaerobic reactors was evaluated and its adjustment to current environmental standards for discharge was established...

  14. 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)

  15. 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.

  16. 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.

  17. 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.

  18. 75 FR 1831 - Seeks Qualified Candidates for the Advisory Committee on Reactor Safeguards

    Science.gov (United States)

    2010-01-13

    ... on the adequacy of proposed reactor safety standards. Of primary importance are the safety issues... initiatives, including risk-informed and performance-based regulations, license renewal, power uprates, and... multinational design certifications, material and structural engineering, nuclear analysis and reactor core...

  19. Standards not that standard.

    Science.gov (United States)

    Vilanova, Cristina; Tanner, Kristie; Dorado-Morales, Pedro; Villaescusa, Paula; Chugani, Divya; Frías, Alba; Segredo, Ernesto; Molero, Xavier; Fritschi, Marco; Morales, Lucas; Ramón, Daniel; Peña, Carlos; Peretó, Juli; Porcar, Manuel

    2015-01-01

    There is a general assent on the key role of standards in Synthetic Biology. In two consecutive letters to this journal, suggestions on the assembly methods for the Registry of standard biological parts have been described. We fully agree with those authors on the need of a more flexible building strategy and we highlight in the present work two major functional challenges standardization efforts have to deal with: the need of both universal and orthogonal behaviors. We provide experimental data that clearly indicate that such engineering requirements should not be taken for granted in Synthetic Biology.

  20. Accelerator based fusion reactor

    Science.gov (United States)

    Liu, Keh-Fei; Chao, Alexander Wu

    2017-08-01

    A feasibility study of fusion reactors based on accelerators is carried out. We consider a novel scheme where a beam from the accelerator hits the target plasma on the resonance of the fusion reaction and establish characteristic criteria for a workable reactor. We consider the reactions d+t\\to n+α,d+{{}3}{{H}\\text{e}}\\to p+α , and p+{{}11}B\\to 3α in this study. The critical temperature of the plasma is determined from overcoming the stopping power of the beam with the fusion energy gain. The needed plasma lifetime is determined from the width of the resonance, the beam velocity and the plasma density. We estimate the critical beam flux by balancing the energy of fusion production against the plasma thermo-energy and the loss due to stopping power for the case of an inert plasma. The product of critical flux and plasma lifetime is independent of plasma density and has a weak dependence on temperature. Even though the critical temperatures for these reactions are lower than those for the thermonuclear reactors, the critical flux is in the range of {{10}22}-{{10}24}~\\text{c}{{\\text{m}}-2}~{{\\text{s}}-1} for the plasma density {ρt}={{10}15}~\\text{c}{{\\text{m}}-3} in the case of an inert plasma. Several approaches to control the growth of the two-stream instability are discussed. We have also considered several scenarios for practical implementation which will require further studies. Finally, we consider the case where the injected beam at the resonance energy maintains the plasma temperature and prolongs its lifetime to reach a steady state. The equations for power balance and particle number conservation are given for this case.

  1. Nuclear Reactor Safety; (USA)

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

    This publication announces on an monthly basis the current worldwide information available on all safety-related aspects of reactors, including: accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures. 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 other US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Technology Data Exchange, the International Atomic Energy Agency's International Nuclear Information System, or government-to-government agreements.

  2. Iris reactor development

    Energy Technology Data Exchange (ETDEWEB)

    Paramonov, D.V.; Carelli, M.D. [Westinghouse Electric Corp., Baltimore, MD (United States); Miller, K. [BNFL, Inc., (United Kingdom); Lombardi, C.V.; Ricotti, M.E. [Polytechnic of Milan, Polimi (Italy); Todreas, N.E. [Masachussets Institute of Technology, MIT (United States); Greenspan, E. [University of California at Berkeley, UCB (United States); Yamamoto, K. [JAPC Japan Atomic Power Co., Tokyo (Japan); Nagano, A. [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Ninokata, H. [Tokyo Institut of Technology, TIT (Japan); Robertson, J. [Westinghouse and bechtel (United States); Oriolo, F. [Pisa Univ. (Italy)

    2001-07-01

    The development progress of the IRIS (International Reactor Innovative and Secure) nuclear power system is presented. IRIS is currently being developed by an international consortium of industry, laboratory, university and utility establishments, led by Westinghouse. It is aimed at achieving the four major objectives of the Generation IV nuclear systems, i.e., proliferation resistance, enhanced safety, economic competitiveness and reduced waste. The project first year activities, which are summarized here, were focused on core neutronics, in-vessel configuration, steam generator and containment design, safety approach and economic performance. Details of these studies are provided in parallel papers in these proceedings. (author)

  3. NEUTRONIC REACTOR CONSTRUCTION

    Science.gov (United States)

    Vernon, H.C.; Goett, J.J.

    1958-09-01

    A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

  4. Reactor coolant pump flywheel

    Science.gov (United States)

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  5. SODIUM DEUTERIUM REACTOR

    Science.gov (United States)

    Oppenheimer, E.D.; Weisberg, R.A.

    1963-02-26

    This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

  6. 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.

  7. Fast Reactor Fuel Type and Reactor Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  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. Decommissioning of commercial reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yui, Kohei [Japan Atomic Power Co., Tokyo (Japan)

    1997-02-01

    In the case of nuclear reactors, the diversion is often difficult as they are highly purposive, the disassembling is not easy as they are robust, and attention is required to handle the equipment containing radioactive substances. Decommissioning is defined as all the measures taken from the state that facilities become unused to the state of becoming green field. In Japan, already 40 years have elapsed since the effort for nuclear power was begun, and in this paper, the present state and future subjects of the decommissioning of nuclear power stations are summarized at the opportunity that the stop of commercial operation of Tokai Nuclear Power Station was decided recently. In the Tokai Nuclear Power Station, 166 MWe graphite-moderated, carbon dioxide-cooled reactor called improved Calder Hall type is installed, which started the operation in 1966. The circumstances of the decision to stop its operation are explained. The basic policy of the decommissioning of commercial nuclear power stations has been already published by the Advisory Committee for Energy. The state of the decommissioning in various foreign countries is reported. In Japan, the state of green field was realized in 1996 in the decommissioning of the JPDR in Japan Atomic Energy Research institute, and the decommissioning of the atomic powered ship ``Mutsu`` was completed. (K.I.)

  11. Survey of mirror machine reactors

    Energy Technology Data Exchange (ETDEWEB)

    Condit, W.C.

    1978-08-11

    The Magnetic Mirror Fusion Program is one of the two main-line fusion efforts in the United States. Starting from the simple axisymmetric mirror concept in the 1950's, the program has successfully overcome gross flute-type instabilities (using minimum-B magnetic fields), and the most serious of the micro-instabilities which plagued it (the drift-cyclotron loss-cone mode). Dense plasmas approaching the temperature range of interest for fusion have been created (n/sub p/ = 10/sup 14//cc at 10 to 12 keV). At the same time, rather extensive conceptual design studies of possible mirror configurations have led to three principle designs of interest: the standard mirror fission-fusion hybrid, tandem mirror, and the field-reversed mirror. The lectures will discuss these three concepts in turn. There will be no discussion of diagnostics for the mirror machine in these lectures, but typical plasma parameters will be given for each type of machine, and the diagnostic requirements will be apparent. In a working fusion reactor, diagnostics will be required for operational control, and remarks will be made on this subject.

  12. 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.

  13. Solvent refined coal reactor quench system

    Science.gov (United States)

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  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. Rotary reactor and use thereof

    NARCIS (Netherlands)

    Bakker Wridzer, J.W.; Kapteijn, F.; Moulijn, J.A.

    1998-01-01

    The invention relates to a rotary reactor consisting of a number of tubular reaction compartments (A), each provided with a first end and a second end, a ceramic first reactor end plate (B) in which said first ends are received, and a second end plate (B) in which said second ends are received,

  16. 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.

  17. Multilevel transport solution of LWR reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Cassiano R.E. de Oliveira; HyeonKae Park

    2008-09-01

    This work presents a multilevel approach for the solution of the transport equation in typical LWR assemblies and core configurations. It is based on the second-order, even-parity formulation of the transport equation, which is solved within the framework provided by the finite element-spherical harmonics code EVENT. The performance of the new solver has been compared with that of the standard conjugate gradient solver for diffusion and transport problems on structured and unstruc-tured grids. Numerical results demonstrate the potential of the multilevel scheme for realistic reactor calculations.

  18. Radiation Shielding for Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Santoro, R.T.

    1999-10-01

    Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor

  19. 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)

  20. Core design analysis of the supercritical water fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mori, M.

    2005-10-01

    Light Water Reactor technology is nowadays the most successful commercial application of fission reactors for the production of electricity. However, in the next years, nuclear industry will have to face new and demanding challenges. The need for sustainable and cheap sources of energy, the need for public acceptance, the need for even higher safety standards, the need to minimize waste production are only a few examples. It is for these very reasons that a few next generation nuclear reactor concepts were selected for extensive research and development. Super critical water cooled reactors are one of them. The use of a supercritical coolant would in fact allow for higher thermal efficiencies and a more compact plant design. As a matter of fact, steam generators, or steam separators and driers would not be needed thus, significantly reducing construction costs. Moreover, because of the high heat capacity of supercritical water, comparatively less coolant would be needed to refrigerate the reactor. Consequently, a water-cooled reactor with a fast neutron spectrum could potentially be designed: the SuperCritical water Fast Reactor. This system presents unique features combining well-known fast and light water reactor characteristics in one design (e.g. the tendency to a positive void reactivity coefficient together with Loss Of Coolant Accidents, as design basis). The core is in fact loaded with highly enriched Mixed Oxide fuel (average plutonium content of {approx}23%), and presents a peculiar and significant geometrical and material heterogeneity (use of radial and axial blankets, solid moderator layers, several enrichment zones). The safety analysis of this very complex core layout, the development of suitable tools of investigation, and the optimization of the void reactivity effect through core design, is the main objective of this work. (orig.)

  1. HIV infection abrogates the functional advantage of natural killer cells educated through KIR3DL1/HLA-Bw4 interactions to mediate anti-HIV antibody-dependent cellular cytotoxicity.

    Science.gov (United States)

    Parsons, Matthew S; Wren, Leia; Isitman, Gamze; Navis, Marjon; Stratov, Ivan; Bernard, Nicole F; Kent, Stephen J

    2012-04-01

    Combinations of KIR3DL1 and HLA-Bw4 alleles protect against HIV infection and/or disease progression. These combinations enhance NK cell responsiveness through the ontological process of education. However, educated KIR3DL1(+) NK cells do not have enhanced degranulation upon direct recognition of autologous HIV-infected cells. Since antibody-dependent cellular cytotoxicity (ADCC) is associated with improved HIV infection outcomes and NK cells overcome inhibition through killer cell immunoglobulin-like receptors (KIR) to mediate ADCC, we hypothesized that KIR3DL1-educated NK cells mediate anti-HIV ADCC against autologous cells. A whole-blood flow cytometry assay was used to evaluate ADCC-induced activation of NK cells. This assay assessed activation (gamma interferon [IFN-γ] production and/or CD107a expression) of KIR3DL1(+) and KIR3DL1(-) NK cells, from HLA-Bw4(+) and HLA-Bw4(-) HIV-positive and HIV-negative individuals, in response to autologous HIV-specific ADCC targets. KIR3DL1(+) NK cells were more functional than KIR3DL1(-) NK cells from HLA-Bw4(+), but not HLA-Bw4(-), healthy controls. In HIV-infected individuals, no differences in NK cell functionality were observed between KIR3DL1(+) and KIR3DL1(-) NK cells in HLA-Bw4(+) individuals, consistent with dysfunction of NK cells in the setting of HIV infection. Reflecting the partial normalization of NK cell responsiveness following initiation of antiretroviral therapy, a significant correlation was observed between the peripheral CD4(+) T-lymphocyte counts in antiretroviral therapy-treated subjects and the functionality of NK cells. However, peripheral CD4(+) T-lymphocyte counts were not correlated with an anti-HIV ADCC functional advantage in educated KIR3DL1(+) NK cells. The abrogation of the functional advantage of educated NK cells may enhance HIV disease progression. Strategies to enhance the potency of NK cell-mediated ADCC may improve HIV therapies and vaccines.

  2. 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.

  3. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2003-04-01

    The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

  4. 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

  5. Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1998-01-01

    This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

  6. 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

  7. Nuclear reactor downcomer flow deflector

    Science.gov (United States)

    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.

  8. Reactor neutrons in nuclear astrophysics

    Science.gov (United States)

    Reifarth, René; Glorius, Jan; Göbel, Kathrin; Heftrich, Tanja; Jentschel, Michael; Jurado, Beatriz; Käppeler, Franz; Köster, Ulli; Langer, Christoph; Litvinov, Yuri A.; Weigand, Mario

    2017-09-01

    The huge neutron fluxes offer the possibility to use research reactors to produce isotopes of interest, which can be investigated afterwards. An example is the half-lives of long-lived isotopes like 129I. A direct usage of reactor neutrons in the astrophysical energy regime is only possible, if the corresponding ions are not at rest in the laboratory frame. The combination of an ion storage ring with a reactor and a neutron guide could open the path to direct measurements of neutron-induced cross sections on short-lived radioactive isotopes in the astrophysically interesting energy regime.

  9. 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

  10. 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.

  11. 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)

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

    An outline is given of containment concepts, sources and release rates of energy, responses of containment structures, effects of projectiles, and leakage rates of radioisotopes, with particular regard to major reactor accidents. (T.F.H.)

  17. Actinide transmutation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ganev, I.K.; Lopatkin, A.V.; Naumov, V.V.; Tocheny, L.V.

    1993-12-31

    Of some interest is the comparison between the actinide nuclide burning up (fission) rates such as americium 241, americium 242, curium 244, and neptunium 237, in the reactors with fast or thermal neutron spectra.

  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. Combustion synthesis continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maupin, G.D.; Chick, L.A.; Kurosky, R.P.

    1998-01-06

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor. 10 figs.

  20. Combustion synthesis continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maupin, Gary D. (Richland, WA); Chick, Lawrence A. (West Richland, WA); Kurosky, Randal P. (Maple Valley, WA)

    1998-01-01

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor.

  1. 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...

  2. 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.

  3. Engineering reactors for catalytic reactions

    Indian Academy of Sciences (India)

    Chemical Engineering and Process Development Division, CSIR - National Chemical Laboratory,. Pune 411 008, India ... Abstract. Catalytic reactions are ubiquitous in chemical and allied industries. ... strategies and recent advances in process intensification/ multifunctional reactors are discussed to illustrate the approach.

  4. 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)

  5. 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

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

    Science.gov (United States)

    Rohanda, Anis; Waris, Abdul

    2015-04-01

    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 16O(n,p)16N 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.

  7. Generation III reactors safety requirements and the design solutions

    Science.gov (United States)

    Felten, P.

    2009-03-01

    Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called "generation III" safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

  8. Background studies for the MINER Coherent Neutrino Scattering reactor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Agnolet, G.; Baker, W. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Barker, D. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Beck, R. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Carroll, T.J.; Cesar, J. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Cushman, P. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Dent, J.B. [Department of Physics, University of Louisiana at Lafayette, Lafayette, LA 70504 (United States); De Rijck, S. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Dutta, B. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Flanagan, W. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Fritts, M. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Gao, Y. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Department of Physics & Astronomy, Wayne State University, Detroit 48201 (United States); Harris, H.R.; Hays, C.C. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Iyer, V. [School of Physical Sciences, National Institute of Science Education and Research, Jatni - 752050 (India); and others

    2017-05-01

    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 m) 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 5–20 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.

  9. 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.

  10. Reactor antineutrinos and nuclear physics

    Energy Technology Data Exchange (ETDEWEB)

    Balantekin, A.B. [University of Wisconsin, Department of Physics, Madison, WI (United States)

    2016-11-15

    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. (orig.)

  11. 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.

  12. 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

  13. Propellant actuated nuclear reactor steam depressurization valve

    Science.gov (United States)

    Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

    1992-01-01

    A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

  14. Impact of amino acid and CP restriction from 20 to 140 kg BW on performance and dynamics in empty body protein and lipid deposition of entire male, castrated and female pigs.

    Science.gov (United States)

    Ruiz-Ascacibar, I; Stoll, P; Kreuzer, M; Boillat, V; Spring, P; Bee, G

    2017-03-01

    Breeding leaner pigs during the last decades may have changed pig's empty body (EB) composition, a key trait for elaborating feeding recommendations. This research aimed to provide new experimental data on changes in the chemical composition of the EB of pigs from 20 to 140 kg BW. In addition, the impact of a reduction in the dietary CP associated with lower lysine, methionine+cystine, threonine and tryptophan levels was determined. In total, 48 males, castrates and females weighing 20 kg BW were allocated either to a control grower-finisher diet formulated according to current Swiss feeding recommendations, or a low CP grower-finisher diet (80% of control). Feed intake was monitored and pigs were weighed weekly. The chemical composition of EB (blood, hairs and hoofs, offals, bile, carcass) was determined at 20, 40, 60, 80, 100, 120 and 140 kg BW on four pigs per gender and diet (eight pigs per gender at 20 kg). The five fractions were weighed and samples were analysed for dry matter, protein, fat and energy. Nutrient deposition rates and N efficiency were calculated by using the 20 kg BW category as reference. Analysis revealed an accurate feed optimisation for the aforementioned essential amino acids (EAA), whereas digestible isoleucine content in the low CP diet was at 70% of the control diet. Despite similar feed intake, daily gain and feed efficiency were impaired (PCP compared with the control pigs. In the same growth period, castrates had the greatest feed intake but, together with females, displayed the lowest (PCP diet and was lower (PCP diet and castrates. N efficiency improved (PCP diet from 100 to 140 kg. The males displayed the greatest (PCP content of finisher II diets could be reduced to 102, 102 and 104 g/kg for females, castrates and males, respectively, without a negative impact on protein deposition or growth. It remains unclear whether the negative effects found in the BW range from 20 to 100 kg on the EB deposition were due to the 20% reduction

  15. Reactor Simulator Testing Overview

    Science.gov (United States)

    Schoenfeld, Michael P.

    2013-01-01

    Test Objectives Summary: a) Verify operation of the core simulator, the instrumentation & control system, and the ground support gas and vacuum test equipment. b) Examine cooling & heat regeneration performance of the cold trap purification. c) Test the ALIP pump at voltages beyond 120V to see if the targeted mass flow rate of 1.75 kg/s can be obtained in the RxSim. Testing Highlights: a) Gas and vacuum ground support test equipment performed effectively for operations (NaK fill, loop pressurization, and NaK drain). b) Instrumentation & Control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings and ramped within prescribed constraints. It effectively interacted with reactor simulator control model and defaulted back to temperature control mode if the transient fluctuations didn't dampen. c) 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 minimum temperature indicating the design provided some heat regeneration. d) ALIP produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  16. 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.

  17. 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.

  18. REACTOR CONTROL MECHANISM

    Science.gov (United States)

    Lane, J.A.; Engberg, R.E.; Welch, J.M.

    1959-05-12

    A quick-releasing mechanism is described which may be used to rapidiy drop a device supported from beneath during normal use, such as a safety rod in a nuclear reactor. In accordance with this invention an electrical control signal, such as may be provided by radiation detection or other alarm condition sensing devices, is delivered to an electromagnetic solenoid, the armature of which is coupled to an actuating mechanism. The solenoid is energized when the mechanism is in its upper or cocked position. In such position, the mechanism engages a plurality of retaining balls, forcing them outward into engagement with a shoulder or recess in a corresponding section of a tubular extension on the upheld device. When the control signal to the solenoid suddenly ceases, the armature drops out, allowing the actuating mechanism to move slightly but rapidly under the force of a compressed spring. The weight of the device will urge the balls inward against a beveled portion of the actuating mechanism and away from the engaging section on the tubular extension, thus allowing the upheld device to fall freely under the influence of gravity.

  19. 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

  20. 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.

  1. Effect of DUPIC Cycle on CANDU Reactor Safety Parameters

    Directory of Open Access Journals (Sweden)

    Nader M.A. Mohamed

    2016-10-01

    Full Text Available 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 UO2 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.

  2. Modeling residence-time distribution in horizontal screw hydrolysis reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sievers, David A.; Stickel, Jonathan J.

    2018-01-01

    The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. This study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters. Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.

  3. 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.

  4. 78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

    Science.gov (United States)

    2013-11-29

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence... Regulatory Commission. Michele G. Evans, Director, Division of Operating Reactor Licensing, Office of Nuclear...

  5. 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)

  6. Entropy Production in Chemical Reactors

    Science.gov (United States)

    Kingston, Diego; Razzitte, Adrián C.

    2017-06-01

    We have analyzed entropy production in chemically reacting systems and extended previous results to the two limiting cases of ideal reactors, namely continuous stirred tank reactor (CSTR) and plug flow reactor (PFR). We have found upper and lower bounds for the entropy production in isothermal systems and given expressions for non-isothermal operation and analyzed the influence of pressure and temperature in entropy generation minimization in reactors with a fixed volume and production. We also give a graphical picture of entropy production in chemical reactions subject to constant volume, which allows us to easily assess different options. We show that by dividing a reactor into two smaller ones, operating at different temperatures, the entropy production is lowered, going as near as 48 % less in the case of a CSTR and PFR in series, and reaching 58 % with two CSTR. Finally, we study the optimal pressure and temperature for a single isothermal PFR, taking into account the irreversibility introduced by a compressor and a heat exchanger, decreasing the entropy generation by as much as 30 %.

  7. Antineutrino monitoring of thorium reactors

    Science.gov (United States)

    Akindele, Oluwatomi A.; Bernstein, Adam; Norman, Eric B.

    2016-09-01

    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 fuel 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 breed 233U, that if diverted constitute a direct use material as defined by the International Atomic Energy Agency (IAEA). The antineutrino spectrum from the fission of 233U has been estimated for the first time, and the feasibility of detecting the diversion of 8 kg of 233U, within a 30 day timeliness goal has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos at a 25 m standoff. It was found that the diversion of a significant quantity of 233U could not be detected within the current IAEA timeliness detection goal using either tests. A rate-time based analysis exceeded the timeliness goal by 23 days, while a spectral based analysis exceeds this goal by 31 days.

  8. Simplifying Microbial Electrosynthesis Reactor Design

    Directory of Open Access Journals (Sweden)

    Cloelle G.S. Giddings

    2015-05-01

    Full Text Available Microbial electrosynthesis, an artificial form of photosynthesis, can efficiently convert carbon dioxide into organic commodities; however, this process has only previously been demonstrated in reactors that have features likely to be a barrier to scale-up. Therefore, the possibility of simplifying reactor design by both eliminating potentiostatic control of the cathode and removing the membrane separating the anode and cathode was investigated with biofilms of Sporomusa ovata, which reduces carbon dioxide to acetate. In traditional ‘H-cell’ reactors, where the anode and cathode chambers were separated with a proton-selective membrane, the rates and columbic efficiencies of microbial electrosynthesis remained high when electron delivery at the cathode was powered with a direct current power source rather than with a poteniostat-poised cathode utilized in previous studies. A membrane-less reactor with a direct-current power source with the cathode and anode positioned to avoid oxygen exposure at the cathode, retained high rates of acetate production as well as high columbic and energetic efficiencies. The finding that microbial electrosynthesis is feasible without a membrane separating the anode from the cathode, coupled with a direct current power source supplying the energy for electron delivery, is expected to greatly simplify future reactor design and lower construction costs.

  9. Rapid starting methanol reactor system

    Science.gov (United States)

    Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.

    1984-01-01

    The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

  10. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  11. 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.

  12. ASME Material Challenges for Advanced Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Ali Siahpush

    2013-07-01

    This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

  13. 10 CFR 170.21 - Schedule of fees for production and utilization facilities, review of standard referenced design...

    Science.gov (United States)

    2010-01-01

    .... Amendment, Renewal, Other Approvals Full cost. C. Test Facility/Research Reactor/Critical Facility..., approvals of facility standard reference designs, re-qualification and replacement examinations for reactor... of fees Fees 1,2 A. Nuclear Power Reactors Application for Construction Permit Full cost. Early Site...

  14. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  15. 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.

  16. KIR3DS1/HLA-B Bw4-80Ile Genotype Is Correlated with the IFN-α Therapy Response in hepatitis B e antigen-Positive Chronic Hepatitis B

    Directory of Open Access Journals (Sweden)

    Wenting Li

    2017-10-01

    Full Text Available To date, several on-treatment-level virological and serological indices that may predict the response to interferon alpha (IFN-α have been reported. However, no effective predictors, such as drug–response genes, that can be detected before administration of anti-hepatitis B virus (HBV therapy with IFN-α, have been found. In the diverse range of chronic viral infection, genes that affect human immunity play important roles in understanding host and viral co-evolution. Killer-cell immunoglobulin-like receptors (KIRs, which are highly polymorphic at the allele and haplotype levels, participate in the antiviral function of natural killer (NK cells via fine-tuning inhibition and activation of NK-cell responses that occur when the NK cells interact with human leukocyte antigen (HLA class I molecules on target cells. For each individual, the pairing of KIR and HLA ligand is genetically determined. To investigate whether a particular KIR and HLA repertoire influences the risk of HBV infection and response to IFN-α treatment for chronic hepatitis B (CHB, we genotyped the KIRs and HLA ligands of 119 hepatitis B e antigen (HBeAg-positive CHB patients. These patients included 43 patients who achieved sustained response (SR induced by IFN-α treatment for 48 weeks, 76 patients who achieved no response (NR, and 96 healthy subjects as controls. SR was defined as HBeAg loss with HBV DNA < 2,000 IU/ml and alanine aminotransferase normalization at 24 weeks posttreatment (week 72. In this study, we showed that activating KIR genes were less prevalent in Han Chinese, especially in Han Chinese with CHB, than in Caucasians. Furthermore, the KIR3DS1 gene, in combination with HLA-B Bw4-80Ile, strongly influenced the therapeutic outcomes for CHB patients who were treated with IFN-α. The frequency of the combination of genes encoding KIR3DS1 and HLA-B Bw4-80Ile was higher in patients who had a sustained treatment response than in patients who had NR [35

  17. 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.)

  18. 3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

    Energy Technology Data Exchange (ETDEWEB)

    Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

    2016-09-15

    The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

  19. STARFIRE: a commercial tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor.

  20. 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

  1. 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.

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

    Science.gov (United States)

    Nelson, Kyle A; Geuther, Jeffrey A; Neihart, James L; Riedel, Todd A; Rojeski, Ronald A; Saddler, Jeffrey L; Schmidt, Aaron J; McGregor, Douglas S

    2012-07-01

    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. Copyright © 2012 Elsevier Ltd. All rights reserved.

  3. 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.

  4. United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

    Energy Technology Data Exchange (ETDEWEB)

    Morrell, Douglas [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2008-10-29

    The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room

  5. 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.

  6. 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...

  7. Reactivity control assembly for nuclear reactor

    Science.gov (United States)

    Bollinger, Lawrence R.

    1984-01-01

    Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

  8. The Westinghouse Small Modular Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez Garcia, F. J.; Memmott, M. J.

    2012-07-01

    Westinghouse has developed a small modular reactor (SMR) which incorporates an integral pressurized water reactor (PWR) design. The Westinghouse SMR design also utilizes many of the key features and innovative concepts from the AP1000 plant, including passive safety systems.

  9. What can Recycling in Thermal Reactors Accomplish?

    Energy Technology Data Exchange (ETDEWEB)

    Steven Piet; Gretchen E. Matthern; Jacob J. Jacobson

    2007-09-01

    Thermal recycle provides several potential benefits when used as stop-gap, mixed, or backup recycling to recycling in fast reactors. These three roles involve a mixture of thermal and fast recycling; fast reactors are required to some degree at some time. Stop-gap uses thermal reactors only until fast reactors are adequately deployed and until any thermal-recycle-only facilities have met their economic lifetime. Mixed uses thermal and fast reactors symbiotically for an extended period of time. Backup uses thermal reactors only if problems later develop in the fast reactor portion of a recycling system. Thermal recycle can also provide benefits when used as pure thermal recycling, with no intention to use fast reactors. However, long term, the pure thermal recycling approach is inadequate to meet several objectives.

  10. How much does the MSW effect contribute to the reactor antineutrino anomaly?

    Science.gov (United States)

    Valdiviesso, G. A.

    2015-05-01

    It has been pointed out that there is a 5.7 ± 2.3 discrepancy between the predicted and the observed reactor antineutrino flux in very short baseline experiments. Several causes for this anomaly have been discussed, including a possible non-standard forth sterile neutrino. In order to quantify how much non-standard this anomaly really is, the standard MSW effect is reviewed. Knowing that reactor antineutrinos are produced in a dense medium (the nuclear fuel) and is usually detected in a less dense one (water, or scintillator), non-adiabatic effects are expected to happen, creating a difference between the creation and detection mixing angles.

  11. ['Gold standard', not 'golden standard'

    NARCIS (Netherlands)

    Claassen, J.A.H.R.

    2005-01-01

    In medical literature, both 'gold standard' and 'golden standard' are employed to describe a reference test used for comparison with a novel method. The term 'gold standard' in its current sense in medical research was coined by Rudd in 1979, in reference to the monetary gold standard. In the same

  12. An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

    NARCIS (Netherlands)

    Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

    2009-01-01

    In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

  13. Operating limits Hanford Production Reactors. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Owsley, G.F. [comp.

    1963-05-20

    This report is applicable to the eight operating production reactors, B, C, D, DR, F, H, KE, and KW. It covers the following: operating parameter limitations; reactivity limitations; control and safety systems; reactor fuel loading; coolant requirements with irradiated fuel in reactor; reactor confinement; test facilities; code compliance; safety instrumentation and set points; and control criteria. Also discussed are administrative procedures for process control, training, audits and inspection, and reports and records.

  14. Digital computer operation of a nuclear reactor

    Science.gov (United States)

    Colley, R.W.

    1982-06-29

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  15. Monolithic reactor: higher yield, less energy

    OpenAIRE

    Mols, B.

    2004-01-01

    The production of margarine, the desulphurisation of crude oil, and the manufacture of synthetic diesel fuel, these are only three of the many industrial processes in which a three-phase reactor is used. Traditionally, this type of reactor is rather ill-defined. Success with a lab scale set-up is no guarantee that a large commercial reactor will work. Scalability is less than perfect, one might say. Researchers at the Reactor & Catalysis Engineering epartment of the Chemical Technology facult...

  16. Reference worldwide model for antineutrinos from reactors

    OpenAIRE

    Baldoncini, Marica; Callegari, Ivan; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2014-01-01

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency (IAEA). We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Bor...

  17. Liquid metal cooled nuclear reactor plant system

    Science.gov (United States)

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  18. 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

  19. 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

  20. Comparative Study on Cyber Securities between Power Reactor and Research Reactor with Bayesian Update

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jinsoo; Heo, Gyunyoung [Kyung Hee University, Yongin (Korea, Republic of); Son, Han Seong [Joongbu Univiersity, Geumsan (Korea, Republic of)

    2016-10-15

    The Stuxnet has shown that nuclear facilities are no more safe from cyber-attack. Due to practical experiences and concerns on increasing of digital system application, cyber security has become the important issue in nuclear industry. Korea Institute of Nuclear Nonproliferation and control (KINAC) published a regulatory standard (KINAC/RS-015) to establish cyber security framework for nuclear facilities. However, it is difficult to research about cyber security. It is hard to quantify cyber-attack which has malicious activity which is different from existing design basis accidents (DBAs). We previously proposed a methodology on development of a cyber security risk model with BBN. However, the methodology had a limitation in which the input data as prior information was solely on expert opinions. In this study, we propose a cyber security risk model for instrumentation and control (I and C) system of nuclear facilities with some equation for quantification by using Bayesian Belief Network (BBN) in order to overcome the limitation of previous research. The proposed model has been used for comparative study on cyber securities between large-sized nuclear power plants (NPPs) and small-sized Research Reactors (RR). In this study, we proposed the cyber security risk evaluation model with BBN. It includes I and C architecture, which is a target system of cyber-attack, malicious activity, which causes cyber-attack from attacker, and mitigation measure, which mitigates the cyber-attack risk. Likelihood and consequence as prior information are evaluated by considering characteristics of I and C architecture and malicious activity. The BBN model provides posterior information with Bayesian update by adding any of assumed cyber-attack scenarios as evidence. Cyber security risk for nuclear facilities is analyzed by comparing between prior information and posterior information of each node. In this study, we conducted comparative study on cyber securities between power reactor

  1. Hydrodynamics of a Monolithic Stirrer Reactor

    NARCIS (Netherlands)

    Kritzinger, H.P.

    2011-01-01

    The Monolithic Stirrer Reactor (MSR) is a novel concept for heterogeneously catalyzed reactors and is presented as an alternative device to slurry reactors. It uses a modified stirrer on which structured catalyst supports (monoliths) are fixed to form permeable blades. The monoliths consist of small

  2. 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...

  3. Laminar Entrained Flow Reactor (Fact Sheet)

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    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. New usage for old reactor

    NARCIS (Netherlands)

    Wassink, J.

    2015-01-01

    The latest measurement instrument of the TU Delft measures the crystal structures of many different materials and is unique within the Netherlands. The so-called Pearl neutron powder diffractometer was opened on 24 September at the RID reactor institute. “It is difficult to overestimate the

  5. 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.

  6. REACTOR CONTROL ROD OPERATING SYSTEM

    Science.gov (United States)

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  7. STEAM STIRRED HOMOGENEOUS NUCLEAR REACTOR

    Science.gov (United States)

    Busey, H.M.

    1958-06-01

    A homogeneous nuclear reactor utilizing a selfcirculating liquid fuel is described. The reactor vessel is in the form of a vertically disposed tubular member having the lower end closed by the tube walls and the upper end closed by a removal fianged assembly. A spherical reaction shell is located in the lower end of the vessel and spaced from the inside walls. The reaction shell is perforated on its lower surface and is provided with a bundle of small-diameter tubes extending vertically upward from its top central portion. The reactor vessel is surrounded in the region of the reaction shell by a neutron reflector. The liquid fuel, which may be a solution of enriched uranyl sulfate in ordinary or heavy water, is mainiained at a level within the reactor vessel of approximately the top of the tubes. The heat of the reaction which is created in the critical region within the spherical reaction shell forms steam bubbles which more upwardly through the tubes. The upward movement of these bubbles results in the forcing of the liquid fuel out of the top of these tubes, from where the fuel passes downwardly in the space between the tubes and the vessel wall where it is cooled by heat exchangers. The fuel then re-enters the critical region in the reaction shell through the perforations in the bottom. The upper portion of the reactor vessel is provided with baffles to prevent the liquid fuel from splashing into this region which is also provided with a recombiner apparatus for recombining the radiolytically dissociated moderator vapor and a control means.

  8. 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.

  9. 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

  10. 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.

  11. Shutdown system for a nuclear reactor

    Science.gov (United States)

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

    1984-06-05

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  12. Fast-acting nuclear reactor control device

    Science.gov (United States)

    Kotlyar, Oleg M.; West, Phillip B.

    1993-01-01

    A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

  13. Tandem Mirror Reactor Systems Code (Version I)

    Energy Technology Data Exchange (ETDEWEB)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  14. 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.

  15. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  16. Savannah River Site production reactor technical specifications. K Production Reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

  17. Nuclear reactor vessel fuel thermal insulating barrier

    Science.gov (United States)

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  18. Video Coding Technique using MPEG Compression Standards

    Directory of Open Access Journals (Sweden)

    A. J. Falade

    2013-06-01

    Full Text Available Digital video compression technologies have become part of life, in the way visual information is created, communicated and consumed. Some application areas of video compression focused on the problem of optimizing storage space and transmission bandwidth (BW. The two dimensional discrete cosine transform (2-D DCT is an integral part of video and image compression, which is used in Moving Picture Expert Group (MPEG encoding standards. Thus, several video compression algorithms had been developed to reduce the data quantity and provide the acceptable quality standard. In the proposed study, the Matlab Simulink Model (MSM has been used for video coding/compression. The approach is more modern and reduces error resilience image distortion.

  19. Aerial Photography and Imagery, Ortho-Corrected, April 2012, color and b/w and NIR, tiff and MrSID, section tiles or countywide mosaic- plan to refly in 2017 at same resolution (6" pixel), Published in 2012, 1:1200 (1in=100ft) scale, Dodge County Government.

    Data.gov (United States)

    NSGIC Local Govt | GIS Inventory — Aerial Photography and Imagery, Ortho-Corrected dataset current as of 2012. April 2012, color and b/w and NIR, tiff and MrSID, section tiles or countywide mosaic-...

  20. Neutronics of a mixed-flow gas-core reactor

    Energy Technology Data Exchange (ETDEWEB)

    Soran, P.D.; Hansen, G.E.

    1977-11-01

    The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF/sub 6/ (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation.

  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. 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

  3. The PRISM reactor as a possible option to deal with the british civilian plutonium stockpile

    Energy Technology Data Exchange (ETDEWEB)

    Fichtlscherer, Christopher [IANUS, TU Darmstadt (Germany); Friess, Friederike [IANUS, TU Darmstadt (Germany); ISR, Universitaet fuer Bodenkultur Wien (Boku) (Austria)

    2017-07-01

    Dealing with stocks of separated weapon-usable plutonium is a big challenge for our modern society. This work focuses on the British civil plutonium stockpiles, which amount to 103.3 tons. One option is seen in irradiating the plutonium in a fast reactor under development, namely the GE PRISM reactor. The PRISM reactor is a small modular, fast reactor which has a thermal power of 840 MW and an electrical output of 311 MW. It is intended to use MOX fuel and proponents claim, that it thus would be possible to produce clean energy, while making the plutonium proliferation resistant. A MCNP model of the reactor is built and depletion calculations with different target burnups of the fuel were conducted to check whether the burned material would fulfil the Spent-Fuel Standard. Particularly it was checked whether the spent fuel is self protecting, meaning that the dose rate does not fall below a limit of 1 Sv/h in 1 meter distance after a cooling period of 30 years. Based on the reactor model calculations the irradiation time to fulfill this limit for the spent fuel is calculated. Based on the needed target burnup, it can be verified, whether it is possible for the PRISM reactor to render the civil plutonium proliferation resistant in only 20 years as is is claimed by its proponents.

  4. Automatically scramming nuclear reactor system

    Science.gov (United States)

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.

    2004-10-12

    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

  5. HEAVY WATER MODERATED NEUTRONIC REACTOR

    Science.gov (United States)

    Szilard, L.

    1958-04-29

    A nuclear reactor of the type which utilizes uranium fuel elements and a liquid coolant is described. The fuel elements are in the form of elongated tubes and are disposed within outer tubes extending through a tank containing heavy water, which acts as a moderator. The ends of the fuel tubes are connected by inlet and discharge headers, and liquid bismuth is circulated between the headers and through the fuel tubes for cooling. Helium is circulated through the annular space between the outer tubes in the tank and the fuel tubes to cool the water moderator to prevent boiling. The fuel tubes are covered with a steel lining, and suitable control means, heat exchange means, and pumping means for the coolants are provided to complete the reactor assembly.

  6. 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.

  7. CONTROL MEANS FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Tonks, L.

    1962-08-01

    A control device surrounding the active portion of a nuclear reactor is described. The control device consists of a plurality of contiguous cylinders partly filled with a neutron absorbing material and partly filled with a neutron reflecting material, each cylinder having a longitudinal reentrant surface into which a portion of an adjacent cylinder extends, one of the cylinders having two re-entrant surfaces, and means for rotating the cylinders one at a time. (AEC)

  8. APPARATUS FOR CONTROLLING NEUTRONIC REACTORS

    Science.gov (United States)

    Dietrich, J.R.; Harrer, J.M.

    1958-09-16

    A device is described for rapidly cortrolling the reactivity of an active portion of a reactor. The inveniion consists of coaxially disposed members each having circumferenital sections of material having dlfferent neutron absorbing characteristics and means fur moving the members rotatably and translatably relative to each other within the active portion to vary the neutron flux therein. The angular and translational movements of any member change the neutron flux shadowing effect of that member upon the other member.

  9. Reactor safeguards against insider sabotage

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, H.A.

    1982-03-01

    A conceptual safeguards system is structured to show how both reactor operations and physical protection resources could be integrated to prevent release of radioactive material caused by insider sabotage. Operational recovery capabilities are addressed from the viewpoint of both detection of and response to disabled components. Physical protection capabilities for preventing insider sabotage through the application of work rules are analyzed. Recommendations for further development of safeguards system structures, operational recovery, and sabotage prevention are suggested.

  10. 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

  11. 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

  12. 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.

  13. 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. Copyright © 2011 Elsevier Ltd. All rights reserved.

  14. Nuclear reactor power monitoring device

    Energy Technology Data Exchange (ETDEWEB)

    Tarumi, Teruji; Oda, Naotaka; Goto, Yasushi; Ito, Toshiaki [Toshiba Corp., Kawasaki, Kanagawa (Japan); Mitsubori, Minehisa

    1997-07-11

    The present invention provides a nuclear power monitoring device which does not lose a safety protection function even upon occurrence of a single failure in an APRM system of a BWR type reactor. Namely, an APRM for inputting signals of local power region monitors (LPRM) has four channels. Each of the channels is constituted so as to be bypassed. With such a constitution, LPRM detector signals can be inputted one by one to each of the four channels of the APRM from each of the LPRM detector assembly. Accordingly, a common channel for LPRM detectors can be eliminated in a small-sized reactor. The number of signals of the LPRM detectors inputted to each of the channels of the APRM is increased in a large-scaled reactor. Since each of the APRM can be bypassed, even if a single failure of one APRM is caused during a predetermined maintenance, the monitoring can be conducted smoothly by bypassing other channel. As a result, a multiple safe-protection function can be ensured. (I.S.)

  15. 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).

  16. 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.

  17. Prospects for Tokamak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  18. Thermal power evaluation of the TRIGA nuclear reactor at CDTN in operations of long duration

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Zacarias; Ferreira, Andrea Vidal; Rezende, Hugo Cesar, E-mail: amir@cdtn.b, E-mail: avf@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2010-07-01

    The standard operations of nuclear research reactor IPR-R1 TRIGA located at CDTN (Belo Horizonte) usually have duration of not more than 8h. However in 2009 two operations for samples irradiations lasted about 12 hours each at a power of 100 kW. These long lasting operations started in the evening and most of them were carried out at night, when there are only small fluctuations in atmosphere temperature. Therefore the conditions were ideal for evaluating the thermal balance of the power dissipated by the reactor core through the forced cooling system. Heat balance is the standard methodology for power calibration of the IPR-R1 reactor. As in any reactor operation, the main operating parameters were monitored and stored by the Data Acquisition System developed for the reactor. These data have been used for the analysis and calculation of the evolution of several neutronic and thermalhydraulic parameters involved in the reactor operation. This paper analyzes the two long lasting operations of the IPR-R1 TRIGA and compares the recorded results for the power dissipated through the primary cooling loop with the results of the power calibration conducted in March 2009. The results corresponded to those of the thermal power calibration within the uncertainty of this methodology, indicating system stability over a period of six months. (author)

  19. A FAST BREEDER REACTOR SPENT FUEL MEASUREMENTS PROGRAM FOR BN-350 REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    P. STAPLES; J. HALBIG; ET AL

    1999-04-01

    A project to verify the fissile content of fast breeder reactor spent nuclear fuel is underway in the Republic of Kasakhstan. There are a variety of assembly types with different irradiation histories and profiles in the reactor that require a variety of measurement and analysis procedures. These procedures will be discussed and compared as will the general process that has been designed to resolve any potential measurement discrepancies. The underwater counter is part of a system that is designed to assist the International Atomic Energy Agency (IAEA) in maintaining continuity of knowledge from the time of measurement until the measured item is placed in a welded container with a unique identification. In addition to satisfying IAEA requirements for the spent nuclear fuel, this measurement program is able to satisfy some of the measurement requirements for the Kasakhstan Atomic Energy Agency concerning the repackaging of the spent nuclear fuel into a standard canister. The project is currently operational in a mode requiring the IAEA's continuous presence.

  20. The Development of the Advanced Light Water Reactor in Korea - The Korean Next Generation Reactor -

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H.G.; Park, Y.S.; Kim, B.S.; Cho, S.J. [Korea Electric Power Research Institute, Taejon (Korea)

    1999-07-01

    Korean next generation reactor (KNGR), which is to be designed as a standard evolutionary advanced light water reactor (ALWR) in Korea, has been developed from 1992 as one of long-term government projects. The major characteristics of the KNGR are as follows; KNGR is 2-loop PWR and its design lift time is 60 years. The CDF and the CFF will be much lower than 10{sup -5}/RY and 10{sup -6}/RY, respectively. For the design improvement, KNGR adopted inconel-690 as a steam generator tube material, four train ECCS, refueling water storage tank inside containment, and double cylindrical concrete containment. For more reliable and easier control, compact workstations have been adopted in the design of main control complex and digital I and C technology is used for protection, control, and monitoring. In addition, KNGR has some passive design features such as fluidic device in safety injection tank, passive secondary condensing system, and passive auto-catalytic hydrogen recombiner to enhance safety. (author). 4 refs., 4 figs.

  1. Reference worldwide model for antineutrinos from reactors

    Science.gov (United States)

    Baldoncini, Marica; Callegari, Ivan; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2015-03-01

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework, we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency. We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO +), and proposed (Juno, RENO-50, LENA, and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation, and detection processes are estimated using a Monte Carlo-based approach, which provides an overall site-dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes, and we estimate a 2.4% increase of the unoscillated event rate in the geoneutrino energy window due to the storage of spent nuclear fuels in the cooling pools. We predict that the research reactors contribute to less than 0.2% to the commercial reactor signal in the investigated 14 sites. We perform a multitemporal analysis of the expected reactor signal over a time lapse of ten years using reactor operational records collected in a comprehensive database published at www.fe.infn.it/antineutrino.

  2. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  3. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  4. 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.

  5. Design of a reactor core in the Oma Full MOX-ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Hama, Teruo [Electric Power Development Co. Ltd., Tokyo (Japan)

    1999-09-01

    The Electric Power Development Co., Ltd. has progressed a construction plan on an improved boiling-water reactor aiming at loading of MOX fuel in all reactor cores (full MOX-ABWR) at Oma-cho, Aomori prefecture, which is a last stage on application of approval on establishment at present. Here were described on outlines of reactor core in the full MOX-ABWR and its safety evaluation. For the full MOX-ABWR loading MOX fuel assembly into all reactor core, thermal and mechanical design analysis of fuel bars and core design analysis were conducted. As a result, it was confirmed that judgement standards in mixed core of MOX fuel and uranium fuel were also applicable as well as that in uranium fuel. (G.K.)

  6. 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

  7. 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.)

  8. A simple model of reactor cores for reactor neutrino flux calculations for the KamLAND experiment

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, K. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan)]. E-mail: kyo@awa.tohoku.ac.jp; Inoue, K. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Owada, K. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Suekane, F. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Suzuki, A. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Hirano, G. [TEPCO Systems Corporation, Tokyo 135-0034 (Japan); Kosaka, S. [TEPCO Systems Corporation, Tokyo 135-0034 (Japan); Ohta, T. [Tokyo Electric Power Company, Tokyo 100-8560 (Japan); Tanaka, H. [Tokyo Electric Power Company, Tokyo 100-8560 (Japan)

    2006-12-21

    KamLAND is a reactor neutrino oscillation experiment with a very long baseline. This experiment successfully measured oscillation phenomena of reactor antineutrinos coming mainly from 53 reactors in Japan. In order to extract the results, it is necessary to accurately calculate time-dependent antineutrino spectra from all the reactors. A simple model of reactor cores and code implementing it were developed for this purpose. This paper describes the model of the reactor cores used in the KamLAND reactor analysis.

  9. 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.

  10. 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.

  11. 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

  12. Mechanical systems development of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Keun Bae; Chang, M. H.; Kim, J. I.; Choi, S.; Kim, K. S.; Kim, T. W.; Jeong, K. H.; Kim, J. H.; Kim, Y. W.; Lee, G. M.

    1997-07-01

    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 applications such as small capacity power generation, co-generation and sea water desalination. This in mind, survey has been made on the worldwide small and medium integral reactors under development. Reviewed are their technical characteristics, development status, design features, application plans, etc. For the mechanical design scope of work, the structural concept compatible with the characteristics and requirements of integral reactor has been established. Types of major components were evaluated and selected. Functional and structural concept, equipment layout and supporting concept within the reactor pressure vessel have also been established. Preliminary mechanical design requirements were developed considering the reactor lifetime, operation conditions, and the expected loading combinations. To embody the concurrent design approach, recent CAD technology and team engineering concept were evaluated. (author). 31 refs.,16 tabs., 35 figs.

  13. Profiling a reactor component using ultrasonics

    Energy Technology Data Exchange (ETDEWEB)

    Pathak, L.; Seshadri, V.R.; Kumaravadivelu, C.; Sreenivasan, G.; Raghunathan, V.S. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-04-01

    Nuclear reactors have many components within the reactor vessel. During the life of a reactor it is possible for these components to be displaced or deformed because of the thermal cycles to which they are subject. Also, these components in situ therefore becomes important for the upkeep of the reactor. However, high radiation levels make it difficult to monitor using optical methods. This paper describes an ultrasonic method which was successfully employed in profiling a deformed guide tube of a reactor. The method uses the well-known ultrasonic ranging technique. However, the specialty of the method is the use of air transducers at 40 kHz to overcome the inherent divergence problems and the difficulties associated with high temperatures inherent in a sodium cooled reactor.

  14. 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.

  15. 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)

  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. Structures and Materials of Reactor Internals for PWR in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Kim, W. S.; Kwon, S. C.; Kwon, J. H.; Kim, Y. S.; Kim, H. P.; Yoo, C. S.; Lee, S. R.; Jung, M. K.; Hwang, S. S

    2007-10-15

    Nuclear reactor types in Korea are PWR type reactor (Westinghouse, Combustion Engineering, Farmatome type) and CANDU type reactor. Structures and Materials for reactor internal of PWR type were investigated. Reactor internal was composed of lower core support structure, upper core support assembly, incore instrumentation support structure. Lower core support structure of these structures is the most important. The major material for the reactor internal is type 304 and 316 stainless steel and radial support clevis bolts are made of Inconel. The main damage mechanism for reactor internal was IASCC and the effect of IASCC on reactor internal was investigated. The accident for reactor internal was also investigate.

  18. Contribution of reactor physics in past and future. Is reactor physics useful?

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan); Kosaka, Shinya [TEPCO Systems Co. (Japan); Tatsumi, Masahiro [Nuclear Fuel Industries Ltd., Tokyo (Japan)] (and others)

    2003-02-01

    Reactor Physics is a science to create rector and to play an important role in application to calculation science and safety evaluation. This feature articles contains topics, interested problems and development problems in the following field of reactor physics such as theory and experiment of reactor physics, core control, safety evaluation, criticality safety, accelerator driven subcritical reactor (ADS), new type reactor and evaluation of reactor physics. An original nuclear calculation method developed in Japan has been applied to design and analysis of fast breeder reactor. Interested problems are a proposal of fundamental principles of progressive reactor, development of calculation science, new knowledge by application of best estimate method to safety evaluation and investigation of complicated phenomena of criticality safety. (S.Y.)

  19. 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

  20. Savannah River Site peer evaluator standards: Operator assessment for restart

    Energy Technology Data Exchange (ETDEWEB)

    1990-06-01

    Savannah River Site has implemented a Peer Evaluator program for the assessment of certified Central Control Room Operators, Central Control Room Supervisors and Shift Technical Engineers prior to restart. This program is modeled after the nuclear Regulatory Commission`s (NRC`s) Examiner Standard, ES-601, for the requalification of licensed operators in the commercial utility industry. It has been tailored to reflect the unique differences between Savannah River production reactors and commercial power reactors.

  1. Savannah River Site peer evaluator standards: Operator assessment for restart

    Energy Technology Data Exchange (ETDEWEB)

    1990-06-01

    Savannah River Site has implemented a Peer Evaluator program for the assessment of certified Central Control Room Operators, Central Control Room Supervisors and Shift Technical Engineers prior to restart. This program is modeled after the nuclear Regulatory Commission's (NRC's) Examiner Standard, ES-601, for the requalification of licensed operators in the commercial utility industry. It has been tailored to reflect the unique differences between Savannah River production reactors and commercial power reactors.

  2. Fission control system for nuclear reactor

    Science.gov (United States)

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  3. Nuclear reactor shield including magnesium oxide

    Science.gov (United States)

    Rouse, Carl A.; Simnad, Massoud T.

    1981-01-01

    An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

  4. Fuel handling apparatus for a nuclear reactor

    Science.gov (United States)

    Hawke, Basil C.

    1987-01-01

    Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

  5. SUPERHEATING IN A BOILING WATER REACTOR

    Science.gov (United States)

    Treshow, M.

    1960-05-31

    A boiling-water reactor is described in which the steam developed in the reactor is superheated in the reactor. This is accomplished by providing means for separating the steam from the water and passing the steam over a surface of the fissionable material which is not in contact with the water. Specifically water is boiled on the outside of tubular fuel elements and the steam is superheated on the inside of the fuel elements.

  6. 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.

  7. Reactors are indispensable for radioisotope production.

    Science.gov (United States)

    Mushtaq, Ahmad

    2010-12-01

    Radioisotopes can be produced by reactors and accelerators. For certain isotopes there could be an advantage to a certain production method. However, nowadays many reports suggest, that useful isotopes needed in medicine, industry and research could be produced efficiently and dependence on reactors using enriched U-235 may be eliminated. In my view reactors and accelerators will continue to play their role side by side in the supply of suitable and economical sources of isotopes.

  8. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  9. 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.

  10. The gas turbine-modular helium reactor (GT-MHR), high efficiency, cost competitive, nuclear energy for the next century

    Energy Technology Data Exchange (ETDEWEB)

    Zgliczynski, J.B.; Silady, F.A.; Neylan, A.J.

    1994-04-01

    The Gas Turbine-Modular Helium Reactor (GT-MHR) is the result of coupling the evolution of a small passively safe reactor with key technology developments in the US during the last decade: large industrial gas turbines, large active magnetic bearings, and compact, highly effective plate-fin heat exchangers. The GT-MHR is the only reactor concept which provides a step increase in economic performance combined with increased safety. This is accomplished through its unique utilization of the Brayton cycle to produce electricity directly with the high temperature helium primary coolant from the reactor directly driving the gas turbine electrical generator. This cannot be accomplished with another reactor concept. It retains the high levels of passive safety and the standardized modular design of the steam cycle MHTGR, while showing promise for a significant reduction in power generating costs by increasing plant net efficiency to a remarkable 47%.

  11. 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.

  12. Biofilm carrier migration model describes reactor performance.

    Science.gov (United States)

    Boltz, Joshua P; Johnson, Bruce R; Takács, Imre; Daigger, Glen T; Morgenroth, Eberhard; Brockmann, Doris; Kovács, Róbert; Calhoun, Jason M; Choubert, Jean-Marc; Derlon, Nicolas

    2017-06-01

    The accuracy of a biofilm reactor model depends on the extent to which physical system conditions (particularly bulk-liquid hydrodynamics and their influence on biofilm dynamics) deviate from the ideal conditions upon which the model is based. It follows that an improved capacity to model a biofilm reactor does not necessarily rely on an improved biofilm model, but does rely on an improved mathematical description of the biofilm reactor and its components. Existing biofilm reactor models typically include a one-dimensional biofilm model, a process (biokinetic and stoichiometric) model, and a continuous flow stirred tank reactor (CFSTR) mass balance that [when organizing CFSTRs in series] creates a pseudo two-dimensional (2-D) model of bulk-liquid hydrodynamics approaching plug flow. In such a biofilm reactor model, the user-defined biofilm area is specified for each CFSTR; thereby, Xcarrier does not exit the boundaries of the CFSTR to which they are assigned or exchange boundaries with other CFSTRs in the series. The error introduced by this pseudo 2-D biofilm reactor modeling approach may adversely affect model results and limit model-user capacity to accurately calibrate a model. This paper presents a new sub-model that describes the migration of Xcarrier and associated biofilms, and evaluates the impact that Xcarrier migration and axial dispersion has on simulated system performance. Relevance of the new biofilm reactor model to engineering situations is discussed by applying it to known biofilm reactor types and operational conditions.

  13. 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....

  14. Non-equilibrium radiation nuclear reactor

    Science.gov (United States)

    Thom, K.; Schneider, R. T. (Inventor)

    1978-01-01

    An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

  15. 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.

  16. 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.

  17. 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...

  18. 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...

  19. 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...

  20. 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...

  1. 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

  2. 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

  3. NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR

    Science.gov (United States)

    Young, G.J.

    1959-06-30

    The method and means for controlling the neutron density in a nuclear reactor is described. It describes the method and means for flattening the neutron density distribution curve across the reactor by spacing the absorbing control members to varying depths in the central region closer to the center than to the periphery of the active portion of the reactor to provide a smaller neutron reproduction ratio in the region wherein the members are inserted, than in the remainder of the reactor thereby increasing the over-all potential power output.

  4. 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

  5. Heat dissipating nuclear reactor with metal liner

    Science.gov (United States)

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  6. 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.

  7. 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.

  8. Standard specification for nuclear-grade hafnium oxide powder

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This specification defines the physical and chemical requirements for hafnium oxide powder intended for fabrication into shapes for use in a nuclear reactor core. 1.2 The material described herein shall be particulate in nature. 1.3 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  9. (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.

  10. 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.

  11. CHIMNEY FOR BOILING WATER REACTOR

    Science.gov (United States)

    Petrick, M.

    1961-08-01

    A boiling-water reactor is described which has vertical fuel-containing channels for forming steam from water. Risers above the channels increase the head of water radially outward, whereby water is moved upward through the channels with greater force. The risers are concentric and the radial width of the space between them is somewhat small. There is a relatively low rate of flow of water up through the radially outer fuel-containing channels, with which the space between the risers is in communication. (AE C)

  12. Synfuel production in nuclear reactors

    Science.gov (United States)

    Henning, C.D.

    Apparatus and method for producing synthetic fuels and synthetic fuel components by using a neutron source as the energy source, such as a fusion reactor. Neutron absorbers are disposed inside a reaction pipe and are heated by capturing neutrons from the neutron source. Synthetic fuel feedstock is then placed into contact with the heated neutron absorbers. The feedstock is heated and dissociates into its constituent synfuel components, or alternatively is at least preheated sufficiently to use in a subsequent electrolysis process to produce synthetic fuels and synthetic fuel components.

  13. 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.

  14. Autonomous Control of Space Nuclear Reactors

    Science.gov (United States)

    Merk, John

    2013-01-01

    Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the

  15. Investigation of molten salt fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kubota, Kenichi; Enuma, Yasuhiro; Tanaka, Yoshihiko; Konomura, Mamoru; Ichimiya, Masakazu [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2000-06-01

    Phase I of Feasibility Studies on Commercialized Fast Reactor System is being performed for two years from Japanese Fiscal Year 1999. In this report, results of the study on fluid fuel reactors (especially a molten salt fast breeder reactor concept) are described from the viewpoint of technical and economical concerns of the plant system design. In JFY1999, we have started to investigate the fluid fuel reactors as alternative concepts of sodium cooled FBR systems with MOX fuel, and selected the unique concept of a molten chloride fast breeder reactor, whose U-Pu fuel cycle can be related to both light water reactors and fast breeder reactors on the basis of present technical data and design experiences. We selected a preliminary composition of molten fuel and conceptual plant design through evaluation of technical and economical issues essential for the molten salt reactors and then compared them with reference design concepts of sodium cooled FBR systems under limited information on the molten chloride fast breeder reactors. The following results were obtained. (1) The molten chloride fast breeder reactors have inherent safety features in the core and plant performances, ad the fluid fuel is quite promising for cost reduction of the fuel fabrication and reprocessing. (2) On the other hand, the inventory of the molten chloride fuel becomes high and thermal conductivity of the coolant is inferior compared to those of sodium cooled FBR systems, then, the size of main components such as IHX's becomes larger and the amount of construction materials is seems to be increased. (3) Furthermore economical vessel and piping materials which contact with the molten chloride salts are required to be developed. From the results, it is concluded that further steps to investigate the molten chloride fast breeder reactor concepts are too early to be conducted. (author)

  16. 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...

  17. Simulated nuclear reactor fuel assembly

    Science.gov (United States)

    Berta, Victor T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  18. Dissecting Reactor Antineutrino Flux Calculations

    Science.gov (United States)

    Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

    2017-09-01

    Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235U, 239Pu, 241Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6 % MeV-1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.

  19. New Production Reactors Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-01

    Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs.

  20. Research on plasma core reactors

    Science.gov (United States)

    Jarvis, G. A.; Barton, D. M.; Helmick, H. H.; Bernard, W.; White, R. H.

    1976-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  1. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  2. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  3. 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)

  4. Development of system integration technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Moon Hee; Kang, D. J.; Kim, K. K. and others

    1999-03-01

    The objective of this report is to integrate the conceptual design of an integral reactor, SMART producing thermal energy of 330 MW, which will be utilized to supply energy for seawater desalination and small-scale power generation. This project also aims to develop system integration technology for effective design of the reactor. For the conceptual design of SMART, preliminary design requirements including the top-tier requirements and design bases were evaluated and established. Furthermore, in the view of the application of codes and standards to the SMART design, existing laws, codes and standards were analyzed and evaluated with respect to its applicability. As a part of this evaluation, directions and guidelines were proposed for the development of new codes and standards which shall be applied to the SMART design. Regarding the integration of SMART conceptual designs, major design activities and interfaces between design departments were established and coordinated through the design process. For the effective management of all design schedules, a work performance evaluation system was developed and applied to the design process. As the results of this activity, an integrated output of SMART designs was produced. Two additional scopes performed in this project include the preliminary economic analysis on the SMART utilization for seawater desalination, and the planning of verification tests for technology implemented into SMART and establishing development plan of the computer codes to be used for SMART design in the next phase. The technical cooperation with foreign country and international organization for securing technologies for integral reactor design and its application was coordinated and managed through this project. (author)

  5. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes.

  6. Prospects for improved understanding of isotopic reactor antineutrino fluxes

    Science.gov (United States)

    Gebre, Y.; Littlejohn, B. R.; Surukuchi, P. T.

    2018-01-01

    Predictions of antineutrino fluxes produced by fission isotopes in a nuclear reactor have recently received increased scrutiny due to observed differences in predicted and measured inverse beta decay (IBD) yields, referred to as the "reactor antineutrino flux anomaly." In this paper, global fits are applied to existing IBD yield measurements to produce constraints on antineutrino production by individual plutonium and uranium fission isotopes. We find that fits including measurements from highly U 235 -enriched cores and fits including Daya Bay's new fuel evolution result produce discrepant best-fit IBD yields for U 235 and Pu 239 . This discrepancy can be alleviated in a global analysis of all data sets through simultaneous fitting of Pu 239 , U 235 , and U 238 yields. The measured IBD yield of U 238 in this analysis is (7.02 ±1.65 )×10-43 cm2/fission , nearly two standard deviations below existing predictions. Future hypothetical IBD yield measurements by short-baseline reactor experiments are examined to determine their possible impact on the global understanding of isotopic IBD yields. It is found that future improved short-baseline IBD yield measurements at both high-enriched and low-enriched cores can significantly improve constraints for U 235 , U 238 , and Pu 239 , providing comparable or superior precision to existing conversion- and summation-based antineutrino flux predictions. Systematic and experimental requirements for these future measurements are also investigated.

  7. Demonstration irradiation of CANFLEX in a CANDU 6 power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Inch, Wayne W. R; Cormier, Mike A. [Atomic Energy of Canada Limited (Canada); Thompson, Paul D. [Poing Lepreau Generating Station, New Brunswick Power (Canada); Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-07-01

    The CANFLEX fuel bundle is the latest design in the evolution of CANDU fuel. Its 43-element fuel bundle assembly and its patented critical-heat-flux (CHF) enhancement appendages offer higher operating and safety margins than the standard 37-element bundle; in addition, it is fully compatible with operating CANDU reactors. Since 1991, Atomic Energy of Canda Limited has partnered with the Korea Atomic Energy Rearch Institute to complete the development, qualification testing and analysis of the CANFLEX fuel bundle. A 2-channel, 24-bundle demonstration irradiation of CANFLEX was started on 1998 September 03 at New Rrunswick Power's Point Lepreau Generating Station. All 24 bundles have been loaded into the reactor and 16 bundles have completed their planned irradiation. The irradiated bundles will be examined to verify bundle integrity and condition. The irradiation data and in-bay photographs are expected to show that the in-reactor performance of the CANFLEX fuel has met all design criteria and that it is fully compatible with existing CANDU reactors. These results will be important in supporting the use of CANFLEX as the production fuel in CANDU 6 reactors. Several bundles irradiated in this demonstration irradiation will be destructively examined in post-irradiation examination in the hot cells at Chalk River Laboratories of Atomic Energy of Canada Limited. CANFLEX fuel development is part of an overall strategy on plant aging, which recognized that certain processes were taking place within the heat-transport system which, if left unabated, could result in a decrease in the margin to fuel-sheath dryout. Because of the increase in critical channel power brought about by the improved CANFLEX design, it was recognized that this new fuel, when used in combination with other remedial actions, could counterbalance the adverse effects of aging within the heat-transport system. Water CHF tests have been completed to firmly establish the thermalhydraulic performance

  8. 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

  9. 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.

  10. 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.

  11. Design and test of a continuous reactor for palm oil transesterification

    Directory of Open Access Journals (Sweden)

    Michael Allen

    2006-07-01

    Full Text Available The continuous reactor for transesterification of refined palm oil with methanol was designed and tested. The reaction condition was focused at ambient pressure, temperature of 60ºC, molar ratio of alcohol to oil of 6:1, and NaOH of 1.0 %wt of oil. The designed reactor was in a form of a 6-stage mechanically stirred tank. Rushton turbines, with 4 standard baffles, and plates with a small opening were installed inside. The reactor has a simple form which could be conveniently constructed and operated. The reactor could produce methyl esters (ME with purities ranging from 97.5-99.2 %wt within residence times of 6-12 minutes in which its production performance was equivalent to a plug flow reactor and the power consumption of a stirrer in the range of 0.2-0.6 kW/m3 was required. The reaction modeling based on a homogeneous concentration field with reaction kinetics could accurately predict the produced purities of ME. The production yields by weight of final product and of ME to the fed oil were 94.7 and 92.3%, respectively. The developed continuous reactor has good potential for producing ME to be used as biodiesel.

  12. A New In-core Production Method of Co-60 in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lyu, Jinqi; Kim, Woosong; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of); Park, Younwon [BEES Inc, Daejeon (Korea, Republic of)

    2016-05-15

    This study introduces an innovative method for Co-60 production in the CANDU6 core. In this new scheme, the central fuel element is replaced by a Co-59 target and Co-60 is obtained after the fuel bundle is discharged. It has been shown that the new method can produce significantly higher amount of Co-60 than the conventional Co production method in CANDU6 reactors without compromising the fuel burnup by removing some (<50%) of the adjuster rods in the whole core. The coolant void reactivity is noticeably reduced when a Co-59 target is loaded into the central pin of the fuel bundle. Meanwhile, the peak power in a fuel bundle is just a little higher due to the central Co-59 target than in conventional CANDU6 fuel design. The basic technology for Co-60 producing was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) in 1946 and the same technology was adapted and applied in CANDU6 power reactors. The standard CANDU6 reactor has 21 adjuster rods which are fully inserted into the core during normal operation. The stainless steel adjuster rods are replaced with neutronically-equivalent Co-59 adjusters to produce Co-60. Nowadays, the roles of the adjuster rods are rather vague since nuclear reactors cannot be quickly restarted after a sudden reactor trip due to more stringent regulations. In some Canadian CANDU6 reactors, some or all the adjuster rods are removed from the core to maximize the uranium utilization.

  13. Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

    Science.gov (United States)

    Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

    2017-11-01

    The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

  14. 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

  15. Design of an Organic Simplified Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-08-01

    Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  16. Monolithic reactor : Higher yield, less energy

    NARCIS (Netherlands)

    Kreutzer, M.T.; Moulijn, J.A.; Kapteijn, F.; Mols, B.

    2004-01-01

    The production of margarine, the desulphurisation of crude oil, and the manufacture of synthetic diesel fuel, these are only three of the many industrial processes in which a three-phase reactor is used. Traditionally, this type of reactor is rather ill-defined. Success with a lab scale set-up is no

  17. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  18. Reactor antineutrino spectra and forbidden beta decays

    Science.gov (United States)

    Štefánik, Dušan; Dvornický, Rastislav; Šimkovic, Fedor

    2017-10-01

    The exact relativistic shape factors, associated with the nuclear matrix elements governing the first forbidden beta decays, are presented. It is expected that their consideration can allow a more accurate theoretical description of antineutrino fluxes from the power reactor. A qualitative analysis of the uncertainty of reactor antineutrino flux from 235U within the electron spectrum conversion method is performed.

  19. Helium-cooled high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Trauger, D.B.

    1985-01-01

    Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.

  20. Rotor for a pyrolysis centrifuge reactor

    DEFF Research Database (Denmark)

    2015-01-01

    The present invention relates to a rotor for a pyrolysis centrifuge reactor, said rotor comprising a rotor body having a longitudinal centre axis, and at least one pivotally mounted blade being adapted to pivot around a pivot axis under rotation of the rotor body around the longitudinal centre axis....... Moreover, the present invention relates to a pyrolysis centrifuge reactor applying such a rotor....

  1. The Design of a Nuclear Reactor

    Indian Academy of Sciences (India)

    IAS Admin

    The aim of this largely pedagogical article is to employ pre-college physics to arrive at an un- derstanding of a system as complex as a nuclear reactor. We focus on three key issues: the fuel pin, the moderator, and lastly the dimensions of the nuclear reactor. 1. Introduction. Design considerations have engaged human ...

  2. Advances in Tandem Mirror fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, L.J.; Logan, B.G.

    1986-05-20

    The Tandem Mirror exhibits several distinctive features which make the reactor embodiment of the principle very attractive: Simple low-technology linear central cell; steady-state operation; high-..beta.. operation; no driven current or disruptions; divertorless operation; direction conversion of end-loss power; low-surface heat loads; and advanced fusion fuel capability. In this paper, we examine these features in connection with two tandem mirror reactor designs, MARS and MINIMARS, and several advanced reactor concepts including the wall-stabilized reactor and the field-reversed mirror. With a novel compact end plug scheme employing octopole stabilization, MINIMARS is expressly designed for short construction times, factory-built modules, and a small (600 MWe) but economic reactor size. We have also configured the design for low radioactive afterheat and inherent/passive safety under LOCA/LOFA conditions, thereby obviating the need for expensive engineered safety systems. In contrast to the complex and expensive double-quadrupole end-cell of the MARS reactor, the compact octopole end-cell of MINIMARS enables ignition to be achieved with much shorter central cell lengths and considerably improves the economy of scale for small (approx.250 to 600 MWe) tandem mirror reactors. Finally, we examine the prospects for realizing the ultimate potential of the tandem mirror with regard to both innovative configurations and novel neutron energy conversion schemes, and stress that advanced fuel applications could exploit its unique reactor features.

  3. 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.

  4. Utilisation of British University Research Reactors.

    Science.gov (United States)

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  5. The First Reactor, 40th Anniversary (rev.)

    Energy Technology Data Exchange (ETDEWEB)

    Allardice, Corbin; Trapnell, Edward R; Fermi, Enrico; Fermi, Laura; Williams, Robert C

    1982-12-01

    This booklet, an updated version of the original booklet describing the first nuclear reactor, was written in honor of the 40th anniversary of the first reactor or "pile". It is based on firsthand accounts told to Corbin Allardice and Edward R. Trapnell, and includes recollections of Enrico and Laura Fermi.

  6. Helix reactor: great potential for flow chemistry

    NARCIS (Netherlands)

    Geerdink, P.; Runstraat, A. van den; Roelands, C.P.M.; Goetheer, E.L.V.

    2009-01-01

    The Helix reactor is highly suited for precise reaction control based on good hydrodynamics. The hydrodynamics are controlled by the Dean vortices, which create excellent heat transfer properties, approach plug flow and avoid turbulence. The flexibility of this reactor has been demonstrated using a

  7. 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

  8. 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

  9. Quality management in BNCT at a nuclear research reactor.

    Science.gov (United States)

    Sauerwein, Wolfgang; Moss, Raymond; Stecher-Rasmussen, Finn; Rassow, Jürgen; Wittig, Andrea

    2011-12-01

    Each medical intervention must be performed respecting Health Protection directives, with special attention to Quality Assurance (QA) and Quality Control (QC). This is the basis of safe and reliable treatments. BNCT must apply QA programs as required for performance and safety in (conventional) radiotherapy facilities, including regular testing of performance characteristics (QC). Furthermore, the well-established Quality Management (QM) system of the nuclear reactor used has to be followed. Organization of these complex QM procedures is offered by the international standard ISO 9001:2008. Copyright © 2011 Elsevier Ltd. All rights reserved.

  10. 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.

  11. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  12. Reactor core and plant design concepts of the Canadian supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Bailey, J.; Rhodes, D.; Guzonas, D.; Hamilton, H.; Haque, Z.; Pencer, J.; Sartipi, A. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada is developing a 1200 MWe supercritical water-cooled reactor (SCWR), which has evolved from the well-established pressure-tube type CANDU{sup 1} reactor. This SCWR reactor concept, which is often referred to as the Canadian SCWR, uses supercritical water as a coolant, has a low-pressure heavy water moderator and a direct cycle for power production. The reactor concept incorporates advanced safety features, such as passive emergency core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce the core damage frequency beyond existing nuclear reactors. This paper presents a description of the Canadian SCWR core design concept, the integration of in-core and out-of-core components and the mechanical plant design concept. Supporting systems for reactor safety, reactor control and moderator cooling are also described. (author)

  13. Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

    2017-04-01

    Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

  14. Cooling system for a nuclear reactor

    Science.gov (United States)

    Amtmann, Hans H.

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  15. Slurry Bubble Column Reactor Optimization (book chapter)

    Energy Technology Data Exchange (ETDEWEB)

    Gamwo, I.K.; Gidaspow, D. (Illinois Inst. of Technology, Chicago, IL); Jung, J. (ANL)

    2007-03-01

    Slurry bubble column reactors (SBCR) are the preferred contactors for the conversion of syngas to fules and chemicals partially due to their superior heat and mass transfer characteristics. The multiphase fluid dynamics in these systems greatly affect the reactor volumetric productivity. Here, we have developed a computational fluid dynamics (CFD) assisted design methodology for searching the optimum particle size for maximum production in a SBCR. Reactor optimization due to heat exchanger configuration was also investigated. We have rearranged the heat exchangers in a SBCR and constructed a CFD model for a baffled reactor. The novel arrangement of the exchangers prevents the unfavorable high catalysts concentration at the lower stage of the reactor. Thus an optimum catalyst concentration is maintained during the course of the production of liquid fuels.

  16. Scanning tunneling microscope assembly, reactor, and system

    Science.gov (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  17. Cleaning natural water in the clarifier reactor

    Science.gov (United States)

    Skolubovich, Yuriy; Skolubovich, Aleksandr; Voitov, Evgeniy; Soppa, Mikhail; Chirkunov, Yuriy

    2017-10-01

    The problems of cleaning low turbidity high-color surface waters for drinking water supply are considered. A new design of the clarifier reactor is proposed, which increases the efficiency of water purification and at the same time reduces its operating costs. A detailed description of clarifier reactor design and its operation is given. The study results of the clarifier reactor operation in real conditions for the purification of low turbidity high-color waters are shown. Due to the weighted layer of dense loading use in a process of water purification, the structure productivity can be increased by 2-3 times in comparison with conventional clarifiers with suspended sediment. Using reagents for water purification, the clarifier reactor, due to the processes of contact coagulation, allows reducing the consumption of reagents up to 50%. Investigations of the clarifier reactor operation in technological schemes for various waters purification, including sewage, showed their effectiveness and prospects.

  18. Seismic attenuation system for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liszkai, Tamas; Cadell, Seth

    2018-01-30

    A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

  19. TRIGA research reactors; Reacteurs de recherche Triga

    Energy Technology Data Exchange (ETDEWEB)

    Fouquet, D.M.; Razvi, J.; Whittemore, W.L. [Triga General Atomics, San Diego, CA (United States); Duban, B.; Harbonnier, G.; Du Limbert, P.; Durand, J.P. [AREVA/FRAMATOME ANP/CERCA, 92 - Paris-La-Defence (France)

    2004-02-01

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  20. Nuclear propulsion apparatus with alternate reactor segments

    Science.gov (United States)

    Szekely, Thomas

    1979-04-03

    1. Nuclear propulsion apparatus comprising: A. means for compressing incoming air; B. nuclear fission reactor means for heating said air; C. means for expanding a portion of the heated air to drive said compressing means; D. said nuclear fission reactor means being divided into a plurality of radially extending segments; E. means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and F. means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus.

  1. Reactivity control assembly for nuclear reactor. [LMFBR

    Science.gov (United States)

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  2. Gobioecetes biwaensis n. g., n. sp. (Monogenea: Dactylogyridae) from the gills of a freshwater gobiid fish, Rhinogobius sp. BW Takahashi & Okazaki, 2002, with a redescription of Parancyrocephaloides daicoci Yamaguti, 1938.

    Science.gov (United States)

    Ogawa, Kazuo; Itoh, Naoki

    2017-06-01

    A description is given of Gobioecetes biwaensis n. g., n. sp. (Monogenea: Dactylogyridae) on the gills of a freshwater gobiid fish, Rhinogobius sp. BW Takahashi & Okazaki, 2002 from Lake Biwa, Shiga Prefecture and Rhinogobius sp. from Sento Imperial Palace, Kyoto Prefecture, Japan. This monogenean is most similar to Parancyrocephaloides rhinogobius Ling, 1973, proposed here as Gobioecetes rhinogobius n. comb., collected from the gills of Rhinogobius giurinus (Rutter, 1897) in Hupei Province, China, but can be differentiated from it in that ventral hamuli of G. biwaensis are larger than those of G. rhinogobius and in that in G. biwaensis, ventral hamuli are larger than dorsal ones, but in G. rhinogobius, ventral and dorsal hamuli are almost equal in size. The new genus is characterized by the absence of a bar supporting dorsal hamuli. Parancyrocephaloides daicoci Yamaguti, 1938, the type species of the genus, is redescribed based on newly collected specimens from a marine dactylopterid, Dactyloptena peterseni (Nyström, 1887) and the type specimens, revealing that it has a pair of vestigial bars supporting dorsal hamuli. The generic diagnoses of Gobioecetes n. g. and Parancyrocephaloides are given and systematic position of these genera in Dactylogyridae is discussed in this paper. Copyright © 2017 Elsevier Ireland Ltd. All rights reserved.

  3. Ceramic oxygen transport membrane array reactor and reforming method

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.; Christie, Gervase Maxwell; Robinson, Charles; Wilson, Jamie R.; Gonzalez, Javier E.; Doraswami, Uttam R.

    2016-11-08

    The invention relates to a commercially viable modular ceramic oxygen transport membrane reforming reactor configured using repeating assemblies of oxygen transport membrane tubes and catalytic reforming reactors.

  4. Reactor process water (PW) piping inspections, 1984--1990

    Energy Technology Data Exchange (ETDEWEB)

    Ehrhart, W.S.; Elder, J.B.; Sprayberry, R.E.; Vande Kamp, R.W.

    1990-12-31

    In July 1983, the NRC ordered the shutdown of five boiling water reactors (BWR`s) because of concerns about reliability of ultrasonic examination for detecting intergranular stress corrosion cracking (IGSCC). These concerns arose because of leaking piping at Niagara Mohawk`s Nine Mile Point which was attributed to IGSCC. The leaks were detected shortly after completion of ultrasonic examinations of the piping. At that time, the Dupont plant manager at Savannah River (SR) directed that investigations be performed to determine if similar problems could exist in SR reactors. Investigation determined that all conditions believed necessary for the initiation and propagation of IGSCC in austenitic stainless steel exist in SR reactor process water (PW) systems. Sensitized, high carbon, austenitic stainless steel, a high purity water system with high levels of dissolved oxygen, and the residual stresses associated with welding during construction combine to provide the necessary conditions. A periodic UT inspection program is now in place to monitor the condition of the reactor PW piping systems. The program is patterned after NRC NUREG 0313, i.e., welds are placed in categories based on their history. Welds in upgraded or replacement piping are examined on a standard schedule (at least every five years) while welds with evidence of IGSCC, evaluated as acceptable for service, are inspected at every extended outage (15 to 18 months). This includes all welds in PW systems three inches in diameter and above. Welds are replaced when MSCC exceeds the replacement criteria of more than twenty percent of pipe circumference of fifty percent of through-wall depth. In the future, we intend to perform flow sizing with automated UT techniques in addition to manual sizing to provide more information for comparison with future examinations.

  5. Reactor process water (PW) piping inspections, 1984--1990

    Energy Technology Data Exchange (ETDEWEB)

    Ehrhart, W.S.; Elder, J.B.; Sprayberry, R.E.; Vande Kamp, R.W.

    1990-01-01

    In July 1983, the NRC ordered the shutdown of five boiling water reactors (BWR's) because of concerns about reliability of ultrasonic examination for detecting intergranular stress corrosion cracking (IGSCC). These concerns arose because of leaking piping at Niagara Mohawk's Nine Mile Point which was attributed to IGSCC. The leaks were detected shortly after completion of ultrasonic examinations of the piping. At that time, the Dupont plant manager at Savannah River (SR) directed that investigations be performed to determine if similar problems could exist in SR reactors. Investigation determined that all conditions believed necessary for the initiation and propagation of IGSCC in austenitic stainless steel exist in SR reactor process water (PW) systems. Sensitized, high carbon, austenitic stainless steel, a high purity water system with high levels of dissolved oxygen, and the residual stresses associated with welding during construction combine to provide the necessary conditions. A periodic UT inspection program is now in place to monitor the condition of the reactor PW piping systems. The program is patterned after NRC NUREG 0313, i.e., welds are placed in categories based on their history. Welds in upgraded or replacement piping are examined on a standard schedule (at least every five years) while welds with evidence of IGSCC, evaluated as acceptable for service, are inspected at every extended outage (15 to 18 months). This includes all welds in PW systems three inches in diameter and above. Welds are replaced when MSCC exceeds the replacement criteria of more than twenty percent of pipe circumference of fifty percent of through-wall depth. In the future, we intend to perform flow sizing with automated UT techniques in addition to manual sizing to provide more information for comparison with future examinations.

  6. Achieving Standardization

    DEFF Research Database (Denmark)

    Henningsson, Stefan

    2016-01-01

    International e-Customs is going through a standardization process. Driven by the need to increase control in the trade process to address security challenges stemming from threats of terrorists, diseases, and counterfeit products, and to lower the administrative burdens on traders to stay...

  7. Achieving Standardization

    DEFF Research Database (Denmark)

    Henningsson, Stefan

    2014-01-01

    International e-Customs is going through a standardization process. Driven by the need to increase control in the trade process to address security challenges stemming from threats of terrorists, diseases, and counterfeit products, and to lower the administrative burdens on traders to stay...

  8. PKI standards

    OpenAIRE

    Mitchell, Chris J

    2000-01-01

    This paper provides a review of the current state of the art in standards for Public Key Infrastructures. The main focus of the paper is the recent work by the Internet Engineering Task Force, ITU-T, and ISO/IEC.

  9. 77 FR 61447 - Seeks Qualified Candidates for the Advisory Committee on Reactor Safeguards

    Science.gov (United States)

    2012-10-09

    ... adequacy of proposed reactor safety standards. Of primary importance are the safety issues associated with... mixed oxide and high burnup fuels. An increased emphasis is being given to safety issues associated with... made without regard to factors such as race, color, religion, national origin, sex, age, or...

  10. 75 FR 66168 - Seeks Qualified Candidates for the Advisory Committee on Reactor Safeguards

    Science.gov (United States)

    2010-10-27

    ... adequacy of proposed reactor safety standards. Of primary importance are the safety issues associated with... mixed oxide and high burnup fuels. An increased emphasis is being given to safety issues associated with... factors such as race, color, religion, national origin, sex, age, or disabilities. Candidates must be...

  11. Residence time distribution of the gas phase in a mechanically agitated gas-liquid reactor

    NARCIS (Netherlands)

    Thijert, M.P.G.; Oyevaar, M.H.; Kuper, W.J.; Westerterp, K.R.

    1992-01-01

    In this study we present a measuring method and extensive experimental data on the gas phase RTD in a mechanically agitated gas-liquid reactor with standard dimensions over a wide range of superficial gas velocities, agitation rates and agitator sizes. The results are modelled successfully, using

  12. Fast Spectrum Molten Salt Reactor Options

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Patton, Bruce W [ORNL; Howard, Rob L [ORNL; Harrison, Thomas J [ORNL

    2011-07-01

    During 2010, fast-spectrum molten-salt reactors (FS-MSRs) were selected as a transformational reactor concept for light-water reactor (LWR)-derived heavy actinide disposition by the Department of Energy-Nuclear Energy Advanced Reactor Concepts (ARC) program and were the subject of a preliminary scoping investigation. Much of the reactor description information presented in this report derives from the preliminary studies performed for the ARC project. This report, however, has a somewhat broader scope-providing a conceptual overview of the characteristics and design options for FS-MSRs. It does not present in-depth evaluation of any FS-MSR particular characteristic, but instead provides an overview of all of the major reactor system technologies and characteristics, including the technology developments since the end of major molten salt reactor (MSR) development efforts in the 1970s. This report first presents a historical overview of the FS-MSR technology and describes the innovative characteristics of an FS-MSR. Next, it provides an overview of possible reactor configurations. The following design features/options and performance considerations are described including: (1) reactor salt options-both chloride and fluoride salts; (2) the impact of changing the carrier salt and actinide concentration on conversion ratio; (3) the conversion ratio; (4) an overview of the fuel salt chemical processing; (5) potential power cycles and hydrogen production options; and (6) overview of the performance characteristics of FS-MSRs, including general comparative metrics with LWRs. The conceptual-level evaluation includes resource sustainability, proliferation resistance, economics, and safety. The report concludes with a description of the work necessary to begin more detailed evaluation of FS-MSRs as a realistic reactor and fuel cycle option.

  13. The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G. [CEA/Saclay, DEN, 91 - Gif sur Yvette (France)] [and others

    2000-07-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  14. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors.

  15. 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).

  16. Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nielsen, Joseph W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Norman, Daren R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-01

    The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

  17. Replacement reactor to revolutionise magnets

    CERN Document Server

    Atkins, G

    2002-01-01

    Electric motors, hearing aids and magnetic resonance imaging are only some of the applications that will benefit from the first advances in magnets in a quarter of a century. Magnets achieve their characteristics when electrons align themselves to produce a unified magnetic field. Neutrons can probe these magnetic structures. The focus is not just on making more powerful magnets, but also identifying the characteristics that make magnets cheaper and easier for industry to manufacture. Staff from the ANSTO's Neutron Scattering Group have already performed a number of studies on the properties of magnets using using HIFAR, but the Replacement Research Reactor that will produce cold neutrons would allow scientists to investigate the atomic properties of materials with large molecules. A suite of equipment will enable studies at different temperatures, pressures and magnetic fields

  18. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Wike, L.D.; Specht, W.L.; Mackey, H.E.; Paller, M.H.; Wilde, E.W.; Dicks, A.S.

    1989-12-01

    The Savannah River Site (SRS) is a large United States Department of Energy installation on the upper Atlantic Coastal Plain of South Carolina. The SRS contains diverse habitats, flora, and fauna. Habitats include upland terrestrial areas, varied wetlands including Carolina Bays, the Savannah River swamp system, and impoundment related and riparian wetlands, and the aquatic habitats of several stream systems, two large cooling reservoirs, and the Savannah River. These diverse habitats support a large variety of plants and animals including many commercially or recreational valuable species and several rare, threatened or endangered species. This volume describes the major habitats and their biota found on the SRS, and discuss the impacts of continued operation of the K, L, and P production reactors.

  19. FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Bassett, C.H.

    1961-11-21

    A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

  20. Dynamic analysis of process reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shadle, L.J.; Lawson, L.O.; Noel, S.D.

    1995-06-01

    The approach and methodology of conducting a dynamic analysis is presented in this poster session in order to describe how this type of analysis can be used to evaluate the operation and control of process reactors. Dynamic analysis of the PyGas{trademark} gasification process is used to illustrate the utility of this approach. PyGas{trademark} is the gasifier being developed for the Gasification Product Improvement Facility (GPIF) by Jacobs-Siffine Engineering and Riley Stoker. In the first step of the analysis, process models are used to calculate the steady-state conditions and associated sensitivities for the process. For the PyGas{trademark} gasifier, the process models are non-linear mechanistic models of the jetting fluidized-bed pyrolyzer and the fixed-bed gasifier. These process sensitivities are key input, in the form of gain parameters or transfer functions, to the dynamic engineering models.