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Sample records for gen iv spent

  1. Gen IV Materials Handbook Implementation Plan

    International Nuclear Information System (INIS)

    Rittenhouse, P.; Ren, W.

    2005-01-01

    A Gen IV Materials Handbook is being developed to provide an authoritative single source of highly qualified structural materials information and materials properties data for use in design and analyses of all Generation IV Reactor Systems. The Handbook will be responsive to the needs expressed by all of the principal government, national laboratory, and private company stakeholders of Gen IV Reactor Systems. The Gen IV Materials Handbook Implementation Plan provided here addresses the purpose, rationale, attributes, and benefits of the Handbook and will detail its content, format, quality assurance, applicability, and access. Structural materials, both metallic and ceramic, for all Gen IV reactor types currently supported by the Department of Energy (DOE) will be included in the Gen IV Materials Handbook. However, initial emphasis will be on materials for the Very High Temperature Reactor (VHTR). Descriptive information (e.g., chemical composition and applicable technical specifications and codes) will be provided for each material along with an extensive presentation of mechanical and physical property data including consideration of temperature, irradiation, environment, etc. effects on properties. Access to the Gen IV Materials Handbook will be internet-based with appropriate levels of control. Information and data in the Handbook will be configured to allow search by material classes, specific materials, specific information or property class, specific property, data parameters, and individual data points identified with materials parameters, test conditions, and data source. Details on all of these as well as proposed applicability and consideration of data quality classes are provided in the Implementation Plan. Website development for the Handbook is divided into six phases including (1) detailed product analysis and specification, (2) simulation and design, (3) implementation and testing, (4) product release, (5) project/product evaluation, and (6) product

  2. Gen IV Materials Handbook Implementation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Rittenhouse, P.; Ren, W.

    2005-03-29

    A Gen IV Materials Handbook is being developed to provide an authoritative single source of highly qualified structural materials information and materials properties data for use in design and analyses of all Generation IV Reactor Systems. The Handbook will be responsive to the needs expressed by all of the principal government, national laboratory, and private company stakeholders of Gen IV Reactor Systems. The Gen IV Materials Handbook Implementation Plan provided here addresses the purpose, rationale, attributes, and benefits of the Handbook and will detail its content, format, quality assurance, applicability, and access. Structural materials, both metallic and ceramic, for all Gen IV reactor types currently supported by the Department of Energy (DOE) will be included in the Gen IV Materials Handbook. However, initial emphasis will be on materials for the Very High Temperature Reactor (VHTR). Descriptive information (e.g., chemical composition and applicable technical specifications and codes) will be provided for each material along with an extensive presentation of mechanical and physical property data including consideration of temperature, irradiation, environment, etc. effects on properties. Access to the Gen IV Materials Handbook will be internet-based with appropriate levels of control. Information and data in the Handbook will be configured to allow search by material classes, specific materials, specific information or property class, specific property, data parameters, and individual data points identified with materials parameters, test conditions, and data source. Details on all of these as well as proposed applicability and consideration of data quality classes are provided in the Implementation Plan. Website development for the Handbook is divided into six phases including (1) detailed product analysis and specification, (2) simulation and design, (3) implementation and testing, (4) product release, (5) project/product evaluation, and (6) product

  3. Gen IV. Technical and economical aspects

    International Nuclear Information System (INIS)

    Kaluzny, Y.; Legee, F.

    2010-01-01

    In this presentation author deals with development of nuclear reactor type of Generation IV. He concluded that: - Nuclear energy is competitive with regards to the other generation sources; Its competitiveness also increases with CO 2 cost. Considering the nuclear cost breakdown of LWR reactors, it turns out that the uranium is currently not in the range of a threshold for FBR deployment; - The global balance of uranium supply and demand and also innovation required to fulfil GEN IV objectives would probably imply the emergence of fast reactor competitiveness after the turn of the mid-century; - We shall need fast reactors in the coming decade.

  4. Gen IV Materials Handbook Functionalities and Operation

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL

    2009-12-01

    This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

  5. New Materials for NGNP/Gen IV

    International Nuclear Information System (INIS)

    Swindeman, Robert W.; Marriott, Douglas L.

    2009-01-01

    The bounding conditions were briefly summarized for the Next Generation Nuclear Plant (NGNP) that is the leading candidate in the Department of Energy Generation IV reactor program. Metallic materials essential to the successful development and proof of concept for the NGNP were identified. The literature bearing on the materials technology for high-temperature gas-cooled reactors was reviewed with emphasis on the needs identified for the NGNP. Several materials were identified for a more thorough study of their databases and behavioral features relative to the requirements ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NH.

  6. New Materials for NGNP/Gen IV

    Energy Technology Data Exchange (ETDEWEB)

    Robert W. Swindeman; Douglas L. Marriott

    2009-12-18

    The bounding conditions were briefly summarized for the Next Generation Nuclear Plant (NGNP) that is the leading candidate in the Department of Energy Generation IV reactor program. Metallic materials essential to the successful development and proof of concept for the NGNP were identified. The literature bearing on the materials technology for high-temperature gas-cooled reactors was reviewed with emphasis on the needs identified for the NGNP. Several materials were identified for a more thorough study of their databases and behavioral features relative to the requirements ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NH.

  7. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors

    International Nuclear Information System (INIS)

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-01-01

    Many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important criterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals

  8. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-04-23

    many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

  9. Developing new nuclear curricula for GEN IV needs

    International Nuclear Information System (INIS)

    Ghitescu, P.; Pavel, G.L.

    2014-01-01

    States who wish to start and develop a nuclear program must take into consideration a strong proven strategy for developing a sustainable program. A complete nuclear research program must include: a good national strategy and support on the topic; strong research laboratories supported by good personnel; education component to provide sustainable and qualified workforce; national/international interest from stakeholders and governments and a well informed society. New demonstrators are foreseen for the next period to be built in Europe and skilled supporting personnel is strongly needed. Current situation in nuclear higher education with perspective will be analysed. EURATOM strongly supports development of multidisciplinary co-operational projects in order to built such novel initiatives. An example of such program supported by European Commission, ARCADIA, will be given. The project is based on the cooperation of a large number of participants all over Europe and the main purpose is to develop a road-map for Gen IV reactor. (authors)

  10. Effect on introduction of GEN IV nuclear reactor on environment friendliness in Korea

    International Nuclear Information System (INIS)

    Kim, Y. H.; Oh, W. J.; Lim, C. Y.; Lee, K. J.

    2002-01-01

    Nowadays GEN IV - international project for future type(Generation IV) nuclear reactor - is studied actively. Introduction of Fast Reactor and Acceleration Driven System which are GEN IV type reactors can reduce High Level Waste through the nuclear fuel cycle including them. In this study, some fuel cycle schemes including these types of reactors are to be analyzed in a view of material flow by new made calculation tool compared to scheme by only thermal reactors. In view of the results so far achieved in this study, introduction of Gen IV type reactors gives an advantage over another cases

  11. Safety Design Criteria (SDC) for Gen-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Nakai, Ryodai

    2013-01-01

    SDC Development Background & Objectives: • Safety Design Criteria (SDC) Development for Gen-IV SFR: – Proposed at the GIF Policy Group (PG) meeting in October 2010 –SDC “harmonization” is increasingly important for: • Realization of enhanced safety designs meeting to Gen-IV safety goals and safety approach common to SFR systems; • Preparation for the forthcoming licensing in the near future; • Because Gen-IV SFR are progressing into conceptual design stage. • The SDC is the Reference criteria: – Of the designs of safety-related Structures, Systems & Components that are specific to the SFR system; – For clarifying the requisites systematically & comprehensively; – When the technology developers apply the basic safety approach and use the codes & standards for conceptual design of the Gen-IV SFR system

  12. European project SARGEN IV: safety approach and assessment of GEN IV reactors

    International Nuclear Information System (INIS)

    Ammirabile, L.

    2013-01-01

    • SARGEN I V has elaborated a proposal for the harmonization of safety assessment practices for GEN IV NPP. • An overall reinforcement of DiD is expected for GEN I V NPP, including improved independence between all levels of DiD. • An inherent approach should reinforce the fulfillment of fundamental safety functions e.g. the consequences for some situations should be reduced and the grace periods should be extended. For the same reason, the use of passive systems can be envisaged. • The need of complementary and integrated deterministic and probabilistic approaches is reiterated. • Methodologies: Some of them are not yet applied. • Assessment of hazards would be a challenging aspect of next generation of NPP safety assessment and should be improved, which is confirmed by the first insights of Fukushima Daiichi TEPCO reactors accidents. • Provisions to cope with extreme events notably to improve the grace period before cliff-edge effects and thus allowing back-up measures to be implemented have to be defined and should be considered as hardened equipments

  13. The Atalante facility at CEA/Marcoule: towards Gen IV systems fuel cycle

    International Nuclear Information System (INIS)

    Bordier, Gilles; Warin, Dominique; Masson, Michel

    2008-01-01

    The Atalante facility is a complete set of 18 hot labs and 9 shielded cells devoted to the research and development on fuel cycle. The activities correspond to 4 major sectors of nuclear research: -) to support the operation of actual reprocessing plants with the aim of adapting the head of the process to the increase of the spend fuel burn-up and to different types of new burnt fuels to be reprocessed (including MOX, USi or UMo fuels); -) to develop the COEX TM process that jointly manages uranium and plutonium from the dissolution of spent fuel to the production of UPuO 2 powder and the fabrication of MOX fuel pellets; -) to prepare the recycling of minor actinides (MA) by partitioning or by grouped actinide extraction, and by MA bearing fuel fabrication; -) to study the long term behavior of high level waste conditioning matrices and especially self irradiation and leaching of vitrified waste. The first hot lab of Atalante was operated in 1992, the process shielded cell (CBP) in 2003 and the last LN1 lab in 2005, while at the same time a large scale demonstration test on the DIAMEX-SANEX MA partitioning process was performed. Now some new challenges involve further necessary evolutions of the facility. Some are related to safety assessment and operating flexibility; the major evolutions will come from new scientific goals and research programs. Furthermore, minor actinides materials irradiation tests in fast reactors will be prepared in the framework of a large international cooperation (GACID program) and need the production of significant amounts of MA bearing mixed U-Pu oxide compounds in new hot labs and shielded cells equipment. The major new research tools are presented and we highlight how Atalante is a unique facility which brings a real opportunity to reinforce the European and international scientific cooperation in order to prepare the next Gen IV fuel cycle. (authors)

  14. Gen IV International Forum - GIF, 2010 Annual Report

    International Nuclear Information System (INIS)

    Anon.

    2011-01-01

    The Generation IV International Forum (GIF), created in 2000 to foster international collaboration at a detailed level of actual R and D, is a cooperative international endeavor, organized to develop the research necessary to test the feasibility and performance capabilities of fourth generation nuclear systems, with the goal of making such systems deployable in large numbers around 2030. Since its beginning, GIF members stated the following goals for the fourth generation of nuclear power plants when compared to previous generations: a) improve sustainability (including effective fuel utilization and minimization of waste); b) improve economics (competitiveness with respect to other energy sources); c) improve safety and reliability (e.g. no need for offsite emergency response); and d) improve proliferation resistance and physical protection. After an in-depth analysis of the different available concepts, whatever their level of development, the Forum selected six concepts as the most promising, and decided to focus R and D on these systems: - the very-high-temperature reactor (VHTR); - the sodium-cooled fast reactor (SFR); - the supercritical-water-cooled reactor (SCWR); - the gas-cooled fast reactor (GFR); - the lead-cooled fast reactor (LFR); - the molten salt reactor (MSR). Active members of the GIF are Canada, Euratom, France, Japan, People's Republic of China, Republic of Korea, Republic of South Africa, Russian Federation, Switzerland and the United States. Altogether, they represent around 90% of the world installed nuclear capacity for producing electricity, and all key technology holders. The forum is led by the policy group, where all members are represented, and currently chaired by Japan since December 2009, assisted by vice-chairs from France and United States. The year 2010 has seen some important achievements and decisions regarding these six systems. For example, two sodium-cooled fast reactors (re)started this year: Monju in Japan restarted after

  15. Development of basic key technologies for Gen IV SFR

    International Nuclear Information System (INIS)

    Kim, Yeongil; Kim, Sungoh; Choi, Sukgi

    2012-04-01

    The advanced concepts, for the breakeven reactor(1,200MWe) and TRU burner(600MWe), were defined to satisfy the technology goals of Generation IV nuclear systems. Based on the advanced design concepts, a conceptual design of the demonstration SFR has been developed using the available licensing technology. The conceptual core design has been developed for the TRU burner in which an initial core is fueled with less than 20wt% enriched U235, and finally transformed to a self-recycled TRU core. The passive decay heat removal circuit ensuring reactor safety even in case of loss of emergency power has been developed and minimization of a reactor vessel and simplification of reactor internals have been conducted in the conceptual design. For development of advanced technologies, control logics for various power levels and the optimal design concept of heat exchanger applicable to supercritical CO 2 Brayton cycle as an energy conversion system was developed. A novel under-sodium waveguide sensor and a prototype under-sodium inspection system have been developed for under-sodium viewing of in-vessel structures submerged in an opaque liquid sodium. The fabrication technology of fuel slugs using the advanced fuel slug casting system was developed, and U-Zr alloy fuel rods were fabricated and examined. And a HT 9 cladding tube was manufactured using the developed cladding tube fabrication technology. For development of basic technologies, the cross section adjustment code ATCROSS and the MATRA-LMR code with HCFs have been developed to reduce core design uncertainties. The SIE ASME-NH computer program to evaluate high temperature structural design for 60 years design life, and the safety analysis code MARS-LMR with thermal-hydraulic and reactivity feedback models have been developed and validated. In addition, the sodium impurity measurement and control technology, the sodium water reaction event propagation model to predict the sodium leak propagation in a steam generator, and

  16. Safeguards Licensing Aspects of a Future Gen IV Test Facility - a Case Study

    International Nuclear Information System (INIS)

    Lindell, M. Aberg; Grape, S.; Hakansson, A.; Svaerd, S. Jacobsson

    2010-01-01

    The scope of this study covers safeguards licensing aspects of a possible future Gen IV demonstration facility. As a basis for the investigation, the facility was assumed to be located in Sweden, comprising a lead-cooled fast reactor and a reprocessing plant with fuel fabrication. The aim has been to identify safeguards requirements that may be set by the IAEA and the Swedish Radiation Safety Authority, and also to suggest how the safeguards system could be implemented in practice. The changed usage and handling of nuclear fuel, as compared to that of today, has been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. This work is part of GENIUS, the Swedish Gen IV research and development programme, which emphasizes lead-cooled fast reactors. (author)

  17. Thermal stability study for candidate stainless steels of GEN IV reactors

    Energy Technology Data Exchange (ETDEWEB)

    Simeg Veternikova, J., E-mail: jana.veternikova@stuba.sk [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Degmova, J. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pekarcikova, M. [Institute of Materials Science, Faculty of Materials Science and Technology, Slovak University of Technology, Paulinska 16, 917 24 Trnava (Slovakia); Simko, F. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia); Petriska, M. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Skarba, M. [Slovak University of Technology, Vazovova 5, 812 43 Bratislava (Slovakia); Mikula, P. [Institute of Nuclear and Physical Engineering, Faculty of Electrical and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Pupala, M. [Department of Molten Salts, Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dubravska cesta 9, 845 36 Bratislava (Slovakia)

    2016-11-30

    Highlights: • Thermal resistance of advanced stainless steels were observed at 1000 °C. • GEN IV candidate steels were confronted to classic AISI steels. • ODS AISI 316 has weaker thermal resistance than classic AISI steel. • Ferritic ODS steels and NF 709 has better thermal resistance than AISI steels. - Abstract: Candidate stainless steels for GEN IV reactors were investigated in term of thermal and corrosion stability at high temperatures. New austenitic steel (NF 709), austenitic ODS steel (ODS 316) and two ferritic ODS steels (MA 956 and MA 957) were exposed to around 1000 °C in inert argon atmosphere at pressure of ∼8 MPa. The steels were further studied in a light of vacancy defects presence by positron annihilation spectroscopy and their thermal resistance was confronted to classic AISI steels. The thermal strain supported a creation of oxide layers observed by scanning electron microscopy (SEM).

  18. Gen IV Materials Handbook Functionalities and Operation (2B) Handbook Version 2.0

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL

    2011-08-01

    This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

  19. Gen IV Materials Handbook Functionalities and Operation (4A) Handbook Version 4.0

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Weiju [ORNL

    2013-09-01

    This document is prepared for navigation and operation of the Gen IV Materials Handbook, with architecture description and new user access initiation instructions. Development rationale and history of the Handbook is summarized. The major development aspects, architecture, and design principles of the Handbook are briefly introduced to provide an overview of its past evolution and future prospects. Detailed instructions are given with examples for navigating the constructed Handbook components and using the main functionalities. Procedures are provided in a step-by-step fashion for Data Upload Managers to upload reports and data files, as well as for new users to initiate Handbook access.

  20. Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations

    International Nuclear Information System (INIS)

    Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Bruna, Giovanni; Hache, Georges; Repussard, Jacques

    2012-01-01

    The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

  1. Study on high temperature design methodology of heat-resistant materials for GEN-IV systems

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Seog; Kim, D. W.; Kim, S. H.; Kim, W. G.; Kim, J. H.; Park, D. G.; Yoon, J. H.; Lee, H. Y.; Hing, J. H

    2005-08-15

    Analysis of the existing high temperature design and assessment codes such as US(ASME-NH,Draft Code Case for Alloy 617), France(RCC-MR), UK(R5), Japan(BDS/DDS/FDS) for Gen IV reactor structure has been carried out. In addition the scope and fields for research and development is needed in the future have been defined. For assessing the high temperature creep cracks, time dependent fracture mechanics (TDFM) parameters of the C and Ct were analyzed. The creep propagation data were obtained from the creep crack growth tests for type 316LN stainless steels, and creep crack growth testing machine for Gen-IV system up to 950 .deg. C was set up. Damage mechanism and causes for creep-fatigue were investigated. The difference between prediction creep-fatigue life and experimental life were investigated. Material properties for analysis creep-fatigue damage were recommended. The assessment procedure (Draft) on creep-fatigue crack initiation has been developed based on the technical appendix A16 of French RCC-MR code. Ultrasonic wave signal against creep ruptured specimens of type 316LN stainless steel was obtained. It was identified that creep damage can be evaluated by ultrasonic method. The NDT techniques evaluated include Barkhausen noise, magnetic hysteresis parameters, positron annihilation, X-ray diffraction and small angle neutron scattering. Experimental procedure and evaluation method of material integrity were developed through the fracture toughness test of Cr-Mo steel.

  2. ASAMPSA2 project: Appliance of LWR PSA2 methodology to GEN IV reactors

    International Nuclear Information System (INIS)

    Bonneville, H.; Raimond, E.; Bassi, C.; Bertrand, F.; Serre, F.; Brinkman, J.L.; Burgazzi, L.; Jouve, S.; Vincon, L.; Polidoro, F.

    2012-01-01

    The European project ASAMPSA2 (Advanced Safety Assessment Methodology: level 2 PSA) of the 7. Framework Program aims at writing practical guidelines for conducting PSA level 2 studies on Light Water Reactors (PWR and BWR). The project includes also a supplemental task dealing with GEN IV reactors. Two main objectives are assigned to this task: 1) the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors with Generation IV concepts; 2) a brief survey of the modelling needs to describe the new features of GEN IV reactor concepts in terms of performing a level 2 PSA. Taking into account the ASAMPSA2 partners knowledge, the project has focused on four concepts: SFR, LFR, GFR and VHTR. For each of those concepts, a conceptual design was selected as reference: the European Fast Reactor (EFR) for sodium cooled fast reactors, the ELSY project for lead cooled fast reactor, the CEA GFR2400 project and the ANTARES project for VHTR. As a first stage, relevant data for each concept have been collected when available. These included: 1) basic general parameters and design characteristics relevant for safety studies with a specific attention given to passive devices; 2) information about former PSA2 studies on such concepts; 3) expert reviews about accident phenomenology knowledge (like PIRT); 4) list of computational tools developed or used for accident progression studies with, if possible, some basic information about the tools (availability, level of development, validation, documentation). In a second stage, the collected data were used to evaluate the compliance of the LWR guideline chapters with GEN IV concepts. The LWR guidelines may be divided into two main sections: chapters dealing with a specific phenomenon induced by core degradation and chapters dealing with general PSA methodology (like interface between PSA1 and PSA2, human risk assessment, system modelling and the role of expert opinion). The overall conclusion is that methodology

  3. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    Directory of Open Access Journals (Sweden)

    Lap-Yan Cheng

    2009-01-01

    Full Text Available The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR in a GEN IV direct-cycle gas-cooled fast reactor (GFR which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  4. Study on Structural Concept Design Improvement for a Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Joo, Young Sang; Lee, Hyeong Yeon; Kim, Jong Bum; Kim, Seok Hoon; Park, Chang Kyu; Koo, Gyeong Hoi

    2010-02-15

    An economic improvement is a hot issue as one of the Gen-IV nuclear plant goals. To secure economic competitiveness of a SFR compared to a pressurized water reactor, several structural design concepts are adapted in without loosing the reactor safety level. One is the increase of the plant capacity with the minimum number of component and loop, which leads the reduction of the plant maintenance, repair, and construction costs by a large-size scale effect. Another is the simple system arrangement, compact reactor size for only two loop system for a 1200MWe capacity of a pool type SFR, and the minimization of IHTS piping length through the properly locating the SG and secondary pump. Several researches are also studied to attain the economic improvement target of the NSSS in structural point of view; for example, an integrated concept of a refueling machine and inspection device with a long waveguide sensor for reactor internals, a high temperature LBB (leak before break) technology application, which can minimize the large protection facility against a large sodium leak like as a guillotine pipe break

  5. The ENEN-III project: Technical Training on the Concepts and Design of GEN IV nuclear reactors

    International Nuclear Information System (INIS)

    Berkvens, T.; Renault, C.; Alonso, M.; Salomaa, R.; Schönfelder, C.

    2013-01-01

    Some conclusions: • Not enough training courses to cover the LO’s: – Especially GEN IV; – Many introductory courses, little specific courses; – Reach out to other partners for more courses. • Skills and Attitudes: – Much more difficult to train/measure; – To be treated in a separate project. • Use of Learning Outcomes must be promoted; • Involvement of human resources necessary for the successful implementation of the schemes: – End of project workshop

  6. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yun, Sunghwan; Kim, Sang Ji

    2014-01-01

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a 23 Na(n,g) 24 Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B 4 C shielding inside the subassembly

  7. The safety R and D for GEN-IV reactors in the European nuclear energy technology platform strategic research agenda

    International Nuclear Information System (INIS)

    Bruna, G.

    2009-01-01

    In the fall 2007 EC launched the Sustainable Nuclear Energy Technology Platform (SNE-TP). The SNE-TP governing board set-up three working groups (WG): 1) Strategic Research Agenda (SRA) WG, in charge of drafting road-maps to support research, development and demonstration for current and future NPPs; 2) Deployment Strategy (DS) WG, in charge of defining the research road-map implementation and 3) Education, Training and Knowledge management (ETKM) WG, which was aimed at issuing proposal to reinforce European education and attract young in the nuclear field. The SRA WG was mandated to prepare the SRA vision document based on the preliminary road-map sketched in the document published by the Commission earlier in 2007. The SRA WG was originally organized in 5 sub-groups covering specific topics (1) GEN II and III, III+, including Advanced LWR, 2) Advanced Fuel Cycle for waste minimization and resource optimization; 3) GEN IV Fast Systems (SFR, LFR, GFR, ADS); 4) GEN IV (V) HTR and non-electricity-production applications; 5) New Nuclear Large Research Infrastructures) and 5 other sub-groups dealing with more generic cross-cutting research activities applicable to many specific topics, namely: 1) Structural material research; 2) modeling, simulation and methods, including physical data and tools and means for qualification and validation; 3) Reactor Safety, including severe accidents and human factor; 4) Advanced Driver and Minor Actinide Fuels: science and properties; 5) Pre-normative Research, Codes and Standards.The present paper is mainly aimed at summarizing the content of the SRA Safety sub-chapter focusing on GEN-IV aspects

  8. Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Koo, Gyeong Hoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm.

  9. Progress reports for Gen IV sodium fast reactor activities FY 2007

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Tentner, A. M.

    2007-01-01

    for prevention of progression into severe accident conditions (prevention of core melting) or for mitigation of severe accident consequences (mitigation of the impact of core melting to protect public health and safety). Because design measures for severe accident prevention and mitigation are beyond the normal design basis, established regulatory guidelines and codes do not provide explicit identification of the design performance requirements for severe accident accommodation. The treatment of severe accidents is one of the key issues of R and D plans for the Gen IV systems in general, and for the Sodium Fast Reactor (SFR) in particular. Despite the lack of an unambiguous definition of safety approach applicable for severe accidents, there is an emerging consensus on the need for their consideration for the design. The US SFR program and Argonne National Laboratory (ANL) in particular have actively studied the potential scenarios and consequences of Hypothetical Core Disruptive Accidents (HCDA) for SFRs with oxide fuel during the Fast Flux Test Facility (FFTF) and Clinch River Breeder Reactor Plant (CRBRP) programs in the 70s and 80s. Later, the focus of the US SFR safety R and D activities shifted to the prevention of all HCDAs through passive safety features of the SFRs with metal fuel in the Integral Fast Reactor (IFR) program, and the study of severe accident consequences was de-emphasized. The goal of this paper is to provide an overview of the current SFR safety approach and the role of severe accidents in Japan and France, in preparation for an expected and more active collaboration in this area between the US, Japan, and France

  10. Conversion of actinide solutions for the production of MA bearing fuels for Gen IV fast reactor systems

    International Nuclear Information System (INIS)

    Fernandez, A.; McGinley, J.; Somers, J.

    2008-01-01

    The conversion of the solution to solid for fuels containing minor actinides for accelerator driven systems or Gen IV fast reactors cannot be made by conventional ammonia or oxalate precipitation as is the case in today's reprocessing plant. The small particle size and concomitant dust that is produced in subsequent processing steps will not permit use of these processes on industrial scale. Innovation is needed to avoid dust generating powders, and indeed to simplify the processes themselves. Two such processing routes have been developed at the JRC-ITU. The sol gel route has been used to produce fuel containing Am and Np for the SUPERFACT, TRABANT and other irradiation experiments. The infiltration process has also been established and fuels have been produced for the FUTURIX and HELIOS experiments. (authors)

  11. Bolivian Rhinotragini IV: Paraeclipta gen. nov. (Coleoptera, Cerambycidae, new species and new combinations

    Directory of Open Access Journals (Sweden)

    Robin O. S. Clarke

    2011-01-01

    Full Text Available Paraeclipta gen. nov. is described to allocate five new species, and ten transferred from Eclipta Bates, 1873: P. cabrujai sp. nov.; P. clementecruzi sp. nov.; P. melgarae sp. nov.; P. tomhacketti sp. nov.; P. moscosoi sp. nov.; P. bicoloripes (Zajciw, 1965, comb. nov.; P. croceicornis (Gounelle, 1911, comb. nov.; P. flavipes (Melzer, 1922, comb. nov.; P. jejuna (Gounelle, 1911, comb. nov.; P. kawensis (Peñaherrera-Leiva & Tavakilian, 2004, comb. nov.; P. longipennis (Fisher, 1947, comb. nov.; P. rectipennis (Zajciw, 1965, comb. nov.; P. soumourouensis (Tavakilian & Peñaherrera-Leiva, 2003, comb. nov.; P. tenuis (Burmeister, 1865, comb. nov.; and P. unicoloripes (Zajciw, 1965, comb. nov. The Bolivian species are illustrated. A key to their identification and host flower records are provided.

  12. Reprocessing techniques of LWR spent fuel for reutilization in hybrid systems and IV generation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aruquipa, Wilmer; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany de P. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Since the era of nuclear technology begins, nuclear reactors have been produced spent fuel. This spent fuel contains material that could be recycle and reprocessed by different processes. All these processes aim to reduce the contribution to the final repository through the re-utilization of the nuclear material. Therefore, some new reprocessing options with non-proliferation characteristics have been proposed and the goal is to compare the different techniques used to maximize the effectiveness of the spent fuel utilization and to reduce the volume and long-term radiotoxicity of high-level waste by irradiation with neutron with high energy such as the ones created in hybrid reactors. In order to compare different recovery methods, the cross sections of fuels are calculated with de MCNP code, the first set consists of thorium-232 spiked with the reprocessed material and the second set in depleted uranium that containing 4.5% of U-235 spiked with the reprocessed material; These sets in turn are compared with the cross section of the UO{sub 2} in order to evaluate the efficiency of the reprocessed fuel as nuclear fuel. (author)

  13. Experimental observations to the electrical field for electrorefining of spent nuclear fuel in the Mark-IV electrorefiner

    International Nuclear Information System (INIS)

    Li, S. X.

    1998-01-01

    Experimental results from the pilot scale electrorefiner (Mark-IV ER) treating spent nuclear fuel are reported in this article. The electrorefining processes were carried out in a LiCl-KCl-UCl 3 electrolyte. It has been noted that spool of molten cadmium below the electrolyte plays an important role in the electrorefining operations. In addition, formations of electrical shorting path between anode baskets and the electrorefiner vessel were observed, which lessened the uranium dissolution process from anode baskets, however appeared to improve the morphology of cathode deposit. The FIDAP simulation code was used to calculate the electrical potential field distributions and the potential gradient near the cathode. The effect of the electrical shorting between anode baskets and electrorefiner vessel on the morphology of cathode products is discussed

  14. A Stochastic Proof of the Resonant Scattering Kernel and its Applications for Gen IV Reactors Type

    International Nuclear Information System (INIS)

    Becker, B.; Dagan, R.; Broeders, C.H.M.; Lohnert, G.

    2008-01-01

    Monte Carlo codes such as MCNP are widely accepted as almost-reference for reactor analysis. The Monte Carlo Code should therefore use as few as possible approximations in order to produce 'experimental-level' calculations. In this study we deal with one of the most problematic approximations done in MCNP in which the resonances are ignored for the secondary neutron energy distribution, namely the change of the energy and angular direction of the neutron after interaction with a heavy isotope with pronounced resonances. The endeavour of exploiting the influence of the resonances on the scattering kernel goes back to 1944 where E. Wigner and J. Wilkins developed the first temperature dependent scattering kernel. However only in 1998, the full analytical solution for the double differential resonant dependent scattering kernel was suggested by W. Rothenstein and R. Dagan. An independent stochastic approach is presented for the first time to confirm the above analytical kernel with a complete different methodology. Moreover, by manipulating in a subtle manner the scattering subroutine COLIDN of MCNP, it is proven that this very subroutine is, to some extent, inappropriate as well as the relevant explanation in the MCNP manual. The impact of this improved resonance dependent scattering kernel on diverse types of reactors, in particular for the Generation IV innovative core design HTR, is shown to be significant. (authors)

  15. Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  16. Summary of Structural Concept Development and High Temperature Structural Integrity Evaluation Technology for a Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Joo, Young Sang; Lee, Hyeong Yeon (and others)

    2008-04-15

    The economic improvement is a hot issue as one of Gen IV nuclear plant goals. It requires many researches and development works to meet the goal by securing the same level of plant safety. One of the key research items is the increase of the plant capacity with the minimum number of components and loops. Through the successful conceptual design experience for the KALIMER-600, the structural design study for a 1200MWe large capacity of sodium-cooled fast reactor has been performed to achieve the above plant size effects. The component number and reactor structural sizing were determined based on the core and fluid system design information. Several researches were performed to reduce the construction cost of NSSS in structural point of view, for example, a simplified component arrangement, concept proposals of integrated components, a high temperature LBB application technology, and an innovative in-service inspection (ISI) tool, and a computer program development of the ASME-NH design procedure of the class 1 structure and component under high temperature over 500 .deg. C. The IHTS piping arrangement was also proposed to minimize the length through the properly locating the SG and pump by 126m. Further studies of these concepts are required to confirm on the fabricability and the structural integrity for the operating and design loads. The proposed concepts will be optimized to a unified conceptual design through several trade-off studies.

  17. Enhanced radiation resistance through interface modification of nano-structured steels for Gen IV in-core applications

    International Nuclear Information System (INIS)

    Jang, Jinsung; Kang, Suk Hoon; Kim, Min Chul

    2013-06-01

    This project is to increase radiation tolerance of candidate alloys for Gen IV core component through the optimization of grain size and grain boundary characteristics. The focus is on nanocrystalline metal alloys with a fcc crystal structure. The long-term goal is to design and develop bulk nanostructured austenitic steels with enhanced void swelling resistance and substantial ductility, and to enhance their creep resistance at elevated temperatures via grain boundary engineering. An austenitic stainless steel, HT-UPS (high temperature ultra-fine precipitates strengthened) was developed at ORNL, and is expected to show enhanced void swelling resistance through the trapping of point defects at nanometer-sized carbides. Reducing the grain size and increasing the fraction-induced point defects (due to the increased sink area of the grain boundaries), to make bubble nucleation at the boundaries less likely (by reducing the fraction of high-energy boundaries), and to improve the strength and ductility under radiation by producing a higher density of nanometer sized carbides on the boundaries

  18. Status of the Gen-IV Proliferation Resistance and Physical Protection (PRPP) Evaluation Methodology

    International Nuclear Information System (INIS)

    Whitlock, J.; Bari, R.; Peterson, P.; Padoani, F.; Cojazzi, G.G.M.; Renda, G.; ); Cazalet, J.; Haas, E.; Hori, K.; Kawakubo, Y.; Chang, S.; Kim, H.; Kwon, E.-H.; Yoo, H.; Chebeskov, A.; Pshakin, G.; Pilat, J.F.; Therios, I.; Bertel, E.

    2015-01-01

    Methodologies have been developed within the Generation IV International Forum (GIF) to support the assessment and improvement of system performance in the areas safeguards, security, economics and safety. Of these four areas, safeguards and security are the subjects of the GIF working group on Proliferation Resistance and Physical Protection (PRPP). Since the PRPP methodology (now at Revision 6) represents a mature, generic, and comprehensive evaluation approach, and is freely available on the GIF public website, several non-GIF technical groups have chosen to utilize the PRPP methodology for their own goals. Indeed, the results of the evaluations performed with the methodology are intended for three types of generic users: system designers, programme policy makers, and external stakeholders. The PRPP Working Group developed the methodology through a series of demonstration and case studies. In addition, over the past few years various national and international groups have applied the methodology to inform nuclear energy system designs, as well as to support the development of approaches to advanced safeguards. A number of international workshops have also been held which have introduced the methodology to design groups and other stakeholders. In this paper we summarize the technical progress and accomplishments of the PRPP evaluation methodology, including applications outside GIF, and we outline the PRPP methodology's relationship with the IAEA's INPRO methodology. Current challenges with the efficient implementation of the methodology are outlined, along with our path forward for increasing its accessibility to a broader stakeholder audience - including supporting the next generation of skilled professionals in the nuclear non-proliferation field. (author)

  19. Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

    International Nuclear Information System (INIS)

    Stauff, N.

    2011-01-01

    Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

  20. Effects of NH₄⁺ on Ce(IV) electro-regeneration in simulated and real spent TFT-LCD Cr-etching solutions.

    Science.gov (United States)

    Chen, Te-San; Huang, Kuo-Lin; Lai, Yu-Chan; Kuo, Yi-Ming

    2012-08-15

    This investigation studies the electro-regeneration of Ce(IV) from Ce(III) in 4 M HNO(3) in the presence/absence of NH(4)(+) and real spent thin-film transistor liquid-crystal display (TFT-LCD) Cr-etching solutions. On Pt, at 2 A and 70 °C for 100 min, the Ce(IV) yield and apparent rate constant of Ce(III) oxidation in 4 M HNO(3) without NH(4)(+) were 100% and 5.54 × 10(-4) s(-1), respectively (and the activation energy was 13.1 kJ mol(-1)). Cyclic voltammetric and electrolytic measurements consistently support the noticeable inhibition by NH(4)(+) of Ce(III) oxidation and lowering of the Ce(IV) yield, respectively. The apparent diffusion coefficients for 0.2 and 0.02 M Ce(III) oxidation in 4 M HNO(3) that contained 0-0.6 M NH(4)(+) were (0.38-0.25) × 10(-5) and (1.6-0.9) × 10(-5) cm(2) s(-1), respectively. Because of combined effects of NH(4)(+) and anion impurities, the 100 min Ce(IV) yield of a real spent TFT-LCD Cr-etching solution (with [NH(4)(+)]/[Ce(III)] = 0.74 M/0.39 M) was 82%, lower than that of 4 M HNO(3) without NH(4)(+), but higher than those of 4 M HNO(3) that contained anion impurities with/without 0.4 M NH(4)(+). Copyright © 2012 Elsevier Ltd. All rights reserved.

  1. Gens Fera. The Wild men in the system of border decoration of the Bible of Wenceslas IV

    Czech Academy of Sciences Publication Activity Database

    Studničková, Milada

    2014-01-01

    Roč. 62, č. 3 (2014), s. 214-239 ISSN 0049-5123 R&D Projects: GA ČR GA13-39192S Institutional support: RVO:68378033 Keywords : book illumination * Bible of Wenceslas IV Subject RIV: AL - Art, Architecture, Cultural Heritage http://www.umeni-art.cz/admin/fileGet.aspx?v=issue-issue-2267-category-2268-paragraph-2269-pdf&l=cz

  2. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - IV: DUPIC Fuel Cycle Cost

    International Nuclear Information System (INIS)

    Ko, Won Il; Choi, Hangbok; Yang, Myung Seung

    2001-01-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.21 to 6.34 mills/kW.h for DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.07 to 0.27 mills/kW.h. Considering the uncertainty (0.40 to 0.44 mills/kW.h) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by ∼20% and reduce the spent fuel arising by ∼65% compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle is comparable with the once-through fuel cycle from the viewpoint of FCC. In the future, it should be important to consider factors such as the environmental benefit owing to natural uranium savings, the capability of reusing spent pressurized water reactor fuel, and the safeguardability of the fuel cycle when deciding on an advanced nuclear fuel cycle option

  3. A Qualitative Assessment Of Diversion Scenarios For A Example Sodium Fast Reactor Using The Gen IV PR And PP Methodology

    International Nuclear Information System (INIS)

    Zentner, Michael D.

    2008-01-01

    A working group was created in 2002 by the Generation IV International Forum (GIF) for the purpose of developing an internationally accepted methodology for assessing the Proliferation Resistance of a nuclear energy system (NES) and its individual elements. A two year case study is being performed by the experts group using this methodology to assess the proliferation resistance of a hypothetical NES called the Example Sodium Fast Reactor (ESFR). This work demonstrates how the PR and PP methodology can be used to provide important information at various levels of details to NES designers, safeguard administrators and decision makers. The study analyzes the response of the complete ESFR nuclear energy system to different proliferation and theft strategies. The challenges considered include concealed diversion, concealed misuse and 'break out' strategies. This paper describes the work done in performing a qualitative assessment of concealed diversion scenarios from the ESFR.

  4. ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

    International Nuclear Information System (INIS)

    Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

    2010-01-01

    The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

  5. Economic, energy and GHG emissions performance evaluation of a WhisperGen Mk IV Stirling engine μ-CHP unit in a domestic dwelling

    International Nuclear Information System (INIS)

    Conroy, G.; Duffy, A.; Ayompe, L.M.

    2014-01-01

    Highlights: • The performance of a Stirling engine MK IV micro-CHP unit was evaluated in a domestic dwelling in Ireland. • The performance of the micro-CHP was compare to that of a condensing gas boiler. • The micro-CHP unit resulted in an annual cost saving of €180 compared to the condensing gas boiler. • Electricity imported from the grid decreased by 20.8% while CO 2 emissions decreased by 16.1%. • The micro-CHP unit used 2889 kW h of gas more than the condensing gas boiler during one year of operation. - Abstract: This paper presents an assessment of the energy, economic and greenhouse gas emissions performances of a WhisperGen Mk IV Stirling engine μ-CHP unit for use in a conventional house in the Republic of Ireland. The energy performance data used in this study was obtained from a field trial carried out in Belfast, Northern Ireland during the period June 2004–July 2005 by Northern Ireland Electricity and Phoenix Gas working in collaboration with Whispertech UK. A comparative performance analysis between the μ-CHP unit and a condensing gas boiler revealed that the μ-CHP unit resulted in an annual cost saving of €180 with an incremental simple payback period of 13.8 years when compared to a condensing gas boiler. Electricity imported from the grid decreased by 20.8% while CO 2 emissions decreased by 16.1%. The μ-CHP unit used 2889 kW h of gas more than the condensing gas boiler

  6. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  7. Small angle neutron scattering study of nano sized microstructure in Fe-Cr ODS steels for gen IV in-core applications.

    Science.gov (United States)

    Han, Young-Soo; Mao, Xiadong; Jang, Jinsung

    2013-11-01

    The nano-sized microstructures in Fe-Cr oxide dispersion strengthened steel for Gen IV in-core applications were studied using small angle neutron scattering. The oxide dispersion strengthened steel was manufactured through hot isostatic pressing with various chemical compositions and fabrication conditions. Small angle neutron scattering experiments were performed using a 40 m small angle neutron scattering instrument at HANARO. Nano sized microstructures, namely, yttrium oxides and Cr-oxides were quantitatively analyzed by small angle neutron scattering. The yttrium oxides and Cr-oxides were also observed by transmission electron microscopy. The microstructural analysis results from small angle neutron scattering were compared with those obtained by transmission electron microscopy. The effects of the chemical compositions and fabrication conditions on the microstructure were investigated in relation to the quantitative microstructural analysis results obtained by small angle neutron scattering. The volume fraction of Y-oxide increases after fabrication, and this result is considered to be due to the formation of non-stochiometric Y-Ti-oxides.

  8. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  9. Study of molybdenum (VI) complexation and precipitation by zirconium (IV) in strongly acid medium. Application to nuclear spent fuel dissolution; Etude de la complexation et de la precipitation du molybdene (VI) par le zirconium (IV) en milieu tres acide. Application a la dissolution du combustible nucleaire irradie

    Energy Technology Data Exchange (ETDEWEB)

    Esbelin, E

    1999-07-01

    These last years the formation of solid deposits has been observed in the dissolution workshops of the La Hague plant. A sample of the solid was withdrawn for expertise: molybdenum and zirconium are the two major components of the solid, identified as zirconium molybdate. This thesis consisted in the approach of the mechanisms in solution liable to induce precipitate formation. After a bibliographical overview on the chemistry of Mo(VI) in highly acidic solution, this system was studied by absorption spectrophotometry in perchloric medium. The implication of two major forms of Mo(VI) in a dimerization equilibrium was confirmed by this way and by {sup 95}Mo NMR. The principal parameters governing this equilibrium were identified. It is thus shown that the molybdenum dimerization reaction is exothermic. Disturbance of the Mo(VI) system in highly acidic solution by Zr(IV) was also studied. In a restricted experimental field, for which 'conventional' exploitation methodologies had to be adapted to the system, a main complex of stoichiometry 1:1 between Mo(VI) and Zr(IV) was found. The precipitation study of Mo(VI) by Zr(IV) under conditions close to those of the dissolution medium of nuclear spent fuel was undertaken. The main parameters which control precipitation kinetics were identified. The results obtained reveal that precipitation is controlled by a single macroscopic process and therefore can be described by a single equation. The solid obtained is composed of only one phase presenting a Mo:Zr non-stoichiometry when compared to the theoretical formula ZrMo{sub 2}O{sub 7}(OH){sub 2},2H{sub 2}O. At last, on the basis of the research results, a descriptive mechanism of the system is proposed in which intervenes a 1:1 intermediate complex, much more soluble than a probable 2:1 precipitation precursor. (author)

  10. GenBank

    Data.gov (United States)

    U.S. Department of Health & Human Services — GenBank is the NIH genetic sequence database, an annotated collection of all publicly available DNA sequences. GenBank is designed to provide and encourage access...

  11. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Paul Y [Los Alamos National Laboratory

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  12. GenBank

    OpenAIRE

    Benson, Dennis A.; Karsch-Mizrachi, Ilene; Lipman, David J.; Ostell, James; Rapp, Barbara A.; Wheeler, David L.

    2002-01-01

    The GenBank sequence database incorporates publicly available DNA sequences of more than 105 000 different organisms, primarily through direct submission of sequence data from individual laboratories and large-scale sequencing projects. Most submissions are made using the BankIt (web) or Sequin programs and accession numbers are assigned by GenBank staff upon receipt. Data exchange with the EMBL Data Library and the DNA Data Bank of Japan helps ensure comprehensive worldwide coverage. GenBank...

  13. Materials for generation-IV nuclear reactors

    International Nuclear Information System (INIS)

    Alvarez, M. G.

    2009-01-01

    Materials science and materials development are key issues for the implementation of innovative reactor systems such as those defined in the framework of the Generation IV. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. The structural materials need to resist much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. For this reason, the first consideration in the development of Generation-IV concepts is selection and deployment of materials that operate successfully in the aggressive operating environments expected in the Gen-IV concepts. This paper summarizes the Gen-IV operating environments and describes the various candidate materials under consideration for use in different structural applications. (author)

  14. GenBank

    OpenAIRE

    Benson, Dennis A.; Karsch-Mizrachi, Ilene; Lipman, David J.; Ostell, James; Wheeler, David L.

    2006-01-01

    GenBank (R) is a comprehensive database that contains publicly available nucleotide sequences for more than 240 000 named organisms, obtained primarily through submissions from individual laboratories and batch submissions from large-scale sequencing projects. Most submissions are made using the web-based BankIt or standalone Sequin programs and accession numbers are assigned by GenBank staff upon receipt. Daily data exchange with the EMBL Data Library in Europe and the DNA Data Bank of Japan...

  15. GenBank

    OpenAIRE

    Benson, Dennis A.; Karsch-Mizrachi, Ilene; Lipman, David J.; Ostell, James; Sayers, Eric W.

    2008-01-01

    GenBank? is a comprehensive database that contains publicly available nucleotide sequences for more than 300 000 organisms named at the genus level or lower, obtained primarily through submissions from individual laboratories and batch submissions from large-scale sequencing projects. Most submissions are made using the web-based BankIt or standalone Sequin programs, and accession numbers are assigned by GenBank? staff upon receipt. Daily data exchange with the European Molecular Biology Labo...

  16. GenBank

    OpenAIRE

    Benson, Dennis A.; Cavanaugh, Mark; Clark, Karen; Karsch-Mizrachi, Ilene; Lipman, David J.; Ostell, James; Sayers, Eric W.

    2012-01-01

    GenBank? (http://www.ncbi.nlm.nih.gov) is a comprehensive database that contains publicly available nucleotide sequences for almost 260 000 formally described species. These sequences are obtained primarily through submissions from individual laboratories and batch submissions from large-scale sequencing projects, including whole-genome shotgun (WGS) and environmental sampling projects. Most submissions are made using the web-based BankIt or standalone Sequin programs, and GenBank staff assig...

  17. Asteroids IV

    Science.gov (United States)

    Michel, Patrick; DeMeo, Francesca E.; Bottke, William F.

    . Asteroids, like planets, are driven by a great variety of both dynamical and physical mechanisms. In fact, images sent back by space missions show a collection of small worlds whose characteristics seem designed to overthrow our preconceived notions. Given their wide range of sizes and surface compositions, it is clear that many formed in very different places and at different times within the solar nebula. These characteristics make them an exciting challenge for researchers who crave complex problems. The return of samples from these bodies may ultimately be needed to provide us with solutions. In the book Asteroids IV, the editors and authors have taken major strides in the long journey toward a much deeper understanding of our fascinating planetary ancestors. This book reviews major advances in 43 chapters that have been written and reviewed by a team of more than 200 international authorities in asteroids. It is aimed to be as comprehensive as possible while also remaining accessible to students and researchers who are interested in learning about these small but nonetheless important worlds. We hope this volume will serve as a leading reference on the topic of asteroids for the decade to come. We are deeply indebted to the many authors and referees for their tremendous efforts in helping us create Asteroids IV. We also thank the members of the Asteroids IV scientific organizing committee for helping us shape the structure and content of the book. The conference associated with the book, "Asteroids Comets Meteors 2014" held June 30-July 4, 2014, in Helsinki, Finland, did an outstanding job of demonstrating how much progress we have made in the field over the last decade. We are extremely grateful to our host Karri Muinonnen and his team. The editors are also grateful to the Asteroids IV production staff, namely Renée Dotson and her colleagues at the Lunar and Planetary Institute, for their efforts, their invaluable assistance, and their enthusiasm; they made life as

  18. Next Gen One Portal Usability Evaluation

    Science.gov (United States)

    Cross, E. V., III; Perera, J. S.; Hanson, A. M.; English, K.; Vu, L.; Amonette, W.

    2018-01-01

    Each exercise device on the International Space Station (ISS) has a unique, customized software system interface with unique layouts / hierarchy, and operational principles that require significant crew training. Furthermore, the software programs are not adaptable and provide no real-time feedback or motivation to enhance the exercise experience and/or prevent injuries. Additionally, the graphical user interfaces (GUI) of these systems present information through multiple layers resulting in difficulty navigating to the desired screens and functions. These limitations of current exercise device GUI's lead to increased crew time spent on initiating, loading, performing exercises, logging data and exiting the system. To address these limitations a Next Generation One Portal (NextGen One Portal) Crew Countermeasure System (CMS) was developed, which utilizes the latest industry guidelines in GUI designs to provide an intuitive ease of use approach (i.e., 80% of the functionality gained within 5-10 minutes of initial use without/limited formal training required). This is accomplished by providing a consistent interface using common software to reduce crew training, increase efficiency & user satisfaction while also reducing development & maintenance costs. Results from the usability evaluations showed the NextGen One Portal UI having greater efficiency, learnability, memorability, usability and overall user experience than the current Advanced Resistive Exercise Device (ARED) UI used by astronauts on ISS. Specifically, the design of the One-Portal UI as an app interface similar to those found on the Apple and Google's App Store, assisted many of the participants in grasping the concepts of the interface with minimum training. Although the NextGen One-Portal UI was shown to be an overall better interface, observations by the test facilitators noted specific exercise tasks appeared to have a significant impact on the NextGen One-Portal UI efficiency. Future updates to

  19. Spent fuels program

    International Nuclear Information System (INIS)

    Shappert, L.B.

    1983-01-01

    The goal of this task is to support the Domestic Spent Fuel Storage Program through studies involving the transport of spent fuel. A catalog was developed to provide authoritative, timely, and accessible transportation information for persons involved in the transport of irradiated reactor fuel. The catalog, drafted and submitted to the Transportation Technology Center, Sandia National Laboratories, for their review and approval, covers such topics as federal, state, and local regulations, spent fuel characteristics, cask characteristics, transportation costs, and emergency response information

  20. (IV) phosphates

    Indian Academy of Sciences (India)

    M(IV) phosphates of the class of tetravalent metal acid (TMA) salts where M (IV) = Zr, Ti, Sn has been synthesized by the sol-gel method. These materials have been characterized for elemental analysis (ICP-AES), thermal analysis (TGA, DSC), X-ray analysis and FTIR spectroscopy. Chemical resistivity of these materials ...

  1. DETEKSI GEN-GEN PENYANDI FAKTOR VIRULENSI PADA BAKTERI VIBRIO

    Directory of Open Access Journals (Sweden)

    Ince Ayu Khairani Kadriah

    2011-04-01

    menggunakan isolat bakteri yang diisolasi dari budidaya udang windu di berbagai daerah di Sulawesi Selatan dan Jawa. Pada penelitian ini digunakan primer spesifik untuk mendeteksi gen-gen virulen toxR gene, hemolysin (vvh gene, dan GyrB gene dengan metode PCR. Dari 35 isolat yang diisolasi, 20 isolat terdeteksi memiliki gen virulensi dan 8 di antaranya memiliki dua gen virulen. Spesies bakteri yang memiliki gen virulen adalah: V.harveyi, V. parahaemolyticus, V. mimicus, dan V. campbelli

  2. Spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2005-01-01

    When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that will not be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country. (author)

  3. Generation IV nuclear reactors: Current status and future prospects

    International Nuclear Information System (INIS)

    Locatelli, Giorgio; Mancini, Mauro; Todeschini, Nicola

    2013-01-01

    Generation IV nuclear power plants (GEN IV NPPs) are supposed to become, in many countries, an important source of base load power in the middle–long term (2030–2050). Nowadays there are many designs of these NPPs but for political, strategic and economic reasons only few of them will be deployed. International literature proposes many papers and reports dealing with GEN IV NPPs, but there is an evident difference in the types and structures of the information and a general unbiased overview is missing. This paper fills the gap, presenting the state-of-the-art for GEN IV NPPs technologies (VHTR, SFR, SCWR, GFR, LFR and MSR) providing a comprehensive literature review of the different designs, discussing the major R and D challenges and comparing them with other advanced technologies available for the middle- and long-term energy market. The result of this research shows that the possible applications for GEN IV technologies are wider than current NPPs. The economics of some GEN IV NPPs is similar to actual NPPs but the “carbon cost” for fossil-fired power plants would increase the relative valuation. However, GEN IV NPPs still require substantial R and D effort, preventing short-term commercial adoption. - Highlights: • Generation IV reactors are the middle–long term technology for nuclear energy. • This paper provides an overview and a taxonomy for the designs under consideration. • R and D efforts are in the material, heat exchangers, power conversion unit and fuel. • The life cycle costs are competitive with other innovative technologies. • The hydrogen economy will foster the development of Generation IV reactors

  4. The spent fuel fate

    International Nuclear Information System (INIS)

    2001-01-01

    The spent fuel is not a waste. It can be upgrade by a reprocessing which extracts all products able to produce energy. The today situation is presented and economically analyzed and future alternatives are discussed. (A.L.B.)

  5. Spent fuel workshop'2002

    International Nuclear Information System (INIS)

    Poinssot, Ch.

    2002-01-01

    This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO 2 fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO 2 dissolution determined from electrochemical experiments with 238 Pu doped UO 2 M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO 2 studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with α doped UO 2 in Boom clay conditions (K. Lemmens), Studies of the behavior of UO 2 / water interfaces under He 2+ beam (C. Corbel), Alpha and gamma radiolysis effects on UO 2 alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines (M. Kelm), On the potential catalytic behavior of

  6. Spent fuel centralized storage

    International Nuclear Information System (INIS)

    Chometon, P.L.

    1985-01-01

    Nuclear energy producer countries have felt the need to build a centralized spent fuel storage before reprocessing (for example, COGEMA in FRANCE), either in an adjoining plant on an appropriate site, or isolated. More rarely, this storage enables to decide whether to reprocess or to definitely store spent fuel considered as being waste: for example CLAB in Sweden. Our Company is specialized in the design and construction of spent fuel centralized storage plants. Storage generally takes place in a pool in order to facilitate handling operations and retrieving of these fuels, but these operations may also be effected in a dry way, either in concrete structures or in storage casks. With respect to pools, which might currently be the most appropriate and flexible system, several improvements have recently been made in the design of cask reception facilities and spent fuel storage. These improvements are presented, hereafter [fr

  7. Spent fuel centralized storage

    International Nuclear Information System (INIS)

    Baillif, L.; Chometon, P.L.

    1986-01-01

    Nuclear energy producer countries have felt the need to build a centralized spent fuel storage before reprocessing, either in an adjoining plant on an appropriate site, or isolated. More rarely, this storage enables to decide whether to reprocess or to definitely store spent fuel considered as being waste: for example CLAB in Sweden. Our Company SGN is specialized among others in the design and construction of spent fuel centralized storage plants. Storage generally takes place in a pool in order to facilitate handling operations and retrieving of these fuels, but these operations may also be effected in a dry way, either in concrete structures or in storage casks. With respect to pools, which might currently be the most appropriate and flexible system, several improvements have recently been made in the design of cask reception facilities and spent fuel storage. These improvements are presented, hereafter [fr

  8. Spent fuel management

    International Nuclear Information System (INIS)

    2005-01-01

    The production of nuclear electricity results in the generation of spent fuel that requires safe, secure and efficient management. Appropriate management of the resulting spent fuel is a key issue for the steady and sustainable growth of nuclear energy. Currently about 10,000 tonnes heavy metal (HM) of spent fuel are unloaded every year from nuclear power reactors worldwide, of which 8,500 t HM need to be stored (after accounting for reprocessed fuel). This is the largest continuous source of civil radioactive material generated, and needs to be managed appropriately. Member States have referred to storage periods of 100 years and even beyond, and as storage quantities and durations extend, new challenges arise in the institutional as well as in the technical area. The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs

  9. GenBank.

    OpenAIRE

    Benson, D; Lipman, D J; Ostell, J

    1993-01-01

    The GenBank sequence database has undergone an expansion in data coverage, annotation content and the development of new services for the scientific community. In addition to nucleotide sequences, data from the major protein sequence and structural databases, and from U.S. and European patents is now included in an integrated system. MEDLINE abstracts from published articles describing the sequences provide an important new source of biological annotation for sequence entries. In addition to ...

  10. Safety, Security And safeguards In GEN IV sodium fast reactors

    OpenAIRE

    Rossi, Fabiana

    2015-01-01

    This work presents first a study of the national and international laws in the fields of safety, security and safeguards. The international treaties and the recommendations issued by the IAEA as well as the national regulations in force in France, the United States and Italy are analyzed. As a result of this, a comparison among them is presented. Given the interest of the Japan Atomic Energy Agency for the aspects of criminal penalties and monetary, also the Japanese case is analyzed. The ...

  11. Disposal of spent fuel

    International Nuclear Information System (INIS)

    Blomeke, J.O.; Ferguson, D.E.; Croff, A.G.

    1978-01-01

    Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate materials for this container need to be studied. 3. It is highly likely that some ''filler'' material will be needed between the fuel elements and the can. 4. Leachability, stability, and waste-rock interaction studies should be carried out on the fuels. The major disadvantages of spent fuel as a disposal form are the lower maximum heat loading, 60 kW/acre versus 150 kW/acre for high-level waste from a reprocessing plant; the greater long-term potential hazard due to the larger quantities of plutonium and uranium introduced into a repository; and the possibility of criticality in case the repository is breached. The major advantages are the lower cost and increased near-term safety resulting from eliminating reprocessing and the treatment and handling of the wastes therefrom

  12. Spent fuel storage and isolation

    International Nuclear Information System (INIS)

    Bensky, M.S.; Kurzeka, W.J.; Bauer, A.A.; Carr, J.A.; Matthews, S.C.

    1979-02-01

    The principal spent fuel activities conducted within the commercial waste and spent fuel within the Commercial Waste and Spent Fuel Packaging Program are: simulated near-surface (drywell) storage demonstrations at Hanford and the Nevada Test Site; surface (sealed storage cask) and drywell demonstrations at the Nevada Test Site; and spent fuel receiving and packaging facility conceptual design. These investigations are described

  13. Spent Fuel in Chile

    International Nuclear Information System (INIS)

    López Lizana, F.

    2015-01-01

    The government has made a complete and serious study of many different aspects and possible road maps for nuclear electric power with strong emphasis on safety and energy independence. In the study, the chapter of SFM has not been a relevant issue at this early stage due to the fact that it has been left for later implementation stage. This paper deals with the options Chile might consider in managing its Spent Fuel taking into account foreign experience and factors related to safety, economics, public acceptance and possible novel approaches in spent fuel treatment. The country’s distinctiveness and past experience in this area taking into account that Chile has two research reactors which will have an influence in the design of the Spent Fuel option. (author)

  14. Spent fuel pyroprocessing demonstration

    International Nuclear Information System (INIS)

    McFarlane, L.F.; Lineberry, M.J.

    1995-01-01

    A major element of the shutdown of the US liquid metal reactor development program is managing the sodium-bonded spent metallic fuel from the Experimental Breeder Reactor-II to meet US environmental laws. Argonne National Laboratory has refurbished and equipped an existing hot cell facility for treating the spent fuel by a high-temperature electrochemical process commonly called pyroprocessing. Four products will be produced for storage and disposal. Two high-level waste forms will be produced and qualified for disposal of the fission and activation products. Uranium and transuranium alloys will be produced for storage pending a decision by the US Department of Energy on the fate of its plutonium and enriched uranium. Together these activities will demonstrate a unique electrochemical treatment technology for spent nuclear fuel. This technology potentially has significant economic and technical advantages over either conventional reprocessing or direct disposal as a high-level waste option

  15. Guidebook on spent fuel storage

    International Nuclear Information System (INIS)

    1984-01-01

    The Guidebook summarizes the experience and information in various areas related to spent fuel storage: technological aspects, the transport of spent fuel, economical, regulatory and institutional aspects, international safeguards, evaluation criteria for the selection of a specific spent fuel storage concept, international cooperation on spent fuel storage. The last part of the Guidebook presents specific problems on the spent fuel storage in the United Kingdom, Sweden, USSR, USA, Federal Republic of Germany and Switzerland

  16. Spent fuel reprocessing options

    International Nuclear Information System (INIS)

    2008-08-01

    The objective of this publication is to provide an update on the latest developments in nuclear reprocessing technologies in the light of new developments on the global nuclear scene. The background information on spent fuel reprocessing is provided in Section One. Substantial global growth of nuclear electricity generation is expected to occur during this century, in response to environmental issues and to assure the sustainability of the electrical energy supply in both industrial and less-developed countries. This growth carries with it an increasing responsibility to ensure that nuclear fuel cycle technologies are used only for peaceful purposes. In Section Two, an overview of the options for spent fuel reprocessing and their level of development are provided. A number of options exist for the treatment of spent fuel. Some, including those that avoid separation of a pure plutonium stream, are at an advanced level of technological maturity. These could be deployed in the next generation of industrial-scale reprocessing plants, while others (such as dry methods) are at a pilot scale, laboratory scale or conceptual stage of development. In Section Three, research and development in support of advanced reprocessing options is described. Next-generation spent fuel reprocessing plants are likely to be based on aqueous extraction processes that can be designed to a country specific set of spent fuel partitioning criteria for recycling of fissile materials to advanced light water reactors or fast spectrum reactors. The physical design of these plants must incorporate effective means for materials accountancy, safeguards and physical protection. Section four deals with issues and challenges related to spent fuel reprocessing. The spent fuel reprocessing options assessment of economics, proliferation resistance, and environmental impact are discussed. The importance of public acceptance for a reprocessing strategy is discussed. A review of modelling tools to support the

  17. Reprocessing of spent plasma

    International Nuclear Information System (INIS)

    Pierini, G.

    1981-01-01

    This invention relates to a process for removing helium and other impurities from a mixture containing deuterium and tritium, a deuterium/tritium mixture when purified in accordance with such a process and, more particularly, to a process for the reprocessing of spent plasma removed from a thermofusion reactor. (U.K.)

  18. Spent fuel counter

    International Nuclear Information System (INIS)

    Drayer, D.D.

    1988-09-01

    In many cases the IAEA must inspect spent fuel shipping casks before they leave facilities. Similarly, inspections may be required at the location where a cask is received and unloaded. In order to reduce the number of inspections required, it would be desirable to develop a system to count spent fuel assemblies as they are loaded or removed from shipping casks. This report discusses several methods which potentially could be used for performing this function. A concept for a Spent Fuel Counter System is proposed which uses a Laser Surveillance System (LASSY), Cerenkov Viewing Device (CVD), and Modular Integrated Video System (MIVS), all coupled together. In the proposed system, LASSY would provide an indication that an object is being placed into or removed from the cask, the CVD would be used to determine if the object has the radiation characteristics of a spent fuel assembly, and the MIVS would record the information. The system may need to be designed so that the operator could determine that it was operating correctly during the loading operations. This would help prevent anomalies from occurring which could only be resolved through reverification measures. Before such a system could be implemented testing would be necessary to determine that the individual components would each work adequately in this application. The issues of reliability, intrusiveness, and cask sealing should also be addressed before a development program is undertaken. 12 refs., 1 fig

  19. Time well spent

    DEFF Research Database (Denmark)

    Fallesen, Peter

    2013-01-01

    Individuals who spent time in foster care as children fare on average worse than non-placed peers in early adult life. Recent research on the effect of foster care placement on early adult life outcomes provides mixed evidence. Some studies suggest negative effects of foster care placement on early...

  20. Nordic Forum for Generation IV Reactors,Status and activities in 2012

    DEFF Research Database (Denmark)

    Van Nieuwenhove, Rudi; Lauritzen, Bent; Nonbøl, Erik

    The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risø, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many as-pects in Gen-IV reactor research and gave a good overview of the activi-ties within this fi...

  1. Generation-IV nuclear reactors, SFR concept

    International Nuclear Information System (INIS)

    Dufour, P.

    2010-01-01

    In this presentation author deals with development of sodium-cooled fast reactors and lead-cooled fast reactors. He concluded that: - SFR is a proved concept that has never achieved industrial deployment; - The GEN IV objectives need to reconsider the design of both the core and the reactor design : innovations are being analysed; Future design will benefit from considerable feedback of design, licensing, construction and operation of PX, SPX, etc.

  2. Encapsulating spent nuclear fuel

    International Nuclear Information System (INIS)

    Fleischer, L.R.; Gunasekaran, M.

    1979-01-01

    A system is described for encapsulating spent nuclear fuel discharged from nuclear reactors in the form of rods or multi-rod assemblies. The rods are completely and contiguously enclosed in concrete in which metallic fibres are incorporated to increase thermal conductivity and polymers to decrease fluid permeability. This technique provides the advantage of acceptable long-term stability for storage over the conventional underwater storage method. Examples are given of suitable concrete compositions. (UK)

  3. Storage of Spent Nuclear Fuel. Specific Safety Guide

    International Nuclear Information System (INIS)

    2012-01-01

    This Safety Guide provides recommendations and guidance on the storage of spent nuclear fuel. It covers all types of storage facilities and all types of spent fuel from nuclear power plants and research reactors. It takes into consideration the longer storage periods that have become necessary owing to delays in the development of disposal facilities and the decrease in reprocessing activities. It also considers developments associated with nuclear fuel, such as higher enrichment, mixed oxide fuels and higher burnup. The Safety Guide is not intended to cover the storage of spent fuel if this is part of the operation of a nuclear power plant or spent fuel reprocessing facility. Guidance is provided on all stages for spent fuel storage facilities, from planning through siting and design to operation and decommissioning, and in particular retrieval of spent fuel. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. Management system; 5. Safety case and safety assessment; 6. General safety considerations for storage of spent fuel. Appendix I: Specific safety considerations for wet or dry storage of spent fuel; Appendix II: Conditions for specific types of fuel and additional considerations; Annex: I: Short term and long term storage; Annex II: Operational and safety considerations for wet and dry spent fuel storage facilities; Annex III: Examples of sections of operating procedures for a spent fuel storage facility; Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex VI: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex VII: Postulated initiating events for consideration in a safety assessment (internal phenomena).

  4. Spent fuel transportation problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.A.

    1977-01-01

    In this paper, problems of transportation of nuclear spent fuel to reprocessing plants are discussed. The solutions proposed are directed toward the achievement of the transportation as economic and safe as possible. The increase of the nuclear power plants number in the USSR and the great distances between these plants and the reprocessing plants involve an intensification of the spent fuel transportation. Higher burnup and holdup time reduction cause the necessity of more bulky casks. In this connection, the economic problems become still more important. One of the ways of the problem solution is the development of rational and cheap cask designs. Also, the enforcement in the world of the environmental and personnel health protection requires to increase the transportation reliability and safety. The paper summarizes safe transportation rules with clarifying the following questions: the increase of the transport unit quantity of the spent fuel; rational shipment organization that minimizes vehicle turnover cycle duration; development of the reliable calculation methods to determine strength, thermal conditions and nuclear safety of transport packaging as applied to the vehicles of high capacity; maximum unification of vehicles, calculation methods and documents; and cask testing on models and in pilot scale on specific test rigs to assure that they meet the international safe fuel shipment rules. Besides, some considerations on the choice and use of structural materials for casks are given, and problems of manufacturing such casks from uranium and lead are considered, as well as problems of the development of fireproof shells, control instrumentation, vehicles decontamination, etc. All the problems are considered from the point of view of normal and accidental shipment conditions. Conclusions are presented [ru

  5. Nordic forum for generation IV reactors, status and activities in 2012

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.; Lauritzen, B.; Nonboel, E.

    2012-12-01

    The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risoe, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many aspects in Gen-IV reactor research and gave a good overview of the activities within this field at the various institutes and universities. The present report contains book of abstracts. The individual Power Point presentations are indexed in INIS and may be found at http://nordic-gen4.org/seminars/nordic-gen4-riso-2012-2/ (LN)

  6. Nordic forum for generation IV reactors, status and activities in 2012

    Energy Technology Data Exchange (ETDEWEB)

    Van Nieuwenhove, R. [Institutt for Energiteknikk, OECD Halden Reactor Project, Kjeller (Norway); Lauritzen, B.; Nonboel, E. [Technical Univ. of Denmark. DTU Nutech, Roskilde (Denmark)

    2012-12-15

    The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risoe, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many aspects in Gen-IV reactor research and gave a good overview of the activities within this field at the various institutes and universities. The present report contains book of abstracts. The individual Power Point presentations are indexed in INIS and may be found at http://nordic-gen4.org/seminars/nordic-gen4-riso-2012-2/ (LN)

  7. Spent fuel reprocessing method

    International Nuclear Information System (INIS)

    Shoji, Hirokazu; Mizuguchi, Koji; Kobayashi, Tsuguyuki.

    1996-01-01

    Spent oxide fuels containing oxides of uranium and transuranium elements are dismantled and sheared, then oxide fuels are reduced into metals of uranium and transuranium elements in a molten salt with or without mechanical removal of coatings. The reduced metals of uranium and transuranium elements and the molten salts are subjected to phase separation. From the metals of uranium and transuranium elements subjected to phase separation, uranium is separated to a solid cathode and transuranium elements are separated to a cadmium cathode by an electrolytic method. Molten salts deposited together with uranium to the solid cathode, and uranium and transuranium elements deposited to the cadmium cathode are distilled to remove deposited molten salts and cadmium. As a result, TRU oxides (solid) such as UO 2 , Pu 2 in spent fuels can be reduced to U and TRU by a high temperature metallurgical method not using an aqueous solution to separate them in the form of metal from other ingredients, and further, metal fuels can be obtained through an injection molding step depending on the purpose. (N.H.)

  8. Spent fuel interim storage

    International Nuclear Information System (INIS)

    Bilegan, Iosif C.

    2003-01-01

    The official inauguration of the spent fuel interim storage took place on Monday July 28, 2003 at Cernavoda NNP. The inaugural event was attended by local and central public authority representatives, a Canadian Government delegation as well as newsmen from local and central mass media and numerous specialists from Cernavoda NPP compound. Mr Andrei Grigorescu, State Secretary with the Economy and Commerce Ministry, underlined in his talk the importance of this objective for the continuous development of nuclear power in Romania as well as for Romania's complying with the EU practice in this field. Also the excellent collaboration between the Canadian contractor AECL and the Romanian partners Nuclear Montaj, CITON, UTI, General Concret in the accomplishment of this unit at the planned terms and costs. On behalf of Canadian delegation, spoke Minister Don Boudria. He underlined the importance which the Canadian Government affords to the cooperation with Romania aiming at specific objectives in the field of nuclear power such as the Cernavoda NPP Unit 2 and spent fuel interim storage. After traditional cutting of the inaugural ribbon by the two Ministers the festivities continued on the Cernavoda NPP Compound with undersigning the documents regarding the project completion and a press conference

  9. Decitabine (SuperGen).

    Science.gov (United States)

    Manoharan, A

    2000-12-01

    Decitabine, a potent DNA methyltransferase inhibitor, which was originally under development by Pharmachemie, is being developed by SuperGen. Pharmachemie had been studying decitabine in phase II clinical trials for several leukemia indications in Europe and the US. Preliminary results indicated that the compound was active in the treatment of myelodysplasia, relapsed leukemia, acute myeloid leukemia and postallogeneic progenitor cell transplant relapse. The compound is in phase II clinical trials with phase III trials scheduled to begin shortly. Decitabine has been used to treat myelodysplastic syndrome in a total of 125 patients, with an overall response rate of 49%. In a study using decitabine to treat chronic myelogenous leukemia in 81 patients, a response rate of 62% among patients in chronic phase of the disease was achieved. In a phase I/II trial designed to establish safety and efficacy in the treatment of sickle cell anemias treatment with decitabine generated a response in 100% of the patients tested: a total of eight patients were enrolled, each experienced elevated levels of fetal hemoglobin. Side effects were minimal and the drug was well tolerated. Plans for additional clinical studies of decitabine as a treatment for sickle cell anemia are underway. A phase II trial using a low dose of decitabine in patients with myelodysplastic syndrome has been completed. Of 66 patients entered, 62 were evaluable. The response rate was 48%, with a median response duration of 40 weeks. The mean survival from the start of therapy was 13 months. In a study with 37 CML patients, a 25% overall response rate was seen in those patients in the blastic phase of the disease, and a 52% response rate was observed in the accelerated phase patients. The most significant side effect was prolonged myelosuppression. The drug suppresses cellular growth in seven human tumor cell lines, possibly by reactivation of certain growth suppressor genes.

  10. Compacting spent fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    A method and apparatus for compacting spent fuel rods comprises transferring the rods from a nuclear fuel rod assembly into a different nuclear fuel rod container having a smaller cross section than the assembly. The individual rods are moved from a fuel assembly and through a transition funnel by movable grippers at opposite ends of the funnel. One movable gripper reciprocates between gripping and release positions in a gap between the fuel assembly and the transition funnel. All of the fuel rods are withdrawn concurrently and are merged towards one another into a tighter array within the transition funnel and emerge as a bundle. A movable and a stationary bundle gripper are provided between the funnel and the storage container to advance the bundle of fuel rods into the container. (author)

  11. Spent fuel transporting vessel

    International Nuclear Information System (INIS)

    Kumagaya, Takeshi.

    1995-01-01

    A large number of annular cooling fins are disposed each at an equal distance on the outer circumferential surface of a vessel main body. An electric power generation module is disposed on the surface of the cooling fins. The electric power generation module comprises a plurality of thermoelectric power generation elements. In each of the thermoelectric generation elements, the inner side thereof in contact with the surface of the cooling fin is at a high temperature while the outer side thereof is at a low temperature nearly equal with an atmospheric temperature. A predetermined amount of electric power is generated by seebeck effect due to the temperature difference. The electric power is always stored in a battery. Accordingly, even if a power generator of a ship should fail and power supply is stopped during transportation of the vessels for spent nuclear fuels, an appropriate amount of electric power can be supplied to a cooling device of the ship. (I.N.)

  12. Spent fuel storage at KURRI

    International Nuclear Information System (INIS)

    Nakagome, Y.; Fujita, Y.; Kanda, K.

    2004-01-01

    The Research Reactor Institute, Kyoto University (KURRI) has more than 200 MTR-type spent fuel elements stored in water pools. The longest pool residence time is 21 years at present. The integrity of spent fuel elements have been confirmed by a visual inspection and a sipping test. The spent fuel elements should be reprocessed in accordance with KURRI's policy. KURRI is now negotiating with a reprocessing plant to make a contract, as considering the consequences in U.S. (author)

  13. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Mineo, H.; Nomura, Y.; Sakamoto, K.

    1998-01-01

    In Japan 52 commercial nuclear power units are now operated, and the total power generation capacity is about 45 GWe. The cumulative amount of spent fuel arising is about 13,500 tU as of March 1997. Spent fuel is reprocessed, and recovered nuclear materials are to be recycled in LWRs and FBRs. In February 1997 short-term policy measures were announced by the Atomic Energy Commission, which addressed promotion of reprocessing programme in Rokkasho, plutonium utilization in LWRs, spent fuel management, backend measures and FBR development. With regard to the spent fuel management, the policy measures included expansion of spent fuel storage capacity at reactor sites and a study on spent fuel storage away from reactor sites, considering the increasing amount of spent fuel arising. Research and development on spent fuel storage has been carried out, particularly on dry storage technology. Fundamental studies are also conducted to implement the burnup credit into the criticality safety design of storage and transportation casks. Rokkasho reprocessing plant is being constructed towards its commencement in 2003, and Pu utilization in LWRs will be started in 1999. Research and development of future recycling technology are also continued for the establishment of nuclear fuel cycle based on FBRs and LWRs. (author)

  14. WWER spent fuel storage

    International Nuclear Information System (INIS)

    Bower, C.C.; Lettington, C.

    1994-01-01

    Selection criteria for PAKS NPP dry storage system are outlined. They include the following: fuel temperature in storage; sub-criticality assurance (avoidance of criticality for fuel in the unirradiated condition without having to take credit for burn-up); assurance of decay heat removal; dose uptake to the operators and public; protection of environment; volume of waste produced during operation and decommissioning; physical protection of stored irradiated fuel assemblies; IAEA safeguards assurance; storage system versus final disposal route; cost of construction and extent of technology transfer to Hungarian industry. Several available systems are evaluated against these criteria, and as a result the GEC ALSTHOM Modular Vault Dry Store (MVDS) system has been selected. The MVDS is a passively cooled dry storage facility. Its most important technical, safety, licensing and technology transfer characteristics are outlined. On the basis of the experience gained some key questions and considerations related to the East European perspective in the field of spent fuel storage are discussed. 8 figs

  15. Containing method for spent fuel and spent fuel containing vessel

    International Nuclear Information System (INIS)

    Maekawa, Hiromichi; Hanada, Yoshine.

    1996-01-01

    Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

  16. Mucopolysaccharidosis type IV

    Science.gov (United States)

    MPS IV; Morquio syndrome; Mucopolysaccharidosis type IVA; MPS IVA; Galactosamine-6-sulfatase deficiency; Mucopolysaccharidosis type IVB; MPS IVB; Beta galactosidase deficiency; Lysosomal storage disease - mucopolysaccharidosis type IV

  17. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    International Nuclear Information System (INIS)

    Kastenberg, William E.; Blandford, Edward; Kim, Lance

    2009-01-01

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public

  18. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    William E. Kastenberg; Edward Blandford; Lance Kim

    2009-03-31

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public.

  19. Disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste

  20. Disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste.

  1. FutureGen Project Report

    Energy Technology Data Exchange (ETDEWEB)

    Cabe, Jim; Elliott, Mike

    2010-09-30

    This report summarizes the comprehensive siting, permitting, engineering, design, and costing activities completed by the FutureGen Industrial Alliance, the Department of Energy, and associated supporting subcontractors to develop a first of a kind near zero emissions integrated gasification combined cycle power plant and carbon capture and storage project (IGCC-CCS). With the goal to design, build, and reliably operate the first IGCC-CCS facility, FutureGen would have been the lowest emitting pulverized coal power plant in the world, while providing a timely and relevant basis for coal combustion power plants deploying carbon capture in the future. The content of this report summarizes key findings and results of applicable project evaluations; modeling, design, and engineering assessments; cost estimate reports; and schedule and risk mitigation from initiation of the FutureGen project through final flow sheet analyses including capital and operating reports completed under DOE award DE-FE0000587. This project report necessarily builds upon previously completed siting, design, and development work executed under DOE award DE-FC26- 06NT4207 which included the siting process; environmental permitting, compliance, and mitigation under the National Environmental Policy Act; and development of conceptual and design basis documentation for the FutureGen plant. For completeness, the report includes as attachments the siting and design basis documents, as well as the source documentation for the following: • Site evaluation and selection process and environmental characterization • Underground Injection Control (UIC) Permit Application including well design and subsurface modeling • FutureGen IGCC-CCS Design Basis Document • Process evaluations and technology selection via Illinois Clean Coal Review Board Technical Report • Process flow diagrams and heat/material balance for slurry-fed gasifier configuration • Process flow diagrams and heat/material balance

  2. Nordic Nuclear Materials Forum for Generation IV Reactors

    International Nuclear Information System (INIS)

    Anghel, C.; Penttilae, S.

    2010-03-01

    A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

  3. Nordic Nuclear Materials Forum for Generation IV Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anghel, C. (Studsvik Nuclear AB, Nykoeping (Sweden)); Penttilae, S. (Technical Research Centre of Finland, VTT (Finland))

    2010-03-15

    A network for material issues for Generation IV nuclear power has been initiated within the Nordic countries. The objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) are to put the basis of a sustainable forum for Gen IV issues, especially focussing on fuels, cladding, structural materials and coolant interaction. Other issues include reactor physics, dynamics and diagnostics, core and fuel design. The present report summarizes the work performed during the year 2009. The efforts made include identification of organisations involved in Gen IV issues in the Nordic countries, update of the forum website, http://www.studsvik.se/GenerationIV, and investigation of capabilities for research within the area of Gen IV. Within the NOMAGE4 project a seminar on Generation IV Nuclear Energy Systems has been organized during 15-16th of October 2009. The aim of the seminar was to provide a forum for exchange of information, discussion on future research needs and networking of experts on Generation IV reactor concepts. As an outcome of the NOMAGE4, a few collaboration project proposals have been prepared/planned in 2009. The network was welcomed by the European Commission and was mentioned as an exemplary network with representatives from industries, universities, power companies and research institutes. NOMAGE4 has been invited to participate to the 'European Energy Research Alliance, EERA, workshop for nuclear structural materials' http://www.eera-set.eu/index.php?index=41 as external observers. Future plans include a new Nordic application for continuation of NOMAGE4 network. (author)

  4. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gonzalez, J.L.

    2002-01-01

    The spent fuel management strategy in Spain is presented. The strategy includes temporary solutions and plans for final disposal. The need for R and D including partitioning and transmutation, as well as the financial constraints are also addressed. (author)

  5. Intermodal transportation of spent fuel

    International Nuclear Information System (INIS)

    Elder, H.K.

    1983-09-01

    Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate

  6. Current status of NPP generation IV

    International Nuclear Information System (INIS)

    Yohanes Dwi Anggoro; Dharu Dewi; Nurlaila; Arief Tris Yuliyanto

    2013-01-01

    Today development of nuclear technology has reached the stage of research and development of Generation IV nuclear power plants (advanced reactor systems) which is an innovative development from the previous generation of nuclear power plants. There are six types of power generation IV reactors, namely: Very High Temperature Reactor (VHTR), Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), and Super Critical Water-cooled Reactor (SCWR). The purpose of this study is to know the development of Generation IV nuclear power plants that have been done by the thirteen countries that are members of the Gen IV International Forum (GIF). The method used is review study and refers to various studies related to the current status of research and development of generation IV nuclear power. The result of this study showed that the systems and technology on Generation IV nuclear power plants offer significant advances in sustainability, safety and reliability, economics, and proliferation resistance and physical protection. In addition, based on the research and development experience is estimated that: SFR can be used optimally in 2015, VHTR in 2020, while NPP types GFR, LFR, MSR, and SCWR in 2025. Utilization of NPP generation IV said to be optimal if fulfill the goal of NPP generation IV, such as: capable to generate energy sustainability and promote long-term availability of nuclear fuel, minimize nuclear waste and reduce the long term stewardship burden, has an advantage in the field of safety and reliability compared to the previous generation of NPP and VHTR technology have a good prospects in Indonesia. (author)

  7. IAEA spent fuel storage glossary

    International Nuclear Information System (INIS)

    1985-10-01

    The aim of this glossary is to provide a basis for improved international understanding of terms used in the important area of spent fuel storage technology. The glossary is the product of an IAEA Consultant Group with valuable input from a substantial list of reviewers. The glossary emphasizes fuel storage relevant to power reactors, but is also widely applicable to research reactors. The intention is to define terms from current technologies. Terms are limited to those directly related to spent fuel storage

  8. Spent-fuel-storage alternatives

    International Nuclear Information System (INIS)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed

  9. GenLab, Laboratorio Virtual de Genética

    Directory of Open Access Journals (Sweden)

    Fidel Ramírez

    2000-07-01

    Full Text Available GenLab es el nombre que tiene el software diseñado por nosotros, en el cual se modela el proceso meiótico y la fecundación en organismos diploides. El objetivo de esta aplicación es ilustrar el resultado de un cruce determinado, tratando de ser lo más ajustados a la realidad. La modelación de la reproducción sexual se realiza internamente y el GenLab se limita a presentar los resultados según el número de descendencia seleccionado para un cruce específico, esto significa que se puede escoger una gran cantidad de características para los parentales y se puede estudiar la frecuencia de estos en la descendencia. El modelo cuenta con base de datos donde están almacenados algunos de los locus de Drosophila melanogaster junto con su ubicación en centimorgans 1. EI propósito de este modelo es servir como herramienta pedagógica  y didáctica tanto en universidades como en colegios, facilitando el aprendizaje de algunos principios básicos de la genética, por lo cual puede ser usado si se cuenta con una conexión a Internet y un navegador visitando http://biologia.unal.edu.co/fidel.

  10. Taxonomic dissection of the genus Micrococcus: Kocuria gen. nov., Nesterenkonia gen. nov., Kytococcus gen. nov., Dermacoccus gen. nov., and Micrococcus Cohn 1872 gen. emend.

    Science.gov (United States)

    Stackebrandt, E; Koch, C; Gvozdiak, O; Schumann, P

    1995-10-01

    The results of a phylogenetic and chemotaxonomic analysis of the genus Micrococcus indicated that it is significantly heterogeneous. Except for Micrococcus lylae, no species groups phylogenetically with the type species of the genus, Micrococcus luteus. The other members of the genus form three separate phylogenetic lines which on the basis of chemotaxonomic properties can be assigned to four genera. These genera are the genus Kocuria gen. nov. for Micrococcus roseus, Micrococcus varians, and Micrococcus kristinae, described as Kocuria rosea comb. nov., Kocuria varians comb. nov., and Kocuria kristinae comb. nov., respectively; the genus Nesterenkonia gen. nov. for Micrococcus halobius, described as Nesterenkonia halobia comb. nov.; the genus Nesterenkonia gen. nov. for Micrococcus halobius, described as Nesterenkonia halobia comb. nov.; the genus Dermacoccus gen. nov. for Micrococcus nishinomiyaensis, described as Dermacoccus nishinomiyaensis comb. nov.; and the genus Kytocossus gen. nov. for Micrococcus sedentarius, described as Kytococcus sedentarius comb. nov. M. luteus and M. lylae, which are closely related phylogenetically but differ in some chemotaxonomic properties, are the only species that remain in the genus Micrococcus Cohn 1872. An emended description of the genus Micrococcus is given [corrected].

  11. Transportation of spent nuclear fuels

    International Nuclear Information System (INIS)

    Meguro, Toshiichi

    1976-01-01

    The spent nuclear fuel taken out of reactors is cooled in the cooling pool in each power station for a definite time, then transported to a reprocessing plant. At present, there is no reprocessing plant in Japan, therefore the spent nuclear fuel is shipped abroad. In this paper, the experiences and the present situation in Japan are described on the transport of the spent nuclear fuel from light water reactors, centering around the works in Tsuruga Power Station, Japan Atomic Power Co. The spent nuclear fuel in Tsuruga Power Station was first transported in Apr. 1973, and since then, about 36 tons were shipped to Britain by 5 times of transport. The reprocessing plant in Japan is expected to start operation in Apr. 1977, accordingly the spent nuclear fuel used for the trial will be transported in Japan in the latter half of this year. Among the permission and approval required for the transport of spent nuclear fuel, the acquisition of the certificate for transport casks and the approval of land and sea transports are main tasks. The relevant laws are the law concerning the regulations of nuclear raw material, nuclear fuel and reactors and the law concerning the safety of ships. The casks used in Tsuruga Power Station and EXL III type, and the charging of spent nuclear fuel, the decontamination of the casks, the leak test, land transport with a self-running vehicle, loading on board an exclusive carrier and sea transport are briefly explained. The casks and the ship for domestic transport are being prepared. (Kato, I.)

  12. Intermodal transfer of spent fuel

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Weiner, R.F.

    1991-01-01

    As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handier exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. A study of the movement of spent fuel casks through ports, including the loading and unloading of containers from cargo vessels, afforded an opportunity to estimate the radiation doses to those individuals handling the spent fuels with doses to the public along subsequent transportation routes of the fuel. A number of states require redundant inspections and for escorts over long distances on highways; thus handlers, inspectors, escort personnel, and others who are not normally classified as radiation workers may sustain doses high enough to warrant concern about occupational safety. This paper addresses the question of radiation safety for these workers. Data were obtained during, observation of the offloading of reactor spent fuel (research reactor spent fuel, in this instance) which included estimates of exposure times and distances for handlers, inspectors and other workers during offloading and overnight storage. Exposure times and distance were also for other workers, including crane operators, scale operators, security personnel and truck drivers. RADTRAN calculational models and parameter values then facilitated estimation of the dose to workers during incident-free ship-to-truck transfer of spent fuel

  13. Container for spent fuel assembly

    International Nuclear Information System (INIS)

    Sawai, Takeshi.

    1996-01-01

    The container of the present invention comprises a container main body having a body portion which can contain spent fuel assemblies and a lid, and heat pipes having an evaporation portion disposed along the outer surface of the spent fuel assemblies to be contained and a condensation portion exposed to the outside of the container main body. Further, the heat pipe is formed spirally at the evaporation portions so as to surround the outer circumference of the spent fuel assemblies, branched into a plurality of portions at the condensation portion, each of the branched portion of the condensation portion being exposed to the outside of the container main body, and is tightly in contact with the periphery of the slit portions disposed to the container main body. Then, since released after heat is transferred to the outside of the container main body from the evaporation portion of the heat pipe along the outer surface of the spent fuel assemblies by way of the condensation portion of the heat pipes exposed to the outside of the container main body, the efficiency of the heat transfer is extremely improved to enhance the effect of removing heat of spent fuel assemblies. Further, cooling effect is enhanced by the spiral form of the evaporation portion and the branched condensation portion. (N.H.)

  14. Spent Nuclear Fuel project, project management plan

    International Nuclear Information System (INIS)

    Fuquay, B.J.

    1995-01-01

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  15. IV treatment at home

    Science.gov (United States)

    ... Other IV treatments you may receive after you leave the hospital include: Treatment for hormone deficiencies Medicines for severe nausea that cancer chemotherapy or pregnancy may cause Patient-controlled analgesia (PCA) for pain (this is IV ...

  16. Creep in generation IV nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Rissanen, L. (VTT Technical Research Centre of Finland, Espoo (Finland))

    2010-05-15

    Nuclear power has an important role in fulfilling the world's growing energy needs and reducing the carbon dioxide emission. Six new, innovative nuclear energy systems have been identified and selected for further development by the international Generation Four International Forum (GIF). These generation four (Gen IV) nuclear energy systems include a variety of reactor, energy conversion and fuel cycle technologies. The successful development and deployment of these largely depend on the performance and reliability of the available structural materials. These potential materials need to sustain their mechanical properties up to high temperatures, high neutron doses and corrosive environments of the new or enhanced types of coolants. Current knowledge on material properties, material-coolant interaction and especially material degradation processes in these new environments are limited. This paper gives an overview of the Gen IV material issues with special emphasis on European design of supercritical light water reactor concept high performance light water reactor (HPLWR). The challenges for the structural materials and the components most likely to suffer from creep and creep-irradiation are highlighted. Some results from relatively short term creep testing in supercritical water are presented for AISI 316NG, 347H and 1.4970 steels. The 1.4970 steel was superior in creep and oxidation resistance (orig.)

  17. Regular Advisory Group on Spent Fuel Management

    International Nuclear Information System (INIS)

    1993-01-01

    The Regular Advisory Group on Spent Fuel Management (RAGSFM) was established in accordance with the recommendations of the Expert Group on International Spent Fuel Management in 1982. The Advisory Group consists of nominated experts from countries with considerable experience and/or requirements in such aspects of the back-end of the fuel cycle as storage, safety, transportation and treatment of spent fuel. The RAGSFM activities cover the following main topics: a) Analysis and summary of spent fuel arisings and storage facilities; b) Interface between spent fuel storage and transportation activities; c) Spent fuel storage process and technology and related safety issues; d)Treatment of spent fuel

  18. Spent Pot Lining Characterization Framework

    Science.gov (United States)

    Ospina, Gustavo; Hassan, Mohamed I.

    2017-09-01

    Spent pot lining (SPL) management represents a major concern for aluminum smelters. There are two key elements for spent pot lining management: recycling and safe storage. Spent pot lining waste can potentially have beneficial uses in co-firing in cement plants. Also, safe storage of SPL is of utmost importance. Gas generation of SPL reaction with water and ignition sensitivity must be studied. However, determining the feasibility of SPL co-firing and developing the required procedures for safe storage rely on determining experimentally all the necessary SPL properties along with the appropriate test methods, recognized by emissions standards and fire safety design codes. The applicable regulations and relevant SPL properties for this purpose are presented along with the corresponding test methods.

  19. Spent fuel management in Japan

    International Nuclear Information System (INIS)

    Shirahashi, K.; Maeda, M.; Nakai, T.

    1996-01-01

    Japan has scarce energy resources and depends on foreign resources for 84% of its energy needs. Therefore, Japan has made efforts to utilize nuclear power as a key energy source since mid-1950's. Today, the nuclear energy produced from 49 nuclear power plants is responsible for about 31% of Japan's total electricity supply. The cumulative amount of spent fuel generated as of March 1995 was about 11,600 Mg U. Japan's policy of spent fuel management is to reprocess spent nuclear fuel and recycle recovered plutonium and uranium as nuclear fuel. The Tokai reprocessing plant continues stable operation keeping the annual treatment capacity or around 90 Mg U. A commercial reprocessing plant is under construction at Rokkasho, northern part of Japan. Although FBR is the principal reactor to use plutonium, LWR will be a major power source for some time and recycling of the fuel in LWRs will be prompted. (author). 3 figs

  20. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H.

    1997-12-01

    Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  1. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H

    1997-12-01

    Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  2. Spent-fuel-storage alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  3. Spent fuel management in Spain

    International Nuclear Information System (INIS)

    Gago, J.A.; Gravalos, J.M.

    1996-01-01

    There are presently nine Light Water Reactors in operation, representing around a 34% of the overall electricity production. In the early years, a small amount of spent fuel was sent to be reprocessed, although this policy was cancelled in favor of the open cycle option. A state owned company, ENRESA, was created in 1984, which was given the mandate to manage all kinds of radioactive wastes generated in the country. Under the present scenario, a rough overall amount of 7000 tU of spent fuel will be produced during the lifetime of the plants, which will go into final disposal. (author)

  4. Spent nuclear fuel transport problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.I.

    1977-01-01

    The paper considers the problems of shipping spent fuel from nuclear power stations to reprocessing plants and also the principal ways of solving these problems with a view to achieving maximum economy and safety in transport. The increase in the number of nuclear power plants in the USSR will entail an intensification of spent-fuel shipments. Higher burnup and the need to reduce cooling time call for heavier and more complex shipping containers. The problem of shipping spent fuel should be tackled comprehensively, bearing in mind the requirements of safety and economy. One solution to these problems is to develop rational and cheap designs of such containers. In addition, the world-wide trend towards more thorough protection of the environment against pollution and of the health of the population requires the devotion of constant attention to improving the reliability and safety of shipments. The paper considers the prospects for nuclear power development in the USSR and in other member countries of the CMEA (1976-1980), the composition and design of some Soviet packaging assemblies, the appropriate cooling time for spent fuel from thermal reactor power stations, procedures for reducing fuel-shipping costs, some methodological problems of container calculation and design, and finally problems of testing and checking containers on test rigs. (author)

  5. Characteristics of spent nuclear fuel

    International Nuclear Information System (INIS)

    Notz, K.J.

    1988-04-01

    The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will, or may, eventually be disposed of in a geological repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. This report deals with spent fuels, but for completeness, the other sources are described briefly. Detailed characterizations are required for all of these potential repository wastes. These characteristics include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. In addition, the present inventories and projected quantities of the various wastes are needed. This information has been assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. 5 refs., 3 figs., 4 tabs

  6. Worldwide spent fuel transportation logistics

    International Nuclear Information System (INIS)

    Best, R.E.; Garrison, R.F.

    1978-01-01

    This paper presents an overview of the worldwide transportation requirements for spent fuel. Included are estimates of numbers and types of shipments by mode and cask type for 1985 and the year 2000. In addition, projected capital and transportation costs are presented. For the year 1977 and prior years inclusive, there is a cumulative worldwide requirement for approximately 300 MTU of spent fuel storage at away-from-reactor (AFR) facilities. The cumulative requirements for years through 1985 are projected to be nearly 10,000 MTU, and for the years through 2000 the requirements are conservatively expected to exceed 60,000 MTU. These AFR requirements may be related directly to spent fuel transportation requirements. In total nearly 77,000 total cask shipments of spent fuel will be required between 1977 and 2000. These shipments will include truck, rail, and intermodal moves with many ocean and coastal water shipments. A limited number of shipments by air may also occur. The US fraction of these is expected to include 39,000 truck shipments and 14,000 rail shipments. European shipments to regional facilities are expected to be primarily by rail or water mode and are projected to account for 16,000 moves. Pacific basin shipments will account for 4500 moves. The remaining are from other regions. Over 400 casks will be needed to meet the transportation demands. Capital investment is expected to reach $800,000,000 in 1977 dollars. Cumulative transport costs will be a staggering $4.4 billion dollars

  7. Algoritmos genéticos

    Directory of Open Access Journals (Sweden)

    José Jesús Martínez Páez

    1998-10-01

    Full Text Available Esta técnica se basa en el concepto de evolución a través de selección de los mejores individuos, y de los operadores genéticos de selección, reproducción y mutación. Se trata entonces, de definir un espacio de soluciones para el problema que se quiere solucionar, en una cadena de bits. A esto se le conoce como la codificación del cromosoma, donde cada bit, denominado gen  tiene cierto significado especial. Inicialmente el algoritmo genera al azar muchas de estas cadenas o seres, es decir, una población, que luego confronta can un ambiente, que es el problema solucionar o función que se quiere optimizar. De esta confrontación  o evaluación a que se somete cada ser. Se obtiene información sobre cómo se comporto cada uno. A través de métodos aleatorios, pero con probabilidad de selección proporcional a su comportamiento, es decir, a mejor comportamiento mayor probabilidad, se selecciona una nueva población de seres supuestamente mejores que la generación anterior.

  8. Metals removal from spent salts

    Science.gov (United States)

    Hsu, Peter C.; Von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Brummond, William A.; Adamson, Martyn G.

    2002-01-01

    A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

  9. Actinide removal from spent salts

    Science.gov (United States)

    Hsu, Peter C.; von Holtz, Erica H.; Hipple, David L.; Summers, Leslie J.; Adamson, Martyn G.

    2002-01-01

    A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration salt solutions that contain less than 0.1 ppm of thorium or uranium.

  10. Thermal analysis of cold vacuum drying of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Piepho, M.G.

    1998-07-20

    The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4).

  11. TMI-2 spent fuel shipping

    International Nuclear Information System (INIS)

    Quinn, G.J.; Burton, H.M.

    1985-01-01

    TMI-2 failed fuel will be shipped to the Idaho National Engineering Laboratory for use in the DOE Core Examination Program. The fuel debris will be loaded into three types of canisters during defueling and dry loaded into a spent fuel shipping cask. The cask design accommodates seven canisters per cask and has two separate containment vessels with ''leaktight'' seals. Shipments are expectd to begin in early 1986

  12. Spent nuclear fuel in Bulgaria

    International Nuclear Information System (INIS)

    Peev, P.; Kalimanov, N.

    1999-01-01

    The development of the nuclear energy sector in Bulgaria is characterized by two major stages. The first stage consisted of providing a scientific basis for the programme for development of the nuclear energy sector in the country and was completed with the construction of an experimental water-water reactor. At present, spent nuclear fuel from this reactor is placed in a water filled storage facility and will be transported back to Russia. The second stage consisted of the construction of the 6 NPP units at the Kozloduy site. The spent nuclear fuel from the six units is stored in at reactor pools and in an additional on-site storage facility which is nearly full. In order to engage the government of the country with the on-site storage problems, the new management of the National Electric Company elaborated a policy on nuclear fuel cycle and radioactive waste management. The underlying policy is de facto the selection of the 'deferred decision' option for its spent fuel management. (author)

  13. Reuse of Hydrotreating Spent Catalyst

    International Nuclear Information System (INIS)

    Habib, A.M.; Menoufy, M.F.; Amhed, S.H.

    2004-01-01

    All hydro treating catalysts used in petroleum refining processes gradually lose activity through coking, poisoning by metal, sulfur or halides or lose surface area from sintering at high process temperatures. Waste hydrotreating catalyst, which have been used in re-refining of waste lube oil at Alexandria Petroleum Company (after 5 years lifetime) compared with the same fresh catalyst were used in the present work. Studies are conducted on partial extraction of the active metals of spent catalyst (Mo and Ni) using three leaching solvents,4% oxidized oxalic acid, 10% aqueous sodium hydroxide and 10% citric acid. The leaching experiments are conducting on the de coked extrude [un crushed] spent catalyst samples. These steps are carried out in order to rejuvenate the spent catalyst to be reused in other reactions. The results indicated that 4% oxidized oxalic acid leaching solution gave total metal removal 45.6 for de coked catalyst samples while NaOH gave 35% and citric acid gave 31.9 % The oxidized leaching agent was the most efficient leaching solvent to facilitate the metal removal, and the rejuvenated catalyst was characterized by the unchanged crystalline phase The rejuvenated catalyst was applied for hydrodesulfurization (HDS) of vacuum gas oil as a feedstock, under different hydrogen pressure 20-80 bar in order to compare its HDS activity

  14. Spent Fuel Working Group Report

    International Nuclear Information System (INIS)

    O'Toole, T.

    1993-11-01

    The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety. To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary's initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group's Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities

  15. Genève Reconnaissante

    CERN Multimedia

    2001-01-01

    Robert Cailliau (centre), with Geneva's Mayor Alain Vaissade (left) and Jean Erhardt, Secretary General of the Administrative Council of Geneva (right). Geneva recognised the contribution of two CERN people to the reputation of the city last Tuesday when Mayor Alain Vaissade presented the Genève Reconaissante Medal to Tim Berners-Lee and Robert Cailliau. Berners-Lee, who was not able to be present in person, invented the World Wide Web at CERN just over a decade ago, while Cailliau was his first collaborator. Quoting Cailliau, Vaissade said that whilst there is no doubt that something like the Web would have appeared sooner or later, the fact that it happened at CERN, in Geneva, was no accident. Both the Laboratory and the city are places where people from around the world meet and work in harmony.

  16. Stark Broadening of Se IV, Sn IV, Sb IV and Te IV Spectral Lines

    Directory of Open Access Journals (Sweden)

    Milan S. Dimitrijević

    2018-03-01

    Full Text Available Stark broadening parameters, line width and shift, are needed for investigations, analysis and modelling of astrophysical, laboratory, laser produced and technological plasmas. Especially in astrophysics, due to constantly increasing resolution of satellite borne spectrographs, and large terrestrial telescopes, data on trace elements, which were previously insignificant, now have increasing importance. Using the modified semiempirical method of Dimitrijević and Konjević, here, Stark widths have been calculated for 2 Se IV, 6 Sn IV, 2 Sb IV and 1 Te IV transitions. Results have been compared with existing theoretical data for Sn IV. Obtained results will be implemented in the STARK-B database, which is also a part of Virtual atomic and molecular data center (VAMDC.

  17. Near surface spent fuel storage: environmental issues

    International Nuclear Information System (INIS)

    Nelson, I.C.; Shipler, D.B.; McKee, R.W.; Glenn, R.D.

    1979-01-01

    Interim storage of spent fuel appears inevitable because of the lack of reprocessing plants and spent fuel repositories. This paper examines the environmental issues potentially associated with management of spent fuel before disposal or reprocessing in a reference scenario. The radiological impacts of spent fuel storage are limited to low-level releases of noble gases and iodine. Water needed for water basin storage of spent fuel and transportation accidents are considered; the need to minimize the distance travelled is pointed out. Resource commitments for construction of the storage facilities are analyzed

  18. Overview of spent fuel management and problems

    International Nuclear Information System (INIS)

    Ritchie, I.G.; Ernst, P.C.

    1998-01-01

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel worldwide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialized and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. Some projections of spent fuel inventories to the year 2006 are presented and discussed. (author)

  19. On-ward observations in neonatal intensive care : Towards safer supplemental oxygen & IV therapy

    NARCIS (Netherlands)

    Van der Eijk, A.C.

    2012-01-01

    The focus of this thesis is on supplemental oxygen therapy and intrave nous (IV) therapy in neonatal intensive care. Both therapies are essential, but potentially dangerous for (preterm) newborn infants. Supplemental oxygen therapy refers to the therapy where a gas mixture with >21% of oxy gen is

  20. Probability of spent fuel transportation accidents

    International Nuclear Information System (INIS)

    McClure, J.D.

    1981-07-01

    The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10 -7 spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10 -9 /mile

  1. HTGR spent fuel storage study

    International Nuclear Information System (INIS)

    Burgoyne, R.M.; Holder, N.D.

    1979-04-01

    This report documents a study of alternate methods of storing high-temperature gas-cooled reactor (HTGR) spent fuel. General requirements and design considerations are defined for a storage facility integral to a fuel recycle plant. Requirements for stand-alone storage are briefly considered. Three alternate water-cooled storage conceptual designs (plug well, portable well, and monolith) are considered and compared to a previous air-cooled design. A concept using portable storage wells in racks appears to be the most favorable, subject to seismic analysis and economic evaluation verification

  2. The solidification of spent resin

    International Nuclear Information System (INIS)

    Shiao, S. J.; Tsai, C. M.; Shyu, Y. H.

    1991-01-01

    A quasi-steady apparatus was applied to measure the thermal conductivity of solids ranging in size for 0.3 to 200 L, and temperature distributions in the solids were recorded during the curing, and theoretical equation for conduction in a cylindrical form with uniform energy generation was established to define the thermal state of reaction. The heat of reaction calculated from the theoretical equation with experimental values for the maximum temperature and thermal conductivity agrees very well with the data reported. The relationships among heat of reaction and amount of curing agent, retardant, loading of spent resin, and water were established

  3. Reduction of uranium in disposal conditions of spent nuclear fuel

    International Nuclear Information System (INIS)

    Myllykylae, E.

    2008-02-01

    This literature study is a summary of publications, in which the reduction of uranium by iron has been investigated in anaerobic groundwater conditions or in aqueous solution in general. The basics of the reduction phenomena and the oxidation states, complexes and solubilities of uranium and iron in groundwaters are discussed as an introduction to the subject, as well as, the Finnish disposal concept of spent nuclear fuel. The spent fuel itself mainly (∼96 %) consists of a sparingly soluble uranium(IV) dioxide, UO 2 (s), which is stable phase in the anticipated reducing disposal conditions. If spent fuel gets in contact with groundwater, oxidizing conditions might be induced by the radiolysis of water, or by the intrusion of oxidizing glacial melting water. Under these conditions, the oxidation and dissolution of uranium dioxide to more soluble U(VI) species could occur. This could lead to the mobilization of uranium and other components of spent fuel matrix including fission products and transuranium elements. The reduction of uranium back to oxidation state U(IV) can be considered as a favourable immobilization mechanism in a long-term, leading to precipitation due to the low solubility of U(IV) species. The cast iron insert of the disposal canister and its anaerobic corrosion products are the most important reductants under disposal conditions, but dissolved ferrous iron may also function as reductant. Other iron sources in the buffer or near-field rock, are also considered as possible reductants. The reduction of uranium is a very challenging phenomenon to investigate. The experimental studies need e.g. well-controlled anoxic conditions and measurements of oxidation states. Reduction and other simultaneous phenomena are difficult to distinghuish. The groundwater conditions (pH, Eh and ions) influence on the prevailing complexes of U and Fe and on forming corrosion products of iron and, thus they determine also the redox chemistry. The partial reduction of

  4. Sealed can of spent fuel

    International Nuclear Information System (INIS)

    Suzuki, Yasuyuki.

    1976-01-01

    Object: To provide a seal plug cover with a gripping portion fitted to a canning machine and a gripping portion fitted to a gripper of the same configuration as a fuel body for handling the fuel body so as to facilitate the handling work. Structure: A sealed can comprises a vessel and a seal plug cover, said cover being substantially in the form of a bottomed cylinder, which is slipped on the vessel and air-tightly secured by a fastening bolt between it and a flange. The spent fuel body is received into the vessel together with coolant during the step of canning operation. Said seal plug cover has two gripping portions, one for opening and closing the plug cover of the canning machine as an exclusive use member, the other being in the form of a hook-shaped peripheral groove, whereby the gripping portions may be effectively used using the same gripper when the spent fuel body is transported while being received in the sealed can or when the fuel body is removed from the sealed can. (Kawakami, Y.)

  5. Intermodal transfer of spent fuel

    International Nuclear Information System (INIS)

    Neuhauser, K.S.; Weiner, R.F.

    1993-01-01

    This paper discusses RADTRAN calculational models and parameter values for describing dose to workers during incident-free ship-to-truck transfer of spent fuel. Data obtained during observation of the offloading of research reactor spent fuel at Newport News Terminal in the Port of Hampton Roads, Virginia, are described. These data include estimates of exposure times and distances for handlers, inspectors, and other workers during offloading and overnight storage. Other workers include crane operators, scale operators, security personnel, and truck drivers. The data are compared to the default data in RADTRAN 4, and the latter are found to be conservative. The casks were loaded under IAEA supervision at their point of origin, and three separate radiological inspections of each cask were performed at the entry to the port (Hampton Roads) by the U.S. Coast Guard, the state of Virginia, and the shipping firm. As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handler exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. (author)

  6. NNDSS - Table IV. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table IV. Tuberculosis - 2016.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  7. NNDSS - Table IV. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table IV. Tuberculosis - 2014.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  8. NNDSS - Table IV. Tuberculosis

    Data.gov (United States)

    U.S. Department of Health & Human Services — NNDSS - Table IV. Tuberculosis - 2015.This Table includes total number of cases reported in the United States, by region and by states, in accordance with the...

  9. SAGE IV Pathfinder

    Data.gov (United States)

    National Aeronautics and Space Administration — Utilizing a unique, new occultation technique involving imaging, the SAGE IV concept will meet or exceed the quality of previous SAGE measurements at a small...

  10. Microscopic Examination of a Corrosion Front in Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    J.A. Fortner; A.J. Kropf; R.J. Finch; J.C. Cunnane

    2006-01-01

    Spent uranium oxide nuclear fuel hosts a variety of trace chemical constituents, many of which must be sequestered from the biosphere during fuel storage and disposal. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within spent uranium oxide nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the plutonium and neptunium in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO 2 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl U(VI)O 2 2+ mineral assemblage that is depleted in plutonium and neptunium relative to the parent fuel. At the corrosion front interface between intact fuel and the uranyl-mineral corrosion layer, we find evidence of a thin (∼20 micrometer) layer that is enriched in plutonium and neptunium within a predominantly U 4+ environment. Available data for the standard reduction potentials for NpO 2+ /Np 4+ and UO 2 2+ /U 4+ couples indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potentials of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. Neptunium is an important radionuclide in dose contribution according to performance assessment models of the proposed U. S. repository at Yucca Mountain, Nevada. A scientific understanding of how the UO 2 matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of neptunium is needed to predict its behavior at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel [1]. In the immediate vicinity of the spent fuel's surface the redox and nucleation behavior is likely to promote

  11. The quality of silage of corn grain and spent P. ostreatus mushroom substrate

    Directory of Open Access Journals (Sweden)

    Adamović Milan J.

    2007-01-01

    Full Text Available The chemical composition, fermentation quality, mycological and mycotoxicological analyses of silage mixture, made of ground corn grain and spent P. ostreatus mushroom substrate, were investigated in this paper. Dry matter content in high moisture ground corn, at the time of ensiling was 70%, and in the spent substrate (on the Salt Cedar wood shaving basis was 52.7%. Corn grain to spent substrate ratio in trials was: 100:0% (I, 90:10% (II, 80:20% (III and 70:30% (IV respectively. Content of the lignocellulose fractions in silage was slightly increased, and protein content was slightly decreased with the increase of spent substrate content. Contents of the VFA (volatile fatty acids in silage, pH value, and NH3-N content were for the silage of very good quality. In the spent substrate 9 mold species were found, from which the most frequent were genus Penicillium, Paecilomyces variotii, and Trichoderma harzianum. In ground corn grain silage (I presence of the yeasts was dominant (90.000/g. In combined trials (II-IV only Penicillium (P. brevicompactum and P. echinulatum mold species were found. Presence of molds and yeasts in investigated trials was within tolerated values for ensiled feedstuffs. Mycotoxin presence in silage was not determined.

  12. Spent fuel. Dissolution and oxidation

    International Nuclear Information System (INIS)

    Grambow, B.

    1989-03-01

    Data from studies of the low temperature air oxidation of spent fuel were retrieved in order to provide a basis for comparison between the mechanism of oxidation in air and corrosion in water. U 3 O 7 is formed by diffusion of oxygen into the UO 2 lattice. A diffusion coefficient of oxygen in the fuel matric was calculated for 25 degree C to be in the range of 10 -23 to 10 -25 m 2 /s. The initial rates of U release from spent fuel and from UO 2 appear to be similar. The lowest rates (at 25 degree c >10 -4 g/(m 2 d)) were observed under reducing conditions. Under oxidizing conditions the rates depend mainly of the nature and concentraion of the oxidant and/or on corbonate. In contact with air, typical initial rates at room temperature were in the range between 0.001 and 0.1 g/(m 2 d). A study of apparent U solubility under oxidizing conditions was performed and it was suggested that the controlling factor is the redox potential at the UO 2 surface rather than the E h of the bulk solution. Electrochemical arguments were used to predict that at saturation, the surface potential will eventually reach a value given by the boundaries at either the U 3 O 7 /U 3 O 8 or the U 3 O 7 /schoepite stability field, and a comparison with spent fuel leach data showed that the solution concentration of uranium is close to the calculated U solubility at the U 3 O 7 /U 3 O 8 boundary. The difference in the cumulative Sr and U release was calculated from data from Studsvik laboratory. The results reveal that the rate of Sr release decreases with the square root of time under U-saturated conditions. This time dependence may be rationalized either by grain boundary diffusion or by diffusion into the fuel matrix. Hence, there seems to be a possibility of an agreement between the Sr release data, structural information and data for oxygen diffusion in UO 2 . (G.B.)

  13. Spent fuel shipping cask accident evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Fields, S.R.

    1975-12-01

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel.

  14. Unleashing Gen Y: Marketing Mars to Millennials

    Science.gov (United States)

    Leahy, Bart D.; Hidalgo, Loretta; Kloberdanz, Cassie

    2007-01-01

    Space advocates need to engage Generation Y (born 1977-1999).This outreach is necessary to recruit the next generation of scientists and engineers to explore Mars. Space advocates in the non-profit, private, and government sectors need to use a combination of technical communication, marketing, and politics, to develop messages that resonate with Gen Y. Until now, space messages have been generated by and for college-educated white males; Gen Y is much more diverse, including as much as one third minorities. Young women, too, need to be reached. My research has shown that messages emphasizing technology, fun, humor, and opportunity are the best means of reaching the Gen Y audience of 60 million (US population is 300 million). The important things space advocates must avoid are talking down to this generation, making false promises, or expecting them to "wait their turn" before they can participate. This is the MTV generation! We need to find ways of engaging Gen Y now to build a future where human beings can live and work on the planet Mars. In addition to the messages themselves, advocates need to keep up with Gen Y' s social networking and use of iPods, cell phones, and the Internet. NASA and space advocacy groups can use these tools for "viral marketing," where young people share targeted space-related information via cell phones or the Internet because they like it. Overall, Gen Y is a socially dynamic and media-savvy group; advocates' space messages need to be sincere, creative, and placed in locations where Gen Y lives. Mars messages must be memorable!

  15. Spent fuels transportation coming from Australia

    International Nuclear Information System (INIS)

    2002-01-01

    Maritime transportation of spent fuels from Australia to France fits into the contract between COGEMA and ANSTO, signed in 1999. This document proposes nine information cards in this domain: HIFAR a key tool of the nuclear, scientific and technological australian program; a presentation of the ANSTO Australian Nuclear Science and Technology Organization; the HIFAR spent fuel management problem; the COGEMA expertise in favor of the research reactor spent fuel; the spent fuel reprocessing at La Hague; the transports management; the transport safety (2 cards); the regulatory framework of the transports. (A.L.B.)

  16. Thermal model of spent fuel transport cask

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.; Sultan, G.F.; Khalil, E.E.

    1996-01-01

    The investigation provides a theoretical model to represent the thermal behaviour of the spent fuel elements when transported in a dry shipping cask under normal transport conditions. The heat transfer process in the spent fuel elements and within the cask are modeled which include the radiant heat transfer within the cask and the heat transfer by thermal conduction within the spent fuel element. The model considers the net radiant method for radiant heat transfer process from the inner most heated element to the surrounding spent elements. The heat conduction through fuel interior, fuel-clad interface and on clad surface are also presented. (author) 6 figs., 9 refs

  17. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

    Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

  18. Three new anascosporic genera of the Saccharomycotina: Danielozyma gen. nov., Deakozyma gen. nov. and Middelhovenomyces gen. nov.

    Science.gov (United States)

    Kurtzman, Cletus P; Robnett, Christie J

    2014-05-01

    Three new non-ascosporic, ascomycetous yeast genera are proposed based on their isolation from currently described species and genera. Phylogenetic placement of the genera was determined from analysis of nuclear gene sequences for D1/D2 large subunit rRNA, small subunit rRNA, translation elongation factor-1α and RNA polymerase II, subunits B1 and B2. The new taxa are: Deakozyma gen. nov., type species Deakozyma indianensis sp. nov. (type strain NRRL YB-1937, CBS 12903); Danielozyma gen. nov., type species Danielozyma ontarioensis comb. nov. (type strain NRRL YB-1246, CBS 8502); D. litseae comb. nov. (type strain NRRL YB-3246, CBS 8799); Middelhovenomyces gen. nov., type species Middelhovenomyces tepae comb. nov. (type strain NRRL Y-17670, CBS 5115) and M. petrohuensis comb. nov. (type strain NRRL Y-17663, CBS 8173).

  19. IV access in dental practice.

    LENUS (Irish Health Repository)

    Fitzpatrick, J J

    2009-04-01

    Intravenous (IV) access is a valuable skill for dental practitioners in emergency situations and in IV sedation. However, many people feel some apprehension about performing this procedure. This article explains the basic principles behind IV access, and the relevant anatomy and physiology, as well as giving a step-by-step guide to placing an IV cannula.

  20. Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

    International Nuclear Information System (INIS)

    Black, J.E.; King, F.D.

    1979-01-01

    Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

  1. Description of Guyruita gen. nov. and two new species (Ischnocolinae, Theraphosidae Descrição de Guyruita gen. nov. e duas novas espécies (Ischnocolinae, Theraphosidae

    Directory of Open Access Journals (Sweden)

    José P.L. Guadanucci

    2007-12-01

    Full Text Available The genus Guyruita gen. nov. and two new species from Brazil are described. Holothele waikoshiemi (Bertani & Araújo, 2005 from Venezuela is transferred here to the new genus. Guyruita gen. nov. differs from the remaining Ischnocolinae by the following features: labium densely occupied by a lot of cuspules (more than 100, intercheliceral intumescence absent, posterior sternal sigilla remote from margin, tarsal claws without teeth, tarsal scopula I-II undivided (tarsus II with a line of sparse setae, which does not divide the scopula, III-IV divided.É descrito o gênero Guyruita gen. nov. e duas espécies novas do Brasil. Holothele waikoshiemi (Bertani & Araújo, 2005 da Venezuela é transferido para o novo gênero. Guyruita gen. nov. difere dos outros Ischnocolinae pelas seguintes caracterísicas: lábio densamente ocupado por muitas cúspides (mais de 100, tumescência interqueliceral ausente, sigilla esternal posterior distante da margem, unhas tarsais sem dentes, escópula tarsal I e II inteiras (tarso II com uma fileira de cerdas esparsas, as quais não dividem a escópula, III e IV divididas.

  2. Genética dos transtornos afetivos

    Directory of Open Access Journals (Sweden)

    Ivanor Velloso Meira Lima

    2004-01-01

    Full Text Available Os autores revisam neste artigo o conjunto de evidências genético-epidemiológicas que indicam a presença de fatores genéticos na vulnerabilidade para os transtornos afetivos. Apresentam também os dados obtidos até o momento, através de estratégias de genética molecular na busca de genes de susceptibilidade para o transtorno afetivo bipolar e para a depressão.In this article the authors review the body of genetic-epidemiological evidences for the role of genes in mood disorders. Current molecular genetic studies searching for susceptibility genes for bipolar affective disorder and depression are also presented.

  3. Research reactor spent fuel in Ukraine

    International Nuclear Information System (INIS)

    Trofimenko, A.P.

    1996-01-01

    This paper describes the research reactors in Ukraine, their spent fuel facilities and spent fuel management problems. Nuclear sciences, technology and industry are highly developed in Ukraine. There are 5 NPPs in the country with 14 operating reactors which have total power capacity of 12,800 MW

  4. Prospects of spent management in Spain

    International Nuclear Information System (INIS)

    Melches, C.; Ramirez, E.; Selgas, F.; Cabanilles, P.A.; Lopez Perez, B.; Uriarte, A.

    1978-01-01

    The purpose of this paper is to outline the forecast on spent fuel management in Spain, taking into account the international developments produced during the last years and specially on LWR fuels. This forecast is based on the following actions: increase of the storage capacity in the reactors: construction of an independent spent fuel storage installation (ISFSI) and a fuel reprocessing pilot plant. (author)

  5. Spent fuel storage for ISER plant

    International Nuclear Information System (INIS)

    Nakajima, Takasuke; Kimura, Yuzi

    1987-01-01

    ISER is an intrinsically safe reactor basing its safety only on physical laws, and uses a steel reactor vessel in order to be economical. For such a new type reactor, it is essentially important to be accepted by the society by showing that the reactor is more profitable than conventional reactors to the public in both technical and economic viewpoint. It is also important that the reactor raises no serious problem in the total fuel cycle. Reprocessing seems one of the major worldwide fuel cycle issues. Spent fuel storage is also one of the key technologies for fuel cycle back end. Various systems for ISER spent fuel storages are examined in the present report. Spent fuel specifications of ISER are similar to those of LWR and therefore, most of LWR spent fuel technologies are basically applicable to ISER spent fuel. Design requirements and examples of storage facilities are also discussed. Dry storage seems to be preferable for the relatively long cooling time spent fuel like ISER's one from economical viewpoint. Vault storage will possibly be the most advantageous for large storage capacity. Another point for discussion is the location and international collaboration for spent fuel storages: ISER expected to be a worldwide energy source and therefore, international spent fuel management seems to be fairly attractive way for an energy recipient country. (Nogami, K.)

  6. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  7. Spent Nuclear Fuel (SNF) Project Execution Plan

    International Nuclear Information System (INIS)

    LEROY, P.G.

    2000-01-01

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities

  8. Costing of spent nuclear fuel storage

    International Nuclear Information System (INIS)

    2009-01-01

    This report deals with economic analysis and cost estimation, based on exploration of relevant issues, including a survey of analytical tools for assessment and updated information on the market and financial issues associated with spent fuel storage. The development of new storage technologies and changes in some of the circumstances affecting the costs of spent fuel storage are also incorporated. This report aims to provide comprehensive information on spent fuel storage costs to engineers and nuclear professionals as well as other stakeholders in the nuclear industry. This report is meant to provide informative guidance on economic aspects involved in selecting a spent fuel storage system, including basic methods of analysis and cost data for project evaluation and comparison of storage options, together with financial and business aspects associated with spent fuel storage. After the review of technical options for spent fuel storage in Section 2, cost categories and components involved in the lifecycle of a storage facility are identified in Section 3 and factors affecting costs of spent fuel storage are then reviewed in the Section 4. Methods for cost estimation and analysis are introduced in Section 5, and other financial and business aspects associated with spent fuel storage are discussed in Section 6.

  9. What does time spent on searching indicate?

    DEFF Research Database (Denmark)

    Borlund, Pia; Dreier, Sabine; Byström, Katriina

    2012-01-01

    In this paper, we report a comparative study on what users’ time spent on searching for information is an indication of. Time spent is commonly interpreted as an implicit measure of interest, but might indeed describe other circumstances of the information retrieval (IR) interaction. This phenome......In this paper, we report a comparative study on what users’ time spent on searching for information is an indication of. Time spent is commonly interpreted as an implicit measure of interest, but might indeed describe other circumstances of the information retrieval (IR) interaction....... This phenomenon of time spent is interesting from an IR evaluation point of view with reference to how time spent is to be interpreted. A comparison of time spent between a semi-lab interactive IR (IIR) study using simulated work task situations and a naturalistic IIR study is presented. The findings...... of this comparison are further related to a study on information searching and seeking in the real work environment that provides a resonance board for the reported IIR studies. The main conclusion is that time spent searching depends not only on interest, but also on circumstances such as prior knowledge...

  10. Dibromidodimethyldipyridineplatinum(IV

    Directory of Open Access Journals (Sweden)

    Mairéad E. Kelly

    2008-11-01

    Full Text Available In the title complex, [PtBr2(CH32(C5H5N2], the PtIV metal centre lies on a twofold rotation axis and adopts a slightly distorted octahedral coordination geometry. The structure displays weak intramolecular C—H...Br hydrogen-bonding interactions.

  11. Spent fuel management in India

    International Nuclear Information System (INIS)

    Balu, K.

    1998-01-01

    From Indian point of view, the spent fuel management by the reprocessing and plutonium recycle option is considered to be a superior and an inevitable option. The nuclear energy programme in Indian envisages three stages of implementation involving installation of thermal reactors in the first phase followed by recycling of plutonium from reprocessed fuel in fast breeder reactors and in the third phase utilization of its large thorium reserves in reactor system based on U-233-Th cycle. The Indian programme for Waste Management envisages disposal of low and intermediate level radioactive waste in near surface disposal facilities and deep geological disposal for high level and alpha bearing wastes. A Waste Immobilization Plant (WHIP), employing metallic melter for HLW vitrification is operational at Tarapur. Two more WIPs are being set up at Kalpakkam and Tarapur. A Solid waste Storage Surveillance Facility (SSSF) is also set up for interim storage of vitrified HLW. Site investigations are in progress for selecting site for ultimate disposal in igneous rock formations. R and D works is taken up on partitioning of HLW. Solvent extraction and extraction chromatographic studies are in progress. Presently emphasis is on separation of heat generating short lived nuclides like strontium and alpha emitters. (author)

  12. A engenharia genética

    OpenAIRE

    Candeias,José Alberto Neves

    1991-01-01

    São abordados os progressos havidos com as técnicas de engenharia genética, capazes de alterar o potencial genético de um organismo, quer pela introdução, quer pela supressão de novos genes estruturais. São mencionadas algumas das aplicações em geral e, em particular, possibilidades de uso no campo da medicina. É feita uma análise crítica dos benefícios e riscos envolvidos.

  13. Advantages of co-located spent fuel reprocessing, repository and underground reactor facilities

    International Nuclear Information System (INIS)

    Mahar, James M.; Kunze, Jay F.; Wes Myers, Carl; Loveland, Ryan

    2007-01-01

    The purpose of this work is to extend the discussion of potential advantages of the underground nuclear park (UNP) concept by making specific concept design and cost estimate comparisons for both present Generation III types of reactors and for some of the modular Gen IV or the GNEP modular concept. For the present Gen III types, we propose co-locating reprocessing and (re)fabrication facilities along with disposal facilities in the underground park. The goal is to determine the site costs and facility construction costs of such a complex which incorporates the advantages of a closed fuel cycle, nuclear waste repository, and ultimate decommissioning activities all within the UNP. Modular power generation units are also well-suited for placement underground and have the added advantage of construction using current and future tunnel boring machine technology. (authors)

  14. Enhanced Design Alternative IV

    International Nuclear Information System (INIS)

    Kramer, N.E.

    1999-01-01

    This report evaluates Enhanced Design Alternative (EDA) IV as part of the second phase of the License Application Design Selection (LADS) effort. The EDA IV concept was compared to the VA reference design using criteria from the Design Input Request for LADS Phase II EDA Evaluations (CRWMS M and O 1999b) and (CRWMS M and O 1999f). Briefly, the EDA IV concept arranges the waste packages close together in an emplacement configuration known as line load. Continuous pre-closure ventilation keeps the waste packages from exceeding their 350 C cladding and 200 C (4.3.6) drift wall temperature limits. This EDA concept keeps relatively high, uniform emplacement drift temperatures (post-closure) to drive water away from the repository and thus dry out the pillars between emplacement drifts. The waste package is shielded to permit human access to emplacement drifts and includes an integral filler inside the package to reduce the amount of water that can contact the waste form. Closure of the repository is desired 50 years after first waste is emplaced. Both backfill and drip shields will be emplaced at closure to improve post-closure performance. The EDA IV concept includes more defense-in-depth layers than the VA reference design because of its backfill, drip shield, waste package shielding, and integral filler features. These features contribute to the low dose-rate to the public achieved during the first 10,000 years of repository life as shown in Figure 3. Investigation of the EDA IV concept has led to the following general conclusions: (1) The total life cycle cost for EDA IV is about $21.7 billion which equates to a $11.3 billion net present value (both figures rounded up). (2) The incidence of design basis events for EDA IV is similar to the VA reference design. (3) The emplacement of the waste packages in drifts will be similar to the VA reference design. However, heavier equipment may be required because the shielded waste package will be heavier. (4) The heavier

  15. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning.

  16. A Study on the Planning of Technology Development and Research for Generation IV Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Moon Hee; Kim, H. R.; Kim, H. J. and others

    2005-08-15

    This study aimed at the planning the domestic technology development of the Gen IV and the formulating the international collaborative project contents and executive plan for 'A Validity Assessment and Policies of the R and D of Generation IV Nuclear Energy Systems'. The results of the study include follows; - Survey of the technology state in the fields of the Gen IV system specific technologies and the common technologies, and the plans of the international collaborative research - Drawing up the executive research and development plan by the experts of the relevant technology field for the systems which Korean will participate in. - Formulating the effective conduction plan of the program reflecting the view of the experts from the industry, the university and the research institute. - Establishing the plan for estimation of the research fund and the manpower for the efficient utilization of the domestic available resources. This study can be useful material for evaluating the appropriateness of the Korea's participation in the international collaborative development of the Gen IV, and can be valuably utilized to establish the strategy for the effective conduction of the program. The executive plan of the research and development which was produced in this study will be used to the basic materials for the establishing the guiding direction and the strategic conduction of the program when the research and development is launched in the future.

  17. The management strategy of spent nuclear fuel

    International Nuclear Information System (INIS)

    Bandi Parapak; Siti Alimah

    2010-01-01

    The assessment of management strategy of spent nuclear fuel has been carried out. Spent nuclear fuel is one of the by-products of nuclear power plant. The technical operations related to the management of spent fuel discharged from reactors are called the back-end fuel cycle. It can be largely divided into three option s : the once-through cycle, the closed cycle and the so-called ‟wait and see” policy. Whatever strategy is selected for the back-end of the nuclear fuel cycle, Away-from-Reactor (AFR) storage facilities has to be constructed. For the once through cycle, the entire content of spent fuel is considered as waste, and is subject to be disposed of into a deep underground repository. In the closed cycle, however, can be divided into: (1) uranium and plutonium are recovered from spent fuel by reprocessing and recycled to manufacture mixed oxide (MOX) fuel rods, (2) waste transmutation in accelerator-driven subcritical reactors, (3) DUPIC (Direct Use of Spent PWR Fuel In CANDU) concept. In wait and see policy, which means first storing the spent fuel and deciding at a later stage on reprocessing or disposal. (author)

  18. Rock cavern storage of spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Won Jin; Kim, Kyung Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Sang Ki [Inha University, Incheon (Korea, Republic of)

    2015-12-15

    The rock cavern storage for spent fuel has been assessed to apply in Korea with reviewing the state of the art of the technologies for surface storage and rock cavern storage of spent fuel. The technical feasibility and economic aspects of the rock cavern storage of spent fuel were also analyzed. A considerable area of flat land isolated from the exterior are needed to meet the requirement for the site of the surface storage facilities. It may, however, not be easy to secure such areas in the mountainous region of Korea. Instead, the spent fuel storage facilities constructed in the rock cavern moderate their demands for the suitable site. As a result, the rock cavern storage is a promising alternative for the storage of spent fuel in the aspect of natural and social environments. The rock cavern storage of spent fuel has several advantages compared with the surface storage, and there is no significant difference on the viewpoint of economy between the two alternatives. In addition, no great technical difficulties are present to apply the rock cavern storage technologies to the storage of domestic spent fuel.

  19. Spent nuclear fuel disposal liability insurance

    International Nuclear Information System (INIS)

    Martin, D.W.

    1984-01-01

    This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides

  20. The cost of spent fuel storage

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-15

    Spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments, constructing repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution?, or What is the best technology for an specific solution? Many countries have deferred the decision on selecting an option, while others works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However currently, the plants are under a process for extended power up-rate to 20% of original power and also there are plans to extended operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. (Author)

  1. Els gens i el cervell social

    OpenAIRE

    Bueno i Torrens, David, 1965-

    2008-01-01

    Fa unes setmanes un grup suec va relacionar la funció d'un gen anomenat receptor de la vasopressina amb el comportament monògam dels humans. La seva funció és que determinades neurones rebin un senyal neuroquímic, la vasopressina, molt més abundant en cervells masculins [...].

  2. Review on Korea Participation of Generation IV International Forum (GIF)

    International Nuclear Information System (INIS)

    Lee, Jewhan; Jeong, Ji-Young; Hahn, Dohee

    2015-01-01

    Generation IV International Forum (GIF) originates from US proposal of an initiative in 2000. The vision was to leapfrog LWR technology and collaborate with international partners to share R and D on advanced nuclear systems. Nine countries and EU joined the initiative and Gen IV concept was defined via technology goals and legal framework. Two years study with more than 100 experts worldwide has evaluated nearly 100 reactor designs and down selected six most promising concepts. In 2005, the first signatures on Framework Agreement were collected and the first research projects were defined in 2006. Korea is one of the founding members of GIF and actively participating in various areas. In 2013, TD was assigned to Korean expert and Korea is endeavoring to enhance the benefit of participation since this turning point. In this paper, pros and cons of engaging with GIF were briefly introduced and items to maximize the benefit were suggested

  3. Divergência genética entre genótipos de frangos tipo caipira

    Directory of Open Access Journals (Sweden)

    R. C. Veloso

    2015-10-01

    Full Text Available RESUMOObjetivou-se com este trabalho verificar a divergência genética entre sete genótipos de frangos tipo caipira da linhagem Redbro utilizando as características de desempenho por meio de técnicas de análise multivariada. Foram utilizados 840 pintos de um dia, machos, distribuídos em delineamento inteiramente ao acaso, dos seguintes genótipos: Caboclo, Carijó, Colorpak, Gigante Negro, Pesadão Vermelho, Pescoço Pelado e Tricolor. Após a consistência dos dados, foram avaliadas as seguintes variáveis: ganho em peso médio diário, consumo de ração médio diário e conversão alimentar, para os períodos: 1 a 28, 1 a 56, 1 a 70 e 1 a 84 dias de idade; peso corporal ao nascimento, aos 28, 56, 70 e aos 84 dias de idade. O desempenho dos genótipos foi avaliado por meio da análise de variância multivariada e da função discriminante linear de Fisher, usando os testes do maior autovalor de Roy e da união-interseção de Roy para as comparações múltiplas. O estudo da divergência genética foi feito por meio da análise por variáveis canônicas e pelo método de otimização de Tocher. Os genótipos Caboclo e Gigante Negro apresentaram médias canônicas diferentes dos demais genótipos. As duas primeiras variáveis canônicas explicaram 97,41% da variação entre os genótipos. A divergência genética entre os genótipos avaliados permitiu a formação de quatro grupos com os seguintes genótipos: grupo 1 - Colorpak; grupo 2 - Pesadão Vermelho e Pescoço Pelado; grupo 3 - Carijó e Tricolor; e grupo 4 - Caboclo e Gigante Negro.

  4. Proceedings of the third spent fuel workshop

    International Nuclear Information System (INIS)

    Werme, L.

    1984-03-01

    The third workshop, held in Boston, Mass. November 10-11, 1983 was organized by Battelle PNL. Questions concerning spent fuel behaviour in nuclear waste repositories were discussed. The following three lectures were presented. The corrosion of Spent UO 2 -Fuel in Synthetic Groundwater, R.S. Forsyth, K. Svanberg and L.O. Werme. Leaching and Radiolysis Studies on UO 2 Fuel, L.H. Johnson, S. Stroes-Gascoyne, D.W. Shoesmith, M.G. Bailey and D.M. Sellinger. Comparison of Spent Fuel and UO 2 Release in Salt Brines, W.J. Gray and G.L. McVay. (G.B.)

  5. Immobilization of spent resin with epoxy resin

    International Nuclear Information System (INIS)

    Gultom, O.; Suryanto; Sayogo; Ramdan

    1997-01-01

    immobilization of spent resin using epoxy resin has been conducted. The spent resin was mixtured with epoxy resin in variation of concentration, i.e., 30, 40, 50, 60, 70 weight percent of spent resin. The mixture were pour into the plastic tube, with a diameter of 40 mm and height of 40 mm. The density, compressive strength and leaching rate were respectively measured by quanta chrome, paul weber apparatus and gamma spectrometer. The results showed that the increasing of waste concentration would be decreased the compressive strength, and increased density by immobilized waste. The leaching rate of 137 Cs from waste product was not detected in experiment (author)

  6. The European gen-set market: growth and consolidation mean joy and pain

    International Nuclear Information System (INIS)

    French, Ian

    2000-01-01

    The changes in the European gen-set market are discussed. In recent years the market has undergone a period of increasing consolidation: prices fell and some companies folded. However, the market is not dead and continued growth is expected over the next five years although the compound rate is forecast to be only 1.5%. The article is presented under the sub-headings of (i) current market situation; (ii) product lifecycle; (iii) shipments by technology; (iv) market deregulation; (v) technology overview (spark ignition, compression ignition and gas turbines) (vi) European market: national overview and (vii) key market challenges (competition, emissions and over capacity)

  7. ATR Spent Fuel Options Study

    International Nuclear Information System (INIS)

    Connolly, Michael James; Bean, Thomas E.; Brower, Jeffrey O.; Luke, Dale E.; Patterson, M. W.; Robb, Alan K.; Sindelar, Robert; Smith, Rebecca E.; Tonc, Vincent F.; Tripp, Julia L.; Winston, Philip L.

    2017-01-01

    The Advanced Test Reactor (ATR) is a materials and fuels test nuclear reactor that performs irradiation services for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Naval Reactors, the National Nuclear Security Administration (NNSA), and other research programs. ATR achieved initial criticality in 1967 and is expected to operate in support of needed missions until the year 2050 or beyond. It is anticipated that ATR will generate approximately 105 spent nuclear fuel (SNF) elements per year through the year 2050. Idaho National Laboratory (INL) currently stores 2,008 ATR SNF elements in dry storage, 976 in wet storage, and expects to have 1,000 elements in wet storage before January 2017. A capability gap exists at INL for long-term (greater than the year 2050) management, in compliance with the Idaho Settlement Agreement (ISA), of ATR SNF until a monitored retrievable geological repository is open. INL has significant wet and dry storage capabilities that are owned by the DOE Office of Environmental Management (EM) and operated and managed by Fluor Idaho, which include the Idaho Nuclear Technology and Engineering Center's (INTEC's) CPP-666, CPP-749, and CPP-603. In addition, INL has other capabilities owned by DOE-NE and operated and managed by Battelle Energy Alliance, LLC (BEA), which are located at the Materials and Fuel Complex (MFC). Additional storage capabilities are located on the INL Site at the Naval Reactors Facility (NRF). Current INL SNF management planning, as defined in the Fluor Idaho contract, shows INTEC dry fuel storage, which is currently used for ATR SNF, will be nearly full after transfer of an additional 1,000 ATR SNF from wet storage. DOE-NE tasked BEA with identifying and analyzing options that have the potential to fulfill this capability gap. BEA assembled a team comprised of SNF management experts from Fluor Idaho, Savannah River Site (SRS), INL/BEA, and the MITRE Corp with an objective of developing and

  8. ATR Spent Fuel Options Study

    Energy Technology Data Exchange (ETDEWEB)

    Connolly, Michael James [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bean, Thomas E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Brower, Jeffrey O. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Luke, Dale E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Patterson, M. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, Alan K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sindelar, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Rebecca E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tonc, Vincent F. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tripp, Julia L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    The Advanced Test Reactor (ATR) is a materials and fuels test nuclear reactor that performs irradiation services for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Naval Reactors, the National Nuclear Security Administration (NNSA), and other research programs. ATR achieved initial criticality in 1967 and is expected to operate in support of needed missions until the year 2050 or beyond. It is anticipated that ATR will generate approximately 105 spent nuclear fuel (SNF) elements per year through the year 2050. Idaho National Laboratory (INL) currently stores 2,008 ATR SNF elements in dry storage, 976 in wet storage, and expects to have 1,000 elements in wet storage before January 2017. A capability gap exists at INL for long-term (greater than the year 2050) management, in compliance with the Idaho Settlement Agreement (ISA), of ATR SNF until a monitored retrievable geological repository is open. INL has significant wet and dry storage capabilities that are owned by the DOE Office of Environmental Management (EM) and operated and managed by Fluor Idaho, which include the Idaho Nuclear Technology and Engineering Center’s (INTEC’s) CPP-666, CPP-749, and CPP-603. In addition, INL has other capabilities owned by DOE-NE and operated and managed by Battelle Energy Alliance, LLC (BEA), which are located at the Materials and Fuel Complex (MFC). Additional storage capabilities are located on the INL Site at the Naval Reactors Facility (NRF). Current INL SNF management planning, as defined in the Fluor Idaho contract, shows INTEC dry fuel storage, which is currently used for ATR SNF, will be nearly full after transfer of an additional 1,000 ATR SNF from wet storage. DOE-NE tasked BEA with identifying and analyzing options that have the potential to fulfill this capability gap. BEA assembled a team comprised of SNF management experts from Fluor Idaho, Savannah River Site (SRS), INL/BEA, and the MITRE Corp with an objective of developing and analyzing

  9. GEN-IV Benchmarking of Triso Fuel Performance Models under accident conditions modeling input data

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise Paul [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: • The modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release. • The modeling of the AGR-1 and HFR-EU1bis safety testing experiments. • The comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from “Case 5” of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. “Case 5” of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to “effects of the numerical calculation method rather than the physical model” [IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary. 09/2016: Tables 6 and 8 updated. AGR-2 input data added

  10. Optimized, Competitive Supercritical-CO2 Cycle GFR for Gen IV Service

    International Nuclear Information System (INIS)

    M.J. Driscoll; P. Hejzlar; G. Apostolakis

    2008-01-01

    An overall plant design was developed for a gas-cooled fast reactor employing a direct supercritical Brayton power conversion system. The most important findings were that (1) the concept could be capital-cost competitive, but startup fuel cycle costs are penalized by the low core power density, specified in large part to satisfy the goal of significant post-accident passive natural convection cooling; (2) active decay heat removal is preferable as the first line of defense, with passive performance in a backup role; (3) an innovative tube-in-duct fuel assembly, vented to the primary coolant, appears to be practicable; and (4) use of the S-Co2 GFR to support hydrogen production is a synergistic application, since sufficient energy can be recuperated from the product H2 and 02 to allow the electrolysis cell to run 250 C hotter than the reactor coolant, and the water boilers can be used for reactor decay heat removal. Increasing core power density is identified as the top priority for future work on GFRs of this type

  11. Consistent Multigroup Theory Enabling Accurate Course-Group Simulation of Gen IV Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rahnema, Farzad; Haghighat, Alireza; Ougouag, Abderrafi

    2013-11-29

    The objective of this proposal is the development of a consistent multi-group theory that accurately accounts for the energy-angle coupling associated with collapsed-group cross sections. This will allow for coarse-group transport and diffusion theory calculations that exhibit continuous energy accuracy and implicitly treat cross- section resonances. This is of particular importance when considering the highly heterogeneous and optically thin reactor designs within the Next Generation Nuclear Plant (NGNP) framework. In such reactors, ignoring the influence of anisotropy in the angular flux on the collapsed cross section, especially at the interface between core and reflector near which control rods are located, results in inaccurate estimates of the rod worth, a serious safety concern. The scope of this project will include the development and verification of a new multi-group theory enabling high-fidelity transport and diffusion calculations in coarse groups, as well as a methodology for the implementation of this method in existing codes. This will allow for a higher accuracy solution of reactor problems while using fewer groups and will reduce the computational expense. The proposed research represents a fundamental advancement in the understanding and improvement of multi- group theory for reactor analysis.

  12. Millimeter-Wave Thermal Analysis Development and Application to GEN IV Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Wosko, Paul; Sundram, S. K.

    2012-10-16

    New millimeter-wave thermal analysis instrumentation has been developed and studied for characterization of materials required for diverse fuel and structural needs in high temperature reactor environments such as the Next Generation Nuclear Plant (NGNP). A two-receiver 137 GHz system with orthogonal polarizations for anisotropic resolution of material properties has been implemented at MIT. The system was tested with graphite and silicon carbide specimens at temperatures up to 1300 ºC inside an electric furnace. The analytic and hardware basis for active millimeter-wave radiometry of reactor materials at high temperature has been established. Real-time, non contact measurement sensitivity to anisotropic surface emissivity and submillimeter surface displacement was demonstrated. The 137 GHz emissivity of reactor grade graphite (NBG17) from SGL Group was found to be low, ~ 5 %, in the 500 – 1200 °C range and increases by a factor of 2 to 4 with small linear grooves simulating fracturing. The low graphite emissivity would make millimeter-wave active radiometry a sensitive diagnostic of graphite changes due to environmentally induced stress fracturing, swelling, or corrosion. The silicon carbide tested from Ortek, Inc. was found to have a much higher emissivity at 137 GHz of ~90% Thin coatings of silicon carbide on reactor grade graphite supplied by SGL Group were found to be mostly transparent to millimeter-waves, increasing the 137 GHz emissivity of the coated reactor grade graphite to about ~14% at 1250 ºC.

  13. A Review of Alloy 800H for Applications in the Gen IV Nuclear Energy Systems

    International Nuclear Information System (INIS)

    Ren, Weiju; Swindeman, Robert W.

    2010-01-01

    Alloy 800H is currently under consideration for applications in the Next Generation Nuclear Plant at operational temperatures above 750 C. To provide supporting information in this paper at the attempt to facilitate the consideration, service requirements of the nuclear system for structural materials is first described; and then an extensive review of Alloy 800H is given on its codification with respect to development and research history, mechanical behavior and design allowables, metallurgical aging resistance, environmental effect considerations, data requirements and availability, weldments, as well as many other aspects relevant to the intended nuclear application; an finally further research and development activities to support the materials qualification are suggested.

  14. Database of thermophysical properties of liquid metal coolants for GEN-IV

    International Nuclear Information System (INIS)

    Sobolev, V.

    2011-01-01

    The report presents an Fe-Cr interatomic potential to model high-Cr ferritic alloys. The potential is fitted to thermodynamic and point-defect properties obtained from density functional theory (DFT) calculations and experiments. The developed potential is also benchmarked against other potentials available in literature. It shows particularly good agreement with the DFT obtained mixing enthalpy of the random alloy, the formation energy of intermetallics and experimental excess vibrational entropy and phase diagram. In addition, DFT calculated point-defect properties, both interstitial and substitutional, are well reproduced, as is the screw dislocation core structure. As a first validation of the potential, we study the precipitation hardening of Fe-Cr alloys via static simulations of the interaction between Cr precipitates and screw dislocations. It is concluded that the description of the dislocation core modification near a precipitate might have a significant influence on the interaction mechanisms observed in dynamic simulations.

  15. GEN IV: Carbide Fuel Elaboration for the 'Futurix Concepts' experiment

    Energy Technology Data Exchange (ETDEWEB)

    Vaudez, Stephane; Riglet-Martial, Chantal; Paret, Laurent; Abonneau, Eric [Commissariat a l' Energie Atomique (C.E.A.), Direction de l' Energie Nucleaire, Centre d' Etudes de Cadarache, 13108 Saint Paul lez Durance Cedex (France)

    2008-07-01

    In order to collect information on the behaviour of the future GFR (Gas Fast Reactor) fuel under fast neutron irradiation, an experimental irradiation program, called 'Futurix-concepts' has been launched at the CEA. The considered concept is a composite material made of a fissile fuel embedded in an inert ceramic matrix. Fissile fuel pellets are made of UPuN or UPuC while ceramics are SiC for the carbide fuel and TiN for the nitride fuel. This paper focuses on the description of the carbide composite fabrication. The UPuC pellets are manufactured using a metallurgical powder process. Fabrication and handling of the fuels are carried out in glove boxes under a nitrogen atmosphere. Carbide fuel is synthesized by carbo-thermic reduction under vacuum of a mixture of actinide oxide and graphitic carbon up to 1550 deg. C. After ball milling, the UPuC powder is pressed to create hexagonal or spherical compacts. They are then sintered up to 1750 deg. C in order to obtain a density of 85 % of the theoretical one. The sintered pellets are inserted into an inert and tight capsule of SiC. In order to control the gap between the fuel and the matrix precisely, the pellets are abraded. The inert matrix is then filled with the pellets and the whole system is sealed by a BRASiC{sup R} process at high temperature under a helium atmosphere. Fabrication of the sample to be irradiated was done in 2006 and the irradiation began in May 2007 in the Phenix reactor. This presentation will detail and discuss the results obtained during this fabrication phase. (authors)

  16. Project planning of Gen-IV sodium cooled fast reactor technology

    International Nuclear Information System (INIS)

    Yoo, Jaewoon; Joo, H. K.; Cho, C. H.; Kim, Y. G.; Lee, D. U.; Jin, M. W.

    2013-05-01

    The project program will be established to shorten the design schedule by sharing the design man power and experimental facility, and by introducing the proven technology through international collaboration and the project plan including preliminary specific design, technology validation and fuel design validation plan will be more detail by reviewing the plan at the International Technical Review Meeting (ITRM). Periodic project progress review meeting will be held to find the technical issues and to resolve them. The results of the progress review meeting will be reflected into the final assessment of research project. The project progress review meeting will be held every quarter and external expert will also participate in the meeting. In parallel with the PGSFR development, innovative small modular SFR will be developed aiming to the international nuclear market. The system and component technologies of both system can be shared but innovative concept will be implemented into the design. Ultra long life core design concept and supercritical CO 2 Brayton cycle will be considered as the innovative concept for enhancing the plant economy and safety

  17. Tritium Sequestration in Gen IV NGNP Gas Stream via Proton Conducting Ceramic Pumps

    International Nuclear Information System (INIS)

    Chen, Franglin Frank; Adams, Thad M.; Brinkman, Kyle; Reifsnider, Kenneth

    2011-01-01

    Several perovskite structured proton conductors based on SrCeO 3 and BaCeO 3 have been investigated in the project. The solid solutions for SrCeO 3 and BaCeO 3 were first investigated. The morphological and electrical properties of Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ with x varying from 0 to 1 prepared by a modified Pechini method were investigated as potential high temperature proton conductors. Dense microstructures were achieved for all the samples upon sintering at 1500ees)C for 5 h. The phase structure analysis indicated that perovskite phase was formed for 0≤x≤0.2, while for x larger than 0.5, impurity phases of Sr 2 CeO 4 and Y 2 O 3 appeared. The stability tests indicated that the resistance to boiling water for Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ was between that of BaCe 0.8 Y 0.2 O 3-δ and SrCe 0.8 Y 0.2 O 3-δ Due to the tendency of the reaction with CO 2 for both BaCe 0.8 Y 0.2 O 3-δ and SrCe 0.8 Y 0.2 O 3-δ , it was not surprising that Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ was also not stable in CO 2 containing atmospheres. The conductivity tests indicated that Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ possessed the electrical conductivity between BaCe 0.8 Y 0.2 O 3-δ and SrCe 0.8 Y 0.2 O 3-δ . The conductivity decreased and the activation energy increased with the increase in Sr content in Ba 1-x Sr x Ce 0.8 Y 0.2 O 3-δ .

  18. Tritium Sequestration in Gen IV NGNP Gas Stream via Proton Conducting Ceramic Pumps

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Fanglin Frank [Univ. of South Carolina, Columbia, SC (United States); Adams, Thad M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Brinkman, Kyle [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Reifsnider, Kenneth [Univ. of South Carolina, Columbia, SC (United States)

    2011-09-30

    Several types of high-temperature proton conductors based on SrCeO3 and BaCeO3 have been systematically investigated in this project for tritium separation in NGNP applications. One obstacle for the field application is the chemical stability issues in the presence of steam and CO2 for these proton conductors. Several strategies to overcome such issues have been evaluated, including A site doping and B site co-doping method for perovskite-structured proton conductors. Novel zirconium-free proton conductors have also been developed with improved electrical conductivity and enhanced chemical stability. Novel catalytic materials for the proton-conducting separation membranes have been investigated. A tubular geometry proton-conducting membrane has been developed for the proton separation membranes. Total dose rate estimated from tritium decay (beta emission) under realistic membrane operating conditions, combined with electron irradiation experiments, indicates that proton ceramic materials possess the appropriate radiation stability for this application.

  19. Electromagnet Response Time Tests on Primary CRDM of a Prototype Gen-IV SFR

    International Nuclear Information System (INIS)

    Lee, Jae-Han; Koo, Gyeong-Hoi

    2015-01-01

    This paper identifies the electromagnetic response characteristics of the electromagnet of a primary control rod drive mechanism (CRDM) used for the reactor scram function. The test measures the electromagnet response time required to release an armature from a stator controlled by a loss of an electromagnetic force on an armature after shorting a power supply to an electromagnet coil. These tests are carried out while changing the electromagnet core material, an assist spring, and an armature holding current. The main factors influencing the test parameters on the response are found to be the armature holding current for holding the armature loads, and the material type of the electromagnet cores. The minimum response time is 0.13 seconds in the case of using SS410 material as an armature, while the S10C material as an armature has a response time of 0.21 seconds. Electromagnet response time characteristics from the test results will be evaluated by comparing the precise moving data of an electromagnet armature through the use of a high-speed camera and a potentiometer in the future

  20. Studies on woloszynskioid dinoflagellates IV: the genus Biecheleria gen. nov

    DEFF Research Database (Denmark)

    Moestrup, Øjvind; Lindberg, Karin; Daugbjerg, Niels

    2009-01-01

    in the dinoflagellates. Biecheleria also comprises the brackish water species Biecheleria baltica sp. nov. (presently identified as Woloszynskia halophila) and the marine species Biecheleria natalensis (syn. Gymnodinium natalense). Gymnodinium halophilum described in 1952 by B. Biecheler but apparently not subsequently...... refound, is transferred to Biecheleria. The Suessiaceae further includes the marine species Protodinium simplex, described by Lohmann in 1908 but shortly afterwards (1921) transferred to Gymnodinium by Kofoid and Swezy and subsequently known as Gymnodinium simplex. It only distantly related to Gymnodinium...

  1. Optimized, Competitive Supercritical-CO2 Cycle GFR for Gen IV Service

    Energy Technology Data Exchange (ETDEWEB)

    M.J. Driscoll; P. Hejzlar; G. Apostolakis

    2008-09-08

    An overall plant design was developed for a gas-cooled fast reactor employing a direct supercritical Brayton power conversion system. The most important findings were that (1) the concept could be capital-cost competitive, but startup fuel cycle costs are penalized by the low core power density, specified in large part to satisfy the goal of significatn post-accident passive natural convection cooling; (2) active decay heat removal is preferable as the first line of defense, with passive performance in a backup role; (3) an innovative tube-in-duct fuel assembly, vented to the primpary coolant, appears to be practicable; and (4) use of the S-Co2 GFR to support hydrogen production is a synergistic application, since sufficient energy can be recuperated from the product H2 and 02 to allow the electrolysis cell to run 250 C hotter than the reactor coolant, and the water boilers can be used for reactor decay heat removal. Increasing core poer density is identified as the top priority for future work on GFRs of this type.

  2. Control Rod Withdrawal Events Analyses for the Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseo; Jeong, Taekyeong; Jeong, Jaeho; Chang, Wonpyo; Lee, Seungwon; An, Sangjun; Lee, Kwilim [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    To confirm the limiting condition, based on the maximum allowable reactivity insertion of 0.3 $, three cases from the end of cycle (EOC) are selected. In addition, assuming the failure of CRSS by earthquake, additional cases is defined at beginning of cycle (BOC). When the CRW occurs, the reactor can be protected by plant protection system (PPS). In this study, PPS mechanism is sequentially studied for all initiating events. For design basis accidents (DBA), the reactor can be scrammed by reactor protection system (RPS). The first and seconds RPS signals are checked during transients. When RPS is failed, so called as anticipated transient without scram (ATWS), the reactor will be protected by diverse protection system (DPS). In this study, in order to analyze various initiating events related control rod withdrawal, four kinds of operating condition is defined. TOP events are analyzed using MARS-LMR. The influence of various plant protection system such as RPS and DPS are investigated.

  3. Preliminary Comparative Evaluation Study on Reference Design of GEN-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Yoon Sub; Kim, Yeong Il; Hong, Ser Gi (and others)

    2005-11-15

    A fast reactor has a good transmutation capability and it enables breeding of fuel and use of a closed fuel cycle. By these characteristics of a fast reactor, the limited uranium resources of the world can be much more effectively utilized and the nuclear wastes of a high level of radioactivity and toxicity from the current nuclear power reactors of LWRs and HWRs can be drastically reduced in its volume and the management of the wastes can be easily treated. Also electricity can be generated more effectively since a fast reactor has the feature of high operation temperature. These features of a fast reactor makes it inevitable on a long term basis to construct fast reactors in Korea. The domestic fast reactor technology level, however, is at the level of coming out of a beginning stage and needs utilization of international expertise. Recently an international cooperation program called GIF has been formulated and our KALIMER was selected as one of the two reference designs for the international joint R and D works with JSFR of Japan. In the current frame of the GIF program, the two selected reference designs are supposed to be evaluated against each other in future and one design is to be finally selected. To make the international cooperation program directed more useful to our fast reactor technology development, it is required to strengthen the competitiveness of KALIMER so that it can be selected. To meet the necessity, a study was made in this research for pre-evaluation of the GIF reference designs and setting up plans for development of designs and technology that will enhance the competitiveness of KALIMER.

  4. Effects of environments on spent fuel

    International Nuclear Information System (INIS)

    Funk, C.W.; Jacobson, L.D.; Menon, M.N.

    1979-07-01

    This report describes the influence of water storage environment and transportation on spent light water reactor (LWR) fuel assemblies. It also estimates the storage duration and capacity requirements for several assumed scenarios

  5. Long term wet spent nuclear fuel storage

    International Nuclear Information System (INIS)

    1987-04-01

    The meeting showed that there is continuing confidence in the use of wet storage for spent nuclear fuel and that long-term wet storage of fuel clad in zirconium alloys can be readily achieved. The importance of maintaining good water chemistry has been identified. The long-term wet storage behaviour of sensitized stainless steel clad fuel involves, as yet, some uncertainties. However, great reliance will be placed on long-term wet storage of spent fuel into the future. The following topics were treated to some extent: Oxidation of the external surface of fuel clad, rod consolidation, radiation protection, optimum methods of treating spent fuel storage water, physical radiation effects, and the behaviour of spent fuel assemblies of long-term wet storage conditions. A number of papers on national experience are included

  6. Spent fuel workshop'2002

    Energy Technology Data Exchange (ETDEWEB)

    Poinssot, Ch

    2002-07-01

    This document gathers the transparencies of the presentations given at the 2002 spent fuel workshop: Session 1 - Research Projects: Overview on the IN CAN PROCESSES European project (M. Cowper), Overview on the SPENT FUEL STABILITY European project (C. Poinssot), Overview on the French R and D project on spent fuel long term evolution, PRECCI (C. Poinssot); Session 2 - Spent Fuel Oxidation: Oxidation of uranium dioxide single crystals (F. Garrido), Experimental results on SF oxidation and new modeling approach (L. Desgranges), LWR spent fuel oxidation - effects of burn-up and humidity (B. Hanson), An approach to modeling CANDU fuel oxidation under dry storage conditions (P. Taylor); Session 3 - Spent Fuel Dissolution Experiments: Overview on high burnup spent fuel dissolution studies at FZK/INE (A. Loida), Results on the influence of hydrogen on spent fuel leaching (K. Spahiu), Leaching of spent UO{sub 2} fuel under inert and reducing conditions (Y. Albinsson), Fuel corrosion investigation by electrochemical techniques (D. Wegen), A reanalysis of LWR spent fuel flow through dissolution tests (B. Hanson), U-bearing secondary phases formed during fuel corrosion (R. Finch), The near-field chemical conditions and spent fuel leaching (D. Cui), The release of radionuclides from spent fuel in bentonite block (S.S. Kim), Trace actinide behavior in altered spent fuel (E. Buck, B. Hanson); Session 4 - Radiolysis Issues: The effect of radiolysis on UO{sub 2} dissolution determined from electrochemical experiments with {sup 238}Pu doped UO{sub 2} M. Stroess-Gascoyne (F. King, J.S. Betteridge, F. Garisto), doped UO{sub 2} studies (V. Rondinella), Preliminary results of static and dynamic dissolution tests with {alpha} doped UO{sub 2} in Boom clay conditions (K. Lemmens), Studies of the behavior of UO{sub 2} / water interfaces under He{sup 2+} beam (C. Corbel), Alpha and gamma radiolysis effects on UO{sub 2} alteration in water (C. Jegou), Behavior of Pu-doped pellets in brines

  7. Spent Nuclear Fuel Project Safety Management Plan

    International Nuclear Information System (INIS)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities

  8. Electrodialytic decontamination of spent ion exchange resins

    International Nuclear Information System (INIS)

    Nott, B.R.

    1982-01-01

    Development of a novel electrodialytic decontamination process for the selective removal of radioactive Cs from spent ion exchange resins containing large amounts of Li is described. The process involves passage of a dc electric current through a bed of the spent ion exchange resin in a specially designed electrodialytic cell. The radiocesium so removed from a volume of the spent resin is concentrated onto a much smaller volume of a Cs selective sorbent to achieve a significant radioactive waste volume reduction. Technical feasibility of the electrodialytic resin decontamination process has been demonstrated on a bench scale with a batch of simulated spent ion exchange resin and using potassium cobalt ferrocyanide as the Cs selective sorbent. A volume reduction factor between 10 and 17 has been estimated. The process appears to be economically attractive. Improvements in process economics can be expected from optimization of the process. Other possible applications of the EDRD process have been identified

  9. Spent fuel storage requirements 1993--2040

    International Nuclear Information System (INIS)

    1994-09-01

    Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges

  10. Management of spent sealed sources in Indonesia

    International Nuclear Information System (INIS)

    Wisnubroto, D.S.

    2002-01-01

    This paper describes the effort of the Center for Development of Radioactive Waste Management (CDRWM) to develop and implement activities in maintaining and improving the safety of spent sealed radiation sources and the security of radioactive materials over their life cycle. There is a wide variety of uses of radiation sources and radioactive materials in Indonesia, while the CDRWM plan to cover all spent radiation sources. Primary consideration is given to sealed radiation sources with relatively high levels of radioactivity which might necessitate interventional measures should control over them be lost. The policy of the Government of Indonesia for spent radiation sources is, whenever possible, spent sealed sources should be returned to the supplier. CDRWM has a general principle that sealed sources should not be removed from their holders, or the holders physically modified (except for Ra-226 needles, smoke detector and lighting preventer). (author)

  11. Development of Pyrochemical Reprocessing of the Spent Nuclear Fuel and Prospects of Closed Fuel Cycle

    OpenAIRE

    Tulackova, R; Chuchvalcova-Bimova, K; Precek, M; Marecek, M; Uhlir, J

    2007-01-01

    Molten-Salt Reactor (MSR) is a design of an advanced reactor system from the GEN IV family working in thermal or epithermal neutron spectrum and using thorium or transuranium fuel in the form of molten fluorides. It is based on the experience with the development of the molten-salt reactor technology in the Oak-Ridge National Laboratory in the United States. The MSR fuel cycle with integrated reprocessing represents one of the potential ways both for significant decrease of total amount of ra...

  12. Nuclear spent fuel management. Experience and options

    International Nuclear Information System (INIS)

    1986-01-01

    Spent nuclear fuel can be stored safely for long periods at relatively low cost, but some form of permanent disposal will eventually be necessary. This report examines the options for spent fuel management, explores the future prospects for each stage of the back-end of the fuel cycle and provides a thorough review of past experience and the technical status of the alternatives. Current policies and practices in twelve OECD countries are surveyed

  13. Rack for storing spent nuclear fuel elements

    Science.gov (United States)

    Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

    1978-06-20

    A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

  14. Transport and storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Lung, M.; Lenail, B.

    1987-01-01

    From a safety standpoint, spent fuel is clearly not ideal for permanent disposal and reprocessing is the best method of preparing wastes for long-term storage in a repository. Furthermore, the future may demonstrate that some fission products recovered in reprocessing have economic applications. Many countries have in fact reached the point at which the recycling of plutonium and uranium from spent fuel is economical in LWR's. Even in countries where this is not yet evident, (i.e., the United States), the French example shows that the day will come when spent fuel will be retrieved for reprocessing and recycle. It is highly questionable whether spent fuel will ever be considered and treated as waste in the same sense as fission products and processed as such, i.e., packaged in a waste form for permanent disposal. Even when recycled fuel material can no longer be reused in LWR's because of poor reactivity, it will be usable in FBR's. Based on the considerable experience gained by SGN and Cogema, this paper has provided practical discussion and illustrations of spent fuel transport and storage of a very important step in the nuclear fuel management process. The best of spent fuel storage depends on technical, economic and policy considerations. Each design has a role to play and we hope that the above discussion will help clarify certain issues

  15. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Park, Seong Won; Shin, Y. J.; Cho, S. H.

    2004-03-01

    The research on spent fuel management focuses on the maximization of the disposal efficiency by a volume reduction, the improvement of the environmental friendliness by the partitioning and transmutation of the long lived nuclides, and the recycling of the spent fuel for an efficient utilization of the uranium source. In the second phase which started in 2001, the performance test of the advanced spent fuel management process consisting of voloxidation, reduction of spent fuel and the lithium recovery process has been completed successfully on a laboratory scale. The world-premier spent fuel reduction hot test of a 5 kgHM/batch has been performed successfully by joint research with Russia and the valuable data on the actinides and FPs material balance and the characteristics of the metal product were obtained with experience to help design an engineering scale reduction system. The electrolytic reduction technology which integrates uranium oxide reduction in a molten LiCl-Li 2 O system and Li 2 O electrolysis is developed and a unique reaction system is also devised. Design data such as the treatment capacity, current density and mass transfer behavior obtained from the performance test of a 5 kgU/batch electrolytic reduction system pave the way for the third phase of the hot cell demonstration of the advanced spent fuel management technology

  16. Spent Fuel Management Newsletter. No. 1

    International Nuclear Information System (INIS)

    1990-03-01

    This Newsletter has been prepared in accordance with the recommendations of the International Regular Advisory Group on Spent Fuel Management and the Agency's programme (GC XXXII/837, Table 76, item 14). The main purpose of the Newsletter is to provide Member States with new information about the state-of-the-art in one of the most important parts of the nuclear fuel cycle - Spent Fuel Management. The contents of this publication consists of two parts: (1) IAEA Secretariat contribution -work and programme of the Nuclear Materials and Fuel Cycle Technology Section of the Division of Nuclear Fuel Cycle and Waste Management, recent and planned meetings and publications, Technical Co-operation projects, Co-ordinated Research programmes, etc. (2) Country reports - national programmes on spent fuel management: current and planned storage and reprocessing capacities, spent fuel arisings, safety, transportation, storage, treatment of spent fuel, some aspects of uranium and plutonium recycling, etc. The IAEA expects to publish the Newsletter once every two years between the publications of the Regular Advisory Group on Spent Fuel Management. Figs and tabs

  17. Pyrochemical processing of DOE spent nuclear fuel

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1995-01-01

    A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or open-quotes pyroprocessing,close quotes provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory

  18. Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

    International Nuclear Information System (INIS)

    Tang, J.S.

    1980-11-01

    Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm

  19. An analysis of spent fuel characteristics for reactor RA spent fuel elements

    International Nuclear Information System (INIS)

    Milosevic, M.; Vukadin, Z.

    2001-01-01

    The need for reducing the intrinsic conservatism in the criticality safety assessment has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the work being performed to investigate the spent fuel characteristics and nuclear criticality safety of the reactor RA spent fuel storage. An analysis methodology is presented along with information representing the validation of the methods and geometrical models. Finally, the criticality safety analysis of the stainless steel containers and aluminum barrels, filled with spent fuel elements, is presented to demonstrate that an adequate margin of subcriticality is proved for the reactor RA spent fuel storage.(author)

  20. Genética humana e sociedade

    OpenAIRE

    Rosa, Vivian Leyser da

    2000-01-01

    Tese (doutorado) - Universidade Federal de Santa Catarina, Centro de Ciências da Educação. Análise do campo de estudos sobre o entendimento público da ciência, distinguindo os modelos de deficit cognitivo e interativo, bem como suas implicações na esfera educacional. Estudo do panorama dos avanços atuais da genética humana, do ponto de vista científico, ético e social. Análise de aspectos relativos ao ensino de genética humana nos cursos de graduação da área da saúde, em nove Universidades...

  1. Preserving Accuracy in GenBank

    DEFF Research Database (Denmark)

    Bidartondo, M.I.; Bruns, T. D.; Blackwell, M.

    2008-01-01

    GenBank, the public repository for nucleotide and protein sequences, is a critical resource for molecular biology, evolutionary biology, and ecology. While some attention has been drawn to sequence errors (1), common annotation errors also reduce the value of this database. In fact, for organisms...... such as fungi, which are notoriously difficult to identify, up to 20% of DNA sequence records may have erroneous lineage designations in GenBank (2). Gene function annotation in protein sequence databases is similarly error-prone (3, 4). Because identity and function of new sequences are often determined...... by bioinformatic analyses, both types of errors are propagated into new accessions, leading to long-term degradation of the quality of the database. Currently, primary sequence data are annotated by the authors of those data, and can only be reannotated by the same authors. This is inefficient and unsustainable...

  2. Multiscale Modeling of the Deformation of Advanced Ferritic Steels for Generation IV Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Nasr M. Ghoniem; Nick Kioussis

    2009-04-18

    The objective of this project is to use the multi-scale modeling of materials (MMM) approach to develop an improved understanding of the effects of neutron irradiation on the mechanical properties of high-temperature structural materials that are being developed or proposed for Gen IV applications. In particular, the research focuses on advanced ferritic/ martensitic steels to enable operation up to 650-700°C, compared to the current 550°C limit on high-temperature steels.

  3. Thermal analysis of spent nuclear fuels repository

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, F.; Salome, J.; Cardoso, F.; Velasquez, C.E.; Pereira, C. [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP 31270-901 (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores - CNPq, Asa Norte, Brazilia (Brazil); Viana, C. [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Pampulha, Belo Horizonte MG, CEP 31270-901 (Brazil); Barros, G.P. [Comissao Nacional de Energia Nuclear-CNEN, Rua Gal Severiano, n 90 - Botafogo, 22290-901, Rio de Janeiro, RJ (Brazil)

    2016-07-01

    In the first part, Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) spent fuels (SF) were evaluated to the thermal of the spent fuel pool (SFP) without an external cooling system. The goal is to compare the water boiling time of the pool storing different types of spent nuclear fuels. This study used the software ANSYS Workbench 16.2 - student version. For the VHTR, two types of fuel were analyzed: (Th,TRU)O{sub 2} and UO{sub 2}. This part of the studies were performed for wet storage condition using a single type of SF and decay heat values at times t=0 and t=10 years after the reactor discharge. The ANSYS CFX module was used and the results show that the time that water takes to reach the boiling point varies from 2.4 minutes for the case of VHTR-(Th,TRU)O{sub 2} SF at time t=0 year after reactor discharge until 32.4 hours for the case of PWR SF at time t=10 years after the discharge reactor. The second part of this work consists of modeling a geological repository. Firstly, the temperature evaluation of the spent fuel from a PWR was analyzed. A PWR canister was simulated using the ANSYS transient thermal module. Then the temperature of canister could be computed during the time spent on a portion of a geological repository. The mean temperature on the canister surface increased during the first nine years, reaching a plateau at 35.5 C. degrees between the tenth and twentieth years after the geological disposal. The idea is to extend this study for the other systems analyzed in the first part. The idea is to include in the study, the spent fuels from VHTR and ADS and to compare the canister behavior using different spent fuels. (authors)

  4. Final environmental statement: US Spent Fuel Policy. Storage of foreign spent power reactor fuel

    International Nuclear Information System (INIS)

    1980-05-01

    In October 1977, the Department of Energy (DOE) announced a Spent Fuel Storage Policy for nuclear power reactors. Under this policy, as approved by the President, US utilities will be given the opportunity to deliver spent fuel to US Government custody in exchange for payment of a fee. The US Government will also be prepared to accept a limited amount of spent fuel from foreign sources when such action would contribute to meeting nonproliferation goals. Under the new policy, spent fuel transferred to the US Government will be delivered - at user expense - to a US Government-approved site. Foreign spent fuel would be stored in Interim Spent Fuel Storage (ISFS) facilities with domestic fuel. This volume of the environmental impact statement includes effects associated with implementing or not implementing the Spent Fuel Storage Policy for the foreign fuels. The analyses show that there are no substantial radiological health impacts whether the policy is implemented or not. In no case considered does the population dose commitment exceed 0.000006% of the world population dose commitment from natural radiation sources over the period analyzed. Full implementation of the US offer to accept a limited amount of foreign spent fuel for storage provides the greatest benefits for US nonproliferation policy. Acceptance of lesser quantities of foreign spent fuel in the US or less US support of foreign spent fuel storage abroad provides some nonproliferation benefits, but at a significantly lower level than full implementation of the offer. Not implementing the policy in regard to foreign spent fuel will be least productive in the context of US nonproliferation objectives. The remainder of the summary provides a brief description of the options that are evaluated, the facilities involved in these options, and the environmental impacts, including nonproliferation considerations, associated with each option

  5. Fact sheet on spent fuel management

    International Nuclear Information System (INIS)

    2006-01-01

    The IAEA gives high priority to safe and effective spent fuel management. As an example of continuing efforts, the 2003 International Conference on Storage of Spent Fuel from Power Reactors gathered 125 participants from 35 member states to exchange information on this important subject. With its large number of Member States, the IAEA is well-positioned to gather and share information useful in addressing Member State priorities. IAEA activities on this topic include plans to produce technical documents as resources for a range of priority topics: spent fuel performance assessment and research, burnup credit applications, cask maintenance, cask loading optimization, long term storage requirements including records maintenance, economics, spent fuel treatment, remote technology, and influence of fuel design on spent fuel storage. In addition to broader topics, the IAEA supports coordinated research projects and technical cooperation projects focused on specific needs. The proceedings of the 2003 IAEA conference on storage of spent fuel from power reactors has been ranked in the top twenty most accessed IAEA publications. These proceedings are available for free downloads at http://www-pub.iaea.org/MTCD/publications/PubDetails.asp?pubId=6924]. The IAEA organized and held a 2004 meeting focused on long term spent fuel storage provisions in Central and Eastern Europe, using technical cooperation funds to support participation by these Member States. Over ninety percent of the participants in this meeting rated its value as good or excellent, with participants noting that the IAEA is having a positive effect in stimulating communication, cooperation, and information dissemination on this important topic. The IAEA was advised in 2004 that results from a recent coordinated research project (IAEA-TECDOC-1343) were used by one Member State to justify higher clad temperatures for spent fuel in dry storage, leading to more efficient storage and reduced costs. Long term

  6. Storing the world's spent nuclear fuel

    International Nuclear Information System (INIS)

    Barkenbus, J.N.; Weinberg, A.M.; Alonso, M.

    1985-01-01

    Given the world's prodigious future energy requirements and the inevitable depletion of oil and gas, it would be foolhardy consciously to seek limitations on the growth of nuclear power. Indeed, the authors continue to believe that the global nuclear power enterprise, as measured by installed reactor capacity, can become much larger in the future without increasing proliferation risks. To accomplish this objective will require renewed dedication to the non-proliferation regime, and it will require some new initiatives. Foremost among these would be the establishment of a spent fuel take-back service, in which one or a few states would retrieve spent nuclear fuel from nations generating it. The centralized retrieval of spent fuel would remove accessible plutonium from the control of national leaders in non-nuclear-weapons states, thereby eliminating the temptation to use this material for weapons. The Soviets already implement a retrieval policy with the spent fuel generated by East European allies. The authors believe that it is time for the US to reopen the issue of spent-fuel retrieval, and thus to strengthen its non-proliferation policies and the nonproliferation regime in general. 7 references

  7. Programme on spent fuel management in India

    International Nuclear Information System (INIS)

    Patil, M.P.; Rao, M.K.; Prasad, A.N.

    1996-01-01

    The Indian Atomic Energy Programme aims at harnessing the natural resources in the most optimal manner. To achieve this end, a three phase Nuclear Power Programme was devised in the early days of establishment of the Department of Atomic Energy. It envisages the utilization of the modest uranium reserves and rich thorium deposits in the country. The limited natural reserves of fuel materials have prompted India to pay increased attention to the back end of the fuel cycle, which consists of Reprocessing and Waste Management. The spent fuel is valued in India as a source for fuel and is treated accordingly, to recover the important fissile materials. Thus, the route of reprocessing of spent fuel in order to recycle uranium and plutonium in future reactors was opted for. Today, India possesses the capability and facilities, catering to the entire fuel cycle, i.e., starting from the mining of the ore, through fabrication of fuel and its application in reactors, to reprocessing of the spent fuel and appropriate waste management. With the third Reprocessing Plant almost ready for commissioning and second and third waste Immobilization Plants under constructions, the Spent Fuel Management Programme has come of age in India. This paper presents an overview of the status of the spent fuel management programme and its future perspective. (author)

  8. Diaquatetrabromidotin(IV trihydrate

    Directory of Open Access Journals (Sweden)

    Fei Ye

    2012-09-01

    Full Text Available The title compound, [SnBr4(H2O2]·3H2O, forms large colourless crystals in originally sealed samples of tin tetrabromide. It constitutes the first structurally characterized hydrate of SnBr4 and is isostructural with the corresponding hydrate of SnCl4. It is composed of SnIV atoms octahedrally coordinated by four Br atoms and two cis-related water molecules. The octahedra exhibit site symmetry 2. They are arranged into columns along [001] via medium–strong O—H...O hydrogen bonds involving the two lattice water molecules (one situated on a twofold rotation axis while the chains are interconnected via longer O—H...Br hydrogen bonds, forming a three-dimensional network.

  9. Analysis of a Spanish energy scenario with Generation IV nuclear reactors

    International Nuclear Information System (INIS)

    Ochoa, Raquel; Jimenez, Gonzalo; Perez-Martin, Sara

    2013-01-01

    Highlights: • Spanish energy scenario for the hypothetical deployment of Gen-IV SFR reactors. • Availability of national resources is assessed, considering SFR’s breeding. • An assessment of the impact of transmuting MA on the final repository. • SERPENT code with own pre- and post-processing tools were employed. • The employed SFR core design is based on the specifications of the CP-ESFR. - Abstract: The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed

  10. Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

    2014-02-01

    This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

  11. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  12. Status of spent fuel shipping cask development

    International Nuclear Information System (INIS)

    Hall, I.K.; Hinschberger, S.T.

    1989-01-01

    This paper discusses how several new-generation shopping cask systems are being developed for safe and economical transport of commercial spent nuclear fuel and other radioactive wastes for the generating sites to a federal geologic repository or monitored retrievable storage (MRS) facility. Primary objectives of the from-reactor spent fuel cask development work are: to increase cask payloads by taking advantage of the increased at-reactor storage time under the current spent fuel management scenario, to facilitate more efficient cask handling operations with reduced occupational radiation exposure, and to promote standardization of the physical interfaces between casks and the shipping and receiving facilities. Increased cask payloads will significantly reduce the numbers of shipments, with corresponding reductions in transportation costs and risks to transportation workers, cask handling personnel, and the general public

  13. Spent fuel storage requirements, 1990--2040

    International Nuclear Information System (INIS)

    Walling, R.; Bierschbach, M.

    1990-11-01

    Historical inventories of spent fuel are combined with US Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements over the next 51 years, through the year 2040. The needs for storage capacity beyond that presently available in the pools are estimated. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. Historical data through December 1989 are derived from the 1990 Form RW-859 data survey of nuclear utilities. Projected discharges through the end of reactor life are based on DOE estimates of future nuclear capacity, generation, and spent fuel discharges. 15 refs., 3 figs., 11 tabs

  14. Laser surveillance system for spent fuel

    International Nuclear Information System (INIS)

    Fiarman, S.; Zucker, M.S.; Bieber, A.M. Jr.

    1980-01-01

    A laser surveillance system installed at spent fuel storage pools (SFSP's) will provide the safeguard inspector with specific knowledge of spent fuel movement that cannot be obtained with current surveillance systems. The laser system will allow for the division of the pool's spent fuel inventory into two populations - those assemblies which have been moved and those which haven't - which is essential for maximizing the efficiency and effectiveness of the inspection effort. We have designed, constructed, and tested a full size laser system operating in air and have used an array of 6 zircaloy BWR tubes to simulate an assembly. The reflective signal from the zircaloy rods is a strong function of position of the assembly, but in all cases is easily discernable from the reference scan of the background with no assembly. A design for a SFSP laser surveillance system incorporating laser ranging is discussed. 10 figures

  15. Spent fuel storage and transportation - ANSTO experience

    International Nuclear Information System (INIS)

    Irwin, Tony

    2002-01-01

    The Australian Nuclear Science and Technology Organisation (ANSTO) has operated the 10 MW DIDO class High Flux Materials Test Reactor (HIFAR) since 1958. Refuelling the reactor produces about 38 spent fuel elements each year. Australia has no power reactors and only one operating research reactor so that a reprocessing plant in Australia is not an economic proposition. The HEU fuel for HIFAR is manufactured at Dounreay using UK or US origin enriched uranium. Spent fuel was originally sent to Dounreay, UK for reprocessing but this plant was shutdown in 1998. ANSTO participates in the US Foreign Research Reactor Spent Fuel Return program and also has a contract with COGEMA for the reprocessing of non-US origin fuel

  16. Spent Nuclear Fuel Alternative Technology Decision Analysis

    International Nuclear Information System (INIS)

    Shedrow, C.B.

    1999-01-01

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology

  17. Evolution of spent fuel dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Standring, Paul Nicholas [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Fuel Cycle and Waste Technology; Takats, Ferenc [TS ENERCON KFT, Budapest (Hungary)

    2016-11-15

    Around 10,000 tHM of spent fuel is discharged per year from the nuclear power plants in operation. Whilst the bulk of spent fuel is still held in at reactor pools, 24 countries have developed storage facilities; either on the reactor site or away from the reactor site. Of the 146 operational AFR storage facilities about 80 % employ dry storage; the majority being deployed over the last 20 years. This reflects both the development of dry storage technology as well as changes in politics and trading relationships that have affected spent fuel management policies. The paper describes the various approaches to the back-end of the nuclear fuel cycle for power reactor fuels and provides data on deployed storage technologies.

  18. A combined XAFS, ESI TOF-MS and LIBD study on the formation of polynuclear Zr(IV), Th(IV) and Pu(IV) species

    Science.gov (United States)

    Rothe, J.; Walther, C.; Brendebach, B.; Büchner, S.; Fuss, M.; Denecke, M. A.; Geckeis, H.

    2009-11-01

    The long term radiotoxicity of spent nuclear fuel disposed of in deep underground repositories after discharge from nuclear power reactors is determined by actinide elements, mainly plutonium. Water intrusion into the repository might cause container corrosion and leaching of the waste matrices, leading to the release of Pu and other actinides into the geological environment. Performance assessment for a future nuclear waste repository requires detailed knowledge on actinide aqueous chemistry in the aquifer surrounding the disposal site. Tetravalent actinides exhibit a strong tendency towards hydrolysis and subsequent polymerization and/or colloid formation. These species provide a potential pathway for migration of actinides away from the repository. Therefore, it is of fundamental interest to study their generation and properties in-situ. To this end, X-ray Absorption Fine Structure Spectroscopy (XAFS) at the INE-Beamline for actinide research at ANKA, Electrospray Mass-Spectrometry (ESI TOF-MS) and Laser Induced Breakdown Detection (LIBD) are combined at FZK-INE in a comprehensive attempt to characterize Zr(IV) (An(IV) analogue), Th(IV) and Pu(IV) polymerization and colloid formation.

  19. Application of the 'GammaGen' Computer Code for NORM Synthetic Spectra Analysis

    International Nuclear Information System (INIS)

    Sarusi, B.; Levinson, S.; German, U.; Antropov, S.; Kovler, K.

    2014-01-01

    'GammaGen' (GG) is a computer software developed to simulate gamma ray spectra obtained from NaI(Tl) or Ge detectors. The detector efficiency, resolution and peak to Compton ratios are used to generate synthetic pulse height spectra for specific detector and geometrical configurations. GG was used in the past for several applications, as to predict the detector response to different radio-nuclides mixtures, to predict the pulse height spectra near a nuclear spent fuel site as a function of the shielding thickness and cooling time and for homeland security purposes. An updated version of the GG program was developed, and the application presented in the present work consists of building a library and simulating the detection and analysis of synthetic spectra to check the performance of a commercial spectrometry system based on a NaI(Tl) detector intended for NORM analysis, which was developed by Amplituda/Russia

  20. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  1. Acceptance of spent fuel of varying characteristics

    International Nuclear Information System (INIS)

    Short, S.M.

    1990-03-01

    This paper is a preliminary overview of a study with the primary objective of establishing a set of acceptance selection criteria and corresponding spent fuel characteristics to be incorporated as a component of requirements for the Federal Waste Management System (FWMS). A number of alternative acceptance allocations and selection rules were analyzed to determine the operational sensitivity of each element of the FWMS to the resultant spent fuel characteristics. Preliminary recommendations of the study include three different sets of selection rules to be included in the FWMS design basis. 2 refs., 4 figs., 4 tabs

  2. Classification of spent nuclear fuel (SNF)

    International Nuclear Information System (INIS)

    1990-03-01

    This report is one of a series of eight prepared by E. R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high-priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. This document discusses the classification of spent nuclear fuels

  3. Transporting spent nuclear fuel: an overview

    International Nuclear Information System (INIS)

    1986-03-01

    Although high-level radioactive waste from both commercial and defense activities will be shipped to the repository, this booklet focuses on various aspects of transporting commercial spent fuel, which accounts for the majority of the material to be shipped. The booklet is intended to give the reader a basic understanding of the following: the reasons for transportation of spent nuclear fuel, the methods by which it is shipped, the safety and security precautions taken for its transportation, emergency response procedures in the event of an accident, and the DOE program to develop a system uniquely appropriate to NWPA transportation requirements

  4. Conditioning of spent mercury by amalgamation

    International Nuclear Information System (INIS)

    Yim, S. P.; Shon, J. S.; An, B. G.; Lee, H. J.; Lee, J. W.; Ji, C. G.; Kim, S. H.; Yoon, J. H.; Yang, M. S.

    2002-01-01

    Solidification by amalgamation was performed to immobilize and stabilize the liquid spent mercury. First, the appropriate metal and alloy which can convert liquid mercury into a solid form of amalgam were selected through initial tests. The amalgam form, formulated in optimum composition, was characterized and subjected to performance tests including compressive strength, water immersion, leachability and initial vaporization rate to evaluate mechanical integrity, durability and leaching properties. Finally, bench scale amalgamation trial was conducted with about 1 kg of spent mercury to verify the feasibility of amalgamation method

  5. Biodiesel Production from Spent Coffee Grounds

    Directory of Open Access Journals (Sweden)

    Blinová Lenka

    2017-06-01

    Full Text Available The residue after brewing the spent coffee grounds is an oil-containing waste material having a potential of being used as biodiesel feedstock. Biodiesel production from the waste coffee grounds oil involves collection and transportation of coffee residue, drying, oil extraction, and finally production of biodiesel. Different methods of oil extraction with organic solvents under different conditions show significant differences in the extraction yields. In the manufacturing of biodiesel from coffee oil, the level of reaction completion strongly depends on the quality of the feedstock oil. This paper presents an overview of oil extraction and a method of biodiesel production from spent coffee grounds.

  6. Integrated spent nuclear fuel database system

    International Nuclear Information System (INIS)

    Henline, S.P.; Klingler, K.G.; Schierman, B.H.

    1994-01-01

    The Distributed Information Systems software Unit at the Idaho National Engineering Laboratory has designed and developed an Integrated Spent Nuclear Fuel Database System (ISNFDS), which maintains a computerized inventory of all US Department of Energy (DOE) spent nuclear fuel (SNF). Commercial SNF is not included in the ISNFDS unless it is owned or stored by DOE. The ISNFDS is an integrated, single data source containing accurate, traceable, and consistent data and provides extensive data for each fuel, extensive facility data for every facility, and numerous data reports and queries

  7. Biodiesel Production from Spent Coffee Grounds

    Science.gov (United States)

    Blinová, Lenka; Bartošová, Alica; Sirotiak, Maroš

    2017-06-01

    The residue after brewing the spent coffee grounds is an oil-containing waste material having a potential of being used as biodiesel feedstock. Biodiesel production from the waste coffee grounds oil involves collection and transportation of coffee residue, drying, oil extraction, and finally production of biodiesel. Different methods of oil extraction with organic solvents under different conditions show significant differences in the extraction yields. In the manufacturing of biodiesel from coffee oil, the level of reaction completion strongly depends on the quality of the feedstock oil. This paper presents an overview of oil extraction and a method of biodiesel production from spent coffee grounds.

  8. The influence of near field hydrogen on actinide solubilities and spent fuel leaching

    International Nuclear Information System (INIS)

    Spahiu, K.; Werme, L.; Eklund, U.B.

    2000-01-01

    Large amounts of hydrogen are produced as a result of the anoxic corrosion of iron in the proposed container materials for some geologic repositories. Another hydrogen source, less important than the anoxic corrosion of iron, is the radiolysis of water by the spent fuel radiation. Gas phase formation occurs when the pressure of the hydrogen equals at least the hydrostatic pressure, around 5 MPa at 500 meters depth. The effects of 5 MPa hydrogen pressure on spent PWR fuel leaching and on uranium oxide solubility have been studied in carbonated solutions at 70 C. The experiments were performed in a 1 liter autoclave, filled with 950 ml of a solution 10 mM NaCl, 2 mM NaHCO 3 and with hydrogen at a pressure of 5 MPa in the remaining 50 ml free volume. The leaching behavior of 2 g PWR spent fuel powder of the 0.25-0.50 mm fraction, placed in a gold basket was studied during several months by analyzing 10 ml solution samples taken after regular time intervals. A few experiments were performed also with unirradiated U(IV) oxide. In both cases extremely low concentrations of uranium (less than 10 -9 M) were measured in the solution samples. Furthermore the uranium levels in solution remained practically constant during the whole leaching period (more than one year), indicating the absence of any oxidative dissolution of the spent fuel matrix. The same conclusion is confirmed by the constant (within analytical errors) levels of strontium, cesium, molybdenum, iodine and technetium during the whole leaching period. These results have been compared with the ones obtained during the leaching of a spent fuel pin in anoxic conditions, where the uranium and other radionuclides levels are several orders of magnitude higher. The surface of spent fuel or U(IV) oxide is partially oxidized during storage, giving rise to relatively high levels of U(VI) in solution even during leaching in anoxic conditions. No such effect could be observed in the presence of 5 MPa hydrogen, indicating

  9. Sorption of radionuclides from spent fuel in crystalline rocks

    International Nuclear Information System (INIS)

    Nikula, A.

    1982-10-01

    The safe disposal of spent nuclear fuel or reprocessed waste is an essential element in the expansion of the nuclear power industry. Stable rock formations e.g. granite are considered to be potential sites for disposal. A major factor in evaluating the degree of safety of the disposal is the sorption of radionuclides in rock, which affects their retardation. The report considers the chemical forms of the hazardous radionuclides of spent nuclear fuel in groundwater and the effects of the water's properties on them. In the groundwater near the Olkiluoto power plant site cesium, strontium and radium are in cationic form, iodine as I - . Technetium would occur as TcO +2 , but the pertechnetate form is also possible. Uranium most probably would be as U(VI) plutonium and neptunium as Np(IV) or Np(V). The valences for thorium, americium and curium are not changed in this groundwater and would be +4, +3 and +3, respectively. The actinides in groundwater are all in hydrated or complex form. An increase on the ionic stregth of the groundwater in most instances causes a decrease in the sorption of nuclides since the ion exchange capacity of the rock is limited. Anionic ligands also decrease sorption of cations by complex formation. In some case, on the other hand, high salt concentrations may cause formation of radiocolloids of lanthanides and neptunium and thus increase sorption. In all cases the degree of sorption described by the distribution ratio Ksub(d) was influenced by the pH of the groundwater. Sorption of cesium and strontium increased with growing pH. The sorption behaviour of actinides was in positive correlation with formation of hydroxide complexes at different pH values. The Ksub(d) values of Cs, Sr, Co, Ni and Am for Olkiluoto granites were found to agree with Swedish values, also determined at ambient atmospheric conditions

  10. Aconselhamento genético Genetic counseling

    Directory of Open Access Journals (Sweden)

    João Monteiro de Pina-Neto

    2008-08-01

    Full Text Available OBJETIVO: Esta revisão sobre aconselhamento genético (AG teve o objetivo de mostrar os conceitos atuais e os princípios filosóficos e éticos aceitos na grande maioria dos países e recomendados pela Organização Mundial da Saúde, as fases do processo, seus resultados e o impacto psicológico de uma doença genética em uma família. FONTES DOS DADOS: Os conceitos apresentados são baseados em uma síntese histórica da literatura sobre AG desde a década de 1930 até o momento atual, sendo que os artigos citados representam os principais trabalhos publicados e que hoje fundamentam a teoria e a prática do AG. SÍNTESE DOS DADOS: O AG modernamente é definido como um processo de comunicação que trata dos problemas humanos relacionados à ocorrência de uma doença genética em uma família. É fundamental que os profissionais da saúde conheçam os aspectos psicológicos desencadeados pela doença genética e como estes aspectos podem ser manejados. Vivemos ainda na genética humana e médica uma fase de predomínio dos aspectos técnicos e científicos e de pouca ênfase no estudo das reações emocionais e dos processos de adaptação das pessoas a estas doenças, o que leva ao baixo entendimento dos clientes sobre os fatos ocorridos, com conseqüências negativas sobre a vida familiar e para a sociedade. CONCLUSÕES: Conclui-se pela necessidade de que as famílias com doenças genéticas sejam encaminhadas para AG e que os profissionais desta área invistam mais na humanização do atendimento, desenvolvendo mais as técnicas do AG psicológico não-diretivo.OBJECTIVE: The objective of this review of genetic counseling (GC is to describe the current concepts and philosophical and ethical principles accepted by the great majority of countries and recommended by the World Health Organization, the stages of the process, its results and the psychological impact that a genetic disease has on a family. SOURCES: The concepts presented are

  11. TidGen Power System Commercialization Project

    Energy Technology Data Exchange (ETDEWEB)

    Sauer, Christopher R. [President & CEO; McEntee, Jarlath [VP Engineering & CTO

    2013-12-30

    ORPC Maine, LLC, a wholly-owned subsidiary of Ocean Renewable Power Company, LLC (collectively ORPC), submits this Final Technical Report for the TidGen® Power System Commercialization Project (Project), partially funded by the U.S. Department of Energy (DE-EE0003647). The Project was built and operated in compliance with the Federal Energy Regulatory Commission (FERC) pilot project license (P-12711) and other permits and approvals needed for the Project. This report documents the methodologies, activities and results of the various phases of the Project, including design, engineering, procurement, assembly, installation, operation, licensing, environmental monitoring, retrieval, maintenance and repair. The Project represents a significant achievement for the renewable energy portfolio of the U.S. in general, and for the U.S. marine hydrokinetic (MHK) industry in particular. The stated Project goal was to advance, demonstrate and accelerate deployment and commercialization of ORPC’s tidal-current based hydrokinetic power generation system, including the energy extraction and conversion technology, associated power electronics, and interconnection equipment capable of reliably delivering electricity to the domestic power grid. ORPC achieved this goal by designing, building and operating the TidGen® Power System in 2012 and becoming the first federally licensed hydrokinetic tidal energy project to deliver electricity to a power grid under a power purchase agreement in North America. Located in Cobscook Bay between Eastport and Lubec, Maine, the TidGen® Power System was connected to the Bangor Hydro Electric utility grid at an on-shore station in North Lubec on September 13, 2012. ORPC obtained a FERC pilot project license for the Project on February 12, 2012 and the first Maine Department of Environmental Protection General Permit issued for a tidal energy project on January 31, 2012. In addition, ORPC entered into a 20-year agreement with Bangor Hydro Electric

  12. Genética molecular del alcoholismo

    OpenAIRE

    Mauricio Rey-Buitrago

    2015-01-01

    El alcoholismo es una patología psiquiátrica compleja y de origen multifactorial en la que el factor genético explica alrededor del 50 % del fenómeno. Son numerosos los genes que se han asociado a esta enfermedad, pero su aporte individual es mínimo y contradictorio. Estos genes operan a través de características intermedias como la impulsividad y la sensibilidad al alcohol, lo que hace compleja la definición del fenotipo del alcoholismo. Los estudios de asociación de SNPs, de asociación a to...

  13. Casette for storage of spent fuel assemblies

    International Nuclear Information System (INIS)

    Ericsson, S.

    1992-01-01

    Describes a design of a casette for spent fuel storage in a fuelstorage pool. The new design, based on flexible spacers, allows the fuel assemblies to be packed more compact and the fuel storage pool used in a more economic way

  14. Total quality in spent fuel pool reracking

    International Nuclear Information System (INIS)

    Cranston, J.S.; Bradbury, R.B.; Cacciapouti, R.J.

    1993-01-01

    The nuclear utility environment is one of strict cost control under prescriptive regulations and increasing public scrutiny. This paper presents the results of A Total Quality approach, by a dedicated team, that addresses the need for increased on-site spent fuel storage in this environment. Innovations to spent fuel pool reracking, driven by utilities' specific technical needs and shrinking budgets, have resulted in both product improvements and lower prices. A Total Quality approach to the entire turnkey project is taken, thereby creating synergism and process efficiency in each of the major phases of the project: design and analysis, licensing, fabrication, installation and disposal. Specific technical advances and the proven quality of the team members minimizes risk to the utility and its shareholders and provides a complete, cost effective service. Proper evaluation of spent fuel storage methods and vendors requires a full understanding of currently available customer driven initiatives that reduce cost while improving quality. In all phases of a spent fuel reracking project, from new rack design and analysis through old rack disposal, the integration of diverse experts, at all levels and throughout all phases of a reracking project, better serves utility needs. This Total Quality environment in conjunction with many technical improvements results in a higher quality product at a lower cost

  15. PROTEIN ENRICHMENT OF SPENT SORGHUM RESIDUE USING ...

    African Journals Online (AJOL)

    BSN

    The optimum concentration of spent sorghum for protein enrichment with S. cerevisiae was 7.Sg/100 ml. Th.: protein ... production of single sell protein using Candida utilis and cassava starch effluem as substrate. ... wastes as substrates, Kluyveromyces fragilis and milk whey coconut water as substrate (Rahmat et al.,. 1995 ...

  16. Magnetically modified spent grain for dye removal

    Czech Academy of Sciences Publication Activity Database

    Šafařík, Ivo; Horská, Kateřina; Šafaříková, Miroslava

    2011-01-01

    Roč. 53, č. 1 (2011), s. 78-80 ISSN 0733-5210 R&D Projects: GA MŠk OC09052; GA MPO 2A-1TP1/094 Institutional research plan: CEZ:AV0Z60870520 Keywords : Spent grain * Magnetic fluid * Adsorption * Dyes Subject RIV: GM - Food Processing Impact factor: 2.073, year: 2011

  17. IMPACTS OF DIFFERENT CONCENTRATIONS OF SPENT ...

    African Journals Online (AJOL)

    DJFLEX

    The impacts of different concentrations of spent carbide waste on the growth and yield of Zea mays Linn. (maize) and Arachis hypogea Linn. (groundnut) were studied at the screen house, Botanic garden, University of Port. Harcourt, Nigeria. The crops were planted in carbide waste concentrations of 40g, 80g, 120g, and ...

  18. ADSORPTION ON HEAT REGENERATED SPENT BLEACHING ...

    African Journals Online (AJOL)

    Preferred Customer

    KINETICS AND THERMODYNAMICS OF AQUEOUS Cu(II) ADSORPTION ON. HEAT REGENERATED SPENT BLEACHING EARTH. Enos W. Wambu1*, Gerald K. Muthakia2*, Joseph K. wa-Thiong'o1 and Paul M. Shiundu3. 1Department of Chemistry, Kenyatta University, P.O. Box 43844-00100, Nairobi, Kenya.

  19. Spent fuel management: Current status and prospects

    International Nuclear Information System (INIS)

    1988-12-01

    The main objective of the Advisory Group on Spent Fuel Management is to review the world-wide situation in Spent Fuel Management, to define the most important directions of national efforts and international cooperation in this area, to exchange information on the present status and progress in performing the back-end of Nuclear Fuel Cycle and to elaborate the general recommendations for future Agency programmes in the field of spent fuel management. This report which is a result of the third IAEA Advisory Group Meeting (the first and second were held in 1984 and 1986) is intended to provide the reader with an overview of the status of spent fuel management programmes in a number of leading countries, with a description of the past and present IAEA activities in this field of Nuclear Fuel Cycle and with the Agency's plans for the next years, based on the proposals and recommendations of Member States. A separate abstract was prepared for each of 14 papers presented at the advisory group meeting. Refs, figs and tabs

  20. Safeguardability of advanced spent fuel conditioning process

    Energy Technology Data Exchange (ETDEWEB)

    Li, T. K. (Tien K.); Lee, S. Y. (Sang Yoon); Burr, Tom; Russo, P. A. (Phyllis A.); Menlove, Howard O.; Kim, H. D.; Ko, W. I. (Won Il); Park, S. W.; Park, H. S.

    2004-01-01

    The Advanced Spent Fuel Conditioning Process (ACP) is an electro-metallurgical treatment technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology since 1977 for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. By using of this technology, a significant reduction of the volume and heat load of spent fuel is expected, which would lighten the burden of final disposal in terms of disposal size, safety and economics. In the framework of collaboration agreement to develop the safeguards system for the ACP, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and the KAERI since 2002. In this study, the safeguardability of the ACP technology was examined for the pilot-scale facility. The process and material flows were conceptually designed, and the uncertainties in material accounting were estimated with international target values.

  1. Drying of mock spent nuclear fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Crepeau, J.C.; Reese, S.; McIlroy, H.M. Jr. [Univ. of Idaho, Idaho Falls, ID (United States). Dept. of Mechanical Engineering; Lords, R.E. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States)

    1998-03-01

    Spent nuclear fuel elements are stored in underwater cooling pools until the elements can be safely handled and prepared for interim dry storage. The fuel was intended for short-term storage in water before it was to be reprocessed. However, the fuel will no longer be reprocessed, and extended storage in water has caused many of the aluminum-clad elements to degrade, exposing the uranium fuel. In addition, sludge, comprised of corroded aluminum and sediment, has accumulated in and around the fuel plates. The water in the sludge must be removed before the spent fuel elements can be placed in dry storage. Experiments have been performed on mock spent fuel elements with simulated corrosion product applied between the plates. A series of vacuum and heating cycles were used to dry the elements, and a mixture of clay and aluminum oxide was used to simulate corrosion products on the elements. The procedures used in the experiments were determined to be adequate to dry the mock spent fuel elements, and the temperature behavior of the simulated corrosion product within the fuel elements could be used to determine when the element was dry. On plates where areas of wet simulant were found, a sharp frying front was observed that separated the wet and dry parts of the simulated corrosion product. The drying front propagated inward towards the center of the mock fuel elements over time.

  2. Spent fuel and waste inventories and projections

    International Nuclear Information System (INIS)

    Carter, W.L.; Finney, B.C.; Alexander, C.W.; Blomeke, J.O.; McNair, J.M.

    1980-08-01

    Current inventories of commercial spent fuels and both commercial and US Department of Energy radioactive wastes were compiled, based on judgments of the most reliable information available from Government sources and the open literature. Future waste generation rates and quantities to be accumulated over the remainder of this century are also presented, based on a present projection of US commercial nuclear power growth and expected defense-related activities. Spent fuel projections are based on the current DOE/EIA estimate of nuclear growth, which projects 180 GW(e) in the year 2000. It is recognized that the calculated spent fuel discharges are probably high in view of recent reactor cancellations; hence adjustments will be made in future updates of this report. Wastes considered, on a chapter-by-chapter basis, are: spent fuel, high-level wastes, transuranic wastes, low-level wastes, mill tailings (active sites), and remedial action wastes. The latter category includes mill tailings (inactive sites), surplus facilities, formerly utilized sites, and the Grand Junction Project. For each category, waste volume inventories and projections are given through the year 2000. The land usage requirements are given for storage/disposal of low-level and transuranic wastes, and for present inventories of mill tailings

  3. Older peoples' perspectives on time spent alone.

    Science.gov (United States)

    Stanley, Mandy; Richard, Ashley; Williams, Shoshannah

    2017-06-01

    Large amounts of time spent alone by older people have been associated with loneliness and poor mental and physical health. There is a paucity of research, however, that examines time alone from an occupational perspective. In this exploratory study we explored the perspectives of older people on their time spent alone. A qualitative descriptive study design was selected. With the aim of maximising variation, five participants were recruited from retirement villages and seven participants who lived independently in the community. Participants recorded time spent alone in a time diary for three days as priming for a semi-structured in-depth interview. Transcripts were analysed thematically. Three key themes were identified: 'it is a matter of getting some balance'; 'keeping busy'; and 'the nights are the worst'. The study highlights the importance older people place on the need to manage time alone so that it is a positive and nourishing experience and to avoid experiencing extended periods of boredom potentially leading to loneliness. Older people utilise occupations to keep busy and achieve an individually acceptable level of time alone. Enabling older people to balance time spent alone by addressing barriers to participation in the community in addition to finding engaging occupations to occupy time has the potential to prevent boredom, loneliness and improve wellbeing. © 2016 Occupational Therapy Australia.

  4. Time spent on television in European countries

    NARCIS (Netherlands)

    Vergeer, M.R.M.; Coenders, M.T.A.; Scheepers, P.L.H.; Konig, R.P.; Nelissen, P.W.M.; Huysmans, F.J.M.

    2009-01-01

    This study aims to explain the variation in time spent on watching television in 15 European Union countries, using determinants defined at the individual level, and characteristics defined at the national level, such as the number of channels and nature of the television supply. The results of the

  5. Regional spent fuel storage facility (RSFSF)

    International Nuclear Information System (INIS)

    Dyck, H.P.

    1999-01-01

    The paper gives an overview of the meetings held on the technology and safety aspects of regional spent fuel storage facilities. The questions of technique, economy and key public and political issues will be covered as well as the aspects to be considered for implementation of a regional facility. (author)

  6. Operation of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide was prepared as part of the IAEA's programme on safety of spent fuel storage. This is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes key activities in the operation of spent fuel storage facilities. Section 3 lists the basic safety considerations for storage facility operation, the fundamental safety objectives being subcriticality, heat removal and radiation protection. Recommendations for organizing the management of a facility are contained in Section 4. Section 5 deals with aspects of training and qualification; Section 6 describes the phases of the commissioning of a spent fuel storage facility. Section 7 describes operational limits and conditions, while Section 8 deals with operating procedures and instructions. Section 9 deals with maintenance, testing, examination and inspection. Section 10 presents recommendations for radiation and environmental protection. Recommendations for the quality assurance (QA) system are presented in Section 11. Section 12 describes the aspects of safeguards and physical protection to be taken into account during operations; Section 13 gives guidance for decommissioning. 15 refs, 5 tabs

  7. Spent nuclear fuel project product specification

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    1999-01-01

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project

  8. Conditioning experience for spent radium sources

    International Nuclear Information System (INIS)

    Kang, I. S.; Shon, J. S.; Kim, K. J.; Min, D. K.

    2001-01-01

    In order to avoid accidents that could be resulted from improper storage of spent radium sources, it is necessary to condition and store them safely. The program for safe conditioning of spent radium sources by IAEA has been established to assist the developing countries. The main object of this paper is to apply the technology that was adapted by IAEA for the conditioning the national inventory of Ra-226 sources in member states, as a part of IAEA's project with the Korean expert team. This paper is the result that the Korean expert team carried out spent radium conditioning, under the project title 'Radium Conditioning in Myanmar(INT4131-06646C)'. The whole inventory of spent radium sources 1,429.5mCi, was safely conditioned by the Korean expert team according to the manual under the supervision of IAEA's technical officer and the control of Myanmar authority on behalf of Myanmar. These sources were encapsuled and welded into 27 small capsules and 3 large capsules, and conditioned in 3 lead shields, producing 3 concrete-shielded drums. The inventories were distributed into 3 shielding devices, holding 500mCi, 459.5mCi, and 470mCi

  9. Biodegradation of hydrocarbons exploiting spent substrate from ...

    African Journals Online (AJOL)

    In Acatzingo, Puebla, Mexico (east-central), oil spills have mainly affected agricultural fields. Pleurotus ostreatus is a white rot basidiomycete and produces extracellular enzymes (lacasses, manganese peroxidases, versatile peroxidases and veratryl alcohol oxidases). The production of edible mushrooms generates spent ...

  10. Development of advanced spent fuel management process. System analysis of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Ro, S.G.; Kang, D.S.; Seo, C.S.; Lee, H.H.; Shin, Y.J.; Park, S.W.

    1999-03-01

    The system analysis of an advanced spent fuel management process to establish a non-proliferation model for the long-term spent fuel management is performed by comparing the several dry processes, such as a salt transport process, a lithium process, the IFR process developed in America, and DDP developed in Russia. In our system analysis, the non-proliferation concept is focused on the separation factor between uranium and plutonium and decontamination factors of products in each process, and the non-proliferation model for the long-term spent fuel management has finally been introduced. (Author). 29 refs., 17 tabs., 12 figs

  11. Spent Fuel Storage Operation - Lessons Learned

    International Nuclear Information System (INIS)

    2013-12-01

    Experience gained in planning, constructing, licensing, operating, managing and modifying spent fuel storage facilities in some Member States now exceeds 50 years. Continual improvement is only achieved through post-project review and ongoing evaluation of operations and processes. This publication is aimed at collating and sharing lessons learned. Hopefully, the information provided will assist Member States that already have a developed storage capability and also those considering development of a spent nuclear fuel storage capability in making informed decisions when managing their spent nuclear fuel. This publication is expected to complement the ongoing Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III); the scope of which prioritizes facility operational practices in lieu of fuel and structural components behaviour over extended durations. The origins of the current publication stem from a consultants meeting held on 10-12 December 2007 in Vienna, with three participants from the IAEA, Slovenia and USA, where an initial questionnaire on spent fuel storage was formulated (Annex I). The resultant questionnaire was circulated to participants of a technical meeting, Spent Fuel Storage Operations - Lessons Learned. The technical meeting was held in Vienna on 13-16 October 2008, and sixteen participants from ten countries attended. A consultants meeting took place on 18-20 May 2009 in Vienna, with five participants from the IAEA, Slovenia, UK and USA. The participants reviewed the completed questionnaires and produced an initial draft of this publication. A third consultants meeting took place on 9-11 March 2010, which six participants from Canada, Hungary, IAEA, Slovenia and the USA attended. The meeting formulated a second questionnaire (Annex II) as a mechanism for gaining further input for this publication. A final consultants meeting was arranged on 20-22 June 2011 in Vienna. Six participants from Hungary, IAEA, Japan

  12. Policy-induced market introduction of Generation IV reactor systems

    International Nuclear Information System (INIS)

    Heek, Aliki Irina van; Roelofs, Ferry

    2011-01-01

    Almost 10 years ago the U.S. Department of Energy (DOE) started the Generation IV Initiative (GenIV) with 9 other national governments with a positive ground attitude towards nuclear energy. Some of these Generation IV systems, like the fast reactors, are nearing the demonstration stage. The question on how their market introduction will be implemented becomes increasingly urgent. One main topic for future reactor technologies is the treatment of radioactive waste products. Technological solutions to this issue are being developed. One possible process is the transformation of long-living radioactive nuclides into short living ones; a process known as transmutation, which can be done in a nuclear reactor only. Various Generation IV reactor concepts are suitable for this process, and of these systems most experience has been gained with the sodium-cooled fast reactor (SFR). However, both these first generation SFR plants and their Generation IV successors are designed as electricity generating plants, and therefore supposed to be commercially viable in the electricity markets. Various studies indicate that the generation costs of a combined LWR-(S)FR nuclear generating park (LWR: light water reactor) will be higher than that of an LWR-only park. To investigate the effects of the deployment of the different reactors and fuel cycles on the waste produced, resources used and costs incurred as a function of time, a dynamic fuel cycle assessment is performed. This study will focus on the waste impact of the introduction of a fraction of fast reactors in the European nuclear reactor park with a cost increase as described in the previous paragraph. The nuclear fuel cycle scenario code DANESS is used for this, as well as the nuclear park model of the EU-27 used for the previous study. (orig.)

  13. The psychosocial consequences of spent fuel disposal

    International Nuclear Information System (INIS)

    Paavola, J.; Eraenen, L.

    1999-03-01

    In this report the potential psychosocial consequences of spent fuel disposal to inhabitants of a community are assessed on the basis of earlier research. In studying the situation, different interpretations and meanings given to nuclear power are considered. First, spent fuel disposal is studied as fear-arousing and consequently stressful situation. Psychosomatic effects of stress and coping strategies used by an individual are presented. Stress as a collective phenomenon and coping mechanisms available for a community are also assessed. Stress reactions caused by natural disasters and technological disasters are compared. Consequences of nuclear power plant accidents are reviewed, e.g. research done on the accident at Three Mile Island power plant. Reasons for the disorganising effect on a community caused by a technological disaster are compared to the altruistic community often seen after natural disasters. The potential reactions that a spent fuel disposal plant can arouse in inhabitants are evaluated. Both short-term and long-term reactions are evaluated as well as reactions under normal functioning, after an incident and as a consequence of an accident. Finally an evaluation of how the decision-making system and citizens' opportunity to influence the decision-making affect the experience of threat is expressed. As a conclusion we see that spent fuel disposal can arouse fear and stress in people. However, the level of the stress is probably low. The stress is at strongest at the time of the starting of the spent fuel disposal plant. With time people get used to the presence of the plant and the threat experienced gets smaller. (orig.)

  14. Numerical Estimation of the Spent Fuel Ratio

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilke, Jason [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Margraf, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunn, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-01-01

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. However, the validity of these large- scale results remain in question due to the lack of a defensible spent fuel ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

  15. Genética molecular del alcoholismo

    Directory of Open Access Journals (Sweden)

    Mauricio Rey Buitrago

    2015-07-01

    Full Text Available El alcoholismo es una patología psiquiátrica compleja y de origen multifactorial en donde el factor genético explica alrededor del 50 % del fenómeno. Son numerosos los genes que se han asociado a la enfermedad, pero su aporte individual es mínimo y contradictorio. Estos genes operan a través de características intermedias como la impulsividad y sensibilidad al alcohol lo que hace compleja la definición del fenotipo del alcoholismo. Los estudios de asociación de SNP`s, asociación a todo el genoma, de expresión y epigenéticos han identificado una amplia gama de variantes genéticas y epigenéticas blancos para los estudios de susceptibilidad, diagnóstico y tratamiento farmacológico. Actualmente se comprende mucho más estas relaciones y el rápido desarrollo de nuevas metodologías de estudio promete continuar este proceso y la generación de algoritmos de diagnóstico, prevención y tratamientos más acertados y confiables.

  16. Caracterização genético-clínica de pacientes com fenilcetonúria no Estado de Alagoas = Genetic and clinical characterization of patients with phenylketonuria in Alagoas state, Brazil

    Directory of Open Access Journals (Sweden)

    Santos, Emerson Santana

    2012-01-01

    Conclusões: O genótipo V388M/IVS10nt11G>A foi o mais prevalente. Trinta por cento dos pacientes foram sintomáticos, provavelmente pela natureza das mutações, não adesão ao tratamento, tratamento inadequado e/ou diagnóstico tardio

  17. The Spent Fuel Management in Finland and Modifications of Spent Fuel Storages

    International Nuclear Information System (INIS)

    Maaranen, Paeivi

    2014-01-01

    The objective of this presentation is to share the Finnish regulator's (STUK) experiences on regulatory oversight of the enlargement of a spent fuel interim storage. An overview of the current situation of spent fuel management in Finland will also be given. In addition, the planned modifications and requirements set for spent fuel storages due to the Fukushima accident are discussed. In Finland, there are four operating reactors, one under construction and two reactors that have a Council of State's Decision-in-Principle to proceed with the planning and licensing of a new reactor. In Olkiluoto, the two operating ASEA-Atom BWR units and the Areva EPR under construction have a shared interim storage for the spent fuel. The storage was designed and constructed in 1980's. The option for enlarging the storage was foreseen in the original design. Considering three operating units to produce their spent fuel and the final disposal to begin in 2022, extra space in the spent fuel storage is estimated to be needed in around 2014. The operator decided to double the number of the spent fuel pools of the storage and the construction began in 2010. The capacity of the enlarged spent fuel storage is considered to be sufficient for the three Olkiluoto units. The enlargement of the interim storage was included in Olkiluoto NPP 1 and 2 operating license. The licensing of the enlargement was conducted as a major plant modification. The operator needed the approval from STUK to conduct the enlargement. Prior to the construction of this modification, the operator was required to submit the similar documentation as needed for applying for the construction license of a nuclear facility. When conducting changes in an old nuclear facility, the new safety requirements have to be followed. The major challenge in the designing the enlargement of the spent fuel storage was to modify it to withstand a large airplane crash. The operator chose to cover the pools with protecting slabs and also to

  18. National report of the Slovak Republic compiled in terms of the Join on the safety of spent fuel management and on the safety of radwaste management

    International Nuclear Information System (INIS)

    Jurina, V.; Viktory, D.; Petrik, T.; Sovcik, J.; Suess, J.; Tomek, J.; Lukacovic, J.; Ivan, J.; Ziakova, M.; Metke, E.; Pospisil, M.; Turner, M.; Homola, J.; Vaclav, J.; Bystricka, S.; Barbaric, M.; Horvath, J.; Betak, J.; Mihaly, B.; Adamovsky, V.; Baloghova, A.; Orihel, M.; Vasina, D.; Balaz, J.; Misovicova, D.; Vrtoch, M.; Mlcuch, J.; Granak, P.; Meleg, J.; Bardy, M.; Gogoliak, J.

    2011-08-01

    The National safety report of the Slovak Republic on the safety of spent fuel management and on the safety of radwaste management in 2011 is presented. These activities in the safety of spent fuel management and radioactive waste management in the Slovak Republic are reported under the headings: (A) Introduction; B) Concept for spent nuclear fuel management (SNF) and radwaste management (RAW); (C) Scope of application of the convention; (D) Spent fuel management and radioactive waste (RAW) management facilities; (E) Legislation and regulation; (F) General safety provisions; (G) Safety of spent fuel management; (H) Safety of radioactive waste (RAW) management; (I) Transboundary movement of spent nuclear fuel and radioactive waste; (J) Disused sealed sources; (K) Planned measures to improve safety; (L) Communication with the public; (M) Annexes. Annexes consists of following parts: I. List of nuclear facilities for spent fuel and RAW management. II. Limits of radioactive material discharges into atmosphere and hydrosphere. III. List of nuclear installations in decommissioning. IV. Inventory of stored spent nuclear fuel. V. Inventory of stored RAW. VI. List of national laws, decrees and guidelines. VII. List of international expert reports (including safety reports). VIII. List of authors.

  19. National report of the Slovak Republic compiled in terms of the join convention on the safety of spent fuel management and on the safety of radwaste management

    International Nuclear Information System (INIS)

    Jurina, V.; Viktory, D.; Petrik, T.; Sovcik, J.; Suess, J.; Tomek, J.; Lukacovic, J.; Ivan, J.; Ziakova, M.; Metke, E.; Pospisil, M.; Turner, M.; Homola, J.; Vaclav, J.; Bystricka, S.; Barbaric, M.; Horvath, J.; Betak, J.; Mihaly, B.; Adamovsky, V.; Baloghova, A.; Orihel, M.; Vasina, D.; Balaz, J.; Misovicova, D.; Vrtoch, M.; Mlcuch, J.; Granak, P.; Meleg, J.; Bardy, M.; Gogoliak, J.

    2011-08-01

    The National safety report of the Slovak Republic on the safety of spent fuel management and on the safety of radwaste management in 2011 is presented. These activities in the safety of spent fuel management and radioactive waste management in the Slovak Republic are reported under the headings: (A) Introduction; B) Concept for spent nuclear fuel management (SNF) and radwaste management (RAW); (C) Scope of application of the convention; (D) Spent fuel management and radioactive waste (RAW) management facilities; (E) Legislation and regulation; (F) General safety provisions; (G) Safety of spent fuel management; (H) Safety of radioactive waste (RAW) management; (I) Transboundary movement of spent nuclear fuel and radioactive waste; (J) Disused sealed sources; (K) Planned measures to improve safety; (L) Communication with the public; (M) Annexes. Annexes consists of following parts: I. List of nuclear facilities for spent fuel and RAW management. II. Limits of radioactive material discharges into atmosphere and hydrosphere. III. List of nuclear installations in decommissioning. IV. Inventory of stored spent nuclear fuel. V. Inventory of stored RAW. VI. List of national laws, decrees and guidelines. VII. List of international expert reports (including safety reports). VIII. List of authors.

  20. Towards an International Culture: Gen Y Students and SNS?

    Science.gov (United States)

    Lichy, Jessica

    2012-01-01

    This article reports the findings of a small-scale investigation into the Internet user behaviour of generation Y (Gen Y) students, with particular reference to social networking sites. The study adds to the literature on cross-cultural Internet user behaviour with specific reference to Gen Y and social networking. It compares how a cohort of…

  1. Ética y genética

    OpenAIRE

    2002-01-01

    Este documento contiene una transcripción de las ponencias presentadas en la Mesa Redonda sobre Ética y Genética celebrada en Oviedo, el día 10 de octubre de 2001, y organizada por el Real Patronato sobre Discapacidad en el marco del XXI Congreso Nacional de Genética Humana.

  2. Biometris GenStat Procedure Library Manual 12th Edition

    NARCIS (Netherlands)

    Goedhart, P.W.; Thissen, J.T.N.M.

    2009-01-01

    The Biometris GenStat Procedure Library contains procedures in which new and existing statistical methodology is implemented, as well as procedures to make GenStat more user-friendly. The Biometris library can be used and distributed freely. The Library is standardly distributed over the Dutch

  3. Engenharia Genética: busca da mercadoria perfeita?

    OpenAIRE

    Souza, José Neivaldo de; FEPAR

    2014-01-01

    Resenha:Engenharia Genética: busca da mercadoria perfeita? -SANDEL, Michael J. Contra a Perfeição: ética na era da engenharia genética. 1ª.ed. Tradução por Ana Carolina Mesquita. Rio de janeiro: Civilização Brasileira, 2013. 160p. 

  4. Engenharia Genética: busca da mercadoria perfeita?

    Directory of Open Access Journals (Sweden)

    José Neivaldo de Souza

    2014-06-01

    Full Text Available Resenha:Engenharia Genética: busca da mercadoria perfeita? -SANDEL, Michael J. Contra a Perfeição: ética na era da engenharia genética. 1ª.ed. Tradução por Ana Carolina Mesquita. Rio de janeiro: Civilização Brasileira, 2013. 160p. 

  5. Centralized disassembly and packaging of spent fuel in the DOE spent fuel management system

    International Nuclear Information System (INIS)

    Johnson, E.R.

    1986-01-01

    In October 1984, E.R. Johnson Associates, Inc. (JAI) initiated a study of the prospective use of a centralized facility for the disassembly and packaging of spent fuel to support the various elements of the US Dept. of Energy (DOE) spent fuel management system, including facilities for monitored retrievable storage (MRS) and repositories. It was DOE's original plan to receive spent fuel at each repository where it would be disassembled and packaged (overpacked) for disposal purposes. Subsequently, DOE considered the prospective use of MRS of spent fuel as an option for providing safe and reliable management of spent fuel. This study was designed to consider possible advantages of the use of centralized facilities for disassembly and packaging of spent fuel at whose location storage facilities could be added as required. The study was divided into three principal technical tasks that covered: (a) development of requirements and criteria for the central disassembly and packaging facility and associated systems. (2) Development of conceptual designs for the central disassembly and packaging facility and associated systems. (3) Estimation of capital and operating costs involved for all system facilities and determination of life cycle costs for various scenarios of operation - for comparison with the reference system

  6. Variabilidad genética de Plasmodium falciparum en pacientes con malaria grave y malaria no complicada en Iquitos - Perú

    Directory of Open Access Journals (Sweden)

    Gisely Hijar G

    2002-07-01

    Full Text Available Objetivo: Determinar la diversidad genética del gen que codifica la proteína rica en glutamato (GLURP de Plasmodium falciparum en pacientes con malaria complicada y no complicada circulante en un área del departamento de Loreto, distrito de Maynas. Materiales y métodos: La diversidad genética fue analizada usando reacción en cadena de la polimerasa (PCR en 30 muestras sanguíneas de pacientes con malaria no complicada (MNC y 46 con malaria grave complicada (MGC. Resultados: Ocho genotipos fueron detectados en pacientes con MNC (Genotipo I,II,III, IV,V, VI,VII y VIII y cuatro genotipos en los pacientes con MGC (Genotipo V,VI,VII,VIII. Asimismo, en 50% de las muestras con MNC fueron detectadas infecciones múltiples, a diferencia de las muestras de MGC en donde no se detectó infecciones múltiples. Conclusión: Existe una diversidad genética en esta región del gen GLURP de P. falciparum, para esa época (marzo 1998 - abril 1999 y esa área del país. En tal sentido, nuestros resultados podrían servir de base para llevar a cabo estudios epidemiológicos posteriores, ya que permitiría conocer la distribución de las cepas circulantes en nuestro país.

  7. The Modern RPG IV Language

    CERN Document Server

    Cozzi, Robert

    2006-01-01

    This updated, classic work on the RPG language covers all the new functions and features that have been added since 2003, including new op codes and built-in functions, new chapters on free-format RPG IV and Web programming interfaces, information on implementing XML within RPG IV, and expanded information on procedures. This reference guide takes both novice and experienced RPG IV programmers through the language, from its foundation to its most advanced techniques. More than 100 charts and tables, as well as 350 real-life code samples of functions and operations are included, showing readers

  8. Historical overview of domestic spent fuel shipments

    International Nuclear Information System (INIS)

    Pope, R.B.; Wankerl, M.W.; Armstrong, S.; Hamberger, C.; Schmid, S.

    1991-01-01

    The purpose of this paper is to provide available historical data on most commercial and research reactor spent fuel shipments that have been completed in the United States between 1964 and 1989. This information includes data on the sources of spent fuel that has been shipped, the types of shipping casks used, the number of fuel assemblies that have been shipped, and the number of shipments that have been made. The data are updated periodically to keep abreast of changes. Information on shipments is provided for planning purposes; to support program decisions of the US Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM); and to inform interested members of the public, federal, state, and local government, Indian tribes, and the transportation community. 5 refs., 7 figs., 2 tabs

  9. Corrosion surveillance in spent fuel storage pools

    International Nuclear Information System (INIS)

    Howell, J.P.

    1996-01-01

    In mid-1991, corrosion of aluminum-clad spent nuclear fuel was observed in the light-water filled basins at the Savannah River site. A corrosion surveillance program was initiated in the P, K, L-Reactor basins and in the Receiving Basin for Offsite Fuels (RBOF). This program verified the aggressive nature of the pitting corrosion and provided recommendations for changes in basin operations to permit extended longer term interim storage. The changes were implemented during 1994--1996 and have resulted in significantly improved basin water quality with conductivity in the 1--3 microS/cm range. Under these improved conditions, no new pitting has been observed over the last three years. This paper describes the corrosion surveillance program at SRS and what has been learned about the corrosion of aluminum-clad in spent fuel storage pools

  10. Actinide chemistry and spent fuel disposal

    International Nuclear Information System (INIS)

    Werme, L.O.

    1990-01-01

    Spent fuel as a high level waste form has been studied for over a decade. Although the fuel contains a multitude of radionuclides, only a few of them will actually constitute a long term potential radiotoxic hazard. These elements are, after some thousand years, plutonium and neptunium. Up to about 100,000 years plutonium is by far dominating the radiotoxicity after which neptunium becomes the dominating element. In this paper, the results obtained on plutonium and neptunium releases from spent fuel are reviewed and discussed. Attempts are made to interpret the data obtained for different fuel types and groundwaters in terms of the chemistry of the respective elements as well as in terms of release kinetics

  11. Impact analysis of spent fuel jacket assemblies

    International Nuclear Information System (INIS)

    Aramayo, G.A.

    1994-01-01

    As part of the analyses performed in support of the reracking of the High Flux Isotope Reactor pool, it became necessary to prove the structural integrity of the spent fuel jacket assemblies subjected to gravity drop that result from postulated accidents associated with the handling of these assemblies while submerged in the pool. The spent fuel jacket assemblies are an integral part of the reracking project, and serve to house fuel assemblies. The structure integrity of the jacket assemblies from loads that result from impact from a height of 10 feet onto specified targets has been performed analytically using the computer program LS-DYNA3D. Nine attitudes of the assembly at the time of impact have been considered. Results of the analyses show that there is no failure of the assemblies as a result of the impact scenarios considered

  12. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Ro, Seung Gy; Shin, Y. J.; Do, J. B.; You, G. S.; Seo, J. S.; Lee, H. G.

    1998-03-01

    This study is to develop an advanced spent fuel management process for countries which have not yet decided a back-end nuclear fuel cycle policy. The aims of this process development based on the pyroreduction technology of PWR spent fuels with molten lithium, are to reduce the storage volume by a quarter and to reduce the storage cooling load in half by the preferential removal of highly radioactive decay-heat elements such as Cs-137 and Sr-90 only. From the experimental results which confirm the feasibility of metallization technology, it is concluded that there are no problems in aspects of reaction kinetics and equilibrium. However, the operating performance test of each equipment on an engineering scale still remain and will be conducted in 1999. (author). 21 refs., 45 tabs., 119 figs

  13. Design of spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide is for interim spent fuel storage facilities that are not integral part of an operating nuclear power plant. Following the introduction, Section 2 describes the general safety requirements applicable to the design of both wet and dry spent fuel storage facilities; Section 3 deals with the design requirements specific to either wet or dry storage. Recommendations for the auxiliary systems of any storage facility are contained in Section 4; these are necessary to ensure the safety of the system and its safe operation. Section 5 provides recommendations for establishing the quality assurance system for a storage facility. Section 6 discusses the requirements for inspection and maintenance that must be considered during the design. Finally, Section 7 provides guidance on design features to be considered to facilitate eventual decommissioning. 18 refs

  14. Robotic cleaning of a spent fuel pool

    International Nuclear Information System (INIS)

    Roman, H.T.; Marian, F.A.; Silverman, E.B.; Barkley, V.P.

    1987-01-01

    Spent fuel pools at nuclear power plants are not cleaned routinely, other than by purifying the water that they contain. Yet, debris can collect on the bottom of a pool and should be removed prior to fuel transfer. At Public Service Electric and Gas Company's Hope Creek Nuclear Power Plant, a submersible mobile robot - ARD Corporation's SCAVENGER - was used to clean the bottom of the spent fuel pool prior to initial fuel loading. The robotic device was operated remotely (as opposed to autonomously) with a simple forward/reverse control, and it cleaned 70-80% of the pool bottom. This paper reports that a simple cost-benefit analysis shows that the robotic device would be less expensive, on a per mission basis, than other cleaning alternatives, especially if it were used for other similar cleaning operations throughout the plant

  15. Cost analysis methodology of spent fuel storage

    International Nuclear Information System (INIS)

    1994-01-01

    The report deals with the cost analysis of interim spent fuel storage; however, it is not intended either to give a detailed cost analysis or to compare the costs of the different options. This report provides a methodology for calculating the costs of different options for interim storage of the spent fuel produced in the reactor cores. Different technical features and storage options (dry and wet, away from reactor and at reactor) are considered and the factors affecting all options defined. The major cost categories are analysed. Then the net present value of each option is calculated and the levelized cost determined. Finally, a sensitivity analysis is conducted taking into account the uncertainty in the different cost estimates. Examples of current storage practices in some countries are included in the Appendices, with description of the most relevant technical and economic aspects. 16 figs, 14 tabs

  16. Spent nuclear fuel project product specification

    International Nuclear Information System (INIS)

    Pajunen, A.L.

    1998-01-01

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification

  17. Factors Influencing Retention of Gen Y and Non-Gen Y Teachers Working at International Schools in Asia

    Science.gov (United States)

    Fong, Hoi Wah Benny

    2018-01-01

    Quantitative studies on international-school teacher retention are few, especially studies that differentiate between Gen Y and non-Gen Y teachers. This article reports on the findings of a study that examined the relationship of job satisfaction factors to the likelihood of contract renewal by international-school teachers. Results from the study…

  18. Status and prospects for spent fuel management in France

    International Nuclear Information System (INIS)

    Portal, R.; L'Epine, P. de

    1996-01-01

    The spent fuel arisings and storage capacities, the interface between fuel storage and transportation activities, the spent fuel storage technology, the reprocessing and recycling industrial activities in France are described in the paper. (author). 6 figs, 8 tabs

  19. Spent nuclear fuel project integrated schedule plan

    International Nuclear Information System (INIS)

    Squires, K.G.

    1995-01-01

    The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel

  20. Natural convection cooling of spent fuels depository

    International Nuclear Information System (INIS)

    Menant, B.

    2000-01-01

    The operating CASCAD Facility was commissioned at Cadarache since 1990. Spent fuels are being storage for a 50 years period. The heat giving by the wastes is evacuated essentially by natural convection. The Trio U software is applied to the thermohydraulic operating of the system. The results allow to illustrate the installation and show system instabilities effects which appear at many scales. (A.L.B.)

  1. Towards a Swedish repository for spent fuel

    International Nuclear Information System (INIS)

    Ahlstroem, P.-E.

    1997-01-01

    Nuclear power is producing electricity for the benefit of society but is also leaving radioactive residues behind. It is our responsibility to handle these residues in a safe and proper manner. The development of a system for handling spent fuel from nuclear power plants has proceeded in steps. The same is true for the actual construction of facilities and will continue to be the case for the final repository for spent fuel and other types of long-lived wastes. The primary objective in constructing the repository will be to isolate and contain the radioactive waste. In case the isolation fails for some reason the multibarrier system should retain and retard the radionuclides that might come into contact with the groundwater. A repository is now planned to be built in two steps where the first step will include deposition of about 400 canisters with spent fuel. This first step should be finished in about 20 years from now and be followed by an extensive evaluation of the results from not only this particular step but also from the development of alternative routes before deciding on how to proceed. A special facility to encapsulate the spent fuel is also required. Such an encapsulation plant is proposed to be constructed as an extension of the existing interim storage CLAB. Finding a site for the repository is a critical issue in the implementation of any repository. The siting process started a few years ago and made some progress but is by no means yet completed. It will go on at least into the early part of the next decade. When the present nuclear power plants begin to be due for retirement there should also be some facilities in place to take permanent care of the long-lived radioactive residues. Progress in siting will be a prerequisite for success in our responsibility to make progress towards a safe permanent solution of the waste issue. (orig.)

  2. Concrete encapsulation for spent nuclear fuel storage

    International Nuclear Information System (INIS)

    Fleischer, L.R.; Gunasekaran, M.

    1981-01-01

    Concrete systems, mixtures and methods for encapsulating and storing spent nuclear fuel. Fuel discharged from nuclear reactors in the form of rods or multi-rod assemblies is completely and contiguously enclosed in concrete having incorporated therein metallic fibers to increase thermal conductivity and polymers to decrease fluid permeability. The metallic fibers and the polymers can be distributed in a single concrete layer, or separate contiguous layers can be utilized for the conductivity and impermeability characteristics

  3. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Practice has been prepared as part of the IAEA's programme on the safety assessment of interim spent fuel storage facilities which are not an integral part of an operating nuclear power plant. This report provides general guidance on the safety assessment process, discussing both deterministic and probabilistic assessment methods. It describes the safety assessment process for normal operation and anticipated operational occurrences and also related to accident conditions. 10 refs, 2 tabs

  4. Biodiesel Production from Spent Coffee Grounds

    OpenAIRE

    Blinová Lenka; Bartošová Alica; Sirotiak Maroš

    2017-01-01

    The residue after brewing the spent coffee grounds is an oil-containing waste material having a potential of being used as biodiesel feedstock. Biodiesel production from the waste coffee grounds oil involves collection and transportation of coffee residue, drying, oil extraction, and finally production of biodiesel. Different methods of oil extraction with organic solvents under different conditions show significant differences in the extraction yields. In the manufacturing of biodiesel from ...

  5. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Park, B. S.; Yoon, J. S.; Hong, H. D. (and others)

    2007-02-15

    In this research, the remote handling technology was developed for the ACP application. The ACP gives a possible solution to reduce the rapidly cumulative amount of spent fuels generated from the nuclear power plants in Korea. The remote technologies developed in this work are a slitting device, a voloxidizer, a modified telescopic servo manipulator and a digital mock-up. A slitting device was developed to declad the spent fuel rod-cuts and collect the spent fuel UO{sub 2} pellets. A voloxidizer was developed to convert the spent fuel UO{sub 2} pellets obtained from the slitting process in to U{sub 3}O{sub 8} powder. Experiments were performed to test the capabilities and remote operation of the developed slitting device and voloxidizer by using simulated rod-cuts and fuel in the ACP hot cell. A telescopic servo manipulator was redesigned and manufactured improving the structure of the prototype. This servo manipulator was installed in the ACP hot cell, and the target module for maintenance of the process equipment was selected. The optimal procedures for remote operation were made through the maintenance tests by using the servo manipulator. The ACP digital mockup in a virtual environment was established to secure a reliability and safety of remote operation and maintenance. The simulation for the remote operation and maintenance was implemented and the operability was analyzed. A digital mockup about the preliminary conceptual design of an enginnering-scale ACP was established, and an analysis about a scale of facility and remote handling was accomplished. The real-time diagnostic technique was developed to detect the possible fault accidents of the slitting device. An assessment of radiation effect for various sensors was also conducted in the radiation environment.

  6. Spent fuel packaging and its safety analysis

    International Nuclear Information System (INIS)

    Takada, Kimitaka; Nakaoki, Kozo; Tamamura, Tadao; Matsuda, Fumio; Fukudome, Kazuyuki

    1983-01-01

    An all stainless steel B(U) type packaging is proposed to transport spent fuels discharged from research reactors and other radioactive materials. The package is used dry and provided with surface fins to absorb drop shock and to dissipate decay heat. Safety was analyzed for structural, thermal, containment shielding and criticality factors, and the integrity of the package was confirmed with the MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, and KENO computer codes. (author)

  7. Management and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    1987-05-01

    The programme consists of the long-term and short-term programme, the continued bedrock investigations, the underground research laboratory, the decision-making procedure in the site selection process and information questions during the site selection process. The National Board for Spent Nuclear Fuel hereby subunits both the SKB's R and D Programme 86 and the Board's statement concerning the programme. Decisions in the matter have been made by the Board's executive committee. (DG)

  8. Re-fermentation of washed spent solids from batch hydrogenogenic fermentation for additional production of biohydrogen from the organic fraction of municipal solid waste.

    Science.gov (United States)

    Muñoz-Páez, Karla M; Ríos-Leal, Elvira; Valdez-Vazquez, Idania; Rinderknecht-Seijas, Noemí; Poggi-Varaldo, Héctor M

    2012-03-01

    In the first batch solid substrate anaerobic hydrogenogenic fermentation with intermittent venting (SSAHF-IV) of the organic fraction of municipal solid waste (OFMSW), a cumulative production of 16.6 mmol H(2)/reactor was obtained. Releases of hydrogen partial pressure first by intermittent venting and afterward by flushing headspace of reactors with inert gas N(2) allowed for further hydrogen production in a second to fourth incubation cycle, with no new inoculum nor substrate nor inhibitor added. After the fourth cycle, no more H(2) could be harvested. Interestingly, accumulated hydrogen in 4 cycles was 100% higher than that produced in the first cycle alone. At the end of incubation, partial pressure of H(2) was near zero whereas high concentrations of organic acids and solvents remained in the spent solids. So, since approximate mass balances indicated that there was still a moderate amount of biodegradable matter in the spent solids we hypothesized that the organic metabolites imposed some kind of inhibition on further fermentation of digestates. Spent solids were washed to eliminate organic metabolites and they were used in a second SSAHF-IV. Two more cycles of H(2) production were obtained, with a cumulative production of ca. 2.4 mmol H(2)/mini-reactor. As a conclusion, washing of spent solids of a previous SSAHF-IV allowed for an increase of hydrogen production by 15% in a second run of SSAHF-IV, leading to the validation of our hypothesis. Copyright © 2011 Elsevier Ltd. All rights reserved.

  9. Conditioning technology of spent radium sources

    International Nuclear Information System (INIS)

    Kang, Il Sik; Kim, K. J.; Jang, K. D.

    2001-03-01

    In order to avoid accidents that could be resulted from improper storage of spent radium sources, it is necessary to condition and store them safely. The program for safe conditioning of spent radium sources by IAEA has been established to assist the developing countries. The main object of this report is to understand well and apply the technology that was applied in conditioning the national inventory of Ra-226 sources in Myanmar, as a part of IAEA's project by the Korean expert team. The report is the result that the Korean expert team carried out in Myanmar under the project title 'Radium Conditioning Service in Myanmar(INT4131-06646C)'. As a result of the mission, a whole inventory, 1,429.5 mCi of spent radium sources was safely conditioned by the Korean expert team according to the manual under the supervision of IAEA's technical officer, Mr. Al-Mughrabi, and under the control of DAE authority. These sources were encapsuled in 27 small capsules and 3 large capsules, and conditioned in 3 lead shields, producing 3 packages. The inventories were distributed into 3 shielding devices, holding 500, 459.5, and 470 mCi

  10. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Shin, Young Joon; Cho, S. H.; You, G. S.

    2001-04-01

    Currently, the economic advantage of any known approach to the back end fuel cycle of a nuclear power reactor has not been well established. Thus the long term storage of the spent fuel in a safe manner is one of the important issues to be resolved in countries where the nuclear power has a relatively heavy weight in power production of that country. At KAERI, as a solution to this particular issue midterm storage of the spent fuel, an alternative approach has been developed. This approach includes the decladding and pulverization process of the spent PWR fuel rod, the reducing process from the uranium oxide to a metallic uranium powder using Li metal in a LiCl salt, the continuous casting process of the reduced metal, and the recovery process of Li from mixed salts by the electrolysis. We conducted the laboratory scale tests of each processes for the technical feasibility and determination for the operational conditions for this approach. Also, we performed the theoretical safety analysis and conducted integral tests for the equipment integration through the Mock-up facility with non-radioactive samples. There were no major issues in the approach, however, material incompatibility of the alkaline metal and oxide in a salt at a high temperature and the reactor that contains the salt became a show stopper of the process. Also the difficulty of the clear separation of the salt with metals reduced from the oxide became a major issue

  11. Considerations for the transportation of spent fuel

    International Nuclear Information System (INIS)

    Jefferson, R.M.

    1984-01-01

    In our society today the transportation of radioactive materials, and most particularly spent reactor fuel, is surrounded by considerable emotion and a wealth of information, good and bad. The transportation of these materials is viewed as unique and distinct from the transportation of other hazardous materials and as a particularly vulnerable component of the nuclear power activities of this nation. Added to this is the concept, widely held, that almost everyone is an expert on the transportation of radioactive materials. One significant contribution to this level of emotion is the notion that all roads (rail and highway), on which these goods will be transported, somehow traverse everyone's backyard. The issue of the transportation of spent fuel has thus become a political battleground. Perhaps this should not be surprising since it has all of the right characteristics for such politicization in that it is pervasive, emotional, and visible. In order that those involved in the discussion of this activity might be able to reach some rational conclusions, this paper offers some background information which might be useful to a broad range of individuals in developing their own perspectives. The intent is to address the safety of transporting spent fuel from a technical standpoint without the emotional content which is frequently a part of this argument

  12. Past and future IAEA spent fuel management activities

    International Nuclear Information System (INIS)

    Grigoriev, A.

    1993-01-01

    The main objectives and strategies of the Agency's activities in the area of spent fuel management are to promote the exchange of information between Member States on technical, safety, environmental and economic aspects of spent fuel management technology, including storage, transport and treatment of spent fuel, and to provide assistance to Member States in the planning, implementation and operation of nuclear fuel cycle facilities. This paper give a list of the meetings held since the last issue of the Spent Fuel Management Newsletter

  13. NextGen Future Safety Assessment Game

    Science.gov (United States)

    Ancel, Ersin; Gheorghe, Adrian; Jones, Sharon Monica

    2011-01-01

    The successful implementation of the next generation infrastructure systems requires solid understanding of their technical, social, political and economic aspects along with their interactions. The lack of historical data that relate to the long-term planning of complex systems introduces unique challenges for decision makers and involved stakeholders which in turn result in unsustainable systems. Also, the need to understand the infrastructure at the societal level and capture the interaction between multiple stakeholders becomes important. This paper proposes a methodology in order to develop a holistic approach aiming to provide an alternative subject-matter expert (SME) elicitation and data collection method for future sociotechnical systems. The methodology is adapted to Next Generation Air Transportation System (NextGen) decision making environment in order to demonstrate the benefits of this holistic approach.

  14. 5 CFR 551.425 - Time spent receiving medical attention.

    Science.gov (United States)

    2010-01-01

    ... 5 Administrative Personnel 1 2010-01-01 2010-01-01 false Time spent receiving medical attention... Relation to Other Activities § 551.425 Time spent receiving medical attention. (a) Time spent waiting for and receiving medical attention for illness or injury shall be considered hours of work if: (1) The...

  15. 5 CFR 551.422 - Time spent traveling.

    Science.gov (United States)

    2010-01-01

    ... 5 Administrative Personnel 1 2010-01-01 2010-01-01 false Time spent traveling. 551.422 Section 551... Activities § 551.422 Time spent traveling. (a) Time spent traveling shall be considered hours of work if: (1... who is permitted to use an alternative mode of transportation, or an employee who travels at a time...

  16. Effects of Spent Engine Oil Polluted Soil and Organic Amendment ...

    African Journals Online (AJOL)

    A trial was conducted at the Screen House of the Department of Crop Science, University of Benin, Benin City, Nigeria to evaluate the efficacy of using organic fertilizer as bioremediant for spent engine oil polluted soils. Three concentrations of spent engine oil (0, 5 and 10 % w/w, spent engine oil in soil) and three ...

  17. Evaluating spent CCA residential decks for second-life products

    Science.gov (United States)

    Robert Smith; David Bailey; Phil Araman

    2003-01-01

    The amount of CCA treated wood being removed from spent residential decks is increasing at a tremendous rate. While most spent CCA treated wood is being disposed in landfills, further useful and environmentally beneficial alternatives have to be met. This research estimated the percentage of recoverable lumber from spent CCA decks that can be recycled into other usable...

  18. Final environmental impact statement: US Spent Fuel Policy. Charge for spent fuel storage

    International Nuclear Information System (INIS)

    1980-05-01

    The United States Government policy relating to nuclear fuel reprocessing, which was announced by President Carter on April 7, 1977, provides for an indefinite deferral of reprocessing, and thus commits light water reactor (LWR) plants to a once-through fuel cycle during that indefinite period. In a subsequent action implementing that policy, the Department of Energy (DOE) on October 18, 1977 announced a spent fuel policy which would enable domestic, and on a selective basis, foreign utilities to deliver spent fuel to the US Government for interim storage and final geologic disposal, and pay the Government a fee for such services. This volume addresses itself to whether the fee charged for these services, by its level or its structure, would have any effect on the environmental impacts of implementing the Spent Fuel Policy itself. This volume thus analyzes the fee and various alternatives to determine the interaction between the fee and the degree of participation by domestic utilities and foreign countries in the proposed spent fuel program for implementing the Spent Fuel Policy. It also analyzes the effect, if any, of the fee on the growth of nuclear power

  19. Metode Transfer Asam Nukleat sebagai Dasar Terapi Gen

    Directory of Open Access Journals (Sweden)

    Novi Silvia Hardiany

    2017-01-01

    Full Text Available Kemajuan ilmu biologi molekuler memberikan manfaat dalam bidang kedokteran untuk mengembangkanterapi gen. Tujuan terapi gen adalah untuk memperbaiki kerusakan gen atau mengganti gen yang rusakdengan gen yang normal. Pemindahan gen dilakukan dengan teknik transfeksi. Transfeksi merupakanproses pemindahan asam nukleat baik menggunakan vektor virus (transduksi atau menggunakan metodenonviral yaitu zat kimia, lipid dan metode fisik. Vektor virus yang digunakan pada transduksi adalahretrovirus, adenovirus, adeno-associated virus (AAV dan herpes simplex virus (HSV. Keberhasilantransfeksi ditentukan oleh berbagai faktor yang dapat dapat dinilai dengan menggunakan reporter sepertigreen fluorescence protein (GFP. Kata Kunci: terapi gen, transfeksi non viral, transduksi, vektor virus   Methods of Nucleic Acid Transfer as Basic Gene Therapy Abstract The advancement of molecular biology provides benefit in the field of medicine to develop genetherapy. The aim of gene therapy is to repair the genetic damage or to replace damaged gene with thenormal gene. Delivery of gene is carried out by transfection technique, a technique to transfer nucleic acidinto eukaryote cells either using viral vectors (known as transduction, and also using non viral methodsuch as chemical substance, lipid and physical method. Some of the viral vectors used in the transductionare retrovirus, adenovirus, Adeno-associated virus (AAV and Herpes Simplex Virus (HSV. The success oftransfection is determined by various factors which can be assessed using several reporters such as GreenFluorescence Protein (GFP. Key words: gene therapy, non viral transfection, transduction, viral vector. Normal 0 false false false IN X-NONE X-NONE

  20. Modelo poblacional con algoritmos genéticos

    OpenAIRE

    Veliz Quintero, Eduardo; Rodriguez Ojeda, Luis

    2009-01-01

    Para el desarrollo de este trabajo, “MODELO POBLACIONAL CON ALGORITMOS GENÉTICOS”, he investigado la rama de la inteligencia artificial, como son los algoritmos genéticos. Primero presento en forma general los aspectos que envuelven los algoritmos genéticos, parto de la necesidad de optimizar, así como su historia y posibles aplicaciones y luego he cubierto detalladamente todo lo que pude investigar sobre la teoría de los algoritmos genéticos, sus fundamentos matemáticos, tipos de algoritmos ...

  1. Best-practices guidelines for L2PSA development and applications. Volume 2 - Best practices for the Gen II PWR, Gen II BWR L2PSAs. Extension to Gen III reactors

    International Nuclear Information System (INIS)

    Raimond, E.; Durin, T.; Rahni, N.; Meignen, R.; Cranga, M.; Pichereau, F.; Bentaib, A.; Guigueno, Y.; Loeffler, H.; Mildenberger, O.; Lajtha, G.; Santamaria, C.S.; Dienstbier, J.; Rydl, A.; Holmberg, J.E.; Lindholm, I.; Maennistoe, I.; Pauli, E.M.; Dirksen, G.; Grindon, L.; Peers, K.; Hulqvist, G.; Parozzi, F.; Polidoro, F.; Cazzoli, E.; Vitazkova, J.; Burgazzi, L.; Oury, L.; Ngatchou, C.; Siltanen, S.; Niemela, I.; Routamo, T.; Helstroem, P.; Bassi, C.; Brinkman, H.; Seidel, A.; Schubert, B.; Wohlstein, R.; Guentay, S.; Vincon, L.

    2010-01-01

    The objective of this coordinated action was to develop best practice guidelines for the performance of Level 2 PSA methodologies with a view of harmonisation at EU level and to allow meaningful and practical uncertainty evaluations in a Level 2 PSA. Specific relationships with community in charge of nuclear reactor safety (utilities, safety authorities, vendors, and research or services companies) have been established in order to define the current needs in terms of guidelines for level 2 PSA development and applications. An international workshop was organised in Hamburg, with the support of VATTENFALL, in November 2008. The level 2 PSA experts from the ASAMPSA2 project partners have proposed some guidelines for the development and application of L2PSA based on their experience and on information available from international cooperation (EC Severe Accident network of Excellence - SARNET, IAEA standards, OECD-NEA publications and workshop) or open literature. The number of technical issues addressed in the guideline is very large and all are not covered with the same relevancy in the first version of the guideline. This version is submitted for external review in November 2010 by severe accident experts and PSA, especially, from SARNET and OECD-NEA members. The feedback of the external review will be dis cussed during an international open works hop planned in March 2011 and all outcomes will be taken into consideration in the final version of this guideline (June 2011). The guideline includes 3 volumes: - Volume 1 - General considerations on L2PSA. - Volume 2 - Technical recommendations for Gen II and III reactors. - Volume 3 - Specific considerations for future reactor (Gen IV). The recommendations formulated in the guideline should not be considered as 'mandatory' but should help the L2PSA developers to achieve high quality studies with limited time and resources. It may also help the L2PSA reviewers by positioning one specific study in comparison

  2. Hafnium(IV) complexation with oxalate at variable temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Friend, Mitchell T.; Wall, Nathalie A. [Washington State Univ., Pullmanm, WA (United States). Dept. of Chemistry

    2017-08-01

    Appropriate management of fission products in the reprocessing of spent nuclear fuel (SNF) is crucial in developing advanced reprocessing schemes. The addition of aqueous phase complexing agents can prevent the co-extraction of these fission products. A solvent extraction technique was used to study the complexation of Hf(IV) - an analog to fission product Zr(IV) - with oxalate at 15, 25, and 35 C in 1 M HClO{sub 4} utilizing a {sup 175+181}Hf radiotracer. The mechanism of the solvent extraction system of 10{sup -5} M Hf(IV) in 1 M HClO{sub 4} to thenoyltrifluoroacetone (TTA) in toluene demonstrated a 4{sup th}-power dependence in both TTA and H{sup +}, with Hf(TTA){sub 4} the only extractable species. The equilibrium constant for the extraction of Hf(TTA){sub 4} was determined to be log K{sub ex}=7.67±0.07 (25±1 C, 1 M HClO{sub 4}). The addition of oxalate to the aqueous phase decreased the distribution ratio, indicating aqueous Hf(IV)-oxalate complex formation. Polynomial fits to the distribution data identified the formation of Hf(ox){sup 2+} and Hf(ox){sub 2(aq)} and their stability constants were measured at 15, 25, and 35 C in 1 M HClO{sub 4}. van't Hoff analysis was used to calculate Δ{sub r}G, Δ{sub r}H, and Δ{sub r}S for these species. Stability constants were observed to increase at higher temperature, an indication that Hf(IV)-oxalate complexation is endothermic and driven by entropy.

  3. Plutonium(IV) and thorium(IV) hydrous polymer chemistry

    International Nuclear Information System (INIS)

    Johnson, G.L.; Toth, L.M.

    1978-05-01

    The recent attention given to Pu(IV) polymers has warranted a review of plutonium and thorium hydrolysis chemistry with respect to the various experimental approaches and insights gained therein. Differing terminologies used in the experimental procedures have often confused the understanding of the chemical processes which occur between the first hydrolysis reaction of the tetravalent actinide and its final dehydration to form the crystalline oxide. This report focuses on the polymer aging reaction which is defined here in terms of A. W. Thomas' ol to oxo conversion reaction and involves simply the conversion of hydroxyl-bridged polymer links to oxygen-bridged linkages. Thorium(IV) hydrolytic reactions are included because they are analogous in many respects to those of Pu(IV) and offer a simpler chemical system for experimental study. Future work using spectroscopic techniques should significantly improve the description of this aging phenomenon

  4. Current status of spent fuel management in China

    International Nuclear Information System (INIS)

    Liu Yuansong

    2005-01-01

    The administration organization and the laws and regulations on spent fuel management in China are described at first. In the year of 2000, CNN C and GNPJVC signed a contract of take-over mode for spent fuel, in which CNNC will take-over the spent fuel annually since 2003. In the year of 2001, CNNC and QNPC signed another contract of take-over mode for spent fuel. Transportation by road is the realistic selections of the transportation of the spent fuel from the nuclear power stations to LNFC. As for spent fuel management, the construction of CNNC's reprocessing pilot plant is going on and R and D of spent fuel management are launching continuously. (author)

  5. Overview of spent fuel management options: Technology and cost

    International Nuclear Information System (INIS)

    Beeman, G.H.

    1986-01-01

    The options for spent fuel management discussed in this paper. There are several spent fuel management options available to utilities beyond the traditional options of reracking and transshipment. Two different dry storage options have been licensed by NRC and it appears that NRC will issue a license amendment to allow for the storage of consolidated spent fuel in the very near future. The use of either of these spent fuel management options is highly dependent upon the individual requirements of the utility. There are several factors that must be considered when choosing a specific fuel management option. Either of these spent fuel management options, dry storage or rod consolidation, is capable of meeting the needs of utilities for additional spent fuel storage until DOE begins accepting spent fuel for final disposal in 1998

  6. Using "The Simpsons" to Teach Humanities with Gen X and Gen Y Adult Students

    Science.gov (United States)

    Fink, Maxwell A.; Foote, Deborah C.

    2007-01-01

    Many educators lament the UNESCO study showing that by the time the average teen graduates from high school he or she has spent more than fifteen thousand hours watching television and only eleven thousand in the classroom (Gorebel, 1998). Rather than regretting this "condition," colleges, universities, and educators of adults and children should…

  7. Tipos de tricomas em genótipos de Lycopersicon Types of trichomes in Lycopersicon genotypes

    Directory of Open Access Journals (Sweden)

    Luciana C. Toscano

    2001-11-01

    Full Text Available Foram identificados os tipos de tricomas em alguns genótipos de Lycopersicon, utilizando microscópio eletrônico de varredura. Foram utilizados os genótipos L. hirsutum (PI 127826 e PI 127827, L. hirsutum var. glabratum (PI 134417, L. pennellii (LA 716 e L. esculentum (híbrido Bruna VFN e Santa Clara. Os tricomas glandulares encontrados foram tipo IV em LA 716, PI 127826, PI 127827 e PI 134417; tipo VIc em PI 127826, PI 127827, PI 134417 e Santa Clara; tipo VII em PI 127826, PI 127827 e PI 134417; tipo I em PI 127826, PI 127827, PI 134417, híbrido Bruna VFN e Santa Clara; tipo VIa em híbrido Bruna VFN. Os tricomas não glandulares observados foram: tipo Va em PI 127826, PI 127827, PI 134417, híbrido Bruna VFN e Santa Clara; VIII em híbrido Bruna VNF e os tipos III e Vb em Santa Clara.The types of trichomes in some Lycopersicon genotypes were identified using the scanning electron microscope. The tomato genotypes observed were L. hirsutum (PI 127826 and PI 127827, L. hirsutum f. glabratum (PI 134417, L. pennellii (LA 716 and L. esculentum (Bruna VFN hybrid and Santa Clara. The following glandular trichomes were found: type IV in LA 716, PI 127826, PI 127827 and PI 134417; type VIc in PI 127826, PI 127827, PI 134417 and Santa Clara; type VII in PI 127826, PI 127827 and PI 134417; type I in PI 127826, PI 127827, PI 134417 , Bruna VFN hybrid and Santa Clara; type VIa in Bruna VFN hybrid. The following nonglandular trichomes were observed: type Va in PI 127826, PI 127827, PI 134417, Bruna VFN Hybrid and Santa Clara; type VIII in Bruna VNF hybrid and types III and Vb in Santa Clara.

  8. Trunnions for spent fuel element shipping casks

    International Nuclear Information System (INIS)

    Cooke, B.

    1989-01-01

    Trunnions are used on spent fuel element shipping casks for one or more of a combination of lifting, tilting or securing to a transport vehicle. Within the nuclear transportation industry there are many different philosophies on trunnions, concerning the shape, manufacture, attachment, inspection, maintenance and repair. With the volume of international transport of spent fuel now taking place, it is recognized that problems are occurring with casks in international traffic due to the variance of the philosophies, national standards, and the lack of an international standard. It was agreed through the ISO that an international standard was required to harmonize. It was not possible to evolve an international standard. It was only possible to evolve an international guide. To evolve a standard would mean superseding any existing national standards which already cover particular aspects of trunnions i.e. deceleration forces imposed on trunnions used as tie down features. Therefore the document is a guide only and allows existing national standards to take precedence where they exist. The guide covers design, manufacture, maintenance, repair and quality assurance. The guide covers trunnions used on spent fuel casks transported by road, rail and sea. The guide details the considerations which should be taken account of by cask designers, i.e. stress intensity, design features, inspection and test methods etc. Manufacture, attachment and pre-service testing is also covered. The guide details user requirements which should also be taken account of, i.e. servicing frequency, content, maintenance and repair. The application of quality assurance is described separately although the principles are used throughout the guide

  9. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-06-01

    This report summarizes the conceptual framework for a Pacific Basin Spent Fuel Logistics System (PBSFLS); and preliminarily describes programatic steps that might be taken to implement such a system. The PBSFLS concept is described in terms of its technical and institutional components. The preferred PBSFLS concept embodies the rationale of emplacing a fuel cycle system which can adjust to the technical and institutional non-proliferation solutions as they are developed and accepted by nations. The concept is structured on the basis of initially implementing a regional spent fuel storage and transportation system which can technically and institutionally accommodate downstream needs for energy recovery and long-term waste management solutions.

  10. Global Spent Fuel Logistics Systems Study (GSFLS). Volume 4. Pacific basin spent fuel logistics system

    International Nuclear Information System (INIS)

    1978-06-01

    This report summarizes the conceptual framework for a Pacific Basin Spent Fuel Logistics System (PBSFLS); and preliminarily describes programatic steps that might be taken to implement such a system. The PBSFLS concept is described in terms of its technical and institutional components. The preferred PBSFLS concept embodies the rationale of emplacing a fuel cycle system which can adjust to the technical and institutional non-proliferation solutions as they are developed and accepted by nations. The concept is structured on the basis of initially implementing a regional spent fuel storage and transportation system which can technically and institutionally accommodate downstream needs for energy recovery and long-term waste management solutions

  11. Recovery of decommissioning and spent fuel charges

    International Nuclear Information System (INIS)

    Bermanis, G.

    1982-01-01

    The licensing and financial aspects of NPP decommissioning, deactivation and dismantling of radioactive equipment in the USA are considered. Data on the costs of spent fuel transport and conservation are given. The state of the problem development in other countries is briefly described. It is pointed out that the technical aspect of the problem is much better studied than that of license-financial problem. At the same time in contrast to TPP NPP use is connected with considerable expenses after the end of a power plant sevice time

  12. Radionuclide characterization of spent purification filters

    International Nuclear Information System (INIS)

    Sinclair, A.N.; Spooner, P.

    1984-01-01

    A remote handling facility is described which will aid in the assessment of the radionuclide mix and total activity content of spent purification filters. The facility will also dry filters as part of their preparation for off-site shipment. One possible system design, consisting of a robot with five degrees of freedom, a shielded Ge-Li detector, and a vacuum drying system, is described. Benefits could include reduced operating and waste storage or burial costs, feedback on filter performance, and an improved understanding of environmental protection requirements

  13. On the Spent Coffee Grounds Biogas Production

    Directory of Open Access Journals (Sweden)

    Tomáš Vítěz

    2016-01-01

    Full Text Available Due to the strict legislation currently in use for landfilling, anaerobic digestion has a strong potential as an alternative treatment for biodegradable waste. Coffee is one of the most consumed beverages in the world and spent coffee grounds (SCG are generated in a considerable amount as a processing waste during making the coffee beverage. Chemical composition of SCG, presence of polysaccharides, proteins, and minerals makes from the SCG substrates with high biotechnological value, which might be used as valuable input material in fermentation process. The methane production ranged from 0.271–0.325 m3/kg dry organic matter.

  14. Concretes characterization for spent radioactive sources

    International Nuclear Information System (INIS)

    Martinez B, J.; Monroy G, F. P.

    2013-10-01

    The present work includes the preparation and characterization of the concrete used as conditioning matrix of spent radioactive sources in the Treatment Plant of Radioactive Wastes of the Instituto Nacional de Investigaciones Nucleares (ININ). The concrete tests tubes were subjected to resistance assays to the compression, leaching, resistance to the radiation and porosity, and later on characterized by means of X rays diffraction, scanning electron microscopy and infrared spectrometry, with the purpose of evaluating if this concrete accredits the established tests by the NOM-019-Nucl-1995. The results show that the concrete use in the Treatment Plant fulfills the requirements established by the NOM-019-Nucl-1995. (author)

  15. Subsurface storage of commercial spent nuclear fuel

    International Nuclear Information System (INIS)

    Richards, L.M.; Szulinski, M.J.

    1979-01-01

    The Atlantic Richfield Company has developed the concept of storing spent fuel in dry caissons. Cooling is passive; safety and safeguard features appear promising. The capacity of a caisson to dissipate heat depends on site-specific soil characteristics and on the diameter of the caisson. It is estimated that approx. 2 kW can be dissipated in the length of one fuel element. Fuel elements can be stacked with little effect on temperature. A spacing of approx. 7.5 m (25 ft) between caissons appears rasonable. Business planning indicates a cost of approx. 0.2 mill/kWh for a 15-yr storage period. 12 figures, 4 tables

  16. Separation Nanotechnology of Diethylenetriaminepentaacetic Acid Bonded Magnetic Nanoparticles for Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kaur, Maninder [Idaho Univ., Moscow, ID (United States); Johnson, Andrew [Idaho Univ., Moscow, ID (United States); Tian, Guoxin [Lawrence Berkeley National Laboratory (LBNL), Berkeley, CA (United States); Jiang, Weilin [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Rao, Linfeng [Lawrence Berkeley National Laboratory (LBNL), Berkeley, CA (United States); Paszczynski, Andrzej [Idaho Univ., Moscow, ID (United States); Qiang, You [Center for Advanced Energy Studies, Idaho Falls, ID (United States); Idaho Univ., Moscow, ID (United States)

    2013-01-01

    A nanomagnetic separation method based on Diethylenetriaminepentaacetic acid (DTPA) conjugated with magnetic nanoparticles (MNPs) is studied for application in spent nuclear fuel separation. The high affinity of DTPA towards actinides aids in separation from the highly acidic medium of nuclear waste. The solubility and magnetization of particles at low pH is protected by encapsulating them in silica layer. Surface functionalization of silica coated particles with polyamines enhances the loading capacity of the chelators on MNPs. The particles were characterized before and after surface modification using transmission electron microscopy (TEM), helium ion microscopy (HIM), Fourier transform-infrared (FT-IR) spectrometry, and X-ray diffractometry. The coated and uncoated samples were studied using vibrating sample magnetometer (VSM) to understand the change in magnetic properties due to the influence of the surface functionalization. The hydrodynamic size and surface charge of the particles are investigated using Dynamic Light Scattering (DLS). The uptake behavior of Am(III), Pu(IV), U(VI), and Np(V) from 0.1M NaNO3 solution was investigated. The sorption result shows the strong affinity of DTPA towards Am(III) and Pu(IV) by extracting 97% and 80% of actinides, respectively. The high removal efficiency and fast uptake of actinides make the chelator conjugated MNPs an effective method for spent nuclear fuel separation.

  17. Experimental observations on electrorefining spent nuclear fuel in molten LiCl-KCl/liquid cadmium system

    International Nuclear Information System (INIS)

    Johnson, T. A.; Laug, D. V.; Li, S. X.; Sofu, T.

    1999-01-01

    Argonne National Laboratory (ANL) is currently performing a demonstration program for the Department of Energy (DOE) which processes spent nuclear fuel from the Experimental Breeder Reactor (EBR-II). One of the key steps in this demonstration program is electrorefining of the spent fuel in a molten LiCl-KCl/liquid cadmium system using a pilot scale electrorefiner (Mk-IV ER). This article summarizes experimental observations and engineering aspects for electrorefining spent fuel in the molten LiCl-KCl/liquid cadmium system. It was found that the liquid cadmium pool acted as an intermediate electrode during the electrorefining process in the ER. The cadmium level was gradually decreased due to its high vapor pressure and vaporization rate at the ER operational temperature. The low cadmium level caused the anode assembly momentarily to touch the ER vessel hardware, which generated a periodic current change at the salt/cathode interface and improved uranium recovery efficiency for the process. The primary current distributions calculated by numerical simulations were used in interpreting the experimental results

  18. An experimental study on Sodalite and SAP matrices for immobilization of spent chloride salt waste

    Science.gov (United States)

    Giacobbo, Francesca; Da Ros, Mirko; Macerata, Elena; Mariani, Mario; Giola, Marco; De Angelis, Giorgio; Capone, Mauro; Fedeli, Carlo

    2018-02-01

    In the frame of Generation IV reactors a renewed interest in pyro-processing of spent nuclear fuel is underway. Molten chloride salt waste arising from the recovering of uranium and plutonium through pyro-processing is one of the problematic wastes for direct application of vitrification or ceramization. In this work, Sodalite and SAP have been evaluated and compared as potential matrices for confinement of spent chloride salt waste coming from pyro-processing. To this aim Sodalite and SAP were synthesized both in pure form and mixed with different glass matrices, i.e. commercially available glass frit and borosilicate glass. The confining matrices were loaded with mixed chloride salts to study their retention capacities with respect to the elements of interest. The matrices were characterized and leached for contact times up to 150 days at room temperature and at 90 °C. SEM analyses were also performed in order to compare the matrix surface before and after leaching. Leaching results are discussed and compared in terms of normalized releases with similar results reported in literature. According to this comparative study the SAP matrix with glass frit binder resulted in the best matrix among the ones studied, with respect to retention capacities for both matrix and spent fuel elements.

  19. 77 FR 2342 - Fifth Meeting: RTCA, Next Gen Advisory Committee

    Science.gov (United States)

    2012-01-17

    ... Metrics--an executive-level set of metrics that capture an overall status of NextGen implementation Other..., Washington Adjourn Attendance is open to the interested public but limited to space availability. With the...

  20. EPCGen2 Pseudorandom Number Generators: Analysis of J3Gen

    Directory of Open Access Journals (Sweden)

    Alberto Peinado

    2014-04-01

    Full Text Available This paper analyzes the cryptographic security of J3Gen, a promising pseudo random number generator for low-cost passive Radio Frequency Identification (RFID tags. Although J3Gen has been shown to fulfill the randomness criteria set by the EPCglobal Gen2 standard and is intended for security applications, we describe here two cryptanalytic attacks that question its security claims: (i a probabilistic attack based on solving linear equation systems; and (ii a deterministic attack based on the decimation of the output sequence. Numerical results, supported by simulations, show that for the specific recommended values of the configurable parameters, a low number of intercepted output bits are enough to break J3Gen. We then make some recommendations that address these issues.

  1. EPCGen2 pseudorandom number generators: analysis of J3Gen.

    Science.gov (United States)

    Peinado, Alberto; Munilla, Jorge; Fúster-Sabater, Amparo

    2014-04-09

    This paper analyzes the cryptographic security of J3Gen, a promising pseudo random number generator for low-cost passive Radio Frequency Identification (RFID) tags. Although J3Gen has been shown to fulfill the randomness criteria set by the EPCglobal Gen2 standard and is intended for security applications, we describe here two cryptanalytic attacks that question its security claims: (i) a probabilistic attack based on solving linear equation systems; and (ii) a deterministic attack based on the decimation of the output sequence. Numerical results, supported by simulations, show that for the specific recommended values of the configurable parameters, a low number of intercepted output bits are enough to break J3Gen. We then make some recommendations that address these issues.

  2. Ekspresi Gen Family Bcl-2 dan Ekspresi Gen Protein Kanal Ion Vdac1 pada Oligozoospermia

    Directory of Open Access Journals (Sweden)

    Arni Amir

    2014-05-01

    Full Text Available AbstrakSalah satu bentuk infertilitas pada pria adalah jumlah spermatozoa yang kurang dari normal (Oligozoospermia. Berbagai mekanisme diduga berperan dalam terjadinya oligozoospermia, mulai dari faktor fisik, stress, makanan dan molekuler. Penelitian ini merupakan kajian molekuler terhadap pengaruh ekspresi gen Bax, Bcl-2 dan VDAC terhadap terjadinya oligozoospermia. Penelitian dilakukan terhadap 40 sampel oligozoospermia dan sebagai kontrol adalah golongan normozoospermia dengan jumlah yang sama banyak. Sperma diisolasi menggunakan Percoll hypaque. Isolasi RNA menggunakan High Pure RNA isolation kit (Roche. Selanjutnya dilakukan pembuatan cDNA menggunakan First Strand cDNA synthetis (Roche. Tingkat ekspresi gen dinilai menggunakan mesin Light Cycler 2.0 (Roche. Primer didesain menggunakan software primer 3 plus dan spesifikasi primer dianalisis dengan Basic Local Alignment Search Tool (Blast. Tingkat ekspresi dibandingkan menggunakan Housekeeping gene (β-actin. Berdasarkan data penelitian diketahui tingkat ekspresi VDAC dan BAX lebih tinggi pada kelompok oligozoospermia dibanding normo (p < 0.05, namun tidak ditemukan perbedaan pada gen Bcl-2 (p > 0.05. Berdasarkan proporsi ekspresi ditemukan sebagian besar kelompok oligozoospermia mengalami overekspresi pada gen VDAC dan BAX (90% dan 93.3%, namun tidak ada overkekspresi pada normo (p>0.05. Pada penelitian ini dapat disimpulkan bahwa overekspresi VDAC dan BAC berkaitan dengan kejadian oligozoospermiaKata kunci: Oligozoospermia, VDAC, Bax, Bcl-2, ekspresi genAbstractOne form of infertility in men is the number of spermatozoa which is less than normal (Oligozoospermia. Various mechanisms though have a role in the occurrence of oligozoospermia, it can be from physical factors, stress, food and molecular.This research is a molecular study of gen expression, Bax, Bcl2 and VDAC influence of the oligozoospermia’s occurrence.The research is conducted on 40 samples of oligozoospermia and as a

  3. Confirmatory factor analysis of the WAIS-IV/WMS-IV.

    Science.gov (United States)

    Holdnack, James A; Xiaobin Zhou; Larrabee, Glenn J; Millis, Scott R; Salthouse, Timothy A

    2011-06-01

    The Wechsler Adult Intelligence Scale-fourth edition (WAIS-IV) and the Wechsler Memory Scale-fourth edition (WMS-IV) were co-developed to be used individually or as a combined battery of tests. The independent factor structure of each of the tests has been identified; however, the combined factor structure has yet to be determined. Confirmatory factor analysis was applied to the WAIS-IV/WMS-IV Adult battery (i.e., age 16-69 years) co-norming sample (n = 900) to test 13 measurement models. The results indicated that two models fit the data equally well. One model is a seven-factor solution without a hierarchical general ability factor: Verbal Comprehension, Perceptual Reasoning, Processing Speed, Auditory Working Memory, Visual Working Memory, Auditory Memory, and Visual Memory. The second model is a five-factor model composed of Verbal Comprehension, Perceptual Reasoning, Processing Speed, Working Memory, and Memory with a hierarchical general ability factor. Interpretative implications for each model are discussed.

  4. Semiautomatic spent-fuel-handling machine

    International Nuclear Information System (INIS)

    Stromberg, M.

    1990-01-01

    The technology for the total automation of the entire fuel handling operation, has been in existence for several years. The simplest form, or first phase of modernization, is the semiautomatic fuel handling positioning system. Several of these types of platforms are in existence today, and recently CIMCORP/PaR systems completed a semiautomatic spent-fuel handling machine (SFHM) built for Calvert Cliffs, owned by Baltimore Gas and Electric. CIMCORP has provided a semiautomated spent nuclear fuel handling system consisting of the following: (1) newly designed refueling platform bridge and trolley; (2) CIMCORP CIMROC 4000 based automatic controls technology; (3) closed circuit TV surveillance of fuel grappling operations; and (4) direct replacement of the original system provided in 1971. All SFHM motions are driven under computer control, with fully automatic bridge and trolley traverse and manually activated hoisting and grappling. Position feedback for motion control and position indication is provided by resolvers. In operation, the technician selects machine destination on a touch screen and the control system automatically positions the bridge and trolley at the desired location. Future automation of fuel grappling and hoisting can be preformed with relatively few machine modifications

  5. Antineutrino Monitoring of Spent Nuclear Fuel

    Science.gov (United States)

    Brdar, Vedran; Huber, Patrick; Kopp, Joachim

    2017-11-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries worldwide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this paper, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear-waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel elements as a function of time, and we illustrate the usefulness of antineutrino detectors in several benchmark scenarios. In particular, we demonstrate how a measurement of the antineutrino flux can help to reverify the contents of a dry storage cask in case the monitoring chain by conventional means gets disrupted. We then comment on the usefulness of antineutrino detectors at long-term storage facilities such as Yucca mountain. Finally, we put forward antineutrino detection as a tool in locating underground "hot spots" in contaminated areas such as the Hanford site in Washington state.

  6. European experience with spent fuel transport

    International Nuclear Information System (INIS)

    Hunter, I.A.

    1995-01-01

    Nuclear Transport Ltd has transported 5000 tonnes of spent fuel from 35 reactors in 8 European countries since 1972. Transport management is governed by the Quality Plan for: transport administration, packaging and shipment procedures at the shipping plant, operations at the power plant, and packaging and shipment organization at the power plant. Selection of a suitable carrier device is made with regard to the shipping plant requirements, physical limitations of the reactor, fuel characteristics, and transport route constraints. The transport plan is set up taking into account exploitation of the casks, reactor shut-down requirements, fuel acceptance plans at the reprocessing plant, and cask maintenance periods. A transport cycle involving spent fuel shipment to La Hague or to Sellafield takes typically two or four weeks, respectively. Most transports through Europe are by rail. A special-design railway ferry boat serves transports to the United Kingdom. Both wet or dry casks are employed. Modern casks are designed for high burnups and for oxide fuels. (J.B.)

  7. Development of Spent Fuel Examination Technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Park, K. J.; Shin, H. S. (and others)

    2007-04-15

    For the official operation of ACPF Facility Attachment based on facility declared DIQ was issued by IAEA and officialized upon ROK government approval. This procedure gives an essential ground to negotiate Joint Determination between governments of ROK and US. For ACPF process material accountability a neutron coincidence counting system was developed and calibrated with Cf-252 source. Its performance test demonstrated that over-all counting efficiency was about 21% with random error, 1.5% against calibration source, which found to be satisfactory to the expected design specification. A calibration curve derived by MCNP code with relationship between ASNC doublet counts vs. neutron activity of Cm-244 showed calibration constant to be 2.78x10E5 counts/s.g which would be used for initial ACP hot operation test. Nuclear material transportation and temporary storage system was established for active demonstration of advanced spent fuel management process line and would be directly applied to the effective management of wastes arising from active demonstration and would later contribute as a base data to development of inter hot-cell movement system in pyro-processing line. In addition, an optimal spent fuel for the ACP demonstration was selected and a computer code was developed as a tool to estimate the expected source term at each key measurement point of ACP.

  8. Present status of JMTR spent fuel shipment

    International Nuclear Information System (INIS)

    Miyazawa, Masataka; Watanabe, Masao; Yokokawa, Makoto; Sato, Hiroshi; Ito, Haruhiko

    2002-01-01

    The Japan Atomic Energy Research Institute (JAERI) has been consistently making the enrichment reduction of reactor fuels in cooperation with RERTR Program and FRR SNF Acceptance Program both conducted along with the U.S. Nuclear Non-Proliferation Policy and JMTR, 50 MW test reactor in Oarai Research Establishment, has achieved core conversion, from its initial 93% enriched UAl alloy to 45% enriched uranium-aluminide fuel, and then to the current 19.8% enriched uranium-silicide fuel. In order to return all of JMTR spent fuels, to be discharged from the reactor by May 12, 2006, to the U.S.A. by May 12, 2009, JAERI is planning the transportation schedule based on one shipment per year. The sixth shipment of spent fuels to U.S. was carried out as scheduled this year, where the total number of fuels shipped amounts to 651 elements. All of the UAl alloy elements have so far been shipped and now shipments of 45% enriched uranium-aluminide type fuels are in progress. Thus far the JMTR SFs have been transported on schedule. From 2003 onward are scheduled more then 850 elements to be shipped. In this paper, we describe our activities on the transportation in general and the schedule for the SFs shipments. (author)

  9. Spent fuel pool cleanup and stabilization

    International Nuclear Information System (INIS)

    Miller, R.L.

    1987-06-01

    Each of the plutonium production reactors at Hanford had a large water-filled spent fuel pool to provide interim storage of irradiated fuel while awaiting shipment to the separation facilities. After cessation of reactor operations the fuel was removed from the pools and the water levels were drawn down to a 5- to 10-foot depth. The pools were maintained with the water to provide shielding and radiological control. What appeared to be a straightforward project to process the water, remove the sediments from the basin, and stabilize the contamination on the floors and walls became a very complex and time consuming operation. The sediment characteristics varied from pool to pool, the ion exchange system required modification, areas of hard-pack sediments were discovered on the floors, special arrangements to handle and package high dose rate items for shipment were required, and contract problems ensued with the subcontractor. The original schedule to complete the project from preliminary engineering to final stabilization of the pools was 15 months. The actual time required was about 25 months. The original cost estimate to perform the work was $2,651,000. The actual cost of the project was $5,120,000, which included $150,000 for payment of claims to the subcontractor. This paper summarizes the experiences associated with the cleanup and radiological stabilization of the 100-B, -C, -D, and -DR spent fuel pools, and discusses a number of lessons learned items

  10. Systems impacts of spent fuel disassembly alternatives

    International Nuclear Information System (INIS)

    1984-07-01

    Three studies were completed to evaluate four alternatives to the disposal of intact spent fuel assemblies in a geologic repository. A preferred spent fuel waste form for disposal was recommended on consideration of (1) package design and fuel/package interaction, (2) long-term, in-repository performance of the waste form, and (3) overall process performance and costs for packaging, handling, and emplacement. The four basic alternative waste forms considered were (1) end fitting removal, (2) fission gas venting, (3) disassembly and close packing, and (4) shearing/immobilization. None of the findings ruled out any alternative on the basis of waste package considerations or long-term performance of the waste form. The third alternative offers flexibility in loading that may prove attractive in the various geologic media under consideration, greatly reduces the number of packages, and has the lowest unit cost. These studies were completed in October, 1981. Since then Westinghouse Electric Corporation and the Office of Nuclear Waste Isolation have completed studies in related fields. This report is now being published to provide publicly the background material that is contained within. 47 references, 28 figures, 31 tables

  11. Free-format RPG IV

    CERN Document Server

    Martin, Jim

    2013-01-01

    This how-to guide offers a concise and thorough introduction to the increased productivity, better readability, and easier program maintenance that comes with the free-format style of programming in RPG IV. Although free-format information is available in IBM manuals, it is not separated from everything else, thereby requiring hours of tedious research to track down the information needed. This book provides everything one needs to know to write RPG IV in the free-format style, and author Jim Martin not only teaches rules and syntax but also explains how this new style of coding has the pot

  12. Nuclear fission, today and tomorrow. From renaissance to technological breakthrough (generation IV) - Part II

    International Nuclear Information System (INIS)

    Van Goethem, Georges

    2010-01-01

    This paper is an overview of the current Euratom FP-7 research and training actions in innovative nuclear fission reactors and fuel cycle technologies, including partitioning and transmutation. It is based on the more than 40 invited lectures that were delivered by research project coordinators and by keynote speakers at the FISA-2009 Conference, organised by the European Commission DG Research/Euratom. The education and training programmes in nuclear fission and radiation protection are also discussed, aiming at continuously increasing the level of nuclear competences across the EU. It is necessary to consider the most recent nuclear fission technologies (Generations of Nuclear Power Plants): - GEN II: safety and reliability of nuclear facilities and energy independence; - GEN III: continuous improvement of safety and reliability, and increased industrial competitiveness in a growing energy market; - GEN IV: for increased sustainability, and proliferation resistance. The focus in this paper is on the design objectives and research issues associated to Generations IV systems that have been agreed upon internationally. Their benefits are discussed according to a series of ambitious criteria or technology goals established at the international level. One will have to produce not only electricity at lower costs but also heat at very high temperatures, while exploiting a maximum of fissile and fertile matters, and recycling all actinides, under safe and reliable conditions. Scientific viability studies and technological performance tests for each Generation IV system are now being carried out in many laboratories world-wide, in line with the intergovernmental GIF agreement. The ultimate phase of commercial deployment is foreseen for 2040. (orig.)

  13. Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report

    International Nuclear Information System (INIS)

    MORGAN, R.G.

    1999-01-01

    The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building)

  14. Estimación de parámetros genéticos para características productivas y reproductivas en los sistemas doble propósito del trópico bajo colombiano

    Directory of Open Access Journals (Sweden)

    A. P. Galeano

    2010-01-01

    Full Text Available Con el objetivo de estimar los componentes de varianza, las heredabilidades, repetibilidades y correlaciones genéticas y fenotípicas para la producción de leche por lactancia (PL, el peso al destete (PD, el intervalo entre partos (IEP y el Índice de Vaca (IV, de las hembras bovinas manejadas en los sistemas de producción de doble propósito del trópico bajo colombiano, se analizaron los registros productivos y reproductivos de 1.687 vacas registradas en la Asociación Colombiana de Criadores de Ganado en Doble Propósito (Asodoble, durante el periodo comprendido entre 1998 y 2007. Se empleó un modelo animal mixto que incluyó los efectos fijos del grupo contemporáneo (finca-sexo-época-año, la composición racial, y la duración de la lactancia como covariable; así como los efectos genéticos aleatorios del animal, el medio ambiente permanente y el residual. Las heredabilidades estimadas para IEP (0,04 y PD (0,11 fueron bajas, y moderadas para PL (0,35 e IV (0,24, respectivamente. La repetibilidad estimada para IEP fue baja (0,08, y para PL (0,41 e IV (0,31 moderada; en el caso de PD este valor fue igual a la heredabilidad (0,11. Las correlaciones genéticas y fenotípicas obtenidas entre PL y PD con respecto a IEP fueron positivas, y se determinó una asociación genética negativa entre PL y PD. Los resultados demostraron que el IV es un buen indicador, desde el punto de vista genético, de la eficiencia productiva y reproductiva de los animales manejados en estos sistemas productivos.

  15. Banque Cantonale de Genève

    CERN Multimedia

    Banque Cantonale de Genève

    2011-01-01

    7e Salon Immobilier BCGE le samedi 3 septembre 2011, de 8 h 30 à 13 h 00, au Centre de formation de Conches À cette occasion, les meilleurs spécialistes professionnels genevois de l’immobilier seront réunis en un seul et même lieu. Si vous le souhaitez, un conseiller spécialisé dans les financements hypothécaires évaluera vos possibilités d’investissement immobilier adaptées à votre situation personnelle. En parallèle, les plus importantes régies immobilières de Genève seront à votre disposition pour vous présenter leurs offres actuelles, ainsi que les projets immobiliers futurs et discuter avec vous de la meilleure stratégie à adopter pour trouver l’objet de vos rêves. De plus, vous aurez la possibilité...

  16. Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

  17. Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevan, Vijay [Univ. of Cincinnati, OH (United States); Carroll, Laura [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-06

    This research project, which includes collaborators from INL and ORNL, focuses on the study of alloy 617 and alloy 800H that are candidates for applications as intermediate heat exchangers in GEN IV nuclear reactors, with an emphasis on the effects of grain size, grain boundaries and second phases on the creep properties; the mechanisms of dislocation creep, diffusional creep and cavitation; the onset of tertiary creep; and theoretical modeling for long-term predictions of materials behavior and for high temperature alloy design.

  18. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Final Report

    International Nuclear Information System (INIS)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-01-01

    Final report of 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Mockups applied to design review of AP600/1000, Construction planning for AP 600, and AP 1000 maintenance evaluation. Proof of concept study also performed for GenIV PBMR models

  19. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Final report of 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Mockups applied to design review of AP600/1000, Construction planning for AP 600, and AP 1000 maintenance evaluation. Proof of concept study also performed for GenIV PBMR models.

  20. Estudios sobre plantas andinas,- IV

    Directory of Open Access Journals (Sweden)

    Cuatrecasas José

    1943-04-01

    Full Text Available Caracteres genéricos:-Capitulo radiado, con flores dimorfas, las externas apenas más largas que las interiores. Invólucro cónico, de brácteas pluriseriadas pero poco numerosas, flojas, lineal-lanceoladas, agudas, de consistencia herbácea incluso en la fructificación, más largas que el resto del capitulo. Receptáculo alveolado ,con el margen de las fositas ,escamoso-lacerado. Corolas exteriores femeninas, liguladas, con tubo muy corto y casi rectas, ,con 4 líneas longitudinales y 3 dientes gruesos y callosos, en 2-3 filas, con frecuencia provistas de un apéndice lineal en la garganta, rudimento de labio superior.

  1. VeVeRa-IV

    NARCIS (Netherlands)

    Evelien Eggink; Debbie Oudijk; Klarita Sadiraj

    2012-01-01

    Original title: VeVeRa-IV The Dutch population is set to age rapidly in the coming years. More and more people will also attain a very great age. This means that the need for home care and care provided in nursing or residential care homes will also increase. As part of the Long-term Care

  2. 11. IV avati Draakoni galeriis...

    Index Scriptorium Estoniae

    2005-01-01

    Tanel Saare (sünd. 1979) näitus "Gott und huhn episode IV: seed shower". Eksponeeritakse väljavõtteid aktsioonidest aastatel 2000-2004 Turus, Nürnbergis, Berliinis, Lohusalus ja Soulis. Osa aktsioone toimus koos rühmitusega Non Grata

  3. Phase IV of Drug Development

    Directory of Open Access Journals (Sweden)

    Viraj Suvarna

    2010-01-01

    Full Text Available Not all Phase IV studies are post-marketing surveillance (PMS studies but every PMS study is a phase IV study. Phase IV is also an important phase of drug development. In particular, the real world effectiveness of a drug as evaluated in an observational, non-interventional trial in a naturalistic setting which complements the efficacy data that emanates from a pre-marketing randomized controlled trial (RCT. No matter how many patients are studied pre-marketing in a controlled environment, the true safety profile of a drug is characterized only by continuing safety surveillance through a spontaneous adverse event monitoring system and a post-marketing surveillance/non-interventional study. Prevalent practice patterns can generate leads that could result in further evaluation of a new indication via the RCT route or even a signal that may necessitate regulatory action (change in labeling, risk management/minimization action plan. Disease registries are another option as are the large simple hybrid trials. Surveillance of spontaneously reported adverse events continues as long as a product is marketed. And so Phase IV in that sense never ends.

  4. A chiral Mn(IV)

    Indian Academy of Sciences (India)

    Singlecrystal X-ray analysis revealed that compound 1 crystallises in the monoclinic 21 space group with six mononuclear [MnIVL2] units in the asymmetric unit along with three solvent DMF molecules. In the crystal structure, each Mn(IV) complex, acting as the building unit, undergoes supramolecular linking through C-H ...

  5. Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors (Workshop Report)

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, RE

    2004-07-15

    The ''Workshop on Advanced Computational Materials Science: Application to Fusion and Generation IV Fission Reactors'' was convened to determine the degree to which an increased effort in modeling and simulation could help bridge the gap between the data that is needed to support the implementation of these advanced nuclear technologies and the data that can be obtained in available experimental facilities. The need to develop materials capable of performing in the severe operating environments expected in fusion and fission (Generation IV) reactors represents a significant challenge in materials science. There is a range of potential Gen-IV fission reactor design concepts and each concept has its own unique demands. Improved economic performance is a major goal of the Gen-IV designs. As a result, most designs call for significantly higher operating temperatures than the current generation of LWRs to obtain higher thermal efficiency. In many cases, the desired operating temperatures rule out the use of the structural alloys employed today. The very high operating temperature (up to 1000 C) associated with the NGNP is a prime example of an attractive new system that will require the development of new structural materials. Fusion power plants represent an even greater challenge to structural materials development and application. The operating temperatures, neutron exposure levels and thermo-mechanical stresses are comparable to or greater than those for proposed Gen-IV fission reactors. In addition, the transmutation products created in the structural materials by the high energy neutrons produced in the DT plasma can profoundly influence the microstructural evolution and mechanical behavior of these materials. Although the workshop addressed issues relevant to both Gen-IV and fusion reactor materials, much of the discussion focused on fusion; the same focus is reflected in this report. Most of the physical models and computational methods

  6. The borehole disposal of spent sources (BOSS)

    International Nuclear Information System (INIS)

    Heard, R.G.

    2002-01-01

    During the International Atomic Energy Agency (IAEA) Regional Training Course on 'The Management of Low-Level Radioactive Waste from Hospitals and Other Nuclear Applications' hosted by the Atomic Energy Corporation of SA Ltd. (AEC), now NECSA, during July/August 1995, the African delegates reviewed their national radioactive waste programmes. Among the issues raised, which are common to most African countries, were the lack of adequate storage facilities, lack of disposal solutions and a lack of equipment to implement widely used disposal concepts to dispose of their spent sources. As a result of this meeting, a Technical Co-operation (TC) project was launched to look at the technical feasibility and economic viability of such a concept. Phase I and II of the project have been completed and the results can be seen in three reports produced by NECSA. The Safety Assessment methodology used in the evaluation of the concept was that developed during the ISAM programme and detailed in Van Blerk's PhD thesis. This methodology is specifically developed for shallow land repositories, but was used in this case as the borehole need not be more than 100m deep and could fit into the definition of a shallow land disposal system. The studies found that the BOSS concept would be suitable for implementation in African countries as the borehole has a large capacity for sources and it is possible that an entire country's disused sources can be placed in a single borehole. The costs are a lot lower than for a shallow land trench, and the concept was evaluated using radium (226) sources as the most limiting inventory. The conclusion of the initial safety assessment was that the BOSS concept is robust, and provides a viable alternative for the disposal of radium needles. The concept is expected to provide good assurance of safety at real sites. The extension of the safety assessment to other types of spent sources is expected to be relatively straightforward. Disposal of radium needles

  7. Spent fuel management fee methodology and computer code user's manual

    International Nuclear Information System (INIS)

    Engel, R.L.; White, M.K.

    1982-01-01

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively

  8. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

    Lima Soares, M.L. de; Oliveira Lopes, M.J. de

    1986-01-01

    A review of the spent fuel management concepts generally adopted in several countries is presented, including an analysis of the brazilian situation. The alternatives are the reprocessing, the interim storage and the final disposal in a repository after appropriate conditioning. The commercial operating reprocessing facilities in the Western World are located in France and in the United Kingdom. In the USA the anti-reprocessing policy from 1977 changed in 1981, when the government supported the resumption of commercial reprocessing and designated the private sector as responsible for providing these services. Small scale facilities are operating in India, Italy, Japan and West Germany. Pilot plants for LWR fuel are being planned by Spain, Pakistan and Argentina. (Author) [pt

  9. Nevada commercial spent nuclear fuel transportation experience

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed.

  10. Management of spent solvents of reprocessing origin

    International Nuclear Information System (INIS)

    Manohar, S.; Srinivas, C.; Vincent, T.; Wattal, P.K.

    2001-01-01

    Spent solvents of reprocessing origin constitute a major portion of radioactive liquid organic wastes arising from nuclear activity. An in-depth study of this waste stream has led to the evolution of a complete management option, which addresses not only the concern of radioactivity but also its organic nature. This is based on alkaline hydrolysis of Tri-n-butyl phosphate (TBP), which converts it into aqueous soluble products, viz. sodium salt of dibutyl phosphoric acid and butanol. During the process of alkaline hydrolysis almost all the activity associated with the waste gets transferred into the aqueous phase. The recovered diluent virtually free of activity and TBP can be recycled, and in case of it not meeting reprocessing standards, can be incinerated. The process generated aqueous waste is found compatible with cement and can be immobilized in cement matrix. (author)

  11. Thermal analysis for spent fuel casks

    International Nuclear Information System (INIS)

    Wells, A.H.

    1986-01-01

    Thermal analyses for spent fuel storage or transportation must demonstrate that the heat produced in the fuel will be removed without causing excessive fuel cladding temperatures or cask surface temperatures. The time required to ship fuel from a reactor to another site is a matter of days so temperature control during shipment is a short term concern and temperatures of 1100-1200 degrees F are acceptable in a hypothetical fire scenario, but storage is envisioned for twenty years or more and temperatures are limited to 380 degrees C (716 F). Elevated temperatures do more that weaken fuel cladding: the materials of the fuel basket also weaken with higher temperature, thermal stresses may cause cracks to form in welds, and impact limiter materials may soften or degrade. The high temperatures of a fire cause loss of liquid neutron shields and may cause solid materials to char and insulate a cask after the fire

  12. Spent fuel management: reprocessing or storage

    International Nuclear Information System (INIS)

    Lima Soares, M.L. de; Oliveira Lopes, M.J. de.

    1986-01-01

    A review of the spent fuel management concepts generally adopted in several countries is presented, including an analysis of the brazilian situation. The alternatives are the reprocessing, the interim storage and the final disposal in a repository after appropriate conditioning. The commercial operating reprocessing facilities in the Western World are located in France and in the United Kingdom. In the USA the anti-reprocessing policy from 1977 changed in 1981, when the Government supported the resumption of commercial reprocessing and designated the private sector as responsible for providing these services. Small scale facilities are operating in India, Italy, Japan and West Germany. Pilot plant for LWR fuel are being planned by Spain, Pakistan and Argentina. (Author) [pt

  13. Characterization plan for Hanford spent nuclear fuel

    International Nuclear Information System (INIS)

    Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

    1994-12-01

    Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF

  14. Characterization plan for Hanford spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

    1994-12-01

    Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF.

  15. Advanced waste forms from spent nuclear fuel

    International Nuclear Information System (INIS)

    Ackerman, J.P.; McPheeters, C.C.

    1995-01-01

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed

  16. Nevada commercial spent nuclear fuel transportation experience

    International Nuclear Information System (INIS)

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed

  17. Calculating the plutonium in spent fuel elements

    International Nuclear Information System (INIS)

    Barnham, Keith

    1992-01-01

    Many members of the public are concerned about plutonium. They are worried about its environmental, health and proliferation risks. Fundamental to all such considerations are two related questions: how much plutonium do nuclear reactors produce ? and how accurately do the relevant authorities know these production figures ? These two questions have been studied with particular reference to the UK civil Magnox reactors. In 1990 these were still the only UK civil reactors whose spent fuel had been reprocessed to extract plutonium in routine production. It has not been possible to conclude that the relevant government industry and safeguard authorities are aware of how much plutonium these reactors produce and that the figures are known to the highest achievable accuracy. To understand why, this chapter will outline some of the history of the attempts to get answers to these two questions. (author)

  18. Potential Interface Issues in Spent Fuel Management

    International Nuclear Information System (INIS)

    2015-10-01

    This publication is an output of a series of meetings to identify and evaluate issues and opportunities associated with interfaces in the back end of the fuel cycle (BEFC) and to describe effective management approaches based on the experience of Member States. During the meetings, participants from Member States and other international organizations shared and evaluated the main interfaces and potential interface issues among the spent fuel storage, transport, reprocessing and disposal of the BEFC, and also reviewed the national approaches to addressing these issues. The aim of this publication is to provide an approach to identify the interfaces in the BEFC as well as the potential issues that should be addressed. It also aims at responding to the solutions Member States most often find to be effective and, in some cases, were adjusted or revisited to reach the fixed target. Most of the interfaces and issues are country specific, as evidenced by the variety and diversity of examples provided in this publication

  19. CONSIDERATIONS REGARDING ROK SPENT NUCLEAR FUEL MANAGEMENT OPTIONS

    OpenAIRE

    CHAIM BRAUN; ROBERT FORREST

    2013-01-01

    In this paper we discuss spent fuel management options in the Republic of Korea (ROK) from two interrelated perspectives: Centralized dry cask storage and spent fuel pyroprocessing and burning in sodium fast reactors (SFRs). We argue that the ROK will run out of space for at-reactors spent fuel storage by about the year 2030 and will thus need to transition centralized dry cask storage. Pyroprocessing plant capacity, even if approved and successfully licensed and constructed by that time, wil...

  20. Expansion of capacity of spent fuel pools and associated problems

    International Nuclear Information System (INIS)

    Francisco, J.L. De; Lopez-Cotarelo, J.; Ramos, J.M.

    1978-01-01

    Expanding the spent fuel storage pool capacity is a good solution for utilities facing the current shortage in fuel reprocessing capacity. The problems more likely to be found when expanding a spent fuel storage facility by using high density storage racks are reviewed. Basically three types of problems arise: 1) Problems related with the characteristics of the new facility. 2) Problems related with the works of expansion. 3) Problems related with the long term storage of large quantities of spent fuel. (author)

  1. Miscibility and oxidation rate of the simulated metallic spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    You, K. S.; Joo, J. S.; Shin, Y. J.; Oh, S. C. [KAERI, Taejon (Korea, Republic of)

    1999-10-01

    The simulated metallic spent fuel was fabricated by using Uranium, Neodymium and Palladium in order to study the miscibility of Neodymium and Palladium with Uranium. For analysis of long-term safty on the metallized spent fuel, the simulated metallic spent fuel was oxidized under pure oxygen environment at 183{approx}250 deg C. From the results, the oxidation rate correlation and activation energy were obtained.

  2. Introduction on the recycling of spent and disused radioactive sources

    International Nuclear Information System (INIS)

    Zhao Mingqiang; Zang Ruihua

    2011-01-01

    It is not only a stress of environment safety, but also a waste of huge resources to send directly to store spent and disused radioactive sources. This article reviews some important aspects of management suggestions recommended by IAEA and requirements of regulations in China for disposing the spent and disused radioactive sources. The present condition and benefit of recycling spent and disused sources are analyzed. Some suggestions on carrying out recycling in China are put forward too. (authors)

  3. Overview of the United States spent nuclear fuel program

    International Nuclear Information System (INIS)

    Hurt, W.L.

    1997-12-01

    As a result of the end of the Cold War, the mission of the US Department of Energy (DOE) has shifted from an emphasis on nuclear weapons development and production to an emphasis on the safe management and disposal of excess nuclear materials including spent nuclear fuel from both production and research reactors. Within the US, there are two groups managing spent nuclear fuel. Commercial nuclear power plants are managing their spent nuclear fuel at the individual reactor sites until the planned repository is opened. All other spent nuclear fuel, including research reactors, university reactors, naval reactors, and legacy material from the Cold War is managed by DOE. DOE's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal. This mission involves correcting existing vulnerabilities in spent fuel storage; moving spent fuel from wet basins to dry storage; processing at-risk spent fuel; and preparing spent fuel in road-ready condition for repository disposal. Most of DOE's spent nuclear fuel is stored in underwater basins (wet storage). Many of these basins are outdated, and spent fuel is to be removed and transferred to more modern basins or to new dry storage facilities. In 1995, DOE completed a complex-wide environmental impact analysis that resulted in spent fuel being sent to one of three principal DOE sites for interim storage (up to 40 years) prior to shipment to a repository. This regionalization by fuel type will allow for economies of scale yet minimize unnecessary transportation. This paper discusses the national SNF program, ultimate disposition of SNF, and the technical challenges that have yet to be resolved, namely, release rate testing, non-destructive assay, alternative treatments, drying, and chemical reactivity

  4. Nuclear Spent Fuel Management in Spain

    International Nuclear Information System (INIS)

    Zuloaga, P.

    2015-01-01

    The radioactive waste management policy is established by the Spanish Government through the Ministry of Industry, Tourism and Commerce. This policy is described in the Cabinet-approved General Radioactive Waste Plan. ENRESA is the Spanish organization in charge of radioactive waste and nuclear SFM and nuclear installations decommissioning. The priority goal in SFM is the construction of the centralized storage facility named Almacén Temporal Centralizado (ATC), whose generic design was approved by the safety authority, Consejo de Seguridad Nuclear. This facility is planned for some 6.700 tons of heavy metal. The ATC site selection process, based on a volunteer community’s scheme, has been launched by the Government in December 2009. After the selection of a site in a participative and transparent process, the site characterization and licensing activities will support the construction of the facility. Meanwhile, extension of the on-site storage capacity has been implemented at the seven nuclear power plants sites, including past reracking at all sites. More recent activities are: reracking performed at Cofrentes NPP; dual purpose casks re-licensing for higher burnup at Trillo NPP; transfer of the spent fuel inventory at Jose Cabrera NPP to a dry-storage system, to allow decommissioning operations; and licence application of a dry-storage installation at Ascó NPP, to provide the needed capacity until the ATC facility operation. For financing planning purposes, the long-term management of spent fuel is based on direct disposal. A final decision about major fuel management options is not made yet. To assist the decision makers a number of activities are under way, including basic designs of a geological disposal facility for clay and granite host rocks, together with associated performance assessment, and supported by a R&D programme, which also includes research projects in other options like advanced separation and transmutation. (author)

  5. Transportation and storage of foreign spent power reactor fuel

    International Nuclear Information System (INIS)

    1979-01-01

    This report describes the generic actions to be taken by the Department of Energy, in cooperation with other US government agencies, foreign governments, and international organizations, in support of the implementation of Administration policies with respect to the following international spent fuel management activities: bilateral cooperation related to expansion of foreign national storage capacities; multilateral and international cooperation related to development of multinational and international spent fuel storage regimes; fee-based transfer of foreign spent power reactor fuel to the US for storage; and emergency transfer of foreign spent power reactor fuel to the US for storage

  6. Status of Away From Reactor spent fuel storage program

    International Nuclear Information System (INIS)

    King, F.D.

    1979-07-01

    The Away From Reactor (AFR) Spent Fuel Program that the US Department of Energy established in 1977 is intended to preclude the shutting down of commercial nuclear power reactors because of lack of storage space for spent fuel. Legislation now being considered by Congress includes plans to provide storage space for commercial spent fuel beginning in 1983. Utilities are being encouraged to provide as much storage space as possible in their existing storage facilities, but projections indicate that a significant amount of AFR storage will be required. The government is evaluating the use of both existing and new storage facilities to solve this forecasted storage problem for commercial spent fuel

  7. Status of research reactor spent fuel world-wide

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    2004-01-01

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)

  8. Remarks on the transportation of spent fuel elements

    International Nuclear Information System (INIS)

    Krull, W.

    1992-01-01

    Information and data are provided on several aspects of the transportation of spent fuel elements. These aspects include contract, transportation, reprocessing batch size, and economical considerations. (author)

  9. Transportation capabilities study of DOE-owned spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1994-10-01

    This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

  10. Seleção preliminar de genótipos de pinheira em Bom Jesus-PI Preliminary selection of sugar apple genotypes in Bom Jesus county, Piauí state, Brazil

    Directory of Open Access Journals (Sweden)

    Ítalo Herbert Lucena Cavalcante

    2011-01-01

    Full Text Available A pinheira (Annona squamosa L. ocorre espontaneamente no Nordeste Brasileiro, onde é explorada de forma extrativista, caracterizando-se pela falta de manejo adequado e material genético selecionado. Nesse sentido, foi realizado um experimento com objetivo de avaliar a produtividade, as características físicas e químicas de frutos de dez genótipos de pinheira no município de Bom Jesus, PI. Adotou-se delineamento inteiramente casualizado, com tratamentos representados por dez genótipos de pinheira e três repetições. Foram avaliadas as seguintes variáveis: vitamina C, acidez titulável, sólidos solúveis, relação SS/AT "ratio", diâmetros longitudinal e transversal, relação DL/DT, número se sementes por fruto, massa dos frutos e produção por planta. Os genótipos apresentam diferenças quanto às características químicas, físicas e produtivas dos frutos. Os genótipos foram agrupados em sete grupos, com destaque para o grupo III (Gen-02 e grupo IV (Gen-05, fato que explicitou as diferenças entre os genótipos de pinheira quanto às características produtivas e químicas e físicas dos frutos. Genótipos Gen-01 e Gen-02 apresentam potencial para instalação em plantios comerciais, pela produtividade, formato do fruto ou por caracterizarem fontes naturais de vitamina C.The sugar apple (Annona squamosa L. is native to tropical America, occurring spontaneously in Northeastern Brazil, where it is exploited mainly as subsistence without adequate management and without genetic material selection. An experiment was developed aiming to evaluate yield, physical and chemical characteristics of the fruits of ten sugar apple genotypes in Bom Jesus, Piauí State, Brazil. A completely randomized design with treatments represented by ten genotypes and three replications was adopted. The following variables were evaluated: vitamin C, titratable acidity, soluble solids, SS/TA ratio, longitudinal diameter and transverse, LD/TD, number of

  11. Glycogen Storage Disease Type IV

    DEFF Research Database (Denmark)

    Bendroth-Asmussen, Lisa; Aksglaede, Lise; Gernow, Anne B

    2016-01-01

    molecular genetic analyses confirmed glycogen storage disease Type IV with the finding of compound heterozygosity for 2 mutations (c.691+2T>C and c.1570C>T, p.R524X) in the GBE1 gene. We conclude that glycogen storage disease Type IV can cause early miscarriage and that diagnosis can initially be made......A 30-yr-old woman presented with 2 consecutive miscarriages within 7 mo. Histopathologic examination of the placental tissue showed intracytoplasmic inclusion vacuoles with a strong reaction in Periodic acid-Schiff staining and a slightly pallor reaction in alcian blue staining. Additional...... on histopathologic examination. Genetic analysis is required to confirm the diagnosis and to offer prenatal genetic testing in future pregnancies....

  12. Survey of wet and dry spent fuel storage

    International Nuclear Information System (INIS)

    1999-07-01

    Spent fuel storage is one of the important stages in the nuclear fuel cycle and stands among the most vital challenges for countries operating nuclear power plants. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and for coordinating and encouraging closer co-operation among Member States. Spent fuel management is recognized as a high priority IAEA activity. In 1997, the annual spent fuel arising from all types of power reactors worldwide amounted to about 10,500 tonnes heavy metal (t HM). The total amount of spent fuel accumulated worldwide at the end of 1997 was about 200,000 t HM of which about 130,000 t HM of spent fuel is presently being stored in at-reactor (AR) or away-from-reactor (AFR) storage facilities awaiting either reprocessing or final disposal and 70,000 t HM has been reprocessed. Projections indicate that the cumulative amount generated by 2010 may surpass 340,000 t HM and by the year 2015 395,000 t HM. Part of the spent fuel will be reprocessed and some countries took the option to dispose their spent fuel in a repository. Most countries with nuclear programmes are using the deferral of a decision approach, a 'wait and see' strategy with interim storage, which provides the ability to monitor the storage continuously and to retrieve the spent fuel later for either direct disposal or reprocessing. Some countries use different approaches for different types of fuel. Today the worldwide reprocessing capacity is only a fraction of the total spent fuel arising and since no final repository has yet been constructed, there will be an increasing demand for interim storage. The present survey contains information on the basic storage technologies and facility types, experience with wet and dry storage of spent fuel and international experience in spent fuel transport. The main aim is to provide spent fuel

  13. Fast reactor development and worldwide cooperation in Generation-IV International Forum

    International Nuclear Information System (INIS)

    Sagayama, Yutaka

    2013-01-01

    Objectives of Gen-IV systems development: Goals: Four challenging technology goals have been defined to be applied to innovative nuclear reactor concepts in the 21st century: 1) Safety and Reliability (safe and reliable operation, no offsite emergency response); 2) Sustainability (effective fuel utilization, minimization of nuclear waste); 3) Proliferation Resistance & Physical Protection (to assure unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism); 4) Economic Competitiveness (life-cycle cost advantage over other energy resources). Phase: Each Generation-IV reactor system is one of three stages. 1) Viability Phase; 2) Performance Phase; 3) Demonstration Phase. Target: Commercial Deployment is expected around 2030s or beyond

  14. Development of spent fuel remote handling technology - Kinematic analysis of bilateral arms for abnormal spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyu Won; Yoo, Ju Sang; Kim, Jong Yoon [Chungbuk National University, Chongju (Korea)

    2000-03-01

    In the project of 'Development of Spent Fuel Remote Handling Technology', Preprocessing technique, mechanism and teleoperation technique are being developed. One of the mechanisms is a device for disassembling of the spent fuel bundle. However, there may be abnormal fuel bar among the fuel bundle, In this case the unpacking task will be difficult and dangerous. So, in that case, a force reflected teleoperation manipulator is desirable. The system is composed of a anthropomorphic input device at control site, power manipulator at remote site and control system. In this research, the forward and inverse kinematic equations of input device and manipulators has been solved, respectively. In addition, the mapping algorithm is proposed and shown using computer simulation. The reaction force of the telemanipulator with the environmental object is reflected through control system. The reaction force is decomposed into joint torque of the input device based on the jacobian equation. The obtained theoretical relations are verified through computer simulation and they will be used effectively in the spent fuel remote handling technology. 6 refs., 26 figs., 7 tabs. (Author)

  15. Uruguay; 2011 Article IV Consultation

    OpenAIRE

    International Monetary Fund

    2011-01-01

    This 2011 Article IV Consultation highlights that the growth momentum in Uruguay has continued into 2011 but a slowdown is under way, led by weaker exports and slower public investment. Uruguay’s economic and financial vulnerabilities are modest, and the government has reduced debt vulnerabilities significantly and built important financial buffers. Executive Directors have commended authorities’ skillful macroeconomic management that has underpinned Uruguay’s excellent economic performance, ...

  16. A Grade IV Isthmic Spondylolisthesis.

    Science.gov (United States)

    Kolber, Morey J; Hanney, William J

    2017-12-01

    A 15-year-old adolescent boy was referred by an orthopaedic surgeon for treatment of right gluteal and thigh pain. Following treatment that reduced his pain, he experienced acute onset of bilateral radiculopathy. He was referred back to the physician, who ordered radiographs and magnetic resonance imaging, which identified a grade IV isthmic spondylolisthesis at the L5-S1 level. J Orthop Sports Phys Ther 2017;47(12):971. doi:10.2519/jospt.2017.7547.

  17. Gen Y learners: just how concerned should we be?

    Science.gov (United States)

    White, Geoff; Kiegaldie, Debra

    2011-12-01

    This article is a response to expressions of concern from a range of sources, including reports of curriculum redesign to accommodate the characteristics of Gen Y, claims made in the press and concerns expressed by educators in the health professions. Are these concerns grounded in research and if so how can educators in the health professions respond? Narrative review and discussion. Generation Y is defined, and beliefs about its defining characteristics are reviewed. The key role of marketing companies in the creation and perpetuation of generalised views of Gen Y is addressed, and a review of recent research related to members of this generation is presented. The more extreme claims of Gen Y as a unique generation are critiqued and implications for teaching members of this generation are reviewed. General advice is provided in relation to teaching and supervising Gen Y. The numerous claims regarding the uniqueness of Gen Y can be found to largely arise from a limited number of poorly designed and implemented studies, often conducted by marketing companies with a vested interest in promoting a belief in the existence of another unique societal subgroup. The limited educational research conducted to date has emphasised the importance of the relational dimension of learner-teacher interaction as a highly valued attribute of quality teaching. © Blackwell Publishing Ltd 2011.

  18. Twenty years of experience in spent fuel shipment from German nuclear power plants - a view of the competent authority

    International Nuclear Information System (INIS)

    Fasten, Ch.; Mueller, U.; Alter, U.

    1994-01-01

    A survey of the transport of spent fuel in and from Germany during the last 20 years is presented. The spent fuel is now transported from the German nuclear power facilities to the reprocessing plants in France and the United Kingdom. In the past, there were also shipments to the former reprocessing plant WAK Karlsruhe (Germany), to the long-term storage facility CLAB (Sweden) and also from the former German Democratic Republic to the USSR. The transport of the spent fuel is carried out in specially built flasks requiring an extensive quality assurance programme. Due to the heavy weight of these packages, the shipments are mostly carried out by rail, but also by road and sea. An overview is given of the following matters: (i) quantities of spent fuel transport, (ii) organisation of transport (iii) licensing matters, and (iv) reported incidents. In addition, an analysis is included of the radiation exposure for normal conditions of transport, especially of the transport workers. Difficulties and hindrances during transport are also reported. (author)

  19. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T; John Mickalonis, J

    2006-09-27

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO{sub 2}) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO{sub 2} layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH{sub 4}F)/ammonium nitrate (NH{sub 4}NO{sub 3}) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO{sub 2} layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH{sub 4}){sub 2}ZrF{sub 6}) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination

  20. ANALISIS GEN HAEMAGGLUTININ PADA VIRUS CAMPAK LIAR

    Directory of Open Access Journals (Sweden)

    Subangkit Subangkit

    2015-05-01

    Full Text Available AbstrakPenyakit Campak disebabkan oleh virus campak yang termasuk genus Morbilivirus dan Family Paramyxoviridae. Penyakit campak masih menjadi masalah kesehatan karena masih ditemukan Kejadian Luar Biasa (KLB di Indonesia. Salah satu penyebab terjadinya KLB tersebut diduga sebagaiakibat perbedaan antigenesitas antara strain vaksin yang digunakan dengan strain virus campak liar yang beredar di Indonesia. Penelitian ini bertujuan mendapatkan gambaran tentang karakteristik genetik gen Haemagglutinin virus campak liar yang ada di Indonesia. Spesimen yang digunakan sebanyak 27 isolat virus penyebab KLB dari 17 propinsi selama periode tahun 2003-2010. Isolat virus dilakukan pemeriksaan secara RT-PCR dan sekuensing dengan metode Sanger. Hasil sekuensing dianalisis dengan menggunakan perangkat lunak Bioedit 7.0 dan MEGA 4.0. Hasil penelitian didapatkan perbedaan 10 asam amino antara virus campak strain vaksin CAM-70 dan virus campak liar pada posisi D416N; K424T; V451M; N455T; V466I; I473T; F476L; Y481S atau Y481N; H495N; G505D. Kesimpulan penelitian ini adalah terdapat perbedaan karakteristik genetik antara virus campak liar di Indonesia berbeda dengan strain virus vaksin CAM-70.Kata kunci : Campak, Analisis Molekuler, Hemagglutinin, CD46AbstractMeasles is caused by virus belonging to the genus Morbilivirus and Family Paramyxoviridae. Measles is still a public health problem because outbreak of measles still found in Indonesia. Outbreak is suspected as a result of differences in antigenicity between vaccine strains used with wild-type measles virus strains circulating in Indonesia. This study aims to get genetic characteristics of wild-type measles virus haemagglutinin gene in Indonesia. The specimens were used 27 viral isolates from 17 provinces period 2003-2010. Viral isolates examined by RT-PCR and sequencing with Sanger method. Sequencing analysis were conducted using Bioedit 7.0 and MEGA 4.0 software. The results showed 10 amino acid differences

  1. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  2. A drying system for spent fuel assemblies

    International Nuclear Information System (INIS)

    Suikki, M.; Warinowski, M.; Nieminen, J.

    2007-06-01

    The report presents a proposed drying apparatus for spent fuel assemblies. The apparatus is used for removing the moisture left in fuel assemblies during intermediate storage and transport. The apparatus shall be installed in connection with the fuel handling cell of an encapsulation plant. The report presents basic requirements for and implementation of the drying system, calculation of the drying process, operation, service and maintenance of the equipment, as well as a cost estimate. Some aspects of the apparatus design are quite specified, but the actual detailed planning and final selection of components have not been included. The report also describes actions for possible malfunction and fault conditions. An objective of the drying system for fuel assemblies is to remove moisture from the assemblies prior to placing the same in a disposal canister for spent nuclear fuel. Drying is performed as a vacuum drying process for vaporizing and draining the moisture present on the surface of the assemblies. The apparatus comprises two pieces of drying equipment. One of the chambers is equipped to take up Lo1-2 fuel assemblies and the other OL1-2 fuel assemblies. The chambers have an internal space sufficient to accommodate also OL3 fuel assemblies, but this requires replacing the internal chamber structure for laying down the assemblies to be dried. The drying chambers can be closed with hatches facing the fuel handling cell. Water vapour pumped out of the chamber is collected in a controlled manner, first by condensing with a heat exchanger and further by freezing in a cold trap. For reasons of safety, the exhaust air of vacuum pumps is further delivered into the ventilation outlet duct of a controlled area. The adequate drying result is ascertained by a low final pressure of about 100 Pa, as well as by a sufficient holding time. The chamber is built for making its cleaning as easy as possible in the event of a fuel rod breaking during a drying, loading or unloading

  3. Biodegradation of polycyclic aromatic hydrocarbons (PAHs) in spent ...

    African Journals Online (AJOL)

    The potential of Pleurotus pulmonarius and Pleurotus ostreatus on the degradation of PAHs in spent and fresh cutting fluids (SCF and FCF) contaminated soils was investigated. Different weights of soil samples were contaminated with varying composition (10, 20 or 30%) of spent and fresh cutting fluids separately then ...

  4. Spent nuclear fuel project quality assurance program plan

    International Nuclear Information System (INIS)

    Lacey, R.E.

    1997-01-01

    This main body of this document describes how the requirements of 10 CFR 830.120 are met by the Spent Nuclear Fuel Project through implementation of WHC-SP-1131. Appendix A describes how the requirements of DOE/RW-0333P are met by the Spent Nuclear Fuel Project through implementation of specific policies, manuals, and procedures

  5. Providing flexibility in spent fuel and vitrified waste management

    International Nuclear Information System (INIS)

    Bradley, N.; O'Tallamhain, C.; Brown, G.A.

    1986-01-01

    The UK Central Electricity Generating Board is pondering a decision to build a dry vault store as a buffer in its overall AGR spent fuel management programme. The application of the dry vault is not limited to fuel from gas cooled reactors, it can be used for spent LWR fuel and vitrified waste. A cutaway diagram of such a vault is presented. (UK)

  6. Adolescent Depression and Time Spent with Parents and Siblings

    Science.gov (United States)

    Desha, Laura N.; Nicholson, Jan M.; Ziviani, Jenny M.

    2011-01-01

    This study examines adolescent depressive symptoms and the quantity and quality of time spent by adolescents with their parents and siblings. We use measures of the quality of relationships with parents and siblings as proxy indicators for the quality of time spent with these social partners. The study emphasizes the salience of parent…

  7. Monitoring instrumentation spent fuel management program. Final report

    International Nuclear Information System (INIS)

    1979-01-01

    Preliminary monitoring system methodologies are identified as an input to the risk assessment of spent fuel management. Conceptual approaches to instrumentation for surveillance of canister position and orientation, vault deformation, spent fuel dissolution, temperature, and health physics conditions are presented. In future studies, the resolution, reliability, and uncertainty associated with these monitoring system methodologies will be evaluated

  8. Nanobiotechnology for the production of biofuels from spent tea ...

    African Journals Online (AJOL)

    Bioenergy is the only alternative and cheap source of energy which can be made easily available to the world. The present experiment included three steps for the conversion of spent tea (Camellia sinensis) into biofuels. In the first step, spent tea was gasified using Co nano catalyst at 300°C and atmospheric pressure.

  9. Brewer's spent grain: A review of its potentials and applications ...

    African Journals Online (AJOL)

    Brewer's spent grain: A review of its potentials and applications. S Aliyu, M Bala. Abstract. Most developing nations continuously produce abundant agro-industrial residues such as brewer's spent grain (BSG), which are underexploited. BSG as the main by-product of brewing industry, representing approximately 85% of ...

  10. Integrated data base for spent fuel and radwaste: inventories

    International Nuclear Information System (INIS)

    Notz, K.J.; Carter, W.L.; Kibbey, A.H.

    1982-01-01

    The Integrated Data Base (IDB) program provides and maintains current, integrated data on spent reactor fuel and radwaste, including historical data, current inventories, projected inventories, and material characteristics. The IDB program collects, organizes, integrates, and - where necessary - reconciles inventory and projection (I/P) and characteristics information to provide a coherent, self-consistent data base on spent fuel and radwaste

  11. The development of spent fuel storage process equipment

    International Nuclear Information System (INIS)

    Yoon, Wan Ki; Kim, Ho Dong; Kim, Ki Joon; Kim, Bum Hoe

    1992-02-01

    A nuclear material accounting system were designed to track the transitions of nuclear materials at the spent-fuel technology research facility. It is embedded in a distributed control system real-time structure of the system gives timely on-line accountancy. And performance of AC servo motor with fuzzy logic control and its applicability to spent fuel management were experimentally evaluated. (Author)

  12. Spent Bleaching Earth as a Pozzolanic Material | Muthengia ...

    African Journals Online (AJOL)

    This paper proposes a cementing material from spent bleaching earth (SBE). SBE consists mainly of residual oil and ... The setting time of a workable paste of the ashed SBE-lime material was well within the Kenya standard 02 1263 requirements for Portland pozzolana cements. Key words: Cement, pozzolana, Spent ...

  13. Electrochemical generation of fentons reagent to treat spent caustic wastewater

    International Nuclear Information System (INIS)

    Hansen, H. K.; Nunez, P.; Rodriguez, N.; Guzman, J.

    2009-01-01

    An important wastewater stream from oil refineries is the spent caustic. Caustic solutions are used as scrubbing agent during the desulphurization process to eliminate sulphur an mercaptans from oil and gasses. Spent caustic is classified as DOO3 (reactive sulphide) hazardous waste under the US Resource Conservation and Recovery Act (RCRA). (Author)

  14. Radwaste management and spent fuel management in JAVYS

    International Nuclear Information System (INIS)

    Bozik, M.; Strazovec, R.

    2010-01-01

    In this work authors present radwaste management and spent fuel management in JAVYS, a.s. Processing of radioactive wastes (RAW) in the Bohunice Radioactive Waste Processing Center and surface storage of RAW in National RAW Repository as well as Interim Spent fuel storage in Jaslovske Bohunice are presented.

  15. Estimación de parámetros genéticos para características productivas y reproductivas en los sistemas doble propósito del trópico bajo colombiano

    Directory of Open Access Journals (Sweden)

    A. P. Galeano

    2010-06-01

    Full Text Available Con el objetivo de estimar los componentes de varianza, las heredabilidades, repetibilidadesy correlaciones genéticas y fenotípicas para la producción de leche por lactancia(PL, el peso al destete (PD, el intervalo entre partos (IEP y el Índice de Vaca (IV,de las hembras bovinas manejadas en los sistemas de producción de doble propósitodel trópico bajo colombiano, se analizaron los registros productivos y reproductivosde 1.687 vacas registradas en la Asociación Colombiana de Criadores de Ganado enDoble Propósito (Asodoble, durante el periodo comprendido entre 1998 y 2007. Seempleó un modelo animal mixto que incluyó los efectos fijos del grupo contemporáneo(finca-sexo-época-año, la composición racial, y la duración de la lactancia comocovariable; así como los efectos genéticos aleatorios del animal, el medio ambientepermanente y el residual. Las heredabilidades estimadas para IEP (0,04 y PD (0,11fueron bajas, y moderadas para PL (0,35 e IV (0,24, respectivamente. La repetibilidadestimada para IEP fue baja (0,08, y para PL (0,41 e IV (0,31 moderada; en el casode PD este valor fue igual a la heredabilidad (0,11. Las correlaciones genéticas y fenotípicasobtenidas entre PL y PD con respecto a IEP fueron positivas, y se determinóuna asociación genética negativa entre PL y PD. Los resultados demostraron que el IVes un buen indicador, desde el punto de vista genético, de la eficiencia productiva yreproductiva de los animales manejados en estos sistemas productivos.

  16. Case histories of West Valley spent fuel shipments: Final report

    International Nuclear Information System (INIS)

    1987-01-01

    In 1983, NRC/FC initiated a study on institutional issues related to spent fuel shipments originating at the former spent fuel processing facility in West Valley, New York. FC staff viewed the shipment campaigns as a one-time opportunity to document the institutional issues that may arise with a substantial increase in spent fuel shipping activity. NRC subsequently contracted with the Aerospace Corporation for the West Valley Study. This report contains a detailed description of the events which took place prior to and during the spent fuel shipments. The report also contains a discussion of the shipment issues that arose, and presents general findings. Most of the institutional issues discussed in the report do not fall under NRC's transportation authority. The case histories provide a reference to agencies and other institutions that may be involved in future spent fuel shipping campaigns. 130 refs., 7 figs., 19 tabs

  17. Some factors to consider in handling and storing spent fuel

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1985-11-01

    This report includes information from various studies performed under the Wet Storage Task of the Behavior of Spent Fuel in Storage Project of the Commercial Spent Fuel Management (CSFM) Program at Pacific Northwest Laboratory. Wet storage experience has been summarized earlier in several other reports. This report summarizes pertinent items noted during FY 1985 concerning recent developments in the handling and storage of spent fuel and associated considerations. The subjects discussed include recent publications, findings, and developments associated with: (1) storage of water reactor spent fuel in water pools, (2) extended-burnup fuel, (3) fuel assembly reconstitution and reinsertion, (4) rod consolidation, (5) variations in the US Nuclear Regulatory Commission's definition of failed fuel, (6) detection of failed fuel rods, and (7) extended integrity of spent fuel. A list of pertinent publications is included

  18. Case histories of West Valley spent fuel shipments: Final report

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    In 1983, NRC/FC initiated a study on institutional issues related to spent fuel shipments originating at the former spent fuel processing facility in West Valley, New York. FC staff viewed the shipment campaigns as a one-time opportunity to document the institutional issues that may arise with a substantial increase in spent fuel shipping activity. NRC subsequently contracted with the Aerospace Corporation for the West Valley Study. This report contains a detailed description of the events which took place prior to and during the spent fuel shipments. The report also contains a discussion of the shipment issues that arose, and presents general findings. Most of the institutional issues discussed in the report do not fall under NRC's transportation authority. The case histories provide a reference to agencies and other institutions that may be involved in future spent fuel shipping campaigns. 130 refs., 7 figs., 19 tabs.

  19. Arrival condition of spent fuel after storage, handling, and transportation

    International Nuclear Information System (INIS)

    Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

    1982-11-01

    This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables

  20. Compaction of spent nuclear fuel cans

    International Nuclear Information System (INIS)

    Sullivan, H.

    1985-01-01

    Hydraulic press apparatus for compacting waste material eg. spent nuclear fuel cans comprises a fixed frame, a movable cross head, a press crown and three groups of piston/cylinder devices; having their pistons connected to the cross head and their cylinders secured to the press crown. A control means connects the first group of devices to hydraulic fluid in a reservoir which is pressurised initially by gas from gas accumulators to move the cross head and a quill secured thereto towards the frame base to compact the waste at a first high rate under a first high loading. Compaction then proceeds at a lower second rate at a lower second loading as the hydraulic fluid in the reservoir is pressurised by a pump. At two subsequent stages of compaction of the waste at which resistance increases causing a pressure rise in cylinders the control means causes hydraulic fluid to be passed to the second group of devices and thence to the third group of devices, the compaction rate reducing at each stage but the compaction force increasing. (author)

  1. Transporting spent reactor fuel: allegations and responses

    International Nuclear Information System (INIS)

    Jefferson, R.M.

    1983-03-01

    A January 1982 monthly newsletter from the Council on Economic Priorities (CEP) was entirely devoted to the presentation of a broad-ranging series of allegations that the transportation of spent fuel in particular, and other high-level radioactive materials by inference is currently being conducted in this country in an unsafe manner. This newsletter preceded the release of a book authored by Marvin Resnikoff on the same subject by over a year. This book titled The Next Nuclear Gamble contained substantially the same allegations as the newsletter, although the book devoted space to a greatly increased number of specific examples. This paper reduces those allegations contained in the executive summary and the recommendations contained in the last chapter of the book to a manageable number by combining the many specific issues into a few topics. Each of these topics is then addressed. As such, this is an abbreviated analysis of The Next Nuclear Gamble and does not address much of the fine detail. In spite of that, it would be possible to address each of the details within the book on a similar basis. The intent of this document is to provide background information for those who are questioned on the validity of the allegations made by the CEp

  2. Processing of spent nickelcatalyst for fat recovery

    Directory of Open Access Journals (Sweden)

    NASIR Mohammad Ibraim

    2001-01-01

    Full Text Available Spent nickel catalyst (SNC has the potential of insulting the quality of the environment in a number of ways. Its disposal has a pollution effect. Optimum recovery of fat from SNC, could save the environment and reduce the oil loss. Hexane has been the solvent of choice for oil extraction. Alternative solvents considered to have been safer have been evaluated. Hexane, isopropanol, ethanol and heptane were examined using soxhlet extraction. While hexane is more efficient in oil recovery from SNC, isopropanol proved to be very good in clear separation of oil from waste material and also provides high solvent recovery compared to other solvents. Isopropanol extraction with chill separation of miscella into lower oil-rich phase, and an upper, solvent-rich recyclable phase save mush energy of vaporization for distilling. An aqueous extraction process with immiscible solvent assisted was tested. Solvent like hexane added to SNC, and water added later with continuous stirring. The mixture was stirred for about 30 minutes, prior to centrifugation. Aqueous process extracted less amount of oil compared to solvent extraction.

  3. Spent fuel transport in fuel cycle

    International Nuclear Information System (INIS)

    Labrousse, M.

    1977-01-01

    The transport of radioactive substances is a minor part of the fuel cycle because the quantities of matter involved are very small. However the length and complexity of the cycle, the weight of the packing, the respective distances between stations, enrichment plants and reprocessing plants are such that the problem is not negligible. In addition these transports have considerable psychological importance. The most interesting is spent fuel transport which requires exceptionally efficient packaging, especially where thermal and mechanical resistance are concerned. To meet the safety criteria necessary for the protection of both public and users it was decided to use the maximum capacity consistent with rail transport and to avoid coolant fluids under pressure. Since no single type of packing is suitable for all existing stations an effort has been made to standardise handling accessories, and future trands are towards maximum automation. A discussion on the various technical solutions available for the construction of these packing systems is followed by a description of those used for the two types of packaging ordered by COGEMA [fr

  4. Spent Nuclear Fuel Project operational staffing plan

    International Nuclear Information System (INIS)

    Debban, B.L.

    1996-03-01

    Using the Spent Nuclear Fuel (SNF) Project's current process flow concepts and knowledge from cognizant engineering and operational personnel, an initial assessment of the SNF Project radiological exposure and resource requirements was completed. A small project team completed a step by step analysis of fuel movement in the K Basins to the new interim storage location, the Canister Storage Building (CSB). This analysis looked at fuel retrieval, conditioning of the fuel, and transportation of the fuel. This plan describes the staffing structure for fuel processing, fuel movement, and the maintenance and operation (M ampersand O) staffing requirements of the facilities. This initial draft does not identify the support function resources required for M ampersand O, i.e., administrative and engineering (technical support). These will be included in future revisions to the plan. This plan looks at the resource requirements for the SNF subprojects, specifically, the operations of the facilities, balances resources where applicable, rotates crews where applicable, and attempts to use individuals in multi-task assignments. This plan does not apply to the construction phase of planned projects that affect staffing levels of K Basins

  5. Spent LWR fuel-storage costs

    International Nuclear Information System (INIS)

    Clark, H.J.

    1981-01-01

    Expanded use of existing storage basins is clearly the most economic solution to the spent fuel storage problem. The use of high-density racks followed by fuel disassembly and rod storage is an order of magnitude cheaper than building new facilities adjacent to the reactor. The choice of a new storage facility is not as obvious; however, if the timing of expenditures and risk allowance are to be considered, then modular concepts such as silos, drywells, and storage casks may cost less than water basins and air-cooled vaults. A comparison of the costs of the various storage techniques without allowances for timing or risk is shown. The impact of allowances for discounting and early resumption of reprocessing is also shown. Economics is not the only issue to be considered in selecting a storage facility. The licensing, environmental impact, timing, and social responses must also be considered. Each utility must assess all of these issues for their particular reactors before the best storage solution can be selected

  6. Electrochemical studies on spent fuel corrosion processes

    International Nuclear Information System (INIS)

    Pablo, J. de; Casas, I.; Clarens, F.; Gimenez, J.; Rovira, M.

    2003-01-01

    This presentations is mainly based on the electrochemical studies carried out by the Canadian team and the research group of the Berlin University. Electrochemical studies allow to study separately both the anodic reaction which corresponds-sources on UO 2 -electrodes response is one of to the UO 2 dissolution and the cathodic reaction that is the reduction of the oxidants. By using intensity current-potential plots a mechanisms of UO 2 corrosion has been established. At-300 mV (vs SCE), irreversible oxidation of UO 2 takes place and dissolution begins. In the absence of complexing agents like carbonate, an oxidised layer is formed at 100 mV a stoichiometry close to UO 2 . In carbonate medium, the oxidized layer is not formed because the U(VI) formed is rapidly dissolved. Results in terms of dissolution rates obtained by electrochemical measurements are similar to the ones obtained in dissolution experiments by using flow through reactors and similar kinetic laws are obtained. The effect of external α and γ-sources on UO 2 -electrodes response is one of the few available data on the effects of radiolysis on the UO 2 dissolution rate and can offer a complementary knowledge to the spent fuel and α-doped pellets dissolution experiments. (Author)

  7. Enzymatic hydrolysis of spent coffee ground.

    Science.gov (United States)

    Jooste, T; García-Aparicio, M P; Brienzo, M; van Zyl, W H; Görgens, J F

    2013-04-01

    Spent coffee ground (SCG) is the main residue generated during the production of instant coffee by thermal water extraction from roasted coffee beans. This waste is composed mainly of polysaccharides such as cellulose and galactomannans that are not solubilised during the extraction process, thus remaining as unextractable, insoluble solids. In this context, the application of an enzyme cocktail (mannanase, endoglucanase, exoglucanase, xylanase and pectinase) with more than one component that acts synergistically with each other is regarded as a promising strategy to solubilise/hydrolyse remaining solids, either to increase the soluble solids yield of instant coffee or for use as raw material in the production of bioethanol and food additives (mannitol). Wild fungi were isolated from both SCG and coffee beans and screened for enzyme production. The enzymes produced from the selected wild fungi and recombinant fungi were then evaluated for enzymatic hydrolysis of SCG, in comparison to commercial enzyme preparations. Out of the enzymes evaluated on SCG, the application of mannanase enzymes gave better yields than when only cellulase or xylanase was utilised for hydrolysis. The recombinant mannanase (Man1) provided the highest increments in soluble solids yield (17 %), even when compared with commercial preparations at the same protein concentration (0.5 mg/g SCG). The combination of Man1 with other enzyme activities revealed an additive effect on the hydrolysis yield, but not synergistic interaction, suggesting that the highest soluble solid yields was mainly due to the hydrolysis action of mannanase.

  8. Malachite Green Adsorption by Spent Coffee Grounds

    Science.gov (United States)

    Syamimie Atirah Mat, Siti; Zati Hanani Syed Zuber, Sharifah; Rahim, Siti Kartini Enche Ab; Sohaimi, Khairunissa Syairah Ahmad; Halim, Noor Amirah Abdul; Fauziah Zainudin, Nor; Aida Yusoff, Nor; Munirah Rohaizad, Nor; Hidayah Ishak, Noor; Anuar, Adilah; Sarip, Mohd Sharizan Md

    2018-03-01

    In this work, the ability of spent coffee grounds (SCG) as a low-cost adsorbent to remove malachite green (MG) from aqueous solutions was studied. Batch adsorption tests were carried out to observe the effect of various experimental parameters such as contact time, initial concentration of malachite green and adsorbent dosage on the removal of dye. The results obtained show that the percentage of dye removal will decreased with the increased of initial concentration of dye in the range of 50 mg/L to 250 mg/L. Besides, percentage removal of dye was also found to be increased as the contact time increased until it reached equilibrium condition. The results also showed that the adsorbent dosage in range of 0.2 g to 1.0 g is proportional to the percentage removal of malachite green dye. Study on the kinetic adsorption and isotherm adsorption has also been investigated. The adsorption isotherm data were described by Langmuir isotherm with high-correlation coefficients while the experimental data showed the pseudo-second-order kinetics model was the best model for the adsorption of MG by SCG with the coefficients of correlation R2 > 0.9978.

  9. Mercury Sorption onto Malt Spent Rootlets

    Science.gov (United States)

    Manariotis, I. D.; Anagnostopoulos, V.; Karapanagioti, H. K.; Chrysikopoulos, C.

    2011-12-01

    Mercury is a metal of particular concern due to its toxicity even at relatively low concentrations. The maximum permissible level for mercury in drinking water set by the European Union is 0.001 mg/L. Mercury is released into the environment via four principal pathways: (1) natural processes; i.e. a volcanic eruption, (2) incidental to some other activity; i.e. coal burning power plants, (3) accidentally during the manufacture, breakage or disposal of products that have mercury put into them deliberately, and (4) direct use in industrial settings. The present study focuses on the removal of mercury (II) from aqueous solutions via sorption onto Malt Spent Rootlets (MSR). Batch experiments were conducted employing MSR with size ranging from 0.18 to 1 mm. The effects of pH, mercury concentration, contact time, and solid to liquid ratio on mercury sorption onto MSR were investigated. The highest mercury removal from the aqueous phase, of 41%, was observed at pH of 5.

  10. Spent fuel storage options: a critical appraisal

    International Nuclear Information System (INIS)

    Singh, K.P.; Bale, M.G.

    1990-01-01

    The delayed decisions on nuclear fuel reprocessing strategies in the USA and other countries have forced the development of new long-term irradiated fuel storage techniques, to allow a larger volume of fuel to be held on the nuclear station site after removal from the reactor. The nuclear power industry has responded to the challenge by developing several viable options for long-term onsite storage, which can be employed individually or in tandem. They are: densification of storage in the existing spent fuel pool; building another fuel pool facility at the plant site; onsite cask park, and on site vault clusters. Desirable attributes of a storage option are: Safety: minimise the number of fuel handling steps; Economy: minimise total installed, and O and M cost; Security: protection from anti-nuclear protesters; Site adaptability: available site space, earthquake characteristics of the region and so on; Non-intrusiveness: minimise required modifications to existing plant systems; Modularisation: afford the option to adapt a modular approach for staged capital outlays; and Maturity: extent of industry experience with the technology. A critical appraisal is made of each of the four aforementioned storage options in the light of these criteria. (2 figures, 1 table, 4 references) (Author)

  11. Development of Advanced Spent Fuel Management Process

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G. (and others)

    2007-06-15

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm{sup 2} and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields.

  12. La genética de las poblaciones centroamericanas

    Directory of Open Access Journals (Sweden)

    Barrantes, Ramiro

    2005-01-01

    Full Text Available Las poblaciones centroamericanas no han sido objeto de muchos estudios genéticos con la excepción de análisis esporádicos de la variación entre y dentro de los grupos amerindios y de origen africano ubicados en el área. No obstante, en los últimos 15 años se efectuaron investigaciones sistemáticas en este sentido incluyendo poblaciones mestizas, particularmente las de Costa Rica y Panamá. En los amerindios se efectuaron estudios detallados de su estructura genética y las relaciones filogenéticas entre poblaciones en un contexto regional y continental. De manera general estos grupos mantienen una baja diversidad genética. Por otra parte, estimativas de la mezcla génica en la población mestiza muestran diferencias dentro y entre los distintos países del área. Al Valle Central de Costa Rica se le atribuye la condición de aislado con una supuesta constitución genética homogénea favorable para los estudios de desequilibrio de ligamiento y asociaciones con algunas enfermedades hereditarias; sin embargo, los resultados obtenidos con relación el flujo génico direccional, la amalgama de etnias y otros componentes de su estructura como la variación temporal y espacial de la consanguinidad y la migración indican una mayor heterogeneidad. Se muestran los resultados y conclusiones obtenidas y se plantean las perspectivas y tendencias al mediano plazo que vinculan los enfoques y métodos de la epidemiología y la antropología genética y el estudio genético de poblaciones.

  13. Spent fuel management perspectives in the Russian Federation

    International Nuclear Information System (INIS)

    Shidlovskii, V.V.; Kudryavtsev, E.G.

    2002-01-01

    Today nuclear power plants in the Russian Federation Produce about 800 tHM of spent nuclear fuel (SNF) per year. About 100 tHM/year of Russian origin spent fuel is being reprocessed at the PA Mayak RT-1 plant. Most part of the spent fuel is being accumulated in wet storage facilities both at-reactor and away-from-reactor types. Due to economic difficulties construction of the new reprocessing plant near Krasnoyarsk (so called RT-2 plant) was delayed for at least 15 years. Near- term storage of the accumulated spent fuel is now the only option considered for the RBMK spent fuel and part of the VVER-1000 spent fuel. Dry storage of the above spent fuel both in casks and in vaults is considered as a preferable storage option for near-term future. First dry storages will be in operation before the year 2005. Besides part of the VVER-1000 spent fuel will be reprocessed at PA Mayak site after PT-1 reprocessing plant modernization. Increased amounts of submarine spent fuel resulted from disarmament process also will be reprocessed at the plant.Both current and delayed reprocessing will be the main spent fuel management option for the Russia in years to come. This will be accompanied by R and D activities to optimize and simplify chemical reprocessing and waste management operations to improve economics of the reprocessing activity. Partitioning of the waste streams, conditioning and long-term storage of immobilized minor actinides will be the most important part of the work to be performed in laboratory and semi-industrial scale. Future deployment of fast neutron reactors will facilitate both separated plutonium and MA utilization in a form of reactor fuel. (author)

  14. Estudios genómicos del dinoflaglado Alexandrium Catenella

    OpenAIRE

    Valenzuela, Pablo D.T.

    2013-01-01

    En este proyecto se usó la metodología de secuenciación de ESTs como parte fundamental de un proyecto destinado al descubrimiento de genes de A. catenella y a crear un recurso de información genómica para investigadores chilenos e internacionales en general. Hay mucha necesidad de contar con herramientas genómicas con las cuales estudiar la expresión génica y su regulación en este organismo para el cual casi no hay información molecular. Fondo de Fomento al Desarrollo Científico y Tecnológ...

  15. Applications of Dredging and Beach Fills in GenCade

    Science.gov (United States)

    2016-06-01

    the nourished beach and the natural beach berm height are identical. GenCade requires the modeler to input a beach fill as an added berm width...step in the simulation. A graphic of the volume of each shoal through time can be viewed by going to Display->Plot Wizard and selecting GenCade Inlet...Station. Kraus, N. C. 2002. Reservoir model for calculating natural sand bypassing and change in volume of ebb-tidal shoals, part I: Description. ERDC

  16. Geometric structures of Gen (n=34-39) clusters

    Science.gov (United States)

    Qin, Wei; Lu, Wen-Cai; Zang, Qing-Jun; Zhao, Li-Zhen; Chen, Guang-Ju; Wang, C. Z.; Ho, K. M.

    2010-06-01

    The structures of Gen (n=34-39) clusters were searched by a genetic algorithm using a tight-binding interatomic potential. First-principles calculations based on density functional theory were performed to further identify the lowest-energy structures. The calculated results show that Gen (n=34-39) clusters favor prolate or Y-shaped three-arm structures consisting of two or three small stable clusters (Ge6, Ge7, Ge9, or Ge10) linked by a Ge6 or Ge9 bulk unit. The calculated results suggest the transition point from prolate to Y-shaped three-arm structures appears at Ge35 or Ge36.

  17. Introducing AstroGen: the Astronomy Genealogy Project

    Science.gov (United States)

    Tenn, Joseph S.

    2016-12-01

    The Astronomy Genealogy Project (AstroGen), a project of the Historical Astronomy Division of the American Astronomical Society (AAS), will soon appear on the AAS website. Ultimately, it will list the world's astronomers with their highest degrees, theses for those who wrote them, academic advisors (supervisors), universities, and links to the astronomers or their obituaries, their theses when online, and more. At present the AstroGen team is working on those who earned doctorates with astronomy-related theses. We show what can be learned already, with just ten countries essentially completed.

  18. Molekularbiologische Untersuchungen am PKD1-Gen der Katze

    OpenAIRE

    Kappe, Eva Christina

    2008-01-01

    1.Ziel dieser Arbeit war es, eine die PKD in der deutschen Perserkatzenpopulation auslösende Mutation im felinen PKD1-Gen oder einen eng mit dem PKD-Phänotyp gekoppelten Marker zu identifizieren. Auf Grundlage der Ergebnisse sollte ein direkter oder indirekter Gentest entwickelt werden. Weiterhin sollten an der Pathogenese der PKD beteiligte Strukturen des PKD1-Gens ausfindig gemacht werden. 2.Die Literaturübersicht stellt die Erkrankung PKD bei der Perserkatze vor und geht auf den gegenwä...

  19. Genes and proteins of Escherichia coli (GenProtEc).

    Science.gov (United States)

    Riley, M; Space, D B

    1996-01-01

    GenProtEc is a database of Escherichia coli genes and their gene products, classified by type of function and physiological role and with citations to the literature for each. Also present are data on sequence similarities among E.coli proteins with PAM values, percent identity of amino acids, length of alignment and percent aligned. The database is available as a PKZip file by ftp from mbl.edu/pub/ecoli.exe. The program runs under MS-DOS on IMB-compatible machines. GenProtEc can also be accessed through the World Wide Web at URL http://mbl.edu/html/ecoli.html.

  20. A engenharia genética Genetic engineering

    OpenAIRE

    José Alberto Neves Candeias

    1991-01-01

    São abordados os progressos havidos com as técnicas de engenharia genética, capazes de alterar o potencial genético de um organismo, quer pela introdução, quer pela supressão de novos genes estruturais. São mencionadas algumas das aplicações em geral e, em particular, possibilidades de uso no campo da medicina. É feita uma análise crítica dos benefícios e riscos envolvidos.This paper deals with the progress made in genetic engineering techniques, capable of altering the genetic potential of a...

  1. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    International Nuclear Information System (INIS)

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs

  2. Determination of U(IV) by spectrophotometry

    International Nuclear Information System (INIS)

    Coutinho, C.M.C.; Bastos, E.T.R.; Bastos, M.B.R.

    1984-01-01

    The determination of uranium (IV) concentration in acid solutions in the presence of U (IV), Fe (II) and Ti (IV) by spectrophotometry is studied, aiming at controlling the solutions in the ion exchange columns, used for uranium enrichment by isotope exchange. (E.G.) [pt

  3. Nitridogermanate Nitrides Sr7[GeN4]N2 and Ca7[GeN4]N2

    OpenAIRE

    Junggeburth, Sebastian C.; Oeckler, Oliver; Johrendt, Dirk; Schnick, Wolfgang

    2008-01-01

    The alkaline earth nitridogermanate nitrides AE7[GeN4]N2 (AE = Ca, Sr) have been synthesized using a Na flux technique in sealed Ta tubes. According to single-crystal X-ray diffraction the isotypic compounds crystallize in space group Pbcn (No. 60) with Z = 4, (Sr7[GeN4]N2: a = 1152.6(2), b = 658.66(13), c = 1383.6(3) pm, V = 1050.5(4) × 106 pm3, R1 = 0.049; Ca7[GeN4]N2: a = 1082.6(2), b = 619.40(12), c = 1312.1(3) pm, V = 879.8(3) × 106 pm3, R1 = 0.016). Owing to the high N/Ge ratio, the com...

  4. Experimental observations on the role of the cadmium pool in Mark-IV ER

    International Nuclear Information System (INIS)

    Li, S. X.; Vaden, D.; Mariani, R. D.; Johnson, T. A.

    2000-01-01

    Argonne National Laboratory (ANL) has developed and demonstrated an electrometallurgical process for the Department of Energy (DOE) to treat sodium bonded spent nuclear fuel. One of the key steps in the demonstration was electrorefining the spent fuel in a molten LiCl-KCl-UCl 3 /liquid cadmium system using a pilot scale electrorefiner (Mark-IV ER). This article summarizes experimental observations and engineering aspects for the roles of the liquid cadmium during electrorefining spent fuel in the Mark-IV ER. It was found that the liquid cadmium pool acted as an intermediate electrode during the electrorefining process. The cadmium level was gradually decreased due to its high vapor pressure and vaporization rate at the ER operational temperature. The low cadmium level caused the anode assembly to electrically short with the ER vessel hardware, which resulted in the difficulties to determine the endpoint of uranium dissolution from the anode baskets and reducing the current efficiency. A reflux cadmium vapor trap was installed and has successfully prevented the cadmium level from decreasing

  5. 76 FR 22162 - Third Meeting RTCA NextGen Advisory Committee (NAC)

    Science.gov (United States)

    2011-04-20

    ... Performance Metrics. Preliminary Reports on Key NextGen Performance Indicators. NextGen Measurement... limited to space availability. With the approval of the chairman, members of the public may present oral...

  6. Spent fuel storage practices and perspectives for WWER fuel in Eastern Europe

    International Nuclear Information System (INIS)

    Takats, F.

    1999-01-01

    In this lecture the general issues and options in spent fuel management and storage are reviewed. Quantities of spent fuel world-wide and spent fuel amounts in storage as well as spent fuel capacities are presented. Selected examples of typical spent fuel storage facilities are discussed. The storage technologies applied for WWER fuel is presented. Description of other relevant storage technologies is included

  7. Existing and near future practices of spent fuel storage in Slovak Republic

    International Nuclear Information System (INIS)

    Mizov, J.

    1999-01-01

    In this paper existing and near future practices of spent fuel storage in Slovak Republic are discussed: (1) Reactor operation and spent fuel production; (2) Past policy in spent fuel storage; (3) Away-from-reactor (AFR) storage facility at Bohunice NPP site; (4) Present policy in spent fuel storage; (5) Final disposal of spent fuel

  8. Nuclear Data Needs for Generation IV Nuclear Energy Systems

    Science.gov (United States)

    Rullhusen, Peter

    2006-04-01

    Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S

  9. Development of pyrochemical spent fuel management in the UK

    Energy Technology Data Exchange (ETDEWEB)

    Banfield, Zara; Cogan, John; Farrant, Dave; Gaubert, Emmanuel; Hopkins, Phil; Lewin, Bob [BNFL - Nexia Solutions Limited, Workington Facility, B291 Trenches, Sellafield (United Kingdom)

    2006-07-01

    Nexia Solutions is undertaking a programme to investigate the role of pyrochemical techniques for spent nuclear fuel and legacy fuel management. The principal UK client is the energy unit, and the other clients are the Nuclear Decommissioning Authority (NDA), for legacy fuel conditioning, and BNFL's corporate investment in advanced reactor systems which is contributing to the Generation IV programme. The emphasis of the programme is pragmatic industrialization, which we have identified as key for the establishment of pyrochemical fuel management. From our experience operating fuel manufacture, power generation and reprocessing plant we know that the areas which require particular attention for successful implementation are: - Plant Interfaces, - Operability, - Process Definition, - Underpinning Science. Plant Interfaces encompass the definition of feeds, products, effluents and wastes and whether the process can meet the constraints they impose. Operability is concerned with the sustainability of the plant processes and is linked to the use of nil-maintenance continuous systems and elimination of batch / mechanical operations and maintenance. Process Definition focuses on the performance, control, recovery and safety of individual unit operations. Together these underpin industrial nuclear plant implementability. As an example, we have built a test rig to demonstrate molten salts transfers, since we consider this to be a capability without which pyrochemical processing will not be viable. Similarly, we have developed pilot scale electro-refiner designs for high continuous throughput and we are building development modules to underpin key features of the designs. Scientific work has been targeted at electro-refiner actinide partitioning and has been expanded to investigate other critical areas of the process which include efficient uranium / salt separation, salt clean up and the development of waste forms which perform at least as well as borosilicate glass

  10. Development of pyrochemical spent fuel management in the UK

    International Nuclear Information System (INIS)

    Banfield, Zara; Cogan, John; Farrant, Dave; Gaubert, Emmanuel; Hopkins, Phil; Lewin, Bob

    2006-01-01

    Nexia Solutions is undertaking a programme to investigate the role of pyrochemical techniques for spent nuclear fuel and legacy fuel management. The principal UK client is the energy unit, and the other clients are the Nuclear Decommissioning Authority (NDA), for legacy fuel conditioning, and BNFL's corporate investment in advanced reactor systems which is contributing to the Generation IV programme. The emphasis of the programme is pragmatic industrialization, which we have identified as key for the establishment of pyrochemical fuel management. From our experience operating fuel manufacture, power generation and reprocessing plant we know that the areas which require particular attention for successful implementation are: - Plant Interfaces, - Operability, - Process Definition, - Underpinning Science. Plant Interfaces encompass the definition of feeds, products, effluents and wastes and whether the process can meet the constraints they impose. Operability is concerned with the sustainability of the plant processes and is linked to the use of nil-maintenance continuous systems and elimination of batch / mechanical operations and maintenance. Process Definition focuses on the performance, control, recovery and safety of individual unit operations. Together these underpin industrial nuclear plant implementability. As an example, we have built a test rig to demonstrate molten salts transfers, since we consider this to be a capability without which pyrochemical processing will not be viable. Similarly, we have developed pilot scale electro-refiner designs for high continuous throughput and we are building development modules to underpin key features of the designs. Scientific work has been targeted at electro-refiner actinide partitioning and has been expanded to investigate other critical areas of the process which include efficient uranium / salt separation, salt clean up and the development of waste forms which perform at least as well as borosilicate glass. Other

  11. Cost analysis of spent nuclear fuel management

    International Nuclear Information System (INIS)

    Robertson, D.L.M.; Ford, L.M.

    1993-01-01

    The Department of Energy Civilian Radioactive Waste Management System (CRWMS) is chartered to develop a waste management system for the safe disposal of spent nuclear fuel (SNF) from the 131 nuclear power reactors in the United States and a certain amount of high level waste (HLW) from reprocessing operations. The current schedule is to begin accepting SNF in 1998 for storage at a Monitored Retrievable Storage (MRS) facility. Subsequently, beginning in 2010, the system is scheduled to begin accepting SNF at a permanent geologic repository in 2010 and HLW in 2015. At this time, a MRS site has not been selected. Yucca Mountain, Nevada is currently being evaluated as the candidate site for the repository for permanent geologic disposal of SNF. All SNF, with the possible exception of the SNF from the western reactors, is currently planned to be shipped to or through the MRS site en route to the repository. The repository will operate in an acceptance and performance confirmation phase for a 50 year period beginning in 2010 with an additional nine year closure and five year decontamination and decommissioning period. The MRS has a statutory maximum capacity of 15,000 Metric Tons Uranium (MTU), with a further restriction that it may not store more than 10,000 MTU until the repository begins accepting waste. The repository is currently scheduled to store 63,000 MTU of SNF and an additional 7,000 MTU equivalent of HLW for a total capacity of 70,000 MTU. The amended act specified the MRS storage limits and identified Yucca Mountain as the only site to be characterized. Also, an Office of the Nuclear Waste Negotiator was established to secure a voluntary host site for the MRS. The MRS, the repository, and all waste containers/casks will go through a Nuclear Regulatory Commission licensing process much like the licensing process for a nuclear power plant. Environmental assessments and impact statements will be prepared for both the MRS and repository

  12. Safe management of spent radiation source

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Sugiura, Nobuyuki; Valdezco, E.M.; Choi, Kwang-Sub

    2003-01-01

    Presented are 8 investigation reports concerning the safe management of spent radiation source (SRS) during the current 2 years. Four reports from Japan are: Scheme for SRS management (approach and present status of the SRS management and consideration toward solving problems); Current International Atomic Energy Agency (IAEA) activities related to safety of radiation sources (Chronology of action plan development, Outline of revised action plan, and Asian regional activities); Current status of SRS management in Japan (Regulation system, Obligations of licensed users, Regulatory system on sealed sources, Status in the incidents on sources occurred, Incident of source loss, and Incidents of orphan sources); and SRS management system in Japan (Current status of using of sealed sources, collection system of SRS-Japan Radioisotope Association (JRIA) services, and Disposal of SRS). Four reports from the Asian countries also concern the current statuses of SRS management in the Philippine (Radioactive waste sources, Waste management strategies, Conditioning of Ra sources, Ra project action plan, as low as reasonably achievable (ALARA) program, Dose assessment, Regulations on radioactive waste, Action plan on the safety and security of sources, IAEA Regional Demonstration Centers, and sitting studies for a near surface disposal facility); Thailand (Current status of using sealed sources, Inventory of SRS, and Current topics of SRS management); Indonesia (Principles of management of radiation sources, Legislative framework of SRS management practices, Regulatory on SRS, management of sealed SRS, management hurdles, and reported incidents); and Korea (Regulatory frame work, Collection systems of SRS, Radioisotope waste generation, Radiation exposure incident, and Scrap monitoring system). (N.I.)

  13. Spent fuel transportation accident: a state's involvement

    International Nuclear Information System (INIS)

    Neuweg, M.

    1978-01-01

    On February 9, 1978 at 8:20 p.m., the duty officer for the Illinois Radiological Assistance Team was notified that a shipment containing uranium and plutonium was involved in an accident near Gibson City, Illinois on Route 54. It was reported that a pig containing an unknown amount of uranium and plutonium was involved. The Illinois District 6A State Police were called to the scene and secured the area. The duty officer in the meantime learned after numerous telephone calls, approximately 1 hour after the first notice was received, that the pig actually was a 48,000 pound cask containing 6 spent fuel rods and the tractor-trailer had split apart and was blocking one lane of the highway. The shipment had departed from Dresden Nuclear Power Station, Morris, Illinois, enroute to Babcox and Wilcox in Lynchburg, Virginia. Initial reports indicated the vehicle had split apart. Actually, the semi-trailer bed had buckled beneath the cask due to apparent excess stress. The cask remained entirely intact and was not damaged, but the state highway was closed to traffic. The State Radiological Assistance Team was dispatched and arrived on the scene at 12:45 a.m. Immediate radiation monitoring revealed a reading of 4 milliroentgen per hour at 10 feet from the cask. No contamination existed nor was anyone exposed to radiation unnecessarily. The cask was transferred to a Tri-State semi-trailer vehicle the following morning at approximately 6:30 a.m. At 9:30 a.m., February 10, the new vehicle was again enroute to its destination. This incident demonstrated typical occurrences involving transportation radiation accident: misinformation and/or lack of information on the initial response notification, inaccuracies of radiation monitorings at the scene of the accident, inconsistencies concerning the occurrences of the accident and unfamiliar terminology utilized by personnel first on the scene, i.e., pig, cask, vehicle split apart, etc

  14. Neutron Generators for Spent Fuel Assay

    International Nuclear Information System (INIS)

    Ludewigt, Bernhard A.

    2010-01-01

    The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel (SNF) assemblies with non-destructive assay (NDA). The 14 NDA techniques being studied include several that require an external neutron source: Delayed Neutrons (DN), Differential Die-Away (DDA), Delayed Gammas (DG), and Lead Slowing-Down Spectroscopy (LSDS). This report provides a survey of currently available neutron sources and their underlying technology that may be suitable for NDA of SNF assemblies. The neutron sources considered here fall into two broad categories. The term 'neutron generator' is commonly used for sealed devices that operate at relatively low acceleration voltages of less than 150 kV. Systems that employ an acceleration structure to produce ion beam energies from hundreds of keV to several MeV, and that are pumped down to vacuum during operation, rather than being sealed units, are usually referred to as 'accelerator-driven neutron sources.' Currently available neutron sources and future options are evaluated within the parameter space of the neutron generator/source requirements as currently understood and summarized in section 2. Applicable neutron source technologies are described in section 3. Commercially available neutron generators and other source options that could be made available in the near future with some further development and customization are discussed in sections 4 and 5, respectively. The pros and cons of the various options and possible ways forward are discussed in section 6. Selection of the best approach must take a number of parameters into account including cost, size, lifetime, and power consumption, as well as neutron flux, neutron energy spectrum, and pulse structure that satisfy the requirements of the NDA instrument to be built.

  15. SACSESS – the EURATOM FP7 project on actinide separation from spent nuclear fuels

    Directory of Open Access Journals (Sweden)

    Bourg Stéphane

    2015-12-01

    Full Text Available Recycling of actinides by their separation from spent nuclear fuel, followed by transmutation in fast neutron reactors of Generation IV, is considered the most promising strategy for nuclear waste management. Closing the fuel cycle and burning long-lived actinides allows optimizing the use of natural resources and minimizing the long-term hazard of high-level nuclear waste. Moreover, improving the safety and sustainability of nuclear power worldwide. This paper presents the activities striving to meet these challenges, carried out under the Euratom FP7 collaborative project SACSESS (Safety of Actinide Separation Processes. Emphasis is put on the safety issues of fuel reprocessing and waste storage. Two types of actinide separation processes, hydrometallurgical and pyrometallurgical, are considered, as well as related aspects of material studies, process modeling and the radiolytic stability of solvent extraction systems. Education and training of young researchers in nuclear chemistry is of particular importance for further development of this field.

  16. Epidemiologia genética: epidemiologia, genética ou nenhuma das anteriores?

    Directory of Open Access Journals (Sweden)

    Aguinaldo Gonçalves

    1990-12-01

    Full Text Available No esforço de contribuir para melhor entendimento da identidade da Epidemiologia Genética, são revistas sua concepção, campo de atuação, métodos e técnicas pertinentes e algumas instâncias de aplicação. Entendendo-a como a área de interesse dos fatores genéticos das doenças e suas interações ambientais, apresenta-se seu campo de atuação como constituído por dois segmentos: um descritivo, que lida com conhecimento da distribuição de tais afecções em famílias e populações, seu impacto a nível do coletivo e sua vigilância epidemiológica, bem como o estudo de seus determinantes; o segundo, caracterizado pela intervenção, refere-se às respectivas medidas preventivas. Em que pese possível limitação pela não-consideração de todas as situações existentes, particular atenção é destinada à revisão de métodos e técnicas que possam ser convergentemente aplicados, a partir de procedimentos genéticos e epidemiológicos. Entre eles, destacam-se como estudos de casos tanto metodologias laboratoriais (como os dermatóglifos quanto quantitativos, como cálculo de herdabilidade e análise multivariada. Alguns objetos de estudo são tomados como instância de aplicação, por contarem com investigações específicas em nosso meio: a hanseníase, o hidrargirismo e a esquizofrenia.In an attempt to contribute to a better undestanding of the identity of Genetic Epidemiology, we review its conception, its field of influence, its appropriate methods and techniques and, at last, some of its applications. Genetic Epidemiology involves the study of genetic factors acting on diseases and on their environmental interactions. These includes two major areas: a descriptive one, related to the distribution of such conditions in families and populations, to the epidemiologic surveillance and to the study of determinants; and another characterized by intervention, which is related to preventive measures. Because of the dificulty in

  17. Loss of spent fuel pool cooling PRA: Model and results

    Energy Technology Data Exchange (ETDEWEB)

    Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

    1996-09-01

    This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 {times} 10{sup {minus}5} and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 {times} 10{sup {minus}3}. Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible.

  18. Potential exposures and health effects from spent fuel transportation

    International Nuclear Information System (INIS)

    Rogers, V.C.; Sandquist, G.M.; Sutherland, A.A.

    1987-01-01

    The radiation exposures and consequent health effects associated with normal operations and postulated accidents during transportation of spent fuel have been analyzed and evaluated and the results have been summarized in the Final Environmental Assessments issued by DOE. For normal, accident-free transport of spent fuel, radiation exposures arise from both gamma and neutron sources within the spent fuel cask. The neutrons result from the spontaneous fission of transuranic nuclides in the spent fuel. The neutron dose flux component was modeled using DISNEL, a generalized, one-dimensional, multi-energy group neutronics code. Computer program PATHRAE-T was then developed from the EPA code PATHRAE and was employed to determine the total, combined dose field, including both ground and sky scatter of neutrons and gamma photons for any position around a truck or rail spent fuel cask. Four activity classes, viz., caravan, traffic obstruction, resident and pedestrian proximity, and servicing of the cask transport vehicle were reviewed for maximum individual exposure assessments. Projected doses for typical activities under maximum exposure conditions were 6 mrem or less per event. A spent fuel rail cask containing up to 14 PWR spent fuel assemblies could conceivably be involved in a variety of rail related transportation accidents. PATHRAE-T was used to estimate doses from rail cask accidents involving the release of radioactive nuclides although a release from such accidents is highly unlikely. The maximum individual exposure, primarily due to inhalation, is about 10 rem and occurs about 70 meters downwind

  19. Loss of spent fuel pool cooling PRA: Model and results

    International Nuclear Information System (INIS)

    Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

    1996-09-01

    This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 x 10 -5 and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 x 10 -3 . Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible

  20. Spent fuel management: Current status and prospects 1993

    International Nuclear Information System (INIS)

    1994-02-01

    Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and it is still one of the most vital problems common to all countries with nuclear reactors. It begins with the discharge of spent fuel from a power or a research reactor and ends with its ultimate disposition, either by direct disposal or by reprocessing of the spent fuel. Two options exist at present - an open, once-through cycle with direct disposal of the spent fuel and a closed cycle with reprocessing of the spent fuel and recycling of plutonium and uranium in new mixed oxide fuels. The selection of a spent fuel strategy is a complex procedure in which many factors have to be weighed, including political, economic and safeguards issues as well as protection of the environment. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for the exchange of information and to co-ordinate and to encourage closer co-operation among Member States in certain research an development activities that are of common interest. Refs, figs and tabs

  1. Spent Fuel Transportation Package Performance Study - Experimental Design Challenges

    International Nuclear Information System (INIS)

    Snyder, A. M.; Murphy, A. J.; Sprung, J. L.; Ammerman, D. J.; Lopez, C.

    2003-01-01

    Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research

  2. Composition calculations by the KARATE code system for the spent-fuel samples from the Novovoronezh reactor

    International Nuclear Information System (INIS)

    Hordosy, G.

    2006-01-01

    KARATE is a code system developed in KFKI AERI. It is routinely used for core calculation. Its depletion module are now tested against the radiochemical measurements of spent fuel samples from the Novovoronezh Unit IV, performed in RIAR, Dimitrovgrad. Due to the insufficient knowledge of operational history of the unit, the irradiation history of the samples was taken from formerly published Russian calculations. The calculation of isotopic composition was performed by the MULTICEL module of program system. The agreement between the calculated and measured values of the concentration of the most important actinides and fission products is investigated (Authors)

  3. New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors

    Directory of Open Access Journals (Sweden)

    P. Balestra

    2016-01-01

    Full Text Available The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D© water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D© default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential, specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead, were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties.

  4. Gen IV Nuclear Energy Systems Interim Status Report on Pre-conceptual LFR Design Studies and Evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Halsey, W G; Brown, N W; Smith, C F; Sienicki, J J; Moisseytsev, A V; Kim, S J; Smith, M A; Yang, W S; Williamson, M; Li, N

    2005-02-09

    Previous pre-conceptual core neutronics and system thermal hydraulics calculations initiated the investigation of viability of a Small Secure Transportable Autonomous Reactor (SSTAR) lead-cooled small modular fast reactor concept.1 The calculations indicated that a single-phase natural circulation SSTAR reactor concept with good core reactor physics performance, good system thermal hydraulics performance, and a high Supercritical Carbon Dioxide (S-CO{sub 2}) Brayton cycle efficiency of 40 % may be viable at an electrical power of 18 MWe (45 MWt). Pre-conceptual studies of SSTAR viability have continued with the objective of improving the system thermal hydraulic performance and raising the plant efficiency as well as extending the neutronics analysis. This effort has been motivated by several considerations. First, the initial Pre-conceptual studies were focused upon a ''pancake'' core having a height-to-diameter of 0.5. It was found that a compact core with high average burn up could be realized with a height-to-diameter ratio of 0.8. Second, the initial assumed reactor vessel height of 12.2 meters limited the height of the Pb-to-CO{sub 2} in-reactor heat exchangers (HXs) which reduced the efficiency of supercritical carbon dioxide (S-CO2) Brayton cycle power converter. It was found that by increasing the reactor vessel height to 18 meters, the greater driving head for single-phase natural circulation would offset both the greater pressure drop of the 0.8 height-to-diameter ratio core as well as the pressure drop of taller HXs. This has enabled the plant efficiency to be increased from 40 to 43 % and the plant electrical power to be raised from 18 to 20 MWe. Third, reactivity feedback coefficients, which had previously not been generated for SSTAR, have now been calculated for the core. The reactivity feedback coefficients provide a basis for future investigation of the autonomous load following and passive shutdown behavior of the reactor. The current status of SSTAR and the Pre-conceptual viability studies are described below.

  5. Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

    2013-06-01

    Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

  6. Conceptual development of a test facility for spent fuel management

    International Nuclear Information System (INIS)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

    1997-01-01

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

  7. ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

    International Nuclear Information System (INIS)

    Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo

    2015-01-01

    The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management

  8. ENVI Model Development for Korean Nuclear Spent Fuel Options Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Sunyoung; Jeong, Yon Hong; Han, Jae-Jun; Lee, Aeri; Hwang, Yong-Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2015-10-15

    The disposal facility of the spent nuclear fuel will be operated from 2051. This paper presents the ENVI code developed by GoldSim Software to simulate options for managing spent nuclear fuel (SNF) in South Korea. The ENVI is a simulator to allow decision-makers to assist to evaluate the performance for spent nuclear fuel management. The multiple options for managing the spent nuclear fuel including the storage and transportation are investigated into interim storage, permanent disposal in geological repositories and overseas and domestic reprocessing. The ENVI code uses the GoldSim software to simulate the logistics of the associated activities. The result by the ENVI model not only produces the total cost to compare among the multiple options but also predict the sizes and timings of different facilities required. In order to decide the policy for spent nuclear management this purpose of this paper is to draw the optimum management plan to solve the nuclear spent fuel issue in the economical aspects. This paper is focused on the development of the ENVI's logic and calculations to simulate four options(No Reprocessing, Overseas Reprocessing, Domestic Reprocessing, and Overseas and Domestic Reprocessing) for managing the spent nuclear fuel in South Korea. The time history of the spent nuclear fuel produced from both the existing and future NPP's can be predicted, based on the Goldsim software made available very user friendly model. The simulation result will be used to suggest the strategic plans for the spent nuclear fuel management.

  9. Revision of Corallinaceae (Corallinales, Rhodophyta): recognizing Dawsoniolithon gen. nov., Parvicellularium gen. nov. and Chamberlainoideae subfam. nov. containing Chamberlainium gen. nov. and Pneophyllum.

    Science.gov (United States)

    Caragnano, Annalisa; Foetisch, Alexandra; Maneveldt, Gavin W; Millet, Laurent; Liu, Li-Chia; Lin, Showe-Mei; Rodondi, Graziella; Payri, Claude E

    2018-03-25

    A multi-gene (SSU, LSU, psbA and COI) molecular phylogeny of the family Corallinaceae (excluding the subfamilies Lithophylloideae and Corallinoideae) showed a paraphyletic grouping of six monophyletic clades. Pneophyllum and Spongites were reassessed and recircumscribed using DNA sequence data integrated with morpho-anatomical comparisons of type material and recently collected specimens. We propose Chamberlainoideae subfam. nov., including the type genus Chamberlainium gen. nov., with C. tumidum comb. nov. as the generitype, and Pneophyllum. Chamberlainium is established to include several taxa previously ascribed to Spongites, the generitype of which currently resides in Neogoniolithoideae. Additionally we propose two new genera, Dawsoniolithon gen. nov. (Metagoniolithoideae), with D. conicum comb. nov. as the generitype and Parvicellularium gen. nov. (subfamily incertae sedis), with P. leonardi sp. nov. as the generitype. Chamberlainoideae has no diagnostic morpho-anatomical features that enable one to assign specimens to it without DNA sequence data, and it is the first subfamily to possess both Type 1 (Chamberlainium) and Type 2 (Pneophyllum) tetra/bisporangial conceptacle roof development. Two characters distinguish Chamberlainium from Spongites: tetra/biasporangial conceptacle chamber diameter (300 μm in Spongites) and tetra/bisporangial conceptacle roof thickness (8 cells in Spongites). Two characters also distinguish Pneophyllum from Dawsoniolithon: tetra/bisporangial conceptacle roof thickness (8 cells in Dawsoniolithon) and thallus construction (dimerous in Pneophyllum vs. monomerous in Dawsoniolithon). This article is protected by copyright. All rights reserved. This article is protected by copyright. All rights reserved.

  10. Update History of This Database - GenLibi | LSDB Archive [Life Science Database Archive metadata

    Lifescience Database Archive (English)

    Full Text Available switchLanguage; BLAST Search Image Search Home About Archive Update History Data ...List Contact us GenLibi Update History of This Database Date Update contents 2014/03/25 GenLibi English arch...base Description Download License Update History of This Database Site Policy | Contact Us Update History of This Database - GenLibi | LSDB Archive ...

  11. 78 FR 5860 - Eighth Meeting: RTCA Next Gen Advisory Committee (NAC)

    Science.gov (United States)

    2013-01-28

    ...] NextGen Implementation Metrics--a recommendation for Key city pairs evaluation of Transcon/Regional City Pairs that can be used for NextGen metrics. Data sources for measuring NextGen fuel Impact [cir... Attendance is open to the interested public but limited to space availability. With the approval of the...

  12. PowerGen plc report and accounts 1994

    International Nuclear Information System (INIS)

    1994-01-01

    The annual report and accounts of PowerGen plc for the year 1994 are presented. Financial highlights are quoted, followed by the Chairman's statement, reviews by the Chief Executive and Financial Directors, reports by the Auditors and Directors, balance sheets and details of the consolidated profit and loss account and principal accounting policies. A four year summary and shareholder information are included. (UK)

  13. GenOVa: a computer program to generate orientational variants

    OpenAIRE

    Cayron, Cyril

    2007-01-01

    A computer program called GenOVa, written in Python, calculates the orientational variants, the operators (special types of misorientations between variants) and the composition table associated with a groupoid structure. The variants can be represented by three-dimensional shapes or by pole figures.

  14. Distributed Generation Market Demand Model (dGen): Documentation

    Energy Technology Data Exchange (ETDEWEB)

    Sigrin, Benjamin [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gleason, Michael [National Renewable Energy Lab. (NREL), Golden, CO (United States); Preus, Robert [National Renewable Energy Lab. (NREL), Golden, CO (United States); Baring-Gould, Ian [National Renewable Energy Lab. (NREL), Golden, CO (United States); Margolis, Robert [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2016-02-01

    The Distributed Generation Market Demand model (dGen) is a geospatially rich, bottom-up, market-penetration model that simulates the potential adoption of distributed energy resources (DERs) for residential, commercial, and industrial entities in the continental United States through 2050. The National Renewable Energy Laboratory (NREL) developed dGen to analyze the key factors that will affect future market demand for distributed solar, wind, storage, and other DER technologies in the United States. The new model builds off, extends, and replaces NREL's SolarDS model (Denholm et al. 2009a), which simulates the market penetration of distributed PV only. Unlike the SolarDS model, dGen can model various DER technologies under one platform--it currently can simulate the adoption of distributed solar (the dSolar module) and distributed wind (the dWind module) and link with the ReEDS capacity expansion model (Appendix C). The underlying algorithms and datasets in dGen, which improve the representation of customer decision making as well as the spatial resolution of analyses (Figure ES-1), also are improvements over SolarDS.

  15. Measuring Gen-Y Customer Experience in the Banking Sector

    Directory of Open Access Journals (Sweden)

    Kyguolienė Asta

    2017-12-01

    Full Text Available The article analyses customer experience as the subject of marketing research and presents methods for assessing customer experience. The results of empirical research revealing the Gen-Y customer experience in using the Lithuanian commercial banks’ services are presented.

  16. LWR spent fuel transmutation in a high power density fusion reactor

    International Nuclear Information System (INIS)

    Sahin, Suemer.; Uebeyli, Mustafa

    2004-01-01

    The prospect of light water reactor (LWR) spent fuel incineration in a high power density fusion reactor has been investigated. The neutron wall load is taken at 10 MW/m 2 and a refractory alloy (W-5Re) is used in the first wall. Neutron transport calculations are conducted over an operation period of 48 months on a simple experimental hybrid blanket in a cylindrical geometry with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S 8 -P 3 approximation. In the neutron rich environment, the tritium breeding ratio remains > 1.05 so that the tritium self-sufficiency is maintained for the fusion reactor. The presence of fissionable nuclear waste fuel in the investigated blanket causes significant energy amplification. The energy multiplication factor is ∼4 at startup and it increases steadily up to 5.55 during power plant operation so that even a modest fusion reactor can supply a significant quantity of electricity. In the course of nuclear waste incineration, most of the fissionable fuel is burnt in situ. In addition to that, excess fissile fuel production enhances the nuclear quality of the nuclear fuel. Starting with an initial cumulative fissile fuel enrichment (CFFE) value of the spent fuel of 2.172%, CFFE can reach 4% after an irradiation period of ∼12 months. Then the spent fuel becomes suitable for a new recharge in an LWR as a regenerated fuel. Further residence in the fusion blanket continues to upgrade the nuclear waste so that after 48 months, CFFE can reach such a high level (9%) that it becomes qualified to be used in a new type of the advanced high temperature reactors for the Generation-IV

  17. REVIEW OF SPENT FUEL INTEGRITY EVALUATION FOR DRY STORAGE

    OpenAIRE

    DONGHAK KOOK; JONGWON CHOI; JUSEONG KIM; YONGSOO KIM

    2013-01-01

    Among the several options to solve PWR spent fuel accumulation problem in Korea, the dry storage method could be the most realistic and applicable solution in the near future. As the basic objectives of dry storage are to prevent a gross rupture of spent fuel during operation and to keep its retrievability until transportation, at the same time the importance of a spent fuel integrity evaluation that can estimate its condition at the final stage of dry storage is very high. According to the n...

  18. Future spent nuclear fuel and radioactive waste infrastructure in Norway

    International Nuclear Information System (INIS)

    Soerlie, A.A.

    2002-01-01

    In Norway a Governmental Committee was appointed in 1991 to make an evaluation of the future steps that need to be taken in Norway to find a final solution for the spent nuclear fuel and for some other radioactive waste for which a disposal option does not exist today. The report from the Committee is now undergoing a formal hearing process. Based on the Committees recommendation and comments during the hearing the responsible Ministry will take a decision on future infrastructure in Norway for the spent nuclear fuel. This will be decisive for the future management of spent nuclear fuel and radioactive waste in Norway. (author)

  19. Analytical methodology and facility description spent fuel policy

    International Nuclear Information System (INIS)

    1978-08-01

    Three generic environmental impact statements (GEISs) on domestic fuels, foreign fuels, and storage charges are being prepared to provide environmental input into decisions on whether, and if so how the 1977 Presidential policy on spent fuel storage should be implmented. This report provides background information for two of these environmental impact statements: Storage of U.S. Spent Power Reactor Fuel and Storage of Foreign Spent Power Reactor Fuel. It includes the analytical methodology used in GEISs to assess the environmental effects and a description of the facilities used in the two GEISs

  20. Comparison of spent fuel management fee collection alternatives

    International Nuclear Information System (INIS)

    White, M.K.; Engel, R.L.

    1979-01-01

    Five alternative methods for recovering the costs of spent fuel management were evaluated. These alternatives consist of collecting the fee for various components of spent fuel management cost (AFR basin storage, transportation from AFR basin to the repository, packaging, repository, R and D, and government overhead) at times ranging from generation of power to delivery of the spent fuel to the government. The five fee collection mechanisms were analyzed to determine how well they serve the interests of the public and the electricity ratepayer

  1. 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.

  2. Handling, conditioning and disposal of spent sealed sources

    International Nuclear Information System (INIS)

    1990-02-01

    The series entitled ''Technical Manual for the Management of Low and Intermediate Level Wastes Generated at Small Nuclear Research Centres and by Radioisotope Users in Medicine, Research and Industry'' will serve as reference material to experts on technical assistance missions and provide ''direct know-how'' for technical staff in developing countries. This document is the first in the series. It provides the technical guidance and know-how necessary to permit developing Member States to safely handle, condition and store spent sealed radiation sources. It covers: characterization of sealed sources, legislation and regulations, management of spent sealed sources, transportation and disposal of spent sealed sources. 5 refs, 10 figs, 6 tabs

  3. The EBR-II spent fuel treatment program

    International Nuclear Information System (INIS)

    Lineberry, M.J.; McFarlane, H.F.

    1995-01-01

    Argonne National Laboratory has refurbished and equipped an existing hot cell facility for demonstrating a high-temperature electrometallurgical process for treating spent nuclear fuel from the Experimental Breeder Reactor-11. Two waste forms will be produced and qualified for geologic disposal of the fission and activation products. Relatively pure uranium will be separated for storage. Following additional development, transuranium elements will be blended into one of the high-level waste streams. The spent fuel treatment program will help assess the viability of electrometallurgical technology as a spent fuel management option

  4. Properties of light water reactor spent fuel cladding. Interim report

    International Nuclear Information System (INIS)

    Farwick, D.G.; Moen, R.A.

    1979-08-01

    The Commercial Waste and Spent Fuel Packaging Program will provide containment packages for the safe storage or disposal of spent Light Water Reactor (LWR) fuel. Maintaining containment of radionuclides during transportation, handling, processing and storage is essential, so the best understanding of the properties of the materials to be stored is necessary. This report provides data collection, assessment and recommendations for spent LWR fuel cladding materials properties. Major emphasis is placed on mechanical properties of the zircaloys and austenitic stainless steels. Limited information on elastic constants, physical properties, and anticipated corrosion behavior is also provided. Work is in progress to revise these evaluations as the program proceeds

  5. Analytical methodology and facility description spent fuel policy

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-01

    Three generic environmental impact statements (GEISs) on domestic fuels, foreign fuels, and storage charges are being prepared to provide environmental input into decisions on whether, and if so how the 1977 Presidential policy on spent fuel storage should be implmented. This report provides background information for two of these environmental impact statements: Storage of U.S. Spent Power Reactor Fuel and Storage of Foreign Spent Power Reactor Fuel. It includes the analytical methodology used in GEISs to assess the environmental effects and a description of the facilities used in the two GEISs.

  6. Transportation 2000. Spent fuel transportation trends in the new millenium

    International Nuclear Information System (INIS)

    Blee, David; Viebrock, James; Patterson, John

    1999-01-01

    The paper will provide a comparison of foreign research reactor spent fuel transportation today verses the assumptions used by the Department of Energy in the Environmental Impact Statement. In addition, it will suggest changes that are likely to occur in transportation logistics through the remainder of the U.S. spent fuel returns program. Cask availability, certification status, shipment strategy, cost issues, and public acceptance are among the topical areas that will be examined. Transportation requirements will be assessed in light of current participation in the returns program and the tendency for shipment plans to shift toward spent fuel return toward the end of the 13 year period of eligibility. (author)

  7. Improving of spent fuel monitoring in condition of Slovak wet interim spent fuel storage facility

    International Nuclear Information System (INIS)

    Miklos, M.; Krsjak, V.; Bozik, M.; Vasina, D.

    2008-01-01

    Monitoring of WWER fuel assemblies condition in Slovakia is presented in the paper. The leak tightness results of fuel assemblies used in Slovak WWER units in last 20 years are analyzed. Good experiences with the 'Sipping system' are described. The Slovak wet interim spent fuel storage facility in NPP Jaslovske Bohunice was build and put in operation in 1986. Since 1999, leak tests of WWER-440 fuel assemblies are provided by special leak tightness detection system 'Sipping in Pool' delivered by Framatome-ANP facility with external heating for the precise detection of active specimens. Another system for monitoring of fuel assemblies condition was implemented in December 2006 under the name 'SVYPP-440'. First non-active tests started at February 2007 and are described in the paper. Although those systems seems to be very effective, the detection time of all fuel assemblies in one storage pool is too long (several months). Therefore, a new 'on-line' detection system, based on new sorbent KNiFC-PAN for effective 134 Cs and 137 Cs activity was developed. This sorbent was compared with another type of sorbent NIFSIL and results are presented. The design of this detection system and its possible application in the Slovak wet spent fuel storage facility is discussed. For completeness, the initial results of the new system are also presented. (authors)

  8. Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Creer, J.M.

    1988-10-01

    REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs

  9. Identificación de mutaciones puntuales del gen de la 21-hidroxilasa en pacientes afectados con hiperplasia suprarrenal congénita.

    Directory of Open Access Journals (Sweden)

    Dora Fonseca

    2005-06-01

    Full Text Available lntroducción. La hiperplasia suprarrenal congénita es un trastorno autosómico recesivo debido a la inadecuada secreción de cortisol. Mas del 95% de los casos de hiperplasia suprarrenal congénita son causados por defectos del gen de la 21 hidroxilasa, CYP21A2 . Las manifestaciones clínicas incluyen la forma clásica y la forma no clásica. Objetivos. Determinar la frecuencia de las mutaciones puntuales P30L, IVS2-12AIC-G, Del 8pb, I172N, cluster Ex 6, V281L, Q318X, R356W y P453S en pacientes con hiperplasia suprarrenal congénita. Materiales y métodos. Se estudiaron 58 pacientes, de los cuales, 48 fueron clásicos y 10 no clásicos. Mediante PCR alelo-especifica y ACRS (Amplified Creation Restriction Sites, se analizaron 9 mutaciones puntuales del gen CYP21A2 y se determinó la frecuencia en la población analizada. Resultados. Los alelos afectados se identificaron en el 82,8% de los cromosomas. Las mutaciones mas frecuentes fueron: IVS2-12AIC-G (26,7%, Q318X (21,5%, V281L (12,1% e I172N (12,1%. Conclusiones. Las mutaciones mas frecuentes en Colombia son similares a las de otros países del mundo, excepto para Q318X que presentó una mayor frecuencia, pero similar a la de otros países latinoamericanos. Este hallazgo y la existencia de 17,2% de alelos no identificados puede indicar diferencia entre el acervo genético de las poblaciones. En la forma clásica perdedora de sal predominaron las mutaciones Q318X e IVS2-12AIC-G; en la virilizante simple, IVS2-12AIC-G e I172N y en la no clásica , V281L, lo cual esta relacionado con el grado de actividad enzimática. En la forma no clásica, se encontraron alelos severos en el 66,7% de los casos, lo que determina el riesgo de tener hijos afectados con la forma grave virilizante simple o perdedora de sal. Los resultados reportados permiten ofrecer asesoramiento genético y diagnóstico prenatal.

  10. Biochar characteristics produced from malt spent rootlets

    Science.gov (United States)

    Fotopoulou, Kalliopi N.; Karapanagioti, Hrissi K.; Manariotis, Ioannis D.

    2013-04-01

    Biochar is a carbon-rich material produced by heating biomass in an oxygen-limited environment. Biochar is mainly used as an additive to soils to sequester carbon and improve soil fertility as well as a sorbent for environmental remediation processes. Surface properties such as point of zero charge, surface area and pore volume, surface topography, surface functional groups and acid-base behavior are important factors, which affect sorption efficiency. Understanding the surface alteration of biochars increases our understanding of the pollutant-sorbent interaction. Malt spent rootlets (MSR) is a by-product formed during beer production, is inexpensive and is produced in high quantities. The objective of the present study was to characterize the surface properties of biochar produced from MSR, and to investigate the effect of thermal treatment conditions on key characteristics that affect sorptive properties. The surface area, the pore volume, and the average pore size of the biochars were determined using gas (N2) adsorption-desorption cycles using the Brunauer, Emmett, and Teller (BET) equation. Isotherms with 30 adsorption and 20 desorption points were conducted at liquid nitrogen temperature (77K). Open surface area and micropore volume were determined using t-plot method and Harkins & Jura equation. Total organic carbon was also determined because it is an important factor that affects sorption. Raw MSR demonstrates low surface area that increases by 1 order of magnitude by thermal treatment up to 750oC. At temperatures from 750 up to 900oC, pyrolysis results to biochars with surface areas 210-340 m2/g. For the same temperature range, a high percentage (46-73%) of the pore volume of the biochars is due to micropores. Similar results were observed for all the grain size fractions of the raw MSR. The up-scaling of the biochar production was easily performed by using increased biomass analogous to the bigger vessels used each time. Positive results were obtained

  11. Spent Nuclear Fuel (SNF) Project Product Specification

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major

  12. NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2012-01-01

    A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

  13. NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

    Energy Technology Data Exchange (ETDEWEB)

    Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (Norway))

    2012-01-15

    A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

  14. Prospect for spent fuel transportation system in China

    International Nuclear Information System (INIS)

    Li, X.Q.; Jiang, Y.Q.

    1998-01-01

    With the arising of spent fuel from nuclear power plants, a national policy of a closed nuclear fuel cycle has been determined. Following being stored at reactor sites for at least 5 years (with storage maximum of 10 years), spent fuel will be transferred into an away-from-reactor wet centralized storage facility, waiting for reprocessing. Therefore, China's spent fuel management activities would be involved in at-reactor storage, transportation, away-from-reactor storage and reprocessing, as well as certainly radioactive waste management. The status of the nuclear fuel cycle will be briefly discussed and the potential spent fuel transportation system in China is discussed in detail, covering administration, flask, terminals and transportation by sea, rail and road. (authors)

  15. Spent fuel storage at Prairie Island: January 1995 status

    International Nuclear Information System (INIS)

    Closs, J.; Kress, L.

    1995-01-01

    The disposal of spent nuclear fuel has been an issue for the US since the inception of the commercial nuclear power industry. In the past decade, it has become a critical factor in the continued operation of some nuclear power plants, including the two units at Prairie Island. As the struggles and litigation over storage alternatives wage on, spent fuel pools continue to fill and plants edge closer to premature shutdown. Due to the delays in the construction of a federal repository, many nuclear power plants have had to seek interim storage alternatives. In the case of Prairie Island, the safest and most feasible option is dry cask storage. This paper discusses the current status of the Independent Spent Fuel Storage Installation (ISFSI) Project at Prairie Island. It provides a historical background to the project, discusses the notable developments over the past year, and presents the projected plans of the Northern States Power Company (NSP) in regards to spent fuel storage

  16. Effects of Ganoderma lucidum Spent Mushroom Substrate Extract on ...

    African Journals Online (AJOL)

    ) from spent mushroom substrate on milk and serum immunoglobulin levels and serum antioxidant capacity. Methods: Forty cows within the same parity and stage of lactation and with similar body weight were randomly divided into four groups ...

  17. An economic analysis of spent fuel management and storage

    International Nuclear Information System (INIS)

    Nagano, Koji

    1998-01-01

    Spent fuel management is becoming a key issue not only in the countries that have already experienced years of nuclear operation but also in the Asian countries that started nuclear utilization rather lately. This paper summarizes the key aspects that essentially determine optimal conditions for desired spent fuel management strategies from the engineering-economic point of view, in both national and regional perspectives. The term 'desired' is intended to highlight positive and beneficial aspects of such strategies, namely mobile and timely exploitation of spent fuel storage. Among all, the economy of scale, the economy of scope, the learning-by-doing effect, and benefits of R and D are reviewed theoretically and empirically, and the paper overviews to what extent these factors are implemented in solving spent fuel management strategy optimization problem. (author)

  18. Method for storing spent nuclear fuel in repositories

    Science.gov (United States)

    Schweitzer, D.G.; Sastre, C.; Winsche, W.

    A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

  19. Storage of spent fuel from light water reactors

    International Nuclear Information System (INIS)

    Wolkenhauer, W.C.

    1976-01-01

    The effects of possible inadequate nuclear fuel reprocessing capability upon a public utility, Washington Public Power Supply System, are studied. The possible alternatives for storing spent fuel are reviewed

  20. International development within the spent nuclear fuel cycle

    International Nuclear Information System (INIS)

    Aggeryd, I.; Broden, K.; Gelin, R.

    1990-06-01

    The report gives a survey of the newest international development of the fuel processing and the spent nuclear fuel cycle. The transmutation technology of long lived nuclides is discussed in more details. (K.A.E)