WorldWideScience

Sample records for reactor interim safe

  1. 105-H Reactor Interim Safe Storage Project Final Report

    International Nuclear Information System (INIS)

    Ison, E.G.

    2008-01-01

    The following information documents the decontamination and decommissioning of the 105-H Reactor facility, and placement of the reactor core into interim safe storage. The D and D of the facility included characterization, engineering, removal of hazardous and radiologically contaminated materials, equipment removal, decontamination, demolition of the structure, and restoration of the site. The ISS work also included construction of the safe storage enclosure, which required the installation of a new roofing system, power and lighting, a remote monitoring system, and ventilation components.

  2. 105-C Reactor interim safe storage project technology integration plan

    International Nuclear Information System (INIS)

    Pulsford, S.K.

    1997-01-01

    The 105-C Reactor Interim Safe Storage Project Technology Integration Plan involves the decontamination, dismantlement, and interim safe storage of a surplus production reactor. A major goal is to identify and demonstrate new and innovative D and D technologies that will reduce costs, shorten schedules, enhance safety, and have the potential for general use across the RL complex. Innovative technologies are to be demonstrated in the following areas: Characterization; Decontamination; Waste Disposition; Dismantlement, Segmentation, and Demolition; Facility Stabilization; and Health and Safety. The evaluation and ranking of innovative technologies has been completed. Demonstrations will be selected from the ranked technologies according to priority. The contractor team members will review and evaluate the demonstration performances and make final recommendations to DOE

  3. Project management plan for Reactor 105-C Interim Safe Storage project

    International Nuclear Information System (INIS)

    Plagge, H.A.

    1996-09-01

    Reactor 105-C (located on the Hanford Site in Richland, Washington) will be placed into an interim safe storage condition such that (1) interim inspection can be limited to a 5-year frequency; (2) containment ensures that releases to the environmental are not credible under design basis conditions; and (3) final safe storage configuration shall not preclude or significantly increase the cost for any decommissioning alternatives for the reactor assembly.This project management plan establishes plans, organizational responsibilities, control systems, and procedures for managing the execution of Reactor 105-C interim safe storage activities to meet programmatic requirements within authorized funding and approved schedules

  4. Final hazard classification and auditable safety analysis for the 105-C Reactor Interim Safe Storage Project

    International Nuclear Information System (INIS)

    Rodovsky, T.J.; Larson, A.R.; Dexheimer, D.

    1996-12-01

    This document summarizes the inventories of radioactive and hazardous materials present in the 105-C Reactor Facility and the operations associated with the Interim Safe Storage Project which includes decontamination and demolition and interim safe storage of the remaining facility. This document also establishes a final hazard classification and verifies that appropriate and adequate safety functions and controls are in place to reduce or mitigate the risk associated with those operations

  5. Analysis of the interim safe storage of reactors at the Hanford site

    International Nuclear Information System (INIS)

    Wang Hailiang

    2014-01-01

    The nine production reactors, i.e. B, C, D, DR, F, H, KE, KW and N, at the Hanford site are all water-cooled and graphite-moderated reactors with natural uranium fuel. In 1993, the U.S. Department of Energy (DOE) decided to put eight production reactors (except for B) into Interim Safe Storage (ISS) for 75 years followed by deferred one-piece removal. Reactor B will remain as a national historical landmark. By the end of 2013, six reactors C, F, D, DR, H and N had been successfully put into the ISS. Reactors KE and KW will be put into the ISS in the coming years. Taking reactor C as an example, this paper mainly talks about how to put the production reactors in the Interim Safe Storage, e.g. how to make site preparation, how to construct the safe storage enclosure (SSE) and how to perform surveillance and maintenance during the ISS period, etc. (authors)

  6. Project management plan for the 105-C Reactor interim safe storage project. Revision 1

    International Nuclear Information System (INIS)

    Miller, R.L.

    1997-01-01

    In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

  7. Engineering and planning for reactor 105-C interim safe storage project subcontract no. 0100C-SC-G0001 conceptual design report. Volume 1

    International Nuclear Information System (INIS)

    1996-04-01

    The 105-C Reactor, one of eight surplus production reactors at the Hanford Site, has been proposed by the U.S. Department of Energy, Richland, Operations Office to be the first large-scale technology demonstration project in the decontamination and decommissioning (D ampersand D) focus area as part of the project for dismantlement and interim safe storage. The 105-C Reactor will be placed in an interim safe storage condition, then undergo the decontamination and decommissioning phase. After D ampersand D, the reactor will be placed in long- term safe storage. This report provides the conceptual design for these activities

  8. DQO Summary Report for 105-N/109-N Interim Safe Storage Project Waste Characterization

    Energy Technology Data Exchange (ETDEWEB)

    T. A. Lee

    2005-09-15

    The DQO summary report provides the results of the DQO process completed for waste characterization activities for the 105-N/109-N Reactor Interim Safe Storage Project including decommission, deactivate, decontaminate, and demolish activities for six associated buildings.

  9. DQO Summary Report for 105-N/109-N Interim Safe Storage Project Waste Characterization

    International Nuclear Information System (INIS)

    Lee, T.A.

    2005-01-01

    The DQO summary report provides the results of the DQO process completed for waste characterization activities for the 105-N/109-N Reactor Interim Safe Storage Project including decommission, deactivate, decontaminate, and demolish activities for six associated buildings.

  10. Inherently safe reactors

    International Nuclear Information System (INIS)

    Maartensson, Anders

    1992-01-01

    A rethinking of nuclear reactor safety has created proposals for new designs based on inherent and passive safety principles. Diverging interpretations of these concepts can be found. This article reviews the key features of proposed advanced power reactors. An evaluation is made of the degree of inherent safety for four different designs: the AP-600, the PIUS, the MHTGR and the PRISM. The inherent hazards of today's most common reactor principles are used as reference for the evaluation. It is concluded that claims for the new designs being inherently, naturally or passively safe are not substantiated by experience. (author)

  11. Prospects for inherently safe reactors

    International Nuclear Information System (INIS)

    Barkenbus, J.N.

    1988-01-01

    Public fears over nuclear safety have led some within the nuclear community to investigate the possibility of producing inherently safe nuclear reactors; that is, reactors that are transparently incapable of producing a core melt. While several promising designs of such reactors have been produced, support for large-scale research and development efforts has not been forthcoming. The prospects for commercialization of inherently safe reactors, therefore, are problematic; possible events such as further nuclear reactor accidents and superpower summits, could alter the present situation significantly. (author)

  12. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Ise, Takeharu

    1987-01-01

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  13. Materials for passively safe reactors

    International Nuclear Information System (INIS)

    Simnad, T.

    1993-01-01

    Future nuclear power capacity will be based on reactor designs that include passive safety features if recent progress in advanced nuclear power developments is realized. There is a high potential for nuclear systems that are smaller and easier to operate than the current generation of reactors, especially when passive or intrinsic characteristics are applied to provide inherent stability of the chain reaction and to minimize the burden on equipment and operating personnel. Taylor, has listed the following common generic technical features as the most important goals for the principal reactor development systems: passive stability, simplification, ruggedness, case of operation, and modularity. Economic competitiveness also depends on standardization and assurance of licensing. The performance of passively safe reactors will be greatly influenced by the successful development of advanced fuels and materials that will provide lower fuel-cycle costs. A dozen new designs of advanced power reactors have been described recently, covering a wide spectrum of reactor types, including pressurized water reactors, boiling water reactors, heavy-water reactors, modular high-temperature gas-cooled reactors (MHTGRs), and fast breeder reactors. These new designs address the need for passive safety features as well as the requirement of economic competitiveness

  14. Small intrinsically safe reactor implications

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1985-01-01

    Reviewing the history of nuclear power, it is found that peaceful uses of nuclear power are children of the war-like atom. Importance of special growth in a shielded environment is emphasized to exploit fully the advantages of nuclear power. Nuclear power reactors must be safe for their assimilation into society from the points of view of both technology and social psychology. ISR/ISER is identified as a missing link in the development of nuclear power reactors from this perspective and advocated for international development and utilization, being unleashed from the concerns of politicization, safety, and proliferation

  15. Method of safely operating nuclear reactor

    International Nuclear Information System (INIS)

    Ochiai, Kanehiro.

    1976-01-01

    Purpose: To provide a method of safely operating an nuclear reactor, comprising supporting a load applied to a reactor container partly with secondary container facilities thereby reducing the load borne by the reactor container when water is injected into the core to submerge the core in an emergency. Method: In a reactor emergency, water is injected into the reactor core thereby to submerge the core. Further, water is injected into a gap between the reactor container and the secondary container facilities. By the injection of water into the gap between the reactor container and the secondary container facilities a large apparent mass is applied to the reactor container, as a result of which the reactor container undergoes the same vibration as that of the secondary container facilities. Therefore, the load borne by the reactor container itself is reduced and stress at the bottom part of the reactor container is released. This permits the reactor to be operated more safely. (Moriyama, K.)

  16. Standards for safe operation of research reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The safety of research reactors is based on many factors such as suitable choice of location, design and construction according to the international standards, it also depends on well trained and qualified operational staff. These standards determine the responsibilities of all who are concerned with the research reactors safe operation, and who are responsible of all related activities in all the administrative and technical stages in a way that insures the safe operation of the reactor

  17. Licensing issues for inherently safe fast reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Lee, S.; Okrent, D.

    1986-01-01

    There has been considerable interest recently in a new generation of liquid metal reactor (LMR) concepts in the US. Some significant changes in regulatory philosophy will be required if the anticipated cost advantages of inherently safe designs are to be achieved. The defense in depth philosophy will need to be significantly re-evaluated in the context of inherently safe reactors. It is the purpose of this paper to begin such a re-evaluation of this regulatory philosophy

  18. SIR - small is safe [in reactor design

    International Nuclear Information System (INIS)

    Hayns, M.

    1989-01-01

    A joint USA-UK venture has been initiated to design a small nuclear reactor which offers low capital cost, greater flexibility and a potentially lower environmental impact. Called Safe Integral Reactor (SIR), the lead unit could be built in the United Kingdom Atomic Energy Authority's (UKAEA's) Winfrith site if the design is accepted by the UK Nuclear Installations Inspectorate (NII). This article describes the 320 MWe reactor unit that is the basis of the design being developed. (author)

  19. Intrinsically Safe and Economical Reactor (ISER)

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki; Asahi, Yoshiro

    1991-01-01

    The Intrinsically Safe and Economical Reactor (ISER) is designed based on the principle of a process inherent ultimate safe reactor, PIUS, a so-called inherently safe reactor (ISR). ISER has been developed joingly by the members of the Kanagawa Institute of Technology, the University of Tokyo, the Japan Atomic Energy Research Institute (JAERI) and several industrial firms in Japan. This paper describes the requirements for the next generation of power reactor, the safety design philosphy of ISR and ISER, the controllability of ISER and the results of analyses of some of the design-based accidents (DBA) of ISER, namely station blackout, accidents in which the pressurizer relief valve becomes jammed and stuck in open position and tube breaks in the steam generator. It is concluded that the ISER can ensure a wide range of contraollabitily and fuel integrity for all the analysed DBAs. (orig.)

  20. Iser: an international inherently safe reactor concept

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1988-01-01

    Iser is a modular standardised 200-300 MWe power reactor based on the PIUS principle. It differs from PIUS in being simpler, and making full use of existing steel-vessel-based LWR technology. Iser is an inherently safe reactor concept under development in Japan. It is a generic concept, not a patented commodity, and it is expected that an international association to develop the concept will be formed. (U.K.)

  1. Mirror Advanced Reactor Study interim design report

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

  2. Mirror Advanced Reactor Study interim design report

    International Nuclear Information System (INIS)

    1983-04-01

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design

  3. Safe actinide disposition in molten salt reactors

    International Nuclear Information System (INIS)

    Gat, U.

    1997-01-01

    Safe molten salt reactors (MSR) can readily accommodate the burning of all fissile actinides. Only minor compromises associated with plutonium are required. The MSRs can dispose safely of actinides and long lived isotopes to result in safer and simpler waste. Disposing of actinides in MSRs does increase the source term of a safety optimized MSR. It is concluded that the burning and transmutation of actinides in MSRs can be done in a safe manner. Development is needed for the processing to handle and separate the actinides. Calculations are needed to establish the neutron economy and the fuel management. 9 refs

  4. Safe new reactor for radionuclide production

    International Nuclear Information System (INIS)

    Gray, P.L.

    1995-01-01

    In late 1995, DOE is schedule to announce a new tritium production unit. Near the end of the last NPR (New Production Reactors) program, work was directed towards eliminating risks in current designs and reducing effects of accidents. In the Heavy Water Reactor Program at Savannah River, the coolant was changed from heavy to light water. An alternative, passively safe concept uses a heavy-water-filled, zircaloy reactor calandria near the bottom of a swimming pool; the calandria is supported on a light-water-coolant inlet plenum and has upflow through assemblies in the calandria tubes. The reactor concept eliminates or reduces significantly most design basis and severe accidents that plague other deigns. The proven, current SRS tritium cycle remains intact; production within the US of medical isotopes such as Mo-99 would also be possible

  5. Interim dry fuel storage for magnox reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, N [National Nuclear Corporation, Risley, Warrington (United Kingdom); Ealing, C [GEC Energy Systems Ltd, Whetstone, Leicester (United Kingdom)

    1985-07-01

    In the UK the practice of short term buffer storage in water ponds prior to chemical reprocessing had already been established on the early gas cooled reactors in Calder Hall. Thus the choice of water pond buffer storage for MGR power plants logically followed the national policy decision to reprocess. The majority of the buffer storage period would take place at the reprocessing plant with only a nominal of 100 days targeted at the station. Since Magnox clad fuel is not suitable for long term pond storage, alternative methods of storage on future stations was considered desirable. In addition to safeguards considerations the economic aspects of the fuel cycle has influenced the conclusion that today the purchase of a MGR power plant with dry spent fuel storage and without commitment to reprocess would be a rational decision for a country initiating a nuclear programme. Dry storage requirements are discussed and two designs of dry storage facilities presented together with a fuel preparation facility.

  6. Interim dry fuel storage for magnox reactors

    International Nuclear Information System (INIS)

    Bradley, N.; Ealing, C.

    1985-01-01

    In the UK the practice of short term buffer storage in water ponds prior to chemical reprocessing had already been established on the early gas cooled reactors in Calder Hall. Thus the choice of water pond buffer storage for MGR power plants logically followed the national policy decision to reprocess. The majority of the buffer storage period would take place at the reprocessing plant with only a nominal of 100 days targeted at the station. Since Magnox clad fuel is not suitable for long term pond storage, alternative methods of storage on future stations was considered desirable. In addition to safeguards considerations the economic aspects of the fuel cycle has influenced the conclusion that today the purchase of a MGR power plant with dry spent fuel storage and without commitment to reprocess would be a rational decision for a country initiating a nuclear programme. Dry storage requirements are discussed and two designs of dry storage facilities presented together with a fuel preparation facility

  7. Brownfields and Urban Agriculture: Interim Guidelines for Safe Gardening Practices

    Science.gov (United States)

    This document is a condensation of the input of experts from the government, the nonprofit sector, and academia who gathered to outline the range of issues which need to be addressed in order to safely grow food on former brownfield sites.

  8. An inherently safe power reactor module

    International Nuclear Information System (INIS)

    Salerno, L.N.

    1985-01-01

    General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

  9. Inherently safe characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    1989-01-01

    This report is based on a detailed study which was carried out by Colenco (a company of the Motor-Columbus Group) on behalf of the Commission of the European Communities (CEC). It presents a summary of this study and concentrates more on the generic issues involved in the subject of inherent safety in nuclear power plants. It is assumed that the reader is reasonably familiar with the design outline of the systems included in the report. The report examines the role of inherent design features in achieving the safety of nuclear power plants as an alternative to the practice, which is largely followed in current reactors, of achieving safety by the addition of engineered safety features. The report examines current reactor systems to identify the extent to which their characteristics are either already inherently safe or, on the other hand, have inherent characteristics that require protective action to be taken. It then considers the advantages of introducing design changes to improve their inherent safety characteristics. Next, it looks at some new reactor types for which claims of inherent safety are made to see to what extent these claims are justified. The general question is then considered whether adoption of the inherently safe reactors would give advantages (by reducing risk in real terms or by improving the public acceptability of nuclear power) which are sufficient to offset the expected high costs and the technical risks associated with any new technology

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

    Energy Technology Data Exchange (ETDEWEB)

    Shedrow, B.

    1998-10-19

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

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

    International Nuclear Information System (INIS)

    Shedrow, B.

    1998-01-01

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

  12. Interim guidance for the safe transport of reprocessed uranium

    International Nuclear Information System (INIS)

    1994-06-01

    Increasingly reprocessed uranium is being used for the fabrication of nuclear fuel elements. Different intermediate reprocessing steps are carried out at different locations. Therefore, transportation of uranium material is necessary. Due to the difference in isotope composition of reprocessed uranium then unirradiated uranium a doubt is casted on the presumption that packages used for the transport of unirradiated uranium are automatically suitable for reprocessed uranium compounds. The Standing Advisory Group on the Safe Transport of Radioactive Material (SAGSTRAM) recommended that the issue be reviewed by consultants and that a document be developed that would give guidance to users of the Regulations. This TECDOC is the result of the endeavors of the experts convened at two Consultants Services meetings. It contains guidance on the provisions in the current Regulations as well as proposals for changes to the new Revised Edition whose publication is planned for 1996. This document demonstrates that under the present Transport Regulations it is possible in most cases to ship reprocessed uranium compounds in the same packages as unirradiated uranium compounds. In few cases a more stringent package type is required. 8 refs, 22 figs, 19 tabs

  13. Interim guidance on the safe transport of uranium hexafluoride

    International Nuclear Information System (INIS)

    1991-06-01

    Uranium hexafluoride (UF 6 ) is a radioactive material that has significant non-radiological hazardous properties. In conformity with international regulatory practice for dangerous goods transport, these properties are classed as ''subsidiary risks'', although they predominate in the cases of depleted and natural UF 6 . UF 6 is transported as a solid material below atmospheric pressure. The IAEA Regulations for the Safe Transport of Radioactive Material, 1985 Edition, Safety Series No. 6, make recommendations that aimed to provide an adequate level of safety against radiological and criticality hazards. The basis for these is that the stringency of package performance requirements, operational procedures and approval and administrative procedures is graded relative to the severity of the hazard. The cylinders used for transporting UF 6 are also used in the production, storage and use of the material and that the fraction of their life cycle in which transport is involved is small. Consideration must also be given to the large number of existing cylinders (estimated to be between 60,000 and 70,000). Specific recommendations provided for UF 6 transport, listed in Section II, are additional to the requirements of the Regulations. The intent of these additional recommendations is to restrict contamination and to provide protection to workers and to the general public against the chemical hazard possibly resulting from a severe accident involving the transport of UF 6 , and in addition against the consequences of explosive rupture of small bare cylinders of UF 6 . 20 refs, figs and tabs

  14. Assuring safe interim storage of Hanford high-level tank wastes

    International Nuclear Information System (INIS)

    Bacon, R.F.; Babad, H.; Lerch, R.E.

    1996-01-01

    The federal government established the Hanford Site in South-Eastern Washington near the City of Richland in 1943 to produce plutonium for national defense purposes. The Hanford Site occupies approximately 1,450 square kilometers (560 square miles) of land North of the City of Richland. The production mission ended in 1988, transforming the Hanford Site mission to waste management, environmental restoration, and waste disposal. Thus the primary site mission has shifted from production to the management and disposal of radioactive, hazardous, and mixed waste that exist at the Hanford Site. This paper describes the focus and challenges facing the Tank Waste Remediation System (TWRS) Program related to the dual and parallel missions of interim safe storage and disposal of the tank associated waste. These wastes are presently stored in 2.08E+05 liters (55,000) to 4.16E+06 liters (1,100,000) gallon low-carbon steel tanks. There are 149 single- and 28 double-shell radioactive underground storage tanks, as well as approximately 40 inactive miscellaneous underground storage tanks. In addition, the TWRS mission includes the storage and disposal of the inventory of 1,929 cesium and strontium capsules created as part of waste management efforts. Tank waste was a by-product of producing plutonium and other defense related materials. From 1944 through 1990, four (4) different major chemical processing facilities at the Hanford Site processed irradiated (spent) fuel from defense reactors to separate and recover plutonium for weapons production. As new and improved processes were developed over the last 50 years, the processing efficiency improved and the waste compositions sent to the tanks for storage changed both chemically and radiologically. The earliest separation processes (e.g., bismuth phosphate coprecipitation) carried out in T Plant (1944-1956) and B Plant (1945-1952) recovered only plutonium

  15. Conceptual design of inherently safe integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Chang, M. H.; Lee, D. J. and others

    1999-03-01

    The design concept of a 300 MWt inherently safe integral reactor(ISIR) for the propulsion of extra large and superhigh speed container ship was developed in this report. The scope and contents of this report are as follows : 1. The state of the art of the technology for ship-mounted reactor 2. Design requirements for ISIR 3. Fuel and core design 4. Conceptual design of fluid system 5. Conceptual design of reactor vessel assembly and primary components 6. Performance analyses and safety analyses. Installation of two ISIRs with total thermal power of 600MWt and efficiency of 21% is capable of generating shaft power of 126,000kW which is sufficient to power a container ship of 8,000TEU with 30knot cruise speed. Larger and speedier ship can be considered by installing 4 ISIRs. Even though the ISIR was developed for ship propulsion, it can be used also for a multi-purpose nuclear power plant for electricity generation, local heating, or seawater desalination by mounting on a movable floating barge. (author)

  16. Advanced nuclear reactor public opinion project. Interim report

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  17. Safe Management Of Fast Reactors: Towards Sustainability

    International Nuclear Information System (INIS)

    Dreimanis, Andrejs

    2015-01-01

    An interdisciplinary systemic approach to socio-technical optimization of nuclear energy management is proposed, by recognizing a) the rising requirements to nuclear safety being realized using fast reactors (FR), b) the actuality to maintain and educate qualified workforce for fast reactors, c) the reactor safety and public awareness as the keystones for improving attitude to implement novel reactors. Knowledge management and informational support firstly is needed in: 1) technical issues: a) nuclear energy safety and reliability, b) to develop safe and economic technologies; 2) societal issues: a) general nuclear awareness, b) personnel education and training, c) reliable staff renascence, public education, stakeholder involvement, e).risk management. The key methodology - the principles being capable to manage knowledge and information issues: 1) a self-organization concept, 2) the principle of the requisite variety. As a primary source of growth of internal variety is considered information and knowledge. Following questions are analyzed indicating the ways of further development: a) threats in peaceful use of nuclear energy, b) basic features of nuclear risks, including terrorism, c) human resource development: basic tasks and instruments, d) safety improvements in technologies, e) advanced research and nuclear awareness improvement There is shown: public education, social learning and the use of mass media are efficient mechanisms forming a knowledge-creating community thereby reasoning to facilitate solution of key socio-technical nuclear issues: a) public acceptance of novel nuclear objects, b) promotion of adequate risk perception, and c) elevation of nuclear safety level and adequate risk management resulting in energetic and ecological sustainability. (author)

  18. Proceedings of the workshop on intrinsically safe and economical reactors

    International Nuclear Information System (INIS)

    Livingston, R.S.

    1985-12-01

    This report presents the proceedings of a workshop concerning the design of inherently safe reactors. This paper emphasizes Japanese contributions to this subject, especially small reactors. Nine analytics were prepared for this report

  19. USING POLYMERIC HYDROGEN GETTERS TO PREVENT COMBUSTIBLE ATMOSPHERES DURING INTERIM SAFE STORAGE OF PLUTONIUM OXIDE

    International Nuclear Information System (INIS)

    Woodsmall, T

    2007-01-01

    Nuclear Materials Management (NMM) of WSRC has recently installed the capability to perform both non-destructive and destructive examination of 3013 containers of Pu oxide in accordance with DOE-STD-3013. The containers will be opened and the oxide will be sampled for analysis. The remaining bulk oxide must then be safely stored in a non-3013-compliant configuration. Available processing equipment and controls cannot prevent the oxide from adsorbing moisture during this process. Subsequent radiolysis of moisture during storage may generate combustible quantities of gases while waiting final processing, and satisfying DOE Interim Safe Storage Criteria (ISSC) would require that storage containers be vented at impractical frequencies. With support from an independent National Laboratory, WSRC/NMM has demonstrated that a commercial hydrogen getter material will effectively prevent the accumulation of combustible gas concentrations. A project overview, including storage requirements and strategies, as well as getter technology, current test results, and anticipated future developments will be addressed

  20. 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

  1. Progress and future direction for the interim safe storage and disposal of Hanford high level waste (HLW)

    International Nuclear Information System (INIS)

    Wodrich, D.D.

    1996-01-01

    This paper describes the progress made at the largest environmental cleanup program in the United States. Substantial advances in methods to start interim safe storage of Hanford Site high-level wastes, waste characterization to support both safety- and disposal-related information needs, and proceeding with cost-effective disposal by the US DOE and its Hanford Site contractors, have been realized. Challenges facing the Tank Waste Remediation System Program, which is charged with the dual and parallel missions of interim safe storage and disposal of the high-level tank waste stored at the Hanford Site, are described

  2. Fail-safe reactivity compensation method for a nuclear reactor

    Science.gov (United States)

    Nygaard, Erik T.; Angelo, Peter L.; Aase, Scott B.

    2018-01-23

    The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on the constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.

  3. A new advanced safe nuclear reactor concept

    International Nuclear Information System (INIS)

    Sefidvash, Farhang

    1999-01-01

    The reactor design is based on fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident condition. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed in order to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another is to modify the shape of the reactor core to produce a non-fluctuating bed and consequently guarantee the dynamic stability of the reactor. The mixing of Tantalum in the fuel is also proposed as an additional inhibition to power excursion. The spent fuel pellets may not be considered nuclear waste since they are in the shape and size that can easily be used as a a radioactive source for food irradiation and industrial applications. The reactor can easily operate with any desired spectrum by varying the porosity in order to be a plutonium burner or utilize a thorium fuel cycle. (author)

  4. Design strategy for control of inherently safe reactors

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1984-01-01

    Reactor power plant safety is assured through a combination of engineered barriers to radiation release (e.g., reactor containment) in combination with active reactor safety systems to shut the reactor down and remove decay heat. While not specifically identified as safety systems, the control systems responsible for continuous operation of plant subsystems are the first line of defense for mitigating radiation releases and for plant protection. Inherently safe reactors take advantage of passive system features for decay-heat removal and reactor shutdown functions normally ascribed to active reactor safety systems. The advent of these reactors may permit restructuring of the present control system design strategy. This restructuring is based on the fact that authority for protection against unlikely accidents is, as much as practical, placed upon the passive features of the system instead of the traditional placement upon the PPS. Consequently, reactor control may be simplified, allowing the reliability of control systems to be improved and more easily defended

  5. Safe operation of critical assemblies and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-09-15

    Some countries have accumulated considerable experience in the operation of these reactors and have in the process developed safe practices. On the other hand, other countries which have recently acquired, or will soon acquire, such reactors do not have sufficient background of experience with them to have developed full knowledge regarding their safe operation. In this situation, the International Atomic Energy Agency has considered that it would be useful to make available to all its Member States a set of recommendations on the safe operation of these reactors, based on the accumulated experience and best practices. The Director General accordingly nominated a Pane Ion Safe Operation of Critical Assemblies and Research Reactors to assist the Agency's Secretariat in drafting such recommendations

  6. Technical and safe development features of modern research reactor

    International Nuclear Information System (INIS)

    Wang Jiaying; Dong Duo

    1998-01-01

    The development trend of research reactor in the world, and development situation in China are introduced. Up to now, some research reactors have serviced for long time and equipment have aged, not to be satisfied for requirement of science and technology development. New research reactors must been developed. The technical features and safe features of new type research reactor in China, for example: multi-pile utilization, compact core of high flux, high automation level of control, reactor two independent shutdown systems, great coefficient of negative temperature, passive safety systems, reliable residual heat removal system are studied

  7. The inherently-safe power reactor DYONISOS (Dynamic Nuclear Inherently-Safe Reactor Operating with Spheres)

    International Nuclear Information System (INIS)

    Taube, M.; Lanfranchi, M.; Weissenfluh, Th. von; Ligou, J.; Yadigaroglu, G.; Taube, P.

    1986-01-01

    A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydro-dynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h. (author)

  8. ISAT promises fail-safe computer-based reactor protection

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    AEA Technology's ISAT system is a multiplexed microprocessor-based reactor protection system which has very extensive self-monitoring capabilities and is inherently fail safe. It provides a way of addressing software reliability problems that have tended to hamper widespread introduction of computer-based reactor protection. (author)

  9. Towards intrinsically safe light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hannerz, K

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification.

  10. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Hannerz, K.

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification

  11. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  12. Safe operation and maintenance of research reactor

    International Nuclear Information System (INIS)

    Munsorn, S.

    1999-01-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U 3 O 8 - Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  13. Application of SAFE to an operating reactor

    International Nuclear Information System (INIS)

    Chapman, L.D.

    1979-01-01

    A method for the evaluation of physical protection systems at nuclear facilities has been developed. The evaluation process consists of five major phases: (1) Facility Characterization, (2) Facility Representation, (3) Component Performance, (4) Adversary Path Analysis, and (5) Effectiveness Evaluation. Each of these phases will be described in some detail and illustrated by examples. The process for evaluation of physical protection system effectiveness against an outside threat will be presented for a reactor facility

  14. Nuclear design of ISER [intrinsically safe and economical reactor

    International Nuclear Information System (INIS)

    Yamano, Naoki; Yokoyama, Takashi

    1985-01-01

    A preliminary core design work on ISER (Intrinsically Safe and Economical Reactor) based on the concept of the PIUS reactor of ASEA-ATOM is performed in order to grasp the characteristics of the reactor core and the fuel management scheme. Certain relations between the fuel specifications and the cycle length are estimated. Items of improvement on the ISER core characteristics and problems to be considered on the nuclear design are presented. Experiments to be considered are also discussed in conjunction with the development of experimental reactor (ISER-E)

  15. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Hannerz, K.

    1983-02-01

    The reactor-safety issue is one of the principal problems threatening the future of the nuclear option, at least in participatory democracies. It has contributed to widespread public distrust and is the direct cause of the escalation in design complexity and quality assurance requirements that are rapidly eroding the competitive advantage of nuclear power. Redesign of the light-water reactor can eliminate those features that leave it open to public distrust and obstructive intervention. This redesign appears feasible within the realm of proven technology in those fields (fuels, materials, water chemistry, waste technology, etc.) in which extended operating experience is essential for confidence in system performance. A pressurized water reactor outline design developed to achieve the above goal is presented. The key feature is the design of the primary system extracting heat from the core so that the latter is protected from damage caused by any credible system failure or any destructive intervention from the outside by either violent means (up to and including nonnuclear warfare) or by mistaken or malicious use of the plant control systems. Such a design objective can be achieved by placing the entire primary circulation system in a large pressurized pool of cold water with a high boric acid content. Enough water is provided in the pool to allow core-decay-heat removal by evaporation for at least one week following any incident with no cooling systems operating. Subsequently it is assumed that a supply of further water (a few cubic meters per hour) from the outside can be arranged, even without the presence of the plant operating personnel

  16. Physical protection of shipments of irradiated reactor fuel; Interim guidance. Regulatory report

    International Nuclear Information System (INIS)

    1980-06-01

    During May, 1979, the U.S. Nuclear Regulatory Commission approved for issuance in effective form new interim regulations for strengthening the protection of spent fuel shipments against sabotage and diversion. The new regulations were issued without benefit of public comment, but comments from the public were solicited after the effective date. Based upon the public comments received, the interim regulations were amended and reissued in effective form as a final interim rule in May, 1980. The present document supersedes a previously issued interim guidance document, NUREG-0561 (June, 1979) which accompanied the original rule. This report has been revised to conform to the new interim regulations on the physical protection of shipments of irradiated reactor fuel which are likely to remain in effect until the completion of an ongoing research program concerning the response of spent fuel to certain forms of sabotage, at which time the regulations may be rescinded, modified or made permanent, as appropriate. This report discusses the amended regulations and provides a basis on which licensees can develop an acceptable interim program for the protection of spent fuel shipments

  17. Safe interim storage of Hanford tank wastes, draft environmental impact statement, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1994-07-01

    This Draft EIS is prepared pursuant to the National Environmental Policy Act (NEPA) and the Washington State Environmental Policy Act (SEPA). DOE and Ecology have identified the need to resolve near-term tank safety issues associated with Watchlist tanks as identified pursuant to Public Law (P.L.) 101-510, Section 3137, ''Safety Measures for Waste Tanks at Hanford Nuclear Reservation,'' of the National Defense Authorization Act for Fiscal Year 1991, while continuing to provide safe storage for other Hanford wastes. This would be an interim action pending other actions that could be taken to convert waste to a more stable form based on decisions resulting from the Tank Waste Remediation System (TWRS) EIS. The purpose for this action is to resolve safety issues concerning the generation of unacceptable levels of hydrogen in two Watchlist tanks, 101-SY and 103-SY. Retrieving waste in dilute form from Tanks 101-SY and 103-SY, hydrogen-generating Watchlist double shell tanks (DSTs) in the 200 West Area, and storage in new tanks is the preferred alternative for resolution of the hydrogen safety issues

  18. Safe interim storage of Hanford tank wastes, draft environmental impact statement, Hanford Site, Richland, Washington

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    This Draft EIS is prepared pursuant to the National Environmental Policy Act (NEPA) and the Washington State Environmental Policy Act (SEPA). DOE and Ecology have identified the need to resolve near-term tank safety issues associated with Watchlist tanks as identified pursuant to Public Law (P.L.) 101-510, Section 3137, ``Safety Measures for Waste Tanks at Hanford Nuclear Reservation,`` of the National Defense Authorization Act for Fiscal Year 1991, while continuing to provide safe storage for other Hanford wastes. This would be an interim action pending other actions that could be taken to convert waste to a more stable form based on decisions resulting from the Tank Waste Remediation System (TWRS) EIS. The purpose for this action is to resolve safety issues concerning the generation of unacceptable levels of hydrogen in two Watchlist tanks, 101-SY and 103-SY. Retrieving waste in dilute form from Tanks 101-SY and 103-SY, hydrogen-generating Watchlist double shell tanks (DSTs) in the 200 West Area, and storage in new tanks is the preferred alternative for resolution of the hydrogen safety issues.

  19. Final Hazard Classification and Auditable Safety Analysis for the 105-F Building Interim Safe Storage Project

    International Nuclear Information System (INIS)

    Rodovsky, T.J.; Bond, S.L.

    1998-07-01

    The auditable safety analysis (ASA) documents the authorization basis for the partial decommissioning and facility modifications to place the 105-F Building into interim safe storage (ISS). Placement into the ISS is consistent with the preferred alternative identified in the Record of Decision (58 FR). Modifications will reduce the potential for release and worker exposure to hazardous and radioactive materials, as well as lower surveillance and maintenance (S ampersand M) costs. This analysis includes the following: A description of the activities to be performed in the course of the 105-F Building ISS Project. An assessment of the inventory of radioactive and other hazardous materials within the 105-F Building. Identification of the hazards associated with the activities of the 105-F Building ISS Project. Identification of internally and externally initiated accident scenarios with the potential to produce significant local or offsite consequences during the 105-F Building ISS Project. Bounding evaluation of the consequences of the potentially significant accident scenarios. Hazard classification based on the bounding consequence evaluation. Associated safety function and controls, including commitments. Radiological and other employee safety and health considerations

  20. International collaboration on inherently safe nuclear reactors

    International Nuclear Information System (INIS)

    Barkenbus, J.N.

    1989-01-01

    Science and technology transcend economic and political ideologies, providing a means of communications and approach common to both the United States and the Soviet Union. This paper suggests that the field of nuclear fission is a logical and productive area for superpower and broader collaboration, but that the kind of collaboration characteristic of past and present activity is less than it optimally could be. The case for cost sharing is compelling with budget constraints and mounting concerns over global warming. The case for collaboration is based on economic, psychological, and political grounds. A collaborative effort in nuclear fission is presented as a near term effort by building and testing of a prototype reactor in the 1990s

  1. A conceptual design of intrinsically safe and economical reactor (ISER)

    International Nuclear Information System (INIS)

    Oda, Junro

    1985-01-01

    The purpose of this paper is to describe the reference conceptual designs of the ISER which were prepared for the ISER development forum in Japan. At the forum, participants from influential utilities, academia, as well as companies in the nuclear industry, discussed the development of the inherently safe reactor over the last two years. The conceptual designs described in this paper are preliminary trial designs at an early stage and essentially versions of the PIUS reactor developed by ASEA-ATOM. A notable feature of the ISER which is different from the original PIUS is its use of a steel reactor pressure vessel for reducing plant construction costs and improving plant performance

  2. The application of probabilistic risk assessment to inherently safe reactors

    International Nuclear Information System (INIS)

    Cave, L.; Kastenberg, W.E.

    1987-01-01

    In the development of safety goals and design criteria for 'inherently safe' reactors a question which arises is 'To what extent is PRA relevant.' To answer this question it is necessary to consider both the risk to the public and the investment risk to the utility. In this paper the factors which are likely to determine safety objectives and their allocation are presented. (orig.)

  3. Pioneering SUPER - Small Unit Passively-safe Enclosed Reactor - 15559

    International Nuclear Information System (INIS)

    Bhownik, P.K.; Gairola, A.; Shamim, J.A.; Suh, K.Y.; Suh, K.S.

    2015-01-01

    This paper presents the basic features of the Small Unit Passively-safe Enclosed Reactor abbreviated as SUPER, a new reactor system that has been designed and proposed at the Seoul National University's Department of Energy Systems Engineering. SUPER is a small modular reactor system or SMR that is cooled by sub-cooled as well as supercritical water. As a new member of SMRs, SUPER is a small-scale nuclear plant that is designed to be factory-manufactured and shipped as modules to be assembled at a site. The concept offers promising answers to many questions about nuclear power including proliferation resistance, waste management, safety, and startup costs. SUPER is a customized paradigm of a supercritical water reactor or SCWR, a type sharing commonalities with the current fleet of light water reactors, or LWRs. SUPER has evolved from the System-integrated Modular Advance Reactor, or SMART, being developed at the Korea Atomic Energy Research Institute, or KAERI. SUPER enhanced the safety features for robustness, design/equipment simplification for natural convection, multi-purpose application for co-generation flexibilities, suitable for isolated or small electrical grids, just-in-time capacity addition, short construction time, and last, but not least, lower capital cost per unit. The primary objectives of SUPER is to develop the conceptual design for a safe and economic small, natural circulation SCWR, to address the economic and safety attributes of the concept, and to demonstrate its technical feasibilities. (authors)

  4. Operator licensing examination standards for power reactors. Interim revision 8

    International Nuclear Information System (INIS)

    1997-01-01

    These examination standards are intended to assist NRC examiners and facility licensees to better understand the processes associated with initial and requalification examinations. The standards also ensure the equitable and consistent administration of examinations for all applicants. These standards are for guidance purposes and are not a substitute for the operator licensing regulations (i.e., 10 CFR Part 55), and they are subject to revision or other changes in internal operator licensing policy. This interim revision permits facility licensees to prepare their initial operator licensing examinations on a voluntary basis pending an amendment to 10 CFR Part 55 that will require facility participation. The NRC intends to solicit comments on this revision during the rulemaking process and to issue a final Revision 8 in conjunction with the final rule

  5. Interim waste storage for the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

    1991-01-01

    The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes that are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig

  6. New safe reactor, but not first of a kind engineering

    International Nuclear Information System (INIS)

    Nurdin, M.

    1997-01-01

    New safe reactor but not a first of a kind engineering is a new reactor concept to fulfill the need on Small Reactor for power generation, both for electricity and for co-generation. Nuclear reactor system of this concept in certain degree has similar design compared to the established and successful reactor systems now in operation; so the material used for the same function and purpose is not the same. The strategy or choice adopted in achieving this concept will be automatically shown by the inspiration or philosophy of ''not to re-invent the wheel.'' Based on the above mentioned strategy, a certain degree of experimental verification and justification are of course needed/necessary to know better the deviations and the differences from the existing nuclear reactor concepts and further to anticipate of course precisely engineering behaviour of the proposed concept. Physical and engineering discussion on the proposed concept are main objectives of this paper in which most of the scope and objectives of this IAEA TCM on Small Reactors with minimized Staffing and/or Remote Monitoring are elaborated. They are discussed in such a way to give the technical and economical background of the proposed concept. (author)

  7. Major outage trends in light water reactors. Interim report

    International Nuclear Information System (INIS)

    Burns, E.T.

    1978-04-01

    The report is a summary of the major outages which occurred in light water reactor plants during the period January 1971 through June 1977. Only those outages greater than 100 hours duration (exclusive of refueling outages) are included in the report. The trends in outages related to various reactor systems and components are presented as a function of plant age, and alternatively, calendar year. The principal contributors to major outages are ranked by their effect on the overall outage time for PWRs and BWRs. In addition, the outage history of each operating nuclear plant greater than 150 MWe is presented, along with a brief summary of those outages greater than two months duration

  8. Fast-mixed spectrum reactor interim report initial feasibility study

    International Nuclear Information System (INIS)

    Fischer, G.J.; Cerbone, R.J.

    1979-01-01

    The report summarizes the results of an initial four-month feasibility study of the Fast-Mixed Spectrum Reactor (FMSR). Reactor physics, fuel cycle, and thermal-hydraulic analyses were performed on a reference design. These results when coupled to a fuel and materials evaluation performed in cooperation with the Argonne National Laboratory indicate that the FMSR is feasible provided the fuels, cladding, and subassembly ducts can survive a peak fuel burnup of 15 to 20 atom percent heavy metal and peak fluences of 8 x 10 23 (nvt > 0.1 MeV). The results of this short study have also provided a basis for exploring alternative designs requiring significantly lower peak burnup and fluences for their operation

  9. Interim Report on Fluid-Fuel Thermal Breeder Reactors (Revised)

    International Nuclear Information System (INIS)

    MacPherson, H. G.; Alexander, L. G.; Carter, W. L.; Chapman, R. H.; Kinyon, B. W.; Miller, J. W.

    1960-01-01

    The merits of aqueous-homogeneous ), graphite-moderated molten salt (MSBR) , and graphite-moderated liquid-bismuth (LBBR) breeder reactors operated at nearly comparable fuel-cycle costs (~1.5 mills/kwhr) were evaluated. The net electrical plant capability was assumed to be 1000 MwE, and the fuel and fertile streams were processed continuously on-site. The specific powers based on fuel were 1.2, 1.2, and 0.5 MwE/kg respectively, and 5.9, 3.7, and 5.3 MwE/tonne based on thorium. Net breeding ratios were 1.10, 1.07, and 1.07, giving doubling times of 5-1/2, 11, and 25 full power years . The fuel-cycle costs at the design points selected were 1.4, 1.3, and 1.6 mills/kwhr . The AHBR has an advantage in breeding ratio and doubling time because D 2 O is superior to graphite as a moderator in breeder reactors. MSBR has an advantage in fuel-cycle costs and in inventory of uranium in the fertile stream as a result of using a solution blanket.

  10. Submersion-Subcritical Safe Space (S4) reactor

    International Nuclear Information System (INIS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The Submersion-Subcritical Safe Space (S 4 ) reactor, developed for future space power applications and avoidance of single point failures, is presented. The S 4 reactor has a Mo-14% Re solid core, loaded with uranium nitride fuel, cooled by He-30% Xe and sized to provide 550 kWth for 7 years of equivalent full power operation. The beryllium oxide reflector of the S 4 reactor is designed to completely disassemble upon impact on water or soil. The potential of using Spectral Shift Absorber (SSA) materials in different forms to ensure that the reactor remains subcritical in the worst-case submersion accident is investigated. Nine potential SSAs are considered in terms of their effect on the thickness of the radial reflector and on the combined mass of the reactor and the radiation shadow shield. The SSA materials are incorporated as a thin (0.1 mm) coating on the outside surface of the reactor core and as core additions in three possible forms: 2.0 mm diameter pins in the interstices of the core block, 0.25 mm thick sleeves around the fuel stacks and/or additions to the uranium nitride fuel. Results show that with a boron carbide coating and 0.25 mm iridium sleeves around the fuel stacks the S 4 reactor has a reflector outer diameter of 43.5 cm with a combined reactor and shadow shield mass of 935.1 kg. The S 4 reactor with 12.5 at.% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide interstitial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating has a slightly smaller reflector outer diameter of 43.0 cm, resulting in a smaller total reactor and shield mass of 901.7 kg. With 8.0 at.% europium-151 added to the fuel, along with europium-151 sesquioxide for the pins and coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

  11. Technological aspects of intrinsically safe and economical reactor (ISER)

    International Nuclear Information System (INIS)

    Yamada, Nobuyuki; Oda, Junro; Yamanaka, Kazuo; Sugawara, Ichiro.

    1987-01-01

    ISER is a modified version of process inherent ultimate safe reactor (PIUS) developed by ASEA-ATOM, Sweden, and follows the same inherent safety principle, that is, passive reactor shutdown through the introduction of borated pool water into a core via an interface, and passive decay heat removal by natural circulation. The most significant deviation from the PIUS is that the ISER employs a steel reactor pressure vessel enclosed in the reactor pit, instead of a prestressed concrete reactor pressure vessel of the PIUS. The merits of using steel pressure vessels are siting versatility including barge-mounted plants, low cost, the standardization and serial production of total NSSSs through the weight reduction and compaction of primary system, as well as the possibility of utilizing current LWR technology, which minimizes R and D effort. In this paper, the design features of the latest version of ISERs are shown, and the specific problems of the key components are discussed. The primary system consists of a primary coolant loop and a borated water pool, which are connected with upper and lower interfaces. The nuclear design and thermohydraulic design, the operation and maintenance, and the design features of a steam generator, a pressurizer, interfaces and so on are described. (Kako, I.)

  12. 76 FR 17160 - Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

    Science.gov (United States)

    2011-03-28

    ... design certification (DC) application for new nuclear power reactors under Title 10 of the Code of... NUCLEAR REGULATORY COMMISSION [NRC-2010-0033; DC/COL-ISG-021] Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby...

  13. A logic scheme for regulating safe operation of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Effat, A; Rahman, F A [Operational Saety Dept, National Center of Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    This investigation presents a logic scheme for regulating the safe operation of research reactor in accordance with the new revision of SS-35 and revised by the 10 CFR. It emphasizes the regulatory inspection and enforcement (RI end E) during the reactor operation phase. IT is developed to provide information, guidance and recommendations to be taken when constructing the RI and E program that could be applied to the operational phase of the egyptian Research Reactors. In the operational phase, the regulatory inspection (RI) means an examination, observation, measurement, or test undertaken or on behalf of the nuclear regulatory body (NRB) during operation to verify that the nuclear materials, components, systems and structures as well operational activities, processes, procedures and personnel competence and performance are in accordance with the requirements established or the provisions approved by NRB or specified in the operational license or contained in regulations. Regulatory inspection includes both routine and non-routine ones. Any of them may be announced or unannounced. The problems identified by the RI must be resolved by the proper RE actions. The RE actions include investigative and corrective RE actions. These RI and E procedures for regulating safe operation of research reactors are presented as flow charts and then developed as a computer logic scheme. The software program is very efficient, very friendly, very simple and is interactive in nature such that the program asks the user certain questions about essential steps that guide the (RI and E) for research reactors, and user responds. The program proceeds based on this response until all the necessary steps for (RI and E) are accomplished. 5 figs.

  14. The accident at the Harrisburg nuclear reactor - Interim conclusions

    International Nuclear Information System (INIS)

    Yiftah, S.

    1979-07-01

    This work describes the first minutes, first day and first week following the Three Mile Island accident. It shows the failures that occurred and the lessons which should be derived. It is pointed out that the doses of radiation that escaped from the TMI plant were at no time large enough to have had any effect on the 2 million people living on a radius of 80 km from the plant. Although no casualties occurred the Harrisburg accident will create an impulse for a new study and understanding of the nuclear plant safety and might serve as a live safety laboratory. After the TMI accident nuclear plants are already safer, one of the conclusions being that a new planning of the operation room is required, with the operators acquiring a better understanding of what is going on during a nuclear reactor accident. (B.G.)

  15. Gas-cooled reactor commercialization study. Interim report

    International Nuclear Information System (INIS)

    1977-01-01

    This report of the gas-cooled reactor commercialization study completes the technical and cost evaluation portions of this study contract. A final report in December will update the status of the incentive analyses and the issues of commercialization. This study was designed to bring together potential industry participants (utilities and suppliers) to evaluate the commercial potential of the HTGR-SC and to build channels of communication among the participating organizations at the same time that technical, economic and institutional issues were being evaluated. RAMCO, Inc., in suggesting and using this study approach, believes its application extends to any commercialization problem involving multi-party involvement in high capital, intensive, high risk energy technologies

  16. Safe decommissioning of the Romanian VVR-S research reactor

    International Nuclear Information System (INIS)

    Garlea, C.; Garlea, I.; Kelerman, C.; Rodna, A.

    2002-01-01

    The VVR-S Romania research reactor was operated between 1957-1997, at 2 MW nominal power, for research and radioisotopical production. The detailed decommissioning plan was developed between 1995-1998, in the frame of the International Atomic Energy Agency Technical assistance project ROM/9/017. The proposed strategy agreed by the counterpart as well as international experts was stage 1. In 1997, an independent analysis performed by European Commission experts, in the frame of PHARE project PH04.1/1994 was dedicated to the 'Study of Soviet Design Research Reactors', had consolidated the development of the project emphasizing technical options of safe management for radioactive wastes and VVR-S spent fuel. The paper presents the main technical aspects as well as those of social impact, which lead to the establishment of strategy for safe management of decommissioning. Technical analysis of the VVR-S reactor and associated radwaste facilities (Radioactive Waste Treatment Plant - Magurele and National Repository Baita-Bihor) proved the possibility of the classical method utilization for dismantling of the facility and treatment-conditioning-disposal of the arrised wastes in safe conditions. The decommissioning plan at stage 2 has been developed based on radiological safety assessment, evaluation of radwaste inventory (removed as well as preserved on site), cost analysis and environmental impact. Technical data were provided by the R and D programme including neutron calculations and experiments, radiological characterizing (for facility and its influence area), seismic analysis and environmental balance during the operation and after shut down of the reactor. A special chapter is dedicated to regulatory issues concerning the development of decommissioning under nuclear safety. Based on the Fundamental Norms of Radiological Safety, the Regulatory Body defined the clearance levels and safety criteria for the process. The development of National Norms for the

  17. Inherently safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Yamada, Masao; Hayakawa, Hitoshi

    1987-01-01

    It is recognized in general that High Temperature Gas-cooled Reactors have remarkable characteristics in inherent safety and it is well known that credits of the time margin have been admitted for accident evaluation in the licensing of the currently operating prototype HTGRs (300 MWe class). Recently, more inherently safe HTGRs are being developed in various countries and drawing attention on their possibility for urban siting. The inherent safety characteristics of these HTRs differ each other depending on their design philosophy and on the features of the components/structures which constitute the plant. At first, the specific features/characteristics of the elemental components/structures of the HTRs are explained one by one and then the overall safety features/characteristics of these HTR plants are explained in connection with their design philosophy and combination of the elemental features. Taking the KWU/Interatom Modular Reactor System as an example, the particular design philosophy and safety characteristics of the inherently safe HTR are explained with a result of preliminary evaluation on the possibility of siting close to densely populated area. (author)

  18. Safe Operation of Critical Assemblies and Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-05-15

    This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

  19. Safe Operation of Critical Assemblies and Research Reactors

    International Nuclear Information System (INIS)

    1961-01-01

    This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

  20. Safe design and operation of fluidized-bed reactors: Choice between reactor models

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1990-01-01

    For three different catalytic fluidized bed reactor models, two models presented by Werther and a model presented by van Deemter, the region of safe and unique operation for a chosen reaction system was investigated. Three reaction systems were used: the oxidation of benzene to maleic anhydride, the

  1. Safety aspect of long-life small safe power reactors

    International Nuclear Information System (INIS)

    Zaki, S.; Sekimoto, H.

    1995-01-01

    Safety aspects of several design options of long-life small safe fast power reactors using nitride fuel and lead-bismuth as coolant are discussed. In the present study hypothetical accidents are simulated for these reactors, i.e., unprotected simultaneous ULOF (total loss of primary pumping system) and UTOP (rod run out transient over power) accidents, caused by the simultaneous withdrawal of all control rods. The proposed designs have some important safety characteristics as low reactivity swing (only 0.2-0.25$), and negative coolant void coefficient over whole burnup period. Effectively negative value of all components of reactivity during an accident is observed. The safety performances of the balance, pancake, and tall slender type of core, each of them satisfy reactivity and negative coolant void coefficient constraint, against the above accident are compared. The simulation results show that all of the design options can survive the above accidents without the help of reactor scram and without the need of operator actions. (author)

  2. Insights from the interim reliability evaluation program pertinent to reactor safety issues

    International Nuclear Information System (INIS)

    Carlson, D.D.

    1983-01-01

    The Interim Reliability Evaluation Program (IREP) consisted of concurrent probabilistic analyses of four operating nuclear power plants. This paper presents and integrated view of the results of the analyses drawing insights pertinent to reactor safety. The importance to risk of accident sequences initiated by transients and small loss-of-coolant accidents was confirmed. Support systems were found to contribute significantly to the sets of dominant accident sequences, either due to single failures which could disable one or more mitigating systems or due to their initiating plant transients. Human errors in response to accidents also were important risk contributors. Consideration of operator recovery actions influences accident sequence frequency estimates, the list of accident sequences dominating core melt, and the set of dominant risk contributors. Accidents involving station blackout, reactor coolant pump seal leaks and ruptures, and loss-of-coolant accidents requiring manual initiation of coolant injection were found to be risk significant

  3. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    International Nuclear Information System (INIS)

    Faghihi, F.; Mirvakili, S.M.

    2009-01-01

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  4. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

    2009-06-15

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  5. Concepts for the interim storage of spent fuel elements from research reactors in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Niephaus, D.; Bensch, D.; Quaassdorff, P.; Plaetzer, S.

    1997-01-01

    Research reactors have been operated in the Federal Republic of Germany since the late fifties. These are Material Test Reactors (MTR) and training, Research and Isotope Facilities of General Atomic (TRIGA). A total of seven research reactors, i.e. three TRIGA and four MTR facilities were still in operation at the beginning of 1996. Provisions to apply to the back-end of the fuel cycle are required for their continued operation and for already decommissioned plants. This was ensured until the end of the eighties by the reprocessing of spent fuel elements abroad. In view of impeding uncertainties in connection with waste management through reprocessing abroad, the development of a national back-end fuel cycle concept was commissioned by the Federal Minister of Education, Science, Research and Technology in early 1990. Development work was oriented along the lines of the disposal concept for irradiated light-water reactor fuel elements from nuclear power plants. Analogously, the fuel elements from research reactors are to be interim-stored on a long-term basis in adequately designed transport and storage casks and then be directly finally disposed without reprocessing after up to forty years of interim storage. As a first step in the development of a concept for interim storage, several sites with nuclear infrastructure were examined and assessed with respect to their suitability for interim storage. A reasonably feasible reference concept for storing the research reactor fuel elements in CASTOR MTR 2 transport and storage casks at the Ahaus interim storage facility (BZA) was evaluated and the hot cell facility and AVR store of Forschungszentrum Juelich (KFA) were proposed as an optional contingency concept for casks that cannot be repaired at Ahaus. Development work was continued with detailed studies on these two conceptual variants and the results are presented in this paper. (author)

  6. SIR (Safe Integral Reactor) - reducing size can reduce cost

    International Nuclear Information System (INIS)

    Hayns, M.R.

    1991-01-01

    Traditional engineering economics have favoured the advantages of larger size as a means of reducing specific capital costs and hence unit generating costs. For large and small plants utilising the same concept, e.g. a small four-loop PWR vs a large four-loop PWR with the same number of components, economies of scale are well established. If, however, a smaller plant is sized to take advantage of features which are only feasible at smaller outputs, is of simpler design, with the advantage taken of the simplified design to produce the most cost-effective layout, and incorporates fewer, more easily replaceable components with minimal assembly on site, it is possible to produce a plant which is competitive with larger plant of more traditional design. When 'system' effects, such as better matching of installed capacity to the growth in demand and the fact that a smaller total capacity will be needed to meet a given demand with a specified level of confidence, are taken into account, it can be shown that a utility's overall cash-flow position can be improved with lower associated absolute financial risks. The UK/US Safe Integral Reactor (SIR) is an integral pressurized water reactor in the 300-400 MW(e) range which utilises conventional water reactor technology in a way not feasible at the very large, sizes of recent years. The SIR concept is briefly explained and its technical and economic advantages in terms of simplicity, construction, maintenance, availability, decommissioning, safety and siting described. The results of system analyses which demonstrate the overall financial advantages to a utility are presented. (author)

  7. Power control of SAFE reactor using fuzzy logic

    International Nuclear Information System (INIS)

    Irvine, Claude

    2002-01-01

    Controlling the 100 kW SAFE (Safe Affordable Fission Engine) reactor consists of design and implementation of a fuzzy logic process control system to regulate dynamic variables related to nuclear system power. The first phase of development concentrates primarily on system power startup and regulation, maintaining core temperature equilibrium, and power profile matching. This paper discusses the experimental work performed in those areas. Nuclear core power from the fuel elements is simulated using resistive heating elements while heat rejection is processed by a series of heat pipes. Both axial and radial nuclear power distributions are determined from neuronic modeling codes. The axial temperature profile of the simulated core is matched to the nuclear power profile by varying the resistance of the heating elements. The SAFE model establishes radial temperature profile equivalence by establishing 32 control zones as the nodal coordinates. Control features also allow for slow warm up, since complete shutoff can occur in the heat pipes if heat-source temperatures drop/rise below a certain minimum value, depending on the specific fluid and gas combination in the heat pipe. The entire system is expected to be self-adaptive, i.e., capable of responding to long-range changes in the space environment. Particular attention in the development of the fuzzy logic algorithm shall ensure that the system process remains at set point, virtually eliminating overshoot on start-up and during in-process disturbances. The controller design will withstand harsh environments and applications where it might come in contact with water, corrosive chemicals, radiation fields, etc

  8. 75 FR 8412 - Office of New Reactors: Interim Staff Guidance on Assessing Ground Water Flow and Transport of...

    Science.gov (United States)

    2010-02-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2010-0047] Office of New Reactors: Interim Staff Guidance on Assessing Ground Water Flow and Transport of Accidental Radionuclide Releases; Solicitation of Public... ground water flow and transport of accidental radionuclide releases necessary to demonstrate compliance...

  9. Interim report on construction of data base for atomic energy science documents (concerning Kyoto University Reactor)

    International Nuclear Information System (INIS)

    Takeuchi, Takayuki

    1984-01-01

    The Kyoto University Research Reactor Institute was established in 1963 as a research institute for all universities in Japan utilizing the facilities in common. The construction of a document data base has been undertaken in commemoration of the 20th anniversary of the institute. The data base concerns the research works performed at the institute and also the publications and reports on the research made by the personnel belonging to the institute. Input data are gathered from concerned researchers. In this interim report, the structure and contents of this data base are shortly described. One of the features of this data base is that it handles data with both Japanese and English at the same time. (Aoki, K.)

  10. Dynamic operator actions analysis for inherently safe fast reactors and light water reactors

    International Nuclear Information System (INIS)

    Ho, V.; Apostolakis, G.

    1988-01-01

    A comparative dynamic human actions analysis of inherently safe fast reactors (ISFRs) and light water reactors (LWRs) in terms of systems response and estimated human error rates is presented. Brief overviews of the ISFR and LWR systems are given to illustrate the design differences. Key operator actions required by the ISFR reactor shutdown and decay heat removal systems are identified and are compared with those of the LWR. It is observed that, because of the passive nature of the ISFR safety-related systems, a large time window is available for operator actions during transient events. Furthermore, these actions are fewer in number, are less complex, and have lower error rates and less severe consequences than those of the LWRs. We expect the ISFR operator errors' contribution to risk is smaller (at least in the context of the existing human reliability models) than that of the LWRs. (author)

  11. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  12. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  13. Summary of advanced LMR [Liquid Metal Reactor] evaluations: PRISM [Power Reactor Inherently Safe Module] and SAFR [Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G.

    1989-10-01

    In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) [Berglund, 1987] and the Sodium Advanced Fast Reactor (SAFR) [Baumeister, 1987], were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the ''inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II [NED, 1986]. The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs

  14. Removal Action Work Plan for 105-DR and 105-F Building Interim Safe Storage Projects and Ancillary Buildings

    International Nuclear Information System (INIS)

    Morton, M.R.

    2000-01-01

    This document contains the removal action work plan for the 105-DR and 105-F Reactor buildings and ancillary facilities. These buildings and facilities are located in the 100-D/DR and 100-F Areas of the Hanford Site, which is owned and operated by the US Department of Energy (DOE), in Benton County, Washington. The 100 Areas (including the 100-D/DR and 100-F Areas) of the Hanford Site were placed on the US Environmental Protection Agency's (EPA's) National Priorities List under the ''Comprehensive Environmental Response, Compensation, and Liability Act of 1980'' (CERCLA). The DOE has determined that hazardous substances in the 105-DR and 105-F Reactor buildings and four ancillary facilities present a potential threat to human health or the environment. The DOE has also determined that a non-time critical removal action is warranted at these facilities. Alternatives for conducting a non-time critical removal action were evaluated in the ''Engineering Evaluation/Cost Analysis for the 105-DR and 105-F Reactor Facilities and Ancillary Facilities'' (DOE-RL 1998a). The engineering evaluation/cost analysis (EE/CA) resulted in the recommendation to decontaminate and demolish the contaminated reactor buildings (except for the reactor blocks) and the ancillary facilities and to construct a safe storage enclosure (SSE) over the reactor blocks. The recommendation was approved in an action memorandum (Ecology et al. 1998) signed by the Washington State Department of Ecology (Ecology), EPA, and DOE. The DOE is the agency responsible for implementing the removal actions in the 105-D/DR and 105-F Areas. Ecology is the lead regulatory agency for facilities in the 100-D/DR Area, and EPA is the lead regulatory agency for facilities in the 100-F Area. The term ''lead regulator agency'' hereinafter, refers to these authorities. This removal action work plan supports implementation of the non-time critical removal action

  15. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    International Nuclear Information System (INIS)

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

  16. Gas reactor international coope--ative program. Interim report: assessment of gas-cooled reactor economics

    Energy Technology Data Exchange (ETDEWEB)

    1979-08-01

    A computer analysis of domestic economic incentive is presented. Included are the sample computer data set for ten combinations of reprocessing and reactor assumptions; basic data set and computer output; higher uranium availability computer output; 50 percent higher GCR fabrication cost computer output; 50 percent higher GCR reprocessing cost computer output; year 1990 and year 2000 GCR introduction scenario computer outputs; 75 percent perceived capacity factor for PBR computer output; and capital cost of GCRs 1.2 times that of LWRs.

  17. Development status of PIUS/ISER - a inherently safe reactor for the international use

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1987-01-01

    It is just in early 1980s that LWR-based nuclear power has become a substantial power source. Though the safety level of nuclear power is always claimed to be sufficiently high by the industry, it rests on the idea of defense in depth, the calculation by probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA). The TMI-2 and Chernobyl-4 accidents occurred in the industrially most advanced countries. In this paper, an alternative way to safe nuclear power is sought in so-called inherently safe reactors (ISR) including the LWR type PIUS/ISER. With proper consideration into the design of nuclear reactor plants, those can be made basically safe through the use of passive safe mechanism for their design. In short, an ISR is a nuclear power reactor which has passive and intrinsic core cooling capability and automatic shutdown capability. As the nuclear power reactors which are currently claimed to be inherently safe, there are the process inherent and ultimately safe reactor (PIUS) of ASEA-ATOM Sweden and the inherently safe and economical reactor (ISER) of the University of Tokyo, Japan, of LWR type. The current status of the development, the reliability, and some technical problems of ISER/PIUS and the attitude of various countries toward ISER/PIUS are described. (Kako, I.)

  18. RETU. The Finnish research programme on reactor safety. Interim report 1995 - May 1997

    International Nuclear Information System (INIS)

    Vanttola, T.; Puska, E.K.

    1997-08-01

    The Finnish national research programme on Reactor Safety (RETU, 1995-1998) concentrates on the search of safe limits of nuclear fuel and the reactor core, accident management methods and risk management of the operation of nuclear power plants. The annual volume of the programme has been about 26 person years and the annual funding FIM 15 million. This report summarises the structure and objectives of the programme, research fields included and the main results obtained during the period 1995 - May 1997. In the field of operational margins of a nuclear reactor, the behaviour of high burnup nuclear fuel is studied both in normal operation and during power transients. The static and dynamic reactor analysis codes are developed and validated to cope with new fuel designs and complicated three-dimensional reactivity transients and accidents. Research on accident management aims at development and validation of calculation methods needed to plan preventive measures and to train the personnel to severe accident mitigation. Other goals are to reduce uncertainties in phenomena important in severe accidents and to study actions planned for accident management. In the field of risk management probabilistic methods are developed for safety related decision making and for complex phenomena and event sequences. Effects of maintenance on nuclear power plant safety are studied and more effective methods for the assessment of human reliability and safety critical organisations are searched

  19. RETU. The Finnish research programme on reactor safety. Interim report 1995 - May 1997

    Energy Technology Data Exchange (ETDEWEB)

    Vanttola, T; Puska, E K [VTT Energy, Espoo (Finland). Nuclear Energy; eds.

    1997-08-01

    The Finnish national research programme on Reactor Safety (RETU, 1995-1998) concentrates on the search of safe limits of nuclear fuel and the reactor core, accident management methods and risk management of the operation of nuclear power plants. The annual volume of the programme has been about 26 person years and the annual funding FIM 15 million. This report summarises the structure and objectives of the programme, research fields included and the main results obtained during the period 1995 - May 1997. In the field of operational margins of a nuclear reactor, the behaviour of high burnup nuclear fuel is studied both in normal operation and during power transients. The static and dynamic reactor analysis codes are developed and validated to cope with new fuel designs and complicated three-dimensional reactivity transients and accidents. Research on accident management aims at development and validation of calculation methods needed to plan preventive measures and to train the personnel to severe accident mitigation. Other goals are to reduce uncertainties in phenomena important in severe accidents and to study actions planned for accident management. In the field of risk management probabilistic methods are developed for safety related decision making and for complex phenomena and event sequences. Effects of maintenance on nuclear power plant safety are studied and more effective methods for the assessment of human reliability and safety critical organisations are searched. 135 refs.

  20. Technology which led to the westinghouse inherently safe liquid metal reactor

    International Nuclear Information System (INIS)

    Schmidt, J.E.; Coffield, R.D.; Doncals, R.A.; Kalinowski, J.E.; Markley, R.A.

    1985-01-01

    The Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor programs resulted in an understanding of liquid metal reactor behavior that is being used to design inherent safety capability into liquid metal reactors. Technological advances give the same beneficial operating characteristics of conventional liquid metal reactors, however, the addition of inherently safe design features precludes the initiation of hypothetical core disruptive accidents. These innovative features permit inherent safety capability to be demonstrated with more than adequate margins. Also, the variety of inherent safety features provides the designers with options in selecting inherent design features for a specific reactor application

  1. The incorporation of boron in fissile transport packages for the transport and interim storage of irradiated light water reactor fuels

    International Nuclear Information System (INIS)

    Hunter, I.J.

    1998-01-01

    Boron is widely used in the nuclear industry for capturing neutrons and it is particularly useful in the criticality control of packages for the transport and interim storage of irradiated light water reactor fuels. Such fuels are typically located in an internal frame of stainless steel or aluminium, referred to as a basket, which locates the fuel assemblies in channels. Transnucleaire has designed and supplied more than 100 baskets of varying types during the past 30 years. Boron has been incorporated in many forms. Early designs of baskets used boron in specific zones to contribute to the control of criticality. Later developments in new materials dispersed boron throughout the basket and gave designers more options for the basic forms which make up the channels. New basket concepts have been developed by Transnucleaire to meet the changing market needs for transport and interim storage and boron continues to play an important role as an efficient thermal neutron absorber. (author)

  2. Fail-safe logic elements for use with reactor safety systems

    International Nuclear Information System (INIS)

    Bobis, J.P.; McDowell, W.P.

    1976-01-01

    A complete fail-safe trip circuit is described which utilizes fail-safe logic elements. The logic elements used are analog multipliers and active bandpass filter networks. These elements perform Boolean operations on a set of AC signals from the output of a reactor safety-channel trip comparator

  3. Surveillance and Maintenance Plan for the 105-C Reactor Safe Storage Enclosure

    International Nuclear Information System (INIS)

    Logan, T. E.

    1998-01-01

    This document provides a plan for implementing surveillance and maintenance activities to ensure that the 105-C Reactor Safe Storage Enclosure is maintained in a safe, environmentally secure, and cost-effective manner until subsequent closure during the final disposition phase of decommissioning

  4. The ultimate safe (US) Reactor: A concept for the third millenium

    International Nuclear Information System (INIS)

    Gat, U.

    1986-01-01

    The Ultimate Safe (U.S.) Reactor is based on a novel safety concept. Fission products in the reactor are allowed to accumulate only to a level at which they would constitute a harmless source term. Removal of fission products also removes the decay heat - the driving force for the source term. The reactor has no excess criticality and is controlled by the reactivity temperature coefficient. Safety is inherent and passive. Waste is removed from the site promptly

  5. Safe operation of research reactors and critical assemblies. Code of practice and annexes. 1984 ed

    International Nuclear Information System (INIS)

    1984-01-01

    The safe operation of research reactors and critical assemblies (hereafter termed 'reactors') requires proper design, construction, management and supervision. This Code of Practice deals mainly with management and supervision. The provisions of the Code apply to the whole life of the reactor, including modification, updating and upgrading. The Code may be subject to revision in the light of experience and the state of technology. The Code is aimed at defining minimum requirements for the safe operation of reactors. Emphasis is placed on which safety requirements should be met rather than on specifying how these requirements may be met. The Code also provides guidance and information to persons and authorities responsible for the operation of reactors. The Code recommends that documents dealing with the operation of reactors and including safety analyses be prepared and submitted for review and approval to a regulatory body. Operation would be authorized on the understanding that it would comply with limits and conditions designed to ensure safety. The Code covers a wide range of reactor types, which gives rise to a variety of safety issues. Safety issues applicable to specific reactor types only (e.g. fast reactors) are not necessarily covered in this Code. Some of the recommendations in the Code are not directly applicable to critical assemblies. A recommendation may therefore be interpreted according to the type of reactor concerned. In such cases the words 'adequate' and 'appropriate' are used to mean 'adequate' or 'appropriate' for the type of reactor under consideration.

  6. Designing visual displays and system models for safe reactor operations

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  7. Designing visual displays and system models for safe reactor operations

    International Nuclear Information System (INIS)

    Brown-VanHoozer, S.A.

    1995-01-01

    The material presented in this paper is based on two studies involving the design of visual displays and the user's prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator's perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors

  8. Progress on PRISM, an inherently safe, economic, and testable advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Tippets, F.E.; Salerno, L.N.; Boardman, C.E.; Kwant, W.; Murata, R.E.; Snyder, C.R.

    1987-01-01

    This paper reports progress on the design of PRISM (Power Reactor Inherently Safe Module) under the DOE-sponsored innovative reactor program now in its third year at General Electric. The purpose of this program is to develop a design for an inherently safe, reliable, and marketable liquid metal fast reactor power plant. The PRISM design approach includes the following key elements: Compact sodium-cooled pool-type reactor modules that are sized to enable factory fabrication, economical shipment to inland as well as water-side sites, and economical full-scale prototype testing for design certification; Nuclear safety-related envelope limited to the reactor modules and their service systems; Inherent, passive shutdown heat removal for loss-of-cooling events; Inherent, passive reactivity shutdown for failure-to-scram events

  9. Safety of intrinsically safe and economical reactor (ISER)

    International Nuclear Information System (INIS)

    Asahi, Y.; Sugawara, I.; Yamanaka, K.

    1988-01-01

    Inherent safety of a reactor may be quantified by the grace period at various safety levels such as maintenance of fuel integrity, maintenance of fuel coolability and avoidance of core-melt. It is important to find out the grace period especially at the safety level of maintenance of fuel integrity. It has been conducted to design the ISER, which is characterized by the steel-made reactor pressure vessel. In addition to the passive nature of the safety design of the reactor itself, the ISER is equipped in the secondary system with a subsystem called the passive safety and shutdown system (PSSS), which will help to increase the grace period. It was found by the null transient analysis that check valves are needed at the top hot/cold interface. The analysis of the station blackout, which is one of the severest accident conceivable for the ISER, was made to examine inherent safety of the ISER with and without the PSSS. This paper reports that found out that the PSSS enhances inherent safety of the ISER

  10. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    International Nuclear Information System (INIS)

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  11. Inherent safe fast breeder reactors and actinide burners, metallic fuel

    International Nuclear Information System (INIS)

    Dorner, S.; Schumacher, G.

    1991-04-01

    Nuclear power without breeder strategy uses the possibilities for the energy supply only to a small extend compared to the possibilities of fast breeder reactors, which offer an energy supply for thousands of years. Moreover, a fast neutron device offers the opportunity to run an actinide-burner that could improve the situation of waste management. Within this concept metallic fuel could play a key role. The present report shows some important aspects of the concept like the pyrometallic reprocessing, the behaviour of metallic fuel during a core meltdown accident and others. The report should contribute to the discussion of these problems and initialize further work

  12. Metal-fuel modeling for inherently safe reactor designs

    International Nuclear Information System (INIS)

    Miles, K.J. Jr.

    1987-01-01

    Current development of breeder reactor systems has led to the renewed interest in metal fuels. These fuels have properties that enhance the inherent safety of the system, such as high thermal conductivity, compatibility with liquid sodium, and low fuel/cladding mechanical interaction. While metal-fuel irradiation behavior is well understood, there are some areas where more information is needed to fully understand the various safety-related phenomena, such as fuel/cladding chemical interaction, eutectic melting and penetration, and axial relocation of molten fuel prior to cladding breach. Because many of these phenomena can cause changes in the reactivity state of the system, their effects on whole-core normal, anticipated, and hypothetical accident scenarios need to be studied. The metal-fuel behavior model DEFORM-5 is being developed to provide the necessary phenomenological basis for these studies. The first stage in the DEFORM-5 development has been completed. Presently, DEFORM-5 calculates the cladding strain, life fraction, and eutectic penetration thinning for Types D9, HT9, or 316 steels. This first stage of DEFORM-5 has been used to analyze the TREAT M2, M3, and M4 transients with irradiated Experimental Breeder Reactor-II driver fuel. The paper shows the DEFORM-5 and experimental results for failure times for the test pins. The results provide confidence and validation of the DEFORM-5 modeling of the cladding behavior

  13. Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

    Science.gov (United States)

    Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

    2018-01-01

    Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been

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

    International Nuclear Information System (INIS)

    1978-09-01

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

  15. Conceptual design of Inherently Safe Fast Reactor (ISFR)

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2003-01-01

    ISFR is a boiling heavy water fast reactor of process inherent ultimate safety (PIUS) type. ISFR may breed fuel in the core. Owing to a positive void coefficient, the application of the PIUS concept to ISFR is not straightforward. Thus, the gap conductance is small so that the time constant τ α of the positive void feedback process is sufficiently large, while the initially-closed two-way check valves to be used as passive switches to the pumps are installed at the lower honeycombs. As a result, the passive shutdown mechanisms can come into effect sufficiently soon to suppress the positive feedback reactivity. Both large τ α and the passive switches also help stabilize the system so that ISFR can perform a constant power operation with a simple control logic for the main coolant pump speed. In a steam generator tube rupture, fuel temperature was found to smoothly decrease to the decay heat level with nucleate boiling. The feasibility of ISFR was proved only to some extent. (author)

  16. Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

    International Nuclear Information System (INIS)

    Schnitzler, Bruce G.

    2012-01-01

    Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse (∼900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as

  17. Assessment methodology applicable to safe decommissioning of Romanian VVR-S research reactor

    International Nuclear Information System (INIS)

    Baniu, O.; Vladescu, G.; Vidican, D.; Penescu, M.

    2002-01-01

    The paper contains the results of research activity performed by CITON specialists regarding the assessment methodology intended to be applied to safe decommissioning of the research reactors, developed taking into account specific conditions of the Romanian VVR-S Research Reactor. The Romanian VVR-S Research Reactor is an old reactor (1957) and its Decommissioning Plan is under study. The main topics of paper are as follows: Safety approach of nuclear facilities decommissioning. Applicable safety principles; Main steps of the proposed assessment methodology; Generic content of Decommissioning Plan. Main decommissioning activities. Discussion about the proposed Decommissioning Plan for Romanian Research Reactor; Safety risks which may occur during decommissioning activities. Normal decommissioning operations. Fault conditions. Internal and external hazards; Typical development of a scenario. Features, Events and Processes List. Exposure pathways. Calculation methodology. (author)

  18. Low-temperature thermionics in space nuclear power systems with the safe-type fast reactor

    International Nuclear Information System (INIS)

    Zrodnikov, A.V.; Yarygin, V.I.; Lazarenko, G.E.; Zabudko, A.N.; Ovcharenko, M.K.; Pyshko, A.P.; Mironov, V.S.; Kuznetsov, R.V.

    2007-01-01

    The potentialities of the use of the low-temperature thermionic converters (TIC) with the emitter temperature ≤ 1500 K in the space nuclear power system (SNPS) with the SAFE-type (Safe Affordable Fission Engine) fast reactor proposed and developed by common efforts of American experts have been considered. The main directions of the 'SAFE-300-TEG' SNPS (300 kW(thermal)) design update by replacing the thermoelectric converters with the low-temperature high-performance thermionic converters (with the barrier index V B ≤ 1.9 eV and efficiency ≥ 10%) meant for a long-term operation (5 years at least) as the components of the SAFE-300-TIC SNPS for a Lunar base have been discussed. The concept of the SNPS with the SAFE-type fast reactor and low-temperature TICs with specific electric power of about 1.45 W/cm 2 as the components of the SAFE-300-TIC system meeting the Nasa's initial requirements to a Lunar base with the electric power demand of about 30 kW(electrical) for robotic mission has been considered. The results, involving optimization and mass-and-size estimation, show that the SAFE-300-TIC system meets the initial requirements by Nasa to the lunar base power supply. The main directions of the system update aimed at the output electric power increase up to 100 kW(electrical) have also been presented. (authors)

  19. Sodium heat pipe module test for the SAFE-30 reactor prototype

    International Nuclear Information System (INIS)

    Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

    2001-01-01

    Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

  20. Concept of an inherently-safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

    2012-01-01

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  1. Ordeals of Chernobyl and the rejustification of the inherently safe reactors

    International Nuclear Information System (INIS)

    Lu, Y.

    1989-01-01

    This paper presents the necessity of developing inherently safe economic reactors (ISERs). Two characteristics which define inherent safety are discussed on the basis of various applications of such a principle in practice. Different design concepts of ISERs are then evaluated and their possible role in the future nuclear program of PRC discussed. A three-stage development strategy of ISERs in PRC is proposed

  2. Proposition of innovative and safe design of grid plate for Tehran research reactor

    International Nuclear Information System (INIS)

    Jalali, H.R.; Fadaei, A.H.

    2017-01-01

    Highlights: • An innovative and safe design for grid plate in research reactors proposed. • New grid plate acts as an independent shutdown system. • Neutronic and transient calculation was done using MTR-PC package. • Calculations show that the performance and safety of new design are acceptable. - Abstract: The purpose of this paper is to propose an innovative and safe design of grid plate for Tehran research reactor (TRR) without any reduction in its performance in comparison with the current operation. The new grid plate consisted of two joined cubic with empty walls which are place of fuels and heavy water, respectively. The proposed design is such that the reactor core is divided into two distinct parts using the heavy water. The heavy water is inserted in the walls of the new grid plate. The new design of grid plate by keeping the characteristics of the previous version creates the possibility of shutting the reactor down in critical condition. In this paper, at initial step, a simulation of acceptable benchmark for Tehran research reactor is performed which could be considered reliable and comparable with SAR (Safety Analysis Report) data. In the next step, two different designs are proposed for grid plate and then are applied to reactor core using simulation tools. For the proposed design: core excess reactivity, shutdown margin, control rod worth, neutron flux and kinetic parameters are calculated. Furthermore, the transient analysis was performed for the new design to check the status of reactor safety. Obtained results show that all neutronic parameters for the first operating core and the new design are comparable, and there is no reduction in the efficiency of reference core. Moreover, in the current design, a diverse and independent shutdown system for TRR was included. Nuclear reactor analysis codes including MTR-PC package were employed to carry out these calculations.

  3. Passive safe small reactor for distributed energy supply system sited in water filled pit at seaside

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Imayoshi, Shou

    2003-01-01

    Japan Atomic Energy Research Institute has developed a Passive Safe Small Reactor for Distributed Energy Supply System (PSRD) concept. The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down device. For improvement of economy, simplification of the reactor system and long operation of the core over five years without refueling with low enriched UO 2 fuel rods are achieved. To avoid releasing the radioactive materials to the circumstance even if a hypothetical accident, the containment is submerged in a pit filled with seawater at a seaside. Refueling or maintenance of the reactor can be conducted using an exclusive barge instead of the reactor building. (author)

  4. Model development for the dynamic analysis of the OSU inherently safe reactor. Part 1

    International Nuclear Information System (INIS)

    Aybar, H.S.

    1992-01-01

    Faculty and students in the Nuclear Engineering Program at the Ohio State University (OSU) have proposed a conceptual design for an inherently safe 340 MWe power reactor. The design is based on the state-of-the-art technology of LWRs and the High Temperature Gas- cooled Reactors (HTGRs). The OSU Inherently Safe Reactor (OSU-ISR) concept uses shorter than standard BWR fuel elements in the reactor core. All the fluid on the primary side is contained within a Prestressed Concrete Reactor Vessel (PCRV). This important feature significantly reduces the probability of a LOCA. A new feature of the OSU-ISR is an operator independent steam driven Emergency Core Cooling System (ECCS) housed within the PCRV. In accident conditions where the steam generators are incapacitated, steam from the core drives a jet injector, which takes water from the suppression pool and pumps it into the core cavity to maintain core coverability. The preliminary analysis of the concept was performed as a design project in the Nuclear Engineering Program at the OSU during the Spring of 1985, and published in ''Nuclear Technology.'' The use of a PCRV for ducting and containment and the replacement of forced recirculation with natural circulation on the primary side significantly improve the inherent safety of the plant. Currently, work is in progress for the refinement of the OSU-ISR concept, partially supported by a grant from the U.S. Department of Energy

  5. Technology development program for safe shipment of spent fuel from liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Freedman, J.M.; Humphreys, J.R.

    1975-10-01

    A comprehensive plan to develop shipping cask technology is described. Technical programs in the disciplines of heat transfer, structures and containment, spent fuel characterization, hot laboratory verification, shielding, and hazards analysis are discussed. Both short- and long-term goals in each discipline are delineated and how the disciplines interrelate is shown. The technologies developed will be used in the design, fabrication, and testing of truck-mounted and rail-car casks. These casks will be used for safely transporting short-cooled, high-burnup Liquid Metal Fast Breeder Reactor (LMFBR) spent fuel from reactors to reprocessing plants

  6. On-site interim storage of spent nuclear fuel: Emerging public issues

    International Nuclear Information System (INIS)

    Feldman, D.L.; Tennessee Univ., Knoxville, TN

    1992-01-01

    Failure to consummate plans for a permanent repository or above- ground interim Monitored Retrievable Storage (MRS) facility for spent nuclear fuel has spurred innovative efforts to ensure at-reactor storage in an environmentally safe and secure manner. This article examines the institutional and socioeconomic impacts of Dry Cask Storage Technology (DCST)-an approach to spent fuel management that is emerging as the preferred method of on-site interim spent fuel storage by utilities that exhaust existing storage capacity

  7. Assessment report of research and development activities in FY2006 activity. 'Fast reactor cycle technology development project' (Interim report)

    International Nuclear Information System (INIS)

    2007-08-01

    Japan Atomic Energy Agency (hereinafter referred to as 'JAEA') asked the advisory committee 'evaluation Committee of Research and Development (R and D) Activities for Advanced Nuclear System/Nuclear Fuel Cycle Technology' (hereinafter referred to as 'Committee') to assess the interim report on Fast Reactor Cycle Technology Development Project ' (former 'Feasibility Study on Commercialized Fast Reactor Cycle Systems') in FY2006, in accordance with 'General Guideline for the Evaluation of Government R and D Activities' by Japanese Cabinet Office, 'Guideline for Evaluation of R and D in Ministry of Education, Culture Sports, Science and Technology' and 'Regulation on Conduct for Evaluation of R and D Activities' by JAEA. In response to JAEA's request, the Committee assessed the R and D program over five years, the criteria for adoption judgment on innovative technologies at the end of 2010 (Project Review), and the organization structure for R and D. etc. (Management Review). As a result of review, the Committee concluded that this R and D program and its organization structure are almost reasonable. (author)

  8. Removal Action Workplan for the 105-DR and 105-F Building Interim Safe Storage Projects and Ancillary Buildings

    International Nuclear Information System (INIS)

    Rodovsky, T. J.

    1999-01-01

    This document is the removal action workplan (RAW) for the 105-DR and 105-F Reactor Buildings and ancillary facilities. These buildings and facilities are located in the 100-D/DR and 100-F Areas of the Hanford Site in Benton County, Washington, which is owned and operated by the U.S. Department of Energy (DOE). The 100 Areas (including 100-D/DR and 100-F Areas) of the Hanford Site were placed on the U.S. Environmental Protection Agency's (EPA) National Priorities List under the Comprehensive Environmental Response, Compensation,and Liability Act of 1980 (CERCLA). DOE has determined that hazardous substances in the 105-DR and 105-F Reactor Buildings and four ancillary facilities present a potential threat to human health or the environment DOE has also determined that a non-time critical removal action is warranted at these facilities

  9. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

  10. Concept of passive safe small reactor for distributed energy supply system

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Nakajima, Nobuya; Sawada, Ken-ichi; Yoritsune, Tsutomu; Shimada, Shoichiro; Nakano, Yoshihiro; Yonomoto, Taisuke; Takahashi, Hiroki

    2003-01-01

    This paper presents a concept of a Passive Safe Small Reactor for Distributed energy supply system (PSRD). The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down. To comply with a severe operation condition of PSRD, material of the ball bearing with graphite retainer has been selected by test. For improvement of economy, simplification of the reactor system and long operation of the core are achieved. Optimization of core design concerning the burnable poison ensures the burn-up of 28 GWd/t for low enriched UO 2 fuel rods. (author)

  11. The PRISM concept for a safe, economic and testable liquid metal fast reactor plant

    International Nuclear Information System (INIS)

    Berglund, R.C.; Salerno, L.N.; Tippets, F.E.

    1987-01-01

    The PRISM project is underway at General Electric as part of an advanced reactor conceptual design program sponsored by the US Department of Energy. The PRISM concept emphasizes inherent safety, modular construction, and factory fabrication. These features are intended to improve the basis for public acceptance, reduce cost,improve licensability, and reduce the risk of schedule delays and cost increases during construction. A PRISM power plant comprises a number of reactor modules. The relatively small size of the reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features for safe accommodation of accidents. These inherent safety features permit simplification and reduction of conventional safety-related systems in the plant. Testing of a full-size prototype reactor module is planned in the late 1990's to demonstrate these inherent safety characteristics. It is intended that the results of the test be used to obtain certification of the design by the US Nuclear Regulatory Commission preparatory to use of reactor modules built to this standard design in licensed commercial plants

  12. Safe use of the Institute of Nuclear Physics reactor with low enriched fuel

    International Nuclear Information System (INIS)

    Baytelesov, S.A.; Dosimbaev, A.A.; Koblik, Yu.N.; Salikhbaev, U.S.; Khalikov, U.A.; Yuldashev, B.S.

    2006-01-01

    Full text: The requirements for safe exploitation of reactor do not accept boiling of water on the surface of fuel elements. At determination of safe thermal regime of reactor (permissible level of power) the regime of the most heat-stressed fuel assembly (FA) in the active core was analyzed. By using ASTRA code [1] the heat-stressed sector is determined by most heat-stressed FA. In calculations the power of reactor was selected so that stock factor prior to the water boiling on the FA surface was not less than 1.45. Besides, in calculations the value of maximal energy density in examined FA is decreased by 10 %. As the part of the energy generated in the FA cores will be lost in constructional materials of the active zone and on the reflector. The stocks of safety before occurrence of instability of flow in gaps between of FA and before crisis of heat exchange are also analyzed. Further, by using the MCNP-4C code [2], densities of fast (E > 0,821 MeV) and thermal flows (E < 0,625 eV) of neutrons were calculated for those experimental channels where the irradiation of samples would be carried out. (author)

  13. Materials considerations for UF6 gas-core reactor. Interim report for preliminary design study

    International Nuclear Information System (INIS)

    Wagner, P.

    1977-04-01

    The limiting materials problem in a high-temperature UF 6 core reactor is the corrosion of the core containment vessel. The UF 6 , the lower fluorides of uranium, and the fluorine that exist at the anticipated reactor operating conditions (1000 K and about one atmosphere UF 6 ) are all corrosive. Because of this, the materials evaluation effort for this reactor design study has concentrated on the identification of a viable system for the containment vessel that meets both the materials and neutronic requirements. A study of the literature has revealed that the most promising corrosion-resistant candidates are Ni or Ni-Al alloys. One of the conclusions of this work is that the containment vessel use a nickel liner or clad since the use of Ni as a structural member is precluded by its relative blackness to thermal neutrons. Estimates of corrosion rates of Ni and Ni-Al alloys, the effects of the pressure and temperature of F 2 on the corrosion rates, calculated equilibrium gas compositions at reactor core operating conditions, suggested methods of fabrication, and recommendations for future research and development are included

  14. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

    2017-01-01

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  15. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  16. A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

    1987-01-01

    A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

  17. Assessment report of research and development activities in JFY2008 activity. 'Fast Reactor Cycle Technology Development Project' (Interim report)

    International Nuclear Information System (INIS)

    2009-08-01

    Japan Atomic Energy Agency (hereinafter referred to as 'JAEA') asked the advisory committee 'Evaluation Committee of Research and Development Activities for Advanced Nuclear System/Nuclear Fuel Cycle Technology' (hereinafter referred to as 'Committee') to assess the interim report on 'Fast Reactor Cycle Technology Development Project (FaCT project)' in JFY2008, in accordance with 'General Guideline for the Evaluation of Government R and D Activities' by Japanese Cabinet Office, 'Guideline for Evaluation of R and D in Ministry of Education, Culture, Sports, Science and Technology' and 'Regulation on Conduct for Evaluation of R and D Activities' by JAEA. In response of JAEA's request, the Committee assessed the 2006-2008 R and D results, the 2009-2010 R and D program and its R and D management. The Committee confirmed the progress status of the R and D for the adoption judgment on innovative technologies scheduled in 2010. As a result of the review, the committee has made suggestions for the future R and D plan and the improvements of the R and D organization structure/management. A CD-ROM is attached as an appendix. (J.P.N.)

  18. Gas Reactor International Cooperative Program. Interim report. Safety and licensing evaluaion of German Pebble Bed Reactor concepts

    International Nuclear Information System (INIS)

    1978-09-01

    The Pebble Bed Gas Cooled Reactor, as developed in the Federal Republic of Germany, was reviewed from a United States Safety and Licensing perspective. The primary concepts considered were the steam cycle electric generating pebble bed (HTR-K) and the process heat pebble bed (PNP), although generic consideration of the direct cycle gas turbine pebble bed (HHT) was included. The study examines potential U.S. licensing issues and offers some suggestions as to required development areas

  19. Interim report of the subcommittee on putting fast breeder reactors in practical use

    International Nuclear Information System (INIS)

    1983-01-01

    Light water reactors are expected to be the main power reactors in Japan for considerably long period hereafter, however, the basic policy of long term energy utilization is to use the plutonium and depleted uranium recovered by the reprocessing of spent fuel for fast breeder reactors, in view of the effective use of energy resources and the energy security. At present, the experimental FBR ''Joyo'' is in operation, and the preparation work was started for the prototype FBR ''Monju'', aiming at the criticality at the end of fiscal 1990. In the discussion, it was confirmed that for the practical use of FBRs, the improvement of their economy is the most important. But the construction cost for the Monju and the type of the demonstration FBR are not yet decided, and there are many uncertain elements. The demand and supply of uranium and the prospect of FBR development, the appropriate way of FBR development in Japan, the trend of FBR development in Japan and foreign countries, the evaluation of the economy of FBRs and the time of putting in practical use, the schedule of FBR development, the cost reduction in FBRs, the fuel cycle and others are reported. (Kako, I.)

  20. Nuclear desalination in the Arab world - Part II: Advanced inherent and passive safe nuclear reactors

    International Nuclear Information System (INIS)

    Karameldin, A.; Samer S. Mekhemar

    2004-01-01

    Rapid increases in population levels have led to greater demands for fresh water and electricity in the Arab World. Different types of energies are needed to contribute to bridging the gap between increased demand and production. Increased levels of safeguards in nuclear power plants have became reliable due to their large operational experience, which now exceeds 11,000 years of operation. Thus, the nuclear power industry should be attracting greater attention. World electricity production from nuclear power has risen from 1.7% in 1970 to 17%-20% today. This ratio had increased in June 2002 to reach more than 30%, 33% and 42% in Europe, Japan, and South Korea respectively. In the Arab World, both the public acceptance and economic viability of nuclear power as a major source of energy are greatly dependent on the achievement of a high level of safety and environmental protection. An assessment of the recent generation of advanced reactor safety criteria requirements has been carried out. The promising reactor designs adapted for the Arab world and other similar developing countries are those that profit from the enhanced and passive safety features of the new generation of reactors, with a stronger focus on the effective use of intrinsic characteristics, simplified plant design, and easy construction, operation and maintenance. In addition, selected advanced reactors with a full spectrum from small to large capacities, and from evolutionary to radical types, which have inherent and passive safety features, are discussed. The relevant economic assessment of these reactors adapted for water/electricity cogeneration have been carried out and compared with non-nuclear desalination methods. This assessment indicates that, water/electricity cogeneration by the nuclear method with advanced inherent and passive safe nuclear power plants, is viable and competitive. (author)

  1. Concept study for interim storage of research reactor fuel elements in transport and storage casks. Transport and storage licensing procedure for the CASTOR MTR 2 cask. Final report

    International Nuclear Information System (INIS)

    Weiss, M.

    2001-01-01

    As a result of the project, a concept was to be developed for managing spent fuel elements from research reactors on the basis of the interim storage technology existing in Germany, in order to make the transition to direct disposal possible in the long term. This final report describes the studies for the spent fuel management concept as well as the development of a transport and storage cask for spent fuel elements from research reactors. The concept analyses were based on data of the fuel to be disposed of, as well as the handling conditions for casks at the German research reactors. Due to the quite different conditions for handling of casks at the individual reactors, it was necessary to examine different cask concepts as well as special solutions for loading the casks outside of the spent fuel pools. As a result of these analyses, a concept was elaborated on the basis of a newly developed transport and storage cask as well as a mobile fuel transfer system for the reactor stations, at which a direct loading of the cask is not possible, as the optimal variant. The cask necessary for this concept with the designation CASTOR trademark MTR 2 follows in ist design the tried and tested principles of the CASTOR trademark casks for transport and interim storage of spent LWR fuel. With the CASTOR trademark MTR 2, it is possible to transport and to place into long term interim storage various fuel element types, which have been and are currently used in German research reactors. The technical development of the cask has been completed, the documents for the transport license as type B(U)F package design and for obtaining the storage license at the interim storage facility of Ahaus have been prepared, submitted to the licensing authorities and to a large degree already evaluated positively. The transport license of the CASTOR trademark MTR 2 has been issued for the shipment of VKTA-contents and FRM II compact fuel elements. (orig.)

  2. Sustainable Solutions for Nuclear used Fuels Interim Storage

    International Nuclear Information System (INIS)

    Arslan, Marc; Favet, Dominique; Issard, Herve; Le Jemtel, Amaury; Drevon, Caroline

    2014-01-01

    AREVA has a unique experience in providing sustainable solutions for used fuel management, fitted with the needs of different customers in the world and with regulation in different countries. These solutions entail both recycling and interim storage technologies. In a first part, we will describe the various types of solutions for Interim Storage of UNF that have been implemented around the world for interim storage at reactor or centralized Pad solution in canisters dry storage, vault type storages for dry storage, dry storage of transportation casks (dual purpose) pools for wet storage, The experience for all these different families of interim storages in which AREVA is involved is extensive and will be discussed with respect to the new challenges: increase of the duration of the interim storage (long term interim storage) increase of burn up of the fuels In a second part of the presentation, special recycling features will be presented. In that case, interim storage of the used fuels is ensured in pools. This provides in the long term good conditions for the behaviour of the fuel and its retrievability. With recycling, the final waste (Universal Canister of vitrified fission products and compacted hulls and end pieces): is stable and licensed in many countries for the final disposal (France, UK, Belgium, NL, Switzerland, Germany, Japan, upcoming: Spain, Australia, Italy). Presents neither safety criticality risks nor proliferation risks (AREVA conditioned HLW and LL-ILW are free of IAEA safeguard constraints thanks to AREVA process high recovery and purification yields). It can therefore be safely stored in interim storage for more than 100 years before final disposal. Some economic considerations will also be discussed. In particular, in the case of long term interim storage of used fuels, there are growing uncertainties regarding the future needs of repackaging and transportation, which can result in future cost overruns. Meanwhile, in the recycling policy

  3. Interim results of the study of control room crew staffing for advanced passive reactor plants

    International Nuclear Information System (INIS)

    Hallbert, B.P.; Sebok, A.; Haugset, K.

    1996-01-01

    Differences in the ways in which vendors expect the operations staff to interact with advanced passive plants by vendors have led to a need for reconsideration of the minimum shift staffing requirements of licensed Reactor Operators and Senior Reactor Operators contained in current federal regulations (i.e., 10 CFR 50.54(m)). A research project is being carried out to evaluate the impact(s) of advanced passive plant design and staffing of control room crews on operator and team performance. The purpose of the project is to contribute to the understanding of potential safety issues and provide data to support the development of design review guidance. Two factors are being evaluated across a range of plant operating conditions: control room crew staffing; and characteristics of the operating facility itself, whether it employs conventional or advanced, passive features. This paper presents the results of the first phase of the study conducted at the Loviisa nuclear power station earlier this year. Loviisa served as the conventional plant in this study. Data collection from four crews were collected from a series of design basis scenarios, each crew serving in either a normal or minimum staffing configuration. Results of data analyses show that crews participating in the minimum shift staffing configuration experienced significantly higher workload, had lower situation awareness, demonstrated significantly less effective team performance, and performed more poorly as a crew than the crews participating in the normal shift staffing configuration. The baseline data on crew configurations from the conventional plant setting will be compared with similar data to be collected from the advanced plant setting, and a report prepared providing the results of the entire study

  4. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    International Nuclear Information System (INIS)

    Ross, Kyle; Cardoni, Jeffrey N.; Wilson, Chisom Shawn; Morrow, Charles; Osborn, Douglas; Gauntt, Randall O.

    2015-01-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  5. Modeling of the Reactor Core Isolation Cooling Response to Beyond Design Basis Operations - Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, Chisom Shawn [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, Douglas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    Efforts are being pursued to develop and qualify a system-level model of a reactor core isolation (RCIC) steam-turbine-driven pump. The model is being developed with the intent of employing it to inform the design of experimental configurations for full-scale RCIC testing. The model is expected to be especially valuable in sizing equipment needed in the testing. An additional intent is to use the model in understanding more fully how RCIC apparently managed to operate far removed from its design envelope in the Fukushima Daiichi Unit 2 accident. RCIC modeling is proceeding along two avenues that are expected to complement each other well. The first avenue is the continued development of the system-level RCIC model that will serve in simulating a full reactor system or full experimental configuration of which a RCIC system is part. The model reasonably represents a RCIC system today, especially given design operating conditions, but lacks specifics that are likely important in representing the off-design conditions a RCIC system might experience in an emergency situation such as a loss of all electrical power. A known specific lacking in the system model, for example, is the efficiency at which a flashing slug of water (as opposed to a concentrated jet of steam) could propel the rotating drive wheel of a RCIC turbine. To address this specific, the second avenue is being pursued wherein computational fluid dynamics (CFD) analyses of such a jet are being carried out. The results of the CFD analyses will thus complement and inform the system modeling. The system modeling will, in turn, complement the CFD analysis by providing the system information needed to impose appropriate boundary conditions on the CFD simulations. The system model will be used to inform the selection of configurations and equipment best suitable of supporting planned RCIC experimental testing. Preliminary investigations with the RCIC model indicate that liquid water ingestion by the turbine

  6. ADVANTAGES, DISADVANTAGES, AND LESSONS LEARNED FROM MULTI-REACTOR DECOMMISSIONING PROJECTS

    International Nuclear Information System (INIS)

    Morton, M.R.; Nielson, R.R.; Trevino, R.A.

    2003-01-01

    This paper discusses the Reactor Interim Safe Storage (ISS) Project within the decommissioning projects at the Hanford Site and reviews the lessons learned from performing four large reactor decommissioning projects sequentially. The advantages and disadvantages of this multi-reactor decommissioning project are highlighted

  7. Stress analysis of pressurized water reactor steam generator tube denting phenomena. Interim report

    International Nuclear Information System (INIS)

    Thomas, J.M.; Cipolla, R.C.; Ranjan, G.V.; Derbalian, G.

    1978-07-01

    In some Pressurized Water Reactor (PWR) steam generators, a corrosion product has formed on the carbon steel support plate in the crevice between the tube and support plate. The corrosion product occupies more volume than the original metal; the tube-to-support plate crevice volume is thus consumed with corrosion product, and further corrosive action results in a radially inward force on the tube and a radially outward force on the corroding support plate. This has resulted in indentation (''denting'') of the tube, accompanied by occasional cracking. Large in-plane deformation and cracking of support plates has also been observed in the most severely affected plants along with some serious side effects, such as deformation and cracking of inner row tube U-bends caused by support plate movement. Mechanical aspects of the tube denting phenomena have been studied using analytical models. The models used ranged from closed form analytical solutions to state-of-the-art numerical elastic-plastic computer program for moderate strains. It was found that tube dents, such as those observed in operating steam generators, are associated with yielding of both the tubes and support plates. Also studied were the stresses in tube U-bends caused by support plate flow slot deformation

  8. Interim Status Report on the Design of the Gas-Cooled Fast Reactor (GFR)

    International Nuclear Information System (INIS)

    Weaver, K. D.

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above

  9. An integrated approach to assessing the fracture safe margins of fusion reactor structures

    International Nuclear Information System (INIS)

    Odette, G.R.

    1996-01-01

    Design and operation of fusion reactor structures will require an appropriate data base closely coupled to a reliable failure analysis method to safely manage irradiation embrittlement. However, ongoing irradiation programs will not provide the information on embrittlement necessary to accomplish these objectives. A new engineering approach is proposed based on the concept of a master toughness-temperature curve indexed on an absolute temperature scale using shifts to account for variables such as size scales, crack geometry and loading rates as well as embrittlement. While providing a simple practical engineering expedient, the proposed method can also be greatly enhanced by fundamental mechanism based models of fracture and embrittlement. Indeed, such understanding is required for the effective use of small specimen test methods, which is a integral element in developing the necessary data base

  10. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 3. Alternatives for interim storage and transportation

    International Nuclear Information System (INIS)

    1976-05-01

    Volume III of the five-volume report contains information on alternatives for interim storage and transportation. Section titles are: interim storage of spent fuel elements; interim storage of chop-leach fuel bundle residues; tank storage of high-level liquid waste; interim storage of solid non-high-level wastes; interim storage of solidified high-level waste; and, transportation alternatives

  11. Preliminary risk assessment of the Integral Inherently-Safe Light Water Reactor

    International Nuclear Information System (INIS)

    McCarroll, Kellen R.; Lee, John C.; Manera, Annalisa; Memmott, Matthew J.; Ferroni, Paolo

    2017-01-01

    The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) concept seeks to significantly increase nuclear power plant safety. The project implements a safety-by-design philosophy, eliminating several initiating events and providing novel, passive safety systems at the conceptual phase. Pursuit of unparalleled safety employs an integrated development process linking design with deterministic and probabilistic safety analyses. Unique aspects of the I 2 S-LWR concept and design process present challenges to the probabilistic risk assessment (PRA), particularly regarding overall flexibility, auditability and resolution of results. Useful approaches to initiating events and conditional failures are presented. To exemplify the risk-informed design process using PRA, a trade-off study of two safety system configurations is presented. Although further optimization is required, preliminary results indicate that the I 2 S-LWR can achieve a core damage frequency (CDF) from internal events less than 1.01 × 10 −8 /ry, including reactor vessel ruptures. Containment bypass frequency due to primary heat exchanger rupture is found to be comparable to non-vessel rupture CDF.

  12. Proceedings of the conference on the Safety in Reactor Operations - TopSafe 2012 Transactions

    International Nuclear Information System (INIS)

    2012-01-01

    TopSafe 2012 provides a forum for addressing the current status and future perspectives with regards to safety at nuclear installations worldwide. In view of the on-going discussions and initiatives that have been taken over the last months the European Nuclear Society (ENS) decided organising this edition of this topical conference from 22 to 26 April 2012 in Helsinki, Finland. TopSafe 2012 focus on three main subjects: Safety and related analyses in operating nuclear power plants and other nuclear installations; Safety and Risk Assessment; Trends in nuclear safety for existing and future installations. The conference is directed at a broad range of experts in the area of nuclear safety, including professionals from the different disciplines involved in the safety of nuclear power plants, fuel cycle installations and research reactors. It is aimed at professionals coming from the research organisations, universities, vendors, operators, regulatory bodies as well as policy makers. Top level representatives of the Countries that are constructing new nuclear power plants are invited. Regulators of all individual Countries with nuclear programme are expected to contribute the Conference. (authors)

  13. Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

  14. An optimized power conversion system concept of the integral, inherently-safe light water reactor

    International Nuclear Information System (INIS)

    Memmott, Matthew J.; Wilding, Paul R.; Petrovic, Bojan

    2017-01-01

    Highlights: • Three power conversion systems (PCS) for the I"2S-LWR are presented. • An optimization analyses was performed to evaluate these PCS alternatives. • The ideal PCS consists of 5 turbines, and obtains an overall efficiency of 35.7%. - Abstract: The integral, inherently safe light water reactor (I"2S-LWR) has been developed to significantly enhance passive safety capabilities while maintaining cost competitiveness relative to the current light water reactor (LWR) fleet. The compact heat exchangers of the I"2S-LWR preclude boiling of the secondary fluid, which decreases the probability of heat exchanger failure, but this requires the addition of a flash drum, which negatively affects the overall plant thermodynamic efficiency. A state of the art Rankine cycle is proposed for the I"2S-LWR to increase the thermodynamic efficiency by utilizing a flash drum with optimized operational parameters. In presenting this option for power conversion in the I"2S-LWR power plant, the key metric used in rating the performance is the overall net thermodynamic efficiency of the cycle. In evaluating the flash-Rankine cycle, three basic industrial concepts are evaluated, one without an intermediate pressure turbine, one with an intermediate turbine and one reheat stream, and one with an intermediate turbine and two reheat streams. For each configuration, a single-path multi-variable optimization is undertaken to maximize the thermal efficiency. The third configuration with an intermediate turbine and 2 reheat streams is the most effective concept, with an optimized efficiency of 35.7%.

  15. Reactivity control system of a passively safe thorium breeder pebble bed reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2014-01-01

    Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF 3 gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW th passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel, between 100 cm

  16. Reactivity control system of a passively safe thorium breeder pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wols, F.J., E-mail: f.j.wols@tudelft.nl; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2014-12-15

    Highlights: • A worth of over 15,000 pcm ensures achieving long-term cold shutdown in thorium PBR. • Control rod worth in side reflector is insufficient due to low-power breeder zone. • 20 control rods, just outside the driver zone, can achieve long-term cold shutdown. • BF{sub 3} gas can be inserted for reactor shutdown, but only in case of emergency. • Perturbation theory accurately predicts absorber gas worth for many concentrations. - Abstract: This work investigates the neutronic design of the reactivity control system for a 100 MW{sub th} passively safe thorium breeder pebble bed reactor (PBR), a conceptual design introduced previously by the authors. The thorium PBR consists of a central driver zone of 100 cm radius, surrounded by a breeder zone with 300 cm outer radius. The fissile content of the breeder zone is low, leading to low fluxes in the radial reflector region. Therefore, a significant decrease of the control rod worth at this position is anticipated. The reactivity worth of control rods in the side reflector and at alternative in-core positions is calculated using different techniques, being 2D neutron diffusion, perturbation theory and more accurate 3D Monte Carlo models. Sensitivity coefficients from perturbation theory provide a first indication of effective control rod positions, while the 2D diffusion models provide an upper limit on the reactivity worth achievable at a certain radial position due to the homogeneous spreading of the absorber material over the azimuthal domain. Three dimensional forward calculations, e.g. in KENO, are needed for an accurate calculation of the total control rod worth. The two dimensional homogeneous calculations indicate that the reactivity worth in the radial reflector is by far insufficient to achieve cold reactor shutdown, which requires a control rod worth of over 15 000 pcm. Three dimensional heterogeneous KENO calculations show that placing 20 control rods just outside the driver channel

  17. Safe operation of critical assemblies and research reactors. Code of practice and Technical appendix. 1971 ed

    International Nuclear Information System (INIS)

    Cox, J.

    1971-01-01

    This book is in two parts. The first is a Code of Practice for the Safe Operation of Critical Assemblies and Research Reactors, prepared as a result of a meeting of experts which took place in Vienna on 20-24 May 1968. The Code has been prepared by the International Atomic Energy Agency in co-operation with the World Health Organization, and its publication is sponsored by both organizations. In addition, the Code was approved by the Board of Governors of the International Atomic Energy Agency on 16 December 1968 as part of the Agency's safety standards, which are applied to operations undertaken by Member States with the assistance of the Agency. The Board, in approving the publication of the present book, also recommended Member States to take the Code into account in the formulation of national regulations and recommendations. The second part of the book is a Technical Appendix to give information and illustrative samples that would be helpful in implementing the Code of Practice. This second part, although published under the same cover, is not part of the Code. An extensive Bibliography, amplifying the Technical Appendix, is included at the end.

  18. Analysis of the running-in phase of a Passively Safe Thorium Breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    Highlights: • This work analyzes important trends of the running-in phase of a thorium breeder PBR. • Depletion equations are solved for important actinides and a fission product pair. • Breeding U-233 is achieved in 7 years by cleverly adjusting the feed fuel enrichment. • A safety analysis shows the thorium PBR is passively safe during the running-in phase. - Abstract: The present work investigates the running-in phase of a 100 MW th Passively Safe Thorium Breeder Pebble Bed Reactor (PBR), a conceptual design introduced in previous equilibrium core design studies by the authors. Since U-233 is not available in nature, an alternative fuel, e.g. U-235/U-238, is required to start such a reactor. This work investigates how long it takes to converge to the equilibrium core composition and to achieve a net production of U-233, and how this can be accelerated. For this purpose, a fast and flexible calculation scheme was developed to analyze these aspects of the running-in phase. Depletion equations with an axial fuel movement term are solved in MATLAB for the most relevant actinides (Th-232, Pa-233, U-233, U-234, U-235, U-236 and U-238) and the fission products are lumped into a fission product pair. A finite difference discretization is used for the axial coordinate in combination with an implicit Euler time discretization scheme. Results show that a time dependent adjustment scheme for the enrichment (in case of U-235/U-238 start-up fuel) or U-233 weight fraction of the feed driver fuel helps to restrict excess reactivity, to improve the fuel economy and to achieve a net production of U-233 faster. After using U-235/U-238 startup fuel for 1300 days, the system starts to work as a breeder, i.e. the U-233 (and Pa-233) extraction rate exceeds the U-233 feed rate, within 7 years after start of reactor operation. The final part of the work presents a basic safety analysis, which shows that the thorium PBR fulfills the same passive safety requirements as the

  19. Advanced Small-Safe Long-Life Lead Cooled Reactor Cores for Future Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Hyeong; Hong, Ser Gi [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    One of the reasons for use of the lead or lead-bismuth alloy coolants is the high boiling temperature that avoids the possibility of coolant voiding. Also, these coolants are compatible with air, steam, and water. Therefore, intermediate coolant loop is not required as in the sodium cooled reactors 3. Lead is considered to be more attractive coolant than lead-bismuth alloy because of its higher availability, lower price, and much lower amount of polonium activity by factor of 104 relatively to lead. On the other hand, lead has higher melting temperature of 601K than that of lead-bismuth (398K), which narrows the operating temperature range and also leads to the possibility of freezing and blockage in fresh cores. Neutronically, the lead and lead-bismuth have very similar characteristics to each other. The lead-alloy coolants have lower moderating power and higher scattering without increasing moderation for neutrons below 0.5MeV, which reduces the leakage of the neutrons through the core and provides an excellent reflecting capability for neutrons. Due to the above features of lead or lead-alloy coolants, there have been lots of studies on the small lead cooled core designs. In this paper, small-safe long-life lead cooled reactor cores having high discharge burnup are designed and neutronically analyzed.. The cores considered in this work rates 110MWt (36.7MWe). In this work, the long-life with high discharge burnup was achieved by using thorium or depleted uranium blanket loaded in the central region of the core. Also, we considered a reference core having no blanket for the comparison. This paper provides the detailed neutronic analyses for these small long-life cores and the detailed analyses of the reactivity coefficients and the composition changes in blankets. The results of the core design and analyses show that our small long-life cores can be operated without refueling over their long-lives longer than 45EFPYs (Effective Full Power Year). In this work

  20. The European Pressurized Water Reactor. A safe and competitive solution for future energy needs

    International Nuclear Information System (INIS)

    Leverenz, R.; Gerhard, L.; Goebel, A.

    2004-01-01

    The European Pressurized Water Reactor - the EPR - is a PWR in the 1600 MW class. Its design is based on experience feedback from several thousand reactors x years of light water reactor operation worldwide, primarily those incorporating the most recent technologies: the French N4 and the German KONVOI reactors. It is an evolutionary design that ensures continuity in the mastery of PWR technology, minimizing the risk for the customer. (author)

  1. Ageing management and knowledge base for safe long-term operation of japanese light water reactors

    International Nuclear Information System (INIS)

    Sekimura, N.

    2008-01-01

    There are 55 operating commercial light water reactor plants (32 BWRs and 23 PWRs) in Japan. Twelve (12) plants have been operating for more than 30 years. Utility companies are required to perform an 'Ageing Management Technical Assessment' be-fore the end of 30 years operation of each plant. The assessments for each plant have been evaluated by the Nuclear and Industry Safety Agency (NISA) of the Ministry of Economy Trade and Industry (WTI) for these 12 plants. The Japan Nuclear Energy Safety Organisation (JNES) has compiled Technical Review Manuals for six major degradation phenomena for the evaluation of Ageing Management Technical Assessment. A 'Road-map for Ageing and Plant Life Management' was established in 2005 by the Special Committee in the Atomic Energy Society of Japan under the commission from the JNES. Within the framework of the road-map, the major research and development fields are divided into the following four categories: 1) engineering information systems; 2) research and development of technologies for inspection, evaluation and repair of the components and materials; 3) development of codes and standards; 4) synthesised maintenance engineering. Continuous revision of the 'Strategy Maps for Ageing Management and Safe Long-term Operation' has been performed under the Coordinating Committee of Ageing Management to promote research and development activities by industries, government and academia, effectively and efficiently. Systematic development of the information basis for database and knowledge-base has been undertaken in addition to the development of codes and standards by academic societies through intensive domestic safety research collaborations and international collaboration. (author)

  2. Interim restorations.

    Science.gov (United States)

    Gratton, David G; Aquilino, Steven A

    2004-04-01

    Interim restorations are a critical component of fixed prosthodontic treatment, biologically and biomechanically. Interim restoration serves an important diagnostic role as a functional and esthetic try-in and as a blueprint for the design of the definitive prosthesis. When selecting materials for any interim restoration, clinicians must consider physical properties, handling properties, patient acceptance, and material cost. Although no single material meets all the requirements and material classification alone of a given product is not a predictor of clinical performance, bis-acryl materials are typically best suited to single-unit restorations, and poly(methylmethacrylate) interim materials are generally ideal for multi-unit, complex, long-term, interim fixed prostheses. As with most dental procedures, the technique used for fabrication has a greater effect on the final result than the specific material chosen.

  3. Supervisory and managerial aspects on the safe operation of the Egyptian second research reactor

    International Nuclear Information System (INIS)

    Abdelrazek, I.D.; Shokr, A.M.

    2000-01-01

    ETRR-2 is a multipurpose reactor for radioisotopes production, neutron beam experiments, basic and applied research in physics and engineering, fuels and material tests, and for training. The reactor is an open pool type, 22 MW power, with average thermal neutron flux of 1.4x10 14 n/cm 2 sec, cooled and moderated by light water and with beryllium reflectors. Various experimental devices and irradiation facilities are integrated with the reactor. The reactor has been licensed for operation in November 1998. Several principles and regulations have been applied to all the reactor project stages to achieve safety. Moreover, other several principals, regulations, and aspects are enforced by the AEA, National Center for Nuclear Safety and Radiation Control, NCNSRC, and reactor management to achieve safety during reactor operation and utilization. Responsibility on Safety and Supervision Aspects AEA chairman has the ultimate responsibility on the reactor safety during operation and utilization. The primary responsible on the safety is the ETRR-2 supervisor, who is supervising the ETRR-2 site that includes the ETRR-2 reactor, Fuel Manufacturing Plant, and other two projects (under execution): Radioisotopes Production Plant and Dry Fuel Storage Facility. ETRR-2 supervisor is responsible to ensure that: the reactor is operated in accordance with the safety requirements by qualified and trained personnel, updating and enforcement of the reactor mandatory documentation, and the services are adequate for the reactor operation. He is responsible, also, to guarantee that: the reactor manager has enough authority and resources to carry out his function effectively and the reactor is kept operated in agreement with established procedures. A Technical Revision Committee, TRC, is formed to advice the ETRR-2 supervisor on the safety of the reactor and experiments. The committee members are from AEA experts with no direct relation to the reactor and experiments being performed. Members

  4. International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13). Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The conference, which was held from 4 to 7 of March 2013 in Paris, provided a forum to exchange information on national and international programmes, and more generally new developments and experience, in the field of fast reactors and related fuel cycle technologies. A first goal was to identify and discuss strategic and technical options that have been proposed by individual countries or companies. Another goal was to promote the development of fast reactors and related fuel cycle technologies in a safe, proliferation resistant and economic way. A third goal was to identify gaps and key issues that need to be addressed in relation to the industrial deployment of fast reactors with a closed fuel cycle. A fourth goal was to engage young scientists and engineers in this field, in particular with sustainability, innovation, simulation, safety, economics and public acceptance

  5. Interim overdentures.

    Science.gov (United States)

    Fenton, A H

    1976-07-01

    The construction of an interim overdenture using existing removable partial dentures with natural tooth crowns and artificial teeth can be a simple and economical method of providing patients with dentures while tissues heal and teeth are prepared and restored. A more definite prognosis for both the patient and his remaining dentition can be established before the final overdenture is completed. The procedures necessary to provide three types of interim overdentures have been outlined. Patients tolerate this method of changing their dentitions extremely well.

  6. Inherent safe design of advanced high temperature reactors - concepts for future nuclear power plants

    International Nuclear Information System (INIS)

    Hodzic, A.; Kugeler, K.

    1997-01-01

    This paper discusses the applicable solutions for a commercial size High Temperature Reactor (HTR) with inherent safety features. It describes the possible realization using an advanced concept which combines newly proposed design characteristics with some well known and proven HTR inherent safety features. The use of the HTR technology offers the conceivably best solution to meet the legal criteria, recently stated in Germany, for the future reactor generation. Both systems, block and pebble bed ,reactor, could be under certain design conditions self regulating in terms of core nuclear heat, mechanical stability and the environmental transfer. 23 refs., 7 figs

  7. Sustainable and safe nuclear fission energy technology and safety of fast and thermal nuclear reactors

    CERN Document Server

    Kessler, Günter

    2012-01-01

    Unlike existing books of nuclear reactor physics, nuclear engineering and nuclear chemical engineering this book covers a complete description and evaluation of nuclear fission power generation. It covers the whole nuclear fuel cycle, from the extraction of natural uranium from ore mines, uranium conversion and enrichment up to the fabrication of fuel elements for the cores of various types of fission reactors. This is followed by the description of the different fuel cycle options and the final storage in nuclear waste repositories. In addition the release of radioactivity under normal and possible accidental conditions is given for all parts of the nuclear fuel cycle and especially for the different fission reactor types.

  8. Designing for a safe response to operational and severe accident initiators in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1994-01-01

    A method is described for optimizing the plant control strategy for a liquid metal reactor with respect to safety margins sustained in unprotected upset events. The optimization is performed subject to the normal requirements for reactor startup, load change and compensation for reactivity changes over the cycle. The method provides a formal approach to the process of exploiting the innate self-regulating property of a metal fueled reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error

  9. Inherently safe shutdown of a liquid metal reactor upon a loss of intermediate cooling

    International Nuclear Information System (INIS)

    Feldman, E.E.; Mohr, D.

    1986-01-01

    Two unprotected (i.e., no scram or plant protection system action) loss-of-heat sink transients are scheduled to be performed on the Experimental Breeder Reactor-II in the Spring of 1986. One is to be initiated from full power (about 60 MW) and the other from half power. The loss-of-heat sink in each test is to be accomplished by essentially stopping the secondary-loop sodium coolant flow in about 20 s. Pretest predictions for the two tests, provided herein, show the reactor to passively shut down with average reactor outlet coolant temperatures rising no more that 5/degree/C before dropping and asymptotically approaching a quenching (or ''smothering'') temperature. This mild behavior is seen to be a direct consequence of the substantial negative reactivity feedback of the reactor coupled with the heat capacitance of the massive sodium pool in the primary tank. 14 refs., 11 figs., 1 tab

  10. The safe, economical operation of a slightly subcritical reactor and transmutor with a small proton accelerator

    International Nuclear Information System (INIS)

    Takahashi, Hiroshi

    1994-01-01

    I suggest that an accelerator can be used to increase the safety and neutron economy of a power reactor and a transmutor of long-lived radioactive wastes, such as minor actinides and fission products, by providing neutrons for its subcritical operation. Instead of the large subcriticality k=0.9-0.95 which we originally proposed for such transmutor, we propose to use a slightly subcritical reactor, such as k=0.99, which will avoid many of the technical difficulties that are associated with large subcriticality, such as localized power peaking, radiation damage due to injection of medium-energy protons, the high current accelerator, and the requirement for a long beam-expansion section. We analyze the power drop that occurred in Phenix reactor, and show that the operating this reactor in subcritical conditions improves safety. (author). 13 refs., 5 figs

  11. The safe, economical operation of a slightly subcritical reactor and transmutor with a small proton accelerator

    International Nuclear Information System (INIS)

    Takahashi, Hiroshi.

    1994-01-01

    This report describes methods in which an accelerator can be used to increase the safety and neutron economy of a power reactor and transmutor of long-lived radioactive wastes, such as minor actinides and fission products, by providing neutrons for its subcritical operation. Instead of the rather large subcriticality of k=0.9--0.95 which we originally proposed for such a transmutor, we propose to use a slightly subcritical reactor, such as k=0.99, which will avoid many of the technical difficulties that are associated with large subcriticality, such as localized power peaking, radiation damage due to the injection of medium-energy protons, the high current accelerator, and the requirement for a long beam-expansion section. We analyzed the power drop that occurred in Phoenix reactor, and show that the operating this reactor in subcritical condition improves its safety

  12. Potential of small nuclear reactors for future clean and safe energy sources

    International Nuclear Information System (INIS)

    Sekimoto, H.

    1992-01-01

    To cope with the various kinds of energy demands expected in the 21st century, it is necessary to explore the potential of small nuclear reactors and to find a way of promoting their introduction to society. The main goal of current research activities is 'the constitution of the self-consistent nuclear energy system'. These activities can be understood by realizing that the nuclear community is facing a turning point for its survival in the 21st century. Self-consistency can be manifested by investigating and developing the potential advantages of the nuclear fission reaction and lessening the potential disadvantages. The contributions in this volume discuss concepts of small reactors, applications of small reactors, and consistency with conventional energy supply systems

  13. The development of a small inherently safe homogeneous reactor for the production of medical isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Carlin, G.E.; Bonin, H.W., E-mail: george.carlin@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2013-07-01

    The use of radioisotopes for various procedures in the health care industry has become one of the most important practices in medicine. New interest has been found in the use of liquid fueled nuclear reactors to produce these isotopes due to the ease of fuel processing and ability to efficiently use LEU as the fuel source. A version of this reactor is being developed at the Royal Military College of Canada to act as a successor to the SLOWPOKE-2 platform. The thermal hydraulic and transient characteristics of a 20 kWt version are being studied to verify inherent safety abilities. (author)

  14. Australia's nuclear reactor: how safe is the old lady of Lucas Heights?

    International Nuclear Information System (INIS)

    Darroch, R.

    1990-01-01

    Safety procedures at the Lucas Heights reactor are discussed based on concerns voiced by a group of senior engineers working at the site. An account is also given of the responses to this claims of the authorities, the courts as well as some of the safety considerations outlined in a recent review of HIFAR management, recently completed by a team of nuclear experts from the Atomic Energy of Canada. It is argued that sufficient doubts have been raised to warrant an independent inquiry on the safety and future of the aging reactor. ills

  15. CANDLE reactor: an option for simple, safe, high nuclear proliferation resistant , small waste and efficient fuel use reactor

    International Nuclear Information System (INIS)

    Sekimoto, H.

    2010-01-01

    The innovative nuclear energy systems have been investigated intensively for long period in COE-INES program and CRINES activities in Tokyo Institute of Technology. Five requirements; sustainability, safety, waste, nuclear-proliferation, and economy; are considered as inevitable requirements for nuclear energy. Characteristics of small LBE cooled CANDLE fast reactor developed in this Institute are discussed for these requirements. It satisfies clearly four requirements; safety, nonproliferation and safeguard, less wastes and sustainability. For the remaining requirement, economy, a high potential to satisfy this requirement is also shown

  16. Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

    International Nuclear Information System (INIS)

    ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

    2017-01-01

    The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

  17. The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    Science.gov (United States)

    Cole, Christopher J. P.

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas

  18. The near boiling reactor : conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    International Nuclear Information System (INIS)

    Cole, C.J.P.

    2005-01-01

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the 'Victoria' Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96 o C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional

  19. Safe design and operation of tank reactors for multiple-reaction networks: uniqueness and multiplicity

    NARCIS (Netherlands)

    Westerterp, K.R.; Westerink, E.J.

    1990-01-01

    A method is developed to design a tank reactor in which a network of reactions is carried out. The network is a combination of parallel and consecutive reactions. The method ensures unique operation. Dimensionless groups are used which are either representative of properties of the reaction system

  20. New fuels for proliferation-safe research reactors. SCK-CEN a kingpin in international research

    International Nuclear Information System (INIS)

    Van den Berghe, S.

    2011-01-01

    A number of research reactors in the world, including BR2, are still working on highly enriched proliferation sensitive nuclear fuel. Extremely high-density fuel is required to make possible the switch to low-enriched fuel without loss of effectiveness. In the past few years, SCK-CEN has developed into the global test station for new, low-enriched fuels.

  1. 75 FR 8411 - Office of New Reactors: Interim Staff Guidance on Assessing the Consequences of an Accidental...

    Science.gov (United States)

    2010-02-24

    ... the public will be better served by being able to review and comment on both documents at this time... Construction Inspection and Operational Programs, Office of New Reactors, U.S. Nuclear Regulatory Commission... conditions for such releases and define acceptable assumptions to describe exposure scenarios and pathways to...

  2. Stylized whole-core benchmark of the Integral Inherently Safe Light Water Reactor (I2S-LWR) concept

    International Nuclear Information System (INIS)

    Hon, Ryan; Kooreman, Gabriel; Rahnema, Farzad; Petrovic, Bojan

    2017-01-01

    Highlights: • A stylized benchmark specification of the I2S-LWR core. • A library of cross sections were generated in both 8 and 47 groups. • Monte Carlo solutions generated for the 8 group library using MCNP5. • Cross sections and pin fission densities provided in journal’s repository. - Abstract: The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) is a pressurized water reactor (PWR) concept under development by a multi-institutional team led by Georgia Tech. The core is similar in size to small 2-loop PWRs while having the power level of current large reactors (∼1000 MWe) but using uranium silicide fuel and advanced stainless steel cladding. A stylized benchmark specification of the I 2 S-LWR core has been developed in order to test whole-core neutronics codes and methods. For simplification the core was split into 57 distinct material regions for cross section generation. Cross sections were generated using the lattice physics code HELIOS version 1.10 in both 8 and 47 groups. Monte Carlo solutions, including eigenvalue and pin fission densities, were generated for the 8 group library using MCNP5. Due to space limitations in this paper, the full cross section library and normalized pin fission density results are provided in the journal’s electronic repository.

  3. On the selfacting safe limitation of fission power and fuel temperature in innovative nuclear reactors

    International Nuclear Information System (INIS)

    Scherer, W.; Brockmann, H.; Drecker, S.; Gerwin, H.; Haas, K.A.; Kugeler, K.; Ohlig, U.; Ruetten, H.J.; Teuchert, E.; Werner, H.; Wolf, L.

    1994-08-01

    Nuclear energy probably will not contribute significantly to the future worldwide energy supply until it can be made catastrophe-free. Therefore it has to be shown, that the consequences of even largest accidents will have no major impact to the environment of a power plant. In this paper one of the basic conditions for such a nuclear technology is discussed. Using mainly the modular pebble-bed high-temperature reactor as an example, the design principles, analytical methods and the level of knowledge as given today in controlling reactivity accidents by inherent safety features of innovative nuclear reactors are described. Complementary possibilities are shown to reach this goal with systems of different types of construction. Questions open today and resulting requirements for future activities are discussed. Today's knowledge credibly supports the possibility of a catastrophe-free nuclear technology with respect to reactivity events. (orig.)

  4. Water treatment for the ISER [intrinsically safe and economical reactor] plant

    International Nuclear Information System (INIS)

    Sugawara, Ichiro.

    1985-01-01

    The ISER reactor assures inherent safety by causing the core, which is submerged in pool water containing a high boric acid concentration, to quickly shut down the nuclear reaction when overheating, pump trip or other problems occur. However, large quantities of pool water may cause difficulties in water quality control and waste management, resulting in higher costs. Therefore, the ISER as a total plant would not be publicly acceptable unless the water treatment and waste management system offer both safety balanced with reactor inherent safety, and economy counterbalanced by large quantities of pool water. This report clarifies the passive safety concept of possible waste treatment and management systems, and the ways to economically construct such facilities

  5. An assessment of the low seismic risk of the inherently safe sodium advanced fast reactor (SAFR)

    International Nuclear Information System (INIS)

    Rutherford, P.D.

    1988-01-01

    A recent probabilistic risk assessment (PRA) of the sodium advanced fast reactor (SAFR) demonstrated the inherently low risk of advanced liquid-metal, pool-type fast reactors with inherent safety systems. As a result, it was recognized that external events, especially seismic events, may not only be a major contributor to risk (as shown in several LWR PRAs) but also may completely dominate the risk. Accordingly, a seismic risk assessment has been completed for SAFR, which resulted in a core damage frequency of 2 x 10 -7 /year and a large release frequency of 4 x 10 -9 /year. This paper reports that public health risk in terms of early fatality risk and latent fatality risk were also several orders of magnitude below the NRC safety goals and below recent LWR risks reported in NUREB/CR1150

  6. Feasibility study on commercialization of fast breeder reactor cycle system. Interim report of phase 2. Technical study report on synthetic evaluation for FBR cycle

    International Nuclear Information System (INIS)

    Shiotani, Hiroki; Ohtaki, Akira; Ono, Kiyoshi; Yasumatsu, Naoto; Kubota, Sadae; Heta, Masanori

    2004-09-01

    This report presents the outline of the development and the results of Synthetic evaluation on the candidate Fast Reactor (FR) cycle system concepts, scenario study on FR cycle deployment and cost-benefit analysis on the candidate FR cycle system concepts in the interim evaluation (FY2001 through FY2003) of the phase 2 of the Japanese 'Feasibility Study on Commercialization of Fast Reactor Cycle System (FS)'. The characteristic evaluation extended to evaluate a new view point of social acceptance besides the viewpoints of safety, economics, reduction of environmental burden, efficient utilization of uranium resource, proliferation resistance, and technical feasibility, which has been considered since the phase 1 of FS. As for the six view points, hierarchy structures and utility functions for quantitative evaluation have been developed and/or improved. Furthermore, the methodology for weighing the viewpoints, which was also developed, made it possible to examine the characteristics of the candidate concepts from all the seven viewpoints. Generally, the FR cycles with sodium-cooled FR were highly evaluated. The characteristic evaluation for alternative power supply systems was also tried in this report for the first time. FR cycle deployment scenarios clarified the necessity of FR cycle deployment and the desirable core features, etc. through the long-term mass flow analysis, which includes comparison among other nuclear fuel cycle schemes and analysis for evaluating the degree to meet future needs, on the typical FR cycle systems. Regarding cost-benefit analysis, both the amount of the cost estimated by the past R and D and the cost in the Road map of FS are used as the investment for FR cycle research and development (R and D), the results showed that the benefit derived from the commercialization of FR cycle will be more than the investment. (author)

  7. Regulatory Considerations for the Long Term Cooling Safe Shutdown Requirements of the Passive Residual Heat Removal Systems in Advanced Reactors

    International Nuclear Information System (INIS)

    Sim, S. K.; Bae, S. H.; Kim, Y. S.; Hwang, Min Jeong; Bang, Young Seok; Hwang, Taesuk

    2016-01-01

    USNRC approved safe shutdown at 215.6 .deg. C for a safe and long term cooling state for the redundant passive RHRSs by SECY-94-084. USNRC issued COLA(Combined Construction and Operating License) for the Levy County NP Unit-1/2 for the AP1000 passive RHRSs in 2014. Korea Hydro and Nuclear Power(KHNP) is developing APR+ and adopted Passive Auxiliary Feedwater System(PAFS) as a new passive RHRS design. Korea Institute of Nuclear Safety(KINS) has been developing regulatory guides for the advanced safety design features of the advanced ALWRs which has plan to construct in near future in Korea[5]. Safety and regulatory issues as well as the safe shut down requirements of the passive RHRS are discussed and considerations in developing regulatory guides for the passive RHRS are presented herein. Passive RHRSs have been introduced as new safety design features for the advanced reactors under development in Korea. These passive RHRSs have potential advantages over existing active RHRS, however, their functions are limited due to inherent ability of passive heat removal processes. It is high time to evaluate the performance of the passive PRHRs and develop regulatory guides for the safety as well as the performance analyses of the passive RHRS

  8. Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Balogun, G.I.; Jonah, S.A.; Umar, I.M.

    2005-01-01

    Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)

  9. Realization of safety culture into a reactor plant-4S (super safe, small and simple) LMR

    International Nuclear Information System (INIS)

    Hattori, S.; Ikemoto, I.; Minato, A.

    1998-01-01

    International Nuclear Safety Advisory Group (INSAG) defines Safety Culture as the following; Safety Culture is that assembly of characteristics and attitudes in organization and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance. While such Safety Culture is certainly a critical element of nuclear safety assurance, it is important to design nuclear power facilities as friendly to operators as possible with minimum dependence on human factors. From the viewpoint of ensuring supply in our global society, it will be necessary to have multiple approaches to further promote the use of nuclear energy worldwide despite various social and cultural restrictions. It should then be considered, as prospective options, to disperse small nuclear power plants throughout the world under technical, social and cultural conditions. Under this circumstance, we have quested for and now propose a a scheme of assuring sheer safety of nuclear power plants by implementing operator-friendly nuclear reactors virtually free from human errors. The scheme specifically includes the measures for improving reliability through fabricating more compact reactors with a continuous in-factory production system, simplify maintenance and inspection of the reactor system using passive systems and further relieving operators of burden of labors. (author)

  10. Interim report on core physics and fuel cycle analysis of the pebble bed reactor power plant concept

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1977-12-01

    Calculations were made to predict the performance of a pebble bed reactor operated in a mode to produce fissile fuel (high conversion or breeding). Both a one pebble design and a design involving large primary feed pebbles and small fertile pebbles were considered. A relatively short residence time of the primary pebbles loaded with 233 U fuel was found to be necessary to achieve a high breeding ratio, but this leads to relatively high fuel costs. A high fissile inventory is associated with a low C/Th ratio and a high thorium loading, causing the doubling time to be long, even though the breeding ratio is high, and the fuel cost of electrical product to be high. Production of 233 U fuel from 235 U feed was studied and performances of the converter and breeder reactor concepts were examined varying the key parameters

  11. PRISM [Power Reactor Inherently Safe Module] design concept enhances waste management

    International Nuclear Information System (INIS)

    Thompson, M.L.; Berglund, R.C.

    1989-01-01

    PRISM, a modular advanced liquid metal reactor (ALMR), has been designed conceptually by GE under the US Department of Energy sponsorship. The concept design and analyses have been primarily focused on passive safety and improved construction and operating costs. Significantly, the unique design of multiple modules and features of PRISM enhance waste management over conventional reactor systems. This paper provides an overview of PRISM of these enhancements. Inherent to the ALMR's, the sodium coolant precludes crud buildup on reactor surfaces and in components and waste for disposal. Preliminary evaluations indicate this fundamental feature results in factors of 2-4 less waste volume and 2-3 orders of magnitude less curies per megawatt-electric for ultimate disposal. For example, the tap designed for sodium cleanup is expected to be exchanged only once every thirty years. Also, inherent to ALMR's, burning waste actinides and selected fission products to preclude their accumulation and burial is very attractive. The hard neutron spectrum of ALMR burns the actinides efficiently and is not poisoned by the actinides and fission products. The modular design of PRISM components (and the fuel cycle equipment) permit replacement without expensive and potentially hazardous volume reduction. For example, the functional components of the reference electromagnetic pump and IHK can be removed intact for waste disposal. Although development of the reference metal fuel is not completed, it is estimated that (low-level) waste from recycle of the fuel will result in significantly less volume than would be generated by aqueous recycle of oxide fuel. 6 refs., 10 figs

  12. A safe private nuclear tool-the miniature neutron source reactor

    International Nuclear Information System (INIS)

    Zhou Yongmao

    1987-01-01

    The prototype miniature neutron source reactor (MNSR) designed by China Institute of Atomic Energy has been operated successfully for more than 3 years and the practical experience enriches the original design idea. The commercial MNSR is under study design and develop in following aspects: 1. Prolonging the control rod cycle duration and core burn-up life; 2. Increasing the neutron flux per unit power. Obviously, the MNSR will show more advantages in extending application area and in providing the core using low envichment fuel. (Liu)

  13. Thermal conductivity of beginning-of-life uranium-plutonium mixed oxide fuel for fast reactor (Interim report)

    International Nuclear Information System (INIS)

    Inoue, Masaki; Mizuno, Tomoyasu; Asaga, Takeo

    1997-11-01

    Thermal conductivity of uranium-plutonium mixed oxide fuel for fast reactor at beginning-of-life was correlated based on the recent results in order to apply to the fuel design and the fuel performance analysis. A number of experimental results of unirradiated fuel specimens were corrected from open literatures and PNC internal reports and examined for the database. In this work two porosity correction factors were needed for high density fuel and low density fuel (around the current Monju specification). The universal porosity correction factor was not determined in this work. In the next step, theoretical and analytical considerations should be taken into account. (J.P.N.)

  14. Gas reactor international cooperative program interim report: United States/Federal Republic of Germany nuclear licensing comparison

    International Nuclear Information System (INIS)

    1978-09-01

    In order to compare US and FRG Nuclear Licensing, a summary description of United States Nuclear Licensing is provided as a basis. This is followed by detailed information on the participants in the Nuclear Licensing process in the Federal Republic of Germany (FRG). FRG licensing procedures are described and the rules and regulations imposed are summarized. The status of gas reactor licensing in both the U.S. and the FRG is outlined and overall conclusions are drawn as to the major licensing differences. An appendix describes the most important technical differences between US and FRG criteria

  15. Fuel Supply Shutdown Facility Interim Operational Safety Requirements

    International Nuclear Information System (INIS)

    BENECKE, M.W.

    2000-01-01

    The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility

  16. Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State

    2016-11-30

    The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup

  17. The gas turbine modular helium reactor. An international project to develop a safe, efficient, flexible product

    International Nuclear Information System (INIS)

    Silberstein, A.J.

    1998-01-01

    As originally scheduled, the Conceptual Design Report of the 600 Mwt Gas Turbine Modular Helium Reactor has been issued in October 1997 by OKBM in Nizhny Novgorod, a keystone Russian Engineering Institute fully involved in the realization of this International Project. The plutonium burning, graphite moderated helium cooled reactor design results from the work done on the basis of General Atomics original concept combined with the goal of optimizing safety power and efficiency with multi contributions in specific fields from the Russian organizations: MINATOM, OKBM, VNIINM, Lutch, Kurchatov Institute, Seversk Chemical Combinat, Fuji Electric and FRAMATOME. The objective to concentrate the engineering work in Russia has met a full success due principally to the quality and experience of the people, to the international support and to the progressive integration of new techniques of communication, of project management culture and utilization of modern computerized design tools and methods. To day the best international standard of quality is reached in the engineering activity and expected to stay at this level for future developments, when including experimental facilities operation and components manufacturing activities, thanks to the diffusion of the common culture, acquired by the main actors during the conceptual design phase, that will be exported to Russian third parties. At this stage we are planning to start design verification and sensitive components and systems qualification, with the same original actors. The European Commission has already shown some significant interest through the MICHELANGELO Initiative in supporting the HTR concepts assessment and identification of the R and D needs. We are looking forward for further support from the International Community and particularly from European Institutions in the frame of the 5th PCRD to pursue the GT MHR R and D program. Furthermore we are looking for funding the building of a prototype in Russia

  18. Feedback control of a primary pump for safe and stable operation of a PIUS-type reactor

    International Nuclear Information System (INIS)

    Tasaka, K.; Imai, S.; Masaoka, H.; Tamaki, M.; Kukita, Y.

    1993-01-01

    A new automatic pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the density lock in order to prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments such as start-up and power ramping tests for the stable normal operation and a loss-of-feedwater test for the safe shutdown in an accident condition, using a small scale atmospheric pressure test loop which simulated the PIUS principle. (orig.)

  19. Thermal-hydraulic experiment for safe and stable operation of a PIUS-type reactor

    International Nuclear Information System (INIS)

    Tasaka, K.; Imai, S.; Masaoka, H.; Irianto, I.D.; Kohketsu, H.; Tamaki, M.; Anoda, Y.; Murata, H.; Kukita, Y.

    1992-01-01

    A new automatic pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed for the PIUS-type reactor. This control system maintains the fluid temperature at the axial center of the lower density lock at the average of the fluid temperatures below and above the density lock in order to prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments such as start-up and power ramping tests for normal operation simulation and a loss of feedwater test for an accident condition simulation, using a small scale atmospheric pressure test loop which simulated the PIUS principle. (author)

  20. Analytical and experimental justification of safe operation of fuel loads of VVER reactors at Rovno NPP

    International Nuclear Information System (INIS)

    Andrianov, A.; Zagrebelny, L.

    2011-01-01

    The main task during the nuclear fuel operation us ensuring of the larger fuel burnup and as a result - reduction of the fuel constituent in the electricity production cost. The neutron-physic calculations are performed using the qualified codes BIPR-7A and PERMAK developed by Kurchatov Institute. The limits for calculated parameters are set by the Ukrainian regulations. Calculation results are documented in reports subject for independent expert review requested by the regulatory authority. In this report the following item have presented: 1) metrological check and calibration of measuring channels; 2) fuel cycles at Rivne NPP; 3) determination of experimental values of thermal-physic and neutron-physic parameters; 4) ICIS equipment check, execution of the program confirming correct connection of the temperature and neutron flux monitoring sensors; 5) monitoring of the core TPh and NPh parameters in all operating modes; 6) monitoring of the fuel condition in the core and 7) FE leak tightness monitoring at the operating reactor

  1. Fail-safe design criteria for computer-based reactor protection systems

    International Nuclear Information System (INIS)

    Keats, A.B.

    1980-01-01

    The increasing quantity and complexity of the instrumentation required in nuclear power plants provides a strong incentive for using on-line computers as the basis of the control and protection systems. On-line computers using multiplexed sampled data are already well established but their application to nuclear reactor protection systems requires special measures to satisfy the very high reliability which is demanded in the interests of safety and availability. Some existing codes of practice relating to segregation of replicated subsysttems continue to be applicable and lead to division of the computer functions into two distinct parts. The first computer, referred to as the Trip Algorithm Computer may also control the multiplexer. Voting on each group of status inputs yielded by the trip algorithm computers is performed by the Vote Algorithm Computer. The conceptual disparities between hardwired reactor-protection systems and those employing computers also rise to a need for some new criteria. An important objective of these criteria, minimising the need for a failure-mode-and-effect-analysis of the computer software, but is achieved almost entirely by 'hardware' properties of the system: the systematic use of hardwired test inputs which cause excursions of the trip algorithms into the tripped state in a uniquely ordered but easily recognisable sequence, and the use of hardwired 'pattern recognition logic' which generates a dynamic 'healthy' stimulus for the shutdown actuators only in response to the unique sequence generated by the hardwired input signal pattern. The adoption of the proposed design criteria ensure not only failure-to-safety in the hardware but the elimination, or at least minimisation, of the dependence on the correct functioning of the computer software for the safety system. (auth)

  2. The Nord interim store

    International Nuclear Information System (INIS)

    Leushacke, D.F.; Rittscher, D.

    1996-01-01

    In line with the decision taken in 1990 to shut down and decommission the Greifswald and Rheinsberg Nuclear Power Stations, the waste management concept of the Energiewerke Nord is based on direct and complete decommissioning of the six shut down reactor units within the next fifteen years. One key element of this concept is the construction and use of the Zwischenlager Nord (Nord Interim Store, ZLN) for holding the existing nuclear fuels and for interim and decay storage of the radioactive materials arising in decommissioning and demolition. The owner and operator of the store is Energiewerke Nord GmbH. The interim store has the functions of a processing and Energiewerke Nord GmbH. The interim store has the functions of a processing and treatment station and buffer store for the flows of residues arising. As a radioactive waste management station, it accommodates nuclear fuels, radioactive waste or residues which are not treated any further. It is used as a buffer store to allow the materials accumulating in disassembly to be stored temporarily before or after treatment in order to ensure continuous loading of the treatment plants. When operated as a processing station, the ZLN is able to handle nearly all types of radioactive waste and residues arising, except for nuclear fuels. These installations allow the treatment of radioactive residues to be separated from the demolition work both physically and in time. The possibilities of interium storage and buffer storage of untreated waste and waste packages make for high flexibility in logistics and waste management strategy. (orig.) [de

  3. Investigation of natural circulation instability and transients in passively safe novel modular reactor

    Science.gov (United States)

    Shi, Shanbin

    The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

  4. A completely automatic operation type super-safe fast reactor, RAPID. Its application to dispersion source on lunar and earth surfaces

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Kawasaki, Akira; Iwamura, Takamichi

    2002-01-01

    At a viewpoint of flexible measures to future electric power demands, expectation onto a small-scale reactor for dispersion source is increasing gradually. This is thought to increase its importance not only for a source at proximity of its market in advanced nations but also for the one in developing nations. A study on development of the completely automatic operation type super-safe fast reactor, RAPID (refueling by all pins integrated design) has been carried out as a part of the nuclear energy basic research promoting system under three years project since 1999 by a trust of the Japan Atomic Energy Research Institute to a group of the Central Research Institute of Electric Power Industry (CRIEPI) and so on. As the reactor is a lithium cooled fast reactor with 200 Kw of electric output supposing to use at lunar surface, it can be applied to a super-small scale nuclear reactor on the earth, and has feasibility to become a new option of future nuclear power generation. On the other hand, CRIEPI has investigated on various types of fast reactors (RAPID series) for fast reactor for dispersion source on the earth. Here was introduced on such super-safe fast reactors at a center of RAPID-L. (G.K.)

  5. Request for interim approval to operate Trench 94 of the 218-E-12B Burial Ground as a chemical waste landfill for disposal of polychlorinated biphenyl waste in submarine reactor compartments. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Cummins, G.D.

    1994-06-01

    This request is submitted to seek interim approval to operate a Toxic Substances Control Act (TSCA) of 1976 chemical waste landfill for the disposal of polychlorinated biphenyl (PCB) waste. Operation of a chemical waste landfill for disposal of PCB waste is subject to the TSCA regulations of 40 CFR 761. Interim approval is requested for a period not to exceed 5 years from the date of approval. This request covers only the disposal of small 10 quantities of solid PCB waste contained in decommissioned, defueled submarine reactor compartments (SRC). In addition, the request applies only to disposal 12 of this waste in Trench 94 of the 218-E-12B Burial Ground (Trench 94) in the 13 200 East Area of the US Department of Energy`s (DOE) Hanford Facility. Disposal of this waste will be conducted in accordance with the Compliance 15 Agreement (Appendix H) between the DOE Richland Operations Office (DOE-RL) and 16 the US Environmental Protection Agency (EPA), Region 10. During the 5-year interim approval period, the DOE-RL will submit an application seeking final 18 approval for operation of Trench 94 as a chemical waste landfill, including 19 any necessary waivers, and also will seek a final dangerous waste permit from 20 the Washington State Department of Ecology (Ecology) for disposal of lead 21 shielding contained in the SRCS.

  6. Request for interim approval to operate Trench 94 of the 218-E-12B Burial Ground as a chemical waste landfill for disposal of polychlorinated biphenyl waste in submarine reactor compartments

    International Nuclear Information System (INIS)

    Cummins, G.D.

    1994-06-01

    This request is submitted to seek interim approval to operate a Toxic Substances Control Act (TSCA) of 1976 chemical waste landfill for the disposal of polychlorinated biphenyl (PCB) waste. Operation of a chemical waste landfill for disposal of PCB waste is subject to the TSCA regulations of 40 CFR 761. Interim approval is requested for a period not to exceed 5 years from the date of approval. This request covers only the disposal of small 10 quantities of solid PCB waste contained in decommissioned, defueled submarine reactor compartments (SRC). In addition, the request applies only to disposal 12 of this waste in Trench 94 of the 218-E-12B Burial Ground (Trench 94) in the 13 200 East Area of the US Department of Energy's (DOE) Hanford Facility. Disposal of this waste will be conducted in accordance with the Compliance 15 Agreement (Appendix H) between the DOE Richland Operations Office (DOE-RL) and 16 the US Environmental Protection Agency (EPA), Region 10. During the 5-year interim approval period, the DOE-RL will submit an application seeking final 18 approval for operation of Trench 94 as a chemical waste landfill, including 19 any necessary waivers, and also will seek a final dangerous waste permit from 20 the Washington State Department of Ecology (Ecology) for disposal of lead 21 shielding contained in the SRCS

  7. Feasibility study on commercialization of fast breeder reactor cycle systems interim report of phase II. Technical study report for nuclear fuel cycle systems

    International Nuclear Information System (INIS)

    Sato, Koji; Amamoto, Ippei; Inoue, Akira

    2004-06-01

    As a part of the feasibility study on commercialization of fast breeder reactor cycle systems, the plant concept concerning the fuel cycle systems (combination of the reprocessing and the fuel fabrication) has been constructed to reduce their total cost by the introduction of various innovative techniques and to apply their utmost superior efficiency from such standpoints of a decrease in the environmental burden, better resource utilization and proliferation resistance improvement by the low decontamination transuranium element (TRU) recycle. This interim report of Phase II describes the results of an on-going study which will cover a five-year period. For oxide fuels, the system which combines the use of the advanced aqueous reprocessing using three main methods such as the crystallization method, the simplified solvent extraction method, and the extraction chromatography method for minor actinide (MA) recovery, as well as the simplified pelletizing fuel fabrication which rationalized a powder mixing process etc., has abundant current results and a high technical feasibility for the basic process. Though this system faces difficulties in the technical development of control technology of the extraction chromatography and the fabrication technology of low decontamination TRU fuel etc., its expected practical use is possible at an early stage. As for the super-critical direct extraction reprocessing, it is necessary to fulfill more basic data although further economical improvement of an advanced aqueous reprocessing is expected. The system which combines the advanced aqueous reprocessing and the gelation sphere packing fuel fabrication has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the fuel fabrication process. However, this system will shoulder additional cost for the reagent recovery process and the waste liquid treatment process due to need to dispose of a large bulk of process waste liquid. The system which

  8. Opening Address [International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), Paris, France, March 4-7, 2013

    International Nuclear Information System (INIS)

    Amano, Yukiya

    2013-01-01

    Public confidence in nuclear power was greatly shaken by the Fukushima Daiichi accident. It will take time to rebuild that confidence. This will only be possible if everyone involved in nuclear power has a total commitment to safety and if the sector is open and transparent. The public need to be reassured that nuclear energy is efficient and safe, can mitigate the effects of climate change and can play a key role in meeting the growing global demand for energy. Fast reactors and related fuel cycles will be important for the long-term sustainability of nuclear power. This innovative technology has the potential to ensure that energy resources which would run out in a few hundred years, using today’s technology, will actually last several thousand years. Fast reactors also reduce the volume and toxicity of the final waste. China’s Experimental Fast Reactor has been connected to the grid. Work is at an advanced stage on construction of India’s 500 MW(e) Prototype Fast Breeder Reactor and of the large BN-800 reactor in the Russian Federation. Interest in fast reactors with closed fuel cycles is increasing steadily. A number of emerging economies are joining the existing fast reactor technology-holders. Considerable R & D work is being done on advanced designs with enhanced safety characteristics. It is important to gather the operational experience of countries with operating fast reactors and related fuel cycle facilities. This can help to achieve higher levels of safety. Events such as the Joint GIF-IAEA Workshop on the safety of sodium-cooled fast reactors last week are a useful way of doing this. They also help to ensure that relevant lessons from the Fukushima Daiichi accident are learned. The IAEA remains the unique collaboration forum for ensuring continued progress in fast reactor technology. We provide an umbrella for knowledge preservation, information exchange and collaborative R&D in which resources and expertise are pooled

  9. Interim report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-06-01

    This Interim Report summarizes the research and development activities of the Superconducting Super Collider project carried out from the completion of the Reference Designs Study (May 1984) to June 1985. It was prepared by the SSC Central Design Group in draft form on the occasion of the DOE Annual Review, June 19--21, 1985. Now largely organized by CDG Divisions, the bulk of each chapter documents the progress and accomplishments to date, while the final section(s) describe plans for future work. Chapter 1, Introduction, provides a basic brief description of the SSC, its physics justification, its origins, and the R&D organization set up to carry out the work. Chapter 2 gives a summary of the main results of the R&D program, the tasks assigned to the four magnet R&D centers, and an overview of the future plans. The reader wishing a quick look at the SSC Phase I effort can skim Chapter 1 and read Chapter 2. Subsequent chapters discuss in more detail the activities on accelerator physics, accelerator systems, magnets and cryostats, injector, detector R&D, conventional facilities, and project planning and management. The magnet chapter (5) documents in text and photographs the impressive progress in successful construction of many model magnets, the development of cryostats with low heat leaks, and the improvement in current-carrying capacity of superconducting strand. Chapter 9 contains the budgets and schedules of the COG Divisions, the overall R&D program, including the laboratories, and also preliminary projections for construction. Appendices provide information on the various panels, task forces and workshops held by the CDG in FY 1985, a bibliography of COG and Laboratory reports on SSC and SSC-related work, and on private industrial involvement in the project.

  10. Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

    International Nuclear Information System (INIS)

    Wittenbrock, N.G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

  11. ITER Conceptual design: Interim report

    International Nuclear Information System (INIS)

    1990-01-01

    This interim report describes the results of the International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activities after the first year of design following the selection of the ITER concept in the autumn of 1988. Using the concept definition as the basis for conceptual design, the Design Phase has been underway since October 1988, and will be completed at the end of 1990, at which time a final report will be issued. This interim report includes an executive summary of ITER activities, a description of the ITER device and facility, an operation and research program summary, and a description of the physics and engineering design bases. Included are preliminary cost estimates and schedule for completion of the project

  12. Materials behavior in interim storage of spent fuel

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Bailey, W.J.; Gilbert, E.R.; Inman, S.C.

    1982-01-01

    Interim storage has emerged as the only current spent-fuel management method in the US and is essential in all countries with nuclear reactors. Materials behavior is a key aspect in licensing interim-storage facilities for several decades of spent-fuel storage. This paper reviews materials behavior in wet storage, which is licensed for light-water reactor (LWR) fuel, and dry storage, for which a licensing position for LWR fuel is developing

  13. Quality assurance plan for the Molten Salt Reactor Experiment Remediation Project at the Oak Ridge National Laboratory. Phase 1 -- Interim corrective measures and Phase 2 -- Purge and trap reactive gases

    International Nuclear Information System (INIS)

    1995-11-01

    This Quality Assurance Plan (QAP) identifies and describes the systems utilized by the Molten Salt Reactor Experiment Remediation Project (MSRERP) personnel to implement the requirements and associated applicable guidance contained in the Quality Program Description Y/QD-15 Rev. 2 (Energy Systems 1995f). This QAP defines the quality assurance (QA) requirements applicable to all activities and operations in and directly pertinent to the MSRERP Phase 1--Interim Corrective Measures and Phase 2--Purge and Trap objectives. This QAP will be reviewed, revised, and approved as necessary for Phase 3 and Phase 4 activities. This QAP identifies and describes the QA activities and procedures implemented by the various Oak Ridge National Laboratory support organizations and personnel to provide confidence that these activities meet the requirements of this project. Specific support organization (Division) quality requirements, including the degree of implementation of each, are contained in the appendixes of this plan

  14. Spent fuel interim management: 1995 update

    International Nuclear Information System (INIS)

    Anderson, C.K.

    1995-01-01

    The problems of interim away-from-reactor spent fuel storage and storage in spent fuel pools at the reactor site are discussed. An overview of the state-of-the-art in the USA, Europe, and Japan is presented. The technical facilities for away-from-reactor storage are briefly described, including wet storage pools, interactive concrete systems, metallic containers, and passive concrete systems. Reprocessing technologies are mostly at the design stage only. It is predicted that during the 20 years to come, about 50 000 tonnes of spent fuel will be stored at reactor sites regardless of the advance of spent fuel reprocessing or interim storage projects. (J.B.). 4 tabs., 2 figs

  15. Actions for continued safe wet storage of spent nuclear fuel at VVR-S reactor in Bucharest-Magurele

    International Nuclear Information System (INIS)

    Isbasescu, M.; Zorliu, A.; Silviu-laurentiu, B.; Stefan, V. . E-mail address of corresponding author: mirifa@ifin.nipne.ro; Isbasescu, M.)

    2005-01-01

    The Romanian VVR-S research reactor is located 8 kilometers from Bucharest in the town of Magurele and was operated by the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH). The reactor first reached criticality in July 1957 and operated until December 1997 when it was permanently shutdown. The VVR - S reactor of IFIN has two repositories for spent fuel elements: (1) Cooling pool located in the reactor room; (2) Long-term repositories located outside the reactor building - SNFW (spent nuclear fuel warehouse). The major factors believed to influence the pitting of aluminium alloys are conductivity, pH, and bicarbonate, chloride, sulphate and oxygen content. Some of these parameters have been analyzed at SNFW-IFIN-HH. (author)

  16. Design and safety aspect of lead and lead-bismuth cooled long-life small safe fast reactors for various core configurations

    International Nuclear Information System (INIS)

    Zaki, S.; Sekimoto, Hiroshi

    1995-01-01

    Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ∼2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25-40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient. (author)

  17. Fusion Breeder Program interim report

    International Nuclear Information System (INIS)

    Moir, R.; Lee, J.D.; Neef, W.

    1982-01-01

    This interim report for the FY82 Fusion Breeder Program covers work performed during the scoping phase of the study, December, 1981-February 1982. The goals for the FY82 study are the identification and development of a reference blanket concept using the fission suppression concept and the definition of a development plan to further the fusion breeder application. The context of the study is the tandem mirror reactor, but emphasis is placed upon blanket engineering. A tokamak driver and blanket concept will be selected and studied in more detail during FY83

  18. 46th Annual meeting on nuclear technology (AMNT 2015). Key topic / Enhanced safety and operation excellence / Sustainable reactor operation management - safe, efficient, valuable

    International Nuclear Information System (INIS)

    Fischer, Erwin

    2015-01-01

    Summary report on the following Topical Session of the 46 th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015: - Sustainable Reactor Operation Management - Safe, Efficient, Valuable (Erwin Fischer) The other Sessions of the Key Topics - ''Outstanding Know-How and Sustainable Innovations'', - ''Enhanced Safety and Operation Excellence'' and - ''Decommissioning Experience and Waste Management Solutions'' have been covered in atw 7 (2015) and will be covered in further issues of atw.

  19. Aspects of reactor dismantling planning following the safe entombment in the NPP Lingen (KWL); Aspekte der Abbauplanung nach dem Sicheren Einschluss im Kernkraftwerk Lingen (KWL)

    Energy Technology Data Exchange (ETDEWEB)

    Priesmeyer, U.; Rojahn, T.; Fries, B. [Kernkraftwerk Lingen GmbH (Germany)

    2009-07-01

    The NPP Lingen (KWL) was shut-down in 1977. Due to the fact that no final repository was available the safe entombment for 25 years was chosen following the decommissioning. The conventional plant components were dismantled and removed from the plant site. The licensing procedure for reactor dismantling with final disposal in Schacht Konrad has been started. The beginning of dismantling operation is scheduled for 2013. The authors describe the preparatory work, the boundary conditions for the dismantling, radiation protection considerations with respect to manual demolition work after the rather long decay time.

  20. 46{sup th} Annual meeting on nuclear technology (AMNT 2015). Key topic / Enhanced safety and operation excellence / Sustainable reactor operation management - safe, efficient, valuable

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Erwin [E.ON Kernkraft GmbH, Global Unit Next Generation, Hannover (Germany)

    2015-08-15

    Summary report on the following Topical Session of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015: - Sustainable Reactor Operation Management - Safe, Efficient, Valuable (Erwin Fischer) The other Sessions of the Key Topics - ''Outstanding Know-How and Sustainable Innovations'', - ''Enhanced Safety and Operation Excellence'' and - ''Decommissioning Experience and Waste Management Solutions'' have been covered in atw 7 (2015) and will be covered in further issues of atw.

  1. Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations. Interim report; Weiterentwicklung moderner Verfahren zu Neutronentransport und Unsicherheitsanalysen fuer Kernberechnungen. Zwischenbericht

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, Winfried; Aures, Alexander; Bostelmann, Friederike; Pasichnyk, Ihor; Perin, Yann; Velkov, Kiril; Zilly, Matias

    2016-12-15

    This report documents the status of the research and development goals reached within the reactor safety research project RS1536 ''Development of modern methods with respect to neutron transport and uncertainty analyses for reactor core calculations'' as of the 3{sup rd} quarter of 2016. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts, in particular fast reactors cooled by liquid metal. The contributing individual goals are the further optimization and validation of deterministic calculation methods with high spatial and energy resolution, the development of a coupled calculation system using the Monte Carlo method for the neutron transport to describe time-dependent reactor core states, the processing and validation of nuclear data, particularly with regard to covariance data, the development, validation, and application of sampling-based methods for uncertainty and sensitivity analyses, the creation of a platform for performing systematic uncertainty analyses for fast reactor systems, as well as the description of states of severe core damage with the Monte Carlo method. Moreover, work regarding the European NURESAFE project, started in the preceding project RS1503, are being continued and completed.

  2. Requirements on fuel management for a safe and optimum operation of the German research reactor FRJ-2

    International Nuclear Information System (INIS)

    Nabbi, R.; Bormann, H.J.; Wolters, J.

    1997-01-01

    In case of a coarse control arm (CCAs) at FRJ-2, reactivity is added to the reactor. The amount of this reactivity depends on the efficiency of the individual CCAs which has been measured as 180% of the average reactivity of the six arms for the central arm. For this design basis accident, it is required that only 4 out of 5 residual arms must be capable of shutting down the reactor. This minimum shutdown reactivity is provided by an optimum fuel management including an experimental reactivity determination. Calculation of fuel burnup and material densities is performed by the depletion code SUSAN, which has been verified by separate calculations using ORIGEN. The difference in the reactivity values (between calculation and measurement) is mainly a consequence of the limitation of the inverse kinetic method, which is not capable of covering the effects of the flux deformation and interaction of the CCAs and core in the process of reactor scram. (author)

  3. Experiences from start-up and use of fail-safe reactor level meter with KNITU probes

    International Nuclear Information System (INIS)

    Badiar, S.; Sipka, J.; Vanco, P.; Slanina, Marek; Liska, L.; Gavora, D.

    2001-01-01

    Within the framework of the reconstruction of both V1 Bohunice units with WWER-440/V-230 reactors, the VUJE-SIEMENS consortium implemented measurement systems for coolant level monitoring in reactor pressure vessels with 1E qualification, resp. with the qualification of category 1 according to NUSS RG1.97. The solution uses KNITU 11 probes made by Russian POZIT company. In operating plants, the installation causes problems in relation to the existing technology and quality assurance system. In the phase of implementation, the most important tasks were to resolve component quality, to commission the system, and to check its performance. (Authors)

  4. Safe sex

    Science.gov (United States)

    ... sex; Sexually transmitted - safe sex; GC - safe sex; Gonorrhea - safe sex; Herpes - safe sex; HIV - safe sex; ... contact. STIs include: Chlamydia Genital herpes Genital warts Gonorrhea Hepatitis HIV HPV Syphilis STIs are also called ...

  5. Current status of the first interim spent fuel storage facility in Japan

    International Nuclear Information System (INIS)

    Shinbo, Hitoshi; Kondo, Mitsuru

    2008-01-01

    In Japan, storage of spent fuels outside nuclear power plants was enabled as a result of partial amendments to the Nuclear Reactor Regulation Law in June 2000. Five months later, Mutsu City in Aomori Prefecture asked the Tokyo Electric Power Company (TEPCO) to conduct technical surveys on siting of the interim spent fuel storage facility (we call it 'Recyclable-Fuel Storage Center'). In April 2003, TEPCO submitted the report on siting feasibility examination, concluded that no improper engineering data for siting, construction of the facility will be possible from engineering viewpoint. Siting Activities for publicity and public acceptance have been continued since then. After these activities, Aomori Prefecture and Mutsu City approved siting of the Recyclable Fuel Storage Center in October 2005. Aomori Prefecture, Mutsu City, TEPCO and Japan Atomic Power Company (JAPC) signed an agreement on the interim spent fuel storage Facility. A month later, TEPCO and JAPC established Recyclable-Fuel Storage Company (RFS) in Mutsu City through joint capital investment, specialized in the first interim spent fuel storage Facility in Japan. In May 2007, we made an application for establishment permit, following safety review by regulatory authorities. In March 2008, we started the preparatory construction. RFS will safely store of spent fuels of TEPCO and JAPC until they will be reprocessed. Final storage capacity will be 5,000 ton-U. First we will construct the storage building of 3,000 ton-U to be followed by second building. We aim to start operation by 2010. (author)

  6. Options for the interim storage of spent fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    1995-01-01

    Different concepts for the interim storage of spent fuel arising from operation of a NPP are discussed. We considered at reactor as well as away from reactor storage options. Included are enhancements of existing storage capabilities and construction of a new wet or dry storage facility. (author)

  7. Fuel supply shutdown facility interim operational safety requirements

    International Nuclear Information System (INIS)

    Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

    1995-01-01

    These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance)

  8. HEXEREI: a multi-channel heat conduction convection code for use in transient thermal hydraulic analysis of high-temperature, gas-cooled reactors. Interim report

    International Nuclear Information System (INIS)

    Giles, G.E.; DeVault, R.M.; Turner, W.D.; Becker, B.R.

    1976-05-01

    A description is given of the development and verification of a generalized coupled conduction-convection, multichannel heat transfer computer program to analyze specific safety questions involving high temperature gas-cooled reactors (HTGR). The HEXEREI code was designed to provide steady-state and transient heat transfer analysis of the HTGR active core using a basic hexagonal mesh and multichannel coolant flow. In addition, the core auxiliary cooling systems were included in the code to provide more complete analysis of the reactor system during accidents involving reactor trip and cooling down on the auxiliary systems. Included are brief descriptions of the components of the HEXEREI code and sample HEXEREI analyses compared with analytical solutions and other heat transfer codes

  9. Designing visual displays and system models for safe reactor operations based on the user`s perspective of the system

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.

    1995-12-31

    Most designers are not schooled in the area of human-interaction psychology and therefore tend to rely on the traditional ergonomic aspects of human factors when designing complex human-interactive workstations related to reactor operations. They do not take into account the differences in user information processing behavior and how these behaviors may affect individual and team performance when accessing visual displays or utilizing system models in process and control room areas. Unfortunately, by ignoring the importance of the integration of the user interface at the information process level, the result can be sub-optimization and inherently error- and failure-prone systems. Therefore, to minimize or eliminate failures in human-interactive systems, it is essential that the designers understand how each user`s processing characteristics affects how the user gathers information, and how the user communicates the information to the designer and other users. A different type of approach in achieving this understanding is Neuro Linguistic Programming (NLP). The material presented in this paper is based on two studies involving the design of visual displays, NLP, and the user`s perspective model of a reactor system. The studies involve the methodology known as NLP, and its use in expanding design choices from the user`s ``model of the world,`` in the areas of virtual reality, workstation design, team structure, decision and learning style patterns, safety operations, pattern recognition, and much, much more.

  10. Designing visual displays and system models for safe reactor operations based on the user's perspective of the system

    International Nuclear Information System (INIS)

    Brown-VanHoozer, S.A.

    1995-01-01

    Most designers are not schooled in the area of human-interaction psychology and therefore tend to rely on the traditional ergonomic aspects of human factors when designing complex human-interactive workstations related to reactor operations. They do not take into account the differences in user information processing behavior and how these behaviors may affect individual and team performance when accessing visual displays or utilizing system models in process and control room areas. Unfortunately, by ignoring the importance of the integration of the user interface at the information process level, the result can be sub-optimization and inherently error- and failure-prone systems. Therefore, to minimize or eliminate failures in human-interactive systems, it is essential that the designers understand how each user's processing characteristics affects how the user gathers information, and how the user communicates the information to the designer and other users. A different type of approach in achieving this understanding is Neuro Linguistic Programming (NLP). The material presented in this paper is based on two studies involving the design of visual displays, NLP, and the user's perspective model of a reactor system. The studies involve the methodology known as NLP, and its use in expanding design choices from the user's ''model of the world,'' in the areas of virtual reality, workstation design, team structure, decision and learning style patterns, safety operations, pattern recognition, and much, much more

  11. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  12. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  13. GT-MHR as economical highly efficient inherently safe modular gas cooled reactor for electric power generation

    International Nuclear Information System (INIS)

    Baydakov, A.G.; Kodochigov, N.G.; Kuzavkov, N.G.; Vorontsov, V.E.

    2001-01-01

    The Gas Turbine Modular Helium Reactor (GT-MHR) project is a joint effort of organisations in the Russian Federation co-ordinated by the Ministry on Atomic Energy (MINATOM), General Atomics in the United States, Framatome in France, and Fuji Electric in Japan. This paper describes the structure, objectives and history of the GT-MHR project, as well as the plant design, including plant layout, components and fuel. It summarises the results of evaluations of the design safety characteristics and economics, for both a plutonium fueled single module and a uranium fueled four module plant. The plans and anticipated structure for an international project to construct the plant are also summarised. (author)

  14. Reference values on safety regulation of land disposal of low level radioactive solid waste (the second interim report) and its incorporation into legal regulations

    International Nuclear Information System (INIS)

    Aoki, Terumi

    1994-01-01

    Safety regulation of land disposal of low level radioactive solid waste in Japan is based on 'the basic philosophy on the safety regulation of land disposal of low level radioactive solid waste' determined by the Nuclear safety Committee (October 1985). The basic philosophy on the upper limit of radioactivity of disposed wastes was published as the reference values in the interim report (February 1987) and in the second interim report (June 1992). In the second interim report, the upper limits of radioactivity are established for three types of solid radioactive wastes: 1) metals, incombustible or flame resistant wastes generated nuclear reactor facilities and solidified in vessels, 2) large metallic structures generated from decommissioning of reactor facilities and difficult to solidify in vessels, and 3) radioactive concrete waste generated from decommissioning of reactor facilities. The upper limits of radioactivity are presented for C-14, Co-60, Ni-63, Sr-90, Cs-137, alfa-emmitters, Ca-41 (for concrete) and Eu-152 (for concrete). Related laws and regulations in Japan on safe disposal of low level wastes are explained. (T.H.)

  15. An approach to the verification of a fault-tolerant, computer-based reactor safety system: A case study using automated reasoning: Volume 1: Interim report

    International Nuclear Information System (INIS)

    Chisholm, G.H.; Kljaich, J.; Smith, B.T.; Wojcik, A.S.

    1987-01-01

    The purpose of this project is to explore the feasibility of automating the verification process for computer systems. The intent is to demonstrate that both the software and hardware that comprise the system meet specified availability and reliability criteria, that is, total design analysis. The approach to automation is based upon the use of Automated Reasoning Software developed at Argonne National Laboratory. This approach is herein referred to as formal analysis and is based on previous work on the formal verification of digital hardware designs. Formal analysis represents a rigorous evaluation which is appropriate for system acceptance in critical applications, such as a Reactor Safety System (RSS). This report describes a formal analysis technique in the context of a case study, that is, demonstrates the feasibility of applying formal analysis via application. The case study described is based on the Reactor Safety System (RSS) for the Experimental Breeder Reactor-II (EBR-II). This is a system where high reliability and availability are tantamount to safety. The conceptual design for this case study incorporates a Fault-Tolerant Processor (FTP) for the computer environment. An FTP is a computer which has the ability to produce correct results even in the presence of any single fault. This technology was selected as it provides a computer-based equivalent to the traditional analog based RSSs. This provides a more conservative design constraint than that imposed by the IEEE Standard, Criteria For Protection Systems For Nuclear Power Generating Stations (ANSI N42.7-1972)

  16. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  17. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  18. German concept and status of the disposal of spent fuel elements from German research reactors

    International Nuclear Information System (INIS)

    Komorowski, K.; Storch, S.; Thamm, G.

    1995-01-01

    Eight research reactors with a power ≥ 100 kW are currently being operated in the Federal Republic of Germany. These comprise three TRIGA-type reactors (power 100 kW to 250 kW), four swimming-pool reactors (power 1 MW to 10 MW) and one DIDO type reactor (power 23 MW). The German research reactors are used for neutron scattering for basic research in the field of solid state research, neutron metrology, for the fabrication of isotopes and for neutron activation analysis for medicine and biology, for investigating the influence of radiation on materials and for nuclear fuel behavior. It will be vital to continue current investigations in the future. Further operation of the German research reactors is therefore indispensable. Safe, regular disposal of the irradiated fuel elements arising now and in future operation is of primary importance. Furthermore, there are several plants with considerable quantities of spent fuel, the safe disposal of which is a matter of urgency. These include above all the VKTA facilities in Rossendorf and also the TRIGA reactors, where disposal will only be necessary upon decommissioning. The present paper report is concerned with the disposal of fuel from the German research reactors. It briefly deals with the situation in the USA since the end of 1988, describes interim solutions for current disposal requirements and then mainly concentrates on the German disposal concept currently being prepared. This concept initially envisages the long-term (25--50 years) dry interim storage of fuel elements in special containers in a central German interim store with subsequent direct final disposal without reprocessing of the irradiated fuel

  19. IAEA Coordinated Research Project on the Establishment of a Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Borio Di Tigliole, A.; Schaaf, Van Der; Barnea, Y.; Bradley, E.; Morris, C.; Rao, D. V. H. [Research Reactor Section, Vianna (Australia); Shokr, A. [Research Reactor Safety Section, Vienna (Australia); Zeman, A. [International Atomic Energy Agency, Vienna (Australia)

    2013-07-01

    Today more than 50% of operating Research Reactors (RRs) are over 45 years old. Thus, ageing management is one of the most important issues to face in order to ensure availability (including life extension), reliability and safe operation of these facilities for the future. Management of the ageing process requires, amongst others, the predictions for the behavior of structural materials of primary components subjected to irradiation such as reactor vessel and core support structures, many of which are extremely difficult or impossible to replace. In fact, age-related material degradation mechanisms resulted in high profile, unplanned and lengthy shutdowns and unique regulatory processes of relicensing the facilities in recent years. These could likely have been prevented by utilizing available data for the implementation of appropriate maintenance and surveillance programmes. This IAEA Coordinated Research Project (CRP) will provide an international forum to establish a material properties Database for irradiated core structural materials and components. It is expected that this Database will be used by research reactor operators and regulators to help predict ageing related degradation. This would be useful to minimize unpredicted outages due to ageing processes of primary components and to mitigate lengthy and costly shutdowns. The Database will be a compilation of data from RRs operators' inputs, comprehensive literature reviews and experimental data from RRs. Moreover, the CRP will specify further activities needed to be addressed in order to bridge the gaps in the new created Database, for potential follow-on activities. As per today, 13 Member States (MS) confirmed their agreement to contribute to the development of the Database, covering a wide number of materials and properties. The present publication incorporates two parts: the first part includes details on the pre-CRP Questionnaire, including the conclusions drawn from the answers received from

  20. IAEA Coordinated Research Project on the Establishment of a Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactors

    International Nuclear Information System (INIS)

    Borio Di Tigliole, A.; Schaaf, Van Der; Barnea, Y.; Bradley, E.; Morris, C.; Rao, D. V. H.; Shokr, A.; Zeman, A.

    2013-01-01

    Today more than 50% of operating Research Reactors (RRs) are over 45 years old. Thus, ageing management is one of the most important issues to face in order to ensure availability (including life extension), reliability and safe operation of these facilities for the future. Management of the ageing process requires, amongst others, the predictions for the behavior of structural materials of primary components subjected to irradiation such as reactor vessel and core support structures, many of which are extremely difficult or impossible to replace. In fact, age-related material degradation mechanisms resulted in high profile, unplanned and lengthy shutdowns and unique regulatory processes of relicensing the facilities in recent years. These could likely have been prevented by utilizing available data for the implementation of appropriate maintenance and surveillance programmes. This IAEA Coordinated Research Project (CRP) will provide an international forum to establish a material properties Database for irradiated core structural materials and components. It is expected that this Database will be used by research reactor operators and regulators to help predict ageing related degradation. This would be useful to minimize unpredicted outages due to ageing processes of primary components and to mitigate lengthy and costly shutdowns. The Database will be a compilation of data from RRs operators' inputs, comprehensive literature reviews and experimental data from RRs. Moreover, the CRP will specify further activities needed to be addressed in order to bridge the gaps in the new created Database, for potential follow-on activities. As per today, 13 Member States (MS) confirmed their agreement to contribute to the development of the Database, covering a wide number of materials and properties. The present publication incorporates two parts: the first part includes details on the pre-CRP Questionnaire, including the conclusions drawn from the answers received from the MS

  1. Hitrex 1: an interim report on experimental and analytical work on BNL's zero power HTR

    Energy Technology Data Exchange (ETDEWEB)

    Beynon, A J; Kitching, S J; Lewis, T A; Waterson, R H

    1972-06-15

    This report presents interim experimental and theoretical results for the BNL Hitrex reactor. Reactivity effects and thermal and fast reaction rate distributions have been measured. Preliminary analysis has been performed, and some initial comparisons between theory and experiments made. (auth)

  2. Fast Flux Test Facility interim examination and maintenance cell: Past, present, and future

    International Nuclear Information System (INIS)

    Vincent, J.R.

    1990-09-01

    The Fast Flux Test Facility Interim Examination and Maintenance Cell was designed to perform interim examination and/or disassembly of experimental core components for final analysis elsewhere, as well as maintenance of sodium-wetted or neutron-activated internal reactor parts and plant support hardware. The Interim Examination and Maintenance Cell equipment developed and used for the first ten years of operation has been primarily devoted to the disassembly and examination of core component test assemblies. While no major reactor equipment has required remote repair or maintenance, the Interim Examina Examination and Maintenance Cell has served as the remote repair facility for its own in-cell equipment, and several innovative remote repairs have been accomplished. The Interim Examination and Maintenance Cell's demonstrated versatility has shown its capability to support a challenging future. 12 refs., 9 figs

  3. How safe is safe?

    International Nuclear Information System (INIS)

    Hughes, C.F.; Flood, M.

    1996-01-01

    60 and 70 degree convexo-concave valve. Nine hundred and one valves were implanted in Australia. Twelve strut fractures were reported. Two other patients have been explanted and have demonstrated 'single leg separation'. This particular problem was only investigated when two patients died of a fractured valve in the same hospital on the same day. A retrospective study of all known patients in Australia has shown poor follow up, lack of knowledge and indeed lack of interest in device failure modes. Consequently, the Australian and New Zealand Heart Valve Registry was established to track all implanted valves and to notify physicians of any new information. This is perhaps the first device-specific register in Australia. The safety of individual devices is often not known by manufacturers, regulators and clinicians alike. No follow up is available and large volume long term studies are yet to be implemented for the majority of devices. Without such studies and without mandatory problem reporting, the relative safety of medical devices will continue to be measured by banner headlines, sensational TV 'grabs' and protracted law suits. At present, only schemes such as the Problem Reporting Scheme can tell us (albeit vaguely) 'how safe is safe'

  4. How safe are nuclear reactors?

    International Nuclear Information System (INIS)

    Krieg, R.

    1988-04-01

    The author, dealing with nuclear safety studies for many years, presents his own view and experience. He gives an interesting description understandable for non-experts. Also delicate problems, pretty often discussed in the public, are included. Starting with the Chernobyl accident he explains the consequences of the radiation exposure and critizes the reaction of relevant social groups including nuclear experts and politicians. The conceivable accident scenarios for German plants are described. Also severe accidents, because of their low probability not considered during the licensing procedure, are discussed. The resulting risks are compared with other known risks. Finally, some hints on the reliability of the assessments are given. Essential negative aspects of nuclear power are due to social political problems. For people the real understanding of risks is difficult. Participation in the work is often impossible. Poor understanding leads to wrong reactions causing political instabilities. On the other hand, energy is the key for all life. A historical view underlines this. Therefore, the potentials but also the environmental impacts of different energy sources - fossil, nuclear and the so-called alternative energies - are compared. Especially the robbery of the fossil energy sources and the severe consequences for our climate are adressed. The author presents a well balanced view of the problems. He asks the reader to think about it and to draw his own conclusions. (orig.) [de

  5. Waste Encapsulation and Storage Facility interim operational safety requirements

    CERN Document Server

    Covey, L I

    2000-01-01

    The Interim Operational Safety Requirements (IOSRs) for the Waste Encapsulation and Storage Facility (WESF) define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt and inspection of cesium and strontium capsules from private irradiators; decontamination of the capsules and equipment; surveillance of the stored capsules; and maintenance activities. Controls required for public safety, significant defense-in-depth, significant worker safety, and for maintaining radiological consequences below risk evaluation guidelines (EGs) are included.

  6. International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), Paris – March 4-7, 2013: Closing Session. Summary of Sustainability of Advanced Fuel Cycles Panel Session II

    International Nuclear Information System (INIS)

    Cameron, R.

    2013-01-01

    Sustainability was discussed in terms of the social, environment and economic perspectives, which arise from the original Brundtland definition of sustainability. The panel presented their perspectives of the need to move towards a sustainable future, involving better use of uranium, reductions in high-level radioactive waste, safe, secure and economic operation of nuclear reactors and the fuel cycle. In all cases, it was considered that sustainability in the long-term must involve fast reactors and a closed nuclear fuel cycle, although both Korea and the IAEA pointed out that these are clearly national decisions and there will not be a single solution for all countries

  7. Central waste complex interim safety basis

    International Nuclear Information System (INIS)

    Cain, F.G.

    1995-01-01

    This interim safety basis provides the necessary information to conclude that hazards at the Central Waste Complex are controlled and that current and planned activities at the CWC can be conducted safely. CWC is a multi-facility complex within the Solid Waste Management Complex that receives and stores most of the solid wastes generated and received at the Hanford Site. The solid wastes that will be handled at CWC include both currently stored and newly generated low-level waste, low-level mixed waste, contact-handled transuranic, and contact-handled TRU mixed waste

  8. CMM Interim Check (U)

    Energy Technology Data Exchange (ETDEWEB)

    Montano, Joshua Daniel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-23

    Coordinate Measuring Machines (CMM) are widely used in industry, throughout the Nuclear Weapons Complex and at Los Alamos National Laboratory (LANL) to verify part conformance to design definition. Calibration cycles for CMMs at LANL are predominantly one year in length. Unfortunately, several nonconformance reports have been generated to document the discovery of a certified machine found out of tolerance during a calibration closeout. In an effort to reduce risk to product quality two solutions were proposed – shorten the calibration cycle which could be costly, or perform an interim check to monitor the machine’s performance between cycles. The CMM interim check discussed makes use of Renishaw’s Machine Checking Gauge. This off-the-shelf product simulates a large sphere within a CMM’s measurement volume and allows for error estimation. Data was gathered, analyzed, and simulated from seven machines in seventeen different configurations to create statistical process control run charts for on-the-floor monitoring.

  9. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  10. Welcoming Address & Opening Remarks [International Conference on Research Reactors: Safe Management and Effective Utilization, Vienna (Austria), 16-20 November 2015

    International Nuclear Information System (INIS)

    Chudakov, M.

    2017-01-01

    For more than 60 years, research reactors have been centres of innovation and productivity for nuclear science and technology programmes in 67 countries around the world. Research reactors provide a multidisciplinary environment to catalyse scientific, industrial, medical and agricultural development. They are facilities for nuclear education and training of young scientists and technicians, and they can contribute to the development of nuclear power programmes. According to the IAEA Research Reactor Database, there are 246 research reactors currently in operation in 55 countries, and close to 30 new research reactor projects are at different stages of implementation. Many of the operating reactors are several decades old and face ageing management issues. These reactors must be operated and maintained with due regard to safety and security. Some reactors face challenges with sustainable supply of fresh fuel. Others are looking to improve utilization, which is linked to justifying adequate resources for operation, maintenance and refurbishment. As some of the fuel return programmes are expected to wind down in the near term, the community will need to find solutions for spent fuel and waste management. And taking into account the large number of reactors, about 140, no longer in operation, as well as ageing reactors coming to the end of their lifecycles, decommissioning is an important area of sharing experience and best practice. You will have an opportunity to discuss these and other issues over the course of the conference.

  11. Interim Hanford Waste Management Plan

    International Nuclear Information System (INIS)

    1985-09-01

    The September 1985 Interim Hanford Waste Management Plan (HWMP) is the third revision of this document. In the future, the HWMP will be updated on an annual basis or as major changes in disposal planning at Hanford Site require. The most significant changes in the program since the last release of this document in December 1984 include: (1) Based on studies done in support of the Hanford Defense Waste Environmental Impact Statement (HDW-EIS), the size of the protective barriers covering contaminated soil sites, solid waste burial sites, and single-shell tanks has been increased to provide a barrier that extends 30 m beyond the waste zone. (2) As a result of extensive laboratory development and plant testing, removal of transuranic (TRU) elements from PUREX cladding removal waste (CRW) has been initiated in PUREX. (3) The level of capital support in years beyond those for which specific budget projections have been prepared (i.e., fiscal year 1992 and later) has been increased to maintain Hanford Site capability to support potential future missions, such as the extension of N Reactor/PUREX operations. The costs for disposal of Hanford Site defense wastes are identified in four major areas in the HWMP: waste storage and surveillance, technology development, disposal operations, and capital expenditures

  12. Safety aspects of long term operation of water moderated reactors. Recommendations on the scope and content of programmes for safe long term operation. Final report of the extrabudgetary programme on safety aspects long term operation of water moderated reactors

    International Nuclear Information System (INIS)

    2007-07-01

    During the last two decades, the number of IAEA Member States giving high priority to continuing the operation of nuclear power plants beyond the time frame originally anticipated is increasing. This is related to the age of nuclear power plants connected to the grid worldwide. The IAEA started to develop guidance on the safety aspects of ageing management in the 1990s. Recognizing the development in a number of its Member States, the IAEA initiated this Extrabudgetary Programme on Safety Aspects of Long Term Operation of Water Moderated Reactors in 2003. The objective of the Programme was to establish recommendations on the scope and content of activities to ensure safe long term operation of water moderated reactors. The term long term operation is used to accommodate various approaches in Member States and is defined as operation beyond an initial time frame set forth in design, standards, licence, and/or regulations, that is justified by safety assessment, considering life limiting processes and features for systems, structures and components. The scope of the Programme included general long term operation framework, mechanical components and materials, electrical components and instrumentation and control, and structural components and structures. The scope of the Programme was limited to physical structures of the NPPs. Four working groups addressed the above indicated technical areas. The Programme steering committee provided coordination and guidance and served as a forum for the exchange of information. The Programme implementation relied on voluntary in kind and financial contributions from the Czech Republic, Hungary, Slovakia, Sweden, the United Kingdom and the USA as well as in kind contributions from Bulgaria, Finland, the Netherlands, the Russian Federation, Spain, the Ukraine, and the European Commission. This report summarizes the main results, conclusions and recommendations of this Programme and provides in the Appendices I-IV detailed

  13. Environmental assessment of radiological effluents from data gathering and maintenance operation on Three Mile Island Unit 2: interim criteria approved by the Commission on April 7,1980

    International Nuclear Information System (INIS)

    1980-05-01

    The staff is currently in the process of preparing a programmatic environmental impact statement (PEIS) for TMI-2 which will address all radiological releases that may occur as a result of the cleanup and recovery operations. These operations will begin after the PEIS is published in final form provided the proposed cleanup programs have been found to be environmentally acceptable. In the interim period it is necessary for the licensee to conduct data gathering and maintenance operations on the damaged reactor. The action of approval of these interim operations does not foreclose any of the options of the PEIS. In addition, regardless of what cleanup choice is made in the PEIS, the approval of these data gathering and maintenance operations enhance the ability of the licensee to maintain the reactor in a safe configuration and to plan effectively for recovery operations. This Environmental Impact Appraisal evaluates the effects on the environment of allowing these data gathering and maintenance operations to be conducted. These data gathering and maintenance operations do not include purging of the containment atmosphere, disposal of EPICOR-II water, or the treatment and disposal of high level radioactively contaminated water in the reactor building

  14. AGR-1 Data Qualification Interim Report

    International Nuclear Information System (INIS)

    Abbott, Machael

    2009-01-01

    Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor (AGR-1) experiment, the processing of these data within NDMAS, and reports the interim FY09 qualification status of the AGR-1 data to date. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category, which is assigned by the data generator, and include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent QA program. The interim qualification status of the following four data streams is reported in this document: (1) fuel fabrication data, (2) fuel irradiation data, (3) fission product monitoring system (FPMS) data, and (4) Advanced Test Reactor (ATR) operating conditions data. A final report giving the NDMAS qualification status of all AGR-1 data (including cycle 145A) is planned for February 2010

  15. Interim storage study report

    Energy Technology Data Exchange (ETDEWEB)

    Rawlins, J.K.

    1998-02-01

    High-level radioactive waste (HLW) stored at the Idaho Chemical Processing Plant (ICPP) in the form of calcine and liquid and liquid sodium-bearing waste (SBW) will be processed to provide a stable waste form and prepare the waste to be transported to a permanent repository. Because a permanent repository will not be available when the waste is processed, the waste must be stored at ICPP in an Interim Storage Facility (ISF). This report documents consideration of an ISF for each of the waste processing options under consideration.

  16. Interim storage report

    International Nuclear Information System (INIS)

    Rawlins, J.K.

    1998-02-01

    High-level radioactive waste (HLW) stored at the Idaho Chemical Processing Plant (ICPP) in the form of calcine and liquid and liquid sodium-bearing waste (SBW) will be processed to provide a stable waste form and prepare the waste to be transported to a permanent repository. Because a permanent repository will not be available when the waste is processed, the waste must be stored at ICPP in an Interim Storage Facility (ISF). This report documents consideration of an ISF for each of the waste processing options under consideration

  17. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  18. Hitrex 1: an interim report on experimental and analytical work on BNL's zero power HTR

    Energy Technology Data Exchange (ETDEWEB)

    Beynon, A.J.; Kitching, S.J.; Lewis, T.A.; Waterson, R.H.

    1972-06-15

    This report presents interim experimental and theoretical results for the BNL Hitrex reactor. Reactivity effects and thermal and fast reaction rate distributions have been measured. Preliminary analysis has been performed, and some initial comparisons between theory and experiments made. (auth)

  19. 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

  20. SAFE Newsletter

    OpenAIRE

    2013-01-01

    The Center of Excellence SAFE – “Sustainable Architecture for Finance in Europe” – is a cooperation of the Center for Financial Studies and Goethe University Frankfurt. It is funded by the LOEWE initiative of the State of Hessen (Landes-Offensive zur Entwicklung wissenschaftlich-ökonomischer Exzellenz). SAFE brings together more than 40 professors and just as many junior researchers who are all dedicated to conducting research in support of a sustainable financial architecture. The Center has...

  1. Federal Administrative Court confirms interim action for the Kruemmel power plant site

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The interim action concerning the reactor site was confirmed in 1972, the action of voidance because of the possible injuriousness to health was rejected in the first instance. In the appeal OVG Lueneburg had sharpened three clauses concerning the contents of the restrictions and injunctions included in the interim action to the disadvantage of the operator. Responding to a new appeal the BVerwG has eliminated these restrictions and dismissed the following appeals of the plaintiffs. (HP) [de

  2. Opening Session [International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), Paris, France, March 4-7, 2013

    International Nuclear Information System (INIS)

    Laurent Michel

    2013-01-01

    For this opening address, I would like to share with you some thoughts about the evolution of the key drivers during the last decades for the development of fast reactors from the very pioneering age till now, taking into account new concerns and major events occurred since the last international conference on fast reactors and related fuel cycles held in Kyoto, Japan (FR 2009). There are three major periods: • The pioneering age (1945-1980) with breeding as a main driver followed by a kind of “winter season“ (1980-2000) for the development of fast reactors worldwide; • The so-called “brainstorming” phase (2000-2010), back to physics and nuclear chemistry, with international rebirth of the research on fast reactors and advanced fuel cycle owing to the GENERATION IV initiative, revisiting various reactor concepts along with 4 main drivers: sustainability, safety, proliferation resistance and costcompetitiveness. • A new era now (started in 2010) with very promising technological options and projects of prototypes with two main key drivers: → Innovation towards enhanced safety which is a major concern for public acceptance of nuclear power, especially after the FUKUSHIMA accident. → Higher flexibility in the management of fissile materials and nuclear waste in order to take into account various possible options for the contribution of nuclear power in the energy mix

  3. ETF interim design review

    International Nuclear Information System (INIS)

    Steiner, D.; Rutherford, P.H.

    1980-01-01

    A three-day ETF Interim Design Review was conducted on July 23-25, 1980, at the Sheraton Potomac Inn in Rockville, Maryland. The intent of the review was to provide a forum for an in-depth assessment and critique of all facets of the ETF design by members of the fusion community. The review began with an opening plenary session at which an overview of the ETF design was presented by D. Steiner, manager of the ETF Design Center, complemented by a physics overview by P.H. Rutherford, chairman of the ETF/INTOR Physics Committee. This was followed by six concurrent review sessions over the next day and a half. The review closed with a plenary session at which the Design Review Board presented its findings. This document consists of the viewgraphs for the opening plenary session and an edited version of the presentations made by Steiner and Rutherford

  4. Interim geotechnical data report

    International Nuclear Information System (INIS)

    1986-01-01

    This issue, the Interim Geotechnical Field Data Report, presents information obtained from the geotechnical activities at the WIPP site underground facilities since the last quarterly report. It also includes cumulative plots which contain all previous data. Finally, it continues the geotechnical analyses and interpretations of the data. The GFDR is organized into two principal parts. The first part, Geotechnical Field Data, presents in graphical form all the data collected since April 1982 from the geomechanical instruments. Presented in the second part, Evaluation and Analyses, are preliminary interpretations and analyses of the data. In this report, continuing geotechnical assessment of all the facility features is presented. Also included in the second part are separate sections on evaluation and interpretation of the instrumentation measurements, and an updated description and evaluation of observed behavior of the underground openings

  5. Federal Interim Storage program

    International Nuclear Information System (INIS)

    Johnson, E.R.; McBride, J.A.

    1984-01-01

    The DOE has developed a program for providing Federal Interim Storage servies for spent nuclear fuel which complies with the requirements of the Nuclear Waste Policy Act of 1982. Although very little constructive activity in providing storage facilities can be undertaken by DOE until fuel has been certified by NRC as eligible for FIS, DOE planning and background information is such as to provide reasonable assurance that its obligations can be fulfilled when the required certifications have been issued. A fee structure providing fuel recovery of all costs associated with the FIS program, as required by the Act, has been developed. It provides for an equitable distribution of costs among users, based on the quantity of fuel requiring storage

  6. Acceptance criteria for interim dry storage of aluminum-clad fuels

    International Nuclear Information System (INIS)

    Sindelar, R.L.; Peacock, H.B. Jr.; Iyer, N.C.; Louthan, M.R. Jr.

    1994-01-01

    Direct repository disposal of foreign and domestic research reactor fuels owned by the United States Department of Energy is an alternative to reprocessing (together with vitrification of the high level waste and storage in an engineered barrier) for ultimate disposition. Neither the storage systems nor the requirements and specifications for acceptable forms for direct repository disposal have been developed; therefore, an interim storage strategy is needed to safely store these fuels. Dry storage (within identified limits) of the fuels received from wet-basin storage would avoid excessive degradation to assure post-storage handleability, a full range of ultimate disposal options, criticality safety, and provide for maintaining confinement by the fuel/clad system. Dry storage requirements and technologies for US commercial fuels, specifically zircaloy-clad fuels under inert cover gas, are well established. Dry storage requirements and technologies for a system with a design life of 40 years for dry storage of aluminum-clad foreign and domestic research reactor fuels are being developed by various groups within programs sponsored by the DOE

  7. Interim Stabilization Equipment Essential and Support Drawing Plan

    International Nuclear Information System (INIS)

    HORNER, T.M.

    2000-01-01

    The purpose of this document is to list the Interim Stabilization equipment drawings that are classified as Essential or Support drawings. Essential Drawings are those drawings identified by the facility staff as necessary to directly support the safe operation of the facility or equipment. [CHG 2000a]. Support Drawings are those drawings identified by the facility staff that further describe the design details of structures, systems or components shown on essential drawings. [CHG 2000a

  8. Interim Stabilization Equipment Essential and Support Drawing Plan

    International Nuclear Information System (INIS)

    KOCH, M.R.

    1999-01-01

    The purpose of this document is to list the Interim Stabilization equipment drawings that are classified as Essential or Support drawings. Essential Drawings: Those drawings identified by the facility staff as necessary to directly support the safe operation of the facility or equipment. Support Drawings: Those drawings identified by the facility staff that further describe the design details of structures, systems or components shown on essential drawings

  9. Safe Haven.

    Science.gov (United States)

    Bush, Gail

    2003-01-01

    Discusses school libraries as safe havens for teenagers and considers elements that foster that atmosphere, including the physical environment, lack of judgments, familiarity, leisure, and a welcoming nature. Focuses on the importance of relationships, and taking the time to listen to teens and encourage them. (LRW)

  10. Safe Grid

    Science.gov (United States)

    Chow, Edward T.; Stewart, Helen; Korsmeyer, David (Technical Monitor)

    2003-01-01

    The biggest users of GRID technologies came from the science and technology communities. These consist of government, industry and academia (national and international). The NASA GRID is moving into a higher technology readiness level (TRL) today; and as a joint effort among these leaders within government, academia, and industry, the NASA GRID plans to extend availability to enable scientists and engineers across these geographical boundaries collaborate to solve important problems facing the world in the 21 st century. In order to enable NASA programs and missions to use IPG resources for program and mission design, the IPG capabilities needs to be accessible from inside the NASA center networks. However, because different NASA centers maintain different security domains, the GRID penetration across different firewalls is a concern for center security people. This is the reason why some IPG resources are been separated from the NASA center network. Also, because of the center network security and ITAR concerns, the NASA IPG resource owner may not have full control over who can access remotely from outside the NASA center. In order to obtain organizational approval for secured remote access, the IPG infrastructure needs to be adapted to work with the NASA business process. Improvements need to be made before the IPG can be used for NASA program and mission development. The Secured Advanced Federated Environment (SAFE) technology is designed to provide federated security across NASA center and NASA partner's security domains. Instead of one giant center firewall which can be difficult to modify for different GRID applications, the SAFE "micro security domain" provide large number of professionally managed "micro firewalls" that can allow NASA centers to accept remote IPG access without the worry of damaging other center resources. The SAFE policy-driven capability-based federated security mechanism can enable joint organizational and resource owner approved remote

  11. Reactor surface contamination stabilization. Innovative technology summary report

    International Nuclear Information System (INIS)

    1998-11-01

    Contaminated surfaces, such as the face of a nuclear reactor, need to be stabilized (fixed) to avoid airborne contamination during decontamination and decommissioning activities, and to prepare for interim safe storage. The traditional (baseline) method of fixing the contamination has been to spray a coating on the surfaces, but ensuring complete coverage over complex shapes, such as nozzles and hoses, is difficult. The Hanford Site C Reactor Technology Demonstration Group demonstrated innovative technologies to assess stabilization properties of various coatings and to achieve complete coverage of complex surfaces on the reactor face. This demonstration was conducted in two phases: the first phase consisted of a series of laboratory assessments of various stabilization coatings on metal coupons. For the second phase, coatings that passed the laboratory tests were applied to the front face of the C Reactor and evaluated. The baseline coating (Rust-Oleum No. 769) and one of the innovative technologies did not completely cover nozzle assemblies on the reactor face, the most critical of the second-phase evaluation criteria. However, one of the innovative coating systems, consisting of a base layer of foam covered by an outer layer of a polymeric film, was successful. The baseline technology would cost approximately 33% as much as the innovative technology cost of $64,000 to stabilize an entire reactor face (196 m 2 or 2116 ft 2 ) with 2,004 nozzle assemblies, but the baseline system failed to provide complete surface coverage

  12. The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

    International Nuclear Information System (INIS)

    Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

    2003-01-01

    Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project

  13. Relationship between general safety requirements and safety culture in the improvement of safe operation of I.N.R. TRIGA reactor facilities

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Preda, M.; Chiritescu, M.; Dumitru, M.

    1996-01-01

    Acquiring of the basic principles of ''safety culture'' by a large number of profesionals in the nuclear field drew the attention of the decision factors in the INR managerial structure, who decided to promote certain practical actions at each level in order to improve nuclear safety. Starting from the ''Republican Standards for Nuclear Safety'' issued by CSEN in 1975, where general safety criteria are defined for nuclear reactors and NPPs, the specialists at the TRIGA reactor originated and implemented a coherent and secure system to ensure nuclear safety over all steps of nuclear activities: research, conception, execution, commissioning and operation. This system has been continuosly corrected so that now it is completely integrated in a modern safety system. The paper presents the way in which a modern system for nuclear safety at the TRIGA reactor has been implemented and developed, in accordance to specific criteria and requirements imposed by related National Regulations and with the principles of safety culture. Starting from the definition of specific responsabilities, there are presented the internal stipulations and practical actions at all levels in order to enhance nuclear safety. (orig.)

  14. Interim Design Report

    CERN Document Server

    Choubey, S.; Goswami, S.; Berg, J.S.; Fernow, R.; Gallardo, J.C.; Gupta, R.; Kirk, H.; Simos, N.; Souchlas, N.; Ellis, M.; Kyberd, P.; Benedetto, E.; Fernandez-Martinez, E.; Efthymiopoulos, I.; Garoby, R.; Gilardoni, S.; Martini, M.; Prior, G.; Ballett, P.; Pascoli, S.; Bross, A.; Geer, S.; Johnstone, C.; Kopp, J.; Mokhov, N.; Morfin, J.; Neuffer, D.; Parke, S.; Popovic, M.; Strait, J.; Striganov, S.; Blondel, A.; Dufour, F.; Laing, A.; Soler, F.J.P; Lindner, M.; Schwetz, T.; Alekou, A.; Apollonio, M.; Aslaninejad, M.; Bontoiu, C.; Dornan, P.; Eccleston, R.; Kurup, A.; Long, K.; Pasternak, J.; Pozimski, J.; Bogacz, A.; Morozov, V.; Roblin, Y.; Bhattacharya, S.; Majumdar, D.; Mori, Y.; Planche, T.; Zisman, M.; Cline, D.; Stratakis, D.; Ding, X.; Coloma, P.; Donini, A.; Gavela, B.; Lopez Pavon, J.; Maltoni, M.; Bromberg, C.; Bonesini, M.; Hart, T.; Kudenko, Y.; Mondal, N.; Antusch, S.; Blennow, M.; Ota, T.; Abrams, R.J.; Ankenbrandt, C.M.; Beard, K.B.; Cummings, M.A.C.; Flanagan, G.; Johnson, R.P.; Roberts, T.J.; Yoshikawa, C.Y.; Migliozzi, P.; Palladino, V.; de Gouvea, A.; Graves, V.B.; Kuno, Y.; Peltoniemi, J.; Blackmore, V.; Cobb, J.; Witte, H.; Mezzetto, M.; Rigolin, S.; McDonald, K.T.; Coney, L.; Hanson, G.; Snopok, P.; Tortora, L.; Andreopoulos, C.; Bennett, J.R.J.; Brooks, S.; Caretta, O.; Davenne, T.; Densham, C.; Edgecock, R.; Kelliher, D.; Loveridge, P.; McFarland, A.; Machida, S.; Prior, C.; Rees, G.; Rogers, C.; Thomason, J.W.G.; Booth, C.; Skoro, G.; Karadzhov, Y.; Matev, R.; Tsenov, R.; Samulyak, R.; Mishra, S.R.; Petti, R.; Dracos, M.; Yasuda, O.; Agarwalla, S.K.; Cervera-Villanueva, A.; Gomez-Cadenas, J.J.; Hernandez, P.; Li, T.; Martin-Albo, J.; Huber, P.; Back, J.; Barker, G.; Harrison, P.; Meloni, D.; Tang, J.; Winter, W.

    2011-01-01

    The International Design Study for the Neutrino Factory (the IDS-NF) was established by the community at the ninth "International Workshop on Neutrino Factories, super-beams, and beta- beams" which was held in Okayama in August 2007. The IDS-NF mandate is to deliver the Reference Design Report (RDR) for the facility on the timescale of 2012/13. In addition, the mandate for the study [3] requires an Interim Design Report to be delivered midway through the project as a step on the way to the RDR. This document, the IDR, has two functions: it marks the point in the IDS-NF at which the emphasis turns to the engineering studies required to deliver the RDR and it documents baseline concepts for the accelerator complex, the neutrino detectors, and the instrumentation systems. The IDS-NF is, in essence, a site-independent study. Example sites, CERN, FNAL, and RAL, have been identified to allow site-specific issues to be addressed in the cost analysis that will be presented in the RDR. The choice of example sites shou...

  15. Scheme of higher-density storage of spent nuclear fuel in Chernobyl NPP interim storage facility no. 1

    International Nuclear Information System (INIS)

    Britan, P.M.

    2008-01-01

    On 29. March 2000 the Cabinet of Ministers of Ukraine issued a decree prescribing that the last operating unit of Chernobyl NPP be shut down before its design lifetime expiry. In accordance with the Contract concluded on 14 June 1999 between the National Energy-generating Company 'Energoatom' and the Consortium of Framatome, Campenon Bernard-SGE and Bouygues, in order to store the spent ChNPP fuel a new interim dry storage facility (ISF-2) for spent ChNPP fuel would be built. Currently the spent nuclear fuel (spent fuel assemblies - SFAs) is stored in reactor cooling pools and in the reactors on Units 1, 2, 3, as well as in the wet Interim Storage Facility (ISF-1). Taking into account the expected delay with the commissioning of ISF-2, and in connection with the scheduled activities to build the New Safe Confinement (including the taking-down of the existing ventilation stack of ChNPP Units 3 and 4) and the expiry of the design operation life of Units 1 and 2, it is expedient to remove the nuclear fuel from Units 1, 2 and 3. This is essential to improve nuclear safety and ensure that the schedule of construction of the New Safe Confinement is met. The design capacity of ISF-1 (17 800 SFAs) is insufficient to store all SFAs (21 284) currently on ChNPP. A technically feasible option that has been applied on other RBMK plants is denser storage of spent nuclear fuel in the cooling ponds of the existing ISF-1. The purpose of the proposed modifications is to introduce changes to the ISF-1 design supported by necessary justifications required by the Ukrainian codes with the objective of enabling the storage of additional SFAs in the existing storage space (cooling pools). The need for the modification is caused by the requirement to remove nuclear fuel from the ChNPP units as soon as possible, before the work begins to decommission these units, as well as to create safe conditions for the construction of the New Safe Confinement over the existing Shelter Unit. (author)

  16. Digital computer operation of a nuclear reactor

    International Nuclear Information System (INIS)

    Colley, R.W.

    1984-01-01

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

  17. Dynamic calculations of a PWR - reactor building for different soil parameters for the safe shutdown earthquake and explosion pressure wave load cases

    International Nuclear Information System (INIS)

    Brandt, K.; Krutzik, N.; Kaiser, A.

    1982-01-01

    For different dynamic soil properties and soil dampings - ranging from very soft to very rigid soil parameters - time histoires of displacements and accelerations as well as response spectra are calculated for several floors for the reactor building of a nuclear power plant using a finite element shell model. As regards the loadcase safety earthquake the computations are carried out for four different soil properties, and the response spectra of different floors are compared. In the loadcase exterior explosion, results for three different soils are obtained. All results are discussed and explained extensively. (Author) [pt

  18. French Strategy and Programs [International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13), Paris, France, March 4-7, 2013

    International Nuclear Information System (INIS)

    Bigot, Bernard

    2013-01-01

    • In spite of the recent Fukushima accident, the use of nuclear energy still remains a necessity for many countries With a mastered and always improved safety, nuclear energy will go on contributing to the world energy needs, reducing fossil fuel dependency and GHG emissions. • With the help of large international collaborations, CEA is committed to demonstrate potential pathways for making significant progress for future fast neutrons reactors and their corresponding fuel cycle. • In synergy with nuclear energy, the renewable energies must be widely developed

  19. 24 CFR 35.1330 - Interim controls.

    Science.gov (United States)

    2010-04-01

    ... Lead-Paint Hazard Evaluation and Hazard Reduction Activities § 35.1330 Interim controls. Interim..., cleanable covering or coating, such as metal coil stock, plastic, polyurethane, or linoleum. (3) Surfaces...

  20. Super safe small reactor RAPID-L conceptual design and R and D. JAERI's nuclear research promotion program, H11-002 (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Kambe, Mitsuru [Central Research Inst. of Electric Power Industry, Tokyo (Japan); Tsunoda, Hirokazu [Mitsubishi Research Inst., Inc., Tokyo (Japan); Mishima, Kaichiro [Kyoto Univ., Kyoto (Japan); Kawasaki, Akira [Tohoku Univ., School of Engineering, Sendai, Miyagi (Japan); Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The 200 kWe uranium nitride fueled lithium cooled fast reactor 'RAPID-L' combined with thermoelectric power conversion system that can be operated unmanned without refueling for up to ten years has been demonstrated. The RAPID refueling concept enables quick and simplified refueling, and achieves plant design lifetime over 20 years. A significant advantage of the RAPID-L design, which does not require the use of control rods - is the introduction of the innovative reactivity control systems: lithium expansion module (LEM), lithium injection module (LIM) and lithium release module (LRM). LEM is the most promising candidate for improving inherent reactivity feedback. LEMs could realize burnup compensation. LIMs assure sufficient negative reactivity feedback in unprotected transients. LRMs enable an automated reactor startup by detecting the hot standby temperature of the primary coolant. All these systems use {sup 6}Li as liquid poison and are actuated by highly reliable physical properties (volume expansion of {sup 6}Li for LEM, and freeze seal melting for LIM and LRM). (author)

  1. Super safe small reactor RAPID-L conceptual design and R and D. JAERI's nuclear research promotion program, H11-002 (Contract research)

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Kawasaki, Akira; Iwamura, Takamichi

    2003-03-01

    The 200 kWe uranium nitride fueled lithium cooled fast reactor 'RAPID-L' combined with thermoelectric power conversion system that can be operated unmanned without refueling for up to ten years has been demonstrated. The RAPID refueling concept enables quick and simplified refueling, and achieves plant design lifetime over 20 years. A significant advantage of the RAPID-L design, which does not require the use of control rods - is the introduction of the innovative reactivity control systems: lithium expansion module (LEM), lithium injection module (LIM) and lithium release module (LRM). LEM is the most promising candidate for improving inherent reactivity feedback. LEMs could realize burnup compensation. LIMs assure sufficient negative reactivity feedback in unprotected transients. LRMs enable an automated reactor startup by detecting the hot standby temperature of the primary coolant. All these systems use 6 Li as liquid poison and are actuated by highly reliable physical properties (volume expansion of 6 Li for LEM, and freeze seal melting for LIM and LRM). (author)

  2. Radioactive waste interim storage in Germany

    International Nuclear Information System (INIS)

    2015-12-01

    The short summary on the radioactive waste interim storage in Germany covers the following issues: importance of interim storage in the frame of radioactive waste management, responsibilities and regulations, waste forms, storage containers, transport of vitrified high-level radioactive wastes from the reprocessing plants, central interim storage facilities (Gorleben, Ahaus, Nord/Lubmin), local interim storage facilities at nuclear power plant sites, federal state collecting facilities, safety, radiation exposure in Germany.

  3. 7 CFR 1735.75 - Interim financing.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 11 2010-01-01 2010-01-01 false Interim financing. 1735.75 Section 1735.75... Involving Loan Funds § 1735.75 Interim financing. (a) A borrower may submit a written request for RUS approval of interim financing if it is necessary to close an acquisition before the loan to finance the...

  4. 7 CFR 1738.21 - Interim financing.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 11 2010-01-01 2010-01-01 false Interim financing. 1738.21 Section 1738.21... Interim financing. (a) Upon notification by RUS that an applicant's application is considered complete, the applicant may enter into an interim financing agreement with a lender other than RUS or use its...

  5. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

    The main objective of this publication is to provide practical guidance and recommendations on operational radiation protection aspects related to the safe handling of tritium in laboratories, industrial-scale nuclear facilities such as heavy-water reactors, tritium removal plants and fission fuel reprocessing plants, and facilities for manufacturing commercial tritium-containing devices and radiochemicals. The requirements of nuclear fusion reactors are not addressed specifically, since there is as yet no tritium handling experience with them. However, much of the material covered is expected to be relevant to them as well. Annex III briefly addresses problems in the comparatively small-scale use of tritium at universities, medical research centres and similar establishments. However, the main subject of this publication is the handling of larger quantities of tritium. Operational aspects include designing for tritium safety, safe handling practice, the selection of tritium-compatible materials and equipment, exposure assessment, monitoring, contamination control and the design and use of personal protective equipment. This publication does not address the technologies involved in tritium control and cleanup of effluents, tritium removal, or immobilization and disposal of tritium wastes, nor does it address the environmental behaviour of tritium. Refs, figs and tabs

  6. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  7. Implications of inherent safe nuclear power system

    International Nuclear Information System (INIS)

    Song, Yo-Taik

    1987-01-01

    The safety of present day nuclear power reactors and research reactors depends on a combination of design features of passive and active systems, and the alert judgement of their operators. A few inherently safe designs of nuclear reactors for power plants are currently under development. In these designs, the passive systems are emphasized, and the active systems are minimized. Also efforts are made to eliminate the potential for human failures that initiate the series of accidents. If a major system fails in these designs, the core is flooded automatically with coolants that flow by gravity, not by mechanical pumps or electromagnetic actuators. Depending on the choice of the coolants--water, liquid metal and helium gas--there are three principal types of inherently safe reactors. In this paper, these inherently safe reactor designs are reviewed and their implications are discussed. Further, future perspectives of their acceptance by nuclear industries are discussed. (author)

  8. Safe cycling!

    CERN Multimedia

    Anaïs Schaeffer

    2012-01-01

    The HSE Unit will be running a cycling safety campaign at the entrances to CERN's restaurants on 14, 15 and 16 May. Pop along to see if they can persuade you to get back in the saddle!   With summer on its way, you might feel like getting your bike out of winter storage. Well, the HSE Unit has come up with some original ideas to remind you of some of the most basic safety rules. This year, the prevention campaign will be focussing on three themes: "Cyclists and their equipment", "The bicycle on the road", and "Other road users". This is an opportunity to think about the condition of your bike as well as how you ride it. From 14 to 16 May, representatives of the Swiss Office of Accident Prevention and the Touring Club Suisse will join members of the HSE Unit at the entrances to CERN's restaurants to give you advice on safe cycling (see box). They will also be organising three activity stands where you can test your knowle...

  9. The Japan Power Demonstration Reactor (JPDR) dismantling activities. Management of JPDR dismantling waste

    International Nuclear Information System (INIS)

    Abe, Masayoshi; Nakata, Susumu; Ito, Shinichi

    1996-01-01

    The management of wastes, both radioactive and non-radioactive, is one of the most important issues for a safe and reasonable dismantling operation of nuclear power plants. A large amount of radioactive wastes is arising from a reactor dismantling operation in a relatively short period time, ranging in a wide variety from very low level to relatively high level. Moreover non-radioactive waste is also in a huge amount. The dismantling operation of Japan Power Demonstration Reactor (JPDR) resulted in 24,440 tons of dismantling wastes, of which about 15% was radioactive and 85% non-radioactive. These wastes were managed successfully implementing a well developed management plan for JPDR dismantling waste. Research and development works for handling of JPDR dismantling wastes were performed, including fixation of loose contamination on surface, volume reduction and waste containers for on-site transportation and interim storage. The JPDR dismantling wastes generated were classified and categorized depending on their materials, characteristics and activity level. Approximately 2,100 tons of radioactive wastes were stored in the interim storage facilities on site using developed containers, and 1,670 tons of radioactive concrete waste were used for a safe demonstration test of a simple near-surface disposal for very low level waste. Other dismantling wastes such as steel and concrete which were categorized as non-radioactive were recycled and reused as useful resources. This paper describes the management of the JPDR dismantling wastes. (author)

  10. Specific provisions applicable to the production, inspection, treatment, packaging and interim storage of low and medium level bituminized wastes resulting from the reprocessing of fuels irradiated in pressurized water reactors

    International Nuclear Information System (INIS)

    1984-04-01

    The Fundamental Safety Rules applicable to certain types of nuclear installation are intended to clarify the conditions of which observance, for the type of installation concerned and for the subject that they deal with, is considered as equivalent to compliance with regulatory French technical practice. These Rules should facilitate safety analysises and the clear understanding between persons interested in matters related to nuclear safety. They in no way reduce the operator's liability and pose no obstacle to statutory provisions in force. For any installation to which a Fundamental Safety Rule applies according to the foregoing paragraph, the operator may be relieved from application of the Rule if he shows proof that the safety objectives set by the Rule are attained by other means that he proposes within the framework of statutory procedures. Furthermore, the Central Service for the Safety of Nuclear Installations reserves the right at all times to alter any Fundamental Safety Rule, as required, should it deem this necessary, while specifying the applicability conditions. This rule sets forth the specific provisions applicable to the production, inspection, treatment, packaging and interim storage of the wastes, resulting from the reprocessing of fuels irradiated in a PWR and coated in bitumen

  11. Particular provisions applicable to the production, inspection, treatment, packaging and interim storage of wastes immobilized in cement, resulting from the reprocessing of fuels irradiated in pressurized light water reactors

    International Nuclear Information System (INIS)

    1985-02-01

    The Fundamental Safety Rules applicable to certain types of nuclear installation are intended to clarify the conditions of which observance, for the type of installation concerned and for the subject that they deal with, is considered as equivalent to compliance with regulatory French technical practice. These Rules should facilitate safety analysises and the clear understanding between persons interested in matters related to nuclear safety. They in no way reduce the operator's liability and pose no obstacle to statutory provisions in force. For any installation to which a Fundamental Safety Rule applies according to the foregoing paragraph, the operator may be relieved from application of the Rule if he shows proof that the safety objectives set by the Rule are attained by other means that he proposes within the framework of statutory procedures. Furthermore, the Central Service for the Safety of Nuclear Installations reserves the right at all times to alter any Fundamental Safety Rule, as required, should it deem this necessary, while specifying the applicability conditions. This rule is intended to stipulate the specific provisions applicable to the production, inspection, treatment, packaging and interim storage of the wastes, resulting from the reprocessing of fuels irradiated in a PWR and immobilized in cement

  12. Spent Fuel Test - Climax: technical measurements. Interim report, fiscal year 1982

    International Nuclear Information System (INIS)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1983-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted for the US Department of Energy (DOE) under the technical direction of the Lawrence Livermore National Laboratory (LLNL). Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April to May 1980, thus initiating a test with a planned 3- to 5-year fuel storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Three exchanges of spent fuel between the SFT-C and a surface storage facility furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and two previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 2-1/2 years of the test on more than 900 channels. Data continue to be acquired from the test. Some data are now available for analysis and are presented here. Highlights of activities this year include analysis of fracture data obtained during site characterization, laboratory studies of radiation effects and drilling damage in Climax granite, improved calculations of near-field heat transfer and thermomechanical response, a ventilation effects study, and further development of the data acquisition and management systems

  13. Management of Hanford Site non-defense production reactor spent nuclear fuel, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1997-03-01

    The US Department of Energy (DOE) needs to provide radiologically, and industrially safe and cost-effective management of the non-defense production reactor spent nuclear fuel (SNF) at the Hanford Site. The proposed action would place the Hanford Site's non-defense production reactor SNF in a radiologically- and industrially-safe, and passive storage condition pending final disposition. The proposed action would also reduce operational costs associated with storage of the non-defense production reactor SNF through consolidation of the SNF and through use of passive rather than active storage systems. Environmental, safety and health vulnerabilities associated with existing non-defense production reactor SNF storage facilities have been identified. DOE has determined that additional activities are required to consolidate non-defense production reactor SNF management activities at the Hanford Site, including cost-effective and safe interim storage, prior to final disposition, to enable deactivation of facilities where the SNF is now stored. Cost-effectiveness would be realized: through reduced operational costs associated with passive rather than active storage systems; removal of SNF from areas undergoing deactivation as part of the Hanford Site remediation effort; and eliminating the need to duplicate future transloading facilities at the 200 and 400 Areas. Radiologically- and industrially-safe storage would be enhanced through: (1) removal from aging facilities requiring substantial upgrades to continue safe storage; (2) utilization of passive rather than active storage systems for SNF; and (3) removal of SNF from some storage containers which have a limited remaining design life. No substantial increase in Hanford Site environmental impacts would be expected from the proposed action. Environmental impacts from postulated accident scenarios also were evaluated, and indicated that the risks associated with the proposed action would be small

  14. Safe and secure: transportation of radioactive materials

    International Nuclear Information System (INIS)

    Howe, D.

    2015-01-01

    Western Waste Management Facility is Central Transportation Facility for Low and Intermediate waste materials. Transportation support for Stations: Reactor inspection tools and heavy water between stations and reactor components and single bundles of irradiated fuel to AECL-Chalk River for examination. Safety Track Record: 3.2 million kilometres safely travelled and no transportation accident - resulting in a radioactive release.

  15. Interim report on the TMI-2 purification filter examination

    International Nuclear Information System (INIS)

    Mason, R.E.; Hobbins, R.R.; Cook, B.A.; MacDonald, P.E.

    1983-02-01

    Filters from the purification/makeup system of the Three Mile Island Unit 2 Reactor were examined after the March 28, 1979, accident to determine the character of the debris transported to the filters. The general condition of the filters is presented. Material was removed from the filters and examined. The elemental and radionuclide makeup of the debris is discussed. Distribution of particle size and shape is presentd for some of the material examined. This is an interim report. When the investigation is completed, another report summarizing all of the data will be issued

  16. How safe is safe enough?

    International Nuclear Information System (INIS)

    Desnoyers, B.; Chanzy, Y.

    2004-01-01

    The IAEA Regulations for the Safe Transport of Radioactive Material, were historically established with the objective to reduce the probability that persons be exposed to unacceptable doses due to normal operation or accident situations during transport of radioactive material. Based on the International Basic Safety Standards for Protection against Ionizing Radiation (BSS), the definition, which was adopted for an unacceptable dose for an accident situation, is the excess of the maximum dose limits permissible in a single year for the occupational exposure of a worker in the BSS. Concerning the severity of accident situations, it has always be clearly stated that the objective of the tests for demonstrating ability to withstand accident conditions of transport was not to cover every accident condition, but solely most of them. The last available evaluations regarding the rate of accidents which are covered by the standardised accident conditions of transport defined in the IAEA Regulations give a range of about 80%, plus or minus 15% which depends on transport mode and studies. Consequently, slight variations in the capabilities of the packages to meet the specified performance would probably not have significant consequences on the protection level in case of accident. In the assessment of the compliance with the regulations, the tendency of experts, taking advantage of the enhanced performances of computer calculation codes, is to ask more and more calculations, with more and more accuracy, leading to more and more restrictions. Consequently, cost and delay are considerably increased without any evidence of an equivalent effect on the level of protection. This paper will initiate a reflection on the general objectives and principles when implementing the Regulations, in such a way that demonstrations remain cost effective, taking into account evolution of the techniques and a high level of safety

  17. Research reactor spent fuel management in Argentina

    International Nuclear Information System (INIS)

    Audero, M.A.; Bevilacqua, A.M.; Mehlich, A.M.; Novara, O.

    2002-01-01

    The research reactor spent fuel (RRSF) management strategy will be presented as well as the interim storage experience. Currently, low-enriched uranium RRSF is in wet interim storage either at reactor site or away from reactor site in a centralized storage facility. High-enriched uranium RRSF from the centralized storage facility has been sent to the USA in the framework of the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The strategy for the management of the RRSF could implement the encapsulation for interim dry storage. As an alternative to encapsulation for dry storage some conditioning processes are being studied which include decladding, isotopic dilution, oxidation and immobilization. The immobilized material will be suitable for final disposal. (author)

  18. Management and storage of spent fuel from CEA research reactors

    International Nuclear Information System (INIS)

    Merchie, F.

    1996-01-01

    CEA research reactors and their interim spent fuel storage facilities are described. Long-term solutions for spent fuel storage problems, involving wet storage at PEGASE or dry storage at CASCAD, are outlined in some detail. (author)

  19. Agenda 21 interim balance, 1997

    Energy Technology Data Exchange (ETDEWEB)

    Van Vuuren, D.P.; Bakkes, J.A.

    1997-10-01

    Five years after the `United Nations Conference on Environment and Development (UNCED)`, an interim balance was drawn up to see what was done to meet the ambitious challenges adopted in Agenda 21 during this conference. Such a balance is presented in this report and the complementary brochure, `Developments in Sustainability 1992-1997`, reflecting societal developments and changes in environmental quality, as well as changes in responses to environmental concerns. 24 figs., 12 tabs., 68 refs.

  20. Radioactivity, radiation protection and monitoring during dismantling of light-water reactors

    International Nuclear Information System (INIS)

    Hummel, L.; Zech, J.B.

    2005-01-01

    Based on the radioactivity inventory in the systems and components of light-water reactors observed during operation, the impact of actions during plant emptying after the conclusion of power operation and possible subsequent long-term safe enclosure concerning the composition of the nuclide inventory of the plant to be dismantled will be described. Derived from this will be the effects on radioactivity monitoring in the plant, physical radiation protection monitoring, and the measured characterization of the residual materials resulting from the dismantling. The impact of long-term interim storage will also be addressed in the discussion. The talk should provide an overview of the interrelationships between source terms, decay times and the radioactivity monitoring requirements of the various dismantling concepts for commercial light-water reactors. (orig.)

  1. Radioactive waste management at a Liquid Metal Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Abrams, C.S.; Fryer, R.H.; Witbeck, L.C.

    1979-01-01

    This paper presents the radioactive waste production and management at a Liquid Metal Fast Breeder Reactor-II (EBR-II), which is operated for the US Department of Energy by the Argonne National Laboratory at the Idaho National Engineering Laboratory (INEL). Since this facility, in addition to supplying power has been used to demonstrate the breeder, fuel cycling, and recently operations with defective fuel elements, various categories of waste have been handled safely over some 14 years of operation. Liquid wastes are processed such that the resulting effluent can be discharged to an uncontrolled area. Solid wastes up to 10,000 R/hr are packaged and shipped contamination-free to a disposal site or interim storage with exposures to personnel approximately 10 mrem. Gaseous waste discharges are low such as 143 Ci of noble gases in 1978 and do not have a significant effect on the environment even with operations with breached fuel

  2. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  3. Application of NUREG/CR-5999 interim fatigue curves to selected nuclear power plant components

    International Nuclear Information System (INIS)

    Ware, A.G.; Morton, D.K.; Nitzel, M.E.

    1995-03-01

    Recent test data indicate that the effects of the light water reactor (LWR) environment could significantly reduce the fatigue resistance of materials used in the reactor coolant pressure boundary components of operating nuclear power plants. Argonne National Laboratory has developed interim fatigue curves based on test data simulating LWR conditions, and published them in NUREG/CR-5999. In order to assess the significance of these interim fatigue curves, fatigue evaluations of a sample of the components in the reactor coolant pressure boundary of LWRs were performed. The sample consists of components from facilities designed by each of the four U.S. nuclear steam supply system vendors. For each facility, six locations were studied, including two locations on the reactor pressure vessel. In addition, there are older vintage plants where components of the reactor coolant pressure boundary were designed to codes that did not require an explicit fatigue analysis of the components. In order to assess the fatigue resistance of the older vintage plants, an evaluation was also conducted on selected components of three of these plants. This report discusses the insights gained from the application of the interim fatigue curves to components of seven operating nuclear power plants

  4. Burn site groundwater interim measures work plan.

    Energy Technology Data Exchange (ETDEWEB)

    Witt, Jonathan L. (North Wind, Inc., Idaho Falls, ID); Hall, Kevin A. (North Wind, Inc., Idaho Falls, ID)

    2005-05-01

    This Work Plan identifies and outlines interim measures to address nitrate contamination in groundwater at the Burn Site, Sandia National Laboratories/New Mexico. The New Mexico Environment Department has required implementation of interim measures for nitrate-contaminated groundwater at the Burn Site. The purpose of interim measures is to prevent human or environmental exposure to nitrate-contaminated groundwater originating from the Burn Site. This Work Plan details a summary of current information about the Burn Site, interim measures activities for stabilization, and project management responsibilities to accomplish this purpose.

  5. High Temperature Materials Interim Data Qualification Report

    International Nuclear Information System (INIS)

    Lybeck, Nancy

    2010-01-01

    Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim FY2010 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under NQA-1 guidelines, and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from two test series within the High Temperature Materials data stream have been entered into the NDMAS vault: (1) Tensile Tests for Sm (i.e., Allowable Stress) Confirmatory Testing - 1,403,994 records have been inserted into the NDMAS database. Capture testing is in process. (2) Creep-Fatigue Testing to Support Determination of Creep-Fatigue Interaction Diagram - 918,854 records have been processed and inserted into the NDMAS database. Capture testing is in process.

  6. Advantages on dry interim storage for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, L.S. [Centro Tecnologico da Marinha em Sao Paulo, Av. Professor Lineu Prestes 2468, 05508-900 Sao Paulo (Brazil); Rzyski, B.M. [IPEN/ CNEN-SP, 05508-000 Sao Paulo (Brazil)]. e-mail: romanato@ctmsp.mar.mil.br

    2006-07-01

    When the nuclear fuel lose its ability to efficiently create energy it is removed from the core reactor and moved to a storage unit waiting for a final destination. Generally, the spent nuclear fuel (SNF) remains inside concrete basins with water within the reactors facility for the radioactive activity decay. Water cools the generated heat and shields radioactivity emissions. After some period of time in water basins the SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing installations, or still wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet installations, depending on the method adopted by the nuclear power plant or other plans of the country. In many SNF wet storage sites the capacity can be fulfilled very quickly. If so, additional area or other alternative storage system should be given. There are many options to provide capacity increase in the wet storage area, but dry storages are worldwide preferred since it reduces corrosion concerns. In the wet storage the temperature and water purity should be constantly controlled whereas in the dry storage the SNF stands protected in specially designed canisters. Dry interim storages are practical and approved in many countries especially that have the 'wait and see' philosophy (wait to see new technologies development). This paper shows the advantages of dry interim storages sites in comparison with the wet ones and the nowadays problems as terrorism. (Author)

  7. Advantages on dry interim storage for spent nuclear fuel

    International Nuclear Information System (INIS)

    Romanato, L.S.; Rzyski, B.M.

    2006-01-01

    When the nuclear fuel lose its ability to efficiently create energy it is removed from the core reactor and moved to a storage unit waiting for a final destination. Generally, the spent nuclear fuel (SNF) remains inside concrete basins with water within the reactors facility for the radioactive activity decay. Water cools the generated heat and shields radioactivity emissions. After some period of time in water basins the SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing installations, or still wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet installations, depending on the method adopted by the nuclear power plant or other plans of the country. In many SNF wet storage sites the capacity can be fulfilled very quickly. If so, additional area or other alternative storage system should be given. There are many options to provide capacity increase in the wet storage area, but dry storages are worldwide preferred since it reduces corrosion concerns. In the wet storage the temperature and water purity should be constantly controlled whereas in the dry storage the SNF stands protected in specially designed canisters. Dry interim storages are practical and approved in many countries especially that have the 'wait and see' philosophy (wait to see new technologies development). This paper shows the advantages of dry interim storages sites in comparison with the wet ones and the nowadays problems as terrorism. (Author)

  8. 76 FR 58790 - Notice of Interim Approval

    Science.gov (United States)

    2011-09-22

    ... to the customers. Rate Scenario 3--Original Cumberland Marketing Policy The third rate alternative... an interim basis to the customers. Rate Scenario 3--Original Cumberland Marketing Policy The third... allocated on an interim basis to the customers. Rate Scenario 3--Original Cumberland Marketing Policy The...

  9. Addendum to IFMIF-CDA interim report

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Hiroshi; Ida, Mizuho [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; eds.

    1996-08-01

    During the second IFMIF-CDA Design Integration Workshop, the conceptual design and contents of `IFMIF-CDA Interim Report` were examined and discussed at both general and group meetings. Based on these discussion, the final IFMIF-CDA Report will be modified from the `Interim Report`. This report describes the outline of these modification. (author)

  10. Addendum to IFMIF-CDA interim report

    International Nuclear Information System (INIS)

    Maekawa, Hiroshi; Ida, Mizuho

    1996-08-01

    During the second IFMIF-CDA Design Integration Workshop, the conceptual design and contents of 'IFMIF-CDA Interim Report' were examined and discussed at both general and group meetings. Based on these discussion, the final IFMIF-CDA Report will be modified from the 'Interim Report'. This report describes the outline of these modification. (author)

  11. Solid waste burial grounds interim safety analysis

    International Nuclear Information System (INIS)

    Saito, G.H.

    1994-01-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment

  12. Solid waste burial grounds interim safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Saito, G.H.

    1994-10-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment.

  13. Safety of interim storage solutions of used nuclear fuel during extended term

    Energy Technology Data Exchange (ETDEWEB)

    Shelton, C.; Bader, S.; Issard, H.; Arslan, M. [AREVA, 7135 Minstrel Way, Suite 300 Columbia, MD 21045 (United States)

    2013-07-01

    In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

  14. Traveling Safely with Medicines

    Science.gov (United States)

    ... Medications Safely My Medicine List How to Administer Traveling Safely with Medicines Planes, trains, cars – even boats ... your trip, ask your pharmacist about how to travel safely with your medicines. Make sure that you ...

  15. Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 and 2

    International Nuclear Information System (INIS)

    CARRELL, R.D.

    2002-01-01

    The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft 2 and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available

  16. Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 & 2

    Energy Technology Data Exchange (ETDEWEB)

    CARRELL, R D

    2002-07-16

    The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft{sup 2} and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available.

  17. Interim and final storage casks

    International Nuclear Information System (INIS)

    Stumpfrock, L.; Kockelmann, H.

    2012-01-01

    The disposal of radioactive waste is a huge social challenge in Germany and all over the world. As is well known the search for a site for a final repository for high-level waste in Germany is not complete. Therefore, interim storage facilities for radioactive waste were built at plant sites in Germany. The waste is stored in these storage facilities in appropriate storage and transport casks until the transport in a final repository can be carried out. Licensing of the storage and transport casks aimed for use in the public space is done according to the traffic laws and for handling in the storage facility according to nuclear law. Taking into account the activity of the waste to be stored, different containers are in use, so that experience is available from the licensing and operation in interim storage facilities. The large volume of radioactive waste to be disposed of after the shut-down of power generation in nuclear power stations makes it necessary for large quantities of licensed storage and transport casks to be provided soon.

  18. Glass packages in interim storage

    International Nuclear Information System (INIS)

    Jacquet-Francillon, N.

    1994-10-01

    This report summarize the current state of knowledge concerning the behavior of type C waste packages consisting of vitrified high-level solutions produced by reprocessing spent fuel. The composition and the physical and chemical properties of the feed solutions are reviewed, and the vitrification process is described. Sodium alumino-borosilicate glass compositions are generally employed - the glass used at la Hague for LWR fuel solutions, for example, contains 45 % SiO 2 . The major physical, chemical, mechanical and thermal properties of the glass are reviewed. In order to allow their thermal power to diminish, the 3630 glass packages produced (as of January 1993) in the vitrification facilities at Marcoule and La Hague are placed in interim storage for several decades. The actual interim storage period has not been defined, as it is closely related to the concept and organization selected for the final destination of the packages: a geological repository. The glass behavior under irradiation is described. Considerable basic and applied research has been conducted to assess the aqueous leaching behavior of nuclear containment glass. The effects of various repository parameters (temperature, flow rate, nature of the environmental materials) have been investigated. The experimental findings have been used to specify a model describing the kinetics of aqueous corrosion of the glass. More generally all the ''source term'' models developed in France by the CEA or by ANDRA are summarized. (author). 152 refs., 33 figs

  19. Inherently safe technologies-chemical and nuclear

    International Nuclear Information System (INIS)

    Weinberg, A.M.

    1984-01-01

    Probabilistic risk assessments show an inverse relationship between the likelihood and the consequences of nuclear and chemical plant accidents, but the Bhopal accident has change public complacency about the safety of chemical plants to such an extent that public confidence is now at the same low level as with nuclear plants. The nuclear industry's response was to strengthen its institutions and improve its technologies, but the public may not be convinced. One solution is to develop reactors which do not depend upon the active intervention of humans of electromechanical devices to deal with emergencies, but which have physical properties that limit the possible temperature and power of a reactor. The Process Inherent Ultimately Safe and the modular High-Temperature Gas-Cooled reactors are two possibilities. the chemical industry needs to develop its own inherently safe design precepts that incorporate smallness, safe processes, and hardening against sabotage. 5 references

  20. A safe workplace

    International Nuclear Information System (INIS)

    Rittsel, Hans; Andersson, Bengt A.

    1993-01-01

    Full text: The video 'A safe workplace' has been produced by ABB Atom in order to create a tool for showing different target audiences that ABB Atom Nuclear Fuel Production Plant is a safe workplace and to 'de-mystify' nuclear fuel production. The main target audiences are visitor groups and employees of the company, but the video also qualifies for use as an information tool for other target groups who ask for a proper explanation of the way nuclear fuel is produced. The summarized content of the video is as follows: All individual steps of the production process are described with focus on the safety, quality and environmental requirements. The first part shows the delivery of UF 6 (uranium hexafluoride) to the plant and the following process for the conversion to UO 2 (uranium dioxide). The conversion method used is wet conversion that includes evaporation, precipitation, filtration, washing, reduction and stabilization. The next part is a description of the fuel pellet manufacture including uranium oxide blending, pellet pressing, sintering, grinding and a final visual inspection. A separate part, describing the manufacture of fuel pellets with a burnable neutron absorber, is included. The third part shows how to produce fuel rods and complete assemblies. Some of the moments of quality supervision that support the entire manufacturing process are also shown. The last part of the video comprises a brief description of the manufacture of fuel channels and other reactor core components like control rods. The video is produced with a Swedish spoken narrative. The playing time is 15 minutes. The video will be delivered with a text printed in English and copies reproduced in the PAL/VHS system may be ordered from ABB Atom Communication Dept. telefax no +4621-11 41 90, at the price of USD 100.- or SEK 750.- each. (author)

  1. Dry interim storage of radioactive material in Germany

    International Nuclear Information System (INIS)

    Drobniewski, Christian; Palmes, Julia

    2013-01-01

    In accordance with the waste management concept in Germany, spent fuel is stored in interim storage facilities for a period of up to 40 years until deposition in a geological repository. In twelve on-site interim storages in the vicinity or directly on the sites of the nuclear power plants, spent fuel elements from reactor operation are stored after the necessary period of decay in wet storage basins inside the reactors. Additionally, three central interim storage facilities for storage of spent fuel of different origin are in operation. The German facilities realize the concept of dry interim storage in metallic transport and storage casks. The confinement of the radioactive material is ensured by the double lid system of the casks, of which the leak tightness is monitored constantly. The casks are constructed to provide adequate heat removal and shielding of gamma and neutron radiation. Usually the storage facilities are halls of thick concrete structures, which ensure the removal of the decay heat by natural convection. The main safety goal of the storage concept is to prevent unnecessary exposure of persons, material goods and environment to ionizing radiation. Moreover any exposure should be kept as low as reasonable achievable. To reach this goal the containment of the radioactive materials, the disposal of decay heat, the sub criticality and the shielding of ionizing radiation has to be demonstrated by the applicant and verified by the licensing authority. In particular accidents, incidents and disasters have to be considered in the facility and cask design. This includes mechanical impacts onto the cask, internal and external fire, and environmental effects like wind, rain, snowfall, flood, earthquakes and landslides. In addition civilizatoric influences like plane crashes and explosions have to be taken into account. In all mentioned cases the secure confinement of the radioactive materials has to be ensured. On-site storage facilities have to consider the

  2. Used Fuel Logistics: Decades of Experience with transportation and Interim storage solutions

    Energy Technology Data Exchange (ETDEWEB)

    Orban, G.; Shelton, C.

    2015-07-01

    Used fuel inventories are growing worldwide. While some countries have opted for a closed cycle with recycling, numerous countries must expand their interim storage solutions as implementation of permanent repositories is taking more time than foreseen. In both cases transportation capabilities will have to be developed. AREVA TN has an unparalleled expertise with transportation of used fuel. For more than 50 years AREVA TN has safely shipped more than 7,000 used fuel transport casks. The transportation model that was initially developed in the 1970s has been adapted and enhanced over the years to meet more restrictive regulatory requirements and evolving customer needs, and to address public concerns. The numerous “lessons learned” have offered data and guidance that have allowed for also efficient and consistent improvement over the decades. AREVA TN has also an extensive experience with interim dry storage solutions in many countries on-site but also is working with partners to developed consolidated interim storage facility. Both expertise with storage and transportation contribute to safe, secure and smooth continuity of the operations. This paper will describe decades of experience with a very successful transportation program as well as interim storage solutions. (Author)

  3. 13 CFR 120.890 - Source of interim financing.

    Science.gov (United States)

    2010-01-01

    ... 13 Business Credit and Assistance 1 2010-01-01 2010-01-01 false Source of interim financing. 120... Development Company Loan Program (504) Interim Financing § 120.890 Source of interim financing. A Project may use interim financing for all Project costs except the Borrower's contribution. Any source (including...

  4. Interim licensing criteria for physical protection of certain storage of spent fuel

    International Nuclear Information System (INIS)

    Dwyer, P.A.

    1994-11-01

    This document presents interim criteria to be used in the physical protection licensing of certain spent fuel storage installations. Installations that will be reviewed under this criteria are those that store power reactor spent fuel at decommissioned power reactor sites; independent spent fuel storage installations located outside of the owner controlled area of operating nuclear power reactors; monitored retrievable storage installations owned by the Department of Energy, designed and constructed specifically for the storage, of spent fuel; the proposed geologic repository operations area; or permanently shutdown power reactors still holding a Part 50 license. This criteria applies to both dry cask and pool storage. However, the criteria in this document does not apply to the storage of spent fuel within the owner-controlled area of operating nuclear power reactors

  5. Interim reliability evaluation program (IREP)

    International Nuclear Information System (INIS)

    Carlson, D.D.; Murphy, J.A.

    1981-01-01

    The Interim Reliability Evaluation Program (IREP), sponsored by the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission, is currently applying probabilistic risk analysis techniques to two PWR and two BWR type power plants. Emphasis was placed on the systems analysis portion of the risk assessment, as opposed to accident phenomenology or consequence analysis, since the identification of risk significant plant features was of primary interest. Traditional event tree/fault tree modeling was used for the analysis. However, the study involved a more thorough investigation of transient initiators and of support system faults than studies in the past and substantially improved techniques were used to quantify accident sequence frequencies. This study also attempted to quantify the potential for operator recovery actions in the course of each significant accident

  6. A collection of publications and articles for a light water ultra-safe plant concept

    International Nuclear Information System (INIS)

    Klevans, E.H.

    1988-01-01

    This collection contains reports titled: ''The Penn State Ultra-Safe Reactor Concept; '' ''Ultra Safe Nuclear Power; '' ''Use of the Modular Modeling System, in the Design of the Penn State Advanced Light Water Reactor; '' ''Use of the Modular Modeling System in Severe Transient Analysis of Penn State Advanced Light Water Reactor; '' ''PSU Engineers' Reactor Design May Stop a Future TMI; '' and ''The Penn State Advanced Light Water reactor Concept.''

  7. Seismic Safety Margins Research Program. Phase I. Interim definition of terms

    International Nuclear Information System (INIS)

    Smith, P.D.; Dong, R.G.

    1980-01-01

    This report documents interim definitions of terms in the Seismic Safety Margins Research Program (SSMRP). Intent is to establish a common-based terminology integral to the probabilistic methods that predict more realistically the behavior of nuclear power plants during an earthquake. These definitions are a response to a request by the Nuclear Regulatory Commission Advisory Committee on Reactor Safeguards at its meeting held November 15-16, 1979

  8. Interim Status Report for Risk Management for SFRs

    International Nuclear Information System (INIS)

    Jankovsky, Zachary Kyle; Denman, Matthew R.; Groth, Katrina; Wheeler, Timothy A.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of passive, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to take advantage of natural phenomena to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a variety of beyond design basis events with the intent of exploring the utility of a Dynamic Bayesian Network to infer the state of the reactor to inform the operator's corrective actions. These inferences also serve to identify the instruments most critical to informing an operator's actions as candidates for hardening against radiation and other extreme environmental conditions that may exist in an accident. This reduction in uncertainty serves to inform ongoing discussions of how small sodium reactors would be licensed and may serve to reduce regulatory risk and cost for such reactors.

  9. Interim Status Report for Risk Management for SFRs

    Energy Technology Data Exchange (ETDEWEB)

    Jankovsky, Zachary Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-10-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of passive, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to take advantage of natural phenomena to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a variety of beyond design basis events with the intent of exploring the utility of a Dynamic Bayesian Network to infer the state of the reactor to inform the operator's corrective actions. These inferences also serve to identify the instruments most critical to informing an operator's actions as candidates for hardening against radiation and other extreme environmental conditions that may exist in an accident. This reduction in uncertainty serves to inform ongoing discussions of how small sodium reactors would be licensed and may serve to reduce regulatory risk and cost for such reactors.

  10. Spent fuel test - Climax: technical measurements. Interim report, Fiscal Year 1983

    International Nuclear Information System (INIS)

    Patrick, W.C.; Butkovich, T.R.; Carlson, R.C.

    1984-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted as part of the Nevada Nuclear Waste Storage Investigations. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April-May 1980. The spent-fuel canisters were retrieved and the thermal sources were de-energized in March-April 1983 when test data indicated that test objectives were met during the 3-year storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. In addition to emplacement and retrieval operations, three exchanges of spent-fuel between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and three previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the 3-1/2 year duration of the test on more than 900 channels. Data acquisition from the test is now limited to instrumentation calibration and evaluation activities. Data now available for analysis are presented here. Highlights of activities this year include a campaign of in situ stress measurements, mineralogical and petrological studies of pretest core samples, microfracture analyses of laboratory irradiated cores, improved calculations of near-field heat transfer and thermomechanical response during the final months of heating as well as during a six-month cool-down period, metallurgical analyses of selected test components, and further development of the data acquisition and data management systems. 27 references, 68 figures, 10 tables

  11. Spent fuel test - Climax: technical measurements. Interim report, Fiscal Year 1983

    Energy Technology Data Exchange (ETDEWEB)

    Patrick, W.C.; Butkovich, T.R.; Carlson, R.C.; Durham, W.B.; Ganow, H.C.; Hage, G.L.; Majer, E.L.; Montan, D.N.; Nyholm, R.A.; Rector, N.L.

    1984-02-01

    The Spent Fuel Test - Climax (SFT-C) is located 420 m below surface in the Climax stock granite on the Nevada Test Site. The test is being conducted as part of the Nevada Nuclear Waste Storage Investigations. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized April-May 1980. The spent-fuel canisters were retrieved and the thermal sources were de-energized in March-April 1983 when test data indicated that test objectives were met during the 3-year storage phase. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. In addition to emplacement and retrieval operations, three exchanges of spent-fuel between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. Technical objectives of the test led to development of a technical measurements program, which is the subject of this and three previous interim reports. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the 3-1/2 year duration of the test on more than 900 channels. Data acquisition from the test is now limited to instrumentation calibration and evaluation activities. Data now available for analysis are presented here. Highlights of activities this year include a campaign of in situ stress measurements, mineralogical and petrological studies of pretest core samples, microfracture analyses of laboratory irradiated cores, improved calculations of near-field heat transfer and thermomechanical response during the final months of heating as well as during a six-month cool-down period, metallurgical analyses of selected test components, and further development of the data acquisition and data management systems. 27 references, 68 figures, 10 tables.

  12. DOE UST interim subsurface barrier technologies workshop

    International Nuclear Information System (INIS)

    1992-09-01

    This document contains information which was presented at a workshop regarding interim subsurface barrier technologies that could be used for underground storage tanks, particularly the tank 241-C-106 at the Hanford Reservation

  13. Research reactor support

    International Nuclear Information System (INIS)

    2005-01-01

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  14. Behavior of spent nuclear fuel and storage system components in dry interim storage.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions.

  15. Behavior of spent nuclear fuel and storage-system components in dry interim storage

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Gilbert, E.R.; Guenther, R.J.

    1982-08-01

    Irradiated nuclear fuel has been handled under dry conditions since the early days of nuclear reactor operation, and use of dry storage facilities for extended management of irradiated fuel began in 1964. Irradiated fuel is currently being stored dry in four types of facilities: dry wells, vaults, silos, and metal casks. Essentially all types of irradiated nuclear fuel are currently stored under dry conditions. Gas-cooled reactor (GCR) and liquid metal fast breeder reactor (LMFBR) fuels are stored in vaults and dry wells. Certain types of fuel are being stored in licensed dry storage facilities: Magnox fuel in vaults in the United Kingdom and organic-cooled reactor (OCR) fuel in silos in Canada. Dry storage demonstrations are under way for Zircaloy-clad fuel from boiling water reactors BWR's, pressurized heavy-water reactors (PHWRs), and pressurized water reactors (PWRs) in all four types of dry storage facilities. The demonstrations and related hot cell and laboratory tests are directed toward expanding the data base and establishing a licensing basis for dry storage of water reactor fuel. This report reviews the scope of dry interim storage technology, the performance of fuel and facility materials, the status of programs in several countries to license dry storage of water reactor fuel, and the characteristics of water reactor fuel that relate to dry storage conditions

  16. Management and storage of nuclear fuel from Belgian research reactors

    International Nuclear Information System (INIS)

    Gubel, P.

    1996-01-01

    Experiences and problems with the storage of irradiated fuel at research reactors in Belgium are described. In particular, interim storage problems exist for spent fuel elements at the BR2 and the shut down BR3 reactors in Mol. (author). 1 ref

  17. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    Ordonez, J.P.

    1990-01-01

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

  18. Centralized interim storage facility for radioactive wastes at Wuerenlingen (ZWILAG)

    International Nuclear Information System (INIS)

    Lutz, H.R.; Schnetzler, U.

    1994-01-01

    Radioactive waste management in Switzerland is the responsibility of the waste producers; in this respect, the law requires permanent, safe management of the wastes by means of final disposal. Nagra is responsible for the research and development work associated with final disposal. Processing of the wastes into a form suitable for disposal, as well as interim storage, remain the responsibility of the waste producers. In order to supplement the existing conditioning and storage facilities at the nuclear power plants and to replace the outdated waste treatment plant at the Paul Scherrer Institute (PSI) at Wuerenlingen, the operators of the Swiss nuclear power plants are planning a joint treatment and storage facility at the PSI-East site. The organisation ''Zwischenlager Wuerenlingen AG'', which was set up at the beginning of 1990, has been entrusted with this task. (author) 4 figs

  19. Savannah River Interim Waste Management Program Plan - FY 1986

    International Nuclear Information System (INIS)

    1985-09-01

    This document provides the program plan as requested by the Savannah River Operations Office of the Department of Energy. The plan was developed to provide a working knowledge of the nature and extent of the interim waste management programs being undertaken by Savannah River (SR) contractors for the Fiscal Year 1986. In addition, the document projects activities for several years beyond 1986 to adequately plan for safe handling and storage of radioactive wastes generated at Savannah River and for developing technology for improved management of low-level solid wastes. A revised plan will be issued prior to the beginning of the first quarter of each fiscal year. In this document, work descriptions and milestone schedules are current as of the date of publication. Budgets are based on available information as of May 1985

  20. Developing new transportable storage casks for interim dry storage

    International Nuclear Information System (INIS)

    Hayashi, K.; Iwasa, K.; Araki, K.; Asano, R.

    2004-01-01

    Transportable storage metal casks are to be consistently used during transport and storage for AFR interim dry storage facilities planning in Japan. The casks are required to comply with the technical standards of regulations for both transport (hereinafter called ''transport regulation'') and storage (hereafter called ''storage regulation'') to maintain safety functions (heat transfer, containment, shielding and sub-critical control). In addition to these requirements, it is not planned in normal state to change the seal materials during storage at the storage facility, therefore it is requested to use same seal materials when the casks are transported after storage period. The dry transportable storage metal casks that satisfy the requirements have been developed to meet the needs of the dry storage facilities. The basic policy of this development is to utilize proven technology achieved from our design and fabrication experience, to carry out necessary verification for new designs and to realize a safe and rational design with higher capacity and efficient fabrication

  1. Savannah River Interim Waste Management Program plan, FY-1987

    International Nuclear Information System (INIS)

    1986-09-01

    This document provides the program plan as requested by the Savannah River Operations office of the Department of Energy. The plan was developed to provide a working knowledge of the nature and extent of the interim waste management programs being undertaken by Savannah River (SR) contractors for the Fiscal Year 1987. In addition, the document projects activities for several years beyond 1987 to adequately plan for safe handling and storage of radioactive wastes generated at Savannah River and for developing technology for improved management of low-level solid wastes. A revised plan will be issued prior to the beginning of the first quarter of each fiscal year. In this document, work descriptions and milestone schedules are current as of the date of publication. Budgets are based on available information as of June 1986

  2. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  3. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  4. Safe Kids Worldwide

    Science.gov (United States)

    ... Blog Videos Newsletter facebook twitter instagram pinterest gplus youtube Search Menu Why It Matters Who We Are What We Do Find Your Safe Kids Safe Kids Day Main menu Keeping All Kids Safe Safety Tips Get Involved 4 Star Charity Donate Text Burns and Scalds 4 tips ...

  5. Methods Data Qualification Interim Report

    International Nuclear Information System (INIS)

    Alessi, R. Sam; Grimmett, Tami; Vang, Leng; McGrath, Dave

    2010-01-01

    The overall goal of the Next Generation Nuclear Plant (NGNP) Data Management and Analysis System (NDMAS) is to maintain data provenance for all NGNP data including the Methods component of NGNP data. Multiple means are available to access data stored in NDMAS. A web portal environment allows users to access data, view the results of qualification tests and view graphs and charts of various attributes of the data. NDMAS also has methods for the management of the data output from VHTR simulation models and data generated from experiments designed to verify and validate the simulation codes. These simulation models represent the outcome of mathematical representation of VHTR components and systems. The methods data management approaches described herein will handle data that arise from experiment, simulation, and external sources for the main purpose of facilitating parameter estimation and model verification and validation (V and V). A model integration environment entitled ModelCenter is used to automate the storing of data from simulation model runs to the NDMAS repository. This approach does not adversely change the why computational scientists conduct their work. The method is to be used mainly to store the results of model runs that need to be preserved for auditing purposes or for display to the NDMAS web portal. This interim report demonstrates the currently development of NDMAS for Methods data and discusses data and its qualification that is currently part of NDMAS.

  6. Methods Data Qualification Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    R. Sam Alessi; Tami Grimmett; Leng Vang; Dave McGrath

    2010-09-01

    The overall goal of the Next Generation Nuclear Plant (NGNP) Data Management and Analysis System (NDMAS) is to maintain data provenance for all NGNP data including the Methods component of NGNP data. Multiple means are available to access data stored in NDMAS. A web portal environment allows users to access data, view the results of qualification tests and view graphs and charts of various attributes of the data. NDMAS also has methods for the management of the data output from VHTR simulation models and data generated from experiments designed to verify and validate the simulation codes. These simulation models represent the outcome of mathematical representation of VHTR components and systems. The methods data management approaches described herein will handle data that arise from experiment, simulation, and external sources for the main purpose of facilitating parameter estimation and model verification and validation (V&V). A model integration environment entitled ModelCenter is used to automate the storing of data from simulation model runs to the NDMAS repository. This approach does not adversely change the why computational scientists conduct their work. The method is to be used mainly to store the results of model runs that need to be preserved for auditing purposes or for display to the NDMAS web portal. This interim report demonstrates the currently development of NDMAS for Methods data and discusses data and its qualification that is currently part of NDMAS.

  7. The long term storage of advanced gas-cooled reactor (AGR) fuel

    International Nuclear Information System (INIS)

    Standring, P.N.

    1999-01-01

    The approach being taken by BNFL in managing the AGR lifetime spent fuel arisings from British Energy reactors is given. Interim storage for up to 80 years is envisaged for fuel delivered beyond the life of the Thorp reprocessing plant. Adopting a policy of using existing facilities, to comply with the principles of waste minimisation, has defined the development requirements to demonstrate that this approach can be undertaken safely and business issues can be addressed. The major safety issues are the long term integrity of both the fuel being stored and structure it is being stored in. Business related issues reflect long term interactions with the rest of the Sellafield site and storage optimisation. Examples of the development programme in each of these areas is given. (author)

  8. Light-water reactor accident classification

    International Nuclear Information System (INIS)

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

  9. 78 FR 68101 - Environmental Issues Associated With New Reactors and Specific Environmental Guidance for...

    Science.gov (United States)

    2013-11-13

    ... New Reactors and Specific Environmental Guidance for Integral Pressurized Water Reactors Reviews... Environmental Issues Associated with New Reactors'' and draft ISG ESP/ COL-ISG-027, ``Interim Staff Guidance on...://www.regulations.gov and search for Docket ID NRC-2013-0212 for ESP/COL- ISG-026 or NRC-2013-0211 for...

  10. SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT

    International Nuclear Information System (INIS)

    Norton, S.H.

    2010-01-01

    CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material. The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford

  11. SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    NORTON SH

    2010-02-23

    CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material. The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford

  12. Auditable Safety Analysis and Final Hazard Classification for the 105-N Reactor Zone and 109-N Steam Generator Zone Facility

    International Nuclear Information System (INIS)

    Kloster, G.L.

    1998-07-01

    This document is a graded auditable safety analysis (ASA) and final hazard classification (FHC) for the Reactor/Steam Generator Zone Segment. The Reactor/Steam Generator Zone Segment, part of the N Reactor Complex, that is also known as the Reactor Building and Steam Generator Cells. The installation of the modifications described within to support surveillance and maintenance activities are to be completed by July 1, 1999. The surveillance and maintenance activities addressed within are assumed to continue for the next 15- 20 years, until the initiation of facility D ampersand D (i.e., Interim Safe Storage). The graded ASA in this document is in accordance with EDPI-4.30-01, Rev. 1, Safety Analysis Documentation, (BHI-DE-1) and is consistent with guidance provided by the U.S. Department of Energy. This ASA describes the hazards within the facility and evaluates the adequacy of the measures taken to reduce, control, or mitigate the identified hazards. This document also serves as the FHC for the Reactor/Steam Generator Zone Segment. This FHC is developed through the use of bounding accident analyses that envelope the potential exposures to personnel

  13. "Same Room, Safe Place".

    Science.gov (United States)

    Keene Woods, Nikki

    2017-04-01

    There are many different professional stances on safe sleep and then there is the reality of caring for a newborn. There is a debate among professionals regarding safe sleep recommendations. The continum of recommendations vary from the American Academy of Pediatrics (AAP) Safe Sleep Guidelines to the bed-sharing recommendations from the Mother-Baby Behavioral Sleep Laboratory. The lack of consistent and uniform safe sleep recommendations from health professionals has been confusing for families but has more recently raised a real professional ethical dilemma. Despite years of focused safe sleep community education and interventions, sleep-related infant deaths are on the rise in many communities. This commentary calls for a united safe sleep message from all health professionals to improve health for mothers and infants most at-risk, "Same Room, Safe Place."

  14. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  15. Reactor core cooling device

    International Nuclear Information System (INIS)

    Kobayashi, Masahiro.

    1986-01-01

    Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

  16. RA research reactor - properties and experimental capabilities

    International Nuclear Information System (INIS)

    Milosevic, M.; Martinc, R.

    1978-01-01

    The brief survey of the Reactor RA exploitation experience, as well as the reactor equipment state, after 18 years of operation is presented. The results of efforts spent on reactor characteristics improvement in order to ensure safe and reliable reactor operation for next 15-20 years, are described [sr

  17. EMCS Retrofit Analysis - Interim Report

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, R.C.; Salsbury, T.I.; Bell, G.C.; Huang, Y.J.; Sezgen, A.O.; Mazzucchi, R.; Romberger, J.

    1999-03-01

    This report presents the interim results of analyses carried out in the Phillip Burton Federal Building in San Francisco from 1996 to 1998. The building is the site of a major demonstration of the BACnet communication protocol. The energy management and control systems (EMCS) in the building were retrofitted with BACnet compatible controllers in order to integrate certain existing systems on one common network. In this respect, the project has been a success. Interoperability of control equipment from different manufacturers has been demonstrated in a real world environment. Besides demonstrating interoperability, the retrofits carried out in the building were also intended to enhance control strategies and capabilities, and to produce energy savings. This report presents analyses of the energy usage of HVAC systems in the building, control performance, and the reaction of the building operators. The report does not present an evaluation of the performance capabilities of the BACnet protocol. A monitoring system was installed in the building that parallels many of the EMCS sensors and data were archived over a three-year period. The authors defined pre-retrofit and post-retrofit periods and analyzed the corresponding data to establish the changes in building performance resulting from the retrofit activities. The authors also used whole-building energy simulation (DOE-2) as a tool for evaluating the effect of the retrofit changes. The results of the simulation were compared with the monitored data. Changes in operator behavior were assessed qualitatively with questionnaires. The report summarizes the findings of the analyses and makes several recommendations as to how to achieve better performance. They maintain that the full potential of the EMCS and associated systems is not being realized. The reasons for this are discussed along with possible ways of addressing this problem. They also describe a number of new technologies that could benefit systems of the type

  18. Physical security at research reactors

    International Nuclear Information System (INIS)

    Clark, R.A.

    1977-01-01

    Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

  19. Design review report FFTF interim storage cask

    International Nuclear Information System (INIS)

    Scott, P.L.

    1995-01-01

    Final Design Review Report for the FFTF Interim Storage Cask. The Interim Storage Cask (ISC) will be used for long term above ground dry storage of FFTF irradiated fuel in Core Component Containers (CCC)s. The CCC has been designed and will house assemblies that have been sodium washed in the IEM Cell. The Solid Waste Cask (SWC) will transfer a full CCC from the IEM Cell to the RSB Cask Loading Station where the ISC will be located to receive it. Once the loaded ISC has been sealed at the RSB Cask Loading Station, it will be transferred by facility crane to the DSWC Transporter. After the ISC has been transferred to the Interim Storage Area (ISA), which is yet to be designed, a mobile crane will be used to place the ISC in its final storage location

  20. CMM Interim Check Design of Experiments (U)

    Energy Technology Data Exchange (ETDEWEB)

    Montano, Joshua Daniel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-07-29

    Coordinate Measuring Machines (CMM) are widely used in industry, throughout the Nuclear Weapons Complex and at Los Alamos National Laboratory (LANL) to verify part conformance to design definition. Calibration cycles for CMMs at LANL are predominantly one year in length and include a weekly interim check to reduce risk. The CMM interim check makes use of Renishaw’s Machine Checking Gauge which is an off-the-shelf product simulates a large sphere within a CMM’s measurement volume and allows for error estimation. As verification on the interim check process a design of experiments investigation was proposed to test a couple of key factors (location and inspector). The results from the two-factor factorial experiment proved that location influenced results more than the inspector or interaction.

  1. Choosing a spent fuel interim storage system

    International Nuclear Information System (INIS)

    Roland, V.; Hunter, I.

    2001-01-01

    The Transnucleaire Group has developed different modular solutions to address spent fuel interim storage needs of NPP. These solutions, that are present in Europe, USA and Asia are metal casks (dual purpose or storage only) of the TN 24 family and the NUHOMS canister based system. It is not always simple for an operator to sort out relevant choice criteria. After explaining the basic designs involved on the examples of the TN 120 WWER dual purpose cask and the NUHOMS 56 WWER for WWER 440 spent fuel, we shall discuss the criteria that govern the choice of a given spent fuel interim storage system from the stand point of the operator. In conclusion, choosing and implementing an interim storage system is a complex process, whose implications can be far reaching for the long-term success of a spent fuel management policy. (author)

  2. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    1979-12-01

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

  3. Safe havens in Europe

    DEFF Research Database (Denmark)

    Paldam, Martin

    2013-01-01

    Eleven safe havens exist in Europe providing offshore banking and low taxes. Ten of these states are very small while Switzerland is moderately small. All 11 countries are richer than their large neighbors. It is shown that causality is from small to safe haven to wealth, and that theoretically...... equilibriums are likely to exist where a certain regulation is substantially lower in a small country than in its big neighbor. This generates a large capital inflow to the safe havens. The pool of funds that may reach the safe havens is shown to be huge. It is far in excess of the absorptive capacity...... of the safe havens, but it still explains, why they are rich. Microstates offer a veil of anonymity to funds passing through, and Switzerland offers safe storage of funds....

  4. Interim results from UO2 fuel oxidation tests in air

    International Nuclear Information System (INIS)

    Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j.

    1987-08-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO 2 , fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO 2 pellets in the temperature range of 135 to 250 0 C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10 5 R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10 5 R/h gamma field. 33 refs., 51 figs., 6 tabs

  5. Interim Administrators in Higher Education: A National Study

    Science.gov (United States)

    Huff, Marie Thielke; Neubrander, Judy

    2015-01-01

    The focus of this paper is on the roles and experiences of interim administrators in higher education. A survey was given to current and recent interim administrators in four-year public universities and colleges across the United States. The goals were to identify the advantages and disadvantages of using and serving as interims, and to solicit…

  6. An Approach for Evaluating the Technical Quality of Interim Assessments

    Science.gov (United States)

    Li, Ying; Marion, Scott; Perie, Marianne; Gong, Brian

    2010-01-01

    Increasing numbers of schools and districts have expressed interest in interim assessment systems to prepare for summative assessments and to improve teaching and learning. However, with so many commercial interim assessments available, schools and districts are struggling to determine which interim assessment is most appropriate to their needs.…

  7. Development of dual-purpose metal cask for interim storage of spent nuclear fuel (1). Outline of cask structure

    International Nuclear Information System (INIS)

    Shimizu, Masashi; Hayashi, Makoto; Kashiwakura, Jun

    2003-01-01

    Spent fuels discharged from nuclear power plants in Japan are planed to be reprocessed at the nuclear fuel recycle plant under construction at Rokkasho-mura. Since the amount of the spent fuels exceeds that of recycled fuel, the spent fuels have to be properly stored and maintained as recycle fuel resource until the beginning of the reprocessing. For that sake, interim storage installations are being constructed outside the nuclear power plants by 2010. The storage dry casks have been practically used as the interim storage in the nuclear power plants. From this reason, the storage system using the storage dry casks is promising as the interim storage installations away form the reactors, which are under discussion. In the interim storage facilities, the storage using the dry cask of the storage metal cask with business showings, having the function of transportation is now under discussion. By employing transportation and storage dual-purpose cask, the repack equipments can be exhausted, and the reliability of the interim storage installations can be increased. Hitachi, Ltd. has been developing the high reliable and economical transportation and storage dry metal cask. In this report, the outline of our developing transportation and storage dry cask is described. (author)

  8. Lessons for outsourcing and interim management relationships.

    Science.gov (United States)

    Macko, W; Kostyack, P T

    1999-01-01

    Few decisions can affect an organization more than the selection of an outsourcing or interim management partner. More and more health care organizations face such decisions in today's competitive market in order to face new business needs. Making these relationships successful can be important for health care organizations seeking competitive advantages or seeking immediately accessible management support. These relationships, however, require careful partner selection and development. Success in outsourcing and interim management relationships is contingent upon a thorough selection process, a strong contract that has clearly and explicitly detailed responsibilities and a culture-sensitive business rapport between the client and selected partner.

  9. 216-T-4 interim stabilization final report

    International Nuclear Information System (INIS)

    Smith, D.L.

    1996-01-01

    This report provides a general description of the activities performed for the interim stabilization of the 216-T-4-1 ditch, 216-T-4-2 ditch, and 216-T-4-2 pond. Interim stabilization was required to reduce the amount of surface-contaminated acres and to minimize the migration of radioactive contamination. Work associated with the 216-T4-1 ditch and 216-T-4-2 pond was performed by the Radiation Area Remedial Action (RARA) Project. Work associated with the 216-T-4-2 ditch was done concurrently but was funded by Westinghouse Hanford Company (WHC) Tank Waste Remediation Systems (TWRS)

  10. Repacking of Cobalt 60 spent sources in the central interim storage

    International Nuclear Information System (INIS)

    Zeleznik, N.

    2003-01-01

    After the transfer of the responsibility for the management of the Central interim storage for waste from small producers, located at the reactor centre in Brinje near Ljubljana, Slovenia, the national Agency for radwaste management (ARAO) started with most urgent activities to improve the utilization of the storage facility. One of the main tasks has also been the rearrangement of the already stored radioactive waste in order to reduce volume of the waste and to collect same radioisotopes in the containers. The latest campaign, performed in 2002/2003, was repacking of all Co-60 spent sealed sources in the storage facility and also at the producer's premises which were after conditioning put into two drums with concrete matrix and stored back to the Central interim storage. The preparation works together with the implementation are described in the paper. (author)

  11. Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

    International Nuclear Information System (INIS)

    Richard, R.F.

    1995-01-01

    It has been postulated that a degradation phenomenon, referred to as ''hot cell rot'', may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ''Hot cell rot'' refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ''hot cell rot'' phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical

  12. Annex D 200 Area Interim Storage Area Final Safety Analysis Report Volume 5 (FSAR) (Section 1 and 2)

    International Nuclear Information System (INIS)

    CARRELL, R.D.

    2003-01-01

    The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft 2 and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped offsite to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility (FFTF) Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the FFTF SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF)TRIGA--One Rad-Vault container stores two DOT-6M 3 containers and six NRF TRIGA casks. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-1 cask with an inner commercial light water reactor (LWR) canister, are used for storing commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available

  13. Are Detox Diets Safe?

    Science.gov (United States)

    ... Safe Videos for Educators Search English Español Are Detox Diets Safe? KidsHealth / For Teens / Are Detox Diets ... seguras las dietas de desintoxicación? What Is a Detox Diet? The name sounds reassuring — everyone knows that ...

  14. Chernobyl new safe confinement

    International Nuclear Information System (INIS)

    Dodd, L.

    2011-01-01

    The author presents the new safe confinement that will be commissioned at Unit 4 of the Chernobyl NPP in 2015. The confinement will ensure that Chernobyl Unit 4 will be placed in an environmentally safe condition for at least next 100 years. The article highlights the current work status, future perspectives and the feasibility of confinement concept [ru

  15. Asymptotically Safe Dark Matter

    DEFF Research Database (Denmark)

    Sannino, Francesco; Shoemaker, Ian M.

    2015-01-01

    We introduce a new paradigm for dark matter (DM) interactions in which the interaction strength is asymptotically safe. In models of this type, the coupling strength is small at low energies but increases at higher energies, and asymptotically approaches a finite constant value. The resulting...... searches are the primary ways to constrain or discover asymptotically safe dark matter....

  16. Effectiveness of interim remedial actions at a radioactive waste facility

    International Nuclear Information System (INIS)

    Devgun, J.S.; Beskid, N.J.; Peterson, J.M.; Seay, W.M.; McNamee, E.

    1989-01-01

    Over the past eight years, several interim remedial actions have been taken at the Niagara Falls Storage Site (NFSS), primarily to reduce radon and gamma radiation exposures and to consolidate radioactive waste into a waste containment facility. Interim remedial actions have included capping of vents, sealing of pipes, relocation of the perimeter fence (to limit radon risk), transfer and consolidation of waste, upgrading of storage buildings, construction of a clay cutoff wall (to limit the potential groundwater transport of contaminants), treatment and release of contaminated water, interim use of a synthetic liner, and emplacement of an interim clay cap. An interim waste containment facility was completed in 1986. 6 refs., 3 figs

  17. Actinide recycling in reactors

    International Nuclear Information System (INIS)

    Kuesters, H.; Wiese, H.W.; Krieg, B.

    1995-01-01

    The objective is an assessment of the transmutation of long-lived actinides and fission products and the incineration of plutonium for reducing the risk potential of radioactive waste from reactors in comparison to direct waste disposal. The contribution gives an interim account on homogeneous and heterogeneous recycling of 'risk nuclides' in thermal and fast reactors. Important results: - A homogeneous 5 percent admixture of minor actinides (MA) from N4-PWRs to EFR fuel would allow a transmutation not only of the EFR MA, but in addition of the MA from 5 or 6 PWRs of equal power. However, the incineration is restricted by safety considerations. - LWR have only a very low MA incineration potential, due to their disadvantageous neutron capture/fission ratio. - In order to keep the Cm inventory at a low level, it is advantageous to concentrate the Am heterogeneously in particular fuel elements or rods. (orig./HP)

  18. Premiering SAFE for Safety Added Fuel Element - 15020

    International Nuclear Information System (INIS)

    Bhowmik, P.K.; Shamim, J.A.; Suh, K.Y.; Suh, K.S.

    2015-01-01

    The impact of the Fukushima accident has been the willingness to implement passive safety measures in reactor design and to simplify reactor design itself. Within this framework, a new fuel element, named SAFE (Safety Added Fuel Element) based on the concept of accident tolerant fuel, is presented. SAFE is a new type of fuel element cooled internally and externally by light water and with stainless steel as the cladding material. The removal of boron may trigger a series of changes which may simplify the system greatly. A simplified thermal analysis of SAFE shows that the fuel centerline temperature is well below the maximal limit during the normal operation of the plant

  19. Overview of symposium on storage of spent fuel from power reactors

    International Nuclear Information System (INIS)

    Bonne, A.; Crijns, M.J.; Dyck, H.P.

    2001-01-01

    An International Symposium on Storage of Spent Fuel from Power Reactors was held in Vienna from 9-13 November 1998. The Symposium was organized by the International Atomic Energy Agency in co-operation with the OECD Nuclear Energy Agency. Of the one hundred sixty participants registered, one hundred twenty-five (including 3 observers) representing 35 countries and 4 international organizations, attended the Symposium. 20 participants from developing countries received Agency's grants. During 4 main Sessions, 44 oral presentations of papers were made and subsequent discussions held. At a poster session 13 papers were presented. This paper will give an overview of the Symposium. The Symposium gave an opportunity to exchange information on the state of art and prospects of spent fuel storage, to discuss the worldwide situation and the major factors influencing the national policies in this field and to identify the most important directions that national efforts and international co-operation in this area should take. It was obvious from the papers presented and the discussions that the handling and storage of spent fuel is continuously taking place safely. Dominant messages retrieved from the Symposium are that the primary spent fuel management solution for the next decades will be interim storage, the duration time of interim storage becomes longer than earlier anticipated and the storage facilities will have to be designed for receiving also spent fuel from advanced fuel cycle practices (i.e. high burnup and MOX spent fuel). It was noted that the handling and storage of spent fuel is a mature technology and meets the stringent safety requirements applicable in the different countries. The changes in nuclear policy and philosophy across the world, and practical considerations, have made interim storage a real necessity in the nuclear power industry. (author)

  20. Interim storage of radioactive waste packages

    International Nuclear Information System (INIS)

    1998-01-01

    This report covers all the principal aspects of production and interim storage of radioactive waste packages. The latest design solutions of waste storage facilities and the operational experiences of developed countries are described and evaluated in order to assist developing Member States in decision making and design and construction of their own storage facilities. This report is applicable to any category of radioactive waste package prepared for interim storage, including conditioned spent fuel, high level waste and sealed radiation sources. This report addresses the following issues: safety principles and requirements for storage of waste packages; treatment and conditioning methods for the main categories of radioactive waste; examples of existing interim storage facilities for LILW, spent fuel and high level waste; operational experience of Member States in waste storage operations including control of storage conditions, surveillance of waste packages and observation of the behaviour of waste packages during storage; retrieval of waste packages from storage facilities; technical and administrative measures that will ensure optimal performance of waste packages subject to various periods of interim storage

  1. Disposal facility data for the interim performance

    International Nuclear Information System (INIS)

    Eiholzer, C.R.

    1995-01-01

    The purpose of this report is to identify and provide information on the waste package and disposal facility concepts to be used for the low-level waste tank interim performance assessment. Current concepts for the low-level waste form, canister, and the disposal facility will be used for the interim performance assessment. The concept for the waste form consists of vitrified glass cullet in a sulfur polymer cement matrix material. The waste form will be contained in a 2 x 2 x 8 meter carbon steel container. Two disposal facility concepts will be used for the interim performance assessment. These facility concepts are based on a preliminary disposal facility concept developed for estimating costs for a disposal options configuration study. These disposal concepts are based on vault type structures. None of the concepts given in this report have been approved by a Tank Waste Remediation Systems (TWRS) decision board. These concepts will only be used in th interim performance assessment. Future performance assessments will be based on approved designs

  2. 33 CFR 385.38 - Interim goals.

    Science.gov (United States)

    2010-07-01

    ..., monitoring and assessment; (ii) Be provided to the independent scientific review panel established in.... The interim goals shall be developed through the use of appropriate models and tools and shall provide... to be required to meet long-term hydrological and ecological restoration goals, based on best...

  3. Interim Storage Facility decommissioning. Final report

    International Nuclear Information System (INIS)

    Johnson, R.P.; Speed, D.L.

    1985-01-01

    Decontamination and decommissioning of the Interim Storage Facility were completed. Activities included performing a detailed radiation survey of the facility, removing surface and imbedded contamination, excavating and removing the fuel storage cells, restoring the site to natural conditions, and shipping waste to Hanford, Washington, for burial. The project was accomplished on schedule and 30% under budget with no measurable exposure to decommissioning personnel

  4. 19 CFR 354.8 - Interim sanctions.

    Science.gov (United States)

    2010-04-01

    ... reconsider imposition of interim sanctions on the basis of new and material evidence or other good cause... Secretary may petition a presiding official to impose such sanctions. (b) The presiding official may impose... person to return material previously provided by the Department and all other materials containing the...

  5. Automotive Mechanics Occupational Performance Survey. Interim Report.

    Science.gov (United States)

    Borcher, Sidney D.; Leiter, Paul B.

    The purpose of this federally-funded interim report is to present the results of a task inventory analysis survey of automotive mechanics completed by project staff within the Instructional Systems Design Program at the Center for Vocational and Technical Education. Intended for use in curriculum development for vocational education programs in…

  6. A Non-Traditional Interim Project.

    Science.gov (United States)

    Brown, Diane; Ward, Dorothy

    1980-01-01

    Describes a project initiated by the Foreign Language Department of Birmingham-Southern College for their Interim term and discusses an interdisciplinary course focusing on Medieval Europe. The course included presentations on German and French language and literature, as well as lectures on the arts, philosophy, and family life of the period.…

  7. 340 waste handling facility interim safety basis

    Energy Technology Data Exchange (ETDEWEB)

    VAIL, T.S.

    1999-04-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people.

  8. ITER interim design report package documents

    International Nuclear Information System (INIS)

    1996-01-01

    This publication contains the Excerpt from the ITER Council (IC-8), the ITER Interim Design Report, Cost Review and Safety Analysis, ITER Site Requirements and ITER Site Design Assumptions and the Excerpt from the ITER Council (IC-9). 8 figs, 2 tabs

  9. LANDFILL BIOREACTOR PERFORMANCE, SECOND INTERIM REPORT

    Science.gov (United States)

    A bioreactor landfill is a landfill that is operated in a manner that is expected to increase the rate and extent of waste decomposition, gas generation, and settlement compared to a traditional landfill. This Second Interim Report was prepared to provide an interpretation of fie...

  10. 340 waste handling facility interim safety basis

    International Nuclear Information System (INIS)

    VAIL, T.S.

    1999-01-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people

  11. Interim Storage of Plutonium in Existing Facilities

    International Nuclear Information System (INIS)

    Woodsmall, T.D.

    1999-01-01

    'In this era of nuclear weapons disarmament and nonproliferation treaties, among many problems being faced by the Department of Energy is the safe disposal of plutonium. There is a large stockpile of plutonium at the Rocky Flats Environmental Technology Center and it remains politically and environmentally strategic to relocate the inventory closer to a processing facility. Savannah River Site has been chosen as the final storage location, and the Actinide Packaging and Storage Facility (APSF) is currently under construction for this purpose. With the ability of APSF to receive Rocky Flats material an estimated ten years away, DOE has decided to use the existing reactor building in K-Area of SRS as temporary storage to accelerate the removal of plutonium from Rocky Flats. There are enormous cost savings to the government that serve as incentive to start this removal as soon as possible, and the KAMS project is scheduled to receive the first shipment of plutonium in January 2000. The reactor building in K-Area was chosen for its hardened structure and upgraded seismic qualification, both resulting from an effort to restart the reactor in 1991. The KAMS project has faced unique challenges from Authorization Basis and Safety Analysis perspectives. Although modifying a reactor building from a production facility to a storage shelter is not technically difficult, the nature of plutonium has caused design and safety analysis engineers to make certain that the design of systems, structures and components included will protect the public, SRS workers, and the environment. A basic overview of the KAMS project follows. Plutonium will be measured and loaded into DOT Type-B shipping packages at Rocky Flats. The packages are 35-gallon stainless steel drums with multiple internal containment boundaries. DOE transportation vehicles will be used to ship the drums to the KAMS facility at SRS. They will then be unloaded, stacked and stored in specific locations throughout the

  12. Gas cooled reactor decommissioning. Packaging of waste for disposal in the United Kingdom deep repository

    International Nuclear Information System (INIS)

    Barlow, S.V.; Wisbey, S.J.; Wood, P.

    1998-01-01

    United Kingdom Nirex Limited has been established to develop and operate a deep underground repository for the disposal of the UK's intermediate and certain low level radioactive waste. The UK has a significant Gas Cooled Reactor (GCR) programme, including both Magnox and AGR (Advanced Gas-cooled Reactor) capacity, amounting to 26 Magnox reactors, 15 AGR reactors as well as research and prototype reactor units such as the Windscale AGR and the Windscale Piles. Some of these units are already undergoing decommissioning and Nirex has estimated that some 15,000 m 3 (conditioned volume) will come forward for disposal from GCR decommissioning before 2060. This volume does not include final stage (Stage 3) decommissioning arisings from commercial reactors since the generating utilities in the UK are proposing to adopt a deferred safe store strategy for these units. Intermediate level wastes arising from GCR decommissioning needs to be packaged in a form suitable for on-site interim storage and eventual deep disposal in the planned repository. In the absence of Conditions for Acceptance for a repository in the UK, the dimensions, key features and minimum performance requirements for waste packages are defined in Waste Package Specifications. These form the basis for all assessments of the suitability of wastes for disposal, including GCR wastes. This paper will describe the nature and characteristics of GCR decommissioning wastes which are intended for disposal in a UK repository. The Nirex Waste Package Specifications and the key technical issues, which have been identified when considering GCR decommissioning waste against the performance requirements within the specifications, are discussed. (author)

  13. 40 Years of Experience of NIRAS / Belgoprocess on the Interim Storage of Low, Intermediate and High Level Waste

    International Nuclear Information System (INIS)

    Braeckeveldt, Marnix; Ghys, Bart

    2016-01-01

    Conclusion: • ONDRAF/NIRAS and Belgoprocess have gained over time an extended experience on the interim storage of Low-Intermediate and High level waste. • An systematic inspection strategy was developed in order the verify the conformity of the different waste-packages and corrective measures were taken to guarantee safe storage conditions. • From 2022 , ONDRAF/NIRAS will operate a surface disposal facility for LLW

  14. Fast Flux Test Facility interim examination and maintenance cell - past, present, and future

    International Nuclear Information System (INIS)

    Vincent, J.R.

    1990-01-01

    The Fast Flux Test Facility (FFTF) interim examination and maintenance (IEM) cell was designed to perform interim examination and/or disassembly of experimental core components for final analysis elsewhere, as well as maintenance of sodium-wetted or neutron-activated internal reactor parts and plant support hardware. The first 10 yr of operation were mainly devoted to the disassembly and examination of core component test assemblies. While some maintenance was performed on reactor support equipment, such as the closed-loop ex-vessel machine (CLEM) sodium-wetted grapple, 90% of IEM cell availability has been devoted to core component tests. Some test assemblies originally considered for processing in the IEM cell have not been irradiated; others, not originally planned, have been designed, irradiated, and processed. While no major reactor equipment has required remote repair or maintenance, the IEM cell has served as the remote repair facility for its own in-cell equipment, and several innovative remote repairs have been accomplished and are described

  15. Safe Sleep for Babies

    Science.gov (United States)

    ... 5 MB] Read the MMWR Science Clips Safe Sleep for Babies Eliminating hazards Recommend on Facebook Tweet ... Page Problem Every year, there are thousands of sleep-related deaths among babies. View large image and ...

  16. Safe operating envelope

    Energy Technology Data Exchange (ETDEWEB)

    Oliva, N [Ontario Hydro, Toronto, ON (Canada)

    1997-12-01

    Safe Operating Envelope is described representing: The outer bound of plant conditions within which day-to-day plant operation must be maintained in order to comply with regulatory requirements, associated safety design criteria and corporate nuclear safety goals. Figs.

  17. The first safe country

    Directory of Open Access Journals (Sweden)

    Raffaela Puggioni

    2012-06-01

    Full Text Available The Dublin II Regulation makes the first safe country of refuge solelyresponsible for refugees and asylum seekers. In the case of Italy, thefirst responsible country has not been acting responsibly.

  18. Buying & Using Medicine Safely

    Science.gov (United States)

    ... Reducers Safe Daily Use of Aspirin Medication Health Fraud Resources for You FDA Consumer Updates (Drugs) Page ... feeds Follow FDA on Twitter Follow FDA on Facebook View FDA videos on YouTube View FDA photos ...

  19. Karate: Keep It Safe.

    Science.gov (United States)

    Jordan, David

    1981-01-01

    Safety guidelines for each phase of a karate practice session are presented to provide an accident-free and safe environment for teaching karate in a physical education or traditional karate training program. (JMF)

  20. Safe operating envelope

    International Nuclear Information System (INIS)

    Oliva, N.

    1997-01-01

    Safe Operating Envelope is described representing: The outer bound of plant conditions within which day-to-day plant operation must be maintained in order to comply with regulatory requirements, associated safety design criteria and corporate nuclear safety goals. Figs

  1. Safe driving for teens

    Science.gov (United States)

    Driving and teenagers; Teens and safe driving; Automobile safety - teenage drivers ... months before taking friends as passengers. Teenage-related driving deaths occur more often in certain conditions. OTHER SAFETY TIPS FOR TEENS Reckless driving is still a ...

  2. Removing Hair Safely

    Science.gov (United States)

    ... For Consumers Home For Consumers Consumer Updates Removing Hair Safely Share Tweet Linkedin Pin it More sharing ... related to common methods of hair removal. Laser Hair Removal In this method, a laser destroys hair ...

  3. Medications: Using Them Safely

    Science.gov (United States)

    ... to Safely Give Ibuprofen Childproofing and Preventing Household Accidents Medicines for Diabetes Complementary and Alternative Medicine How Do Pain Relievers Work? What Medicines Are and What They Do Medicines ...

  4. DroidSafe

    Science.gov (United States)

    2016-12-01

    Massachusetts Avenue, Build E19-750 Cambridge , MA 02139-4307 8. PERFORMING ORGANIZATION REPORT NUMBER 9. SPONSORING/MONITORING AGENCY NAME(S) AND ADDRESS...Activity objects illustrating the challenges of points-to and information flow analysis...measure how many malicious flows Droid- Safe was able to detect). As these results illustrate , DroidSafe implements an analysis of unprece- dented

  5. Safe biodegradable fluorescent particles

    Science.gov (United States)

    Martin, Sue I [Berkeley, CA; Fergenson, David P [Alamo, CA; Srivastava, Abneesh [Santa Clara, CA; Bogan, Michael J [Dublin, CA; Riot, Vincent J [Oakland, CA; Frank, Matthias [Oakland, CA

    2010-08-24

    A human-safe fluorescence particle that can be used for fluorescence detection instruments or act as a safe simulant for mimicking the fluorescence properties of microorganisms. The particle comprises a non-biological carrier and natural fluorophores encapsulated in the non-biological carrier. By doping biodegradable-polymer drug delivery microspheres with natural or synthetic fluorophores, the desired fluorescence can be attained or biological organisms can be simulated without the associated risks and logistical difficulties of live microorganisms.

  6. Recent progress on developments of tritium safe handling techniques in Japan

    International Nuclear Information System (INIS)

    Watanabe, Kuniaki; Matsuyama, Masao

    1993-01-01

    Vast amounts of tritium will be used for thermonuclear fusion reactors. Without establishing safe handling techniques for large amounts of tritium, undoubtedly the fusion reactors will not be accepted. Japanese activity on tritium related research has considerably developed in the last 10 years. This review paper gives a brief summary of safe handling techniques developed by Japanese research groups. (author)

  7. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

  8. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

  9. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    International Nuclear Information System (INIS)

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future

  10. Safety Case for Safe-store

    International Nuclear Information System (INIS)

    Woollam, Paul B.

    2002-01-01

    Magnox Electric plc (Magnox), a wholly owned subsidiary of BNFL, owns 26 gas-cooled, graphite-moderated units on 11 sites in the UK. Eight units have been permanently shutdown and the remainder will shut this decade in a currently declared closure programme. The first of these reactors went to power in 1952 and the fleet has generated typically 9% of the UK's electricity during the last five decades. In accordance with UK Government policy, BNFL aims for a systematic and progressive reduction in hazards on its decommissioning sites. The end-point of the decommissioning process is that the reactors will be dismantled and their sites de-licensed. This will be done through minimising both the risks to the public, workers and the environment and also the lifetime cost, consistent with world class safety. There will be passive safe storage during deferment periods and it is BNFL's clear intent that the reactors will not be Safe-stored indefinitely. The main hazard associated with any decommissioned nuclear site is the spent fuel. Hence the reactors will be de-fuelled as soon as practicable after shutdown. After this work is complete, Cs-137 contaminated plant (e.g. fuel pools, effluent plant, and drains) will be dismantled when it is no longer needed. All other plant and buildings will also be dismantled when they are no longer needed, except for the reactor buildings which will be put into passive safe storage. Co-60 contaminated plant, such as steam generators, will be dismantled with the reactors. The reactors will be dismantled in a sequenced programme, with a notional start time around 100 years from shutdown. Magnox Electric is ensuring that the reactors and primary circuits on all its sites are well characterised. We have carried out a detailed, peer reviewed hazard identification on the lead site from which we have generated a rolling 25-year basic safety case. We have then searched for cliff edge effects and possible long-term changes to generate the 100-year

  11. Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

    International Nuclear Information System (INIS)

    1986-11-01

    Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

  12. Expansion of storage capacity of interim spent fuel storage (MSVP) Bohunice

    International Nuclear Information System (INIS)

    Pilat, P.; Fridrich, V.

    2005-01-01

    This article describes modifications of Interim spent fuel storage, performed with aim of storage capacity expansion, seismic stability enhancement, and overall increase of service life as well as assuring of MSVP safe operation. Uniqueness of adopted technical solutions is based upon the fact that mentioned innovations and modifications were performed without any changes, neither in ground plan nor architecture of MSVP structure. It also important to mention that all modifications were performed during continual operation of MSVP without any breaks of limits or operational regulations. Reconstruction and innovation of existing construction and technological systems of MSVP has assured required quality standard comparable with actual trends. (authors)

  13. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  14. Pathways to deep decarbonization - Interim 2014 Report

    International Nuclear Information System (INIS)

    2014-01-01

    The interim 2014 report by the Deep Decarbonization Pathways Project (DDPP), coordinated and published by IDDRI and the Sustainable Development Solutions Network (SDSN), presents preliminary findings of the pathways developed by the DDPP Country Research Teams with the objective of achieving emission reductions consistent with limiting global warming to less than 2 deg. C. The DDPP is a knowledge network comprising 15 Country Research Teams and several Partner Organizations who develop and share methods, assumptions, and findings related to deep decarbonization. Each DDPP Country Research Team has developed an illustrative road-map for the transition to a low-carbon economy, with the intent of taking into account national socio-economic conditions, development aspirations, infrastructure stocks, resource endowments, and other relevant factors. The interim 2014 report focuses on technically feasible pathways to deep decarbonization

  15. Gaz de France interim financial report 2007

    International Nuclear Information System (INIS)

    2007-01-01

    This financial report contains the unaudited condensed financial statements of Gaz de France Group for the first half ended June 30, 2007, which were reviewed by the audit committee on August 27, 2007 and by the board of directors at its meeting on August 28, 2007. It includes forward-looking statements concerning the objectives, strategies, financial position, future operating results and the operations of Gaz de France Group. These statements reflect the Group's current perception of its activities and the markets in which it operates, as well as various estimates and assumptions considered to be reasonable. Content: interim management report (highlights of the first half of 2007, revenues and results for the period, financial structure, data on outstanding stock, outlook); interim consolidated financial statements (consolidated statements of income, consolidated balance sheets, consolidated statements of cash flows, recognized income and expenses, statements of changes in shareholders' equity, note to the consolidated financial statements); statement by the person responsible for the interim financial report; statutory auditors' report. (J.S.)

  16. Tribal child welfare. Interim final rule.

    Science.gov (United States)

    2012-01-06

    The Administration for Children and Families (ACF) is issuing this interim final rule to implement statutory provisions related to the Tribal title IV-E program. Effective October 1, 2009, section 479B(b) of the Social Security Act (the Act) authorizes direct Federal funding of Indian Tribes, Tribal organizations, and Tribal consortia that choose to operate a foster care, adoption assistance and, at Tribal option, a kinship guardianship assistance program under title IV-E of the Act. The Fostering Connections to Success and Increasing Adoptions Act of 2008 requires that ACF issue interim final regulations which address procedures to ensure that a transfer of responsibility for the placement and care of a child under a State title IV-E plan to a Tribal title IV-E plan occurs in a manner that does not affect the child's eligibility for title IV-E benefits or medical assistance under title XIX of the Act (Medicaid) and such services or payments; in-kind expenditures from third-party sources for the Tribal share of administration and training expenditures under title IV-E; and other provisions to carry out the Tribal-related amendments to title IV-E. This interim final rule includes these provisions and technical amendments necessary to implement a Tribal title IV-E program.

  17. Safeness of radiological machinery

    International Nuclear Information System (INIS)

    Yokoyama, Shun

    1979-01-01

    The human factors affecting the safeness of radiological machinery, which are often very big and complicated machines, are described from the stand point of handling. 20 to 50% of the troubles on equipments seem to be caused by men. This percentage will become even higher in highly developed equipments. Human factors have a great influence on the safeness of radiological equipments. As the human factors, there are sensory factors and knowledge factors as well as psychological factors, and the combination of these factors causes mishandling and danger. Medical services at present are divided in various areas, and consist of the teamwork of the people in various professions. Good human relationship, education and control are highly required to secure the safeness. (Kobatake, H.)

  18. The Interim Financial Statements: The Case of Greece

    OpenAIRE

    Rogdaki, E.I.; Kazantzis, Ch.

    1999-01-01

    The following paper refers to the accounting and auditing issues which emerge in the preparation of the interim financial statements of the companies: Firstly, the interim financial statements are defined as being the financial statements that provide useful information about the financial position and the financial results of a company which are realized and accrued during the fiscal year. The interim financial statements can be prepared on a monthly basis, on a quarterly basis or covering a...

  19. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

    Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

  20. Interim Storage of Spent Nuclear Fuel before Final Disposal in Germany - Regulator's view

    International Nuclear Information System (INIS)

    Arens, G.; Goetz, Ch.; Geupel, Sandra; Gmal, B.; Mester, W.

    2014-01-01

    For spent nuclear fuel management in Germany the concept of dry interim storage in dual purpose casks before direct disposal is applied. The Federal Office for Radiation Protection (BfS) is the competent authority for licensing of interim storage facilities. The competent authority for surveillance of operation is the responsible authority of the respective federal state (Land). Currently operation licenses for storage facilities have been granted for a storage time of 40 years and are based on safety demonstrations for all safety issues as safe enclosure, shielding, sub-criticality and decay heat removal under consideration of operation conditions. In addition, transportability of the casks for the whole storage period has to be provided. Due to current delay in site selection and exploration of a disposal site, an extension of the storage time beyond 40 years could be needed. This will cause appropriate actions by the licensee and the competent authorities as well. A brief description of the regulatory base of licensing and surveillance of interim storage is given from the regulators view. Furthermore the current planning for final disposal of spent nuclear fuel and high level waste and its interconnections between storage and disposal concepts are shortly explained. Finally the relevant aspects for licensing of extended storage time beyond 40 years will be discussed. Current activities on this issue, which have been initiated by the Federal Government, will be addressed. On the regulatory side a review and amendment of the safety guideline for interim storage of spent fuel has been performed and the procedure of periodic safety review is being implemented. A guideline for implementing an ageing management programme is available in a draft version. Regarding safety of long term storage a study focussing on the identification and evaluation of long term effects as well as gaps of knowledge has been finished in 2010. A continuation and update is currently underway

  1. Training and Certification of Research Reactor Personnel

    International Nuclear Information System (INIS)

    Zarina Masood

    2011-01-01

    The safe operation of a research reactor requires that reactor personnel be fully trained and certified by the relevant authorities. Reactor operators at PUSPATI TRIGA Reactor underwent extensive training and are certified, ever since the reactor first started its operation in 1982. With the emphasis on enhancing reactor safety in recent years, reactor operator training and certification have also evolved. This paper discusses the changes that have to be implemented and the challenges encountered in developing a new training programme to be in line with the national standards. (author)

  2. Compilation of interim technical research memoranda. Volume I

    International Nuclear Information System (INIS)

    Shanahan, W.R.

    1984-04-01

    Four interim technical research memoranda are presented that describe the results of numerical simulations designed to investigate the dynamics of energetic plasma beams propagating across magnetic fields

  3. Conceptual design of interim storage facility for CNAI

    International Nuclear Information System (INIS)

    Fuenzalida Troyano, Carlos S.; Bergallo, Juan E.; Nassini, Horacio E.P.; Blanco, Anibal; Delmastro, Dario F.

    2007-01-01

    The reduced storage capacity available in the two spent fuel pools of argentine PHWR Atucha-1 power plant, the current plans for extending the reactor operation beyond its design lifetime, and the government decision on Atucha-2 NPP construction ending, have motivated the evaluation of a dry storage option for the interim management of spent fuel assemblies. Two different designs are presently being analyzed by an expert working group, from both technical and economical points of views. Authors are proposing a modular system consisting of an arrangement of reinforced concrete structures into which welded metallic canisters loaded with 37 spent fuel assemblies each stored in horizontal position. The reinforced concrete module is designed to provide the necessary physical protection and biological shielding to the loaded canisters during long-term storage, as well as passive means to remove the spent fuel decay heat by a combination of radiation, conduction and natural air convection. In this works are presented advances in the conceptual designs for a spent nuclear fuel system to Atucha I nuclear power plant. (author) [es

  4. Safety aspects of spent nuclear fuel interim storage installations

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, Luiz Sergio [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil). Dept. da Qualidade. Div. de Sistemas da Qualidade]. E-mail: romanato@ctmsp.mar.mil.br; Rzyski, Barbara Maria [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Div. de Ensino]. E-mail: bmrzyski@ipen.br

    2007-07-01

    Nowadays safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. In this kind of storage it is necessary to consider the physical security. Nuclear installations have become more vulnerable. New types of accidents must be considered in the design of these installations, which in the early days were not considered like: fissile material stolen, terrorists' acts and war conflicts, and traditional accidents concerning the transport of the spent fuel from the reactor to the storage location, earthquakes occurrence, airplanes crash, etc. Studies related to airplane falling had showed that a collision of big commercials airplanes at velocity of 800 km/h against SNF storage and specially designed concrete casks, do not result in serious structural injury to the casks, and not even radionuclides liberation to the environment. However, it was demonstrated that attacks with modern military ammunitions, against metallic casks, are calamitous. The casks could not support a direct impact of this ammo and the released radioactive materials can expose the workers and public as well the local environment to harmful radiation. This paper deals about the main basic aspects of a dry SNF storage installation, that must be physically well protected, getting barriers that difficult the access of unauthorized persons or vehicles, as well as, must structurally resist to incidents or accidents caused by unauthorized intrusion. (author)

  5. An interim report on the Zenith Exponential Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Absalom, R M; Cameron, I R; Kinchin, G H; Sanders, J E; Wilson, D J [Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1959-06-15

    The following memorandum gives an interim account of the exponential experiments with Zenith-type fuel elements being carried out at Winfrith. Results quoted are still subject to revision: however it is hoped that the description of the work at this stage will stimulate discussion and suggestions for further measurements before the experiment is dismantled later in the year. The measurements are being undertaken in order to form some initial understanding of the reactor physics of uranium 235-thorium-graphite systems of the type later to be studied in Zenith. There have been no previous investigations of this type of system in the U.K. though measurements on enriched uranium-graphite systems have been reported from the U.S.A.. A practical result of the measurements will be a revision of the estimated critical loadings for Zenith, since the exponential systems studied cover the range of loadings proposed for the first critical assemblies. The theoretical work on these systems includes a two-group analysis being carried on in the Zenith group and a multigroup analysis being made by the H.T.G.C. Technical Assessments Group, including a Monte Carlo study of resonance capture.

  6. Safety aspects of spent nuclear fuel interim storage installations

    International Nuclear Information System (INIS)

    Romanato, Luiz Sergio

    2007-01-01

    Nowadays safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. In this kind of storage it is necessary to consider the physical security. Nuclear installations have become more vulnerable. New types of accidents must be considered in the design of these installations, which in the early days were not considered like: fissile material stolen, terrorists' acts and war conflicts, and traditional accidents concerning the transport of the spent fuel from the reactor to the storage location, earthquakes occurrence, airplanes crash, etc. Studies related to airplane falling had showed that a collision of big commercials airplanes at velocity of 800 km/h against SNF storage and specially designed concrete casks, do not result in serious structural injury to the casks, and not even radionuclides liberation to the environment. However, it was demonstrated that attacks with modern military ammunitions, against metallic casks, are calamitous. The casks could not support a direct impact of this ammo and the released radioactive materials can expose the workers and public as well the local environment to harmful radiation. This paper deals about the main basic aspects of a dry SNF storage installation, that must be physically well protected, getting barriers that difficult the access of unauthorized persons or vehicles, as well as, must structurally resist to incidents or accidents caused by unauthorized intrusion. (author)

  7. How the University of Texas system responded to the need for interim storage of low-level radioactive waste materials.

    Science.gov (United States)

    Emery, Robert J

    2012-11-01

    Faced with the prospect of being unable to permanently dispose of low-level radioactive wastes (LLRW) generated from teaching, research, and patient care activities, component institutions of the University of Texas System worked collaboratively to create a dedicated interim storage facility to be used until a permanent disposal facility became available. Located in a remote section of West Texas, the University of Texas System Interim Storage Facility (UTSISF) was licensed and put into operation in 1993, and since then has provided safe and secure interim storage for up to 350 drums of dry solid LLRW at any given time. Interim storage capability provided needed relief to component institutions, whose on-site waste facilities could have possibly become overburdened. Experiences gained from the licensing and operation of the site are described, and as a new permanent LLRW disposal facility emerges in Texas, a potential new role for the storage facility as a surge capacity storage site in times of natural disasters and emergencies is also discussed.

  8. Fabrication and operational experience with the interim storage cask

    International Nuclear Information System (INIS)

    Scott, P.L.

    1998-01-01

    This paper discusses the fabrication and operational experience of the Interim Storage Cask (ISC). The ISC is a dry storage cask which is used to safely store a Core Component Container (CCC) containing up to seven Fast Flux Test Facility (FFTF) spent fuel assemblies at the US Department of Energy's Hanford Site. Under contract to B and W Hanford Company (BWHC), General Atomics (GA) designed and fabricated thirty ISC casks which BWHC is remotely loading at the FFTF facility. BWHC designed and fabricated the CCCS. As of December 1997, thirty ISCs have been fabricated, of which eighteen have been loaded and moved to a storage site adjacent to the FFTF facility. Fabrication consisted of three sets of casks. The first unit was completed and acceptance tested before any other units were fabricated. After the first unit passed all acceptance tests, nine more units were fabricated in the first production run. Before those nine units were completed, GA began a production run of twenty more units. The paper provides an overview of the cask design and discusses the problems encountered in fabrication, their resolution, and changes made in the fabrication processes to improve the quality of the casks. The paper also discusses the loading process and operational experiences with loading and handling of the casks. Information on loading times, worker dose exposure, and total dose for loading are presented

  9. Developing new transportable storage casks for interim dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, K.; Iwasa, K.; Araki, K.; Asano, R. [Hitachi Zosen Diesel and Engineering Co., Ltd., Tokyo (Japan)

    2004-07-01

    Transportable storage metal casks are to be consistently used during transport and storage for AFR interim dry storage facilities planning in Japan. The casks are required to comply with the technical standards of regulations for both transport (hereinafter called ''transport regulation'') and storage (hereafter called ''storage regulation'') to maintain safety functions (heat transfer, containment, shielding and sub-critical control). In addition to these requirements, it is not planned in normal state to change the seal materials during storage at the storage facility, therefore it is requested to use same seal materials when the casks are transported after storage period. The dry transportable storage metal casks that satisfy the requirements have been developed to meet the needs of the dry storage facilities. The basic policy of this development is to utilize proven technology achieved from our design and fabrication experience, to carry out necessary verification for new designs and to realize a safe and rational design with higher capacity and efficient fabrication.

  10. Immunotherapy for recurrent malignant glioma: an interim report on survival.

    Science.gov (United States)

    Ingram, M; Buckwalter, J G; Jacques, D B; Freshwater, D B; Abts, R M; Techy, G B; Miyagi, K; Shelden, C H; Rand, R W; English, L W

    1990-12-01

    We present interim survival data for a group of 83 adult patients with recurrent malignant glioma treated by implanting stimulated autologous lymphocytes into the tumour bed following surgical debulking. The patients were treated 6 months or more prior to data analysis. Fifty-nine patients were male and 24 female. The mean age for the entire group was 48.4 years and the mean Karnofsky rating (KR) was 67.2. Eight of the patients had grade II tumours, 33 had grade III tumours and 42 had grade IV tumours. Statistical analysis focuses on tumour grade, KR and patient age, factors that have been shown to affect survival in previous studies. Multifactorial analyses are employed to identify interrelationships among factors related to survival. Seven patients (8%) did not respond to immunotherapy, 76 (92%) had a good initial response. Twenty-five patients (30.1%) are living and 18 (22%) have shown no evidence of recurrence. Results are evaluated in the light of those obtained in trials of other experimental therapies for recurrent malignant gliomas. It is concluded that the present protocol offers a safe and comparatively effective treatment option.

  11. High Temperature Materials Interim Data Qualification Report FY 2011

    International Nuclear Information System (INIS)

    Lybeck, Nancy

    2011-01-01

    Projects for the very high temperature reactor (VHTR) Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the Next Generation Nuclear Plant (NGNP) Data Management and Analysis System (NDMAS) to ensure that VHTR data are qualified for use, stored in a readily accessible electronic form, and analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the High Temperature Materials characterization data stream, the processing of these data within NDMAS, and reports the interim fiscal year (FY) 2011 qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. The High Temperature Materials data are being collected under the Nuclear Quality Assurance (NQA)-1 guidelines and will be qualified data. For NQA-1 qualified data, the qualification activities include: (1) capture testing to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing to confirm that the data are an accurate representation of the system or object being measured, and (3) documenting that the data were collected under an NQA-1 or equivalent Quality Assurance program. Currently, data from seven test series within the High Temperature Materials data stream have been entered into the NDMAS vault, including tensile tests, creep tests, and cyclic tests. Of the 5,603,682 records currently in the vault, 4,480,444 have been capture passed, and capture testing is in process for the remaining 1,123,238.

  12. Keeping Food Safe

    Centers for Disease Control (CDC) Podcasts

    2009-05-27

    This CDC Kidtastics podcast discusses things kids and parents can do to help prevent illness by keeping food safe.  Created: 5/27/2009 by National Center for Zoonotic, Vector-Borne, and Enteric Diseases (NCZVED).   Date Released: 5/27/2009.

  13. Effective and Safe Ships

    DEFF Research Database (Denmark)

    Pedersen, Preben Terndrup; Amdahl, Jørgen; Rutgersson, Olle

    1996-01-01

    A Joint Nordic Research project "Effecive and Safe Ships" is presented. The project is aiming to develop methods and tools for quantitative evaluation fo ship safety. This report is the report of the preliminary phase where the plan for the main project is developed. The objectives of the project...

  14. Are EU Banks Safe?

    NARCIS (Netherlands)

    R.J. Theissen (Roel)

    2013-01-01

    markdownabstract__Abstract__ What exactly are the rules banks are subject to, and are they fit for purpose? These are the two questions addressed in this book ‘Are EU banks safe?’ and its descriptive companion book ‘EU banking supervision’. The full rulebook on banks is difficult to find

  15. Safe medication management in specialized home healthcare - an observational study.

    Science.gov (United States)

    Lindblad, Marléne; Flink, Maria; Ekstedt, Mirjam

    2017-08-24

    Medication management is a complex, error-prone process. The aim of this study was to explore what constitutes the complexity of the medication management process (MMP) in specialized home healthcare and how healthcare professionals handle this complexity. The study is theoretically based in resilience engineering. Data were collected during the MMP at three specialized home healthcare units in Sweden using two strategies: observation of workplaces and shadowing RNs in everyday work, including interviews. Transcribed material was analysed using grounded theory. The MMP in home healthcare was dynamic and complex with unclear boundaries of responsibilities, inadequate information systems and fluctuating work conditions. Healthcare professionals adapted their everyday clinical work by sharing responsibility and simultaneously being authoritative and preserving patients' active participation, autonomy and integrity. To promote a safe MMP, healthcare professionals constantly re-prioritized goals, handled gaps in communication and information transmission at a distance by creating new bridging solutions. Trade-offs and workarounds were necessary elements, but also posed a threat to patient safety, as these interim solutions were not systematically evaluated or devised learning strategies. To manage a safe medication process in home healthcare, healthcare professionals need to adapt to fluctuating conditions and create bridging strategies through multiple parallel activities distributed over time, space and actors. The healthcare professionals' strategies could be integrated in continuous learning, while preserving boundaries of safety, instead of being more or less interim solutions. Patients' and family caregivers' as active partners in the MMP may be an underestimated resource for a resilient home healthcare.

  16. The safe home project.

    Science.gov (United States)

    Arphorn, Sara; Jiraniratisai, Sopaphan; Rungtakul, Rungsri; Phutta, Nikom

    2011-12-01

    The Thai Health Promotion Foundation supported the Improvement of Quality of Life of Informal Workers project in Ban Luang District, Amphur Photaram, Ratchaburi Province. There were many informal workers in Ban Luang District. Sweet-crispy fish producers in Ban Luang were the largest group among the sweet-crispy fish producers in Thailand. This project was aimed at improving living and working conditions of informal workers, with a focus on the sweet-crispy fish group. Good practices of improved living and working conditions were used to help informal workers build safe, healthy and productive work environments. These informal workers often worked in substandard conditions and were exposed to various hazards in the working area. These hazards included risk of exposure to hot work environment, ergonomics-related injuries, chemical hazards, electrical hazards etc. Ergonomics problems were commonly in the sweet-crispy fish group. Unnatural postures such as prolonged sitting were performed dominantly. One hundred and fifty informal workers participated in this project. Occupational health volunteers were selected to encourage occupational health and safety in four groups of informal workers in 2009. The occupational health volunteers trained in 2008 were farmers, beauty salon workers and doll makers. The occupational health and safety knowledge is extended to a new informal worker group: sweet-crispy fish producer, in 2009. The occupational health and safety training for sweet-crispy fish group is conducted by occupational health volunteers. The occupational health volunteers increased their skills and knowledge assist in to make safe home and safe community through participatory oriented training. The improvement of living and working condition is conducted by using a modified WISH, Work Improvement for Safe Home, checklist. The plans of improvement were recorded. The informal workers showed improvement mostly on material handling and storage. The safe uses and safe

  17. The future of nuclear reactors

    International Nuclear Information System (INIS)

    Teller, E.

    1989-01-01

    The Atomic Energy Commission Advisory Committee on Reactor Safeguards began work in early 1948 with the firm and unanimous conviction that nuclear power could not survive a significant damaging accident. They as a committee felt that their job was to make reactors so safe that no such event would ever occur. However, ambitious reactor planners did not like all the buts and cautions that the committee was raising. They seemed to delay unduly their setting sail into the brave new world of clean, cheap, safe nuclear energy. The committee was soon nicknamed the Committee on Reactor Prevention. Reactors, of course, represented a tremendous step into the future. To an unprecedented extent, they were based on theory. But the committee did not have the luxury of putting a preliminary model into operation and waiting for difficulties to show up. In assessing new designs and developments, they had to anticipate future difficulties. Their proposals in good part were accepted, but their deep emphasis on safety did not become a part of the program. Today, forty years later, the author still believes both in the need for nuclear reactors and in the need of a thorough-going, pervasive emphasis on their safety. Real, understandable safety can be achieved, and that achievement is the key to our nuclear future. The details he gives are only examples. The need for reactors that are not only safe but obviously safe can be ignored only at our peril

  18. Conceptual design of an interim dry storage system for the Atucha nuclear power plant spent fuels

    International Nuclear Information System (INIS)

    Nassini, Horacio E.P.; Fuenzalida Troyano, C.S.; Bevilacqua, Arturo M.; Bergallo, Juan E.

    2005-01-01

    The Atucha I nuclear power station, after completing the rearrangement and consolidation of the spent fuels in the two existing interim wet storage pools, will have enough room for the storage of spent fuel from the operation of the reactor till December 2014. If the operation is extended beyond 2014, or if the reactor is decommissioned, it will be necessary to empty both pools and to transfer the spent fuels to a dry storage facility. This paper shows the progress achieved in the conceptual design of a dry storage system for Atucha I spent fuels, which also has to be adequate, without modifications, for the storage of fuels from the second unity of the nuclear power station, Atucha II, that is now under construction. (author) [es

  19. Advanced 4S (super safe, small and simple) LMR

    International Nuclear Information System (INIS)

    Minato, A.; Handa, N.

    2000-01-01

    This paper describes a new nuclear power system which can be used for a greater variety of applications. The 4S liquid metal reactor has high inherent safety and passive safety characteristics. It is also easy to operate, maintain and inspect, faster to construct, more flexible in location, requires less initial investment, and is better suited to electrical grid management. The reactor offers a new route through which to expand the use of safe nuclear technology in the world. (author)

  20. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    Gaber, F.A.; El Messiry, A.M.

    1988-01-01

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  1. Safe use of nanomaterials

    CERN Multimedia

    2013-01-01

    The use of nanomaterials  is on the increase worldwide, including at CERN. The HSE Unit has established a safety guideline to inform you of the main requirements for the safe handling and disposal of nanomaterials at CERN.   A risk assessment tool has also been developed which guides the user through the process of evaluating the risk for his or her activity. Based on the calculated risk level, the tool provides a list of recommended control measures.   We would therefore like to draw your attention to: Safety Guideline C-0-0-5 - Safe handling and disposal of nanomaterials; and Safety Form C-0-0-2 - Nanomaterial Risk Assessment   You can consult all of CERN’s safety rules and guidelines here. Please contact the HSE Unit for any questions you may have.   The HSE Unit

  2. Plutonium safe handling

    International Nuclear Information System (INIS)

    Tvehlov, Yu.

    2000-01-01

    The abstract, prepared on the basis of materials of the IAEA new leadership on the plutonium safe handling and its storage (the publication no. 9 in the Safety Reports Series), aimed at presenting internationally acknowledged criteria on the radiation danger evaluation and summarizing the experience in the safe management of great quantities of plutonium, accumulated in the nuclear states, is presented. The data on the weapon-class and civil plutonium, the degree of its danger, the measures for provision of its safety, including the data on accident radiation consequences with the fission number 10 18 , are presented. The recommendations, making it possible to eliminate the super- criticality danger, as well as ignition and explosion, to maintain the tightness of the facility, aimed at excluding the radioactive contamination and the possibility of internal irradiation, to provide for the plutonium security, physical protection and to reduce irradiation are given [ru

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  4. The Homestake Interim Laboratory and Homestake DUSEL

    Science.gov (United States)

    Lesko, Kevin T.

    2011-12-01

    The former Homestake gold mine in Lead South Dakota is proposed for the National Science Foundation's Deep Underground Science and Engineering Laboratory (DUSEL). The gold mine provides expedient access to depths in excess of 8000 feet below the surface (>7000 mwe). Homestake's long history of promoting scientific endeavours includes the Davis Solar Neutrino Experiment, a chlorine-based experiment that was hosted at the 4850 Level for more than 30 years. As DUSEL, Homestake would be uncompromised by competition with mining interests or other shared uses. The facility's 600-km of drifts would be available for conversion for scientific and educational uses. The State of South Dakota, under Governor Rounds' leadership, has demonstrated exceptionally strong support for Homestake and the creation of DUSEL. The State has provided funding totalling $46M for the preservation of the site for DUSEL and for the conversion and operation of the Homestake Interim Laboratory. Motivated by the strong educational and outreach potential of Homestake, the State contracted a Conversion Plan by world-recognized mine-engineering contractor to define the process of rehabilitating the facility, establishing the appropriate safety program, and regaining access to the facility. The State of South Dakota has established the South Dakota Science and Technology Authority to oversee the transfer of the Homestake property to the State and the rehabilitation and preservation of the facility. The Homestake Scientific Collaboration and the State of South Dakota's Science and Technology Authority has called for Letters of Interest from scientific, educational and engineering collaborations and institutions that are interested in hosting experiments and uses in the Homestake Interim Facility in advance of the NSF's DUSEL, to define experiments starting as early as 2007. The Homestake Program Advisory Committee has reviewed these Letters and their initial report has been released. Options for

  5. The Homestake Interim Laboratory and Homestake DUSEL

    International Nuclear Information System (INIS)

    Lesko, Kevin T.

    2011-01-01

    The former Homestake gold mine in Lead South Dakota is proposed for the National Science Foundation's Deep Underground Science and Engineering Laboratory (DUSEL). The gold mine provides expedient access to depths in excess of 8000 feet below the surface (>7000 mwe). Homestake's long history of promoting scientific endeavours includes the Davis Solar Neutrino Experiment, a chlorine-based experiment that was hosted at the 4850 Level for more than 30 years. As DUSEL, Homestake would be uncompromised by competition with mining interests or other shared uses. The facility's 600-km of drifts would be available for conversion for scientific and educational uses. The State of South Dakota, under Governor Rounds' leadership, has demonstrated exceptionally strong support for Homestake and the creation of DUSEL. The State has provided funding totalling $46M for the preservation of the site for DUSEL and for the conversion and operation of the Homestake Interim Laboratory. Motivated by the strong educational and outreach potential of Homestake, the State contracted a Conversion Plan by world-recognized mine-engineering contractor to define the process of rehabilitating the facility, establishing the appropriate safety program, and regaining access to the facility. The State of South Dakota has established the South Dakota Science and Technology Authority to oversee the transfer of the Homestake property to the State and the rehabilitation and preservation of the facility. The Homestake Scientific Collaboration and the State of South Dakota's Science and Technology Authority has called for Letters of Interest from scientific, educational and engineering collaborations and institutions that are interested in hosting experiments and uses in the Homestake Interim Facility in advance of the NSF's DUSEL, to define experiments starting as early as 2007. The Homestake Program Advisory Committee has reviewed these Letters and their initial report has been released. Options for

  6. First (Kick-Off) Research Coordination Meeting (RCM) of the IAEA Coordinated Research Project (CRP) on Sodium Properties and Safe Operation of Experimental Facilities in Support of the Development and Deployment of Sodium-Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    General information on recent IAEA activities in the field of fast reactors, including on-going and planned FR-related CRPs, as well as a review of the NAPRO CRP content, objectives, participants, planning and expected outcomes/deliverables, was provided. It was reminded that the NAPRO CRP is being organized in 3 main Work Packages (WP): • WP 1: collection, expert assessment and dissemination of consistent Na property data such as: surface tension, saturation vapor pressure and emissivity, thermo-dynamic behavior of ternary oxides in Na, and solubility - diffusivity of metallic impurities, etc. • WP 2: development of guidelines and best practices for Na facility design and operation, including fill and drain, purification, out-gassing prior to filling, Na storage, component handling, drying of Na piping after repair, etc. • WP 3: development of guidelines and best practices for Na facility safety, including prevention and mitigation of Na leaks, prevention and detection of Na fires, assessment of Na impact in the environment after accidental release, hydrogen hazards in cleaning facilities, etc

  7. 40 CFR 80.141 - Interim detergent gasoline program.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 16 2010-07-01 2010-07-01 false Interim detergent gasoline program. 80... (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Detergent Gasoline § 80.141 Interim detergent gasoline... apply to: (i) All gasoline sold or transferred to a party who sells or transfers gasoline to the...

  8. 50 CFR 660.720 - Interim protection for sea turtles.

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 9 2010-10-01 2010-10-01 false Interim protection for sea turtles. 660.720 Section 660.720 Wildlife and Fisheries FISHERY CONSERVATION AND MANAGEMENT, NATIONAL OCEANIC AND... Migratory Fisheries § 660.720 Interim protection for sea turtles. (a) Until the effective date of §§ 660.707...

  9. 12 CFR 541.18 - Interim Federal savings association.

    Science.gov (United States)

    2010-01-01

    ... an existing savings and loan holding company or to facilitate any other transaction the Office may... 12 Banks and Banking 5 2010-01-01 2010-01-01 false Interim Federal savings association. 541.18... REGULATIONS AFFECTING FEDERAL SAVINGS ASSOCIATIONS § 541.18 Interim Federal savings association. The term...

  10. Transuranic waste storage and assay facility (TRUSAF) interim safety basis

    International Nuclear Information System (INIS)

    Gibson, K.D.

    1995-09-01

    The TRUSAF ISB is based upon current facility configuration and procedures. The purpose of the document is to provide the basis for interim operation or restrictions on interim operations and the authorization basis for the TRUSAF at the Hanford Site. The previous safety analysis document TRUSAF hazards Identification and Evaluation (WHC 1977) is superseded by this document

  11. Presidential Transition: The Experience of Two Community College Interim Presidents

    Science.gov (United States)

    Thompson, Matthew D.

    2010-01-01

    The purpose of this qualitative case study was to understand the experiences of two community college interim presidents, their characteristics, and how they led institutions following an abrupt presidential departure. There were two fundamental questions framing this research study, 1. How do two interim community college presidents lead…

  12. Decision on performing interim analysis for comparative clinical trials.

    Science.gov (United States)

    Pak, Kyongsun; Jacobus, Susanna; Uno, Hajime

    2017-09-01

    In randomized-controlled trials, interim analyses are often planned for possible early trial termination to claim superiority or futility of a new therapy. While unblinding is necessary to conduct the formal interim analysis in blinded studies, blinded data also have information about the potential treatment difference between the groups. We developed a blinded data monitoring tool that enables investigators to predict whether they observe such an unblinded interim analysis results that supports early termination of the trial. Investigators may skip some of the planned interim analyses if an early termination is unlikely. We specifically focused on blinded, randomized-controlled studies to compare binary endpoints of a new treatment with a control. Assuming one interim analysis is planned for early termination for superiority or futility, we conducted extensive simulation studies to assess the impact of the implementation of our tool on the size, power, expected number of interim analyses, and bias in the treatment effect. The numerical study showed the proposed monitoring tool does not affect size or power, but dramatically reduces the expected number of interim analyses when the effect of the treatment difference is small. The tool serves as a useful reference when interpreting the summary of the blinded data throughout the course of the trial, without losing integrity of the study. This tool could potentially save the study resources and budget by avoiding unnecessary interim analyses.

  13. General certification procedure of enterprises and interim job enterprises

    CERN Document Server

    Int. At. Energy Agency, Wien

    2002-01-01

    This procedure defines the certification global process of enterprises employing workers of A or B category for nuclear facilities and interim job enterprises proposing workers of A or B category for nuclear facilities. This certification proves the enterprises ability to satisfy the specification ''E'' of the CEFRI and the interim job enterprises to satisfy the specification ''I'' of the CEFRI. (A.L.B.)

  14. 47 CFR 51.715 - Interim transport and termination pricing.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 3 2010-10-01 2010-10-01 false Interim transport and termination pricing. 51.715 Section 51.715 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) COMMON CARRIER... Telecommunications Traffic § 51.715 Interim transport and termination pricing. (a) Upon request from a...

  15. Do Interim Assessments Reduce the Race and SES Achievement Gaps?

    Science.gov (United States)

    Konstantopoulos, Spyros; Li, Wei; Miller, Shazia R.; van der Ploeg, Arie

    2017-01-01

    The authors examined differential effects of interim assessments on minority and low socioeconomic status students' achievement in Grades K-6. They conducted a large-scale cluster randomized experiment in 2009-2010 to evaluate the impact of Indiana's policy initiative introducing interim assessments statewide. The authors used 2-level models to…

  16. 78 FR 49782 - Interim Staff Guidance on Changes During Construction

    Science.gov (United States)

    2013-08-15

    ... Construction AGENCY: Nuclear Regulatory Commission. ACTION: Draft interim staff guidance; request for comment... During Construction.'' This ISG provides guidance to the NRC staff on the Preliminary Amendment Request...-ISG-025 ``Interim Staff Guidance on Changes during Construction under 10 CFR Part 52'' is available...

  17. Interim supply in the electricity industry

    Energy Technology Data Exchange (ETDEWEB)

    Strassburg, W

    1977-06-01

    The interim supply or the so-called 'condition without a contract' can occur within the framework of energy supply in the relationship between 1) public utility and tariff customer, 2) public utility and the special last-in-line consumer, 3) supplying and distributing public utility, 4) public utility and territorial administrative body. The present contribution deals with the cases named under (3) and (4). Cases (1) and (2) were dealt with in a previous article. Relevant contract clauses and laws and their effects on contracting partners are discussed.

  18. Interim supply in the electricity industry

    Energy Technology Data Exchange (ETDEWEB)

    Strassburg, W

    1977-05-01

    The interim supply or the so-called 'condition without a contract' can occur within the framework of energy supply in the relationship between: 1) public utility and tariff customer, 2) public utility and the special last-in-line consumer, 3) supplying and distributing public utility, 4) public utility and territorial administrative body. The contribution at hand deals with the cases under 1) and 2); cases 3) and 4) are dealt with in an article to be published. Relevant contract clauses and laws and their effects on contracting partners are discussed. 41 references.

  19. Interim Reliability Evaluation Program procedures guide

    International Nuclear Information System (INIS)

    Carlson, D.D.; Gallup, D.R.; Kolaczkowski, A.M.; Kolb, G.J.; Stack, D.W.; Lofgren, E.; Horton, W.H.; Lobner, P.R.

    1983-01-01

    This document presents procedures for conducting analyses of a scope similar to those performed in Phase II of the Interim Reliability Evaluation Program (IREP). It documents the current state of the art in performing the plant systems analysis portion of a probabilistic risk assessment. Insights gained into managing such an analysis are discussed. Step-by-step procedures and methodological guidance constitute the major portion of the document. While not to be viewed as a cookbook, the procedures set forth the principal steps in performing an IREP analysis. Guidance for resolving the problems encountered in previous analyses is offered. Numerous examples and representative products from previous analyses clarify the discussion

  20. National NIF Diagnostic Program Interim Management Plan

    International Nuclear Information System (INIS)

    Warner, B

    2002-01-01

    The National Ignition Facility (NIF) has the mission of supporting Stockpile Stewardship and Basic Science research in high-energy-density plasmas. To execute those missions, the facility must provide diagnostic instrumentation capable of observing and resolving in time events and radiation emissions characteristic of the plasmas of interest. The diagnostic instrumentation must conform to high standards of operability and reliability within the NIF environment. These exacting standards, together with the facility mission of supporting a diverse user base, has led to the need for a central organization charged with delivering diagnostic capability to the NIF. The National NIF Diagnostics Program (NNDP) has been set up under the aegis of the NIF Director to provide that organization authority and accountability to the wide user community for NIF. The funds necessary to perform the work of developing diagnostics for NIF will be allocated from the National NIF Diagnostics Program to the participating laboratories and organizations. The participating laboratories and organizations will design, build, and commission the diagnostics for NIF. Restricted availability of funding has had an adverse impact, unforeseen at the time of the original decision to projectize NIF Core Diagnostics Systems and Cryogenic Target Handing Systems, on the planning and initiation of these efforts. The purpose of this document is to provide an interim project management plan describing the organizational structure and management processes currently in place for NIF Core Diagnostics Systems. Preparation of a Program Execution Plan for NIF Core Diagnostics Systems has been initiated and a current draft is provided as Attachment 1 to this document. The National NIF Diagnostics Program Interim Management Plan provides a summary of primary design criteria and functional requirements, current organizational structure, tracking and reporting procedures, and current planning estimates of project scope

  1. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  2. Introducing Systematic Aging Management for Interim Storage Facilities in Germany

    International Nuclear Information System (INIS)

    Spieth-Achtnich, Angelika; Schmidt, Gerhard

    2014-01-01

    In Germany twelve at-reactor and three central (away from reactor) dry storage facilities are in operation, where the fuel is stored in combined transport-and-storage casks. The safety of the storage casks and facilities has been approved and is licensed for up to 40 years operating time. If the availability of a final disposal facility for the stored wastes (spent fuel and high-level wastes from reprocessing) will be further delayed the renewal of the licenses can become necessary in future. Since 2001 Germany had a regulatory guideline for at-reactor dry interim storage of spent fuel. In this guideline some elements of ageing were implemented, but no systematic approach was made for a state-of-the-art ageing management. Currently the guideline is updated to include all kind of storage facilities (central storages as well) and all kinds of high level waste (also waste from reprocessing). Draft versions of the update are under discussion. In these drafts a systematic ageing management is seen as an instrument to upgrade the available technical knowledge base for possible later regulatory decisions, should it be necessary to prolong storage periods to beyond the currently approved limits. It is further recognized as an instrument to prevent from possible and currently unrecognized ageing mechanisms. The generation of information on ageing can be an important basis for the necessary safety-relevant verifications for long term storage. For the first time, the demands for a systematic monitoring of ageing processes for all safety-related components of the storage system are described. In addition, for inaccessible container components such as the seal system, the neutron shielding, the baskets and the waste inventory, the development of a monitoring program is recommended. The working draft to the revised guideline also contains recommendations on non-technical ageing issues such as the long-term preservation of knowledge, long term personnel planning and long term

  3. Ageing of research reactors

    International Nuclear Information System (INIS)

    Ciocanescu, M.

    2001-01-01

    Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)

  4. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1985-01-01

    During the past two years, scientists from Argonne have developed an advanced breeder reactor with a closed self contained fuel cycle. The Integral Fast Reactor (IFR) is a new reactor concept, adaptable to a variety of designs, that is based on a fuel cycle radically different from the CRBR line of breeder development. The essential features of the IFR are metal fuel, pool layout, and pyro- and electro-reprocessing in a facility integral with the reactor plant. The IFR shows promise to provide an inexhaustible, safe, economic, environmentally acceptable, and diversion resistant source of nuclear power. It shows potential for major improvement in all of the areas that have led to concern about nuclear power

  5. Design, construction and commissioning of an interim spent fuel store for the decommissioning of Ignalina NPP, Lithuania

    International Nuclear Information System (INIS)

    Rainer Goehring; Martin Beverungen; Phil Smith

    2006-01-01

    The contract for the design, construction and commissioning (turn-key) of an interim spent fuel store facility (ISFSF) has been awarded to a Consortium of GNS Gesellschaft fuer Nuklear Service and RWE NUKEM GmbH under the lead of RWE NUKEM. The contract was signed on the 12.01.2005. The Interim Spent Fuel Storage Facility (ISFSF) is financed by the Ignalina Decommissioning Support Fund which is managed by EBRD. All spent fuel assemblies, currently stored in the spent fuel pits at the reactors plus future arising (about 18000 in total) will be loaded in the CONSTOR R RBMK1500/M2 containers, which are stored in the new facility. The initial contract has been awarded for 3500 spent fuel assemblies. (authors)

  6. Decontamination and Decommissioning Project of the TRIGA Mark - 2 and 3 research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jung, K. J.; Baik, S. T.; Chung, U. S.; Park, S. K.; Moon, J. S.; Jung, K. H.; Lee, B. J.; Kim, J. K.; Kim, K. H

    1999-02-01

    Design work for the D and D began in 1998, and expected to be finished at the end of February 1999. Base on the D and D design, the decommissioning plan and the environmental impact assessment report have been completed and submitted to the MOST for licensing at the end of 1998. These documents are being reviewing at the KINS. It is expected that the licensing from authority will be come out at the end of September 1999. Then D and D practical work will be started in the latter half of the year 1999 and completed by 2002. The first practical work of the D and D will be the KRR-2 reactor hall to transform the hall as a temporary storage of radioactive waste produced during the D and D work. Meanwhile, all the spent fuel from KRR-1 and 2 were safely transported to us at the middle of 1998. The D and D project was originally planned, to be finished by the end of the year 1999. This project were lately modified and extended until the end of the year 2002 because of the interim storage of radioactive wastes arising from the D and D work. These radioactive wastes will be stored at the KRR-2 reactor hall until a disposal facility is operational. (author)

  7. German Approach for the Transport of Spent Fuel Packages after Interim Storage

    International Nuclear Information System (INIS)

    Wille, Frank; Wolff, Dietmar; Droste, Bernhard; Voelzke, Holger

    2014-01-01

    In Germany the concept of dry interim storage of spent nuclear fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently (in 2012) revised 'Guidelines for dry cask storage of spent nuclear fuel and heat-generating waste' by the German Waste management Commission (ESK) which are very similar to the former RSK (reactor safety commission) guidelines. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time. In recent years the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependent system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report including evaluation of long term behavior of components and specific operating procedures of the package. Present research and knowledge concerning the long term behavior of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behavior of aged metal and elastomeric seals under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period (aged package). Assessment methods for the material compatibility, the behavior of fuel assemblies and the aging behavior of shielding parts are issues as well. This paper describes the state

  8. Escola segura Safe school

    Directory of Open Access Journals (Sweden)

    Edson Ferreira Liberal

    2005-11-01

    Full Text Available OBJETIVO: Revisão das estratégias para tornar o ambiente escolar seguro. Inicialmente os autores contextualizam a violência e os acidentes no ambiente escolar e fazem recomendações, baseadas em dados da literatura, para a implantação de escolas seguras. FONTE DE DADOS: Artigos publicados entre 1993 e 2005 na base de dados MEDLINE. Dados nacionais epidemiológicos e da literatura também foram pesquisados. SÍNTESE DOS DADOS: Há evidência crescente de que a intervenção tem múltiplos componentes. O foco político é a prática em educação em saúde com o envolvimento de toda a comunidade. O norte dessas intervenções é ajudar estudantes e toda a comunidade a adotar um comportamento seguro e saudável. As escolas estão assumindo um envolvimento crescente na promoção da saúde, prevenção de doenças e prevenção de trauma. Nesse contexto de prevenção de causas externas de morbimortalidade, é importante reconhecer o risco ambiental, locais e comportamentos de risco como favoráveis ao trauma e à violência, além de um novo conceito de acidentes como algo que possa ser evitado. CONCLUSÃO: A implementação da escola segura representa uma nova direção promissora para o trabalho preventivo baseado na escola. É importante notar que uma escola segura deve intervir não meramente na sua estrutura física, mas também torná-la tão segura quanto possível, trabalhando com a comunidade escolar por meio de educação em saúde, discutindo principalmente o comportamento saudável.OBJECTIVE: To review the strategies to make school a safe environment. The paper first addresses the social context of accidents and violence in the school environment, and makes recommendations, based on the literature data, for the implementation of safe schools. SOURCE OF DATA: Articles published between 1993 and 2005 in the MEDLINE database. Brazilian epidemiological and literature data have also been searched. SUMMARY OF THE FINDINGS: There is

  9. Safe motherhood at risk?

    Science.gov (United States)

    Thompson, A

    1996-12-01

    Health professionals' negative attitudes toward clients often exacerbate the problems women face in terms of health status and access to health care. Thus, the health professionals can themselves be obstacles to women seeking the health care they need. A key challenge to midwives, in addition to providing technically competent services, is gaining insight into the people for whom they are responsible so that childbirth traditions are treated with respect and women are offered dignity. Safe motherhood requires intersectoral collaboration. Many innovative approaches to safe motherhood are based on the community's participation in planning services that meet the needs of women. Other approaches are based on decentralization of services. For example, a large university teaching hospital in Lusaka, Zambia, set up birthing centers around the city to take the pressure off the hospital. Midwives head up these centers, which are close to the women's homes. Decentralization of delivery services has improved the physical and emotional outcomes for mothers and newborns. Midwives must be prepared to articulate concerns about inequalities and deficiencies in the health care system in order to persuade the government to change. Women, including midwives, need to form multidisciplinary alliances to work together to effect change. The front-line workers in maternity care are midwives. They should adopt the following strategies to become even more effective in their efforts to make motherhood safer. They should listen to what women say about their needs. They should scale services to a manageable, human scale. They should learn the skills to become politically active advocates. They should work with other midwives, women, leaders, and other professional groups. Motherhood can be safe when women have more control over their own decision making, the education to liberate themselves to make their own decisions, and access to skilled care.

  10. Management of research reactor ageing

    International Nuclear Information System (INIS)

    1995-03-01

    As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs

  11. Management of research reactor ageing

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs.

  12. H Reactor

    Data.gov (United States)

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

  13. Reactor feedwater control device

    International Nuclear Information System (INIS)

    Koshi, Yuji.

    1993-01-01

    In the device of the present invention, an excess response is not caused in a reactor feed water system even when voids are fluctuated by using an actual water level signal as a reactor water level signal. That is, a standard water level signal and a reactor water level signal are inputted to a comparator. An adder adds water level difference signal outputted from the comparator and mismatch flow rate signal prepared by multiplying the difference between a main steam flow rate signal and a feed water flow rate signal by a mismatch gain. A feed water controller integrates the added signal and outputs flow rate demand signal. A feed water system receives the flow rate demand signal as input. A water level calculation means is disposed to such a device for calculating an actual water level based on the change of coolant possessing amount of the reactor, and the output thereof is defined as a reactor water level signal. With such procedures, excessive elevation of water level of the reactor can be prevented even upon occurrence of void fluctuation phenomenon or the like in the reactor such as upon sole scram operation. Accordingly, plant shut down caused thereby can be avoided safely. (I.S.)

  14. Vitamins, Are They Safe?

    Directory of Open Access Journals (Sweden)

    Hadi Hamishehkar

    2016-12-01

    Full Text Available The consumption of a daily multivitamin among people all over the world is dramatically increasing in recent years. Most of the people believe that if vitamins are not effective, at least they are safe. However, the long term health consequences of vitamins consumption are unknown. This study aimed to assess the side effects and possible harmful and detrimental properties of vitamins and to discuss whether vitamins can be used as safe health products or dietary supplements. We performed a MEDLINE/PubMed, EMBASE, Scopus and Google Scholar search and assessed reference lists of the included studies which were published from 1993 through 2015. The studies, with an emphasis on RCTs (randomized controlled clinical trials, were reviewed. As some vitamins such as fat-soluble vitamins (vitamin A, vitamin D, vitamin E, and also some of the water-soluble vitamins like folic acid may cause adverse events and some like vitamin C is widely taken assuming that it has so many benefits and no harm, we included relevant studies with negative or undesired results regarding the effect of these vitamins on health. Our recommendation is that taking high-dose supplements of vitamins A, E, D, C, and folic acid is not always effective for prevention of disease, and it can even be harmful to the health.

  15. Spent fuel management options for research reactors in Latin America

    International Nuclear Information System (INIS)

    2006-06-01

    Research reactors (RRs) have been operated in Latin America since the late 1950s, and a total of 23 RRs have been built in the region. At the time of writing (November 2005), 18 RRs are in operation, 4 have been shut down and 1 has been decommissioned. The number of operating RRs in Latin America represents around 6% of the existing operational RRs worldwide and around 21% of the RRs operating in developing countries. Common to all RRs in the region is a consistent record of safe and successful operation. With the purpose of carrying out a collaborative study of different aspects of the management of spent fuel from RRs, some countries from the region proposed to the IAEA in 2000 the organization of a Regional Project. The project (IAEA TC Regional Project RLA/4/018) that was approved for the biennium 2001-2002 and extended for 2003-2004 included the participation of Argentina, Brazil, Chile, Mexico and Peru. The main objectives of this project were: (a) to define the basic conditions for a regional strategy for managing spent fuel that will provide solutions compatible with the economic and technological realities of the countries involved; and (b) to determine what is needed for the temporary wet and dry storage of spent fuel from the research reactors in the countries of the Latin American region that participated in the project. This TECDOC is based on the results of TC Regional Project RLA/4/018. This project was successful in identifying and assessing a number of viable alternatives for RRSF management in the Latin American region. Options for operational and interim storage, spent fuel conditioning and final disposal have been carefully considered. This report presents the views of Latin American experts on RR spent fuel management and will be useful as reference material for the Latin American RR community, decision making authorities in the region and the public in general

  16. Cool and Safe: Multiplicity in Safe Innovation at Unilever

    Science.gov (United States)

    Penders, Bart

    2011-01-01

    This article presents the making of a safe innovation: the application of ice structuring protein (ISP) in edible ices. It argues that safety is not the absence of risk but is an active accomplishment; innovations are not "made safe afterward" but "safe innovations are made". Furthermore, there are multiple safeties to be accomplished in the…

  17. Modular Stellarator Fusion Reactor concept

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1981-08-01

    A preliminary conceptual study is made of the Modular Stellarator Reactor (MSR). A steady-state ignited, DT-fueled, magnetic fusion reactor is proposed for use as a central electric-power station. The MSR concept combines the physics of the classic stellarator confinement topology with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. The physics basis of the design point is described together with supporting magnetics, coil-force, and stress computations. The approach and results presented herein will be modified in the course of ongoing work to form a firmer basis for a detailed conceptual design of the MSR

  18. The United States Naval Nuclear Propulsion Program - Over 151 Million Miles Safely Steamed on Nuclear Power

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2015-03-01

    NNSA’s third mission pillar is supporting the U.S. Navy’s ability to protect and defend American interests across the globe. The Naval Reactors Program remains at the forefront of technological developments in naval nuclear propulsion and ensures a commanding edge in warfighting capabilities by advancing new technologies and improvements in naval reactor performance and reliability. In 2015, the Naval Nuclear Propulsion Program pioneered advances in nuclear reactor and warship design – such as increasing reactor lifetimes, improving submarine operational effectiveness, and reducing propulsion plant crewing. The Naval Reactors Program continued its record of operational excellence by providing the technical expertise required to resolve emergent issues in the Nation’s nuclear-powered fleet, enabling the Fleet to safely steam more than two million miles. Naval Reactors safely maintains, operates, and oversees the reactors on the Navy’s 82 nuclear-powered warships, constituting more than 45 percent of the Navy’s major combatants.

  19. Retention of long-term interim restorations with sodium fluoride enriched interim cement

    Science.gov (United States)

    Strash, Carolyn

    Purpose: Interim fixed dental prostheses, or "provisional restorations", are fabricated to restore teeth when definitive prostheses are made indirectly. Patients undergoing extensive prosthodontic treatment frequently require provisionalization for several months or years. The ideal interim cement would retain the restoration for as long as needed and still allow for ease of removal. It would also avoid recurrent caries by preventing demineralization of tooth structure. This study aims to determine if adding sodium fluoride varnish to interim cement may assist in the retention of interim restorations. Materials and methods: stainless steel dies representing a crown preparation were fabricated. Provisional crowns were milled for the dies using CAD/CAM technology. Crowns were provisionally cemented onto the dies using TempBond NE and NexTemp provisional cements as well as a mixture of TempBond NE and Duraphat fluoride varnish. Samples were stored for 24h then tested or thermocycled for 2500 or 5000 cycles before being tested. Retentive strength of each cement was recorded using a universal testing machine. Results: TempBond NE and NexTemp cements performed similarly when tested after 24h. The addition of Duraphat significantly decreased the retention when added to TempBond NE. NexTemp cement had high variability in retention over all tested time periods. Thermocycling for 2500 and 5000 cycles significantly decreased the retention of all cements. Conclusions: The addition of Duraphat fluoride varnish significantly decreased the retention of TempBond NE and is therefore not recommended for clinical use. Thermocycling significantly reduced the retention of TempBond NE and NexTemp. This may suggest that use of these cements for three months, as simulated in this study, is not recommended.

  20. Basis for Interim Operation for Fuel Supply Shutdown Facility

    International Nuclear Information System (INIS)

    BENECKE, M.W.

    2003-01-01

    This document establishes the Basis for Interim Operation (BIO) for the Fuel Supply Shutdown Facility (FSS) as managed by the 300 Area Deactivation Project (300 ADP) organization in accordance with the requirements of the Project Hanford Management Contract procedure (PHMC) HNF-PRO-700, ''Safety Analysis and Technical Safety Requirements''. A hazard classification (Benecke 2003a) has been prepared for the facility in accordance with DOE-STD-1027-92 resulting in the assignment of Hazard Category 3 for FSS Facility buildings that store N Reactor fuel materials (303-B, 3712, and 3716). All others are designated Industrial buildings. It is concluded that the risks associated with the current and planned operational mode of the FSS Facility (uranium storage, uranium repackaging and shipment, cleanup, and transition activities, etc.) are acceptable. The potential radiological dose and toxicological consequences for a range of credible uranium storage building have been analyzed using Hanford accepted methods. Risk Class designations are summarized for representative events in Table 1.6-1. Mitigation was not considered for any event except the random fire event that exceeds predicted consequences based on existing source and combustible loading because of an inadvertent increase in combustible loading. For that event, a housekeeping program to manage transient combustibles is credited to reduce the probability. An additional administrative control is established to protect assumptions regarding source term by limiting inventories of fuel and combustible materials. Another is established to maintain the criticality safety program. Additional defense-in-depth controls are established to perform fire protection system testing, inspection, and maintenance to ensure predicted availability of those systems, and to maintain the radiological control program. It is also concluded that because an accidental nuclear criticality is not credible based on the low uranium enrichment

  1. E-Area Performance Assessment Interim Measures Assessment FY2005

    Energy Technology Data Exchange (ETDEWEB)

    Stallings, M

    2006-01-31

    After major changes to the limits for various disposal units of the E-Area Low Level Waste Facility (ELLWF) last year, no major changes have been made during FY2005. A Special Analysis was completed which removes the air pathway {sup 14}C limit from the Intermediate Level Vault (ILV). This analysis will allow the disposal of reactor moderator deionizers which previously had no pathway to disposal. Several studies have also been completed providing groundwater transport input for future special analyses. During the past year, since Slit Trenches No.1 and No.2 were nearing volumetric capacity, they were operationally closed under a preliminary closure analysis. This analysis was performed using as-disposed conditions and data and showed that concrete rubble from the demolition of 232-F was acceptable for disposal in the STs even though the latest special analysis for the STs had reduced the tritium limits so that the inventory in the rubble exceeded limits. A number of special studies are planned during the next years; perhaps the largest of these will be revision of the Performance Assessment (PA) for the ELLWF. The revision will be accomplished by incorporating special analyses performed since the last PA revision as well as revising analyses to include new data. Projected impacts on disposal limits of more recent studies have been estimated. No interim measures will be applied during this year. However, it is being recommended that tritium disposals to the Components-in-Grout (CIG) Trenches be suspended until a limited Special Analysis (SA) currently in progress is completed. This SA will give recommendations for optimum placement of tritiated D-Area tower waste. Further recommendations for tritiated waste placement in the CIG Trenches will be given in the upcoming PA revision.

  2. Nuclear reactor kinetics and control

    International Nuclear Information System (INIS)

    Lewins, J.

    1978-01-01

    A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

  3. Safe nuclear power

    International Nuclear Information System (INIS)

    Cady, K.B.

    1992-01-01

    Nearly 22 percent of the electricity generated in the United States already comes from nuclear power plants, but no new plants have been ordered since 1978. This paper reports that the problems that stand in the way of further development have to do with complexity and perceived risk. Licensing, construction management, and waste disposal are complex matters, and the possibility of accident has alienated a significant portion of the public. But a national poll conducted by Bruskin/Goldring at the beginning of February shows that opposition to nuclear energy is softening. Sixty percent of the American people support (strongly or moderately) the use of nuclear power, and 18 percent moderately oppose it. Only 15 percent remain obstinately opposed. Perhaps they are not aware of recent advances in reactor technology

  4. PROJECT W-551 INTERIM PRETREATMENT SYSTEM PRECONCEPTUAL CANDIDATE TECHNOLOGY DESCRIPTIONS

    Energy Technology Data Exchange (ETDEWEB)

    MAY TH

    2008-08-12

    The Office of River Protection (ORP) has authorized a study to recommend and select options for interim pretreatment of tank waste and support Waste Treatment Plant (WTP) low activity waste (LAW) operations prior to startup of all the WTP facilities. The Interim Pretreatment System (IPS) is to be a moderately sized system which separates entrained solids and 137Cs from tank waste for an interim time period while WTP high level waste vitrification and pretreatment facilities are completed. This study's objective is to prepare pre-conceptual technology descriptions that expand the technical detail for selected solid and cesium separation technologies. This revision includes information on additional feed tanks.

  5. Interim report of the task force on energy policy

    International Nuclear Information System (INIS)

    2001-01-01

    A Task Force was established by the Premier of British Columbia in August 2001 to draft an energy policy framework for the province. Based on best practices worldwide, and keeping in mind the specific energy needs of British Columbia, this framework aims at fostering energy development in British Columbia, in accordance with exemplary environmental practices. This interim report comprises the preliminary findings of the Task Force, and public input is sought before the final report is finalized and presented to government. The energy sector of British Columbia comprises hydroelectric power, oil, gas and coal resources. In addition, green energy and alternative energy technologies are being developed, such as wind, solar, and wave power, and hydrogen fuel cells. Industry and individual consumers are well served by the highly developed transmission and distribution systems for energy. Several strategic directions were identified by the Task Force for inclusion in the energy policy of British Columbia, to meet its full potential. They are: growth to ensure safe, reliable energy and take advantage of economic opportunities; diversification; competitiveness; industry restructuring and expansion; environmental imperative; government leadership; and community and First Nations' involvement. Some changes are also required for the continuing success of the energy sector in British Columbia: a move to fully competitive markets in the electricity system, the development of natural gas storage capacity in the Lower Mainland, additional considerations for coal use, and the development of alternative energy sources. It is expected that private capital and more energy supply will result from a fully competitive energy market, which in turn would lower energy costs. Jobs and income would increase as a result of the growth in the sector. Diversification makes good economic and environmental sense. tabs., figs

  6. Safe pill-dispensing.

    Science.gov (United States)

    Testa, Massimiliano; Pollard, John

    2007-01-01

    Each patient is supplied with a smart-card containing a Radio Frequency IDentification (RFID) chip storing a unique identification code. The patient places the Smart-card on a pill-dispenser unit containing an RFID reader. The RFID chip is read and the code sent to a Base-station via a wireless Bluetooth link. A database containing both patient details and treatment information is queried at the Base-station using the RFID as the search key. The patient's treatment data (i.e., drug names, quantities, time, etc.) are retrieved and sent back to the pill-dispenser unit via Bluetooth. Appropriate quantities of the required medications are automatically dispensed, unless the patient has already taken his/her daily dose. Safe, confidential communication and operation is ensured.

  7. Activity of the RA Reactor Physics group in 1980 - Definition of the Operation conditions for future safe and economical RA reactor operation with 80% enriched fuel; Prilog Ia - Rad sluzbe za fiziku reaktora RA u 1980. godini - Definisanje pogonskih uslova za dalji siguran i ekonomican rad reaktora RA sa 80% obogacenim gorivom

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1980-12-15

    During 1980. the RA reactor was not in operation. That is why this period was devoted to definition of operating conditions for further reactor operation with 80% enriched fuel. The fuel elements which were in the core at the moment of shutdown in March 1979will not be used again (388 80% enriched fuel elements, and 511 2% enriched fuel elements). The reactor will be operated only with 80% enriched fuel, staring with initiat core configuration with 440 elements on the borders gradually changing to equi;librium core with 720 fuel elements. The analyses were concerned with safety issues of future operation. [Serbo-Croat] Reaktor RA tokom 1980 godine nije radio. Zbog toga je ovaj period iskoriscen za intenzivan rad na definisanju pogonskih uslova za dalji rad reaktora sa 80% obogacenim gorivom. Gorivo sa kojim je reaktor prestao da radi marta 1979. godine (388 gorivnih elemenata 80% obogacenoh i 511 elementa 2% obogaceno) nece biti vraceno u jezgro, vec ce reaktor raditi samo sa 80% obogacenim gorivom, pocev od konfiguracije sa 440 elemenata na periferiji do ravnotezne konfiguracije sa 720 elemenata. Sve nalize su se bavile sigurnosnim aspektima rada reaktora sa visokoobogacenim gorivom u jezgru.

  8. Strategies for safe motherhood.

    Science.gov (United States)

    Chatterjee, A

    1995-02-01

    The Safe Motherhood Initiative was launched in 1988 as a global effort to halve maternal mortality and morbidity by the year 2000. The program uses a combination of health and nonhealth strategies to emphasize the need for maternal health services, extend family planning services, and improve the status of women. The maternal mortality rate (per 100,000 live births) is 390 for the world, 20-30 for developed countries, 450 for developing countries, and 420 for Asia. This translates into 308,000 maternal deaths in Asia, of which 100,000 occur in India. The direct causes of maternal mortality include sepsis, hemorrhage, eclampsia, and ruptured uterus. Indirect causes occur when associated medical conditions, such as anemia and jaundice, are exacerbated by pregnancy. Underlying causes are ineffective health services, inadequate obstetric care, unregulated fertility, infections, illiteracy, early marriage, poverty, malnutrition, and ignorance. India's Child Survival and Safe Motherhood Program seeks to achieve immediate improvements by improving health care. Longterm improvements will occur as nutrition, income, education, and the status of women improve. Improvements in health care will occur in through the provision of 1) essential obstetric care for all women (which will be essentially designed for low-risk women), 2) early detection of complications during pregnancy and labor, and 3) emergency services. Services will be provided to pregnant women at their door by field staff, at a first referral hospital, perhaps at maternity villages where high risk cases can be housed in the latter part of their pregnancies, and through the continual accessibility of government vehicles. In addition, family planning services will be improved so that fertility regulation can have its expected beneficial effect on the maternal mortality rate. The professional health organizations in India will also play a vital role in the success of this effort to reduce maternal mortality.

  9. Light water ultra-safe plant concept

    International Nuclear Information System (INIS)

    Klevans, E.

    1989-01-01

    Since the accident at Three Mile Island (TMI), Penn State Nuclear Engineering Department Faculty and Staff have considered various methods to improve already safe reactor designs and public perception of the safety of Nuclear Power. During 1987 and 1988, the Department of Energy provided funds to the Nuclear Engineering Department at Penn State to investigate a plant reconfiguration originated by M.A. Schultz called ''The Light Water Ultra-Safe Plant Concept''. This report presents a final summary of the project with references to several masters' theses and addendum reports for further detail. The two year research effort included design verification with detailed computer simulation of: (a) normal operation characteristics of the unique pressurizing concept, (b) severe transients without loss of coolant, (c) combined primary and secondary system modeling, and (d) small break and large break loss of coolant accidents. Other studies included safety analysis, low power density core design, and control system design to greatly simplify the control room and required operator responses to plant upset conditions. The overall conclusion is that a reconfigured pressurized water reactor can achieve real and perceived safety improvements. Additionally, control system research to produce greatly simplified control rooms and operator requirements should be continued in future projects

  10. Nuclear hydrogen production and its safe handling

    International Nuclear Information System (INIS)

    Chung, Hongsuk; Paek, Seungwoo; Kim, Kwang-Rag; Ahn, Do-Hee; Lee, Minsoo; Chang, Jong Hwa

    2003-01-01

    An overview of the hydrogen related research presently undertaken at the Korea Atomic Energy Research Institute are presented. These encompass nuclear hydrogen production, hydrogen storage, and the safe handling of hydrogen, High temperature gas-cooled reactors can play a significant role, with respect to large-scale hydrogen production, if used as the provider of high temperature heat in fossil fuel conversion or thermochemical cycles. A variety of potential hydrogen production methods for high temperature gas-cooled reactors were analyzed. They are steam reforming of natural gas, thermochemical cycles, etc. The produced hydrogen should be stored safely. Titanium metal was tested primarily because its hydride has very low dissociation pressures at normal storage temperatures and a high capacity for hydrogen, it is easy to prepare and is non-reactive with air in the expected storage conditions. There could be a number of potential sources of hydrogen evolution risk in a nuclear hydrogen production facility. In order to reduce the deflagration detonation it is necessary to develop hydrogen control methods that are capable of dealing with the hydrogen release rate. A series of experiments were conducted to assess the catalytic recombination characteristics of hydrogen in an air stream using palladium catalysts. (author)

  11. Commercial experience with facility deactivation to safe storage

    International Nuclear Information System (INIS)

    Sype, T.T.; Fischer, S.R.; Lee, J.H. Jr.; Sanchez, L.C.; Ottinger, C.A.; Pirtle, G.J.

    1995-09-01

    The Department of Energy (DOE) has shutdown many production reactors; the Department has begun a major effort to also shutdown a wide variety of other nuclear facilities. Because so many facilities are being closed, it is necessary to place many of them into a safe- storage status, i.e., deactivation, before conducting decommissioning- for perhaps as long as 20 years. The challenge is to achieve this safe-storage condition in a cost-effective manner while remaining in compliance with applicable regulations. The DOE Office of Environmental Management, Office of Transition and Management, commissioned a lessons-learned study of commercial experience with safe storage and decommissioning. Although the majority of the commercial experience has been with reactors, many of the lessons learned presented in this document can provide insight into transitioning challenges that Will be faced by the DOE weapons complex

  12. Commercial experience with facility deactivation to safe storage

    Energy Technology Data Exchange (ETDEWEB)

    Sype, T.T. [Sandia National Labs., Albuquerque, NM (United States); Fischer, S.R. [Los Alamos National Lab., NM (United States); Lee, J.H. Jr.; Sanchez, L.C.; Ottinger, C.A.; Pirtle, G.J. [Sandia National Labs., Albuquerque, NM (United States)

    1995-09-01

    The Department of Energy (DOE) has shutdown many production reactors; the Department has begun a major effort to also shutdown a wide variety of other nuclear facilities. Because so many facilities are being closed, it is necessary to place many of them into a safe- storage status, i.e., deactivation, before conducting decommissioning- for perhaps as long as 20 years. The challenge is to achieve this safe-storage condition in a cost-effective manner while remaining in compliance with applicable regulations. The DOE Office of Environmental Management, Office of Transition and Management, commissioned a lessons-learned study of commercial experience with safe storage and decommissioning. Although the majority of the commercial experience has been with reactors, many of the lessons learned presented in this document can provide insight into transitioning challenges that Will be faced by the DOE weapons complex.

  13. Hierarchies of belief and interim rationalizability

    Directory of Open Access Journals (Sweden)

    Jeffrey C. Ely

    2006-03-01

    Full Text Available In games with incomplete information, conventional hierarchies of belief are incomplete as descriptions of the players' information for the purposes of determining a player's behavior. We show by example that this is true for a variety of solution concepts. We then investigate what is essential about a player's information to identify behavior. We specialize to two player games and the solution concept of interim rationalizability. We construct the universal type space for rationalizability and characterize the types in terms of their beliefs. Infinite hierarchies of beliefs over conditional beliefs, which we call Delta-hierarchies, are what turn out to matter. We show that any two types in any two type spaces have the same rationalizable sets in all games if and only if they have the same Delta-hierarchies.

  14. Interim report and accounts 1993/94

    International Nuclear Information System (INIS)

    1993-01-01

    An interim set of accounts and reports is presented here for 1993/1994 for the health science company Amersham International. The company's research programs focus on developments in life science research, nuclear medicine and industrial quality and safety assurance, with particular expertise in the application of radioactivity to labelling and detection at the molecular level. This report which covers the half-year to 30 September 1993 shows promising financial results, with turnover, operating profits and earnings per share all having risen. All life science markets report growth although difficult trading conditions are being reported in Europe. Two products in the Healthcare business have achieved progress, a pain palliation agent for bone metastases has been launched in the United States, and European approval has been gained for a new technetium based heart imaging agent. Further growth is expected for the company. (UK)

  15. Safety regulations for radioisotopes, etc. (interim report)

    International Nuclear Information System (INIS)

    1980-01-01

    An (interim) report by an ad hoc expert committee to the Nuclear Safety Commission, on the safety regulations for radioisotopes, etc., was presented. For the utilization of radioisotopes, etc., there is the Law Concerning Prevention of Radiation Injury Due to Radioisotopes, etc. with the advances in this field and the improvement in international standards, the regulations by the law have been examined. After explaining the basic ideas of the regulations, the problems and countermeasures in the current regulations are described: legal system, rationalization in permission procedures and others, inspection on RI management, the system of the persons in charge of radiation handling, RI transport, low-level radioactive wastes, consumer goods, definitions of RIs, radiation and sealed sources, regulations by group partitioning, RI facilities, system of personnel exposure registration, entrusting of inspection, etc. to private firms, and reduction in the works for permission among governmental offices. (author)

  16. Study of radiation exposure rate on the measurement points in Kartini reactor hall as based to determine operation safety parameters (KBO)

    International Nuclear Information System (INIS)

    Mahrus Salam; Elisabeth Supriyatni; Fajar Panuntun

    2016-01-01

    In the operation of nuclear facility there are safety parameters, which is the value of the conservatively maximum limit to ensure that all of the uncertainty in the analysis of facility operations safety have been considered, such as uncertainty of measurement, response time and uncertainty calculation tool, and is get a long to others value of normal operating condition limits, in other words, there are still allowed or permitted. Calculation of the radiation exposure rate on five measurement points (50 cm above the water surface of reactor pool, above interim storage (bulk shielding), reactor deck, thermal column and sub critical facility) and to be compared to the operation safety parameters (KBO) of Kartini reactor. The exposure rate value is obtained by calculating the source term of radioactivity on the core, attenuation resulting from the radiation shielding and measurement distance. From the calculation obtained that the value of gamma exposure rate of 50 cm above the water surface of reactor pool is 96.91 mR/hr (KBO<100 mR/hr), on the deck of Bulk Shielding amounted to 1.70 mR/h (KBO<2.5 mR/hr), on the reactor deck amounted to 5.73 mR/hr (KBO<10 mR/hr), on the Thermal Column amounted to 2.73 mR/hr (KBO<10 mR/hr) and on the sub critical facility amounted to 1.148 mR/hr (KBO<2.5 mR/hr). The value of gamma exposure rate at 5 locations measurements are still less than the operation safety parameters (KBO), it means that the reactor is safe to be operated. (author)

  17. Interim Hanford Waste Management Technology Plan

    International Nuclear Information System (INIS)

    1985-09-01

    The Interim Hanford Waste Management Technology Plan (HWMTP) is a companion document to the Interim Hanford Waste Management Plan (HWMP). A reference plan for management and disposal of all existing and certain projected future radioactive Hanford Site Defense Wastes (HSDW) is described and discussed in the HWMP. Implementation of the reference plan requires that various open technical issues be satisfactorily resolved. The principal purpose of the HWMTP is to present detailed descriptions of the technology which must be developed to close each of the technical issues associated with the reference plan identified in the HWMP. If alternative plans are followed, however, technology development efforts including costs and schedules must be changed accordingly. Technical issues addressed in the HWMTP and HWMP are those which relate to disposal of single-shell tank wastes, contaminated soil sites, solid waste burial sites, double-shell tank wastes, encapsulated 137 CsCl and 90 SrF 2 , stored and new solid transuranic (TRU) wastes, and miscellaneous wastes such as contaminated sodium metal. Among the high priority issues to be resolved are characterization of various wastes including early determination of the TRU content of future cladding removal wastes; completion of development of vitrification (Hanford Waste Vitrification Plant) and grout technology; control of subsidence in buried waste sites; and development of criteria and standards including performance assessments of systems proposed for disposal of HSDW. Estimates of the technology costs shown in this report are made on the basis that all identified tasks for all issues associated with the reference disposal plan must be performed. Elimination of, consolidation of, or reduction in the scope of individual tasks will, of course, be reflected in corresponding reduction of overall technology costs

  18. Management and storage of spent research reactor fuel within the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Krull, W.

    1996-01-01

    Research reactors in the Federal Republic of Germany and their needs for the interim storage of spent nuclear fuel are described. Existing long-term interim storage facilities are described. Special licensing and legal restrictions imposed by the German Atomic Energy Act are outlined. Possible final solutions for the back end of the nuclear fuel cycle for research reactors, including reprocessing in the United Kingdom or France, return of US-origin fuel and a home-grown German solution are discussed. (author). 2 refs, 5 figs, 4 tabs

  19. Safe transport of spent fuels after long-term storage

    International Nuclear Information System (INIS)

    Aritomi, M.; Takeda, T.; Ozaki, S.

    2004-01-01

    Considering the scarcity of energy resources in Japan, a nuclear energy policy pertaining to the spent fuel storage has been adopted. The nuclear energy policy sets the rules that spent fuels generated from LWRs shall be reprocessed and that plutonium and unburnt uranium shall be recovered and reused. For this purpose, a reprocessing plant, which has a reprocessing capability of 800 ton/yr, is under construction at Rokkasho Village. However, it is anticipated that the start of its operation will be delayed. In addition, the amount of spent fuels generated from nuclear power plants exceeds its reprocessing capability. Therefore, the establishment of storage technology for spent fuels becomes an urgent problem in Japan in order to continue smoothly the LWR operations. In this paper, the background of nuclear power generation in Japan is introduced at first. Next, the policy of spent fuel storage in Japan and circumstances surrounding the spent fuels in Japan are mentioned. Furthermore, the major subjects for discussions to settle and improve 'Standard for Safety Design and Inspection of Metal Casks for Spent Fuel Interim Storage Facility' in Atomic Energy Society of Japan are discussed, such as the integrity of fuel cladding, basket, shielding material and metal gasket for the long term storage for achieving safe transport of spent fuels after the storage. Finally, solutions to the unsolved subject in establishing the spent fuel interim storage technologies ase introduced accordingly

  20. Technical specifications: Health Physics Research Reactor

    International Nuclear Information System (INIS)

    1986-03-01

    These technical specifications define the key limitations that must be observed for safe operation of the Health Physics Research Reactor (HPRR) and an envelope of operation within which there is assurance that these limits will not be exceeded