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Sample records for atw schnellstatistik kernkraftwerke

  1. Nuclear power plants: 2006 atw compact statistics; atw Schnellstatistik Kernkraftwerke 2006

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2007-01-15

    At the turn of 2006/2007, nuclear power plants were available for energy supply, or under construction, in 32 countries of the world. A total of 437 nuclear power plants, which is 7 plants less than at the 2005/2006 turn, were in operation in 31 countries with an aggregate gross power of approx. 388 GWe and an aggregate net power, respectively, of 369 GWe. The available gross power of nuclear power plants dropped by approx. 1.6 GWe, the available net power, by approx. 1.2 GWe. The Tarapur 3 nuclear generating unit was commissioned in India, a D{sub 2}O PWR of 540 MWe gross power. Power operation was discontinued for good in 2006 only in nuclear power plants in Europe: Bohunice 1 (Slovak Republic, 440/408 MWe, VVER PWR); Kozloduy 3 and Kozloduy 4 (Bulgaria, 440/408 MWe each, VVER PWR); Dungeness A1 and Dungeness A2 (United Kingdom, 245/219 MWe each, Magnox GGR); Sizewell A1 and Sizewell A2 (United Kingdom, 236/210 MWe each, Magnox GGR), and Jose Cabrera 1 (Zorita) (Spain, 160/153 MWe, PWR). 29 nuclear generating units, i.e. 8 plants more than at the end of 2005, with an aggregate gross power of approx. 28 GWe, were under construction in 10 countries end of 2006. In China, construction of the Qinshan II-3, Qinshan II-4 nuclear generating units was started. In the Republic of Korea, construction work began on 4 new projects: Shin Kori 1, Shin Kori 2, and Shin Wolsong 1, Shin Wolsong 2. In Russia, work was resumed on the BN-800 sodium-cooled fast breeder reactor project at Beloyarsk and the RBMK Kursk 5. Some 40 new nuclear power plants are in the concrete project design, planning and licensing phases worldwide; on some of them, contracts have already been awarded. Another approximately seventy units are in their preliminary project phases. (orig.)

  2. Accelerator transmutation of wastes (ATW) - Prospects and safety

    International Nuclear Information System (INIS)

    Accelerator transmutation of nuclear waste (ATW) has during last years gained interest as a technologically possible method to transform radioactive wastes into short-lived or stable isotopes. Different ATW-projects are described from the physical and technical point of view. The principal sketch of the safety analysis of the ATW-idea is given. Due to the very limited technical data for existing ATW-projects the safety analysis can cause some risks for the health and environmental safety for the closest environment. General public should not be affected. 35 refs, 22 figs, 4 tabs

  3. Analysis of SBO ATWS for Maanshan PWR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Che-Hao; Chen, Shao-Wen [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Shih, Chunkuan [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Nuclear and New Energy Education and Research Foundation, Hsinchu, Taiwan (China); Lin, Hao-Tzu [Atomic Energy Council, Taoyuan, Taiwan (China). Inst. of Nuclear Energy Research

    2015-11-15

    Station blackout anticipated transient without scram (SBO ATWS) is considered as loss of off-site and on-site power but no credit for automatic reactor trip. SBO ATWS causes reactor coolant pump (RCP) trip, loss of all main feedwater pumps and turbine trip, then the reactor coolant system (RCS) pressure rises rapidly due to loss of heat removal paths. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) is assumed to be an unacceptable plant condition in SECY-83-293. The simulation is performed by TRACE which is a thermal-hydraulic code developed by U.S. NRC. Three different AFW flows are modeled to ensure the pressures will not be beyond the criteria. RCP seal-leakage is concerned as a SBLOCA due to loss of RCP seal-cooling. Four possible leakage flows are modeled to examine the reactor core water level and temperature variation.

  4. A definition of sensor output for the ATWS rule

    International Nuclear Information System (INIS)

    The use and acceptance of probabilistic risk analysis (PRA) to measure the safety of nuclear power plants has grown rapidly over the last decade. PRA has been used to address specific issues in the nuclear industry such as system interactions, technical specification requirements, optimization of limiting conditions for operation, financial risk, and the preparation of emergency response guidelines. This paper presents the discussion of how PRA was used to resolve an anticipated transient without scram (ATWS) issue on component diversity. An introduction to the ATWS issue is presented along with the PRA methods used to define the sensor output for the ATWS rule

  5. Study of safety relief valve operation under ATWS conditions. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.

    1979-06-25

    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  6. An ATWS Analysis with a Realistic Evaluation Methodology

    International Nuclear Information System (INIS)

    Anticipated Transients Without Scram (ATWS) would occur on failure of all the control and shutdown assemblies to insert into the core following an automatic reactor trip. The major concern of the ATWS derives from consequences of the high primary system pressure which is the characteristic of the transients. According to section 2.4 of YVL guides which are Finnish regulations for safety of nuclear power plants (NPP), the acceptance criterion for the ATWS analysis is that the pressure of the protected item does not exceed a pressure limit that is 1.3 times the design pressure. The main purpose of this paper is to assess its impact on the APR1400 preliminarily, for Europe regulatory environments by applying European Utility Requirements (EUR) for Light Water Reactor Nuclear Power Plants

  7. Study of safety relief valve operation under ATWS conditions

    International Nuclear Information System (INIS)

    In March 1979, ETEC published as ETEC-TDR-78-19 a search which updated earlier NRC studies of saturated or subcooled water flow through relief valves, under ATWS conditions. This Supplement expands upon that search to include supercritical steam-water flow. No applicable data for the supercritical conditions was found, nor were any newer data on saturated or subcooled conditions uncovered. The Supplement also updated a look for facilities currently capable of simultaneously imposing all ATWS conditions upon test relief valves. Results confirmed the negative findings of ETEC-TDR-78.19

  8. ATW system impact on high-level waste

    International Nuclear Information System (INIS)

    This report discusses the Accelerator Transmutation of Waste (ATW) concept which aims at destruction of key long-lived radionuclides in high-level nuclear waste (HLW), both fission products and actinides. This focus makes it different from most other transmutation concepts which concentrate primarily on actinide burning. The ATW system uses an accelerator-driven, sub-critical assembly to create an intense thermal neutron environment for radionuclide transmutation. This feature allows rapid transmutation under low-inventory system conditions, which in turn, has a direct impact on the size of chemical separations and materials handling components of the system. Inventories in ATW are factors of eight to thirty times smaller than reactor systems of equivalent thermal power. Chemical separations systems are relatively small in scale and can be optimized to achieve high decontamination factors and minimized waste streams. The low-inventory feature also directly impacts material amounts remaining in the system at its end of life. In addition to its low-inventory operation, the accelerator-driven neutron source features of ATW are key to providing a sufficient level of neutrons to allow transmutation of long-lived fission products

  9. ATW system impact on high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Arthur, E.D.

    1992-12-01

    This report discusses the Accelerator Transmutation of Waste (ATW) concept which aims at destruction of key long-lived radionuclides in high-level nuclear waste (HLW), both fission products and actinides. This focus makes it different from most other transmutation concepts which concentrate primarily on actinide burning. The ATW system uses an accelerator-driven, sub-critical assembly to create an intense thermal neutron environment for radionuclide transmutation. This feature allows rapid transmutation under low-inventory system conditions, which in turn, has a direct impact on the size of chemical separations and materials handling components of the system. Inventories in ATW are factors of eight to thirty times smaller than reactor systems of equivalent thermal power. Chemical separations systems are relatively small in scale and can be optimized to achieve high decontamination factors and minimized waste streams. The low-inventory feature also directly impacts material amounts remaining in the system at its end of life. In addition to its low-inventory operation, the accelerator-driven neutron source features of ATW are key to providing a sufficient level of neutrons to allow transmutation of long-lived fission products.

  10. An ATWS Analysis for EU-APR1400 Following the European Utility Requirement

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Minshin; Lee, Cheolshin; Sohn, Jongjoo [KEPCO Engineering and Construction Company., Inc., Daejeon (Korea, Republic of)

    2013-05-15

    This paper presents the results of the evaluation of the ATWS events with respect to Reactor Coolant System (RCS) overpressure and re-criticality for the European APR1400 (EU-APR1400) according to European Utility Requirement (EUR). This paper evaluates the ATWS impact on the EU-APR1400 by applying EUR. Based on the results of safety analysis for ATWS events, all the acceptance criteria for EUR can be satisfied due to the proper functioning of ATWS mitigation systems. However the four events are investigated only in this paper, and hence the results of this paper can not be concluded that the EU-APR1400 design satisfy all requirements for the EUR. Therefore, a further study for all Design Basis Event Category 2 (DBC2) events with ATWS needs to be performed in order to assess the comprehensive impact of ATWS events for the EU-APR1400 design.

  11. Abnormal grain growth in Ni-5at.%W

    Science.gov (United States)

    Witte, M.; Belde, M.; Barrales Mora, L.; de Boer, N.; Gilges, S.; Klöwer, J.; Gottstein, G.

    2012-12-01

    The growth of abnormally large grains in textured Ni-5at.%W substrates for high-temperature superconductors deteriorates the sharp texture of these materials and thus has to be avoided. Therefore the growth of abnormal grains is investigated and how it is influenced by the grain orientation and the annealing atmosphere. Texture measurements and grain growth simulations show that the grain orientation only matters so far that a high-angle grain boundary exists between an abnormally growing grain and the Cube-orientated matrix grains. The annealing atmosphere has a large influence on abnormal grain growth which is attributed to the differences in oxygen partial pressure.

  12. Reactivity transients during a blowdown in a MSIV closure ATWS

    International Nuclear Information System (INIS)

    Anticipated transients without scram (ATWS) events have received considerable attention in the past and are still a subject of great interest in severe-accident analysis. Of special interest is the effect of the low-pressure emergency core cooling system (ECCS) on the plant response following a blowdown by the automatic depressurization system (ADS). There is a potential for positive reactivity insertion due to the cold water injection of the low-pressure coolant injection (LPCI) system and the low-pressure core spray system in a boiling water reactor (BWR)/4. The main concern is whether a power excursion and pressure oscillation can occur in such an event. Furthermore, since thermal-hydraulic feedback plays an important role in these accidents, the uncertainty of the reactivity feedback coefficients used can impact the outcome of the analysis for such a power excursion. The objectives of the work reported in this paper are to study the consequences of the reactivity transients during a blowdown in an ATWS event with closure of the main steam isolation valves (MSIVs) and to evaluate the effect of the LPCI system and the sensitivity of plant response to the feedback coefficients. This work was performed with the Brookhaven National Laboratory plant analyzer

  13. Quantification of operator actions during ATWS following MSIV closure

    International Nuclear Information System (INIS)

    Brookhaven National Laboratory (BNL) assisted the Accident Sequence Evaluation Program (ASEP) by performing a Human Reliability Analysis (HRA) of the operations crew tasks during the Anticipated Transient Without Scram (ATWS) accident sequence with Main Steam Isolation Valve (MSIV) closure at the Peach Bottom Atomic Power Station, Unit 2. A detailed task analysis was performed based on consideration of staffing, team interaction, and control room layout at Peach Bottom. ATWS scenarios developed by Oak Ridge National Laboratory (ORNL) and Idaho National Engineering Laboratory (INEL) were reviewed. Discussions were held with thermal-hydrodynamic/core neutronics engineers at BNL to determine the success criterion for tasks. Five major operator tasks were identified. After reviewing a computerized data base of human error probabilities (HEPs) from 19 probabilistic risk assessments (PRAs) for tasks similar to those above to establish the historic range of HEPs for such errors, consensus opinion and structured expert judgment was used to quantify each of these tasks at each branch point in the event tree within that range

  14. Contribution of Anticipated Transients Without Scram (ATWS) to core melt at United States nuclear power plants

    International Nuclear Information System (INIS)

    This report looks at WASH-1400 and several other Probabilistic Risk Assessments (PRAs) and Probabilistic Safety Studies (PSSs) to determine the contribution of Anticipated Transients Without Scram (ATWS) events to the total core melt probability at eight nuclear power plants in the United States. After considering each plant individually, the results are compared from plant to plant to see if any generic conclusions regarding ATWS, or core melt in general, can be made. 8 refs., 34 tabs

  15. Kernkraftwerke Lippe-Ems GmbH (KLE). 1996 annual report

    International Nuclear Information System (INIS)

    The Kernkraftwerke Lippe-Ems GmbH (KLE) operates the Emsland nuclear power station at Lingen (Ems), equipped with a 1 300 MWatt PWR. Shareholders are VEW ENERGIE, PreussenElektra, and RWE Energie. The Emsland power plant over the reporting period was operated in the base regime under full-load operating conditions. Gross electricity output was 11 137 million kWh, the highest ever annual output of the plant. Net electricity generation over the reporting period was 10 557 million kWh. The shareholders decided to reduce the share capital by 200 000 thousand Deutschmarks to 900 000 TDM, reduction to become effective in spring 1997. Expenses of KLE are primarily determined by the fuel costs and spent fuel management costs, as well as for decmmissioning activities and by write-offs. The electricity prices determined by agreement with the shareholders cover all costs and include a suitable return on share capital employed. The number of persons employed was 286. (Orig./DG)

  16. Design of an FPGA-based PWR ATWS mitigation system

    International Nuclear Information System (INIS)

    The present research is to explore the feasibility and conceptual design by using triple-redundant FPGA-based system for Anticipated-Transient-Without-Scram (ATWS) Mitigation System and Actuation Circuit (AMSAC) of a pressurized water reactor (PWR) type nuclear power plant (NPP). The Taipower's (Taiwan Power Company) Maanshan NPP was chosen for demonstration. An engineering simulated interface between AMSAC system and reactor/plant systems of Maanshan NPP was developed to provide an environment to validate the triple-redundant FPGA-based system. The software-free FPGA-based nuclear instrumentation and control (I and C) systems can easily be used for the modernization of the Taipower's nuclear power plant analog systems, thus may reduce the safety risk of undetectable software faults and common cause failures, and also minimize the regulatory licensing efforts and cost. (author)

  17. Licensing issues in the context of terrorist attacks on nuclear power plants; Genehmigungsrechtliche Fragen terroristischer Angriffe auf Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Danwitz, T. von

    2002-07-01

    The terrorist attack on the World Trade Center in September 2001 has prompted enhanced nuclear risk awareness among the German population. But in the current public debate about the safety of nuclear power plants in Germany in times of new dimensions of danger, aspects such as the role of the constitutional law, the German Atomic Energy Act, and the regulatory system governing nuclear power plant licensing in the context of protection and safety have not been addressed. The author therefore discusses the German nuclear power plant licensing law and administrative regime, elaborating on the significance attributed in those bodies of law to risks like terrorist attacks on nuclear power plants. (orig./CB) [German] Das allgemeine Risiko von terroristischen Anschlaegen auf Kernkraftwerke ist durch die Ereignisse vom 11. September 2001 wieder verstaerkt in das Bewusstsein der Oeffentlichkeit getreten. Die verfassungsrechtlichen Grundlagen und die atomgesetzliche Einordnung der Risiken von terroristischen Angriffen auf kerntechnische Anlagen bleiben jedoch in der aktuellen Diskussion weithin ungeklaert. Der Beitrag unternimmt es daher, die verfassungs- und verwaltungsrechtliche Bedeutung der Risiken terroristischer Angriffe auf Kernkraftwerke in atomrechtlichen Genehmigungsverfahren zu untersuchen. (orig./CB)

  18. Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group

    International Nuclear Information System (INIS)

    In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD and D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years

  19. Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group.

    Energy Technology Data Exchange (ETDEWEB)

    Collins, E.; Duguid, J.; Henry, R.; Karell, E.; Laidler, J.; McDeavitt, S.; Thompson, M.; Toth, M.; Williamson, M.; Willit, J.

    1999-08-12

    In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD&D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years.

  20. A Los Alamos concept for accelerator transmutation of waste and energy production (ATW)

    International Nuclear Information System (INIS)

    This document contains the diagrams presented at the ATW (Accelerator Transmutation of Waste and Energy Production) External Review, December 10-12, 1990, held at Los Alamos National Laboratory. Included are the charge to the committee and the presentations for the committee's review. Topics of the presentations included an overview of the concept, LINAC technology, near-term application -- high-level defense wastes (intense thermal neutron source, chemistry and materials), advanced application of the ATW concept -- fission energy without a high-level waste stream (overview, advanced technology, and advanced chemistry), and a summary of the research issues

  1. Study of safety relief valve operation under ATWS conditions. [Supercritical flow

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Whitten, S.D.

    1979-09-01

    In March 1979, the NRC published a report (NUREG/CR-0687) prepared by the Energy Technology Engineering Center (ETEC-TDR-78-19). That report presented a literature survey which updated earlier NRC studies of saturated or subcooled water flow through relief valves, under ATWS conditions. This supplement expands upon that search to include supercritical steam-water flow. No applicable data for the supercritical conditions were found, nor were any newer data on saturated or subcooled conditions uncovered. This supplement also updates a look for facilities currently capable of simultaneously imposing all ATWS conditions upon test relief valves. Results confirmed the negative findings of NUREG/CR-0687.

  2. Study of safety relief valve operation under ATWS conditions. [Super critical flow

    Energy Technology Data Exchange (ETDEWEB)

    Hutmacher, E.S.; Whitten, S.D.

    1979-07-25

    In March 1979, ETEC published as ETEC-TDR-78-19 a search which updated earlier NRC studies of saturated or subcooled water flow through relief valves, under ATWS conditions. This Supplement expands upon that search to include supercritical steam-water flow. No applicable data for the supercritical conditions was found, nor were any newer data on saturated or subcooled conditions uncovered. The Supplement also updated a look for facilities currently capable of simultaneously imposing all ATWS conditions upon test relief valves. Results confirmed the negative findings of ETEC-TDR-78.19.

  3. Nuclear energy and politics in Russian ATWS conditions

    International Nuclear Information System (INIS)

    Relations between politics and nuclear power in the countries of sustainable development has been many times discussed during the short history of nuclear energy, and regularly arising new events, even very important (in Sweden, USA, etc.), just add to the formed understanding of the problem. Russia for 10 years lives in conditions of a transition period, which seems similar to ATWS-type accidents at nuclear power plants. In these conditions the effect of politics on nuclear power and vice versa are seen very clearly, and, more important, change swiftly, which may present interest for the countries with smoother public processes. The role of political processes in nuclear power is obvious and may be reduced to three main factors: change of political system and transition to market economy have placed nuclear power, though still within state sector, in an absolutely new economic condition, which determine its today's situation as 'Survival'; new possibilities of political influence and opposition to nuclear power (mainly struggle against construction of new nuclear fuel cycle objects) on a levels of authority (local, regional, federal); impact of the USSR collapse on the situation in Russian nuclear power was due sooner to temporary weakening of control and regulatory structures, than to the fact, that some fuel cycle elements have found themselves abroad (the factor of uranium resources' loss is unimportant at present). Nuclear safety was chosen to be the subject of Moscow 1996 Summit, initiated with the purpose of Russia coming closer to G7. The Summit has confirmed the thesis on the possibility of nuclear power o play an important role in the world energy demand in accordance with sustainable development goals. successful activities of Russia-USA Commission for economic and technological cooperation, known as 'Gore-Chernomyrdin' Commission, is to a large extent determined by positive nuclear decisions. Eastern direction of Russian nuclear export (Iran, China

  4. RAMONA-3B calculations for Browns Ferry ATWS [Anticipated Transient Without Scram] study

    International Nuclear Information System (INIS)

    Several aspects of the Anticipated Transient Without Scram (ATWS) initiated by an inadvertent closure of all Main Steam Isolation Valves (MSIV) in a typical BWR/4 are analyzed in the report. The analysis is performed using the Brookhaven National Laboratory code, RAMONA-3B, which employs a three-dimensional neutron kinetics model coupled with a parallel-channel thermal hydraulics in representing a Boiling Water Reactor (BWR) Core. Four different transient scenarios have been investigated: (a) downcomer water level and reactor pressure control, (b) manual control rod insertion transient, (c) high pressure boil-off, and (d) recirculation pump trip failure. Results of these calculations should provide better understanding of mitigative effects of operator actions during ATWS, thus helping in the development of adequate Emergency Procedure Guidelines (EPG) required for the BWR plant safety. A few unresolved questions subject to future investigations are also discussed

  5. RAMONA-3B calculations for Browns Ferry ATWS (Anticipated Transient Without Scram) study

    Energy Technology Data Exchange (ETDEWEB)

    Saha, P; Slovik, G C; Neymotin, L Y

    1987-02-01

    Several aspects of the Anticipated Transient Without Scram (ATWS) initiated by an inadvertent closure of all Main Steam Isolation Valves (MSIV) in a typical BWR/4 are analyzed in the report. The analysis is performed using the Brookhaven National Laboratory code, RAMONA-3B, which employs a three-dimensional neutron kinetics model coupled with a parallel-channel thermal hydraulics in representing a Boiling Water Reactor (BWR) Core. Four different transient scenarios have been investigated: (a) downcomer water level and reactor pressure control, (b) manual control rod insertion transient, (c) high pressure boil-off, and (d) recirculation pump trip failure. Results of these calculations should provide better understanding of mitigative effects of operator actions during ATWS, thus helping in the development of adequate Emergency Procedure Guidelines (EPG) required for the BWR plant safety. A few unresolved questions subject to future investigations are also discussed.

  6. Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

    International Nuclear Information System (INIS)

    The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available

  7. Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

    Energy Technology Data Exchange (ETDEWEB)

    Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available.

  8. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  9. Enhanced passive safety features against ATWS of fast breeder reactors with capabilities of MA incineration

    Energy Technology Data Exchange (ETDEWEB)

    Ninokata, Hisashi; Sawada, Tetsuo; Sato, Manabu [Tokyo Institute of Technology (Japan)] [and others

    1997-12-01

    The paper gives an outline of the general and simple reactivity correlation method to identify the region of the major design parameters that assures power stabilization and passive shutdown of sodium-cooled large fast reactors under ATWS conditions. Based on the model developed, general design guidelines are shown that enhance passive capabilities being aimed at preventing sodium boiling and fuel failures in the events of ULOF and UTOP. Discussions extend to the influences of minor actinides loading in the core onto the passive safety features. 6 refs., 1 fig., 1 tab.

  10. The feasibility study I on the blanket fuel options for the ATW/HYPER

    International Nuclear Information System (INIS)

    The choice of a blanket fuel cycle technology and the fuel type for HYPER/ATW are important to develop an ADS with better economics, performance and safety. Even though several fuel types have been considered as an alternative of the blanket fuels for HYPER/ATW, the metal alloy and the dispersion fuels were selected as the candidate fuels for ADS, and the technical feasibilities for both fuels are evaluated in this report. General performance characteristics, fabrication abilities, technical aspects, safety aspects, economics, and non-proliferation aspects for each fuel type are reviewed and evaluated. And some technological problems are addressed in this report, focused on the development strategy, the roadmaps, and the flexibility to meet the missions and specific designs. This study has been performed at the first stage of conceptual design. Since it is under the lack of physical properties for each fuel material, no an attempt is made to select the best fuel option, but the more better fuel options are recommended

  11. The feasibility study I on the blanket fuel options for the ATW/HYPER

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok; Meyer, M.K; Hayes, S.L

    2001-01-01

    The choice of a blanket fuel cycle technology and the fuel type for HYPER/ATW are important to develop an ADS with better economics, performance and safety. Even though several fuel types have been considered as an alternative of the blanket fuels for HYPER/ATW, the metal alloy and the dispersion fuels were selected as the candidate fuels for ADS, and the technical feasibilities for both fuels are evaluated in this report. General performance characteristics, fabrication abilities, technical aspects, safety aspects, economics, and non-proliferation aspects for each fuel type are reviewed and evaluated. And some technological problems are addressed in this report, focused on the development strategy, the roadmaps, and the flexibility to meet the missions and specific designs. This study has been performed at the first stage of conceptual design. Since it is under the lack of physical properties for each fuel material, no an attempt is made to select the best fuel option, but the more better fuel options are recommended.

  12. LOCA- and ATWS-calculations for homogeneous and heterogeneous advanced pressurized water reactors

    International Nuclear Information System (INIS)

    LOCA and ATWS calculations have been performed for the two KfK reference designs (homogeneous with p/d=1.2 and heterogeneous reactor) of APWR and for a homogeneous reactor with a tighter fuel rod lattice (p/d=1.123) as well as for a reference PWR. The calculations have been performed with the Ispra version of the code RELAP5/Mod.1. New correlations have been introduced in the code to account for the core geometry, which is different from that of a PWR. The results of the calculations show that during the LOCA the fuel rod cladding hot spot temperatures in the seed of the heterogeneous reactor reach values which are about 2500C higher than the corresponding temperatures for a PWR, and that during the ATWS the pressure inside the primary circuit exceeds the maximal allowable pressure in the case of the homogeneous reactor with p/d=1.123. On the basis of the present calculations only the homogeneous reactor with p/d=1.2 appears to be acceptable from a safety point of view. These results need of course experimental confirmation. (orig.)

  13. Generic implications of ATWS events at the Salem Nuclear Power Plant. Licensee and staff actions

    International Nuclear Information System (INIS)

    This report, Volume 2 of two volumes of NUREG-1000, describes the intermediate term actions to be taken by licensees and applicants of the US Nuclear Regulatory Commission (NRC), on the one hand, and by NRC staff, on the other, to address the generic issues raised by two anticipated transients without scram (ATWS) at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25, 1983. These actions came about as a result of the findings of NUREG-1000, Volume 1, and of reviews by the NRC Committee to Review Generic Requirements, the NRC Program Offices, and the Commission. The actions to be taken by licensees and applicants have been detailed in a letter pursuant to 10 CFR 50.54(f)

  14. Application of expert system and neural network in diagnosis during BWR ATWS sequences

    International Nuclear Information System (INIS)

    A prototype operator aid system employing an expert system and neural network is designed to help the plant operator during a BWR ATWS accident. The expert system is the driver of the inference engine, it consists of IF -- THEN -- and DO -- format rules developed from the knowledge base. A back propagation neural network is used when the operator can not supply the needed information to the expert system. Data of various plant parameters are fed into a pretrained neural network for transient identification. The case signature is then fed into the expert system, where a decision is made regarding the proper operator response. Testing results show that the neural network can retrieve the transients correctly even when random noise is added or the input data is incomplete. The computer simulation of the integrated system has also been demonstrated

  15. CSAU methodology and results for an ATWS event in a BWR using information theory methods

    International Nuclear Information System (INIS)

    Highlights: • We apply the CSAU methodology to an ATWS in a BWR using information theory methods. • We show how to perform the selection of the most influential inputs on the critical safety parameter. • We apply the maximum entropy principle to get the input parameter distribution. • We examine the maximum relative entropy principle to update the input parameter PDF. • We quantify the uncertainty of the critical safety parameter using order statistics and information theory. - Abstract: This paper shows an application of the CSAU methodology to an ATWS in a BWR reactor, when the temperature of the suppression pool is taken as the critical safety parameter. The method combines CSAU methodology with recent techniques of information theory. In this paper we use auxiliary tools to help in the evaluation and improvement of the parameters distribution that enter in the elements II and III of CSAU based methodologies. These tools have been implemented in two FORTRAN programs: GEDIPA (Generation of the Parameter Distribution) and UNTHERCO (Uncertainty in Thermal Hydraulic Codes). The first one analyzes the information data available on a given parameter or parameters with the goal to know all the information about the probability distribution function of these parameters. The second apply information theory methods, as the maximum entropy principle (MEP) and the maximum relative entropy Principle (MREP), in order to build conservative distribution functions for the parameters from the available data. Also, the distribution function of a given parameter can be updated using the MREP principle when new information is provided. UNTHERCO performs the MONTECARLO sampling for a given set of parameters when the distribution function of these parameters is previously known. If the distribution of a parameter is unknown, then, the MEP is applied to deduce the distribution function for this parameter

  16. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant

    International Nuclear Information System (INIS)

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133 degrees F) has a 95-percentile uncertainty of 14.4 K (26 degrees F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175 degrees F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6 degrees F)

  17. An analysis of reactor transient response for boiling water reactor ATWS events

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼21 Btu/lbm. A sensitivity study with regard to the steam separator pressure - loss coefficient, however, indicates that system stability is strongly dependent on the magnitude of this parameter which suggests a significant degree of uncertainty In the results. Under isolation conditions at rated pressure, stability is significantly enhanced by rapid pressure fluctuations generated through cycling of safety/relief valves. Large-amplitude instabilities develop, however, in depressurization events, and SRV cycling no longer stabilizes the system. In a simulated depressurization to ∼500 psia, prompt critical excursions occurred, and oscillation amplitudes reached 1000% of rated power. Implications of the Present Study: With the exception of guidance to avoid SRV cycling, these preliminary results have provided further support for the validity of this response strategy. SABRE calculations have shown that the reactor is probably slightly unstable in natural circulation operation, but relief valve cycling prevents the occurrence of instabilities at or near design pressure. Thus, unstable operation should not be a concern when boron injection and HPCI are available and depressurization is unnecessary. The reactor water level for injection flow corresponding to HPCI operation has been shown to be acceptable and consistent with earlier estimates based on the NSAC results; however, condensation effects have a significant influence on the equilibrium reactor water level. In addition, the SABRE results reinforce PP and L's concerns that

  18. Investigation of sensors and instrument components in boiling water reactors. Results from Oskarshamn 2, Barsebaeck 2 in Sweden and Kernkraftwerk Muehleberg in Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Bergdahl, B.G. [GSM Power Systems AB, Nykoeping (Sweden)

    1998-05-01

    The reactor monitoring instruments are important for the operation and safety of the plants. Static properties of the instruments are controlled annually, but the dynamic properties are rarely, if ever, examined. This study is the result of a project initiated by the Swedish Nuclear Power Inspectorate. The examinations are based on signal analysis and simultaneous measurement of multiple signals. Results from Oskarshamn 2 (O2), Barsebaeck 2 (B2) and Kernkraftwerk Muehleberg (KKM) are discussed in this report. The presentation is focused on reactor pressure and reactor level signals. the analysis of O2 revealed that the dynamics for 3 out of 14 sensors was `filtered`, meaning that a rapid level displacement is registered with delay. Inspection showed that a 1 sec filter was installed instead of 1.2 sec. The study also showed that old pressure-sensors in use both at O2 and B2 could not cope with high frequencies, and that some level-sensors were disturbed by mechanical oscillations at Bw. At KKM, a 2 Hz resonance was observed with 12 pressure and level sensors. The oscillation was created by an old pressure sensor and influenced the other sensors through the common impulse network 11 refs, 46 figs

  19. Investigation of sensors and instrument components in boiling water reactors. Results from Oskarshamn 2, Barsebaeck 2 in Sweden and Kernkraftwerk Muehleberg in Switzerland

    International Nuclear Information System (INIS)

    The reactor monitoring instruments are important for the operation and safety of the plants. Static properties of the instruments are controlled annually, but the dynamic properties are rarely, if ever, examined. This study is the result of a project initiated by the Swedish Nuclear Power Inspectorate. The examinations are based on signal analysis and simultaneous measurement of multiple signals. Results from Oskarshamn 2 (O2), Barsebaeck 2 (B2) and Kernkraftwerk Muehleberg (KKM) are discussed in this report. The presentation is focused on reactor pressure and reactor level signals. the analysis of O2 revealed that the dynamics for 3 out of 14 sensors was 'filtered', meaning that a rapid level displacement is registered with delay. Inspection showed that a 1 sec filter was installed instead of 1.2 sec. The study also showed that old pressure-sensors in use both at O2 and B2 could not cope with high frequencies, and that some level-sensors were disturbed by mechanical oscillations at Bw. At KKM, a 2 Hz resonance was observed with 12 pressure and level sensors. The oscillation was created by an old pressure sensor and influenced the other sensors through the common impulse network

  20. RSM modelling of an ATWS accident simulated by the ALMOD code: methodological and practical achievement

    International Nuclear Information System (INIS)

    A simulation study of a PWR station black-out ATWS has been performed by applying Response Surface Methodology (RSM) on the data obtained by inspecting the ALMOD code. The case under study has shown that the a priori information which alone could be inadequate, is optimally utilized if coupled with a preliminary sensitivity analysis through RSM techniques. In particular the engineering selection of the model variables and the rank order of the remaining ones had to be modified after an RSM preliminary sensitivity analysis. An other qualifying feature of the exercise is the use of randomization of the variables not included in the model in order to coherently exploit the methodology in its full efficiency. This procedure is able to give a figure of merit of the global importance of the neglected variables through the analysis of residuals. Results show that the proposed technique is an effective tool for selecting the most important accident variables and that the body of information gained is significant with respect to the number of observations performed

  1. An analysis of reactor transient response for boiling water reactor ATWS events

    International Nuclear Information System (INIS)

    Numerical simulations of BWR (boiling water reactor) dynamic response under ATWS (anticipated transient without scram) conditions are presented for the case where the reactor is operated at natural circulation conditions. In non-isolation events, reactor stability is strongly influenced by the degree of core inlet subcooling. At normal water level and pressure, instabilities develop if core-inlet subcooling exceeds a critical value of ∼ 21 Btu/Ibm. A sensitivity study with regard to the steam separator pressure-loss coefficient, however, indicates that system stability is strongly dependent on the magnitude of this parameter which suggests a significant degree of uncertainty in the results. Under isolation conditions at rated pressure, stability is significantly enhanced by rapid pressure fluctuations generated through cycling of safety/relief valves. Large-amplitude instabilities develop, however, in depressurization events, and SRV cycling no longer stabilizes the system. In a simulated depressurization to ∼ 500 psia, prompt critical excursions occurred, and oscillation amplitudes reached 1000% of rated power

  2. Reliability and availability considerations in the RF systems of ATW-class accelerators

    Science.gov (United States)

    Tallerico, Paul J.; Lynch, Michael T.; Lawrence, George

    1995-09-01

    In an RF-driven, ion accelerator for waste transmutation or nuclear material production, the overall availability is perhaps the most important specification. The synchronism requirements in an ion accelerator, as contrasted to an electron accelerator, cause a failure of an RF source to have a greater consequence. These large machines also are major capital investments, so the availability determines the return on this capital. RF system design methods to insure a high availability without paying a serious cost penalty are the subject of this paper. The overall availability goal in our present designs is 75% for the entire ATW complex, and from 25 to 35% of the unavailability is allocated to the RF system, since it is one of the most complicated subsystems in the complex. The allowed down time for the RF system (including the linac and all other subsystems) is then only 7 to 9% of the operating time per year, or as little as 613 hours per year, for continuous operation. Since large accelerators consume large amounts of electrical power, excellent efficiency is also required with the excellent availability. The availability also influences the sizes of the RF components: smaller components may fail and yet the accelerator may still meet all specifications. Larger components are also attractive, since the cost of an RF system usually increases as the square root of the number of RF systems utilized. In some cases, there is a reliability penalty that accompanies the cost savings from using larger components. We discuss these factors, and present an availability model that allows one to examine these trade offs, and make rational choices in the RF and accelerator system designs.

  3. Reactivity transients during a blowdown in a MSIV [Main Steam Isolation Valves] closure ATWS [Anticipated Transients Without Scram

    International Nuclear Information System (INIS)

    The objectives of this work are to study the consequences of the reactivity transients during a blowdown in an ATWS event with closure of the Main Steam Isolation Valves (MSIV), and to evaluate the effect of the LPCI (Low Pressure Coolant Injection) system and the sensitivity of plant response to the feedback coefficients. The present work was performed with the BNL Plant Analyzer (BPA). The BPA is a on-line, interactive BWR system code which models the non-homogeneous, non-equilibrium two-phase flow with a drift flux mixture model, the reactor kinetics with a point kinetic model, the thermal conduction with an integral method, and the control and plant protection systems with modern control theory. It also models the balance of plant (BOP) as well as the Mark I containment of a BWR/4. Thus, the BPA is a comprehensive engineering plant analyzer transients as well as accidents (e.g., ATWS and Small Break Loss of Coolant Accidents)

  4. Development of cube textured Ni-5 at.%W alloy substrates for coated conductor application using a melting process

    International Nuclear Information System (INIS)

    Biaxially textured Ni-5 at.%W substrates have been prepared by cold rolling, followed by three different annealing routes. In this paper, the processes of melting Ni and W metals, flat rolling, various annealing methods are described in detail. The Ni-5 at.%W tapes annealed under either high vacuum or flowing Ar (7% H2) gas were characterized by X-ray pole figures, ODF, EBSD as well as AFM analysis. The texture analysis indicated that as fabricated tapes have a sharp cube texture formed after annealing at a wide temperature range of 800-1100 oC. The high quality of cube orientation on tapes was obtained after a two-step annealing (TSA), where the percentage of the cube texture component was as high as 93.5% within a misorientation angle smaller than 8o from EBSD analysis. Furthermore, it was also observed that the number of twin boundaries in this tape decreased with respect to that of tapes annealed both in vacuum and one-step gas annealing. From AFM on 1 μm2 areas, it was concluded that the roughness (RMS) on the tape surface reached 0.98 nm

  5. Spectroscopic classification of Gaia16atw and Gaia16aui with the SEDM (Spectra Energy Distribution Machine) on Palomar 60-inch (P60) telescope

    Science.gov (United States)

    Blagorodnova, N.; Neill, D.; Walters, R.

    2016-07-01

    The Caltech Time Domain Astronomy group reports the classification of Gaia16atw and Gaia16aui, discovered by the Gaia ESA survey. The observations were performed with the Spectral Energy Distribution Machine (SEDM)(http://www.astro.caltech.edu/sedm/, range 350-950nm, spectral resolution R~100) on Palomar 60-inch (P60) telescope.

  6. Evolution of microstructure, texture and topography during additional annealing of cube-textured Ni–5at.%W substrate for coated conductors

    DEFF Research Database (Denmark)

    Wulff, Anders Christian; Mishin, Oleg; Grivel, Jean-Claude

    2012-01-01

    Microstructure, texture and topography have been studied in a recrystallized Ni–5at.%W substrate before and after additional annealing at 1025C for 1 h. The initial recrystallized material contained a strong cube texture and a high fraction of low angle grain boundaries. R3 boundaries were also f...

  7. NRC Fact-Finding Task Force report on the ATWS event at Salem Nuclear Generating Station, Unit 1, on February 25, 1983

    International Nuclear Information System (INIS)

    An NRC Region I Task Force was established on March 1, 1983 to conduct fact finding and data collection with regard to the circumstances which led to an anticipated transient without scram (ATWS) event at the Public Service Electric and Gas Company's Salem Generating Station, Unit 1 on February 25, 1983. The charter of the Task Force was to determine the factual information pertinent to management and administrative controls which should have ensured proper operation of the reactor trip breakers in the solid state protection system. This report documents the findings of the Task Force along with its conclusions

  8. Highly textured Gd2Zr2O7 films grown on textured Ni-5 at.%W substrates by solution deposition route: Growth, texture evolution, and microstructure dependency

    DEFF Research Database (Denmark)

    Yue, Zhao; Grivel, Jean-Claude; Napari, M.;

    2012-01-01

    or crystallization in the thicker films. This work not only demonstrates a route for producing textured Gd2Zr2O7 buffer layers with dense structure directly on technical substrates, but also provides some fundamental understandings related to chemical solution derived films grown on metallic substrates.......Growth, texture evolution and microstructure dependency of solution derived Gd2Zr2O7 films deposited on textured Ni-5 at.%W substrates have been extensively studied. Influence of processing parameters, in particular annealing temperature and dwell time, as well as thickness effect on film texture...

  9. Dismantling reactor pressure vessel internals at the Stade nuclear power station. Another milestone reached on the way to green field conditions; Rueckbau der Reaktordruckbehaelter-Einbauten im Kernkraftwerk Stade. Ein weiterer Meilenstein auf dem Weg zur gruenen Wiese ist realisiert

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, Annette [AREVA NP GmbH, Erlangen (Germany); Knoll, Peter [Kernkraftwerk Stade GmbH und Co. oHG, Stade (Germany)

    2009-08-15

    In November 2003, the Stade nuclear power station (KKS) of E.ON Kernkraft GmbH was shut down for economic reasons. In its history of 31 years of operation up to that point in time KKS generated 152,460,660 MWh (gross) of electricity. From 1984 on, the reactor in addition supplied district heat to an adjacent saltworks. In early 2007 E.ON Kernkraftwerk GmbH, in the course of dismantling phase III, commissioned Areva NP to disassemble and package the reactor pressure vessel (RPV) internals. Within 17 months Areva did the entire engineering for this project, which was demanding in many respects. Over that period of time, detailed planning, drafting and licensing of the documents, design and manufacturing of the machines and facilities as well as their qualification, and personnel training were completed. Activities on site began in May 2008. Twelve months later, the RPV internals had been dismantled. This contractual milestone was reached even ahead of time. The shielding measures taken, ongoing optimization throughout the project phase, and the possibility to do without a containment allowed the estimated collective dose for these activities to be clearly underrun. The whole project was completed in late August 2009, and disassembly and packaging ready for repository storage of the RPV internals were carried out in a minimum of time. (orig.)

  10. Proof of radiation exposure in the vicinity of Kruemmel power plant by chromosomal analysis of the population and by enhanced environmental radioactivity; Nachweis einer Strahlenbelastung beim Kernkraftwerk Kruemmel durch Chromosomenanalyse in der Bevoelkerung und durch erhoehte Umweltradioaktivitaet

    Energy Technology Data Exchange (ETDEWEB)

    Dannheim, B.; Heimers, A.; Schmitz-Feuerhake, I.; Schroeder, H. [Fachbereich 1, Arbeitsgruppe Medizinische Physik, Bremen Univ. (Germany)

    2001-07-01

    The leukaemia cluster in the proximity of the German boiling water reactor Kruemmel was detected by a local physician. 9 cases in children were registered in the period 1990-1996 which corresponds to 5.6 fold increase in the 5 km region around the plant. An incidence study conducted between 1984-93 showed an elevated rate of leukaemias also in adults. Because the supervising ministry had attested undisturbed operation of the plant and no conspiceous radioactivity had been noticed at that time, we started an independent investigation. Radiation exposures during the operation of the plant were proven by chromosome aberration studies in the population and by analyses of the environmental radioactivity. (orig.) [German] Das Leukaemiecluster in unmittelbarer Naehe zum Kernkraftwerk Kruemmel war durch einen einheimischen Arzt entdeckt worden. Im Zeitraum 1990 bis 1996 stieg die Anzahl bei Kindern auf 9 Faelle an, woraus sich eine Erhoehung um den Faktor 5,6 ableitet. Eine Inzidenzstudie, die fuer den Zeitraum 1984-93 ausgefuehrt wurde, zeigte auch fuer Erwachsene eine erhoehte Leukaemierate. Da die Aufsichtsbehoerde in Kiel einen einwandfreien Betrieb konstatierte und keinerlei Hinweis fuer erhoehte Kontaminationen in der Umgebung sah, fuehrten wir eine unabhaengige Untersuchung durch. Anhand von Chromosomenaberrationsstudien in der Bevoelkerung und durch Analysen von Umgebungsueberwachungsmessungen stellten wir eine Strahlenbelastung waehrend der Betriebszeit der Anlage fest. (orig.)

  11. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. Fourth quarterly report 1997; Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. 4. Quartal 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-08-01

    The report presents a brief survey of notifiable events in German nuclear power plants and research reactors of the given output category, covering the last quarter of the year 1997. (orig./CB) [Deutsch] Der vorliegende Bericht enthaelt die Uebersicht ueber die meldepflichtigen Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen (Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet) der Bundesrepublik Deutschland fuer das vierte Vierteljahr 1997. (orig./AJ)

  12. RADDA - Comparison of results of three ATWS/ATWC scenarios simulated with the help of POLCA-T and S3K/RELAP5

    International Nuclear Information System (INIS)

    The effects of ATWS and ATWC-events with control rods failing to enter the core has been evaluated in this project. To understand the uncertainties in using modern 3D-calculation methods two different codes were used in the project. The outputs from the two code packages were compared. Within the project the used code were first evaluated against a real event, pancake core at Forsmark 3. The results give important knowledge of the core responses for such events and on how to use different code to perform such calculations. The NKS report is only one minor part of the total project. The project was sponsored by TVO, Forsmark, OKG, Ringhals, SKI besides the NKS-funding. The results could be used for PSA-studies and for deterministically safety analysis. (au)

  13. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant. An application of the CSAU methodology using the BNL engineering plant analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.; Cheng, H.S.; Mallen, A.N. [Brookhaven National Lab., Upton, NY (United States); Johnsen, G.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Lellouche, G.S. [Technical Data Services, Chicago, IL (United States)

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133{degrees}F) has a 95-percentile uncertainty of 14.4 K (26{degrees}F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175{degrees}F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6{degrees}F).

  14. Study on the Standardized Reliability Calculation Related to ATWS Function Execution%有关ATWS功能执行的可靠性计算标准化研究

    Institute of Scientific and Technical Information of China (English)

    李悠然; 孙伟; 刘爱国; 郭智武

    2015-01-01

    预期瞬态不停堆事故缓解( ATWS)系统是为了确保核电厂在紧急停堆保护发生故障的情况下,相关事故缓解措施能够有效执行的重要系统. 因此,基于ATWS系统的功能执行开展相关的可靠性计算研究是十分必要的. 以某工程技术方案为例,从ATWS系统功能设计要求、信号逻辑处理以及系统结构组成等几个方面开展研究,基于对可靠性计算方案和相关失效数据的分析研究,得出可供参考的计算结果和分析建议,以期为可靠性计算在核电工程中的标准化研究应用提供一定的经验积累和数据参考.%ATWS ( anticipated transient without scram) system is an important system for ensuring related accident mitigation measures can be effectively executed in the case of scram protection of nuclear power plant fails. Thus, the research on related reliability calculation based on function execution of ATWS system is necessary. With certain engineering technical scheme as example, the research is conducted from several aspects, e. g. , the requirements of functional design of ATWS system, signal logical processing, and system compositions, etc. On the basis of analysis and research on reliability calculation schemes and related failure data, the calculation results and analysis recommendations are provided for reference. It is expected that these can give certain accumulated experience and data reference for standardized study of reliability calculation in nuclear power engineering.

  15. Highly reinforced, low magnetic and biaxially textured Ni-7 at.%W/Ni-12 at.%W multi-layer substrates developed for coated conductors

    International Nuclear Information System (INIS)

    Mechanically strengthened, highly cube textured Ni-7 at.%W/Ni-12 at.%W multi-layer substrates developed for coated conductors have been prepared by the advanced spark plasma sintering technique. The key innovation for developing this weakly magnetic and reinforced substrate was to use a new powder metallurgy and sintering route to bond multi-layers of Ni7W/Ni12W/Ni7W together in order to get an initial ingot, followed by the optimized cold working and annealing. Particular efforts were made in view of the optimization of the design, pressing as well as the heat treatment processes of the starting ingots to obtain a chemically gradient composite bulk, thus ensuring the subsequent cold deformation. The produced composite substrates have a strong {100} texture on Ni7W outer layers. The percentage of the biaxially orientated grains within a misorientation angle of 10 deg. is as high as 97.5%, while the length percentage of low-angle grain boundaries ranging from 2 deg. to 10 deg. in the composite substrate reaches 87.2%. Moreover, the yield strength σ0.2 of the tape approaches 333 MPa, and the saturation magnetization is substantially reduced by 81.6% at 77 K when compared to that of a commercial used Ni5W substrate

  16. Expert report of ENSI on the request of KKN AG for a general license - Project 'New nuclear power plant Niederamt'; Gutachten des ENSI zum Rahmenbewilligungsgesuch der KKN AG. Neubauprojekt Kernkraftwerk Niederamt

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-09-15

    The 'Kernkraftwerk Niederamt AG' (KKN) Company submitted to the Swiss Federal Inspectorate of Nuclear Safety (ENSI) a request for a general license for a new power plant to be built near to the Goesgen power plant. According to the law, all damage risks with a probability higher than 10{sup -4}/a must be taken into account through protection measures. The considered risks concern the power plant itself as well as the population in the neighbourhood and the environment. The purpose of the general license is to demonstrate that the site chosen for the foreseen power plant is acceptable and that the risks can be counteracted through adequate measures. The buildings of the power plant and their partition over the two banks of the Aare River are briefly described. The reactor is a Light Water Reactor of third generation with a maximum thermal power of 5.8 GW{sub th}. The main cooling is provided by a hybrid system of water evaporation and air heating, what reduces the plume at the exit of the cooling tower. First, it is demonstrated that, in the case of a very unlikely severe accident in the power plant, the people in the neighbourhood can be evacuated quickly. Then, numerous types of possible accidents in the neighbourhood of the power plant are analyzed in order to settle their possible negative influence on the operation of the power plant: bursting of gas containers on the neighbouring roads and railways, fires of all types of hydrocarbons, air pollution through chloride gas, etc. The check by ENSI of the KKN studies on the potential danger for the power plant through neighbouring industrial plants, roads or railways demonstrated that none of the considered accidents presents an unacceptable risk for the power plant: on the one hand, these plants are located too far from the power plant, so that a sensible injury to the power plant safety can be excluded; on the other, the protection of the power plant can be guaranteed through appropriate technical

  17. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. Second quarterly report 1998; Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. Vierteljahresbericht 2. Quartal 1998

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-31

    The report contains the documentation of notifiable events in the defined reactors recorded over the second quarter of 1998. The documentation is prepared according to the national notification and reporting system prescribed by the relevant law in Germany, and is filed to the national atomic energy supervisory authorities in Germany for documentation in the national record. (orig./CB) [Deutsch] Der vorliegende Bericht enthaelt die Uebersicht ueber die meldepflichtigen Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen (Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistuung 50 kW thermische Dauerleistung ueberschreitet) der Bundesrepublik Deutschland fuer das zweite Vierteljahr 1998. Meldepflichtige Ereignisse in Kernkraftwerken der Bundesrepublik Deutschland werden seit 1975 nach bundeseinheitlichen Meldekriterien in der jeweils gueltigen Fassung an die atomrechtlichen Aufsichtsbehoerden gemeldet und in einer zentral gefuehrten Liste erfasst. (orig.)

  18. Core safety discussion under station blackout ATWS accident of solid fuel molten salt reactor%固态熔盐堆全厂断电ATWS事故工况下的堆芯安全探讨

    Institute of Scientific and Technical Information of China (English)

    焦小伟; 王凯; 何兆忠; 陈堃

    2015-01-01

    利用修改后的适用于固态熔盐堆的RELAP5/MOD4.0系统分析程序,对固态熔盐堆全厂断电ATWS(Anticipated Transient Without Scram)事故进行了分析.主回路系统进行了合理简化建模,模拟系统在全厂断电ATWS事故时非能动余热排出系统有效与否两种情况下的瞬态响应过程.分析结果表明:非能动余热排出系统在全厂断电ATWS事故初期作用不明显,但长期作用较明显,投入使用后最终将使堆芯温度和主冷却剂温度达到稳定;对于固态熔盐堆来说,即使非能动余热排出系统失效,燃料元件温度上升也很缓慢,给人员干预采取必要措施提供了超过20天的宽限时间.分析结果表明了固态熔盐堆在应对极端事件时具有高的安全性.

  19. 中间退火对 Ni-7 at%W 合金基带再结晶织构的影响%Effect of intermediate annealing on recrystallization texture of Ni-7 at%W alloy based-trip

    Institute of Scientific and Technical Information of China (English)

    胡汪洋; 陈纪昌; 刘二微; 王均安

    2014-01-01

    采用真空电磁搅拌电弧熔炼制备的Ni-7at%W合金初始锭,经热锻、热轧、大形变量冷轧和最终再结晶退火制备出100μm厚合金基带。采用X射线衍射(XRD)和电子背散射衍射(EBSD)研究了轧制过程中的中间退火处理对再结晶织构的影响。结果表明,提高中间退火次数可以削弱轧制织构α取向线上的取向强度,提高β取向线上的取向强度,促进轧制组织中立方取向晶核的形成,消除再结晶过程中晶粒异常长大现象,最终提高再结晶立方织构的比例。%The Ni-7at%W ingot was prepared by the vacuum arc melting with electromagnetic stirring , then the Ni-7at%W substrate with a thickness of 100 μm was fabricated by hot forging , hot rolling, cold rolling and recrystallization annealing .The effect of intermediate annealing treatment on the recrystallization texture of Ni-7at%W alloy substrate was investigated by X-ray diffraction ( XRD) and electron backscatter diffraction ( EBSD ) .The results show that the intermediate annealing treatments strengthen the intensity of β-fiber and at meantime reduce the intensity of α-fiber in the deformed alloy .Moreover , it favors the formation of cube nuclei in the rolling structure and the elimination of abnormal grain growth phenomenon during recrystallization . All of these will lead to higher volume fraction of recrystallization cube texture .

  20. Life extension of German nuclear power plants only with the consent of the Federal Council? The importance and extent of the need for consent to an amendment to the German Atomic Energy Act; Laengerer Betrieb der deutschen Kernkraftwerke nur mit Zustimmung des Bundesrates? Bedeutung und Reichweite der Zustimmungsbeduerftigkeit bei Aenderung des Atomgesetzes

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Horst

    2010-08-15

    In its coalition agreement of October 26, 2009, the new German federal government plans ''to extend the service life of German nuclear power plants while, at the same time, complying with the strict German and international safety standards.'' This has triggered a debate not only about (nuclear) energy, as in the past election campaign in the summer of 2009, but also about the constitutional law issue whether an amendment to the Atomic Energy Act resulting in longer operating life of nuclear power plants required the consent of the Federal Council (the ''Bundesrat,'' the second chamber of parliament). After the election to the state parliament in North Rhine-Westphalia on May 9, 2010, majority in the Federal Council changed. As a consequence, no consent to an amendment to the Atomic Energy Act must be expected. In view of the large number of recent statements about constitutional law in opinions for various federal and ministerial accounts as well as firms and associations, the outline by R. Scholz in the May issue of atw 2010 will be followed in this issue by the key points of examination of the need for consent, under aspects of constitutional law, and an attempt will be made to explain the evaluations underlying the generation of a legal concept about these items. The decision by the German Federal Constitutional Court of May 4, 2010, published on June 11, 2010, plays a major role in this respect because it established clarity in some important aspects of a legal subject matter in the field of state admini-stration on behalf of the federation, albeit in the field of air traffic law, not nuclear law. However, the structures of the norms in the German Basic Law (Art. 87c and Art. 87d, para.2) to be applied are almost identical. The energy policy and energy economy aspects of a plant life extension are considered along with the option of an appeal to the Federal Constitutional Court against any plant life extension. Finally

  1. Development of a procedure for qualitative and quantitative evaluation of human factors as a part of probabilistic safety assessments of nuclear power plants. Part B. Technical documentation; Entwicklung eines Verfahrens zur qualitativen und quantitativen Bewertung von Personalhandlungen in der probabilistischen Sicherheitsanalyse fuer Kernkraftwerke. Teil B. Technische Dokumentation

    Energy Technology Data Exchange (ETDEWEB)

    Richei, A.

    1998-06-01

    assignable performance shaping factors are gathered. However, regarding the evaluation of cognitive tasks, criteria are derived from a psychologic pattern which are included in a corresponding rule base. First Calculations for the validation of the procedure are performed based on three examples, reflecting the common practice of probabilistic safety assessments on the one hand and including problems, which cannot be evaluated or can hardly be evaluated with the established human risk analysis procedures on the other one. The results of the HEROS application are plausible, comprehensible and can be compared with results of established procedures. Basic approaches for optimizing the man-machine-system may be also obtained from a sensitivity studies using HEROS. The new procedure may be further developed to be applied to other technical scopes and allow further developments, particularly regarding the `learning` of data using the NeuroFuzzy-Technology. (orig.) [Deutsch] Internationale Studien belegen, dass Unfaelle in technischen Anlagen zunehmend auf Kombinationen von technischem und menschlichem Versagen zurueckzufuehren sind. Daher wird der Untersuchung der Zuverlaessigkeit von Mensch-Maschine-Wechselwirkungen weltweit grosse Beachtung beigemessen. Fuer Kernkraftwerke werden solche Untersuchungen im Rahmen von probabilistischen Sicherheitsanalysen durchgefuehrt. Hierzu existieren zwar eine Vielzahl von Verfahren, jedoch wird keines allen als wichtig erachteten Anforderungen - wie Uebertragbarkeit von Daten auf andere Anlagen, Schwachstellenanalyse und Bewertung kognitiver Handlungen - vollstaendig gerecht. Aufbauend auf einem modernen Modell eines Mensch-Maschine-Systems werden das Human Error Rate Assessment und Optimizing System (HEROS) und ein gleichnamiges Expertensystem vorgestellt, mit dem einerseits die menschliche Fehlerwahrscheinlichkeit fuer Personalhandlungen quantifiziert werden kann und andererseits auch qualitative Aussagen zur Optimierung des Mensch

  2. Kernkraftwerke Lippe-Ems GmbH. Report on the business year 1990

    International Nuclear Information System (INIS)

    The brochure describes the tasks and activities of this nuclear power utility in the form of the year-end report for 1990 laying open numerous financial data (balance-sheet, profit-and-loss accounting, etc.). The society's purpose is the construction and operation of nuclear power plants. At the site Lingen, KLE has operated since 1988 the Emsland nuclear power plant, which has a 1300 megawatt PWR-type reactor. (UA)

  3. Safety culture in nuclear power plants. Proceedings; Sicherheitskultur im Kernkraftwerk. Seminarbericht

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-01

    As a consequence of the INSAG-4 report on `safety culture`, published by the IAEA in 1991, the Federal Commission for the Safety of Nuclear Power Plants (KSA) decided to hold a one-day seminar as a first step in this field. The KSA is an advisory body of the Federal Government and the Federal Department of Transport and Energy (EVED). It comments on applications for licenses, observes the operation of nuclear power plants, assists with the preparation of regulations, monitors the progress of research in the field of nuclear safety, and makes proposals for research tasks. The objective of this seminar was to familiarise the participants with the principles of `safety culture`, with the experiences made in Switzerland and abroad with existing concepts, as well as to eliminate existing prejudices. The main points dealt with at this seminar were: - safety culture from the point of view of operators, - safety culture from the point of view of the authorities, - safety culture: collaboration between power plants, the authorities and research organisations, - trends and developments in the field of safety culture. Invitations to attend this seminar were extended to the management boards of companies operating Swiss nuclear power plants, and to representatives of the Swiss authorities responsible for the safety of nuclear power plants. All these organisations were represented by a large number of executive and specialist staff. We would like to express our sincerest thanks to the Head of the Federal Department of Transport and Energy for his kind patronage of this seminar. (author) figs., tabs., refs.

  4. KWL Lingen nuclear plant. Technical annual report 2015; KWL Kernkraftwerk Lingen. Technischer Jahresbericht 2015

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2016-07-01

    The technical annual report 2015 on the Lingen nuclear plant covers the following issues: report on the segments operation, process engineering, safety engineering, licensing and supervising procedures, operational data, radiation protection, radioactive materials, and in-service inspections.

  5. Stade. Decommissioning and dismantling of the nuclear power plant - from the nuclear power plant to the green lawn. 3. ed.; Stade. Stilllegung und Rueckbau des Kernkraftwerks - vom Kernkraftwerk zur ''Gruenen Wiese''

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    The nuclear power plant Stade (KKS) was shutdown in 2003 and is being dismantled since 2005. The contribution covers the following issues: What means decommissioning and dismantling? What was the reason for decommissioning? What experiences on the dismantling of nuclear power plants are available? What is the dismantling procedure? What challenges for the power plant personal result from dismantling? What happens with the deconstruction material? What happens with the resulting free area (the ''green lawn'')? What is the legal frame work for dismantling?.

  6. Stade nuclear power station (KKS): four giants on tour; Kernkraftwerk Stade - KKS: Vier Riesen gehen auf Reisen

    Energy Technology Data Exchange (ETDEWEB)

    Beverungen, M.; Viermann, J. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2008-04-15

    The Stade nuclear power station was the first nuclear power plant in the Federal Republic of Germany to deliver heat in addition to electricity. Since 1984, district heat was distributed to a saltworks nearby. The power plant, which is situated on the banks of the river Elbe, was commissioned in 1972 after approximately 4 years of construction. Together with the Wuergassen plant, it was among the first commercial nuclear power plants in this country. E.ON Kernkraft holds a 2/3 interest, Vattenfall Europe a 1/3 interest in the nuclear power plant. The Stade nuclear power station was decommissioned on November 14, 2003 for economic reasons which, in part, were also politically motivated. In September 2005, the permit for demolition of the nuclear part was granted. The release from supervision under the Atomic Energy Act is expected for 2014. In the course of demolition, the 4 steam generators of the Stade nuclear power station were removed. These components, which have an aggregate weight of approx. 660 tons, are to be safely re-used in Sweden. In September 2007, the steam generators were loaded on board the Swedish special vessel, MS Sigyn, by means of a floating crane. After shipment to Sweden, heavy-duty trucks carried the components to the processing hall of Studsvik AB for further treatment. After 6 months of treatment, the contaminated inner surfaces of the tube bundles of the steam generators have been decontaminated successfully, among other items. This has increased the volume of material available for recycling and thus decreased the volume of residues. (orig.)

  7. Nuclear knowledge-management. A core competence of VGB; Uebergreifendes Wissensmanagement fuer Kernkraftwerke. Eine VGB-Kernkompetenz

    Energy Technology Data Exchange (ETDEWEB)

    Pamme, Hartmut [RWE Power AG, Essen (Germany). Steuerung Kernkraftwerke

    2009-07-01

    It is a well established expectation that utilities/operators of nuclear power plants communicate their own operational situation and are able to comment promptly on any findings and events in the international nuclear scene. In order to gain synergies on knowledge management, utilities have been using VGB as common platform for many years. The paper describes the generic expectations concerning knowledge management towards an association like VGB. It is analysed which elements and peculiarities of modern knowledge management are already established within VGB in the nuclear field. (orig.)

  8. Harmonisation of licensing processes for decommissioning. Options and limitations; Genehmigungsverfahren fuer die Stilllegung der deutschen Kernkraftwerke. Konvoi oder Kakophonie?

    Energy Technology Data Exchange (ETDEWEB)

    Raetzke, Christian

    2016-03-15

    The shutdown of eight reactors in Germany in the wake of Fukushima 2011 and the scheduled phase-out of the remaining units in several steps ending 2022 has obviously triggered a wave of applications for decommissioning and dismantling licences. It would seem natural to strive for a harmonised handling of these processes, analogous to the 'convoi' concept which was successfully employed for licensing and construction of the three most recent German NPPs in the 1980s. However, a comparative analysis shows that the motivation of all players is much different from that of earlier times and that harmonisation of licensing processes for dismantling is not as crucial for operators, authorities and technical support organisations as it was for construction.

  9. VGH Mannheim: legitimacy of the decommissioning license for a nuclear power plant; VGH Mannheim: Rechtmaessigkeit der Stilllegungsgenehmigung fuer ein Kernkraftwerk

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2015-03-16

    The contribution describes the details of the court (VGH) decision on the legitimacy of the decommissioning license for the NPP Obrigheim. Inhabitants of the neighborhood (3 to 4.5 km distance from the NPP) are suspect hazards for life, health and property due to the dismantling of the nuclear power plant in case of an accident during the licensed measures or a terroristic attack with radioactive matter release.

  10. Source Term Analysis for the Nuclear Power Station Goesgen-Daeniken; Quelltermanalysen fuer das Kernkraftwerk Goesgen-Daeniken

    Energy Technology Data Exchange (ETDEWEB)

    Hosemann, J.P.; Megaritis, G.; Guentay, S.; Hirschmann, H.; Luebbesmeyer, D.; Lieber, K.; Jaeckel, B.; Birchley, J.; Duijvestijn, G

    2001-08-01

    Analyses are performed for three accident scenarios postulated to occur in the Goesgen Nuclear Power Plant, a 900 MWe Pressurised Water Reactor of Siemens design. The scenarios investigated comprise a Station Blackout and two separate cases of small break loss-of-coolant accident which lead, respectively, to high, intermediate and low pressure conditions in the reactor system. In each case the accident assumptions are highly pessimistic, so that the sequences span a large range of plant states and a damage phenomena. Thus the plant is evaluated for a diversity of potential safety challenges. A suite of analysis tools are used to examine the reactor coolant system response, the core heat-up, melting, fission product release from the reactor system, the transport and chemical behaviour of those fission products in the containment building, and the release of radioactivity (source term) to the environment. Comparison with reference values used by the licensing authority shows that the use of modern analysis tools and current knowledge can provide substantial reduction in the estimated source term. Of particular interest are insights gained from the analyses which indicate opportunities for operators to reduce or forestall the release. (author)

  11. Lifetime extension of nuclear power plants. Exclusive competence of the Bundestag?; Laufzeitverlaengerung fuer Kernkraftwerke. Entscheidung zwischen Bundestag und Bundesrat?

    Energy Technology Data Exchange (ETDEWEB)

    Scholz, Rupert [Muenchen Univ. (Germany). Inst. fuer Politik und oeffentliches Recht

    2010-05-15

    With the Act on the structured phase-out of the utilisation of nuclear energy for the commercial generation of electricity (Gesetz zur geordneten Beendigung der Kernenergienutzung zur gewerblichen Erzeugung von Elektrizitaet) of 22 April 2002 (Federal Gazette I p. 1351), the ''nuclear power phase-out'' was implemented into law. Ever since then, section 7 (1a) of the Atomic Energy Act (Atomgesetz - AtG) has provided that the authorisation to operate a nuclear power plant expires once the electricity volume for the respective installation as listed in Appendix 3, column 2 or the electricity volume derived from transfers has been produced. The coalition treaty of the current government factions provides for extending the operating periods of nuclear power plants. To this end, paragraphs 1a to 1d of section 7 AtG could be repealed, thus restoring the legal status prevailing prior to the ''phase-out''. As an alternative it would be conceivable to increase the values set forth in Appendix 3 for the energy volume quantity of a given installation accordingly. Both alternatives require an amendment of the Atomic Energy Act, over which the Deutsche Bundestag has exclusive competence. This is stated in the Grundgesetz (Constitution). Such a amendment would not require the consent of the Bundesrat, since the administrative tasks assigned to the Federal States (Laender) on behalf of the Federal Government pursuant to sec. 7, 24 (2) AtG would not be changing in a qualitative sense. Consequently, it would not constitute interference with the administrative powers of the Federal States from an organizational or procedural point of view. The quantitative change in the tasks to be performed by the Federal States on behalf of the Federal Government that would accompany an extension of the operating periods would not lead to a right of consent on the part of the Bundesrat pursuant to Art. 87c of the Grundgesetz. (orig.)

  12. Anticipated transients without scram for light water reactors. Appendices. Staff report

    International Nuclear Information System (INIS)

    Information is presented concerning scram failure probability, rod drive failure data, ATWS rule and ATWS requirements, treatment of steam generator tube failures in ATWS evaluation, radiological consequences assessments, ATWS study to include parameter variations and equipment reliability in probabilistic accident analysis, PWR MTC for ATWS, safety valve flows, ATWS contribution to risk, fuel integrity, value-impact analysis, and analytical methods

  13. Probability and consequences of severe reactor accidents. 60th year atw

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, Norman Carl [Massachusetts Institute of Technology (MIT), Cambridge, MA (United States). Dept. of Nuclear Engineering

    2015-06-15

    The study carried out on behalf of former USAEC (United States Atomic Energy Commission) led by Prof. Rasmussen and published in reworked form as WASH 1400 by the USNRC (United States Nuclear Regulatory Commission) in 1975, assessed in 3,300 pages the risks that can be deducted from severe accidents in nuclear power plants. The results, often quoted and criticised, were so far the most conclusive statements to this question. In his lecture at the reactor meeting in 1976, Prof. Rasmussen tried to trace back the conclusion of the results to the question: Is the use of larger nuclear power plants, in accordance to experiences and calculations so far, acceptable? His risk assessment, related to American power plants and cites, on behalf of the BMI is currently evaluated by the IRS together with the LRA on specific occurrences within the Federal Republic of Germany.

  14. Reduction of the consequences of accidents whereby the emergency shutdown system in modern reactors fails (ATWS)

    International Nuclear Information System (INIS)

    If a nuclear reactor can not be shutdown by pulling out the control rods, an emergency shutdown system must be used. The events, when such a system fails, have been calculated. Also attention is paid to the chance that both systems fail and the possibility of using an extra independent shutdown system, realized in pressurized water reactors (PWR) or boiling water reactors (BWR). Finally a General Electric developed safety method and an alternative method regarding the failure of an emergency shutdown system are described. The results of this investigation, which were also based on a literature study, can be applied in formulating specifications of new nuclear power plants

  15. Study on Fabrication of Ni-5 at.%W Tapes for Coated Conductors from Cylinder Ingots

    DEFF Research Database (Denmark)

    Ma, L.; Suo, H. L.; Yue, Zhao;

    2015-01-01

    observed that the fraction of cube texture within 10° from the ideal {001}〈100〉 orientation was ~98% and the fraction of LAGBs was ~90%. The as-obtained tapes have a strong cube texture also very close to the edge of the tape and they would therefore increase the fraction of applicable material while......, during heavy cold rolling, be characterized by a lower concentration of stress along the edges of the ingot. It can reduce fabrication costs and increase process efficiency. The fraction of cube texture on the surface of the finally recrystallized tapes was investigated using the EBSD technique. It was...... simplifying the heavy rolling process. Accordingly, it suggests that this fabrication method is a good choice to most small scale research laboratories for achieving long length Ni5W tapes for coated conductors with an easy way and a higher fraction of applicable material....

  16. Fabrication of the Textured Ni-9.3at.%W Alloy Substrate for Coated Conductors

    DEFF Research Database (Denmark)

    Gao, M. M.; Suo, H. L.; Grivel, Jean-Claude;

    2011-01-01

    It is difficult to obtain a sharp cube texture in the Ni-9.3at.% W substrate used for coated conductors due to its low stacking fault energy. In this paper, the traditional cold rolling procedure was optimized by introducing an intermediate recovery annealing. The deformation texture has been imp...

  17. Periodical safety review of the Goesgen-Daeniken nuclear power plant. Summary, results and evaluation; Periodische Sicherheitsueberpruefung fuer das Kernkraftwerk Goesgen-Daeniken. Zusammenfassung, Ergebnisse und Bewertung

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-15

    The Goesgen nuclear power plant (KKG) received its operational licence on September 9, 1978. The operational start-up of the plant went on into the year 1979, but there was a short interruption because of the accident in the Three Mile Island reactor on March 28, 1979. In May 1985 KKG submitted a request for raising the thermal reactor power from the then 2808 MW to 3002 MW. Based on the examination by the Federal Agency for the Safety of Nuclear Installations (HSK), the Swiss Federal Council granted the licence in two steps: in December 1985 for raising the thermal power to 2900 MW, and, in April 1992, to 3002 MW. The licence for the second step was given under the condition that some more experience was to be gained concerning the fuel rod cladding under higher loading. As part of the yearly re-licensing on restart after fuel assembly reloading, HSK confirmed that the plant status conformed to the legal requirements. In November 1986, HSK asked all Swiss nuclear power plant managers to state their opinions on proposed measures concerning severe accidents. Some of the measures were already in discussion; the Chernobyl accident on April 26, 1986, accelerated their implementation and was also a reason for the introduction of the measures against severe accidents. In this context, KKG carried out a risk study which led to the installation of a filtered pressure release system for the containment. Another consequence of the Chernobyl accident was the introduction of technical Periodical Safety Reviews (PSR) for all operating nuclear power plants. Central points of the PSR are: a) comparison with the continuously improving state-of-the-art of science and technology concerning safety precautions; b) a systematic evaluation of operating experience and plant status; c) the taking into account of probabilistic safety analyses in the overall evaluation of the plant. Within the framework of the examination of the overall plant, HSK also checks how its requirements concerning plant safety and radiation protection are taken into account. Even if the plant manager considers the guarantee of plant safety as his duty, an overall investigation by the authorities makes sense because it also looks into rare accident scenarios for which there are, of course, no actual working experience and which can only be considered within the framework of extended plant examinations. The PSRs on the Swiss nuclear power plants therefore complement the continuous control activities of the HSK; they are carried out about every 10 years. For KKG the PSR process was initiated by a letter from the HSK in February 1994. The areas to be considered were: a) examination of design and fulfilment of technical safety systems and comparison with the actual state-of-the-art of science and technology; b) evaluation of operational experience; c) review of the technical precautions against severe accidents including the preparation of emergency measures; d) review of the emergency organisation; e) examination of the plant protection against radioactivity; f) future dismantling at the end of operational life and disposal of the radioactive wastes; g) evaluation of accident analyses and of the KKG probabilistic safety analysis; h) review of plant organisation and plant management. The examination confirmed that, at KKG, there are very many technical safety precautions. KKG operational experience is good, the results show a high degree of operational availability and a very low number of incidental shut-downs. In international comparison the collective doses of the staff are low and the release of radioactive materials to the environment is negligible; on this account KKG is one of the world's best plants operating pressurised water reactors. Up to now the examinations have not brought any ageing deterioration to light concerning the status of safety-relevant components or ducts

  18. Replacement of the feedwater pipe system in reactor building outside containment at the nuclear power plant Philippsburg; Austausch der Speisewasserleitung im Reaktorgebaeude ausserhalb SHB im Kernkraftwerk Philippsburg I

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, A. [Energie-Versorgung Schwaben AG, Stuttgart (Germany); Labes, M. [Siemens AG Unternehmensbereich KWU, Offenbach am Main (Germany); Schwenk, B. [Kernkraftwerk Philippsburg GmbH (Germany)

    1998-11-01

    After full replacement of the feedwater pipe system during the inspection period in 1997, combined with a modern materials, manufacturing and analysis concept, the entire pipe system of the water/steam cycle in the reactor building of KKP 1 now consists of high-toughness materials. The safety level of the entire plant has been increased by leaving aside postulation of F2 breaks in the reactor building and providing for protection against 0.1 leaks. Based on fluid-dynamic calculations for the cases of pump failure and pipe break, as well as pipe system calculations in 5 extensive calculation cycles, about 130 documents were filed for inspection and approval (excluding preliminary test documents on restraints). Points of main interest for safety analysis in this context were the optimised closing performance of the 3rd check valves and the integrity of the nozzle region at the RPV. (oirg./CB) [Deutsch] Durch den Restaustausch der Speisewasserleitungen in der Revision 1997, verbunden mit einem modernen Werkstoff-, Fertigungs- und Nachweiskonzept, sind im Reaktorgebaeude von KKP 1 in den Hauptleitungen des Wasser-Dampf-Kreislaufes nur noch hochzaehe Werkstoffe eingesetzt. Durch den Verzicht auf das Postulat von 2F-Bruechen im Reaktorgebaeude und durch die Auslegung gegen 0,1F-Lecks wird das Sicherheitsniveau der Anlage insgesamt gesteigert. Ausgehend von fluiddynamischen Berechnungen fuer Pumpenausfall und Rohrbruch sowie Rohrsystem-Berechnungen in 5 umfangreichen Berechnungskreisen wurden fuer die Genehmigung und Begutachtung ca. 130 Unterlagen (ohne Halterungs-Vorpruefunterlagen) eingereicht und vom Gutachter geprueft. Schwerpunkte der Nachweisfuehrung waren die Optimierung des Schliessverhaltens der 3. Rueckschlagarmaturen sowie der Integritaetsnachweis des RDB-Anschlusses. (orig./MM)

  19. Successful implementation of ageing management exemplified at the cooling tower of Emsland nuclear power plant; Erfolgreiche Umsetzung von Alterungsmanagement am Beispiel Kuehlturm des Kernkraftwerkes Emsland

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Alexander [Hochtief Solutions AG, Consult IKS Energy, Frankfurt am Main (Germany). Design Kraftwerke; Dueweling, Carsten [Kernkraftwerke Lippe-Ems GmbH, Lingen (Germany). Abschnitt Bautechnik

    2013-07-15

    The present paper describes the successful implementation of the restoration of water-distribution channels at the cooling tower of the Emsland nuclear power plant under the aspect of ageing management. The main challenge of aging management is the determination of potential aging mechanism and to avoid systematically and effectively their damaging influences. In the course of the annual site inspections abnormalities at the lower side of the water-distribution channels of the cooling tower were detected, analysed, and repaired. The extraordinary high chlorine equivalent of the cooling water was identified as main reason of the damages located. Due to extensive infiltration into the concrete structure, chloride-induced corrosion generates a volume expansion of the reinforcement and thereby to a blast off of the concrete covering. According to the restoration concept, the damaged concrete was removed by maximum pressure water jet blasting; where necessary the reinforcement was retrofitted and a layered concrete substitution was applied by synthetic cement mortar. The realised procedures conserve the load bearing reinforcement only for a certain period, because the permanent chloride infiltration could not be stopped. Therefore, the structure has to be monitored permanently. (orig.)

  20. Nuclear power plants in Germany. Recent developments in off-site nuclear emergency preparedness and response; Kernkraftwerke in Deutschland. Neue Entwicklungen im anlagenexternen Notfallschutz

    Energy Technology Data Exchange (ETDEWEB)

    Gering, Florian [Bundesamt fuer Strahlenschutz, Oberschleissheim/Neuherberg (Germany). Abt. SW 2.2 Entscheidungshilfesysteme, Lageermittlung und Kommunikation

    2014-10-15

    The reactor accident in Fukushima, Japan, in 2011 triggered a thorough review of the off-site emergency preparedness and response for nuclear power plants in Germany. ''Off-site emergency preparedness and response'' includes all actions to protect the public outside the fence of a nuclear power plant. This review resulted in several changes in off-site emergency preparedness and response, which are briefly described in this article. Additionally, several recent activities are described which may influence emergency preparedness and response in the future.

  1. Consequences of changed nuclear power plant lifetimes in Germany. Scenario analyses until 2035; Auswirkungen veraenderter Laufzeiten fuer Kernkraftwerke in Deutschland. Szenarioanalysen bis zum Jahre 2035

    Energy Technology Data Exchange (ETDEWEB)

    Blesl, Markus; Bruchof, David; Fahl, Ulrich; Kober, Tom; Kuder, Ralf; Beestermoeller, Robert; Goetz, Birgit; Voss, Alfred

    2011-06-01

    The report is aimed to discuss the implications of changed NPP lifetimes in Germany on energy policy, environment, energy cost and macroeconomics. An extensive scenario analysis is used considering the effects on the German energy system in the frame of the European context. It is shown that a nuclear phase-out until 2017 is technically feasible, but needs adequate replacement options that will change the German energy system in the medium term. The study shows that the time of nuclear phase-out has no significant influence on the use of renewable energies.

  2. No nuclear power plant - now final repository? What to do with small amounts of waste?; Kein Kernkraftwerk - kein Endlager? Wohin mit wenig Abfaellen?

    Energy Technology Data Exchange (ETDEWEB)

    Feinhals, Joerg [DMT GmbH und Co. KG, Hamburg (Germany)

    2015-07-01

    Countries with nuclear power plants try to find a solution for the disposal of radioactive waste. Countries that have no nuclear power plants but produce radioactive waste in medicine, industry and research and operate research reactors have a problem: the challenging question of an appropriate disposal concept. Possibilities for such a concept are discussed in this contribution, for instance a multinational final repository, near-surface disposal of low- and medium-level radioactive wastes or a small scale disposal facility (SSDF). In any case safety analyses are required.

  3. Substitution of cooling tower components in the nuclear power plant Goesgen-Daeniken AG; Ersatz der Kuehlturmeinbauten im Kernkraftwerk Goesgen-Daeniken AG

    Energy Technology Data Exchange (ETDEWEB)

    Rich, H.W. [Kernkraftwerk Goesgen-Daeniken AG (Switzerland)

    2011-07-01

    At the nuclear power plant Goesgen-Daeniken AG (Daeniken, Switzerland), the cooling tower installations of asbestos cement to have been replaced by plastics. The resulting continuous decrease in the cooling capacity is based on a weakly dimensioned wall thickness of the film installations and on a deposition of suspended matter. The deposition of suspended matter additionally was favoured by biofilms on the film surface. Four measures are presented for the remediation of this problematic situation. With this, the contamination of the film installations are minimized. Deformations of foil packets can be avoided. The cooling capacity of the cooling tower significantly has been improved.

  4. Replacement of the cooling tower packing at the Goesgen-Daeniken AG nuclear power plant; Ersatz der Kuehlturmeinbauten im Kernkraftwerk Goesgen-Daeniken

    Energy Technology Data Exchange (ETDEWEB)

    Rich, Hans Walter [Kernkraftwerk Goesgen-Daeniken AG, Daeniken (Switzerland)

    2012-07-01

    In 2005 the asbestos cement cooling tower packing was replaced by plastic material. Two years later, the packing showed strong deformations, deposits of solids and weight gain. At the end of 2007 parts of the packing collapsed into the cooling tower basin. Investigations were made, revealing that the thickness of the packing foil was too low and that packing geometry and biofilms on the surface of the packing favoured deposition of solids. Successful measures were taken to solve the problems. (orig.)

  5. Expertise on the Goesgen-Daeniken nuclear power plant on the granting of a licence for the construction and operation of a water storage pool for fuel assemblies at the site of the power plant; Gutachten zum Gesuch der Kernkraftwerk Goesgen-Daeniken AG um Erteilung der Bewilligung fuer den Bau und Betrieb eines Brennelement-Nasslagers auf dem Areal des Kernkraftwerks

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-04-15

    On June 26, 2002, the Goesgen-Daeniken AG nuclear power plant (KKG) delivered a request to the Swiss Federal Council for the granting of a licence for the construction and operation of a water storage pool for the on-site storage of the power plant's fuel assemblies. The present report contains the results of the examination of the request by the Federal Agency for the Safety of Nuclear Installations (HSK), to check that the projected storage pool satisfies the legal requirements from the point of view of nuclear safety and protection against radioactivity. A water storage pool already exists in the reactor building of KKG. It was conceived for a fuel cycle based on the reprocessing of the spent fuel assemblies. Its capacity is not sufficient when the spent fuel assemblies are no longer reprocessed but have to be transferred and stored in the Central Intermediate Storage Facility (ZWILAG) in Wuerenlingen because their heat production is too high. The capacity of the actual water pool allows a maximum cooling time of 5-6 years, while 7-10 years are required before transfer to ZWILAG. The projected new water storage pool has to be aircraft crash and earthquake proof, in the same way that the reactor building itself has to be. It can store a maximum of 1008 fuel assemblies. The water in the pool as well as the pool walls shield the radiation from of the fuel assemblies almost completely. Each fuel assembly is put into a square steel channel. The channel walls are lined with 6.11 mg/cm{sup 2} of the neutron absorbing nuclide B-10, which guaranties the subcriticality of the water pool even if the storage pool would be entirely filled with non-irradiated fuel assemblies with the maximal allowed enrichment or the maximal allowed content of Plutonium in case of MOX fuel assemblies, which is a very conservative assumption. The heat released by decay in the spent fuel assemblies is transferred to the pool water. Storage pool cooling is carried out by natural circulation through two cooling towers which release the heat to the environment. The cooling system is designed for a maximum cooling power of 1 MW. With this system the temperature of the pool water does not exceed 80 {sup o}C. When they are retrieved from the reactor core, the fuel assemblies are first transferred to the present water storage pool within the reactor building where they remain for at least two years. During this time, most of the short-life radioactive nuclides decay such that their contribution to the production of heat becomes negligible. In the new storage pool, the total radioactivity at full loading will amount to about 10{sup 19} Bq, i.e. one order of magnitude less than the maximal activity in the present pool. As far as the volatile radio-nuclides are concerned, all noble gases except Kr-85 and all iodine isotopes except I-129 have already decayed; as a consequence, the radiological risk in the new storage pool is much lower than in the old one. As the heating rate in the new pool is more than one order of magnitude lower than that of the present one, a possible failure in the heat release system produces only a slow increase of pool water temperature of less than 1 K per hour with the maximum heating power of 1 MW. In the first phase, it is foreseen to limit the cooling power to 0.5 MW and the number of stored fuel assemblies to 504. As the number of retrieved fuel assemblies from the reactor core is about 40 per year, the first phase will last at least 10 years. After closing of the nuclear power plant at the end of its working time and its dismantling, the storage can still work independently. After examination of the whole project for the new water storage pool, HSK concludes that under some additional conditions the concept presented can be the basis for the safe operation of the pool foreseen

  6. Nuclear power: on line; Kernenergie Online

    Energy Technology Data Exchange (ETDEWEB)

    Thieme, Christian [atw Redaktion, Hattingen (Germany)

    2011-04-15

    Presentation of these contents in the World Wide Web (WWW): Joint Research Centre (JRC), Institute for Transuranium Elements (ITU) - itu.jrc.ec.europa.eu Kernkraftwerk Gundremmingen (Germany) - www.kkw-gundremmingen.de Canadian Nuclear Association (Canada) - www.cna.ca Kernkraftwerk Krsko (Slovenia) - www.nek.si. (orig.)

  7. Effects of thermal effluents from the Unterweser reactor (KKU) on biocenoses in the Unterweser. Pt. 3. Final report. Auswirkungen der Abwasserwaerme des Kernkraftwerkes Unterweser (KKU) auf die Biozoenosen in der Unterweser. T. 3. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Vobach, M.; Feldt, W.

    1991-05-01

    Between August 1975 and November 1982, the influence of thermal pollution in the biocenosis of the Unterweser from cooling water dischanged by the Unterweser reactor was investigated. The state of the parts of the river not yet stressed by cooling water is compared with conditions after the start-up of the reactor (September 1978). The seasonal cycle of water temperature has not changed after the start-up as compared to the time before. A warming of the river water because of cooling water discharged from the reactor is recognizable in the area immediately surrounding the month of the discharge system. Benthal investigations show the composition of species and number of individuals to be unchanged after the start-up of the reactor. Phytoplankton, too, continues to have its population maximum in May and August. Zooplankton, being present abundantly and in clusters, has retained its original composition of species. Now as before the reactor's start-up, flounder, smelt, stickleback, sprat and gudgeon which between them account for 97 per cent of the total catch, continue to be the five major fish species. Variations in the composition of catch are not to be explained by changes of temperature. The slight temperature increase does not modify the spectrum of species; there is no temperature stimulus. The seasonal cycle of water temperatures, which are important for a diapause, i.e. a slowing in the developmental cycle with a reduced metabolism, thus safeguarding the survival of certain species, do continue to occur. Any observed changes are to be interpreted as expressions of longer-term biological cycles. (orig./BBR).

  8. Co-60 calculation methods for the waste form documentation. Example: disposal of reactor pressure vessel internals of the NPP Stade (KKS); Co-60 Berechnungsverfahren fuer die Erstellung einer Abfallgebindedokumentation. Beispiel: Entsorgung der RDB-Einbauten des Kernkraftwerks Stade (KKS)

    Energy Technology Data Exchange (ETDEWEB)

    Kaehle, S. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Tittelbach, S. [WTI Wissenschaftlich-Technische Ingenieurleistung GmbH, Juelich (Germany); Bacmeister, G. [E.ON Kernkraft GmbH, Stade (Germany). Kernkraftwerk Stade

    2010-05-15

    In the frame of the reactor dismantling project of the NPP Stade the reactor pressure vessel and the internals are dissected, packed and dried since 2007. Due to the high neutron flux the reactor internals are highly activated and partially show a strong activation gradient. The authors demonstrate an activity calculation technique (esp. for the leading radionuclide Co-60) based on underwater dose rate measurements of the reactor internals that will allow to optimize the maximum utilization of the interim storage plant at the NPP site and ti reach the dose limits for the final repository Schacht Konrad. The authors show that the use of methods based on heterogeneous source term distributions provides a significantly reduced Co-60 activities.

  9. Expected returns from a tax on nuclear fuel elements in the context of longer service lives of German nuclear power plants; Moegliches Aufkommen einer Brennelementesteuer im Kontext der Laufzeitverlaengerung der Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Kondziella, Hendrik; Bruckner, Thomas [Leipzig Univ. (Germany). Inst. fuer Infrastruktur und Ressourcenmanagement; Bode, Sven [arrhenius Institut fuer Energie- und Klimapolitik, Hamburg (Germany)

    2010-10-15

    To what extent will the fuel element tax introduced by the German government in combination with the longer service life of nuclear power stations reduce the profits of public utilities? A qualitative assessment suggests that the tax will not equal the full profits. Using an electricity market model, various scenarios can be calculated for an eight-year prolongation of the residual service life of existing nuclear power plants. (orig.)

  10. How does react power price on a possible lifetime extension for power plants? Nuclear power, power prices and power market models; Wie reagiert der Strompreis auf eine moegliche Verlaengerung der Laufzeiten fuer Kernkraftwerke? Kernkraft, Strompreis und Strommarktmodelle

    Energy Technology Data Exchange (ETDEWEB)

    Nestle, Uwe [Buendnis 90/Die Gruenen, Berlin (Germany). Bundesarbeitsgemeinschaft Energie

    2010-08-23

    Extending the life of the nuclear power plants currently operated in Germany is being discussed in the light of a more likely change in government for a Christian Democrat/Liberal coalition. The reason cited most frequently is the impossibility to meet the objectives of climate protection without raising further the price of electricity if the life of nuclear power plants cannot be extended. The question to be looked into is that of the legal pre-requisites to be established in Germany in order for the existing nuclear power plants to be operated for longer periods of time. So in this contribution some discussion is done wether a possible lifetime extension of nuclear power plants will react on power prices.(GL)

  11. Transfer of financial obligations for the disposal of nuclear waste and decommissioning of German NPP's. Legal aspects of a trust model; Sicherstellung der finanziellen Entsorgungsvorsorge fuer die Stilllegungs- und Rueckbaukosten der deutschen Kernkraftwerke. Rechtliche Randbedingungen eines Stiftungsmodells

    Energy Technology Data Exchange (ETDEWEB)

    Schewe, Markus; Wiesendahl, Stefan [Kuemmerlein Rechtsanwaelte und Notare, Essen (Germany)

    2015-04-15

    The nuclear power plant operators have to bear the costs associated with the closure and the decommissioning of the German nuclear power plants as well as the costs for the disposal of nuclear waste. For that purpose, the operators have to build up sufficient reserves for the decommissioning phase. These reserves at the end of 2013 amounted to approximately 36 billion Euro. Changing this system is discussed very so often. Last in May 2014, a public debate started dealing with the so called trust model (''Stiftungsmodell''). The press published deliberations of several operators to transfer their entire nuclear business to the Federal Republic of Germany. Under this deliberation the current nuclear power plant operations, as well as closure obligations would be contributed to trust. Further, also the reserves should be ''transferred'' to the trust. RAG-Foundation (RAG-Stiftung) - which will assume the financial obligations in connection with Germany's closure of underground coal mining activities - sometimes is cited as a role model. The article covers elements of German trust law and atomic energy law regarding such deliberations. In trust law e.g. it can be debated whether the trust should be established under public or - as in the case of RAG-Foundation - under private law. In this context we will set out the major differences between those two options. In the public law part we will notably address issues arising from individual licensing requirements for nuclear power plants and focus on questions concerning reliability, requisite qualification and organizational structures.

  12. Occupational safety in the nuclear power plant. The contribution of sociology to the development of a communication tool for the elimination of hazardous situations; Arbeitssicherheit im Kernkraftwerk. Der Beitrag der Sozialpsychologie zur Entwicklung eines Kommunikationsinstrumentes fuer die Behebung von Gefaehrdungssituationen

    Energy Technology Data Exchange (ETDEWEB)

    Zedler, Christien [IAOP - Institut fuer Arbeitspsychologie, Organisation und Prozessgestaltung, Berlin (Germany); Huber, Veit [E.ON Kernkraft GmbH (Germany)

    2012-11-01

    Nuclear power plant companies make efforts to enhance the operational safety in the plant. Despite a variety of measures the number of accidents at work is still too high, esp. for external personnel. Social psychological considerations were used to develop communication tools for the elimination of hazardous situations, for instance by safety dialogues between employees. The observation of hazardous situations should trigger communication and discussion on the risk of the specific situation. In the contribution practical experiences and recommendations for the realization of a safety dialogue culture in the NPP Grafenrheinfeld are summarized and illustrated by examples.

  13. Reserves for nuclear power plant decommissioning and radwaste disposal in Germany. An analysis and evaluation from the angle of energy policy; Energiewirtschaftliche Bewertung der Rueckstellungen fuer die Entsorgung der deutschen Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Buerger, V. [comp.

    1998-12-31

    The study, which is the first of its kind in Germany, presents a comprehensive survey of total reserves set up by the German nuclear industry for liabilities and costs for nuclear power plant decommissioning and resulting radwaste disposal, which is a legal and foreseeable responsibility but uncertain in amount. The study looks into the various ways the earmarked money was invested and analyses the funds with respect to their efficiency and reliability to provide financial security for the given tasks and purpose. The question put in this context is: Are the reserves set up so far in line with official cost estimates, i.e. will they cover estimated costs, or do they even exceed the estimated amounts? The conclusions drawn and explained in this document are: The reserves for nuclear decommissioning have been used by the nuclear power plant operators and electricity companies as a significant capital source. Some of the capital accrued is being increasingly used at present to cover expenses arising for restructuring of business and diversification into new business segments of interest in the open national and European electricity markets. Companies such as RWE, Preussen Elektra, and Bayernwerk, which until deregulation of the energy sector were just power supply companies, have been transformed into conglomerate companies and international players in the markets, like RWE Holding, VEBA, and VIAG. It can be safely assumed that the companies would not have been able to reach the important positions they currently hold in the German economy without tapping the reserves for nuclear decommissioning. (orig./CB) 27 refs. [Deutsch] Die Studie gibt erstmals einen vollstaendigen Ueberblick ueber die Summe der in Deutschland gebildeten Rueckstellungen im Kernenergiebereich. Sie geht der Frage nach, wie diese Gelder angelegt sind und ob die praktizierten Anlageformen dem hohen Sicherheitsanspruch entsprechen, den die Gesellschaft an die finanziellen Ressourcen zur Bewaeltigung eines derart gravierenden Problems stellen muss. Weiter wird untersucht, ob die absolute Hoehe der gebildeten Rueckstellungen mit den offiziell diskutierten Kostenschaetzungen in Einklang steht; ob also ausreichend Gelder zurueckgestellt werden oder ob die Rueckstellungen gar zu hoch sind. Ein weiterer wichtiger Aspekt wird hierbei deutlich: Die Rueckstellungen im Kernenergiebereich sind von ihrer absoluten Hoehe her nicht mit Rueckstellungen zu anderen Zwecken z.B. fuer Pensionszahlungen, zu vergleichen. Sie haben fuer die rueckstellungsbildenden Unternehmen eine enorme Bedeutung als Kapitalquelle entwickelt. Dieses Kapital spielte und spielt immer noch eine erhebliche Rolle bei der Umwandlung von frueher vorwiegend auf den Stromsektor orientierten Unternehmen wie RWE, PreussenElektra und Bayernwerk zu modernen, international agierenden Mischkonzernen wie der RWE Holding, VEBA und VIAG. Es darf angenommen werden, dass die herausragende Rolle dieser Konzerne in der deutschen Volkswirtschaft ohne dieses Finanzpolster nicht in gleichem Umfang haette entwickelt werden koennen. (orig.)

  14. Reserves for nuclear power plant decommissioning and radwaste disposal in Germany. An analysis and evaluation from the angle of energy policy. Energiewirtschaftliche Bewertung der Rueckstellungen fuer die Entsorgung der deutschen Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Buerger, V. (comp.)

    1998-01-01

    The study, which is the first of its kind in Germany, presents a comprehensive survey of total reserves set up by the German nuclear industry for liabilities and costs for nuclear power plant decommissioning and resulting radwaste disposal, which is a legal and foreseeable responsibility but uncertain in amount. The study looks into the various ways the earmarked money was invested and analyses the funds with respect to their efficiency and reliability to provide financial security for the given tasks and purpose. The question put in this context is: Are the reserves set up so far in line with official cost estimates, i.e. will they cover estimated costs, or do they even exceed the estimated amounts The conclusions drawn and explained in this document are: The reserves for nuclear decommissioning have been used by the nuclear power plant operators and electricity companies as a significant capital source. Some of the capital accrued is being increasingly used at present to cover expenses arising for restructuring of business and diversification into new business segments of interest in the open national and European electricity markets. Companies such as RWE, Preussen Elektra, and Bayernwerk, which until deregulation of the energy sector were just power supply companies, have been transformed into conglomerate companies and international players in the markets, like RWE Holding, VEBA, and VIAG. It can be safely assumed that the companies would not have been able to reach the important positions they currently hold in the German economy without tapping the reserves for nuclear decommissioning. (orig./CB) 27 refs.

  15. An investigation of the applicability of the new ion exchange resin, Reillex{trademark}-HPQ, in ATW separations. Milestone 4, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ashley, K.R.; Ball, J.; Grissom, M.; Williamson, M.; Cobb, S.; Young, D.; Wu, Yen-Yuan J.

    1993-09-07

    The investigations with the anion exchange resin Reillex{trademark}-HPQ is continuing along several different paths. The topics of current investigations that are reported here are: The sorption behavior of chromium(VI) on Reillex{trademark}-HPQ from nitric acid solutions and from sodium hydroxide/sodium nitrate solutions; sorption behavior of F{sup {minus}} on Reillex{trademark}-HPQ resin in acidic sodium nitrate solution; sorption behavior of Cl{sup {minus}} on Reillex{trademark}-HPQ resin in acidic sodium nitrate solution; sorption behavior of Br{sup {minus}} on Reillex{trademark}-HPQ resin in acidic sodium nitrate solution; and the Honors thesis by one of the students is attached as Appendix II (on ion exchange properties of a new macroperous resin using bromide as the model ion in aqueous nitrate solutions).

  16. Surface engineering of biaxial Gd2Zr2O7 thin films deposited on Ni–5at%W substrates by a chemical solution method

    DEFF Research Database (Denmark)

    Yue, Zhao; Grivel, Jean-Claude; Liu, Min;

    2012-01-01

    The surface texture and morphology of thin films play an essential role in determining their properties. In this study, local features in the film surface of crystallized Gd2Zr2O7 (GZO) films with a thickness gradient are investigated by means of scanning electron microscopy and electron backscat......The surface texture and morphology of thin films play an essential role in determining their properties. In this study, local features in the film surface of crystallized Gd2Zr2O7 (GZO) films with a thickness gradient are investigated by means of scanning electron microscopy and electron...... ordered nanoislands or network) to 3-dimensional domes (equiaxed nanograins), and (ii) the segregation of residual carbon in the surface layer. The epitaxial nuclei forming at the interface hardly further develop by consuming the polycrystalline grains in the surface layer. A two-step annealing procedure...

  17. Development of One Meter Long Double-Sided CeO2 Buffered Ni-5at.%W Templates by Reel-to-Reel Chemical Solution Deposition Route

    DEFF Research Database (Denmark)

    Yue, Zhao; Konstantopoulou, K.; Wulff, Anders Christian;

    2013-01-01

    High performance long-length coated conductors fabricated using various techniques have attracted a lot of interest recently. In this work, a reel-to-reel design for depositing double-sided coatings on long-length flexible metallic tapes via a chemical solution method is proposed and realized...... layer are 7.2◦ and 5.8◦ with standard deviation of 0.26◦ and 0.34◦, respectively, being indicative of the high quality epitaxial growth of the films prepared in the continuous manner. An all chemical solution derived YBCOLow−TFA/Ce0.9La0.1O2/Gd2Zr2O7/CeO2 structure is obtained on a short sample...

  18. Judges as civilization censors

    International Nuclear Information System (INIS)

    Negative criticism on the judgment of the administrative court at Freiburg concerning PV burst protection for the Kernkraftwerk Whyl, and the effects of the judgment on the development of nuclear energy in the FRG. (HP)

  19. Abstracts of papers from the literature on anticipated transients without scram for light water reactors 1. 1975-1979

    International Nuclear Information System (INIS)

    INIS ATOMINDEX abstracts relating to ATWS for light water reactors for the years 1975-1979 are presented under the subject headings of; general, licensing and standards, models and computer codes, frequency of occurrence of ATWS, transient calculations of results including probabilistic analysis, radiological consequences of ATWS, fuel behaviour, and studies of plant components. (U.K.)

  20. 非能动核电厂支持事件树分析的ATWS慢化剂反馈分析%Analysis of Moderator Reactivity for ATWS to Support PSA Success Criteria in Passive Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    徐珍; 梁锋; 徐军

    2013-01-01

    在非能动核电厂的ATWS事故中,可能由于反应堆冷却剂系统超压而导致系统损坏.本文使用系统分析程序对AP1000核电厂各种系统工况下的慢化剂温度系数进行研究分析,确定了事故过程中反应堆冷却剂系统(RCS)不超压的极限慢化剂温度系数.该分析结果为概率安全分析中的ATWS事件树分析提供了必要的支持.

  1. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. Annual report 1997; Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. Jahresbericht 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-10-01

    There were 117 notifiable events reported from nuclear power plants in Germany, and 12 reported from research reactors. These events have been anlysed for the annual report 1997 under a variety of aspects. The results do not indicate any systematics of occurrence. None of the reported events resulted in any release of radioactivity exceeding the regulatory limits, so that there were no off-site risks involved. Among the reported events, there were three belonging to category E (prompt notification), the other 114, or 12, respectively, were at lowest scale, N, and there were none belonging to scale S. (orig./GL) [Deutsch] Im Jahr 1997 wurden aus den Kernkraftwerken der Bundesrepublik Deutschland urspruenglich insgesamt 117 und aus den Forschungsreaktoren 12 Ereignisse gemeldet. Fuer den Jahresbericht wurden diese Ereignisse nach verschiedenen Gesichtspunkten analysiert. Systematische Schwachstellen wurden dabei nicht festgestellt. Bei keinem der gemeldeten Ereignisse traten Abgaben radioaktiver Stoffe oberhalb genehmigter Grenzwerte auf. Eine Gefaehrdung von Personen und der Umgebung war in keinem Fall gegeben. Im Berichtsjahr wurden drei Ereignisse in der Kategorie E (Eilmeldung) gemeldet. Die anderen 114 bzw. 12 Ereignisse lagen in der niedrigsten Meldekategorie N (Normalmeldung). Ereignisse der Kategorie S (Sofortmeldung) traten nicht auf. (orig./GL)

  2. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. Third quarterly report 1997; Meldepflichtige Ereignisse zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. Vierteljahresbericht 3. Quartal 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    The report presents the survey and the scenarios of events reported from nuclear power plant and research reactors with a rated thermal output above 50 kW, covering the 3rd quarter of 1997. (orig./CB) [Deutsch] Der vorliegende Bericht enthaelt die Uebersicht ueber die meldepflichtigen Ereignisse in Kernkraftwerken und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet, fuer das dritte Vierteljahr 1997.

  3. Grohnde. Documentation of the police operation during the demonstration against the NPP Grohnde on 19.03.1977 and the evacuation of the occupied cooling tower site on 23.08.1977; Grohnde. Dokumentation der Polizeieinsaetze anlaesslich der Demonstration gegen das Kernkraftwerk Grohnde am 19.03.1977 und der Raeumung des besetzten Kuehlturmgelaendes am 23.08.1977

    Energy Technology Data Exchange (ETDEWEB)

    Stricker, Michael

    2014-07-01

    The documentation of the police operation during the demonstration against the NPP Grohnde on 16.03.1977 and the evacuation of the occupied cooling tower site on 23.08.1977 covers the following issues: involved action forces: police Niedersachsen, police Nordrhein-Westfalen, police Schleswig-Holstein, police Bremen and the Bundesgrenzschutz; concept of the police operation, provisions (lodging and board) for the police, operating resources, details of the operation sequence; post-processing of the operation; the Grohnde trials.

  4. Comments on conceptual questions concerning the clearance of wastes for disposal on a dump site during the decommissioning and dismantling of the nuclear power plant Obrigheim (KWO); Stellungnahme zu konzeptionellen Fragen der Freigabe zur Beseitigung auf einer Deponie bei Stilllegung und Abbau des Kernkraftwerks Obrigheim (KWO)

    Energy Technology Data Exchange (ETDEWEB)

    Kueppers, Christian

    2015-08-03

    The comments on conceptual questions concerning the clearance of wastes for disposal on a dump site during the decommissioning and dismantling of the nuclear power plant Obrigheim (KWO) cover the following issues: fundamentals of the 10 micro-Sv concept for clearance; specific regulations for the clearance of wastes from the dismantling of KWO for disposal on a dump site; disposal concept at shutdown and dismantling of KWO; measurements and control during clearance for disposal during shutdown and dismantling of KWO; documentation and reports.

  5. Investigation of incident scenarios for BWR-type plant, taking into consideration the effects of the plant emergency protection systems. The Kruemmel reactor station (KKK) as an example. [accident management]. Untersuchungen zu Ereignisablaeufen fuer SWR-Anlagen unter Einbeziehung von Massnahmen des anlageninternen Notfallschutzes am Beispiel des Kernkraftwerkes Kruemmel (KKK)

    Energy Technology Data Exchange (ETDEWEB)

    Erven, U.; Nolte, W.; Linden, J. v.; Weidlich, H.

    1988-01-01

    The hypothetical scenarios discussed are the following: (1) Failure of the main heat sink with simultaneous breakdown of after-heat removal from the condensation chamber. (2) Failure of the entire, not battery-supported a.c. supply system (station blackout). (3) Loss of coolant in the containment, with failure of the sump pumping system. The study in addition to the initial failure assumes failure of further systems (condensation chamber cooling, non-battery-supported standby power supply, containment sump pumping system). In all three cases, further emergency provisions are possible in order to prevent dry-out of the reactor core, or undue pressure buildup in the containment. These provisions are easily feasible, as analyses have shown that there is sufficient time for carrying out the planned emergency protection measures or the plant protection measures, namely: 27 hours in case of 'failure of main heat sink', at least 5 hours in case of 'station blackout', and at least 4 days in case of 'loss of coolant in the containment'. The frequencies of occurrence of inadmissible plant conditions are low that there is consensus on the national and international level that such cases are not relevant for risk assessment, and therefore cannot be taken into consideration for technical protection measures development.

  6. Decommissioning and disposal of nuclear core parts; Abbau und Entsorgung von Kernbauteilen. Strahlenschutzmassnahmen am Beispiel der Stillegungsprojekte Gundremmingen (KRB A) und Kahl (VAK)

    Energy Technology Data Exchange (ETDEWEB)

    Duempelmann, W.; Steiner, H. [Kernkraftwerk Betriebsgesellschaft mbH, Gundremmingen (Germany); Eickelpasch, N.; Hackel, W. [Versuchsatomkraftwerk Kahl GmbH (VAK), Kahl (Germany)

    1997-12-31

    The authors describe the operational procedures, the measures for radiation protection, and the experience gained in decommissioning the shut-down nuclear power plants of Gundremmingen (KRB A) and Kahl (VAK). (orig.) [Deutsch] Die Autoren beschreiben das praktische Vorgehen, die Strahlenschutzmassnahmen und die Erfahrungen beim Abbau der stillgelegten Kernkraftwerke Gundremmingen (KRB A) und Kahl (VAK). (orig.)

  7. 46th Annual meeting on nuclear technology (AMNT 2015). Key Topic / Enhanced safety and operation excellence / Radiation protection

    International Nuclear Information System (INIS)

    Summary report on the Focus Session 'Radiation Protection' of the 46th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 and 8 (2015) and will be covered in further issues of atw.

  8. Accelerator-driven destruction of long-lived radioactive waste and energy production

    International Nuclear Information System (INIS)

    Nuclear waste management involves many issues. ATW is an option that can assist a repository by enhancing its capability and thereby assist nuclear waste management. Technology advances and the recent release of liquid metal coolant information from Russia has had an enormous impact on the viability of an ATW system. It now appears economic with many repository enhancing attributes. In time, an ATW option added to present repository activities will provide the public with a nuclear fuel cycle that is acceptable from economic and environmental points of view

  9. Notifiable events in facilities for fission of nuclear fuels in the Federal Republic of Germany. Nuclear power plants and research reactors with a maximum continuous thermal output of more than 50 kW. First quarterly report 1998; Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen in der Bundesrepublik Deutschland. Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet. Vierteljahresbericht 1. Quartal 1998

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

    In the first quarter of 1998, 22 notifiable events in the nuclear power plants installed in Germany were reported and are listed in the survey. There was no event involving discharge of radioactivity above the licensed limits, so that there was no radiological hazard to the population or the environment. All events reported belong to the lowest category N of the Nuclear Event Scale. 21 events were assigned to category 0 of the INES system, (of no or only slight relevance to safety, no radiological significance), and one to INES category 1 (operating incident, no radiological significance). (orig./CB) [Deutsch] Im I. Quartal 1998 wurden 22 meldepflichtige Ereignisse aus den Kernkraftwerken der Bundesrepublik Deutschland erfasst. Die Uebersichtsliste enthaelt alle 22 Ereignisse, die in diesem Zeitraum gemeldet wurden. Abgaben radioaktiver Stoffe oberhalb genehmigter Grenzwerte traten in diesem Zeitraum nicht auf. Eine Gefaehrdung von Personen und der Umgebung war in keinem Fall gegeben. Alle meldepflichtigen Ereignisse lagen in der niedrigsten behoerdlichen Meldekategorie N (Normalmeldung). Ereignisse der behoerdlichen Meldekategorie E (Eilmeldung) und der Kategorie S (Sofortmeldung) waren nicht zu verzeichnen. 21 meldepflichtigen Ereignisse wurden der INES-Stufe 0 (keine oder sehr geringe sicherheitstechnische, bzw. keine radiologische Bedeutung) zugeordnet. Ein Ereignis wurde der INES-Stufe 1 (betriebliche Stoerung, keine radiologische Bedeutung) zugeordnet. (orig.)

  10. Studies on the deterministic and probabilistic assessment of external effects. Deterministic investigation of the robustness of German nuclear power plants against external effects under consideration of actual findings on the events to be assumed; Untersuchungen zur deterministischen und probabilistischen Bewertung von Einwirkungen von aussen (EVA-Ereignisse). Deterministische Untersuchung der Widerstandsfaehigkeit deutscher Kernkraftwerke gegen Einwirkungen von aussen, unter Beruecksichtigung aktueller Erkenntnisse hinsichtlich der anzusetzenden Einwirkungen

    Energy Technology Data Exchange (ETDEWEB)

    Sperbeck, Silvio; Strack, Christian; Thuma, Gernot

    2013-11-15

    The aim of the analyses on natural hazards described in this report was to evaluate the advantages of innovative hazard assessment methods available today over the hazard assessment methods commonly applied for German nuclear power plant sites in the past. For each hazard under consideration (earthquake, flooding, and wind loads) it has been assessed whether the new methods provide additional insights that could call for their mandatory application in future site specific hazard assessments. If no additional insights are gained, the hitherto applied methods can be considered adequate according to today's standards. In the context of this work, no areas could be identified where the hazard assessment methods stipulated in German (nuclear) regulations are generally inadequate. These methods that are commonly applied in practice do not seem to be prone to significantly underestimate the site specific hazard. Nevertheless, some newer methods allow for more precise (reduction of uncertainties) and more comprehensive (consideration of additional hazard characteristics) hazard assessments. Therefore, depending on the hazard under consideration, it could be advisable to supplement future site specific hazard assessments by some additional analyses. As the methods for some of these additional analyses are not yet fully developed, further research will be necessary to enable these amendments.

  11. Fotografie und atomare Katastrophe

    OpenAIRE

    Bürkner, Daniel

    2015-01-01

    Die Dissertation setzt sich mit den fotografischen Repräsentationen der Atombombenabwürfe auf Hiroshima und Nagasaki sowie der Havarie des Kernkraftwerks Tschernobyl auseinander. Dabei werden künstlerische, dokumentarische und touristische Bilder analysiert, die sich der jeweiligen Strahlenkatastrophe oftmals erst Jahre nach dem Ereignis annehmen und ikonografische oder medial-materielle Bezüge zu ihr aufweisen. Es zeigen sich zentrale Strategien, atomare Katastrophen, seien sie militäri...

  12. BWR core thermal-hydraulic stability in anticipated transients without SCRAM

    International Nuclear Information System (INIS)

    The potential exists for core thermal hydraulic density wave oscillations to occur in the boiling water reactor (BWR) during anticipated transients without scram (ATWS) events. Conditions which may lead to oscillations in ATWS result either from recirculation pump trip (RPT) from appropriate ATWS signals, or from the failure to scram during core oscillations. A program to evaluate these scenarios will assess the adequacy of the system design and emergency operating procedures relative to the reactor system performance. Evaluation of ATWS events with oscillations is extremely complex, and defining the bases for the analyses is critical to the success of the program. Accurate analysis requires detailed modeling of the sequence of events, the plant configuration, and initial conditions

  13. 46{sup th} annual meeting on nuclear technology (AMNT 2015). Key topics / Outstanding know-how and sustainable innovations enhanced safety and operation excellence

    Energy Technology Data Exchange (ETDEWEB)

    Lamm, Matthias [AREVA GmbH, Erlangen (Germany). R and D; Hollands, Thorsten [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Reactor Safety Research Div.

    2016-01-15

    Summary report on the Technical Sessions ''Know-how, New Build and Innovations'' and ''Operation and Safety of Nuclear Installations, Fuel SA: WASA-BOSS + CESAM'' of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 12 (2015) and will be covered in further issues of atw.

  14. 46{sup th} Annual meeting on nuclear technology (AMNT 2015). Key topic / Decommissioning experience and waste management solutions

    Energy Technology Data Exchange (ETDEWEB)

    Buettner, Klaus [NUKEM Technologies Engineering Systems GmbH, Alzenau (Germany). Process Engineering; Klute, Stefan [BKW Energie AG, Bern (Switzerland). Decommissioning Project KKM

    2015-12-15

    Summary report on the Topical Sessions ''Radioactive Waste Management, Storage and Disposal'' and ''Decommissioning of Nuclear Facilities - Challenges and Solutions'' of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 11 (2015) and will be covered in further issues of atw.

  15. 46th Annual meeting on nuclear technology (AMNT 2015). Key topics / Outstanding know-how and sustainable innovations / Enhanced safety and operation excellence

    International Nuclear Information System (INIS)

    Summary report on the Topical Sessions ''CFD Simulations for Reactor Safety Relevant Objectives '' and ''Fuel Management During the Last Cycles and Beyond'' of the 46th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 10 (2015) and will be covered in further issues of atw.

  16. 46{sup th} Annual meeting on nuclear technology (AMNT 2015). Key topics / Outstanding know-how and sustainable innovations / Enhanced safety and operation excellence

    Energy Technology Data Exchange (ETDEWEB)

    Schaffrath, Andreas; Scheuerer, Martina [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany); Oenneby Carina; Benjaminsson, Ulf [Westinghouse Electric Sweden AB, Vaesteraes (Sweden). Europe, Middle East and Africa (EMEA)

    2015-11-15

    Summary report on the Topical Sessions ''CFD Simulations for Reactor Safety Relevant Objectives '' and ''Fuel Management During the Last Cycles and Beyond'' of the 46th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 10 (2015) and will be covered in further issues of atw.

  17. 46{sup th} annual meeting on nuclear technology (AMNT 2015). Key topic / Enhanced safety and operation excellence

    Energy Technology Data Exchange (ETDEWEB)

    Traichel, Anke [NUKEM Technologies, Alzenau (Germany). Department of Safety Engineering and Assessment/Proposals Engineering

    2016-02-15

    Summary report on the Technical Session ''Operation and Safety of Nuclear Installations, Fuel - Special Issues'' of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 12 (2015), 1 (2016) and will be covered in further issues of atw.

  18. 46th annual meeting on nuclear technology (AMNT 2015). Key topic / Enhanced safety and operation excellence

    International Nuclear Information System (INIS)

    Summary report on the Technical Session ''Operation and Safety of Nuclear Installations, Fuel - Special Issues'' of the 46th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 12 (2015), 1 (2016) and will be covered in further issues of atw.

  19. 46th annual meeting on nuclear technology (AMNT 2015). Key topics / Outstanding know-how and sustainable innovations enhanced safety and operation excellence

    International Nuclear Information System (INIS)

    Summary report on the Technical Sessions ''Know-how, New Build and Innovations'' and ''Operation and Safety of Nuclear Installations, Fuel SA: WASA-BOSS + CESAM'' of the 46th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 12 (2015) and will be covered in further issues of atw.

  20. 46th Annual meeting on nuclear technology (AMNT 2015). Key topic / Decommissioning experience and waste management solutions

    International Nuclear Information System (INIS)

    Summary report on the Topical Sessions ''Radioactive Waste Management, Storage and Disposal'' and ''Decommissioning of Nuclear Facilities - Challenges and Solutions'' of the 46th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015. Other Sessions of AMNT 2015 have been covered in atw 7 to 11 (2015) and will be covered in further issues of atw.

  1. Interdecadal change of atmospheric stationary waves and North China drought

    Institute of Scientific and Technical Information of China (English)

    Dai Xin-Gang; Fu Cong-Bin; Wang Ping

    2005-01-01

    The inderdecadal change of atmospheric stationary waves (ATW) has been investigated for the two periods 1956-77 and 1978-99. The trough of ATW in the middle and low layer of the troposphere over the Asian continent has experienced a significant weakening during the past two decades, which exerts a great influence on the North China climate. The ATW in 200 hPa has also exhibited some changes since 1977, as a stationary ridge appeared over the northwestern China while a stationary trough appeared above North China. This leads to an increasing of the upward motion above northwestern China and a decreasing above North China. A west-east section of the stationary waves at 40°N shows that the ATW above North China tilted westward for the period 1956-77, but was almost upright during 1978-99. The composite analysis confirms that the climate mean ATW pattern after 1977 is similar to the dry pattern for North China, while the rainy pattern is similar to that before 1977. In consequence, the North China drought is partly due to the interdecadal change of the ATW over boreal Asia in the recent two decades.

  2. BWR core thermal-hydraulic stability in anticipated transients without SCRAM

    International Nuclear Information System (INIS)

    The potential exists for core thermal hydraulic density wave oscillations to occur in the boiling water reactor (BWR) during anticipated transients without scram (ATWS) events. Conditions which may lead to oscillations in ATWS result either from recirculation pump trip (RPT) from appropriate ATWS signals, or from the failure to scram during core oscillations. A program to evaluate these scenarios will assess the adequacy of the system design and emergency operating procedures relative to the reactor system performance. Evaluation of ATWS events with oscillations is extremely complex, and defining the bases for the analyses is critical to the success of the program. Accurate analysis requires detailed modeling of the sequence of events, the plant configuration, and initial conditions. In conclusion: The likelihood and magnitude of any core thermal-hydraulic oscillations for a given ATWS event sequence depend not only on the choice of initial reactor conditions and fuel characteristics, but also on system dynamics and operator actions. Because of the complexity of the phenomena, accurate analysis requires detailed modeling of all relevant systems and interactions. In addition, selection of parameters must be consistent both with the intended application of the results and across the full spectrum of input conditions. Analysis of ATWS/stability events requires a reasonably limiting but representative set of plant conditions and accurate simulation of operator actions. (authors)

  3. Oceanographic and surface meteorological data collected from station ATW20 by University of Wisconsin-Milwaukee and assembled by Great Lakes Observing System (GLOS) in the Great Lakes region from 2014-07-01 to 2016-06-30 (NODC Accession 0123639)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — NCEI Accession 0123639 contains oceanographic and surface meteorological data in netCDF formatted files, which follow the Climate and Forecast metadata convention...

  4. Irradiated fuel behavior during reactivity initiated accidents in LWR's: Status of research and development studies in France

    International Nuclear Information System (INIS)

    There is much interest in the nuclear industry concerning the ability of training simulators to adequately model severe accident conditions, specifically Anticipated Transient Without Scram (ATWS) events. The Pennsylvania Power and Light Co. has recently installed a new simulator which was provided by S3 Technologies. As part of the licensed operator training program, PP ampersand L provides training on Emergency Operating Procedures (EOPs). Since the ATWS event is challenging from both a computational and operational point of view, the Engineering Department was asked to benchmark the new simulator performance. The purpose of this benchmark was to ensure simulator fidelity with EOP basis calculations which are numerically more rigorous. Once acceptable simulator fidelity had been demonstrated, EOPs were evaluated to ensure they could be implemented by the operators. This paper examines the details of the new simulator response for ATWS events, and exposes the PP ampersand L ATWS procedures to further examination. The simulator benchmark was carried out using the PP ampersand L-developed SABRE code which has been benchmarked against plant data and industry accepted codes. For many ATWS scenarios, the new simulator, which is based upon first principles, provides preditions consistent with SABRE. Reactor power levels, consistent with SABRE results, are significantly higher than predicted by the old simulator, and containment pressurization occurs much more rapidly than previously simulated. Additionally, the new simulated reactor water level, pressure and power are far more responsive to perturbations than predicted by the old simulator. This responsiveness is consistent with SABRE predictions and has helped to define modifications to the ATWS emergency operating procedures. The modified procedures enhance the operators ability to respond to ATWS given the much more realistic reactor model

  5. ATWA Frequency for the Analog I and C System of the OPR-1000 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Seungcheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    An anticipated transient without scram (ATWS) is an anticipated operational occurrence (AOO) that results in a rapid pressure rise of the primary side by no reactor trip. The magnitude and timing of the reactor coolant system (RCS) pressure rise depends on the moderator temperature coefficient (MTC), the pressure relief capacity and the energy removal capacity of the secondary side in the pressurized water reactor (PWR). It is dealt with an important safety issue in the point that the primary pressure over ASME stress C level (3,200psig) can lead to core damage consequently. ATWS risk is simply defined as the multiplication of the ATWS frequency and unfavorable exposure time (UET). This paper focuses the estimation of an ATWS frequency for the OPR-1000 reactor with an analog reactor protection system (RPS). It is an important issue in risk-informed technical specification (RITS) of RPS. The plant-specific ATWS frequency model for the OPR-1000 reactor was developed using more realistic information and the state-of-art technology. The results of the work can be directly used to improve risk-informed surveillance test interval (RI-STI) of the KSNP safety-related I and C systems such as RPS.

  6. Relationship between T-wave amplitude and oxygen pulse in guinea pigs in hyperbaric helium and hydrogen.

    Science.gov (United States)

    Kayar, S R; Parker, E C; Aukhert, E O

    1998-09-01

    Diving is known to induce a change in the amplitude of the T wave (ATw) of electrocardiograms, but it is unknown whether this is linked to a change in cardiovascular performance. We analyzed ATw in guinea pigs at 10-60 atm and 25-36 degreesC, breathing 2% O2 in either helium (heliox; n = 10) or hydrogen (hydrox; n = 9) for 1 h at each pressure. Core temperature and electrocardiograms were detected by using implanted radiotelemeters. O2 consumption rate was measured by using gas chromatography. In a previous study (S. R. Kayar and E. C. Parker. J. Appl. Physiol. 82: 988-997, 1997), we analyzed the O2 pulse, i.e., the O2 consumption rate per heart beat, in the same animals. By multivariate regression analysis, we identified variables that were significant to O2 pulse: body surface area, chamber temperature, core temperature, and pressure. In this study, inclusion of ATw made a significantly better model with fewer variables. After normalizing for chamber temperature and pressure, the O2 pulse increased with increasing ATw in heliox (P = 0.001) but with decreasing ATw in hydrox (P pulse for animals breathing heliox vs. hydrox.

  7. Nitrogen sources and sinks in a wastewater impacted saline aquifer beneath the Florida Keys, USA

    Science.gov (United States)

    Dillon, Kevin S.; Chanton, Jeffrey P.; Smith, Leslie K.

    2007-06-01

    Groundwater wells surrounding a high volume advance treatment wastewater (ATW) disposal well in the Florida Keys were monitored for nitrate, nitrite, and ammonium concentrations over a 14 month period. Nutrient concentrations in the shallow subsurface (9 m) show a bimodal distribution between the low salinity wastewater plume and the ambient brackish to saline groundwaters. High NO 3- concentrations are found within the ATW plume while the highest NH 4+ concentrations are found in shallow wells outside of the plume. Evidence suggests that the overlying mud layer unique to this study site contributes the bulk of the NH 4+ observed in these wells. NO 3- concentrations at 9 m wells varied by a factor of four in response to concurrent variations in ATW NO 3- loads over the coarse of the study. Estimated NO 3- uptake rates varied from 32 ± 29 to 98 ± 69 and did not directly correlate with ATW NO 3- loading as we hypothesized. We estimate that 70 ± 34% of the NO 3- from the treatment plant is removed from solution in the subsurface of the study site. Considerable decreases in NO 3- concentration and enrichment of 15NO 3- was observed in many wells, indicating significant denitrification or anaerobic ammonium oxidation is occurring in the subsurface. Dissolved inorganic nitrogen concentrations, distributions, and 15N compositions indicate that denitrification is likely the dominant mechanism for N removal in the ATW plume at Key Colony Beach, Florida.

  8. Some basic insights into nuclear power plant decommissioning; Einige grundsaetzliche Erkenntnisse fuer die Stillegung von Kernkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    Eickelpasch, N. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH (Germany); Steiner, H. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH (Germany)

    1996-06-01

    There are 14 projects running in Germany for decommissioning of a nuclear power plant, 11 of them are performed under the responsibility of the state, and 3 are projects of industrial enterprises. The two most advanced projects are that for shutdown of unit A of the KRB Gundremmingen station, and the VAK project at Kahl (VAK experimental reactor station). Both plants are operated as subsidiaries, of the utilities RWE and Bayernwerk. The conference paper gives some basic insights obtained in the course of these two projects, covering a period of several years. The results are: The two different disposal strategies allowed by the law, i.e. ``materials recycling`` and ``ultimate disposal``, should be assessed and analysed by two separate studies. Quantities and qualities of the liquid wastes to be managed after final shutdown of a plant differ from those of the preceding phase and require specific waste management planning. It is recommended to perform a radiologic analysis of the task of decontamination of the primary loop prior to dismantling work, as shown by the activities for VAK decommissioning. (orig.) [Deutsch] In Deutschland gibt es 14 stillgelegte Kernkraftwerke, 11 davon sind staatliche Projekte, 3 kommerzielle. Die beiden am weitesten fortgeschrittenen Projekte sind der Block A des Kernkraftwerkes Grundremmingen (KRB) und das Versuchsatomkraftwerk Kahl (VAK) - beides Tochtergesellschaften des RWE und Bayernwerks. Aus der Vielzahl der Erfahrungen aus dem langjaehrigen Abbau dieser Kraftwerke sollen einige wenige grundsaetzliche Erkenntnisse aufgezeigt werden. Dies sind im einzelnen - eine insbesondere wirtschaftliche Bewertung der beiden vom Gesetz her gleichwertigen Materialwege `Wiederverwertung` und `Endlager`, - die Tatsache, dass sich nach der endgueltigen Stillegung eines Kernkraftwerkes die Menge und Qualitaet der fluessigen Abfaelle wesentlich veraendert und besondere Massnahmen erfordert, - eine strahlenschutzmaessige Bewertung der Primaerkreis

  9. Separations technology development to support accelerator-driven transmutation concepts

    International Nuclear Information System (INIS)

    This is the final report of a one-year Laboratory-Directed Research and Development (LDRD) Project at the Los Alamos National Laboratory (LANL). This project investigated separations technology development needed for accelerator-driven transmutation technology (ADTT) concepts, particularly those associated with plutonium disposition (accelerator-based conversion, ABC) and high-level radioactive waste transmutation (accelerator transmutation of waste, ATW). Specific focus areas included separations needed for preparation of feeds to ABC and ATW systems, for example from spent reactor fuel sources, those required within an ABC/ATW system for material recycle and recovery of key long-lived radionuclides for further transmutation, and those required for reuse and cleanup of molten fluoride salts. The project also featured beginning experimental development in areas associated with a small molten-salt test loop and exploratory centrifugal separations systems

  10. Best-estimate methodology for analysis of anticipated transients without scram in pressurized water reactors

    International Nuclear Information System (INIS)

    Union Fenosa, a utility company in Spain, has performed research on pressurized water reactor (PWR) safety with respect to the development of a best-estimate methodology for the analysis of anticipated transients without scram (ATWS), i.e., those anticipated transients for which failure of the reactor protection system is postulated. A scientific and technical approach is adopted with respect to the ATWS phenomenon as it affects a PWR, specifically the Zorita nuclear power plant, a single-loop Westinghouse-designed PWR in Spain. In this respect, an ATWS sequence analysis methodology based on published codes that is generically applicable to any PWR is proposed, which covers all the anticipated phenomena and defines the applicable acceptance criteria. The areas contemplated are cell neutron analysis, core thermal hydraulics, and plant dynamics, which are developed, qualified, and plant dynamics, which are developed, qualified, and validated by comparison with reference calculations and measurements obtained from integral or separate-effects tests

  11. Best-estimate methodology for analysis of anticipated transients without scram in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Fenosa, Madrid (Spain))

    1993-07-01

    Union Fenosa, a utility company in Spain, has performed research on pressurized water reactor (PWR) safety with respect to the development of a best-estimate methodology for the analysis of anticipated transients without scram (ATWS), i.e., those anticipated transients for which failure of the reactor protection system is postulated. A scientific and technical approach is adopted with respect to the ATWS phenomenon as it affects a PWR, specifically the Zorita nuclear power plant, a single-loop Westinghouse-designed PWR in Spain. In this respect, an ATWS sequence analysis methodology based on published codes that is generically applicable to any PWR is proposed, which covers all the anticipated phenomena and defines the applicable acceptance criteria. The areas contemplated are cell neutron analysis, core thermal hydraulics, and plant dynamics, which are developed, qualified, and plant dynamics, which are developed, qualified, and validated by comparison with reference calculations and measurements obtained from integral or separate-effects tests.

  12. What does the phasing out of nuclear energy?; Was bedeutet der Ausstieg aus der Kernenergie?

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2015-07-01

    Since August 6, 2011 the phasing out of nuclear energy in Germany is decided politically. But he is mastered still a long way. In fact, the power economic development until shutdown of nuclear power plants the last in the year 2022 will significantly be affected by the replacement of the contribution to power supply by renewable energy. [German] Seit dem 6.8.2011 ist der Ausstieg aus der Kernenergienutzung in Deutschland politisch beschlossen. Doch bewaeltigt ist er noch laengst nicht. Tatsaechlich wird die stromwirtschaftliche Entwicklung bis zum Abschalten der letzten Kernkraftwerke im Jahr 2022 massgeblich vom Ersatz des entfallenden Versorgungsbeitrages durch erneuerbare Energien gepraegt sein.

  13. Power plant engineering. Vol. 2. Safe and sustainable energy supply; Kraftwerkstechnik. Bd. 2. Sichere und nachhaltige Energieversorgung

    Energy Technology Data Exchange (ETDEWEB)

    Beckmann, Michael; Hurtado, Antonio

    2010-07-01

    Power plant engineering, volume 2 covers scientific, economical and political contributions on the actual development of energy economy. Maint topics are: fossil-fuel power plants, pilot- and new construction projects, CCS-technologies, gas- and steam turbines, coal drying, slugging and corrosion, regenerative energy, nuclear power plants as measuring engineering. (orig./GL) [German] Kraftwerkstechnik, Band 2 umfasst wissenschaftliche, oekonomische und politische Beitraege zu aktuellen Entwicklungen der Energiewirtschaft. Themenschwerpunkte sind fossile Kraftwerke, Pilot- und Neubauprojekte, CCS-Technologien, Gas- und Dampfturbinen, Kohletrocknung, Verschlackung und Korrosion, regenerative Energie, Kernkraftwerke sowie Messtechnik. (orig.)

  14. AMNT 2014. Key Topic: Fuel, decommissioning and disposal - report. Pt. 1

    Energy Technology Data Exchange (ETDEWEB)

    Seipolt, Thomas [NUKEM Technologies Engineering Services GmbH, Alzenau (Germany); Weber, Stefan [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Kock, Ingo [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) GmbH, Koeln (Germany)

    2015-02-15

    Summary report on the following Topical Sessions of the Key Topic 'Fuel, Decommissioning and Disposal' of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - From Pilot Project to an Industrial Service (Thomas Seipolt); - Radioactive Waste Management - Experiences with Interim and Final Storage (Stefan Weber and Ingo Kock). The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 and 12 (2014), 1 (2015) and will be covered in further issues of atw.

  15. Decommissioning and dismantling of the Rossendorf isotope production. Pt. 2. Aspects of implementation; Stilllegung und Rueckbau der Rossendorfer Isotopenproduktion. T. 2. Aspekte der Durchfuehrung

    Energy Technology Data Exchange (ETDEWEB)

    Grahnert, Thomas; Jansen, Sven; Boessert, Wolfgang [VKTA - Strahlenschutz Analytik und Entsorgung Rossendorf e.V., Dresden (Germany); Kniest, Steffen [Siempelkamp NIS Ingenieurgesellschaft mbH, Dresden (Germany)

    2016-06-15

    After just over 40 years of production operation 2000, the operation of the Rossendorf Isotope Production was finally stopped. In the last few years of production already sections of the Rossendorf Isotope Production have been decommissioned. With the end of the isotope production the decommissioning of the entire complex started. In the two-part report, the decommissioning and dismantling of the Rossendorf Isotope production is presented. In part 1 (atw 5/2016) mainly the authorisation procedures and the realised decommissioning concept are presented. Part 2 (atw 6/2016) deals with special selected aspects of the implementation of the decommissioning programme.

  16. Decommissioning and dismantling of the Rossendorf Isotope Production; Stilllegung und Rueckbau der Rossendorfer Isotopenproduktion. T. 1. Betriebshistorie, Genehmigungsverfahren und Planungskonzept

    Energy Technology Data Exchange (ETDEWEB)

    Grahnert, Thomas [VKTA - Strahlenschutz Analytik und Entsorgung Rossendorf e.V., Dresden (Germany). Reststoffbehandlung und Qualitaetswesen; Jansen, Sven [VKTA - Strahlenschutz Analytik und Entsorgung Rossendorf e.V., Dresden (Germany). Betrieblicher Strahlenschutz; Boessert, Wolfgang [VKTA - Strahlenschutz Analytik und Entsorgung Rossendorf e.V., Dresden (Germany). Bereich Rueckbau und Entsorgung; Kniest, Steffen [Siempelkamp NIS Ingenieurgesellschaft mbH, Dresden (Germany)

    2016-05-15

    After just over 40 years of production operation 2000, the operation of the Rossendorf Isotope Production was finally stopped. In the last few years of production already sections of the Rossendorf Isotope Production have been decommissioned. With the end of the isotope production the decommissioning of the entire complex started. In the two-part report, the decommissioning and dismantling of the Rossendorf Isotope production is presented. In part 1 (atw 5/2016) mainly the authorisation procedures and the realised decommissioning concept are presented. Part 2 (atw 6/2016) deals with special selected aspects of the implementation of the decommissioning programme.

  17. 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)

    Summary report on the following Topical Session of the 46th 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.

  18. Evaluation of load case ``switch-off of the high pressure pump of the emergency core cooling system``, measures of verification and in situ-test; Einstufung des Lastfalls ``Ausfall der TH-Hochdruckeinspeisepumpe``, Massnahmen zur Verifikation bis hin zum Grossversuch

    Energy Technology Data Exchange (ETDEWEB)

    Trobitz, M.; Mattheis, A. [Kernkraftwerke Gundremmingen Betriebsgesellschaft m.b.H. (Germany); Kerkhof, K.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Gurr-Beyer, C. [Buero fuer Baudynamik, Stuttgart (Germany); Hofstoetter, P. [Technischer Ueberwachungs-Verein Rheinland e.V., Koeln (Germany)

    1998-11-01

    Within the framework of periodic safety inspection of the Gundremmingen power station (RWE-Bayernwerk - KRB II), the load collectives used for the design of safety-relevant systems and components were checked for their consistency with latest updates of the design basis. It was found that there was no analytical information or study available describing a particular process and its effects, namely switch-off of the high-pressure feedwater pump of the emergency core cooling system. The paper reports the work performed for closing the gap, including preparatory analyses, accompanying measures such as vibration measurements during plant shut-down, as well as the preparation and performance of the in-situ test. The experimental results and the comparative evaluation of calculated and experimental data are presented. (orig./CB) [Deutsch] Im Rahmen der periodischen Sicherheitsueberpruefung des Kernkraftwerkes Gundremmingen (Kernkraftwerke RWE-Bayernwerk - KRB II) wurden u.a. die Lastkollektive, die zur Auslegung sicherheitstechnisch relevanter Systeme und Komponenten herangezogen wurden, auf Aktualitaet ueberprueft. Dabei zeigte sich, dass bislang fuer eine Betriebsweise - naemlich das Abschalten der Hochdruckeinspeisepumpe des nuklearen Not- und Nachkuehlsystems (TH-HD-Pumpe) - keine analytischen Untersuchungen vorliegen. Vorbetrachtungen fuer analytische Untersuchungen, begleitende Massnahmen wie Schwingungsmessungen waehrend des Anlagenstillstandes, sowie der Versuchsaufbau und die Versuchsdurchfuehrung des Anlagenversuches werden hier dargestellt. Die Ergebnisse und der Vergleich Rechnung-Messung zum Grossversuch werden in diesem Beitrag vorgestellt. (orig.)

  19. Nanoindentation and micro-mechanical fracture toughness of electrodeposited nanocrystalline Ni–W alloy films

    International Nuclear Information System (INIS)

    Nanocrystalline nickel–tungsten alloys have great potential in the fabrication of components for microelectromechanical systems. Here the fracture toughness of Ni–12.7 at.%W alloy micro-cantilever beams was investigated. Micro-cantilevers were fabricated by UV lithography and electrodeposition and notched by focused ion beam machining. Load was applied using a nanoindenter and fracture toughness was calculated from the fracture load. Fracture toughness of the Ni–12.7 at.%W was in the range of 1.49–5.14 MPa √m. This is higher than the fracture toughness of Si (another important microelectromechanical systems material), but considerably lower than that of electrodeposited nickel and other nickel based alloys. - Highlights: ► Micro-scale cantilevers manufactured by electro-deposition and focused ion beam machining. ► Nanoindenter used to perform micro-scale fracture test on Ni-13at%W micro-cantilevers. ► Calculation of fracture toughness of electrodeposited Ni-13at%W thin films. ► Fracture toughness values lower than that of nanocrystalline nickel.

  20. 47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key topic / Outstanding know-how and sustainable innovations

    Energy Technology Data Exchange (ETDEWEB)

    Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Bereich Reaktorsicherheitsforschung

    2016-07-15

    Summary report on the Technical Session: ''Reactor Physics, Thermo- and Fluid-Dynamics'' of the 47th Annual Conference on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 will be covered in further issues of atw.

  1. Results of the brugge benchmark study for flooding optimization and history matching

    NARCIS (Netherlands)

    Peters, E.; Arts, R.J.; Brouwer, G.K.; Geel, C.R.; Cullick, S.; Lorentzen, R.J.; Chen, Y.; Dunlop, K.N.B.; Vossepoel, F.C.; Xu, R.; Sarma, P.; Alhutali, A.H.; Reynolds, A.C.

    2010-01-01

    In preparation for the SPE Applied Technology Workshop (ATW) held in Brugge in June 2008, a unique benchmark project was organized to test the combined use of waterflooding-optimization and history-matching methods in a closed-loop workflow. The benchmark was organized in the form of an interactive

  2. Nanoindentation and micro-mechanical fracture toughness of electrodeposited nanocrystalline Ni-W alloy films

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, D.E.J., E-mail: david.armstrong@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford, OX1 3PH (United Kingdom); Haseeb, A.S.M.A. [Department of Mechanical Engineering, University of Malaya, 50603 Kuala Lumpur (Malaysia); Roberts, S.G.; Wilkinson, A.J. [Department of Materials, University of Oxford, Parks Road, Oxford, OX1 3PH (United Kingdom); Bade, K. [Institut fuer Mikrostrukturtechnik (IMT), Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2012-04-30

    Nanocrystalline nickel-tungsten alloys have great potential in the fabrication of components for microelectromechanical systems. Here the fracture toughness of Ni-12.7 at.%W alloy micro-cantilever beams was investigated. Micro-cantilevers were fabricated by UV lithography and electrodeposition and notched by focused ion beam machining. Load was applied using a nanoindenter and fracture toughness was calculated from the fracture load. Fracture toughness of the Ni-12.7 at.%W was in the range of 1.49-5.14 MPa {radical}m. This is higher than the fracture toughness of Si (another important microelectromechanical systems material), but considerably lower than that of electrodeposited nickel and other nickel based alloys. - Highlights: Black-Right-Pointing-Pointer Micro-scale cantilevers manufactured by electro-deposition and focused ion beam machining. Black-Right-Pointing-Pointer Nanoindenter used to perform micro-scale fracture test on Ni-13at%W micro-cantilevers. Black-Right-Pointing-Pointer Calculation of fracture toughness of electrodeposited Ni-13at%W thin films. Black-Right-Pointing-Pointer Fracture toughness values lower than that of nanocrystalline nickel.

  3. Reproductive hormones regulate the selective permeability of the blood-brain barrier

    OpenAIRE

    Wilson, Andrea C.; Clemente, Luca; Liu, Tianbing; Bowen, Richard L.; Meethal, Sivan Vadakkadath; Atwood, Craig S.

    2008-01-01

    Reproductive hormones regulate the selective permeability of the blood-brain barrier : Current address: Department of Biochemistry, Colorado State University, CO, USA. (Clemente, Luca) correspondence: Corresponding author. University of Wisconsin-Madison Medical School, Wm S. Middleton Memorial VA (GRECC 11G), 2500 Overlook Terrace, Madison, WI 53705, USA. Tel.: +1 608 256 1901x11664; fax: +1 608 280 7291. (Atw...

  4. Provisions for nuclear damage - the existing safety philosophy; Atomrechtliche Schadensvorsorge - die bisherige Sicherheitsphilosophie

    Energy Technology Data Exchange (ETDEWEB)

    Winter, U.; Gloeckle, W. [Ministerium fuer Umwelt und Verkehr, Stuttgart (Germany). Abt. 7

    2001-01-01

    The political boundary conditions for the continued operation of nuclear power plants in Germany were defined in the agreement of June 14, 2000 by the present German federal government and the power utilities operating nuclear power plants. The agreement is to be signed finally after the legislative process for the tenth amendment to the German Atomic Energy Act will have been concluded. With reference to the existing high level of safety of nuclear power plants in Germany, the federal government agreed in the paper, inter alia, that the safety standards and the safety philosophy underlying them are not to be changed. Both the safety standards and the safety philosophy are tied to the central legal rule under the German Atomic Energy Act about provisions to be made for damage; as the wording of the law shows, they have remained so-called indefinite legal concepts not specified any further. The necessary specificity so far has been achieved by a number of court rulings and executive decisions in which the constitutionality of the Atomic Energy Act with respect to the provisions for nuclear damage was answered in the affirmative, and a distinction was made between provisions serving to avert danger and the acceptable residual risk. The continued need for specification by the administration has been met in a number of safety criteria for nuclear power plants and in the accident guidelines, both of which were put into practice by the federal states, within the framework of the States Committee on Atomic Energy, and the federal government together with the expert bodies. Also internationally the concept has been confirmed at the first verification meeting of the Convention on Nuclear Safety. (orig.) [German] Die politischen Rahmenbedingungen fuer den weiteren Betrieb der Kernkraftwerke in Deutschland wurden zwischen der jetzigen Bundesregierung und den betreibenden Energieversorgungsunternehmen mit der Vereinbarung vom 14. Juni 2000 festgelegt. Die endgueltige

  5. The Chernobyl reactor accident, ten years on. Teaching projects for mathematics instruction in interdisciplinary working groups; 10 Jahre nach Tschernobyl. Unterrichtsprojekte fuer den Mathematikunterricht in faecheruebergreifenden Kooperationen

    Energy Technology Data Exchange (ETDEWEB)

    Boer, H. [comp.; Delle, E. [comp.; Mies, K. [comp.; Warmeling, A. [comp.

    1996-10-01

    The booklet presents background information and addresses the following aspects: ionizing radiation and radiation effects; safety of German nuclear power plants; statistical evidence of radiation injuries; short-lived and long-lived ionizing radiation; radioactive waste; CO{sub 2} emissions as an argument in favour of nuclear power generation. The material presented is intended for use by a school project team interested in the subjects, or as a basis for collaborative, interdisciplinary teaching in working groups, and it offers information and problems for mathematics teaching. (HP) [Deutsch] Neben vielen Informationen behandelt die Broschuere: Strahlen und Strahlenwirkungen; Sicherheit deutscher Kernkraftwerke; statistischer Nachweis von Strahlenschaeden; Kurz- und Langfestigkeit der Strahlenbelastung; radioaktiver Abfall; CO{sub 2}-Problematik als Argument fuer die Kernenergie. Die Broschuere ist gedacht z.B. fuer eine Projektgruppe, einen Projekttag, fuer eine Lerngruppe in faecheruebergreifender Kooperation. Die Materialien sind ausgearbeitet fuer die Themembearbeitung im Mathematikunterricht mit Uebungsaufgaben. (HP)

  6. Sustainable power generation and utilisation; Verantwortbare Erzeugung und Nutzung von Energie

    Energy Technology Data Exchange (ETDEWEB)

    Hennenhoefer, G.

    2003-07-01

    In 2001, one third of the total power generated in Germany was provided by nuclear power plants. In base load power supply their share is even higher, amounting to more than 50 percent. The decision of the German government to phase out nuclear power raises the question of how this power will be provided in the future. [German] 2001 produzierten die deutschen Kernkraftwerke ein Drittel der oeffentlichen Stromversorgung. In der Grundlast, dem ''Rund-um-die-Uhr-Strom'', bildet die Kernenergie mit einem Anteil von ueber 50% das Rueckgrat unserer Stromerzeugung. In wenigen Jahren muessen wir entscheiden, wie das von der Bundesregierung geforderte Auslaufen kompensiert werden soll. Denn in etwa zehn Jahren erwarten wir Bedarf an neuen Kraftwerkskapazitaeten zum Ersatz auslaufender konventioneller und nuklearer Stromerzeugungsanlagen. (orig.)

  7. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  8. Accelerator transmutation of 129I

    International Nuclear Information System (INIS)

    Iodine-129 is one of several long-lived reactor products that is being considered for transmutation by the Los Alamos Accelerator Transmutation of Waste (ATW) program. A reasonable rate of transmutation of 1291 is possible in this system because of the anticipated high neutron flux generated from the accelerator. This report summarizes previous papers dealing with the transmutation of 1291 where reactor technologies have been employed for neutron sources. The transmutation process is considered marginal under these conditions. Presented here are additional information concerning the final products that could be formed from the transmutation process in the ATW blanket. The transmutation scheme proposes the use of solid iodine as the target material and the escape of product xenon from the containers after van Dincklange (1981). Additional developmental plans are considered

  9. Morphological Atherosclerosis Calcification Distribution (MACD) Index is a Strong Predictor of Cardio-Vascular Death and Include Predictive Power of BMD

    DEFF Research Database (Denmark)

    Christiansen, Claus; Karsdal, Morten; Ganz, Melanie;

    followed for 8.3±0.3 years and CVD deaths were recorded. BMD and several aortic calcification markers were computed: number, morphology, distribution, from outlines of the calcified plaques in lumbar X-rays. These markers were compared to BMD, SCORE card, Framingham score, and the Aortic Calcification...... Severity score - AC24. AC24 adjusted by age, waist circumference, and triglyceride levels (ATW) predicted mortality in postmenopausal women (CVD p=0.03, All-cause p=0.006). The SCORE card and the Framingham score resulted in mortality odds ratios (MOR) of 5.0 and 5.2 - defining high risk as =6 and =18......, respectively. BMD and BMD adjusted for ATW was lower in the group of deceased than in survivors (pscores based on the calcification geometry provided highly significant predictions. The number of calcified deposits...

  10. Pyrochemical separations technologies envisioned for the U.S. accelerator transmutation of waste system

    International Nuclear Information System (INIS)

    A program has been initiated for the purpose of developing the chemical separations technologies necessary to support a large Accelerator Transmutation of Waste (ATW) system capable of dealing with the projected inventory of spent fuel from the commercial nuclear power stations in the United States. The baseline process selected combines aqueous and pyrochemical processes to enable the efficient separation of uranium, technetium, iodine, and the transuranic elements from LWR spent fuel. The diversity of processing methods was chosen for both technical and economic factors. A six-year technology evaluation and development program is foreseen, by the end of which an informed decision can be made on proceeding with demonstration of the ATW system

  11. Three-dimensional time-dependent star reactor kinetics analysis coupled with RETRAN and MCPWR system response

    International Nuclear Information System (INIS)

    The operation of a nuclear power plant must be continually supported by analyses which may include FSAR design basis and best-estimate thermal-hydraulic (T/H) and reactor dynamics analyses. The development and improvement of new analysis techniques provide many advantages including the capability to evaluate the impact of modeling assumptions made in previous rector kinetics and T/H calculations. The methodology presented in this paper shows how the time-dependent, three-dimensional reactor kinetics STAR nodal code can be directly coupled with the overall RCS T/H codes, RETRAN and MCPWR, in a tandem, iterative approach. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective mode for three PWR transients: Loss of Feedwater Anticipated Transient Without Scram (ATWS); Station Blackout ATWS; and a Total Loss of Reactor Coolant System (RCS) Flow with a control rod scram

  12. Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

    International Nuclear Information System (INIS)

    The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scram (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram

  13. Fuel rod behaviour during transients

    International Nuclear Information System (INIS)

    The clad deformation code CANSWELL-2 is described. This is used, either as a stand-alone code or within MABEL-2, to predict and analyse the results of LOCA simulations in the Halden and NRU reactors and in the KfK and PROPAT rigs. Experimental evidence on fuel behaviour in RIA, PCM and ATWS events is presented with inclusion of certain FRAP-T5 results. Published calculations from the accident codes FRAP-T4 and FRAP-T5 are compared with experimental results in simulated loss of coolant tests in the Power Burst Facility. The limitations of this code in its treatment of RIA, PCM and ATWS events are considered. (U.K.)

  14. Recrystallization textures of powder metallurgically prepared pure Ni, Ni-W and Ni-Mo alloy tapes for use as substrates for coated superconductors

    International Nuclear Information System (INIS)

    Development of cube texture after heavy cold deformation and annealing has been studied in powder metallurgically prepared pure Ni, Ni-5at.%Mo and Ni-5at.%W alloys for use as substrates for coated superconductor applications. Two grades of Ni powder with different purities have been used to prepare the initial materials. Addition of W and Mo is found to be beneficial in increasing the volume fraction of the cube component, irrespective of the purity of the Ni powder used. W particularly increases the volume fraction of the cube component in Ni by decreasing the volume fraction of the RD (rolling direction)-rotated cube grains. Studies on partially recrystallized samples indicate that in contrast to pure Ni, in Ni-5at.%W alloy the recrystallized grains are mostly cube oriented right from the beginning of recrystallization

  15. Fabrication of YBCO Coated Conductors on Biaxial Textured Metal Substrate by All-Sputtering

    Institute of Scientific and Technical Information of China (English)

    Xiao Han; Jing-Tan He; Jie Xiong; Bo-Wan Tao

    2008-01-01

    CeO2/YSZ/CeO2 buffer layers were prepared on biaxial textured Ni-5at.%W substrate by direct-current magnetron reactive sputtering with the optimum process. YBCO thin films were deposited on CeO2/YSZ/CeO2 buffered Ni-5at.%W substrate at temperature ranging from 500°C to 700°C by diode dc sputtering. By optimizing substrate temperature, pure c-axis oriented YBCO films were obtained. The microstructures of CeO2/YSZ/CeO2 buffer layers were characterized by X-ray diffraction. A smooth, dense and crack-free surface morphology was observed with scanning electron microscopy. The critical current density Jc about 0.75 MA/cm2 at 77 K was obtained.

  16. 2012 annual meeting on nuclear technology. Pt. 2. Section reports

    International Nuclear Information System (INIS)

    Summary report on 2 out of 12 sessions of the Annual Conference on Nuclear Technology held in Stuttgart, 22 to 24 May 2012: - Fusion technology (Section 9), and - Radiation protection (Section 11). The sessions of the sections: - Reactor physics and methods of calculation (Section 1), - Thermodynamics and fluid dynamics (Section 2), - Safety of nuclear installations - methods, analysis, results (Section 3), - Front end of the fuel cycle, fuel elements and core components (Section 4), - Radioactive waste management, storage (Section 5), - Operation of nuclear installations (Section 6), - New build and innovations (Section 7), - Decommissioning of nuclear installations (Section 8), and - Energy economics (Section 10) will be covered in further issues of atw. The report on the session: - Education, expert knowledge, know-how-transfer (Section 12) has been covered in atw 8/9 (2012). (orig.)

  17. Cube-textured metal substrates for reel-to-reel processing of coated conductors

    DEFF Research Database (Denmark)

    Wulff, Anders Christian

    , migrating boundaries were found to abandon grooves and generate grooves at new positions. Despite the observed changes in the extent of grain boundary grooving, the mean surface roughness was almost identical before and after the additional annealing. Microstructure, texture, hardness and magnetic....... It was observed that the initial mean surface roughness decreased after annealing except after very ne polishing. Additionally, the roughness of the buer layers were found to increase slightly for the ne polished substrates. Grain boundary grooving was observed to impose a lower limit for the mean surface...... properties have been studied in a series of new Ni-Cu-W substrates. Adding 5 at.% copper to Ni-5at.%W was observed to substantially decrease the Curie temperature and the saturation mass magnetisation without signicantly modifying the microstructure and texture compared with Ni-5at.%W. The hardness...

  18. Safety analysis and neutronics of accelerator-driven transmutation of wastes with concurrent energy production. Annual report for the year 1996

    International Nuclear Information System (INIS)

    The research activities have significantly expanded compared to the earlier period and were during 1996 concentrated on the following major objectives: ATW system studies, simulations, optimization and design of spallation targets, benchmarking of the calculational tools in the frame of the IAEA coordinated research project, development of the computer codes for ADS and some experimental activities on the subcritical reactor MASURCA. Under 1996 a very extensive international collaboration network has been further developed and many collaborative research projects have been launched. 31 refs

  19. Evaluation of Inherent Safety Features of the KALIMER-600 Design Concept for Anticipated Transient Without Scram Events

    International Nuclear Information System (INIS)

    KAERI is developing the KALIMER-600 design concept, which is a 600 MWt metallic fuelled pool-type sodium-cooled fast reactor, under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding GEN-IV nuclear power plants such as plant safety, economics, proliferation resistance, and sustainability. In this report, key safety design features are briefly described and safety analyses results for typical ATWS accidents are presented. The design specifications improved during the 4th design phase of the KALIMER program were reflected in the analysis. Three ATWS events as the most relevant for evaluation of passive safety design features were selected. These are unprotected transient overpower (UTOP), unprotected loss of flow (ULOF), and unprotected loss of heat sink (ULOHS). The plant responses and safety margins of the KALIMER-600 during three ATWS events were investigated using the system transient code SSC-K version 1.3 developed by KAERI. The key characteristic necessary for passive safety and inherent self-protection is an overall negative reactivity feedback response to reactor accident initiators. Because the passive safety mechanisms are the result of tightly coupled thermal, hydraulic, neutronic, and mechanical physical phenomena, analysis and investigation methods employing detailed computational models of the phenomena and geometry are required to permit accurate quantification of effects. The analysis results by SSC-K shows that the KALIMER-600 design has inherent safety characteristics and is capable of accommodating selected ATWS events. The passive safety mechanism in the KALIMER-600 design makes the core shutdown with sufficient margin and the passive removal of decay heat. The self-regulation of power without scram is mainly due to the inherent reactivity feedbacks in conjunction with the passive decay heat removal

  20. Current US plans for development of fuels for accelerator transmutation of waste

    International Nuclear Information System (INIS)

    The United States is currently investigating the feasibility of proposed technologies for the Accelerator Transmutation of Waste (ATW) concept, which is funded as part of the U.S. Department of Energy's Advanced Accelerator Applications (AAA) Program. The ATW concept is proposed as a means to transmute transuranic isotopes and, perhaps, long-lived fission products removed from light water reactor spent fuel to shorter-lived fission products. To attain maximum possible transmutation rates, no fertile material (i.e., U-238 or Th-232) is to be incorporated into the fuel. Fuel forms currently proposed for ATW application include non-fertile dispersions of metal alloy or nitride fuel particles in a metal matrix, a non-fertile metal alloy, or non-fertile nitride pellets for a fast-spectrum, liquid metal-cooled transmuter, and non-fertile TRISO-coated particles dispersed in graphite compacts for a thermal-spectrum, gas-cooled transmuter. There is little or no experience with these non-fertile fuels, so an extensive fuel development program is envisioned. Current plans call for initial effort to demonstrate feasibility of the proposed fuel forms by the end of 2005, consistent with AAA program decision milestones. Feasibility research and development will consist of the following: Development of fabrication processes to demonstrate fabricability of the proposed fuel forms; Simple irradiation tests to screen samples of each fuel type for unexpected or poor performance; and Determination of intrinsic properties or characteristics (e.g., out-of pile interdiffusion behavior of fuel and constituents and thermophysical properties). If the decision is made to continue development of the ATW concept beyond 2005, then of the successful candidate forms, one or two will be selected for further development, with more extensive irradiation testing and fuel property characterization. (author)

  1. Rotationally acquired four-dimensional optical coherence tomography of embryonic chick hearts using retrospective gating on the common central A-scan

    DEFF Research Database (Denmark)

    Happel, Christoph M.; Thommes, Jan; Thrane, Lars;

    2011-01-01

    We introduce a new method of rotational image acquisition for four-dimensional (4D) optical coherence tomography (OCT) of beating embryonic chick hearts. The rotational axis and the central A-scan of the OCT are identical. An out-of-phase image sequence covering multiple heartbeats is acquired at.......We demonstrate this approach and provide a video of a beating Hamburger and Hamilton stage 16 embryonic chick heart generated from a 4D OCT data set using rotational image acquisition....

  2. TRACE Model for Simulation of Anticipated Transients Without Scram in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Cheng L. Y.; Baek J.; Cuadra,A.; Aronson, A.; Diamond, D.; Yarsky, P.

    2013-11-10

    A TRACE model has been developed for using theTRACE/PARCS computational package [1, 2] to simulate anticipated transients without scram (ATWS) events in a boiling water reactor (BWR). The model represents a BWR/5 housed in a Mark II containment. The reactor and the balance of plant systems are modeled in sufficient detail to enable the evaluation of plant responses and theeffectiveness of automatic and operator actions tomitigate this beyond design basis accident.The TRACE model implements features thatfacilitate the simulation of ATWS events initiated by turbine trip and closure of the main steam isolation valves (MSIV). It also incorporates control logic to initiate actions to mitigate the ATWS events, such as water levelcontrol, emergency depressurization, and injection of boron via the standby liquid control system (SLCS). Two different approaches have been used to model boron mixing in the lower plenum of the reactor vessel: modulate coolant flow in the lower plenum by a flow valve, and use control logic to modular.

  3. Disposition of nuclear waste using subcritical accelerator-driven systems

    International Nuclear Information System (INIS)

    Studies have shown that the repository long-term radiological risk is from the long-lived transuranics and the fission products Tc-99 and I-129, thermal loading concerns arise mainly form the short-lived fission products Sr-90 and Cs-137. In relation to the disposition of nuclear waste, ATW is expected to accomplish the following: (1) destroy over 99.9% of the actinides; (2) destroy over 99.9% of the Tc and I; (3) separate Sr and Cs (short half-life isotopes); (4) separate uranium; (5) produce electricity. In the ATW concept, spent fuel would be shipped to a ATW site where the plutonium, other transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their only pass through the facility. This approach contrasts with the present-day reprocessing practices in Europe and Japan, during which high purity plutonium is produced and used in the fabrication of fresh mixed-oxide fuel (MOX) that is shipped off-site for use in light water reactors

  4. A realistic anticipated transient without scram evaluation of the Zorita nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Rebollo, L. (Union Fenosa, Madrid (Spain))

    1993-07-01

    A best-estimate methodology for analysis of an anticipated transient without scram (ATWS) in a pressurized water reactor (PWR) is applied to the simulation of the passive response to postulated ATWS scenarios of the Jose Cabrera (Zorita) nuclear power plant (NPP) owned and operated by Union Fenosa, which is the only Westinghouse PWR with a single coolant loop. A justification of the calculation hypotheses is included. The results of the specific studies are evaluated, and the conclusion is that the intrinsic safety margins of the original design of the plant guarantees the integrity of the fuel, primary circuit, and containment, without the need to incorporate an automatic ATWS mitigation system. Finally, a suitable plant-specific prototype emergency operating procedure is designed that is substantially different from the previous Zorita NPP procedure and from the generic procedure applicable to multiloop plants. This procedure is validated by simulating the operator-plant interface by means of a validation matrix including the scenarios presenting the most adverse dynamic modes foreseeable.

  5. A realistic anticipated transient without scram evaluation of the Zorita nuclear power plant

    International Nuclear Information System (INIS)

    A best-estimate methodology for analysis of an anticipated transient without scram (ATWS) in a pressurized water reactor (PWR) is applied to the simulation of the passive response to postulated ATWS scenarios of the Jose Cabrera (Zorita) nuclear power plant (NPP) owned and operated by Union Fenosa, which is the only Westinghouse PWR with a single coolant loop. A justification of the calculation hypotheses is included. The results of the specific studies are evaluated, and the conclusion is that the intrinsic safety margins of the original design of the plant guarantees the integrity of the fuel, primary circuit, and containment, without the need to incorporate an automatic ATWS mitigation system. Finally, a suitable plant-specific prototype emergency operating procedure is designed that is substantially different from the previous Zorita NPP procedure and from the generic procedure applicable to multiloop plants. This procedure is validated by simulating the operator-plant interface by means of a validation matrix including the scenarios presenting the most adverse dynamic modes foreseeable

  6. Dynamic simulation of a sodium-cooled fast reactor power plant

    Energy Technology Data Exchange (ETDEWEB)

    Shinaishin, M.A.M.

    1976-08-01

    Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

  7. Materials compatibility and corrosion issues for accelerator transmutation of waste

    International Nuclear Information System (INIS)

    The need to understand the materials issues in an accelerator transmutation of waste (ATW) system is essential. This report focuses on the spallation container material, as this material is exposed to some of the most crucial environmental conditions of simultaneous radiation and corrosion in the system. The most severe design being considered is that of liquid lead. In previous investigations of lead compatibility with materials, the chemistry of the system was derived solely from the corrosion products; however, in an ATW system, the chemistry of the lead changes not only with the derived corrosion products of the material being tested but also with the buildup of the daughter production with time. Daughter production builds up and introduces elements that may have a great effect on the corrosion activity of the liquid lead. Consequently, data on liquid lead compatibility can be regarded only as a guide and must be reevaluated when particular daughter products are added. This report is intended to be a response to specific materials issues and concerns expressed by the ATW design working group and addresses the compatibility/corrosion concerns

  8. Safety Analysis for Key Design Features of KALIMER-600 Design Concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Jeong, H. Y.; Ha, K. S

    2007-02-15

    This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies.

  9. Accelerator-driven Transmutation of Waste

    Science.gov (United States)

    Venneri, Francesco

    1998-04-01

    Nuclear waste from commercial power plants contains large quantities of plutonium, other fissionable actinides, and long-lived fission products that are potential proliferation concerns and create challenges for the long-term storage. Different strategies for dealing with nuclear waste are being followed by various countries because of their geologic situations and their views on nuclear energy, reprocessing and non-proliferation. The current United States policy is to store unprocessed spent reactor fuel in a geologic repository. Other countries are opting for treatment of nuclear waste, including partial utilization of the fissile material contained in the spent fuel, prior to geologic storage. Long-term uncertainties are hampering the acceptability and eventual licensing of a geologic repository for nuclear spent fuel in the US, and driving up its cost. The greatest concerns are with the potential for radiation release and exposure from the spent fuel for tens of thousands of years and the possible diversion and use of the actinides contained in the waste for weapons construction. Taking advantage of the recent breakthroughs in accelerator technology and of the natural flexibility of subcritical systems, the Accelerator-driven Transmutation of Waste (ATW) concept offers the United States and other countries the possibility to greatly reduce plutonium, higher actinides and environmentally hazardous fission products from the waste stream destined for permanent storage. ATW does not eliminate the need for, but instead enhances the viability of permanent waste repositories. Far from being limited to waste destruction, the ATW concept also brings to the table new technologies that could be relevant for next-generation power producing reactors. In the ATW concept, spent fuel would be shipped to the ATW site where the plutonium, transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their first and only pass through the

  10. Safety management in nuclear power plants as seen by a Regulatory Authority; Das Sicherheitsmanagement von Kernkraftwerken aus Sicht der atomrechtlichen Aufsichtsbehoerde

    Energy Technology Data Exchange (ETDEWEB)

    Seidel, E.R. [Bayerisches Staatsministerium fuer Umwelt, Gesundheit und Verbraucherschutz, Muenchen (Germany); Rauh, H.J. [TUeV Sueddeutschland Bau und Betrieb GmbH, Muenchen (Germany)

    2004-03-01

    Over the past few years, the regulatory authorities supervising the operation of German nuclear power plants on behalf of the government have been faced increasingly by problems of safety management and safety culture. So far, notifiable events have not affected the public or the environment. This is due to the fault-tolerant design of nuclear power plants and their effective supervision by government authorities. Operators and regulatory authorities share the understanding that maximum safety of plants should be ensured as a matter of priority even over economic principles. The most perfect safety management, and the high safety culture it promotes, are indispensable parts of safety philosophy. Instructions about practical measures enable the operators of nuclear power plants to elaborate quality goals for a systematically designed safety management. Continuous observation of indicators based on reference requirements (best-practice levels) puts the regulatory authorities in a position to detect systematically, though only by indirect methods, the beginnings of negative developments in the safety management of a plant. In addition, such regulatory indicators make actions by the regulatory authorities more transparent to the operator and to the public at large, thus contributing greatly to the objective assessment of the safety of nuclear power plants. (orig.) [German] In den letzten Jahren sind bei der staatlichen Aufsicht ueber den Betrieb der deutschen Kernkraftwerke Fragen des Sicherheitsmanagements und der Sicherheitskultur immer mehr in den Vordergrund gerueckt. Meldepflichtige Ereignisse haben wegen der fehlerverzeihenden Auslegung der Kernkraftwerke und ihrer effektiven staatlichen Ueberwachung bisher zu keiner Beeintraechtigung der Bevoelkerung oder der Umwelt gefuehrt. Es ist gemeinsames Verstaendnis der Betreiber und der Aufsichtsbehoerden hoechstmoegliche Sicherheit der Anlagen zu gewaehrleisten, vor jeder wirtschaftlichen Ueberlegung. Ein moeglichst

  11. Proceedings of the workshop on the implementation of severe accident management measures

    International Nuclear Information System (INIS)

    The OECD/NEA Workshop on the Implementation of Severe Accident Management (SAM) Measures was hosted by the PSI (Paul Schemer Institut), by two Swiss Utilities (Kernkraftwerk Beznau and Kernkraftwerk Leibstadt), and by Electricite de France. Eighty specialists from fourteen OECD Member countries attended the meeting, as well as specialists from three non-Member economies and the European Commission. Thirty-three papers were presented in four sessions, preceded by a brief Introductory Session (two invited papers) and followed by a General Discussion. The objectives of the meeting were: 1) to exchange information on activities in the area of SAM implementation and on the rationale for such actions, 2) to monitor progress made, 3) to identify cases of agreement or disagreement, 4) to discuss future orientations of work, 5) to make recommendations to the CSNI. Session summaries prepared by the Chairpersons and discussed by the whole writing group are given in Annex. During the first session, 'SAM Programmes Implementation', papers from one regulator and several utilities and national research institutes were presented to outline the status of implementation of SAM programmes in countries like Switzerland, Russia, Spain, Finland, Belgium and Korea. Also, the contribution of SAM to the safety of Japanese plants (in terms of core damage frequency) was quantified in a paper. One paper gave an overview on the situation regarding SAM implementation in Europe. The second session, 'SAM Approach', provided background and bases for Severe Accident Management in countries like Sweden, Japan, Germany and Switzerland, as well as for hardware features in advanced light water reactor designs, such as the European Pressurised Reactor (EPR), regarding Severe Accident Management. The third session, 'SAM Mitigation Measures', was about hardware measures, in particular those oriented towards hydrogen mitigation where fundamentally different approaches have been taken in Scandinavian

  12. Safety first - safety standards and safety management in Germany; Safety First - Sicherheitsstandards und Sicherheitsmanagement in Deutschland

    Energy Technology Data Exchange (ETDEWEB)

    Brockmeier, U. [EnBW Kraftwerke AG, Stuttgart (Germany)

    2003-03-01

    unter denen sich Risikotechnologien, wie die friedliche Nutzung der Kernenergie heute entwickeln. Einem Objektivierungsprozess hinsichtlich der Wahrnehmung von Risiken kommt damit eine Schluesselrolle im Hinblick auf die weitere Nutzung der Kernenergie in Deutschland zu. Das in Deutschland im Konsens gewachsene Regelwerk wird auch zukuenftig - gerade weil es sich in weiten Bereichen im untergesetzlichen Rahmen bewegt - einen wesentlichen Beitrag zu diesem Objektivierungsprozess leisten. Dies gilt insbesondere auch fuer das Projekt KTA 2000. Ein Schwerpunkt in der Weiterentwicklung des Sicherheitsmanagements deutscher Kernkraftwerke liegt derzeit in der Konzeptionierung und Umsetzung indikator-gestuetzter Sicherheitsmanagementsysteme. Die auf der DIN ISO 9001 basierenden Systeme sind durch ihre Prozessorientierung und Indikatorsteuerung ein entscheidender Beitrag zur Objektivierung der Wahrnehmung des Sicherheitsniveaus der deutschen Kernkraftwerke sowohl nach Innen zu den Betriebsmannschaften als auch nach Aussen zu Gutachtern, Behoerden, Politik und Oeffentlichkeit. (orig.)

  13. Study on depressurization measurements and effect in PWR

    International Nuclear Information System (INIS)

    Implementation of new regulations on nuclear powered plant design and operation raise new design and management requirement for plants, and the operational plants also need accident management to enhance the reactor operation safety. Thus, for sake of reducing risk of high-pressure and mitigating the consequence, depressurization is a measure carried out to reduce primary pressure. With SCDAP/RELAP5 this paper studies the depressurization measurements and effect factors in pressurized water reactor under the important severe accident sequences induced by very small break lost of coolant accident (VSBLOCA), anticipated transient without scram (ATWS) and station blackout (SBO) plus auxiliary feedwater failure. (author)

  14. Nuclear questions

    International Nuclear Information System (INIS)

    This brochure 'nuclear problems' deals with the attitude of the protestant church in the region around the northern Elbe towards further quantitative economic growth, esp. nuclear energy, with the following essays: preaching the Gospel in an environment in danger: the Christian occident and the problems of the third world, facing the problems of exhausted supplies, the role of the prophet, problem of environment - a problem of theology, the political dimension, against ATW, signal Brokdorf, strange effects (defense of the church from unqualified teachings by non-professionals), Christian liberty, church and nuclear energy, violence and robes. (HP)

  15. Dynamic characteristics of a VK-50 reactor operating under conditions of the loss of a normal feedwater flow

    Science.gov (United States)

    Semidotskiy, I. I.; Kurskiy, A. S.

    2013-12-01

    The paper describes the conditions of the ATWS type with virtually complete cessation of the feed-water flow at the operating power level of a reactor of the VK-50 type. Under these conditions, the role of spatial kinetics in the system of feedback between thermohydraulic and nuclear processes with bulk boiling of the coolant in the reactor core is clearly seen. This feature determines the specific character of experimental data obtained and the suitability of their use for verification of the associated codes used for calculating water-water reactors.

  16. Nuclear power risk criteria for Mexico

    International Nuclear Information System (INIS)

    The preliminary sequence of events for three types of LOCAs (low, medium and large) and seven transients, particularly turbine trip, loss of offsite power were developed. All of the systems involved in the sequence were examined and analysed in detail and a success criterion was defined for each system in accordance with the initiating events. The difference of transient with and without SCRAM was discussed and a special sequence for the last case (ATWS) was developed. The quantification of the sequence was performed using some results from the PSA (level 1) for Laguna Verde Nuclear Power Plant (LVNPP) and the most significant sequences were shown. 16 refs, 7 figs, 7 tabs

  17. Trends vs. reactor size of passive reactivity shutdown and control performance

    International Nuclear Information System (INIS)

    In LMR [liquid metal reactor] concepts, favorable passive reactivity shutdown performance in response to ATWS [anticipated transient without scram] events has been shown to be achievable when measurable, integral reactivity parameters satisfy certain requirements. The trends in the integral reactivity parameters with reactor size for both oxide and metal fuel have been developed based on a data base of about two dozen reactor designs in the range 400 to 3,600 MWth. The general conclusion is that the favorable passive reactivity control features which accrue to the metallic-fueled reactors in the modular size range can be achieved as well in the larger commercial sizes

  18. Special MAFIA postprocessors for the analysis of rf structures

    Energy Technology Data Exchange (ETDEWEB)

    Browman, M.J.

    1992-01-01

    This paper describes three stand-alone programs that use the electromagnetic fields generated by the MAFIA 2.04 codes to analyze radio-frequency (RF) cavities. Illustrations are provided that show how these codes are used to (1) analyze the effect of the coupling slots on the electric and magnetic fields of the linacs for the APLE Prototype Experiment (APEX) and the Advanced Free-Electron Laser (AFEL), (2) verify the Panofsky-Wenzel theorem for a high-energy deflecting cavity proposed for the Accelerator Transmutation of Waste (ATW) project, and (3) study the effectiveness of that deflecting cavity.

  19. Special MAFIA postprocessors for the analysis of rf structures

    Energy Technology Data Exchange (ETDEWEB)

    Browman, M.J.

    1992-09-01

    This paper describes three stand-alone programs that use the electromagnetic fields generated by the MAFIA 2.04 codes to analyze radio-frequency (RF) cavities. Illustrations are provided that show how these codes are used to (1) analyze the effect of the coupling slots on the electric and magnetic fields of the linacs for the APLE Prototype Experiment (APEX) and the Advanced Free-Electron Laser (AFEL), (2) verify the Panofsky-Wenzel theorem for a high-energy deflecting cavity proposed for the Accelerator Transmutation of Waste (ATW) project, and (3) study the effectiveness of that deflecting cavity.

  20. Abstracts of the 37. annual aquatic toxicity workshop : big cities, big challenges, great solutions : urbanization and environmental impacts

    International Nuclear Information System (INIS)

    The aquatic toxicity workshop (ATW) is Canada's major annual meeting in the field of aquatic toxicology. It provides a forum to discuss current and emerging topics regarding water quality. Participants included students, academics, regulators, environmental consultants and industry representatives interested in the field of ecotoxicology. Some of the sessions were entitled: sediment and soil toxicity methods; oil sands development and production; impacts of oil spills and oil clean-up; industrial effluent monitoring; general aquatic toxicity; and regional monitoring frameworks. The workshop featured 142 presentations, of which 27 have been catalogued separately for inclusion in this database.

  1. CYCLIC OXIDATION OF Ti-48%Al-2%Cr-2%Nb-(0~1%)W ALLOYS BETWEEN 800 AND 1000°C IN AIR

    OpenAIRE

    SANG-HWAN BAK; DONG YI SEO; SEON-JIN KIM; JAE CHUN LEE; DONG BOK LEE

    2010-01-01

    Ti-48%Al-2%Cr-2%Nb-(0, 0.5, 1) at.%W alloys were synthesized via the powder metallurgical route, and cyclically oxidized at 800, 900, or 1000°C in air for up to 100 h in order to find the effects of W on their oxidation characteristics. At 800°C, they oxidized relatively slowly, and the scales were thin and adherent. At 900°C, the scales began to spall locally. At 1000°C, they spalled repetitively during oxidation. Cr, Nb, and W improved the cyclic oxidation resistance of TiAl alloys. The oxi...

  2. 46th Annual meeting on nuclear technology. Key topic / outstanding know-how and sustainable innovations

    International Nuclear Information System (INIS)

    Summary report on the following Focus Session of the 46th Annual Conference on Nuclear Technology held in Berlin, 5 to 7 May 2015: Implementing New Safety Requirements in Europe (Christian Raetzke) 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'' and will be covered in further issues of atw.

  3. Special MAFIA postprocessors for the analysis of RF structures

    Science.gov (United States)

    Browman, M. J.

    1992-08-01

    This paper describes three stand-alone programs that use the electromagnetic fields generated by the MAFIA 2.04 codes to analyze radio-frequency (RF) cavities. Illustrations are provided that show how these codes are used to do the following: (1) analyze the effect of the coupling slots on the electric and magnetic fields of the linacs for the APLE Prototype Experiment (APEX) and the Advanced Free-Electron Laser (AFEL); (2) verify the Panofsky-Wenzel theorem for a high-energy deflecting cavity proposed for the Accelerator Transmutation of Waste (ATW) project; and (3) study the effectiveness of that deflecting cavity.

  4. Development of cube textured Ni-W alloy substrates used for coated conductors

    DEFF Research Database (Denmark)

    Suo, Hongli; Ma, Lin; Gao, Mangmang;

    2014-01-01

    It is considered as a challenge for RABiTS route to get cube textured Ni-W alloy substrates with high mechanical and magnetic properties for coated conductors. The works of our group in recent years are summarized about different Ni-W substrates with high W content and composite tapes made by RABi......TS technique. The fabrication process and the mechanism of cube texture formation in above different tapes are studied systematically. Compared with commercial Ni-5 at.%W substrate, these alloy substrates show a comparable texture quality and improved mechanical properties as well as reduced or zero...... magnetization especially in the novel composite substrates....

  5. Real-time simulation of neutron space-time kinetics for high-temperature gas-cooled reactor by ILU-GCR algorithm

    International Nuclear Information System (INIS)

    In the neutron space-time kinetics computation program for High-Temperature Gas-Cooled Reactor, Generalized Conjugate Residual algorithm pretreated by incomplete LU decomposition (ILU-GCR) is used for dealing with the shape function. Compared with classical methods, ILU-CCR algorithm has obvious advantages. For the supposable HTCR model. the variance of the core reactivity, the average neutron flux of each group, the relative power and the temperature along with time are computed for the dynamic simulation of the control rod ejection accident under conditions of over power protection and ATWS. (authors)

  6. Nuclear power: Hour of fog producers; Atomkraft: Stunde der Nebelwerfer

    Energy Technology Data Exchange (ETDEWEB)

    Rauner, M.; Schuh, H.

    2004-03-04

    Seven advanced nuclear power plants in Germany can withstand a frontal crash by a full-tanked Jumbo-Jet. But for five older plants even smaller planes can cause an hazard impossible to control. A fog generation around the power plants, favorized by operators and politicians, to camouflage this plants against terroristic flights is absurd because of the possibility of flight automation. However terrorists may attack reactors also from the ground, but how they can do is top secret. (GL) [German] Einem frontalen Aufprall eines voll getankten Jumbo-Jets mit hoher Geschwindigkeit koennen in Deutschland sieben moderne Kernkraftwerke standhalten. Bei fuenf aelteren Modellen kann dagegen selbst ein kleineres Flugzeug ein schwer beherrschbares Unglueck ausloesen. Obwohl Experten eine vorzeitige Schliessung solch alter Meiler fuer realistisch halten, ist die Verhandlungslage aufgrund ideologischer Einfluesse verfahren. Die von Betreibern und Politikern gern gepriesenen Nebelwerfer zur Tarnung der Kraftwerke sind wegen der Moeglichkeit des Instrumentenfluges sinnlos. Die beste Abwehr bieten vorgelagerte Schutzbauten aus Beton. Jedoch koennen Terroristen Reaktoren auch vom Boden aus gefaehrden, wie, ist natuerlich geheim.

  7. Comparison calculation/experiment on the load case ``shutdown of TH high pressure pumps under consideration of fluid structure interaction``; Vergleich Rechnung/Messung zum Lastfall ``Abschaltung der TH-Hochdruckpumpen unter Beruecksichtigung der Fluid-Struktur-Wechselwirkung``

    Energy Technology Data Exchange (ETDEWEB)

    Erath, W.; Nowotny, B.; Maetz, J. [KED, Rodenbach (Germany)

    1998-11-01

    Measurements of an experiment in a pipe system with pump shutdown and valve closing have been performed in the nuclear power plant KRB II. Comparative calculations of fluid and structure including interaction show an excellent agreement with the measured results. Theory and implementation of the fluid/structure interaction and the results of the comparison are described. It turns out that the consideration of the fluid/structure interaction is mostly a significant increase of the effective structural damping. (orig.) [Deutsch] Es wurden Messungen am nuklearen Nachkuehlsystem des Kernkraftwerks Gundremmingen (KRB II) bei einem Versuche mit Pumpenabschalten und Ventilschliessen durchgefuehrt. Vergleichsrechnungen der Fluid-Strukturdynamik unter echter Beruecksichtigung der Wechselwirkung ergaben eine ausgezeichnete Uebereinstimmung der Rechnung mit den Messungen. Es werden Theorie und Implementierung der Koppelung der Fluid- und Struktur-Berechnungen sowie die Vergleiche von Messung und Rechnung beschrieben. Es ergibt sich, dass die Beruecksichtigung der Wechselwirkung notwendig ist zur genaueren Berechnung von `weichen` Rohrleitungsystemen. Eine wichtige Folge der Wechselwirkung ist meist eine deutliche Erhoehung der effektiven Strukturdaempfung. (orig.)

  8. The man-machine-organisation interface; Schnittstelle Mensch-Technik-Organisation

    Energy Technology Data Exchange (ETDEWEB)

    Kociok, B. [Bundesamt fuer Strahlenschutz, Salzgitter (Germany)

    1999-04-01

    The reliable and safety-centred action of man is one crucial factor for safe operation of nuclear power plants, apart from the technical systems and the organisational aspects. Essential factors influencing human performance are: Qualification and competence of the operating personnel, technical conditions and status of systems, including the level of automation, information technology in the control room, and plant organisation. Analyses of documentation of notifiable events in power plant operation or other incidents yield information on available potentials for safety enhancements or reduction of human errors. (orig./CB) [German] Das zuverlaessige und sicherheitsorientierte Handeln des Menschen ist neben den technischen Einrichtungen und der Organisation fuer den sicheren Betrieb der Kernkraftwerke von entscheidender Bedeutung. Wesentliche Einflussfaktoren auf menschliche Handlungen sind: - Die Personalqualifikation, - der technische Zustand der Anlage, einschliesslich ihres Automatisierungsgrades, - die Gestaltung der Warte und - die Betriebsorganisation. Aus der Erfassung von meldepflichtigen und sonstigen Ereignissen und deren Analyse lassen sich Moeglichkeiten fuer Sicherheitsverbesserungen ermitteln und das Auftreten von menschlichen Fehlhandlungen weiter reduzieren. (orig.)

  9. Mechanized radiation testing of austenitic pipe welds. Testing of media filled pipes and determination of the flaw depth by tomosynthesis; Mechanisierte Durchstrahlungspruefung von Rundschweissnaehten. Pruefung mediengefuellter Rohrleitungen und Tiefenlagenbestimmung durch Tomosynthese

    Energy Technology Data Exchange (ETDEWEB)

    Ewert, U.; Redmer, B. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany); Mueller, J. [COMPRA GmbH, Frechen (Germany); Trobitz, M. [Kernkraftwerke Gundremmingen Betriebsgesellschaft mbH, Gundremmingen (Germany); Baranov, V.A. [Institute for Introscopy, Tomsk (Russian Federation)

    1999-08-01

    A compact detection system was built for multi-angle inspection of pipes, consisting of a high-sensitivity radiometric line scanner and an ultrasonic manipulator. Improved flaw imaging quality is achieved with this system as compared to film radiography. Measurements have been carried out on site in a nuclear power plant and in a laboratory. Better flaw imaging quality was also achieved in the testing of water-filled pipes. Non-linear tomosynthesis was applied for processing and interpretation of measured data. The system delivers considerably better images of planary materials inhomogeneitites, (such as cracks and lack-of-bond defects). (orig./CB) [Deutsch] Eine hoch empfindliche radiometrische Zeilenkamera wurde mit einem Ultraschall-Manipulator zu einem Gesamtsystem aufgebaut und fuer Mehrwinkel-Inspektionen von Rohrleitungen angewandt. Bei der Inspektion von Rundschweissnaehten an Rohren mit ca. 8... 20 mm Wanddicke wurde eine Verbesserung der Bildqualitaet im Vergleich zur Filmradiographie erreicht. Diese Messungen wurden in einem Kernkraftwerk unter Vor-Ort-Bedingungen sowie im Labor ausgefuehrt. Ein signifikantes Ansteigen der Bildqualitaet wurde auch bei der Pruefung von wassergefuellten Rohren erzielt. Methoden der nicht-linearen Tomosynthese wurden fuer die Verarbeitung und Interpretation der gemessenen Projektionsdaten genutzt. Das entwickelte System gestattet eine erhebliche Verbesserung der Anzeige von planaren Materialinhomogenitaeten (z.B. Risse und Bindefehler). (orig./DGE)

  10. Information policy of a Swiss nuclear power plant; Information der Medien und der Oeffentlichkeit durch den Anlagenbetreiber

    Energy Technology Data Exchange (ETDEWEB)

    Erne, L. [Kernkraftwerk Leibstadt (Switzerland)

    1997-12-31

    The Leibstadt NPP is situated on the river Rhine between Schaffhausen and Basel. It is the policy of the Leibstadt Plant Management to inform the public and the media in a timely and accurate manner during normal operation as well as in the case of unforeseen eventualities. We also inform the authorities of the communities in Southern Germany in the vicinity of the plant. Detailed procedures have been set up to provide a high standard of media service. The PR group is part of the emergency organisation and is supported by other technical groups. Modern communication technology, i.e. fax, teletext and internet are available. Our experience has shown that the early stages of an incident can be very hectic. The relevant information has to be collected internally and at the same time the media has to be presented with accurate communiques. (orig.) [Deutsch] Das Kernkraftwerk Leibstadt, gelegen auf der Schweizer Seite des Hochrheins, bekennt sich zu einer raschen, aktiven, transparenten und grenzueberschreitenden Informationspolitik sowohl im Normalbetrieb wie bei ungeplanten Ereignissen. Das Informationsverhalten ist festgelegt in detaillierten Vorschriften. Die Informationsstelle ist eingebunden in die Notfallorganisation mit Unterstuetzung von entsprechenden Dienstgruppen. In der Praxis nutzt das Werk moderne Kommunikationsmittel wie Telefax, Teletext oder neu Internet. Erfahrungen zeigen, dass in der Startphase eines Ereignisses der Zeitdruck enorm ist, waehrend intern gleichzeitig zuerst gesicherte Informationen zu beschaffen sind. (orig.)

  11. Development of the Diverse Means for Reactor Shutdown Function of EU-APR1400

    International Nuclear Information System (INIS)

    This paper provides general descriptions of the EBS focusing on basic design characteristics such as system function, configuration and operation, and presents results from the preliminary verification of system performance. The diverse means for the reactor shutdown function of EU-APR1400 have been developed to comply with the diversity principle of the European design requirements of a new nuclear power plant. The preliminary verification of the EBS performance was done by the ATWS analysis. The analysis results show that the EBS was designed properly. Diversity is the fundamental principle in safety system design of a new nuclear power plant, which uses different mitigation measures to provide diverse ways of responding to a significant event. Regarding the diversity principle, EU-APR1400 (European APR1400) safety system should be in accordance with European design requirements. EBS (Emergency Boration System) is designed to provide the diverse means to shut down the reactor against ATWS (Anticipated Transient Without Scram) and to mitigate the event consequences in the EUAPR1400

  12. Status of development of actinide blanket processing flowsheets for accelerator transmutation of nuclear waste

    International Nuclear Information System (INIS)

    An accelerator-driven subcritical nuclear system is briefly described that transmutes actinides and selected long-lived fission products. An application of this accelerator transmutation of nuclear waste (ATW) concept to spent fuel from a commercial nuclear power plant is presented as an example. The emphasis here is on a possible aqueous processing flowsheet to separate the actinides and selected long-lived fission products from the remaining fission products within the transmutation system. In the proposed system the actinides circulate through the thermal neutron flux as a slurry of oxide particles in heavy water in two loops with different average residence times: one loop for neptunium and plutonium and one for americium and curium. Material from the Np/Pu loop is processed with a short cooling time (5-10 days) because of the need to keep the total actinide inventory, low for this particular ATW application. The high radiation and thermal load from the irradiated material places severe constraints on the separation processes that can be used. The oxide particles are dissolved in nitric acid and a quarternary, ammonium anion exchanger is used to extract neptunium, plutonium, technetium, and palladium. After further cooling (about 90 days), the Am, Cm and higher actinides are extracted using a TALSPEAK-type process. The proposed operations were chosen because they have been successfully tested for processing high-level radioactive fuels or wastes in gram to kilogram quantities

  13. Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    H. Yamano

    2012-01-01

    Full Text Available As a next-generation plant, a large-scale Japan sodium-cooled fast reactor (JSFR adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are a passive reactor shutdown system and a recriticality-free concept against anticipated transients without scram (ATWS in design extension conditions (DECs. A fully passive decay heat removal system with natural circulation is also introduced for design-basis events (DBEs and DECs. In this paper, the safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system against a typical ATWS. Severe accident analysis supported by safety experiments and phenomenological consideration led to the feasibility of in-vessel retention without energetic recriticality. Moreover, a probabilistic safety assessment indicated to satisfy the risk target.

  14. Disposition of nuclear waste using subcritical accelerator-driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

    1998-12-31

    Spent fuel from nuclear power plants contains large quantities of Pu, other actinides, and fission products (FP). This creates challenges for permanent disposal because of the long half-lives of some isotopes and the potential for diversion of the fissile material. Two issues of concern for the US repository concept are: (1) long-term radiological risk peaking tens-of-thousands of years in the future; and (2) short-term thermal loading (decay heat) that limits capacity. An accelerator-driven neutron source can destroy actinides through fission, and can convert long-lived fission products to shorter-lived or stable isotopes. Studies over the past decade have established that accelerator transmutation of waste (ATW) can have a major beneficial impact on the nuclear waste problem. Specifically, the ATW concept the authors are evaluating: (1) destroys over 99.9% of the actinides; (2) destroys over 99.9% of the Tc and I; (3) separates Sr-90 and Cs-137; (4) separates uranium from the spent fuel; (5) produces electric power.

  15. Development of 4S and related technologies. (3) Statistical evaluation of safety performance of 4S on ULOF event

    International Nuclear Information System (INIS)

    The purpose of this work is to evaluate quantitatively and statistically the safety performance of Super-Safe, Small, and Simple reactor (4S) by analyzing with ARGO code, a plant dynamics code for a sodium-cooled fast reactor. In this evaluation, an Anticipated Transient Without Scram (ATWS) is assumed, and an Unprotected Loss of Flow (ULOF) event is selected as a typical ATWS case. After a metric concerned with safety design is defined as performance factor a Phenomena Identification Ranking Table (PIRT) is produced in order to select the plausible phenomena that affect the metric. Then a sensitivity analysis is performed for the parameters related to the selected plausible phenomena. Finally the metric is evaluated with statistical methods whether it satisfies the given safety acceptance criteria. The result is as follows: The Cumulative Damage Fraction (CDF) for the cladding is defined as a metric, and the statistical estimation of the one-sided upper tolerance limit of 95 percent probability at a 95 percent confidence level in CDF is within the safety acceptance criterion; CDF < 0.1. The result shows that the 4S safety performance is acceptable in the ULOF event. (author)

  16. Disposition of nuclear waste using subcritical accelerator-driven systems

    International Nuclear Information System (INIS)

    Spent fuel from nuclear power plants contains large quantities of Pu, other actinides, and fission products (FP). This creates challenges for permanent disposal because of the long half-lives of some isotopes and the potential for diversion of the fissile material. Two issues of concern for the US repository concept are: (1) long-term radiological risk peaking tens-of-thousands of years in the future; and (2) short-term thermal loading (decay heat) that limits capacity. An accelerator-driven neutron source can destroy actinides through fission, and can convert long-lived fission products to shorter-lived or stable isotopes. Studies over the past decade have established that accelerator transmutation of waste (ATW) can have a major beneficial impact on the nuclear waste problem. Specifically, the ATW concept the authors are evaluating: (1) destroys over 99.9% of the actinides; (2) destroys over 99.9% of the Tc and I; (3) separates Sr-90 and Cs-137; (4) separates uranium from the spent fuel; (5) produces electric power

  17. Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    As a next-generation plant, a large-scale Japan sodium-cooled fast reactor (JSFR) adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are a passive reactor shutdown system and a recriticality-free concept against anticipated transients without scram (ATWS) in design extension conditions (DECs). A fully passive decay heat removal system with natural circulation is also introduced for design-basis events (DBEs) and DECs. In this paper, the safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system against a typical ATWS. Severe accident analysis supported by safety experiments and phenomenological consideration led to the feasibility of in-vessel retention without energetic recriticality. Moreover, a probabilistic safety assessment indicated to satisfy the risk target.

  18. Level-2 PSA for the Prototype Fast Breeder Reactor MONJU Applied to the Accident Management Review

    International Nuclear Information System (INIS)

    JNES independently evaluated the three events it selected - PLOHS, LORL and ATWS events - and reviewed the results of the Level 2 PSA carried out by JAEA. Regarding ATWS events, the organization carried out a qualitative evaluation of the results of JAEA's evaluation and carried out a quantitative evaluation of the containment failure frequency (CFF) in relation to PLOHS and LORL events. In JNES's independent evaluation of PLOHS and LORL events, accident scenarios in the three phases - the plant response phase, the core damage phase and the containment vessel response phase - were analyzed. The phenomenal event trees were quantified by applying the information about phenomena specific to fast reactors, including plant thermal-hydraulic analysis at the time of core damage, boundary structure analysis, analysis of the characteristics of the disrupted core, the results of sodium-concrete reaction tests, and the results of hydrogen diffusion induced combustion tests, to the PRDs. As the result, the total CFF before the preparation of the AM measures was rated at 9.2E-9/reactor year (CDF at 2.7E-7/reactor year), and it has been confirmed that these numerical values are well below the power reactor performance goal indicator values (CDF: 10-4/year or so; CFF: 10-5/year or so) even before the preparation of the AM measures. (author)

  19. The Clinical Use of Left Ventricular Twist Degree in Chronic Heart Failure Subjects by Three-dimensional Ultrasound Speckle Tracking Imaging%三维斑点追踪技术在慢性心力衰竭患者左室扭转运动中的应用研究

    Institute of Scientific and Technical Information of China (English)

    张艳丽; 王小丛; 赵丽荣; 装莉平; 于微

    2012-01-01

    Objective This study was performed to assess left ventricle twist degree in patients with chronic heart failure by three-dimensional ultrasound speckle tracking imaging. Methods The apical 4-chamber and 2-chamber views were acquired in thirty-two patients with chronic heart failure and thirty-three healthy volunteers .using 3D-trace software to measure values of left ventricle end-diastolic volumes (LVEDV) , end-systolic volumes(LVESV) ,left ventricular ejection fraction (LVEF),basal segment twist degree(BTW),middle segment twist degree(MTW) .apical segment twist degree( ATW) ,left ventricular global twist degree(LVTW). Values were compared in two groups, the correlations between LVEF and LVTW,BTW,MTW, ATW were analyzed respectively. Results LVEF,LVTW,MTW, ATW in CHF patients were lower than the control group .the correlations between BTW,MTW, ATW,LVTW and LVEF were found (0. 557,0. 926,0. 932,0. 945. P<0. 01 for all). Conclusions The left ventricular function was impaired in patients with CHF. The left ventricular twist can be studied by three-dimensional ultrasound speckle tracking imaging, which would be a new tool for the evaluation of left ventricular systolic function.%目的 应用三维斑点追踪显像技术研究慢性心力衰竭患者(CHF)左室扭转的运动特征,探讨其临床价值.方法 CHF组患者30例,年龄匹配的健康志愿者(对照组)33例,采集标准的四腔心及两腔心切面,进行全容积图像存储,应用3D-trace软件进行脱机分析,软件自动分析计算左心室舒张末期容积(LVEDV),左室收缩末期容积(LVESV),左室射血分数(LVEF),左室基底段收缩期扭转角度峰值(BTW),中间段收缩期扭转角度峰值(MTW),心尖段收缩期扭转角度峰值(ATW),左室整体收缩期扭转角度峰值(LVTW).结果 CHF组LVEF,LVTW,MTW,ATW均较正常组减低,BTW、MTW、ATW、LVTW与LVEF之间有明显的相关性,相关系数分别为0.557,0.926,0.932,0.945.结论 CHF患者左心收缩功能明显降

  20. Cube-textured metal substrates for reel-to-reel processing of coated conductors

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, A.C.

    2012-09-15

    This thesis presents the results of a study aimed at investigating important fabrication aspects of reel-to-reel processing of metal substrates for coated conductors and identifying a new substrate candidate material with improved magnetic properties. The effect of mechanical polishing on surface roughness and texture in Ni-5at.%W tapes in the cold-rolled condition was studied as a function of polishing grade. The surface roughness of the tape in the polished and annealed condition, and after subsequent coating with a Gd{sub 2}Zr{sub 2}O{sub 7} buffer layer was investigated taking grain boundaries into account. It was observed that the initial mean surface roughness decreased after annealing except after very fine polishing. Additionally, the roughness of the buffer layers were found to increase slightly for the fine polished substrates. Grain boundary grooving was observed to impose a lower limit for the mean surface roughness. Fractions of cube texture within deviations of 5 deg. from the ideal cube orientation, in the annealed substrates, were found to be very sensitive to the surface roughness before annealing. Microstructure, texture and topography were studied in a strongly cube-textured Ni-5at.%W substrate before and after an additional annealing (condition A1 and A2, respectively) simulating a buffer layer crystallisation heat treatment. Condition A1 was characterised by a high fraction of cube texture, a high fraction of low angle grain boundaries and a low fraction of {Sigma}3 boundaries. A strong correlation was observed between the grain boundary groove depth and boundary type. Coherent twin boundaries and low angle grain boundaries were characterised by the smallest average groove depth while significantly deeper grooves were observed at other boundary types. A similar correlation was observed between the inclination angle at groove walls and the boundary type. The microstructure was slightly coarser in condition A2 and it was accompanied by a cube

  1. Electronic properties of interfaces produced by silicon wafer hydrophilic bonding

    Energy Technology Data Exchange (ETDEWEB)

    Trushin, Maxim

    2011-07-15

    The thesis presents the results of the investigations of electronic properties and defect states of dislocation networks (DNs) in silicon produced by wafers direct bonding technique. A new insight into the understanding of their very attractive properties was succeeded due to the usage of a new, recently developed silicon wafer direct bonding technique, allowing to create regular dislocation networks with predefined dislocation types and densities. Samples for the investigations were prepared by hydrophilic bonding of p-type Si (100) wafers with same small misorientation tilt angle ({proportional_to}0.5 ), but with four different twist misorientation angles Atw (being of < , 3 , 6 and 30 , respectively), thus giving rise to the different DN microstructure on every particular sample. The main experimental approach of this work was the measurements of current and capacitance of Schottky diodes prepared on the samples which contained the dislocation network at a depth that allowed one to realize all capabilities of different methods of space charge region spectroscopy (such as CV/IV, DLTS, ITS, etc.). The key tasks for the investigations were specified as the exploration of the DN-related gap states, their variations with gradually increasing twist angle Atw, investigation of the electrical field impact on the carrier emission from the dislocation-related states, as well as the establishing of the correlation between the electrical (DLTS), optical (photoluminescence PL) and structural (TEM) properties of DNs. The most important conclusions drawn from the experimental investigations and theoretical calculations can be formulated as follows: - DLTS measurements have revealed a great difference in the electronic structure of small-angle (SA) and large-angle (LA) bonded interfaces: dominating shallow level and a set of 6-7 deep levels were found in SA-samples with Atw of 1 and 3 , whereas the prevalent deep levels - in LA-samples with Atw of 6 and 30 . The critical twist

  2. Proposal for geological site selection for L/ILW and HLW repositories. Justification of waste allocation, barrier concept and requirements on geology. Report on safety and technical feasibility. Technical report 08-05; Vorschlag geologischer Standortgebiete fuer das SMA- und das HAA-Lager. Begruendung der Abfallzuteilung, der Barrierensysteme und der Anforderungen an die Geologie. Bericht zur Sicherheit und technischen Machbarkeit. Technischer Bericht 08-05

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-10-15

    production, content of potentially gas-producing components and content of complexants also have to be considered. The waste is divided into the categories high-level waste (HLW), alpha-toxic waste (ATW) and low- and intermediate-level waste (L/ILW). HLW differs significantly from ATW and L/ILW. For this reason HLW is disposed of in a separate repository with a specifically designed barrier system. The ATW and L/ILW differ in terms of specific radiotoxicity, specific activity and specific heat production. However, many of their other properties are very similar, particularly the material inventory. A combined repository for all ATW and L/ILW constructed in a suitable host rock in a favourable geological setting has the potential to fulfil the safety requirements. Calculated doses are dominated by just a few of the ATW and L/ILW waste types. If these dominant waste types could be disposed of elsewhere, the requirements on the geology could be reduced while the level of safety would remain the same. The existing concept: a HLW repository with a facility for long-lived intermediate-level waste (ILW) and a L/ILW repository, has been maintained, with the aim of allocating the dose-dominating ATW and L/ILW to the ILW facility. Nagra's proposal includes two variants, characterised by minimum requirements on the large-scale hydraulic conductivity of the host rock for the L/ILW repository of 10{sup -10} m/s and 10{sup -9} m/s respectively. The volume of waste allocated to the L/ILW repository is somewhat smaller for the 10{sup -9} m/s variant than for the 10{sup -10} m/s variant. All the ATW is allocated to the HLW repository (ILW facility). The safety concept shows how the different engineered and geological barriers contribute to system safety and what safety functions they perform. In the selected safety concept, both the engineered and the geological barriers contribute significantly to the barrier function of the overall system. The concept also describes the contribution

  3. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Wigeland; J. E. Cahalan

    2009-12-01

    Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

  4. Planning and reporting of Russian transmutation research projects within ISTC. Phase 2

    Energy Technology Data Exchange (ETDEWEB)

    Conde, H. [Uppsala Univ. (Sweden). Dept. of Neutron Research; Gudowski, W. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Reactor and Neutron Physics; Liljenzin, J.O. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Chemistry; Mileikovsky, C. [Pully (Switzerland)

    1998-11-01

    The present report about phase 2 of the SKI project on Planning and Reporting of Russian Transmutation Research Projects within ISTC is an update of the information given in the SKI report no 97:15 (Feb 1997) about phase 1 of the same project. The background information is partly repeated in the present report to avoid that the reader has to go back to the report of Phase 1 for information about the basis for the project. USA, EU, Japan, Republic of Korea and Norway are at present supporting the International Scientific and Technical Center (ISTC) in Moscow. The Centre gives funds to research projects of civilian interest to former nuclear weapon laboratories to counteract the risk of nuclear weapon proliferation by the emigration of former USSR technical and scientific experts to `border countries` which are aiming towards the development of nuclear weapons. Before Sweden and Finland entered the EU, both countries gave national support to ISTC, in the case of Sweden 4 MUSD. Some of the projects which were funded by the Swedish national support to ISTC are still in progress. Nuclear technical concepts (i.e. Accelerator Transmutation of Nuclear Waste, ATW) have been proposed to incinerate and transmute long-lived radioactive nuclear waste to relax the time needed to store the waste in a geological repository. The named Russian experts are knowledgeable and well equipped of doing research in the different technical fields of relevance for the transmutation concepts. Thus, a number of ISTC projects have been initiated, and further ones have been proposed, to investigate different technical aspects of ATW with a result that a fair number of former weapon specialists have converted from military to peaceful civilian research. A similar centre STCU (The Scientific and Technical Centre of the Ukraine) has been set up in Kiev. Sweden has been active in promoting this Centre, which is supported by USA, Japan, Canada and recently also by EU. The present report describes the

  5. Methodological aspects of probabilistic analysis and risk management: potentials and limits of an integrated risk assessment strategy encompassing various energy sources for electricity generation; Konzeptionelle Fragen der probabilistischen Analyse und des Risikomanagements: Moeglichkeiten und Grenzen einer Gesamtstrategie zur Risikovorsorge beim Einsatz unterschiedlicher Energietraeger zur Stromversorgung

    Energy Technology Data Exchange (ETDEWEB)

    Koeberlein, K.

    1997-08-01

    The typical risks and impacts, connected with electrical power supply, - accidental risk from nuclear power plants, - local and regional environmental impacts from fossil power plants, - potential damages caused by a global climate change, show characteristics which render a mutual comparison on a common nominator very difficult if not impossible: (a) Accidental risks from nuclear power plants are characterized by very low probabilities of occurence and - potentially - very large damage extent. Applying the technology with due care it can be expected that the risks - which can be evaluated theoretically - even over long periods of time will not be realized in actually occuring damages. (b) The pollution caused by burning of fossil fuels can be measured and is causing damages which can - in part directly - be proven. The extent (and the probability) of hidden damages can be evaluated based on epidemiological analyses or by means of dose-effect relations evaluated experimentally. (c) Risks from a global climate change can not be quantified reliably up to now. Damages are expected to be unavoidable, if massive countermeasures are not taken very soon. The potential damage extent has been estimated very coarsely, but reliable insights do not exist. An approach which pricipally could bring different kinds of risk to a common denominator is the evaluation of external costs of electrical power supply systems. However, even using this instrument, accidental risks from nuclear power plants and potential damages from a global climate change can not (or not yet) be included into a quantitative risk comparison. (orig./DG) [Deutsch] Die typischen, mit der Stromversorgung verbundenen Risiken und Belastungen, wie - Unfallrisiken durch Kernkraftwerke, - lokale und regionale Umweltbelastungen durch fossil befeuerte Kraftwerke, - moegliche Schaeden durch eine globale Klimaaenderung, weisen Charakteristiken auf, die einen gegenseitigen Vergleich auf `gleichem Nenner` sehr erschweren

  6. Reactor safety research against the backdrop of the Energy-Omnibus Law; Reaktorsicherheits-Forschung. Vor dem Hintergrund des Energie-Artikelgesetzes

    Energy Technology Data Exchange (ETDEWEB)

    Kuczera, B. [Projekt Nukleare Sicherheitsforschung, Forschungszentrum Karlsruhe GmbH (Germany)

    1995-05-01

    On July 19, 1994, the German Federal Parliament adopted the Coal/Nuclear Power Omnibus Law, in which a new quality of safety of future nuclear power plants has been laid down. The defense-in-depth safety concept underlying the nuclear power plants currently in operation is derived from the principle of safety precautions made against reactor accidents, and encompasses preventive measures of accident mangement and mitigating measures of containing possible consequences. Accident management leads to the requirement that even in the most unlikely accidents with core meltdown the consequences remain limited to the plant. A new quality in reactor safety is represented by the System 80+ advanced pressurized water reactor and by the European Pressurized Water Reactor, EPR. Despite different views about the approaches used to address individaul aspects in the achievement of safety goals, there is agreement on the principle that risk provisions, by achieving more transparency, are to result in better public acceptance of the peaceful uses of nuclear power. (orig.) [Deutsch] Am 19.7.94 hat der Deutsche Bundestag das Artikelgesetz Kohle/Kernenergie verabschiedet, in dem eine neue Qualitaet bei der Sicherheit von zukuenftigen Kernkraftwerken festgeschrieben ist. Das Defense-in-Depth-Sicherheitskonzept fuer die heute betriebenen Kernkraftwerke leitet sich aus dem Prinzip der Sicherheitsvorsorge gegen Reaktorstoerfaelle ab und umfasst praeventive Massnahmen zur Beherrschung von Stoerfaellen und mitigative Massnahmen zur Eingrenzung von moeglichen Folgen. Das Accident Management fuehrt zu der Forderung, dass selbst bei unwahrscheinlichsten Unfaellen mit Kernschmelzen die Schadensfolgen auf die Anlage beschraenkt bleiben. Eine neue Qualitaet in der Reaktorsicherheit stellen z.B. der fortgeschrittene Druckwasserreaktor System 80{sup +} und der europaeische Druckwasserreaktor EPR dar. Wenn auch Loesungsansaetze zu Einzelaspekten bei der Umsetzung von Sicherheitszielen

  7. Analysis by regulatory safety criteria of notifiable events in Bavarian nuclear power plants; Analyse meldepflichtiger Ereignisse in bayerischen Kernkraftwerken mithilfe von aufsichtlichen Sicherheitsindikatoren

    Energy Technology Data Exchange (ETDEWEB)

    Seidel, E.R. [Bayerisches Staatsministerium fuer Landesentwicklung und Umweltfragen, Muenchen (Germany); Straub, G. [TUeV Sueddeutschland Bau und Betrieb GmbH, Muenchen (Germany)

    2001-03-01

    The holistic analysis of notifiable events by the regulatory authority on the basis of some selected safety indicators, and evaluation of the results, furnish important findings about the safety-related importance of the events examined, and about the current safety status of the plant and its mode of operation. The regulatory evaluation process by far exceeds a mere assessment of causes and consequences of an event and of the measures taken by the operator to remedy the consequences. Other factors investigated seek to determine, on the basis of the current state of the art, whether there are weak spots in plant design, organization, and human factors requiring thorough improvement on a medium or longer term. The evaluations carried out of individual safety criteria thus constitute the basis of further studies and investigations by the regulatory authority and of any improvements which may be ordered. (orig.) [German] Die ganzheitliche Analyse meldepflichtiger Ereignisse durch die Aufsichtsbehoerde an Hand ausgewaehlter Sicherheitsindikatoren und die Auswertung der Ergebnisse liefern wichtige Aussagen ueber die sicherheitstechnische Bedeutung der Ereignisse und den aktuellen Sicherheitszustand des Kernkraftwerks sowie seine Betriebsweise. Der aufsichtliche Bewertungsprozess geht dabei weit ueber die Betrachtung der Ursache und der Auswirkungen des Ereignisses sowie der Massnahmen des Betreibers zur Behebung der Ereignisfolgen hinaus. Vielmehr wird auch geprueft, inwieweit auf der Grundlage des aktuellen Standes von Wissenschaft und Technik Schwaechen in der Anlagenauslegung, bei der Organisation und bei menschlichen Faktoren vorliegen, die mittel- bis laengerfristig nachhaltige Verbesserungen erfordern. Die fuer die einzelnen Sicherheitsindikatoren ermittelten Bewertungen bilden daher die Grundlage fuer weitere Untersuchungen und Pruefungen der Aufsichtsbehoerde sowie die evtl. notwendige Anordnung von Verbesserungen. (orig.)

  8. Nuclear reactors: Notifiable events in 2002; Meldepflichtige Ereignisse 2002

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2003-06-01

    Notifiable events in nuclear power plants in the Federal Republic of Germany are reported to the regulatory authorities under the Atomic Energy Act in accordance with standardized national reporting criteria, and are recorded centrally. The binding legal provisions covering these reports can be found in the Nuclear Safety Commissioner and Reporting Ordinance (AtSMV). On an international level, events are classified in the International Nuclear Event Scale (INES) comprising eight levels. The four quarterly reports covering 2002 include 167 notifiable events for nuclear power plants in operation and in the decommissioning stage. Of these events, 157 are in reporting category N (normal), while ten are in reporting category E (urgent). No events have been reported in category S (immediate). 154 events are INES level 0, 13 events are INES level 1. 13 category-N events were reported for research reactors. All of them are INES level 0. There were no releases of radioactive material above the licensed levels for ex-vent air and liquid effluents. (orig.) [German] Meldepflichtige Ereignisse in Kernkraftwerken in der Bundesrepublik Deutschland werden gemaess bundeseinheitlichen Meldekriterien an die atomrechtlichen Aufsichtsbehoerden gemeldet und zentral erfasst. Rechtsverbindlich sind sie in der Atomrechtlichen Sicherheitsbeauftragten- und Meldeverordnung AtSMV niedergelegt. International werden Ereignisse der insgesamt acht Stufen umfassenden ''International Nuclear Event Scale'' zugeordnet. Nach den vorliegenden Quartalsberichten fuer das Jahr 2002 wurden 167 meldepflichtige Ereignisse fuer Kernkraftwerke (in Betrieb und in Stillegung) mitgeteilt. Von diesen sind 157 der Meldekategorie N (Normalmeldung) und 10 der Meldekategorie E (Eilmeldung) zugeordnet. Es sind keine Ereignisse der Kategorie S (Sofortmeldung) zu verzeichnen. Der INES-Sufe 0 sind 154, der Stufe 1 13 Ereignisse zugeordnet. Fuer Forschungsreaktoren wurden 13 Ereignisse der Kategorie N

  9. 1997: notifiable events; 1997: Meldepflichtige Ereignisse. 117 gemeldete Ereignisse aus deutschen Kernkraftwerken und zwoelf aus Forschungsreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1998-07-01

    In May 1998, the German Federal Ministry for the Environment, Nature Conservation, and Rector Safety (BMU) presented the 1997 survey of `Notifiable events in plants for nuclear fuel fission - nuclear power and research reactors whose maximum power exceeds 50 kW of continuous thermal power - in the Federal Republic of Germany`. Since 1975, the operators of nuclear power plants in the Federal Republic of Germany have been required to report to the nuclear supervisory authorities all notifiable events in accordance with standard national reporting criteria. This official reporting system serves for monitoring the safety status of notifiable plants and use the findings derived from the events reported to improve the safety status of plants within the supervisory procedures where necessary. The reports constitute an important base for the early detection of defects and for preventing the occurrence of similar defects in other plants. In 1997, there were 117 notifiable events in nuclear power plants in the Federal Republic of Germany. None of these events is to be classified as an accident, and in none of the events were dose limits under the German Radiation Protection Ordinance exceeded. (orig.) [Deutsch] Ende Mai 1998 legte das Bundesministerium fuer Umwelt, Naturschutz und Reaktorsicherheit (BMU) die Uebersicht des Jahres 1997 ueber `Meldepflichtige Ereignisse in Anlagen zur Spaltung von Kernbrennstoffen - Kernkraftwerke und Forschungsreaktoren, deren Hoechstleistung 50 kW thermische Dauerleistung ueberschreitet - der Bundesrepublik Deutchland` vor. Die Meldungen stellen eine wesentliche Basis fuer die fruehzeitige Erkennung etwaiger Maengel ebenso wie fuer die Vorbeugung gegen Auftreten aehnlicher Fehler in anderen Anlagen dar. 1997 traten 117 meldepflichtige Ereignisse in Kernkraftwerken der Bundesrepublik Deutschland auf. Bei keinem der gemeldeten Ereignisse traten Abgaben radioaktiver Stoffe oberhalb genehmigter Grenzwerte auf. Eine Gefaehrdung von Personen und

  10. New elements in Baden-Wuerttemberg's nuclear power oversight; Neue Elemente in der baden-wuerttembergischen Kernenergieaufsicht

    Energy Technology Data Exchange (ETDEWEB)

    Winter, U. [Ministerium fuer Umwelt und Verkehr Baden-Wuerttemberg, Stuttgart (Germany)

    2004-07-01

    In the wake of the notifiable events at unit 2 of the Philippsburg Nuclear Power Station in the autumn of 2001 (insufficient boration and filling levels in the flooding tanks of the emergency core cooling and residual heat removal system), criticism had been levelled also against the Baden-Wuerttemberg regulatory authority with the state Ministry for the Environment and Transport. As a consequence, the authority was subjected to a number of external audits, some of them initiated in-house, others launched externally. The outcome of these audits, and optimization and development processes in nuclear oversight in Baden-Wuerttemberg, both as a consequence of the investigations and of in-house initiatives, are outlined. The measures initiated also serve to meet current challenges a regulatory authority is facing as a result of deregulation of the electricity market and opt-out of the use of nuclear power. (orig.) [German] Nach den meldepflichten Ereignissen im Kernkraftwerk Philippsburg, Block 2, vom Herbst 2001 (Unterborierung und Fuellstandsunterschreitung in Flutbehaeltern des Not- und Nachkuehlsystems) war auch die banden-wuerttembergische Aufsichtsbehoerde beim Ministerium fuer Umwelt und Verkehr Baden-Wuerttemberg in die Kritik geraten. Die Aufsichtsbehoerde wurde daraufhin gleich mehreren externen Ueberpruefungen - teils selbst initiiert, teils von aussen veranlasst - unterzogen. Ergebnisse der Ueberpruefungen sowie Optimierungsprozesse und Weiterentwicklungen in der Kernenergieaufsicht in Baden-Wuerttemberg, sowohl aufgrund der Untersuchungen, als auch aufgrund eigener Initiative, werden behandelt. Ziel ist es auch, mit dem Buendel von Massnahmen fuer die aktuellen Herausforderungen, die Strommarktliberalisierung und Atomausstieg an eine Aufsichtsbehoerde stellen, geruestet zu sein. (orig.)

  11. Risk reduction category (RRC-A) accident studies in the safety analysis report of the EPR trademark reactor

    Energy Technology Data Exchange (ETDEWEB)

    Poehlmann, M.; Bleher, G.; Ismaier, A.; Knoll, A.; Levi, P.; Garcia, E. Vera; Schels, A.; Seitz, H.; Lima Campos, L. [AREVA GmbH, Erlangen (Germany)

    2013-07-01

    The Risk Reduction Category (RRC-A) is considered in the safety demonstration of nuclear reactors in addition to design basis operating conditions (Plant Condition Category, PCC), in order to analyze with a risk reduction approach any operating conditions with multiple failures. As extending the operating conditions of the plant 'beyond design basis', the Risk Reduction Category (RRC-A) is also denoted as Design Extension Condition (DEC-A). In the German licensing framework, the RRCA (or DEC-A) transients correspond to safety assessment level '4b' of the 'Sicherheitsanforderungen an Kernkraftwerke' (Safety Requirements for Nuclear Power Plants), Az. RS I 5 - 13303/01 of the German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety. These RRC-A (or DEC-A) operating conditions require specific design provisions (implemented by manual or automatic action), known as RRC-A measures, intended to render consequences of accumulated failures admissible. In contrast, RRC-B constitute severe accidents that lead to core melt. Identification of RRC-A operating conditions and corresponding RRC-A measures is based on the use of results of probabilistic safety assessments. After the Fukushima accident the RRC-A accidents like Station Black Out (SBO) or Loss of Ultimate Heat Sink (LUHS) are of particular interest in the safety assessment of nuclear new builds. In several chapters of the Safety Analysis Report it is demonstrated that the AREVA EPRTM design is resistant at RRC-A accident conditions. (orig.)

  12. Practical decommissioning experience with nuclear installations in the European Community

    International Nuclear Information System (INIS)

    Initiated by the Commission of the European Communities (CEC), this seminar was jointly organized by Kernkraftwerke RWE Bayernwerk GmbH (KRB) and the CEC at Gundremmingen-Guenzburg (D), where the former KRB-A BWR is presently being dismantled. The meeting aimed at gathering a limited number of European experts for the presentation and discussion of operations, the results and conclusions on techniques and procedures presently applied in the dismantling of large-scale nuclear installations in the European Community. Besides the four pilot dismantling projects of the presently running third R and D programme (1989-93) of the European Community on decommissioning of nuclear installations (WAGR, BR-3 PWR, KRB-A BWR and AT-1 FBR fuel reprocessing), the organizers selected the presentation of topics on the following facilities which have a significant scale and/or representative features and are presently being dismantled: the Magnox reprocessing pilot plant at Sellafield, the HWGCR EL4 at Monts d'Arree, the operation of an on-site melting furnace for G2/G3 GCR dismantling waste at Marcoule, an EdF confinement conception of shut-down LWRs for deferred dismantling, and the technical aspects of the Greifswald WWER type NPPs decommissioning. This was completed by a presentation on the decommissioning of material testing reactors in the United Kingdom and by an overview on the conception and implementation of two EC databases on tools, costs and job doses. The seminar concluded with a guided visit of the KRB-A dismantling site. This meeting was attended by managers concerned by the decommissioning of nuclear installations within the European Community, either by practical dismantling work or by decision-making functions. Thereby, the organizers expect to have contributed to the achievement of decommissioning tasks under optimal conditions - with respect to safety and economics - by making available a complete and updated insight into on-going dismantling projects and by

  13. Decommissioning of Greifswald NPP (KGR), Greifswald, Germany: Reconstruction of the Former Turbine Hall into a Manufacturing Site for Large Ship Components. Annex A.I-4

    International Nuclear Information System (INIS)

    The Kernkraftwerk Greifswald (KGR) site is located in the north-east of Germany, at the Baltic Sea. At the time of the unification of the German states in October 1990, the Kombinat 'Bruno Leuschner' incorporated almost all East German nuclear facilities, which included the power stations in Greifswald and Rheinsberg, the construction site in Stendal, as well as the disposal site in Morsleben. Directly after the unification, operation and all construction work were stopped. Serious efforts were made to restart some units in Greifswald or to use the site for new nuclear and/or conventional power plants. However, a decision was finally made to decommission all plants, mainly due to a lack of political acceptance and secured financial basis. On the site in Greifswald, there are eight reactor units of the Russian PWR WWER-440. The units 1-4 are of the model 230 and the units 5-8 of the more recent model 213. The reactors are constructed on a double unit basis, i.e. two reactors are arranged in one reactor hall with supporting mechanical equipment and secondary systems together. There is only one turbine hall (with a length of 1.2 km, a height of 40 m, and a width of 35 m) for 16 turbines. The decommissioning preparation started immediately after shutdown of all operating units in 1989/1990 and the decommissioning licence was applied for in June 1994 and issued a year later in June 1995. In parallel with the decommissioning activities, a major objective is to create and support a new future use of the site in Greifswald in order to give the employees and the region new employment opportunities. Under this framework, successively different site areas and building structures have been exempted from the atomic law for industrial reuse

  14. Power engineering. Systems for energy conversion. Compact knowledge for study and profession. 4. upd. and enl. ed.; Energietechnik. Systeme zur Energieumwandlung. Kompaktwissen fuer Studium und Beruf

    Energy Technology Data Exchange (ETDEWEB)

    Zahoransky, Richard A.

    2009-07-01

    This textbook imparts to the reader a fundamental understanding for relations of energy conversion processes. It comprises the total spectra of energy engineering, starting with fundamentals of energy process engineering via description of operating power plants (all types) to energy distribution and - storage. Main topics are sustainable energy systems from renewable energy sources. combined systems (e.g. Gas/steam turbine power plants) and plants with cogeneration (e.g. modular cogeneration plants). A new chapter Kyoto-Protocol was created as a concept of emissions-free coal-fired power plants. A new wording for deregulation of energy markets was received. Numerous texts and graphs were been revised. Chapter 18 ''Deregulation of Energy Markets'' is newly revised. Due to its didactic concepts the book directs not only to students but also everybody, who is inerested into actual questions of energy engineering. (org./GL) [German] Dieses Lehrbuch vermittelt dem Leser ein grundlegendes Verstaendnis fuer die Zusammenhaenge der Energieumwandlungsprozesse. Es umfasst die gesamte Bandbreite der Energietechnik. Die Schwerpunkte reichen von nachhaltigen, erneuerbaren Energietechniken, Kombianlagen (z.B. Gas- und Dampfturbinen-Kraftwerke) ueber Anlagen mit Kraft-Waerme-Kaelte-Kopplung bis hin zum Kyoto-Protokoll. Die 4. Auflage beinhaltet erstmals Uebungsaufgaben mit ausfuehrlichen Loesungen zu den einzelnen Kapiteln. Mehrere Kapitel sind aktualisiert. Das Kapitel 18 ''Liberalisierung der Energiemaerkte'' ist neu gefasst. Aus dem Inhalt Energietechnische Grundlagen - Dampfkraftwerke - Kernkraftwerke - Gasturbinen - Kombinationskraftwerke - Stationaere Kolbenmotoren - Brennstoffzellen - Kraft-Waerme-Kaelte-Kopplung - Wasserkraft - Solartechnik - Windenergie - Biomasse - Geothermie - Energetische Muellverwertung - Energieverteilung und -speicherung - Liberalisierung der Energiemaerkte - Kyoto-Protokoll. (orig.)

  15. Thorium nuclear fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Eom, Tae Yoon; Do, Jae Bum; Choi, Yoon Dong; Park, Kyoung Kyum; Choi, In Kyu; Lee, Jae Won; Song, Woong Sup; Kim, Heong Woo

    1998-03-01

    Since thorium produces relatively small amount of TRU elements after irradiation in the reactor, it is considered one of possible media to mix with the elements to be transmuted. Both solid and molten-salt thorium fuel cycles were investigated. Transmutation concepts being studied involved fast breeder reactor, accelerator-driven subcritical reactor, and energy amplifier with thorium. Long-lived radionuclides, especially TRU elements, could be separated from spent fuel by a pyrochemical process which is evaluated to be proliferation resistance. Pyrochemical processes of IFR, MSRE and ATW were reviewed and evaluated in detail, regarding technological feasibility, compatibility of thorium with TRU, proliferation resistance, their economy and safety. (author). 26 refs., 22 figs

  16. Study on severe accident mitigation measures for the development of PWR SAMG

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In the development of the Severe Accident Management Guidelines (SAMG), it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Station Blackout (SBO), and Anticipated Transients without Scram (ATWS) in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management (SAM) such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.

  17. Probabilistic risk assessment course documentation. Volume 1: PRA fundamentals

    International Nuclear Information System (INIS)

    The full range of PRA topics is presented, with a special emphasis on systems analysis and PRA applications. Systems analysis topics include system modeling such as fault tree and event tree construction, failure rate data, and human Reliability. The discussion of PRA applications is centered on past and present PRA based programs, such as WASH-1400 and the Interim Reliability Evaluation Program, as well as on some of the potential future applications of PRA. The relationship of PRA to generic safety issues such as station blackout and Anticipated Transient Without Scram (ATWS) is also discussed. In addition to system modeling, the major PRA tasks of accident process analysis, and consequence analysis are presented. An explanation of the results of these activities, and the techniques by which these results are derived, forms the basis for a discussion of these topics. An additional topic which is presented in this course is the topic of PRA management, organization, and evaluation. 84 figs., 41 tabs

  18. Inhibition of interleukin-13 gene expression in T cells through GATA-3 pathway by arsenic trioxide

    Institute of Scientific and Technical Information of China (English)

    YAO Xin; HE Hai-yan; YANG Yan; DAI Shan-lin; SUN Pei-li; YIN Kai-sheng; HUANG Mao

    2008-01-01

    @@ Arsenic trioxide (AT) has a long history of use in both traditional Chinese medicine and in modern medicine in asthma therapy.Recently,Yin et al1 found that AT even at small doses reduced the airway inflammation of sensitized guinea pigs.However the mechanism underlying this is still largely unknown.Interleukin 13 (IL-13),as one of the important TH2 cytokines,plays an important role in asthma pathogenesis through promoting eosinophilic inflammation,mucus secretion and airway hyperresponsiveness.2 To further explore the molecular anti-inflammatory basis of AT,we employed Hut-78 cells,a human T cell line,with activation via CD3/CD28 receptors to mimick in vivo co-stimulation to investigate the effect of AT on IL-13 transcription.

  19. THERMAL HYDRAULIC ANALYSIS OF A LIQUID-METAL-COOLED NEUTRON SPALLATION TARGET

    Energy Technology Data Exchange (ETDEWEB)

    W. GREGORY; R. MARTIN; T. VALACHOVIC

    2000-07-01

    We have carried out numerical simulations of the thermal hydraulic behavior of a neutron spallation target where liquid metal lead-bismuth serves as both coolant and as a neutron spallation source. The target is one of three designs provided by the Institute of Physics and Power Engineering (IPPE) in Russia. This type of target is proposed for Accelerator-driven Transmutation of Waste (ATW) to eliminate plutonium from hazardous fission products. The thermal hydraulic behavior was simulated by use of a commercial CFD computer code called CFX. Maximum temperatures in the diaphragm window and in the liquid lead were determined. In addition the total pressure drop through the target was predicted. The results of the CFX analysis were close to those results predicted by IPPE in their preliminary analysis.

  20. Annual meeting on nuclear technology 2013 workshop. Preserving competence in nuclear technology

    Energy Technology Data Exchange (ETDEWEB)

    Steinwarz, Wolfgang [Siempelkamp Nukleartechnik GmbH, Krefeld (Germany)

    2013-10-15

    Main topics of the actual energy and nuclear energy discussion are presented and discussed during the Plenary Session at the first day of the Annual Meeting on Nuclear Technology. The Topical Sessions and the Technical Session are the outstanding expert panels of the Annual Meeting on Nuclear Technology. Young scientists present results of their work on the Workshop Preserving Competence in Nuclear Technology. The Nuclear Energy Campus leads young people through the world of radioactivity, nuclear technology and radiation protection with informational stands and an interactive exhibition. The event is oriented towards upper high school (Gymnasium) classes and engineering classes from technical colleges, as well as students undergoing careers guidance. The main results of the technical part of the Annual Meeting on Nuclear Technology 2013, Berlin 14 to 16 May 2013, are summarised by the chairs for atw. The following report summarises the presentations of the Workshop Preserving Competence in Nuclear Technology. (orig.)

  1. Comparative study of different models of transportation of boron in the codes Thermohydraulic TRAC-BF1, TRACE and RELAP; Estudio Comparativo de Diferentes Modelos de Transporte de Boro en los Codigos Termohidraulicos TRAC-BF1, TRACE y RELAP

    Energy Technology Data Exchange (ETDEWEB)

    Jambrina, A.; Solar, A.; Barrachina, T.; Miro, R.; Verdu, G.; Concejal, A.

    2013-07-01

    In BWR the importance of boron transport lies in maintaining the core integrity during ATWS-kind severe accidents in which under certain circumstances a boron injection is required. The boron transport model implemented in TRAC-BF1 code is based on a first order accurate upwind difference scheme. Four numerical schemes that solve the boron transport model have been analyzed and compared with the analytical solution that provides the Burgers equation: first order Upwind, second order Godunov, second-order modified Godunov and a third-order QUICKEST using the ULTIMATE universal limiter. The modified Godunov scheme has been implemented in TRAC-BF1 source code. The results using these new schemes are presented in this paper.

  2. A study on the regulatory approach of major technical issues

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Choi, J. T.; Kim, I. J. [Sunmoon Univ., Asan (Korea, Republic of); Lee, S. H.; Baek, W. P.; Yang, S. H. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of)

    1998-06-15

    This project is to provide the regulatory direction of 4 major technical issues for the Korean Next Generation Reactors, which are parts of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are as follows : reliability of risk significant SSCs established during design stage must be maintained through the operating life of the plant, currently used classification method of plant conditions and safety requirement were reviewed, and a quantitative classification method is needed to be developed further, the basic regulatory directions are proposed for multiple failures such as SBO, TLOFW, multiple SGTR and ATWS, safety requirements are proposed for survivability/availability of severe accident mitigation design features by 5 items if basic requirements, selection of initial event, identification of available equipment and instruments, identification of environmental conditions and verification methods.

  3. Probabilistic risk assessment course documentation. Volume 1: PRA fundamentals

    Energy Technology Data Exchange (ETDEWEB)

    Breeding, R J; Leahy, T J; Young, J

    1985-08-01

    The full range of PRA topics is presented, with a special emphasis on systems analysis and PRA applications. Systems analysis topics include system modeling such as fault tree and event tree construction, failure rate data, and human Reliability. The discussion of PRA applications is centered on past and present PRA based programs, such as WASH-1400 and the Interim Reliability Evaluation Program, as well as on some of the potential future applications of PRA. The relationship of PRA to generic safety issues such as station blackout and Anticipated Transient Without Scram (ATWS) is also discussed. In addition to system modeling, the major PRA tasks of accident process analysis, and consequence analysis are presented. An explanation of the results of these activities, and the techniques by which these results are derived, forms the basis for a discussion of these topics. An additional topic which is presented in this course is the topic of PRA management, organization, and evaluation. 84 figs., 41 tabs.

  4. A PWR plant model for the analysis of large amplitude transients

    International Nuclear Information System (INIS)

    The PWR transient code ALMOD has been developed to cover a wide range of transient and accident simulation in safety analysis, comprising failure of safety system components (e.g. analysis of anticipated transients without scram=ATWS). Because of the large amplitudes to be expected during the transients, simplified models such as linearized models, used in control system analysis, are not applicable here. As the transients have to be analyzed over minutes, feedback from the entire coolant system becomes effective, thus requiring the simulation of core and both primary and secondary coolant system. Because of the long duration of the transients special emphasis has been put on computational speed. Key variables of interest in transient analysis are fuel and cladding temperature as well as primary and secondary system pressure. Extreme plant conditions such as two phase flow in the primary coolant system, filling of the pressurizer with water etc. have to be simulated with sufficient accuracy. (orig.)

  5. TRACG: Twenty years of collaboration between ENUSA and GE-HITACHI

    Energy Technology Data Exchange (ETDEWEB)

    Haces, J.; Trueba, M.; Garcia, J.; Barrera, J.

    2011-07-01

    TRACG is the GE Hitachi Nuclear Energy (GEH) proprietary version of the Transient Reactor Analysis Code. It is a best-estimate code for analysis of boiling eater reactors (BWR). Enusa has extensively contributed to the development of TRACG, applying this code to different scenarios and BWR plants: loss-of-coolant accident (LOCA), anticipated operational occurrences (AOO), instability events licensing of GNF fuel for Nordic plants, anticipated transients without scram (ATWS) reactivity insertion accidents (RIA), validation of the simulator for the Advanced BWR (ABWR) plant, the licensing of the TRACG based U. s. Nuclear Regulatory commission (NRC)-approved AOO and LOCA licensing methodologies, and in the licensing of the passively safe generation III+ Economic Simplified Boiling Water Reactor (ESBWR).

  6. Response of actinides to flux changes in high-flux systems

    International Nuclear Information System (INIS)

    When discussing the transmutation of actinides in accelerator-based transmutation of waste (ATW) systems, there has been some concern about the dynamics of the actinides under high transient fluxes. For a pure neptunium feed, it has been estimated that the 238Np/237Np ratio increase due to an increasing flux may lead to an unstable, positive reactivity growth. In this analysis, a perturbation method is used to calculate the response of the entire set of actinides in a general way that allows for more species than just neptunium. The time response of the system can be calculated; i.e., a plot of fuel composition and reactivity versus time after a change in flux can be made. The effects of fission products can also be included. The procedure is extremely accurate on short time scales (∼ 1000 s) for the flux levels we contemplate. Calculational results indicate that the reactivity insertions are always smaller than previously estimated

  7. Three-dimensional reactor dynamics code for VVER type nuclear reactors. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R.

    1995-11-17

    A three-dimensional reactor dynamics computer code HEXTRAN has been developed, thoroughly validated, and extensively applied for transient and accident analyses of VVER type nuclear reactors. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical models in spatial and time discretization of neutronics, heat transfer and two-phase flow hydraulics. The dynamic coupling with the thermal hydraulic system code SMABRE allows also the modelling of cooling circuits. Best-estimate or conservative analyses can be performed for different accidents, e.g., RIA, ATWS or local boron dilutions. The usefulness of the three-dimensionality is shown particularly when there are asymmetric or thermal hydraulic disurbances in the core or cooling circuits.

  8. Comparative study of different models of transportation of boron in the codes Thermohydraulic TRAC-BF1, TRACE and RELAP

    International Nuclear Information System (INIS)

    In BWR the importance of boron transport lies in maintaining the core integrity during ATWS-kind severe accidents in which under certain circumstances a boron injection is required. The boron transport model implemented in TRAC-BF1 code is based on a first order accurate upwind difference scheme. Four numerical schemes that solve the boron transport model have been analyzed and compared with the analytical solution that provides the Burgers equation: first order Upwind, second order Godunov, second-order modified Godunov and a third-order QUICKEST using the ULTIMATE universal limiter. The modified Godunov scheme has been implemented in TRAC-BF1 source code. The results using these new schemes are presented in this paper.

  9. European Nuclear Features

    Energy Technology Data Exchange (ETDEWEB)

    Barre, B.; Gonzalez, E.; Diaz Diaz, J.L.; Jimenez, J.L.; Velarde, G.; Navarro, J.M.; Hittner, D.; Dominguez, M.T.; Bollini, G.; Martin, A.; Suarez, J.; Traini, E.; Lang-Lenton, J.

    2004-09-01

    ''European Nuclear Features - ENF'' is a joint publication of the three specialized technical journals, Nuclear Espana (Spain), Revue General Nucleaire (France), and atw - International Journal of Nuclear Power (Germany). The ENF support the international Europeen exchange of information and news about energy and nuclear power. News items, comments, and scientific and technical contributions will cover important aspects of the field. The second issue of ENF contains contributions about theses topics, among others: Institutional and Political Changes in the EU. - CIEMAT Department of Nuclear Fission: A General Overview. - Inertial Fusion Energy at DENIM. - High Temperature Reactors. European Research Programme. - On Site Assistance to Khmelnitsky NPP 1 and 2 (Ukraine). - Dismantling and Decommissioning of Vandellos I. (orig.)

  10. BMD PREDICTION OF DEATH IS ENCAPSULATED BY THE MORPHOLOGICAL ATHEROSCLEROSIS CALCIFICATION DISTRIBUTION (MACD) INDEX

    DEFF Research Database (Denmark)

    Ganz, Melanie; Nielsen, Mads; Karsdal, Morten;

    2009-01-01

    Background: We investigate the relation between the BMD and the aortic calcification markers AC24 and MACD, triglyceride level, cholesterol level, waist-to-hip ratio and the incidence of cardiovascular death. Methods: Our population consists of 308 women aged 48 to 76 that were followed for 8.......3±0.3 years and of which CVD, cancer, and all cause deaths were recorded. The spine BMD and aortic calcification markers, AC24 and the recently proposed Morphological Atherosclerosis Calcification Distribution (MACD) index, were quantified from DXA scans and lateral X-rays respectively. The MACD...... is constructed to capture the risk of death from the outline of aortic calcifications, and not just from the amount of calcification quantified by the AC24. The relation to death was analysed using markers adjusted for age, triglyceride level, and waist circumference (ATW adjusted). A student's t-test of group...

  11. Fabrication of Ni-5 at. %W Long Tapes with CeO2 Buffer Layer by Reel-to-Reel Method

    DEFF Research Database (Denmark)

    Ma, Lin; Tian, Hui; Yue, Zhao;

    2015-01-01

    A 10-m-long homemade textured Ni-5at.%W (Ni5W) long tape with a CeO2 buffer layer has been prepared successfully by means of rolling-assisted biaxially textured substrate (RABiTS) route followed by a chemical solution deposition method in a reel-to-reel manner. Globally, the Ni5W substrate and CeO2...... film exhibit high homogeneity in terms of biaxial texture over the tape. The average values of full width at half maximum of in-plane and out-of-plane texture are 7.2° and 6.1° in Ni5W substrate, 7.6° and 6.1° in CeO2 buffer layer, respectively, all of those with a small standard deviation...

  12. Resolution of US regulatory issues involving boiling water reactor stability

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) and the Boiling Water Reactor Owners Group (BWROG) have been reexamining BWR instability characteristics and consequences since the March 1988 instability event at LaSalle Unit 2. The NRC and BWROG concluded that existing reactor protection systems do not prevent violation of the critical power ratio (CPR) safety limits caused by large asymmetric oscillations. The studies are also examining the need to modify the automatic and operator actions previously developed for response to an anticipated transient without scram (ATWS) event because of oscillation effects not fully considered in previous studies. This paper presents the current status of these studies and an assessment of actions needed to resolve the issue. (author)

  13. Genusa Bepu methodologies for the safety analysis of BWRs; Metodologias Bepu de Genusa para el analisis de seguridad de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Trueba, M.; Garcia, J.; Goodson, C.; Ibarra, L.

    2016-08-01

    This article describes the BEPU methodologies developed by General Electric-Hitachi (GEH) for the evaluation of the BWR reactor safety analysis based on the TRACG best-estimate code. These methodologies are applicable to a wide range of events, operational transients (AOO), anticipated transients without scram (ATWS), loss of coolant accidents (LOCA) and instability events; to different BWR types operating commercially. General Electric (GE( designs and other vendors, including Generation III+ESBWR; to the new operation strategies, and to all types of BWR fuel. Their application achieves, among other benefits, a better understanding of the overall plant response and an improvement in margins to the operating limits; thus, the increase of flexibility in reactor operation and reduction in generation costs. (Author)

  14. GRSAC Users Manual

    Energy Technology Data Exchange (ETDEWEB)

    Ball, S.J.; Nypaver, D.J.

    1999-02-01

    An interactive workstation-based simulation code (GRSAC) for studying postulated severe accidents in gas-cooled reactors has been developed to accommodate user-generated input with ''smart front-end'' checking. Code features includes on- and off-line plotting, on-line help and documentation, and an automated sensitivity study option. The code and its predecessors have been validated using comparisons with a variety of experimental data and similar codes. GRSAC model features include a three-dimensional representation of the core thermal hydraulics, and optional ATWS (anticipated transients without scram) capabilities. The user manual includes a detailed description of the code features, and includes four case studies which guide the user through four different examples of the major uses of GRSAC: an accident case; an initial conditions setup and run; a sensitivity study; and the setup of a new reactor model.

  15. GRSAC Users Manual

    International Nuclear Information System (INIS)

    An interactive workstation-based simulation code (GRSAC) for studying postulated severe accidents in gas-cooled reactors has been developed to accommodate user-generated input with ''smart front-end'' checking. Code features includes on- and off-line plotting, on-line help and documentation, and an automated sensitivity study option. The code and its predecessors have been validated using comparisons with a variety of experimental data and similar codes. GRSAC model features include a three-dimensional representation of the core thermal hydraulics, and optional ATWS (anticipated transients without scram) capabilities. The user manual includes a detailed description of the code features, and includes four case studies which guide the user through four different examples of the major uses of GRSAC: an accident case; an initial conditions setup and run; a sensitivity study; and the setup of a new reactor model

  16. 2010 ANNUAL MEETING ON NUCLEAR TECHNOLOGY. Pt. 3. Section reports

    International Nuclear Information System (INIS)

    Summary report on these 6 - out of 12 - Sessions of the Annual Conference on Nuclear Technology held in Berlin on May 3 to 6, 2010: - Decommissioning of Nuclear Installations (Session 7), - Fusion Technology (Session 8), - Energy Industry and Economics (Session 10), - Radiation Protection (Session 11), - New Build and Innovations (Session 12), and - Education, Expert Knowledge, Know-how-Transfer (Session 13). The other Sessions: - Reactor Physics and Methods of Calculation (Session 1), - Thermodynamics and Fluid Dynamics (Session 2), - Safety of Nuclear Installations - Methods, Analysis, Results (Session 3), - Front End and Back End of the Fuel Cycle, Radioactive Waste, Storage (Session 4), - Front End of the Fuel Cycle, Fuel Elements and Core Components (Session 5), - Operation of Nuclear Installations (Session 6) have been covered in atw issues 10 and 11 (2010). (orig.)

  17. Calculation of mechanical strength of the bolts of the flanged joints of LWR-type reactor pressure vessels (with particular emphasis on the behaviour under critical internal excess pressure, acting like a ''safety valve''). 5. Technical report

    International Nuclear Information System (INIS)

    The reactor pressure vessel has to be made absolutely fail-safe towards excess primary loads (internal pressure). For this purpose, the vessel is equipped with safety valves (e.g. at the pressurizer) which normally are fully sufficient to master any pressure excursion. Nevertheless, a deterministic safety approach requires additional measures to ensure, in case of safety valve failure, control of the pressure so as to prevent reactor pressure vessel bursting. One way to achieve this is to make the flange joints plastify so strongly in the course of a pressure transient that the flange gap will sufficiently widen and thus behave like a 'safety valve'. In order to keep damage as small as possible, these parts should be easy to replace so that the bolts, or rather the washers, seem to be appropriate for modification. Tests have been made to ascertain whether reduction of bolt cross-sectional area (increase in admissible stress), or insertion of suitable washers is the best way to achieve reliable behaviour under normal conditions and also additional function in terms of a safety valve in case of pressure transients. For this purpose, model calculations have been made for all possible flange joints whith all possible stress variations and bolt dimensions. The event simulated is the ATWS', and the strength computations and stress analyses made for the flange joints of the pressure vessel of the Biblis reactor, unit B, are taken as an example. Main attention has been given to the forces affecting the bolts and to the forces acting between the reactor vessel head flange and the vessel flange under internal excess pressure. For assessment of the thermodynamic processes in case of an ATWS, the calculations made for the Grafenrheinfeld reactor have been taken as a basis. (orig.)

  18. Preliminary safety analysis for key design features of KALIMER

    International Nuclear Information System (INIS)

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions

  19. Preliminary safety analysis for key design features of KALIMER with breakeven core

    International Nuclear Information System (INIS)

    KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model

  20. Safety analysis for key design features of KALIMER with breakeven core

    International Nuclear Information System (INIS)

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term nuclear R and D Program. In this report, key safety design features are described and safety analyses results for typical ATWS accidents in the KALIMER design with breakeven core are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the event categorization and acceptance criteria for the KALIMER safety analysis are described in chapter 2. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER system response to the events. They are categorized as Bounding Events (BEs) because of their low probability of occurrence. In chapter 4, the performance analysis results of the KALIMER containment dome are described along with the HCDA accident scenario and source terms. The major containment parameters of peak pressure and peak temperature have been calculated using the CONTAIN-LMR code. Radiological consequence has been evaluated by the MACCS code. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using SCHAMBETA code developed in the framework of the modified bethe-tait method. Work energy potentials based arising from the sodium expansion as well as the isentropic fuel expansion are then calculated to evaluate the structural integrity of the reactor vessel, reactor internals and primary coolant system of KALIMER

  1. Effect of W additions on the structural and magnetic properties of Ni{sub 50}Ti{sub 50−x}W{sub x} and Ti{sub 50}Ni{sub 50−x}W{sub x} systems obtained by mechanical alloying

    Energy Technology Data Exchange (ETDEWEB)

    Jara, Angelica; Arjona, Jose David; Bautista, Pedro; Gonzalez, Gema, E-mail: gemagonz@ivic.gob.ve

    2014-12-05

    Highlights: • W additions strongly affect the magnetic and structural properties of Ni-Ti. • The saturation magnetization and magnetic remanence decreases with W addition. • W additions induces amophization of Ni-Ti. - Abstract: The effect of tungsten (W{sub x}) additions (x = 0.5, 1.0, 1.5 and 2.0 at.%), on the structural and magnetic properties of the binary systems Ni{sub 50}Ti{sub 50−x} and Ti{sub 50}Ni{sub 50−x} obtained by mechanical alloying was studied. The elementary powders were milled in a Spex 8000 horizontal mill, under N{sub 2} atmosphere, for 5 and 20 h. After 20 h of milling a homogenous microstructure was observed, particularly for small W additions. For this milling time a mixed of nanocrystalline and amorphous structure was obtained. As W concentration increases (1, 1.5 and 2 at.%), in both systems, the presence of small β-W reflections and the presence of very small peaks corresponding to the formation of an incipient new phase, identified as a NiTi(W) solid solution was observed, especially evident for 2 at.%W. The saturation magnetization and magnetic remanence decreases with the addition of W down to a minimum value at 1.5 at.%W, for both systems. The samples were characterized by SEM, EDS, XRD and magnetic measurements by VSM. The structural and magnetic behavior for both ternary alloys was very similar with the W additions.

  2. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

    2000-07-01

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions.

  3. Analyses of Design Extended Condition Events for the Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Jeong, Taekyung; Lee, Kwilim; Jeong, Jaeho; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the sensitivity tests are conducted. In the case of the UTOP event, a sensitivity test for the reactivity insertion amount and rate were conducted. This analysis can give a requirement for margin of control rod stop system (CRSS). For example, the CRSS in the PRISM designed based on the 0.4 $ reactivity insertion, which is analyzed with safety analysis of UTOP event. Moreover, the sensitivity tests for weighting factor in the core radial expansion reactivity feedback model were also carried out for all ATWS events. Currently, the reactivity feedback model for the PGSFR is not validated yet. However, the reactivity feedback models in the MARS-LMR are validating with various plant-based data including EBR-II SHRT. The ATWS events for the PGSFR classified in the design extended condition including UTOP, ULOF, and ULOHS are analyzed with MARS-LMR. In this study, the sensitivity tests for reactivity insertion amount and rate in the UTOP event are conducted. The reactivity insertion amount is obviously an influential parameter. The reactivity insertion amount can give a requirement for design of the CRSS, therefore, this sensitivity result is very important to the CRSS. In addition, sensitivity tests for the weighting factor in the radial expansion reactivity model are carried out. The weighting factor for a grid plate, W{sub GP}, which means contribution of feedback in the grid plate is changed for all unprotected events. The grid plate expansion is governed by a core inlet temperature. As the W{sub GP} is increased, the power in the UTOP and the ULOF is increased, however, the power in the ULOHS is decreased. The higher power during transient means lower reactivity feedback and smaller expansion. Thus, the core outlet temperature rise is dominant in the UTOP and ULOF events, however, the core inlet temperature rise is dominant in the ULOHS. Therefore, the grid plate expansion in the ULOHS is predominant.

  4. Furrow-irrigated chufa crops in Valencia (Spain. I: Productive response to two irrigation strategies

    Directory of Open Access Journals (Sweden)

    N. Pascual-Seva

    2013-01-01

    Full Text Available Chufa (Cyperus esculentus L. var. sativus Boeck. is an important vegetable crop in Valencia (Spain, where its tubers are used to produce a refreshing drink called 'horchata'. Water is relatively inexpensive, there are no data regarding the volumes of water used to grow chufa, and the irrigation water use efficiency (IWUE has neither been determined. The aim of this research was to compare the productive responses of the chufa crop to two irrigation strategies (IS. The volumetric soil water content (VSWC was monitored with capacitance sensors. Trends in VSWC were used to determine the in situ field capacity (FC, beginning each irrigation event when the VSWC reached either approximately 45% (H1 or 60% (H2 of the FC at a soil depth of 0.10 m. The experiments were conducted over three consecutive seasons. An area velocity flow module measured the water flow. The yields, the water volumes used, and the IWUE were calculated. Plants were periodically sampled and the harvest index and relative growth rate were determined. The yield was affected by the year and by the IS. The greatest yields were obtained with the H2 strategy (on average 2.18 kg m-2 for H2 vs. 1.94 kg m-2 for H1; p≤0.01, and the average tuber weight (ATW was affected (p≤0.01 by the year and IS interaction. IWUE was affected by the year, and none of the considered factors affected the harvest index (p≤0.05. It can be concluded that maintaining a higher VSWC would increase both yield and ATW without affecting IWUE.

  5. Ten years of nuclear law development

    International Nuclear Information System (INIS)

    I took over the legal column in atw in early 1998. My second contribution was about the 8th amendment to the German Atomic Energy Act. My last but one article covered the 10th act amending the Atomic Energy Act focusing on the revision of the reliability audit and the regulations about competence for the Asse II mine. What are the changes in German atomic energy law over this ten-year period? What will be the future of atomic energy law in Germany? The term 'Atomic Energy Act' conceals the fact that the Atomic Energy Act of the 8th amendment does not have much in common any more with the Act of the 10th amendment. The dividing line appeared in the 9th amendment, which put into effect one of the key objectives of the red-green coalition government of the autumn of 1998: Terminating the peaceful use of nuclear power 'if possible by consensus' and without any indemnification of licensees. Although the Atomic Energy Act of April 22, 2002 formally kept its name, the original purpose of this piece of legislation was turned into the opposite by mentioning as the first objective the orderly termination of the use of nuclear power for commercial generation of electricity. On a European level, nuclear power has been re-evaluated in the meantime for various obvious reasons, and it is to be hoped that also Germany will find a way back to using nuclear power within the broad energy mix. With this contribution, which is my last one, I say goodbye to the readers of the legal column in atw. Thank you for your interest over all the years. (orig.)

  6. Safety analysis for key design features of KALIMER-600 design concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong-Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Joeng, H. Y.; Ha, K. S.; Heo, S

    2005-03-01

    KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents, containment design basis accidents, and flow blockages in the KALIMER design are presented. First, the basic approach to achieve the safety goal and main design features of KALIMER-600 are introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2, In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER-600 system response to the events. The objectives of Chapter 4, are to assess the response of KALIMER-600 containment to the design basis accidents and to evaluate whether the consequences are acceptable or not in the aspect of structural integrity and the exposure dose rate. In Chapter 5, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly, are described. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed.

  7. Nuclear power plant personnel errors in decision-making as an object of probabilistic risk assessment. Methodological extensions on the basis of a differentiated analysis of safety-relevant goals; Entscheidungsfehler des Betriebspersonals von Kernkraftwerken als Objekt probabilistischer Risikoanalysen; Methodische Erweiterungen auf der Basis einer differenzierten Betrachtungsweise sicherheitsgerichteter Ziele

    Energy Technology Data Exchange (ETDEWEB)

    Reer, B.

    1993-09-01

    wird eine neuartige Fragetechnik entwickelt und gezeigt, dass solche Fehler als Rueckwirkungen untergeordneter Ziele auftreten koennen. Solche Rueckwirkungen sind ueber eine differenzierte Betrachtung sicherheitsgerichteter Ziele identifizierbar. Zur Quantifizierung wird eine neue Methode entwickelt, mit der sich situationsspezifisch Wahrscheinlichkeiten fuer Entscheidungsfehler schaetzen lassen. Es gelingt, die Faktoren Konflikt und Aehnlichkeit so zu operationalisieren, dass sie mit den Informationen, die einem PRA-Anwender ueblicherweise zur Verfuegung stehen, quantitativ zugaenglich sind. Das Quantifizierungs verfahren basiert auf Extra- und Interpolationen zu den wenigen Daten, die zur Zeit ueber Entscheidungsfehler von Operateuren existieren. Ausserdem wird fuer passive Entscheidungsfehler (Unterlassungen notwendiger Handlungen) ein neuartiger Modellansatz vorgestellt, der solche Fehler ueber Verzoegerungszeiten quantifiziert. Die praktische Durchfuehrbarkeit dieses dynamischen Ansatzes wird am Beispiel einer probabilistischen Analyse der konkreten Operateurmassnahmen gezeigt, die waehrend des Stoerfalls im Kernkraftwerk Davis-Besse (1985) angefordert wurden. Die Erweiterungen der klassischen MMSA-Methodik werden am Beispiel der Nachwaermeabfuhr (NWA) des HTR-500 angewendet. Entscheidungsfehler (als Ursachen ungeplanter Handlungen) werden systematisch und umfassend beruecksichtigt. Fuenf zusaetzliche Entscheidungsfehler werden identifiziert.

  8. Change in CRUD deposition, water chemistry and ECP response after the transition to HWC/OLNC at KKL - An update

    International Nuclear Information System (INIS)

    In November 2008, Kernkraftwerk Leibstadt (KKL) started injection of Pt in the reactor water during power operation, On-line NobleChemTM (OLNC), to mitigate IGSCC in the reactor internals. The water chemistry regime in KKL thus changed from NWC to hydrogen injection and OLNC. KKL has since then applied OLNC at several campaigns. An extensive testing and evaluation program was performed following the first OLNC injection, in a joint project between KKL, Westinghouse and EPRI. The aim of the project was to study how platinum injection influences crud behavior and water chemistry in a high duty plant with annual cycles. Crud sample collection, as part of the detailed poolside inspection campaign reported previously [2], [3], was performed on several assemblies that had been exposed to OLNC for the first time. The transition to HWC/OLNC showed an increased availability of iron in core and caused an increased fuel crud deposition, especially at high axial elevations of the rods. The total crud, seen after the introduction of OLNC, seems to contain mainly iron. The average pre-HWC/OLNC fuel crud composition was 76% Fe and 15% Zn. The fuel crud after exposure to one HWC/OLNC cycle consisted in average of 88% Fe and 6% Zn. The amount of Pt on rods exposed to HWC/OLNC corresponds in average to about 0.2% of the total crud. Most Pt is found in the crud at the upper axial locations of the fuel rods, especially for fuel with reasonably long time in operation. Pt was mainly present in the hematite-rich outer crud layer. The axial fuel crud distributions typically show a lower maximum around the level 700 mm, and an upper maximum just below 3000 mm. The upper maximum is most pronounced for Optima and Optima2 fuel that have been exposed to the HWC/OLNC cycle. These results are compared to the previous inspections results before HWC/OLNC In the paper, the crud deposition changes after OLNC, as well as the role of platinum, is analyzed and discussed. (authors)

  9. Real and mythical consequences of Chernobyl accident

    International Nuclear Information System (INIS)

    This presentation describes the public Unacceptance of Nuclear Power as a consequence of Chernobyl Accident, an accident which was a severest event in the history of the nuclear industry. It was a shock for everybody, who has been involved in nuclear power programs. But nobody could expect that it was also the end romantic page in the nuclear story. The scale of the detriment was a great, and it could be compared with other big technological man-made catastrophes. But immediately after an accident mass media and news agencies started to transmit an information with a great exaggerations of the consequences of the event. In a report on the Seminar The lessons of the Chernobyl - 1' in 1996 examples of such incorrect information, were cited. Particularly, in the mass media it was declared that consequences of the accident could be compared with a results of the second world war, the number of victims were more than hundred thousand people, more than million of children have the serious health detriments. Such and other cases of the misconstruction have been called as myths. The real consequences of Chernobyl disaster have been summed on the International Conference 'One decade after Chernobyl' - 2, in April 1996. A very important result of the Chernobyl accident was a dissemination of stable unacceptance of the everything connected with 'the atom'. A mystic horror from invisible mortal radiation has been inspired in the masses. And from such public attitude the Nuclear Power Programs in many countries have changed dramatically. A new more pragmatic and more careful atomic era started with a slogan: 'Kernkraftwerk ? Nein, danke'. No doubt, a Chernobyl accident was a serious technical catastrophe in atomic industry. The scale of detriment is connected with a number of involved peoples, not with a number of real victims. In comparison with Bhopal case, earthquakes, crashes of the airplanes, floods, traffic accidents and other risky events of our life - the Chernobyl is

  10. Piercing of corporate veil of nuclear companies; Durchgriffshaftung der Atomkonzerne

    Energy Technology Data Exchange (ETDEWEB)

    Frenz, Walter [RWTH Aachen Univ. (Germany). Lehr- und Forschungsgebiet Berg-, Umwelt- und Europarecht

    2015-11-15

    Belreibergesellschaften im Fall der Beendigung der Beherrschungs- und Ergebnisabfuehrungsvertraege fuer die Nuklearverbindlichkeiten nur sehr eingeschraenkt, und zwar in zweifacher Hinsicht: Der Anspruch ist lediglich auf Sicherungsleistung gerichtet und nicht auf Kostenuebernahme und zudem entsprechend der Judikatur auf fuenf Jahre nach seiner Begruendung begrenzt; fuer den Bereich des Umwandlungsrechts gelten vergleichbare Regelungen. Dabei dauert der Rueckbau eines Kernkraftwerks allein schon 20 Jahre und ein Endlager duerfte vor 2050 nicht verfuegbar sein.

  11. Assessment of the quality of safety relevant SSC's by analysis of transients and events within the scope of aging management; Qualitaetsbeurteilung von SIWI-Komponenten durch Transienten- und Ereignisanalyse im Rahmen eines prozessorientierten Alterungsmanagements

    Energy Technology Data Exchange (ETDEWEB)

    Ilg, U. [EnBW Kernkraft GmbH, Kernkraftwerk Philippsburg (Germany); Koenig, G. [EnBW Kernkraft GmbH, Kernkraftwerk Neckarwestheim (Germany); Schoeckle, F. [Amtec Messtechnischer Service GmbH, Lauffen (Germany)

    2007-07-01

    In the context of ageing management in the EnBW nuclear power stations Philippsburg and Neckarwestheim, the quality standard of the safety-relevant components must be monitored regularly. In the case of the mechanical components of Category M 1 (integrity concept with assured quality and exclusion of component failure), this means: Monitoring loads and water chemistry as potential damage mechanisms and monitoring the consequences of possible damage mechanisms (non-destructive testing etc.). In Category M 2, quality must be ensured and common mode failure must be excluded. These are mostly the external systems. Here, too, minimum quality must be ensured, e.g. by recurrent inspections, functional testing, inspection and maintenance. In both groups, also results of incidents and failure reports are considered. Quality assurance is component-specific including all measures for ensuring and maintaining quality. It is also checked whether the methods applied are sufficient and appropriate against the background of current knowledge. The summarized results of the evaluations (normally, in the form of so-called 'status sheets' with data tables) are presented and explained annually in a status report or status discussion. The contribution explains the procedure and presents concrete examples. [German] Im Rahmen des an den EnBW-Kernkraftwerks-Standorten Philippsburg und Neckarwestheim eingefuehrten operativen Alterungsmanagements ist der Qualitaetsstand der zu betrachtenden sicherheitstechnisch wichtigen (SIWI-) Komponenten regelmaessig zu bewerten. Fuer die mechanischen Komponenten der Gruppe M 1 (Integritaetskonzept, d.h. die Qualitaet ist zu gewaehrleisten; Komponenten duerfen nicht versagen) wird dabei jeweils einbezogen: Ergebnisse der Ueberwachung der Ursachen moeglicher Schaedigungsmechanismen (Ueberwachung der Belastungen sowie der Wasserchemie) sowie deren Bewertung und alle Ergebnisse aus der Ueberwachung der Folgen moeglicher Schaedigungsmechanismen

  12. Simulator training. Requirements and ways to meet them; Die Simulatorschulung. Anforderungen und deren Realisierung

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, E. [Gesellschaft fuer Simulatorschulung mbH (French Guiana), Essen (Germany) KSG Kraftwerks-Simulator-Gesellschaft mbH, Essen (Germany)

    1997-02-01

    The excellent operating performance of German nuclear power plants year by year attests to the high reliability and safety of these plants. On the one hand, this is due to the mature technical plant concept and, on the other hand, to the safety culture of plant operation, i.e. the qualified and careful work of the plant personnel running and monitoring the plants and keeping them in top condition technically. Special responsibility rests on the shift personnel operating plants around the clock. It is for this reason that operators take very seriously the efforts by these staff members to acquire and preserve a high level of training. Training by simulators, as organized centrally in Essen for twenty-two nuclear generating units in Germany, the Netherlands, and Switzerland, plays a major role in these activities. The special characteristic of simulator training is the combination of two things, namely technology and human behavior. The power plant processes, with all their complexities, must be conveyed to human operators, with all their skills traits and weaknesses, by means of appropriate simulation. The requirements resulting from this need, and the solutions adopted at the Essen Simulator Center, are described in this article. (orig.) [Deutsch] Die deutschen Kernkraftwerke beweisen mit ihren hervorragenden Betriebsergebnissen jedes Jahr erneut ihre hohe Zuverlaessigkeit und Sicherheit. Dies beruht zum einen auf dem technisch ausgereiften Anlagenkonzept und zum anderen auf der Sicherheitskultur des Betriebes, d.h. der qualifizierten und sorgfaeltigen Arbeit des Betriebspersonals, welches das Kraftwerk betreibt, ueberwacht und in technisch einwandfreiem Zustand erhaelt. Eine besondere Verantwortung hat das Personal der Schichten, welche die Anlage rund um die Uhr `fahren`. Enstprechend ernst nehmen deshalb die Betreiber den Erwerb und den dauerhaften Erhalt eines hohen Ausbildungsstandes dieser Mitarbeiter. Die Schulung an Simulatoren - wie sie zentral in Essen

  13. 49-2游泳池式反应堆超设计基准事故的筛选与分析%Screening and Analysis of Beyond Design Basis Accident of 49-2 SPR

    Institute of Scientific and Technical Information of China (English)

    张亚东; 郭玥; 吴园园; 邹耀

    2015-01-01

    The beyond design basis accident was analyzed to ensure safe operation of 49‐2 Swimming Pool Reactor (SPR) after design life .Because it’s difficult to use PSA method ,the unconditional assumed severe accidents were adopted to obtain a conserva‐tive result . The main conclusions were obtained by analyzing anticipated transients without scram in station blackout (SBO ATWS) ,horizontal channel rupture ,core un‐covering after shutdown and emergency response capacity .The results show that the core is safe in SBO ATWS ,and the fuel elements will not melt as long as the core are not exposed in 2.5 h in loss of coolant accident caused by horizontal channel rupture and other factors .The passive siphon breaker function and various ways of emergency core makeup can ensure that the core is not exposed .%为保证49‐2游泳池式反应堆在超寿期下的安全运行,需进行超设计基准事故分析。由于难以采用概率安全评价(PSA )方法进行分析,所以本文无条件假设最严重事故来得到一保守结果。主要分析了全厂断电下未能紧急停堆的预期瞬变(ATWS)、水平孔道断裂和停堆后堆芯完全裸露的事故,以及应急能力。结果表明:在全厂断电A T WS下堆芯是安全的;水平孔道断裂及其他因素造成失水时,只要2.5 h内堆芯不裸露即可保证燃料元件不熔化;非能动破坏虹吸能力和多样的应急补水方式能保证堆芯不裸露。

  14. New Fuel Pin Axial Expansion Reactivity Feedback Model in MARS-LMR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chiwoong; Ha, Kwiseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the sensitivity tests are conducted. In the case of the UTOP event, a sensitivity test for the reactivity insertion amount and rate were conducted. This analysis can give a requirement for margin of control rod stop system (CRSS). For example, the CRSS in the PRISM designed based on the 0.4 $ reactivity insertion, which is analyzed with safety analysis of UTOP event. Moreover, the sensitivity tests for weighting factor in the core radial expansion reactivity feedback model were also carried out for all ATWS events. Currently, the reactivity feedback model for the PGSFR is not validated yet. However, the reactivity feedback models in the MARS-LMR are validating with various plant-based data including EBR-II SHRT. The ATWS events for the PGSFR classified in the design extended condition including UTOP, ULOF, and ULOHS are analyzed with MARS-LMR. In this study, the sensitivity tests for reactivity insertion amount and rate in the UTOP event are conducted. The reactivity insertion amount is obviously an influential parameter. The reactivity insertion amount can give a requirement for design of the CRSS, therefore, this sensitivity result is very important to the CRSS. In addition, sensitivity tests for the weighting factor in the radial expansion reactivity model are carried out. The weighting factor for a grid plate, W{sub GP}, which means contribution of feedback in the grid plate is changed for all unprotected events. The grid plate expansion is governed by a core inlet temperature. As the W{sub GP} is increased, the power in the UTOP and the ULOF is increased, however, the power in the ULOHS is decreased. The higher power during transient means lower reactivity feedback and smaller expansion. Thus, the core outlet temperature rise is dominant in the UTOP and ULOF events, however, the core inlet temperature rise is dominant in the ULOHS. Therefore, the grid plate expansion in the ULOHS is predominant. These analysis results will give better

  15. Application of Advanced Thermal Hydraulic TRACG Model to Preserve Operating Margins in BWRs at Extended Power Up-rate Conditions

    International Nuclear Information System (INIS)

    GE has developed TRACG, a customized BWR version of the TRAC model, for application to BWR analyses. This model was initially applied to special BWR challenges and for benchmarking the official simplified thermal-hydraulic design models. However, in past years extensive additional model development, qualification and application studies have been completed. This development has followed the CSAU methodology, where extensive model evaluation and qualification have been performed to demonstrate the applicability of the model and to quantify the uncertainty in the model parameters as well as in plant parameters and initial conditions. This has then been combined with a statistically based application methodology following the CSAU approach to generate tolerance limits for the critical safety and design parameters. This effort has resulted in application processes that have been reviewed and approved by the US NRC to enable routine application of the TRACG model to the design and licensing analyses and utilize the improved operating margin to optimize the fuel cycle design. These applications have been supported by development of programs that construct specific plant and problem base-decks that utilize BWR plant characteristics and system databases to standardize and streamline the application to several plants. The application of the TRACG model in Transient and LOCA analyses has assisted in allowing similar power peaking at higher power density conditions for BWRs. Also, the application of the TRACG model in Stability analyses has assisted in preserving the setpoints of stability monitoring systems to avoid margin loss for high power density applications. TRACG is being used for analysis of ATWS events. It has been used to support the development of emergency procedure guidelines, and it is currently being used to demonstrate that the suppression pool temperature limits can be met for up-rated conditions. Finally, the application of the TRACG model in Faulted Load

  16. Corrosion product deposition on fuel element surfaces of a boiling water reactor

    International Nuclear Information System (INIS)

    Over the last decade the problem of corrosion products deposition on light water reactor fuel elements has been extensively investigated in relation to the possibility of failures caused by them. The goal of the present study is to understand in a quantitative way the formation of such kind of deposits and to analytically understand the mechanism of formation and deposition with help of the quasi-steady state concentrations of a number of 3d metals in reactor water. Recent investigations on the complex corrosion product deposits on a Boiling Water Reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The buildup of magnetic corrosion product deposits (crud) on the fuel cladding of the BWR, Kernkraftwerk Leibstadt (KKL) Switzerland has hampered the Eddy-current based measurements of ZrO2 layer thickness. The magnetic behavior of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS). Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements’ oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main components identified by 2D μXRD mapping inside the layer were hematite and spinel phases with the common formula (MxFey)[M(1-x)Fe(2-y)]O4, where M = Zn, Ni, Mn. With μXRD it was clearly shown that the cell parameter of analyzed spinel is different from the one of the pure endmembers (ZnFe2O4, NiFe2O4 and MnFe2O4) proving the existence of solid solutions. These

  17. Development of integrated computer code for analysis of risk reduction strategy

    International Nuclear Information System (INIS)

    The development of the MIDAS/TH integrated severe accident code was performed in three main areas: 1) addition of new models derived from the national experimental programs and models for APR-1400 Korea next generation reactor, 2) improvement of the existing models using the recently available results, and 3) code restructuring for user friendliness. The unique MIDAS/TH models include: 1) a kinetics module for core power calculation during ATWS, 2) a gap cooling module between the molten corium pool and the reactor vessel wall, 3) a penetration tube failure module, 4) a PAR analysis module, and 5) a look-up table for the pressure and dynamic load during steam explosion. The improved models include: 1) a debris dispersal module considering the cavity geometry during DCH, 2) hydrogen burn and deflagration-to-detonation transition criteria, 3) a peak pressure estimation module for hydrogen detonation, and 4) the heat transfer module between the molten corium pool and the overlying water. The sparger and the ex-vessel heat transfer module were assessed. To enhance user friendliness, code restructuring was performed. In addition, a sample of severe accident analysis results was organized under the preliminary database structure

  18. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

    Energy Technology Data Exchange (ETDEWEB)

    Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L. (Science Applications International Corp., Albuquerque, NM (USA); Sandia National Labs., Albuquerque, NM (USA))

    1989-08-01

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs.

  19. The 50{sup th} anniversary of the N.S. Otto Hahn. When nuclear power said 'Ahoy'; Als die Kernenergie 'Ahoi' rief. NS Otto Hahn. Nuklearbetriebenes Forschungs- und Frachtschiff

    Energy Technology Data Exchange (ETDEWEB)

    Reinartz, Jerome [atw-Redaktion, Hattingen (Germany)

    2013-11-15

    In September 1963 the construction of the only German nuclear ship to date, NS Otto Hahn, started in Kiel. For the 50{sup th} anniversary, atw remembers an important part of nuclear technology history in Germany: The research and freight ship shows one thing above all in retrospect: The technology ran reliably. But cost pressure and reservations shattered the dream of a nuclear power shipping era. Until it was decommissioned in 1979 the ship travelled a total of 650,000 sea miles and called at 33 harbours in 22 countries. On the research level, the 'Otto Hahn' could satisfy expectations, however, it could not ring in an era of nuclear shipping - the atomic boat could never cover its operating costs with its freight trips and permission to call at foreign ports were rare. However, on the one hand, the ship's journeys, sometimes under hard weather conditions, demonstrated just how robust and durable the 'progressive pressurized water reactor' on board was, on the other hand, the 'Otto Hahn' had by all means been a prototype which under other market conditions could have been a model for nuclear container ships. In any case, it proved the performance capacity of the then still young German and European nuclear technology industry, that did not need to hide behind the Russian and American competition. (orig.)

  20. Chemically deposed layer sytems for the realization of YBa{sub 2}Cu{sub 3}O{sub 7-{delta}} band conductors; Chemisch deponierte Schichtsysteme zur Realisierung von YBa{sub 2}Cu{sub 3}O{sub 7-{delta}}-Bandleitern

    Energy Technology Data Exchange (ETDEWEB)

    Engel, Sebastian

    2009-04-30

    The aim of this thesis was to produce new buffer-layer systems for biaxially texturated Ni5at%W substrates by means of chemical processes. As very promising materials for the buffer layers CaTiO{sub 3} and SrTiO{sub 3} were chosen. The production of the single layers pursued from the organometallic prestage by means of dip coating and subsequent head treatment. During the work first the single precursor solutions were to be developed. A main component of the theses forms the understanding of the texture development during the heat treatment of precursor layers on biaxially texturated metallic substrates. Based on this the growth of thick buffer layers is studied and by means of YBCO layers, which were deposed by beans of a pulsed laser, the functionality of the synthesized buffer layers proved. A further component of this thesis formes the influence of nanoscaling precipitations in thew YBCO on its superconducting properties. The YBCO deposition pursued via a variation of the TFA process, as substrate (001)-oriented SrTiO{sub 3} monocrystals were applied.

  1. YBCO coated conductors prepared by chemical solution deposition: A TEM study

    International Nuclear Information System (INIS)

    Recently large attention has been devoted to chemical solution deposition (CSD) as a promising method for fabricating low-cost YBCO coated conductors. We present an extensive transmission electron microscopy (TEM) cross-section analysis of CSD grown La2Zr2O7 (LZO) buffer layers on flexible Ni-5at%W substrates. The high performance of these chemical solution derived buffer layers was confirmed by a YBCO critical current density Jc of 0.84 MA/cm2 achieved for a coated conductor sample with a layer sequence Ni-5at%W/LZO (CSD)/CeO2 (CSD)/YBCO, where the YBCO film was deposited by pulsed laser deposition (PLD). TEM sample preparation was carried out by conventional mechanical polishing and ion milling techniques. TEM bright-field images of the LZO films and nickel substrates were acquired under two-beam conditions. The layer thicknesses and nanovoid size were determined for the LZO buffer layers. Moreover, the interfaces between the different layers were investigated and identified. Electron diffraction patterns were obtained in order to determine the microscopic texture of the samples. Despite the presence of nanovoids in the LZO buffer layers, they act as efficient Ni diffusion barriers

  2. Restructuring and hierarchisation of component lists subject to QA(Q-Lists)

    International Nuclear Information System (INIS)

    Until now, component lists subject to Quality Assurance (Q-lists) included every structure, system and component (SSC) related to safety, and therefore subject to Quality Assurance, in accordance with Appendix B of 10 CFR 50. This involves applying all Quality Assurance, maintenance, standardization, auditing, inspection, etc, procedures to a large number of components, which causes delay to and hinders to a large extent refuelling and maintenance times. The NRC is currently proposing the restructuring of the Q-lists, categorizing and establishing hierarchies for the SSCs based on their particular contribution to core damage. Significance categorization is made possible by previously determining the importance or risk significance of each SSC. To identify risk significant SSCs, NUMARC 93-01 shall be used as input data. The results of PSA and other deterministic criteria, eg RG 1.97. ATWS, etc, shall also be taken into account. Iberdrola is going to apply all these criteria in the case of Cofrentes NPP, which will lead to a substantial reduction in maintenance costs and the application of a more efficient Quality Assurance Programme, in keeping with the guidelines and latest NRC trends in this item. (Author)

  3. Review and comment on the advanced spent fuel management process (1): Technical aspects and non-proliferation concerns

    International Nuclear Information System (INIS)

    Efforts are made to analyze the project, the Advanced Spent Fuel Management Technology (ASFMT), which is currently carried out at Korea Atomic Energy Research Institute, on the technical feasibility and validity as well as on the nuclear non-proliferation concerns. The project is a part of a program under the 'Long and Midterm Nuclear Development Program'. On the technical analysis, reviewed the papers presented at the national and international meetings on the subject by KAERI staffs, and also participated to various technical discussions on the 'Mock-up Test', currently in progress. On the non-proliferation concerns, the ASFMT project was reviewed and analyzed in reference to various programs currently in progress or in a formulation stages in US, such as the DOE TOPS and ATW. Further reviewed the past JASNEC process and programs for possible application of the ASFMT project for JASNEC project. Provided a few thoughts for effectively carrying out the ASFMT project, and a plan for the next phase is presented

  4. Mitigation of sodium-cooled fast reactor severe accident consequences using inherent safety principles

    International Nuclear Information System (INIS)

    Full text: Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. In the United States, accidents which have the potential for severe consequences usually are of probability less than 1 x 10-4 per reactor year, intended to satisfy the U.S. Nuclear Regulatory Commission (NRC) goal of limiting accidents with any fuel melting to such low probabilities. Such severe accidents include the category of Anticipated Transient Without Scram (ATWS) events mentioned above. Three accidents are usually analyzed to evaluate the reactor response in these cases; the unprotected (unscrammed) loss-of-flow (ULOF), where pumping power is lost and the pumps coast down, reducing coolant flow through the reactor core; the unprotected transient overpower (UTOP), where a control rod is inadvertently withdrawn from the core; and the unprotected loss-of-heat-sink (ULOHS), where the steam generator is isolated from the reactor in response to a turbine trip. For each of these accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection

  5. A study on the regulatory approach of major technical issues

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Oh, S. H.; Kang, H. J.; Kim, G. S. [Sunmoon Univ., Asan (Korea, Republic of); Lee, S. H.; Baek, W. P.; Yang, S. H.; Jeong, Y. H. [Korea Association for Nuclear Technology, Taejon (Korea, Republic of)

    1999-02-15

    This project is to provide the regulatory direction of 4 major technical issues for the Korean Next Generation Reactors, which are parts of major technical issues resulted from the safety regulation R and D on the KNGR. The outstanding results are are as follows : related to the classification and acceptance criteria of plant conditions, currently used classification and safety requirement were reviewed and regulatory direction was proposed. Among multiple failures, it is identified that SBO, TLOFW, multiple SGTR and ATWS are basically to be considered for additional requirements for advanced reactors. This study reviewed risk aspects, design consideration, and trends of safety requirements, and proposed fundamental safety requirements to be applied for KNGR. Multiple steam generator tube failure is a significant safety concern because of the possibility of release of radionuclides to the environment through containment bypass. Proposed safety requirement for this event can be categorized mainly as analysis requirement, design evaluation requirement and PSA requirement; For protection of containment failure, a reasonable safety position is necessary through and integrated review of possibility of severe accident occurrence, effects of sever accident mitigation features and cost effects of these design features. With this consideration safety requirements developed are the analysis requirement, provision of protective measures and survivability/availability of protective measures.

  6. ALMOD-JRC computer program

    International Nuclear Information System (INIS)

    This paper discusses the details concerning the newly developed or modified models of the computer program ALMOD-JRC, originating from ALMOD 3/Rel 4. The most important argument for the implementation of the new models was the need to enlarge the spectrum of the simulated phenomena, and to improve the simulation of experimental facilities such as LOFT or LOBI. This has led to a better formulation of the heat transfer and pressure drops correlations and to the implementation of the treatment of the heat losses to structural materials. In particular a series of test cases on real power plants, a pre-test examination of a LOBI station blackout ATWS experiment and the post test analysis of the L9-3 experiment, show the ability of ALMOD-JRC to correctly simulate PWR incident sequences. Although in ALMOD-JRC the code capabilities have been expanded, the limitations of the original version of the program still hold for what concerns the treatment of the coolant thermohydraulics as homogeneous flow for the two phase conditions in the primary coolant circuit. The other interesting feature of the new code is the remarkably shorter running times obtained with the introduction of simplified numerical treatments for the solving equations, without significant loss of accuracy of results

  7. Nuclear research and nuclear technology in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    The atomwirtschaft-atomtechnik has reflected the development of this quarter century. In this jubilee edition it describes the future lines of development. It has invited the Federal German companies and institutions of the branch to present their performance potential in the form of monography - more detailed than usually. This invitation was accepted by 81 of the most important enterprises. The figure also includes a number of important service companies, the research centres of the country, and last not least, a number of energy supply enterprises. Part 2 of this jubilee edition as a whole offers a crossection of the present performances offered in the German nuclear research, nuclear techniques, and the planning and service belonging to nuclear power operation. For the English-speaking readers, a digest part was set up in part 3 of the present edition. In part 4, the reader will find a product index in German and English. Each key-word indicates an offering firm by the page number allocated. Access to the monographies (part 2) and the digest (part 3) can be found in the listing of the monography-advertisers from page 102 on. The atw-jubilee edition closes with part 5, with product advertisements of companies from home and abroad. (orig./UA)

  8. Development of Integrated Evaluation System for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Ha; Kim, K. R.; Park, S. H.; Park, S. Y.; Park, J. H.; Song, Y. M.; Ahn, K. I.; Choi, Y

    2007-06-15

    The objective of the project is twofold. One is to develop a severe accident database (DB) for the Korean Standard Nuclear Power plant (OPR-1000) and a DB management system, and the other to develop a localized computer code, MIDAS (Multi-purpose IntegrateD Assessment code for Severe accidents). The MELCOR DB has been constructed for the typical representative sequences to support the previous MAAP DB in the previous phase. The MAAP DB has been updated using the recent version of MAAP 4.0.6. The DB management system, SARD, has been upgraded to manage the MELCOR DB in addition to the MAAP DB and the network environment has been constructed for many users to access the SARD simultaneously. The integrated MIDAS 1.0 has been validated after completion of package-wise validation. As the current version of MIDAS cannot simulate the anticipated transient without scram (ATWS) sequence, point-kinetics model has been implemented. Also the gap cooling phenomena after corium relocation into the RPV can be modeled by the user as an input parameter. In addition, the subsystems of the severe accident graphic simulator are complemented for the efficient severe accident management and the engine of the graphic simulator was replaced by the MIDAS instead of the MELCOR code. For the user's convenience, MIDAS input and output processors are upgraded by enhancing the interfacial programs.

  9. An innovative design approach to a cost effective commercial liquid metal reactor

    International Nuclear Information System (INIS)

    the SGHARS (Steam Generator Auxiliary Removal System) which uses the steam generator. In the abnormal event of the loss of feedwater we can use the RVACS (Reactor Vessel Auxiliary Cooling System). In a pool configuration, decay heat is transferred by conduction to the vessel and then by radiation and conduction to the guard vessel. The guard vessel is cooled initially by evaporative cooling followed by natural draft air cooling of the external surface. This approach is similar to the one taken for the AP1000. Finally, another innovative feature is the inclusion of a tertiary shutdown system. Thanks to the inherent reactivity feedbacks response the LMR can be brought to hot shutdown even when both the primary and secondary shutdown systems are unavailable. However, the core will experience high transient temperatures, before they are finally brought down by the reactivity feedbacks. The designer has to account for these high temperatures attained during ATWS (Anticipated Transients Without Scram). Since an ATWS is by definition the absence of both the primary and secondary shutdown systems, adoption of a tertiary system will eliminate the ATWS from design considerations. The system considered is different from the other two shutdown systems both in terms of absorber configuration and mode of insertion. The probability of simultaneous failure of all three diverse systems is less than the 10-8 threshold. Elimination of ATWS design considerations allows a power increase of the order of 10%, while the capital cost increase associated with the tertiary system is less than 1%. In summary our approach to achieve economic competitiveness with advanced LWRs is very simple: design the ARR with the same philosophy and within similar constraints. This is implemented by the innovative components presented here. (authors)

  10. Recent improvements of reactor physics codes in MHI

    Energy Technology Data Exchange (ETDEWEB)

    Kosaka, Shinya, E-mail: shinya-kosaka@mhi.co.jp; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki [Mitsubishi Heavy Industries, Ltd. (Japan)

    2015-12-31

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO’s Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  11. INCOGEN pre-feasibility study. Nuclear cogeneration

    International Nuclear Information System (INIS)

    The Netherlands Programme to Intensify Nuclear Competence (PINK, abbreviated in Dutch) supported the technical and economical evaluation of a direct cycle High Temperature Reactor (HTR) installation for combined heat and power generation. This helium cooled, graphite moderated HTR based on the German HTR-M, is named INCOGEN (Inherently safe Nuclear COGENeration). The INCOGEN reference is a 40 MW HTR design by the US company Longmark Power International (LPI). The energy conversion system comprises a single-shaft helium turbine-compressor (2.3-1.0 MPa) directly coupled with a 16.5 MW generator, a recuperator and low-temperature (150C to 40C) heat exchangers (23 MW). Spherical fuel elements (60 mm diameter) will be added little by little, which keeps the core only marginally critical. Void core volume can accommodate added fuel for several years until defuelling. Analyses of failure scenarios (loss of coolant accident or LOCA, loss of flow accident or LOFA, anticipated transient without scram or ATWS) show no excess of maximum acceptable fuel temperature of 1600C. Scoping analyses indicate no severe graphite fires. Transient analyses of the turbine-compressor system indicate adequate control flexibility. Optimization and endurance testing of the helium turbine-compressor is recommended

  12. Green tea consumption after intense taekwondo training enhances salivary defense factors and antibacterial capacity.

    Directory of Open Access Journals (Sweden)

    Shiuan-Pey Lin

    Full Text Available The aim of this study was to investigate the short-term effects of green tea consumption on selected salivary defense proteins, antibacterial capacity and anti-oxidation activity in taekwondo (TKD athletes, following intensive training. Twenty-two TKD athletes performed a 2-hr TKD training session. After training, participants ingested green tea (T, caffeine 6 mg/kg and catechins 22 mg/kg or an equal volume of water (W. Saliva samples were collected at three time points: before training (BT-T; BT-W, immediately after training (AT-T; AT-W, and 30 min after drinking green tea or water (Rec-T; Rec-W. Salivary total protein, immunoglobulin A (SIgA, lactoferrin, α-amylase activity, free radical scavenger activity (FRSA and antibacterial capacity were measured. Salivary total protein, lactoferrin, SIgA concentrations and α-amylase activity increased significantly immediately after intensive TKD training. After tea drinking and 30 min rest, α-amylase activity and the ratio of α-amylase to total protein were significantly higher than before and after training. In addition, salivary antibacterial capacity was not affected by intense training, but green tea consumption after training enhanced salivary antibacterial capacity. Additionally, we observed that salivary FRSA was markedly suppressed immediately after training and quickly returned to pre-exercise values, regardless of which fluid was consumed. Our results show that green tea consumption significantly enhances the activity of α-amylase and salivary antibacterial capacity.

  13. 3-D space time kinetics of compact high temperature reactor with fuel temperature feedback

    International Nuclear Information System (INIS)

    The Compact High Temperature Reactor (CHTR) is being developed as technology demonstrator for Indian High Temperature Reactor programme. Physics design of conceptual core of (Th-233U) fuelled CHTR is in advance stage and various core configurations have been proposed. Reactor core operation at high temperature necessitates sophisticated safety and anticipated transients analyses including postulated LORA, LOCA, and power set-back transients in CHTR. Recently, efficient IQS module in ARCH with adiabatic fuel temperature feedback capability has been developed. For accounting fuel and coolant temperature feedbacks in the simulation of 3D space time transients in CHTR, module for 1D (radial) heat conduction based module for heat transfer from fuel to coolant has been incorporated in 3D space-time analysis code ARCH. The AER benchmarking results of ARCH-IQS code with Doppler feedback and results of anticipated transient without scram (ATWS) of (Th-233U) fuelled CHTR with the present capability in ARCH-IQS code have been presented in this paper. (author)

  14. Preliminary Investigation of the Soluble Boron Free AP 1000 Core with the BigT Burnable Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Yahya, Mohd-Syukri; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, HyeongHeon [KEPCO Engineering and Construction Company., Inc., Daejeon (Korea, Republic of)

    2014-10-15

    The measurement of the U and Pu peak ratio provides information on the relative concentration of U and Pu elements. Photon measurements of spent nuclear fuel using high resolution spectrometers show a large background continuum in the low energy x-ray region in large part from Compton scattering of energetic gamma-rays. The high Compton continuum can make measurements of plutonium x-rays difficult because the relatively small signal to background ratio produced. According to the performance of the MCNPX simulation, the suppression ratios for the measurements of spent nuclear fuels were more than a factor of five. This result shows the feasibility of a Compton suppression system to the XRF technique. Many advanced PWRs are required to have a 24-month operating cycle to improve plant economy, and to keep the boron concentration low to allow an adequately negative moderator feedback during any ATWS event through 100% core life. Too much boron, typically greater than 1,300 ppm at full power, will make the MTC positive. The optimal design of burnable absorbers is key to the feasibility of this extended cycle and low boron core below the design limit of peak pin power. New concepts for burnable absorbers include changing the materials and geometry in the burnable absorber. k{sub inf}, peaking factor, MTC, and control rod worth of new BAs were compared with those of the conventional BA.

  15. Dependability Evaluation of Advanced Diverse Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Yang Gyun; Lee, Yoon Hee; Sohn, Se Do; Baek, Seung Min [KEPCO, Daejeon (Korea, Republic of); Lee, Sang Jeong [Chungnam National University, Daejeon (Korea, Republic of)

    2014-08-15

    For the mitigation of anticipated transients without scram (ATWS) as well as common cause failure (CCF) within the plant protection system (PPS) and the emergency safety feature . component control system (ESF-CCS), the diverse protection system (DPS) has been designed by KEPCO Engineering and Construction Company. Recently KEPCO E and C has developed the advanced diverse protection system (ADPS), which has four redundant channels, in an attempt to enhance a fault-tolerant capability of the system. For the evaluation of overall system improvement effects of the ADPS compared with the DPS, the dependability evaluation results are described herein. For all dependability attributes, this paper suggests a practical dependability evaluation method which uses quantitative dependability scores and indices. An overall dependability evaluation index (DEI) for the ADPS is evaluated with the average value of reliability/ security/maintainability/safety indices (i.e., RID, SID, MID, and SID') for dependability. The evaluation results show that the DEI value of ADPS can be improved by approximately 23% compared with that of the DPS, thanks to its fault-tolerant system architecture, software design changes, and external interface design features. Several suggestions have been made, in this paper, of an overall quantitative dependability evaluation method for the nuclear instrumentation and control (I and C) systems including the DPS and ADPS, and the usefulness of dependability evaluation on nuclear I and C systems has been confirmed.

  16. Development of Risk Management Technology/Development of Risk-Informed Application Technology

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Joon Eon; Kim, K. Y.; Ahn, K. I.; Lee, Y. H.; Lim, H. G.; Jung, W. S.; Choi, S. Y.; Han, S. J.; Ha, J. J.; Hwang, M. J.; Park, S. Y.; Yoon, C

    2007-06-15

    This project aims at developing risk-informed application technologies to enhance the safety and economy of nuclear power plant altogether. For this, the Integrated Level 1 and 2 PSA model is developed. In addition, the fire and internal flooding PSA models are improved according to the PSA standard of U.S.A. To solve the issues of domestic PSA model, the best-estimate thermal hydraulic analyses are preformed for the ATWS and LSSB. In order to reduce the uncertainty of PSA, several new PSA technologies are developed: (1) more exact quantification of large fault tree, (2) importance measure including the effects of external PSA. As feasibility studies of Option 2 and 3, the class of 6 systems' SSC are re-classified based on the risk information and the sensitivity analyses is performed for the EDG starting time, respectively. It is also improved that the methodology to identify the vital area of NPP. The research results of this project can be used in the regulatory body and the industry projects for risk-informed applications.

  17. ECN contributions to ADTT `96

    Energy Technology Data Exchange (ETDEWEB)

    Koning, A.J.

    1996-07-01

    An outline is presented of the status of nuclear data evaluation for accelerator-driven systems. The international effort consists mainly of measuring, compiling and calculating nuclear data for elements and isotopes relevant for transmutation of radioactive waste (ATW), energy amplification and other accelerator-related nuclear applications. We argue that input for global, macroscopic calculation schemes for hybrid nuclear systems basically should consist of three types of nuclear data: (a) High-energy transport codes for energies above about 150 MeV, (b) neutron and proton transport data files for energies below about 150 MeV and (c) neutron and proton transmutation/activation libraries below about 150 MeV. Our specific contribution to the field concerns (b) and (c). The progress of the evaluation of high-energy nuclear data files for the most important materials and the related compilation of nuclear reaction information is reported. The evaluated data are calculated with the computer codes ECIS95, MINGUS and GNASH and are stored in ENDF6-format. We illustrate the library production with a short outline of the employed physical methods. Finally, we briefly discuss the application of the activation/transmutation library ECNAF96. (orig.).

  18. A study on reactor core failure thresholds to safety operation of LMFBR

    International Nuclear Information System (INIS)

    Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

  19. Recent improvements of reactor physics codes in MHI

    Science.gov (United States)

    Kosaka, Shinya; Yamaji, Kazuya; Kirimura, Kazuki; Kamiyama, Yohei; Matsumoto, Hideki

    2015-12-01

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO's Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented.

  20. YKAe - Research programme on nuclear power plant systems behaviour and operational aspects of safety

    International Nuclear Information System (INIS)

    The major part of nuclear energy research in Finland has been organised as five-year nationally coordinated research programs. The research programme on Systems Behaviour and Operational Aspects of Safety is under way during 1990-1994. Its annual volume has been about 35 person-years and its annual expenditure about FIM 18 million. Studies in the field on safe operational margins of nuclear fuel and reactor core concentrate on fuel high burn-up behaviour, VVER fuel experiments, and reactor core behaviour in complex reactivity transients such as 3-D phenomena and ATWS events. The PACTEL facility is used for the thermal hydraulic studies of the Loviisa type reactors (scaled 1:305). Validation of accident analysis codes is carried out by participation in international standard problems. Advanced foreign computer codes for severe reactor accidents are implemented, modified as needed and applied to level-2 PSAs and the improvement of accident management procedures. Fire simulation methods are tested using data from experiments in the German HDR facility. A nuclear plant analyzer for efficient safety analyses is being developed using the APROS process simulation environment. Computerized operator support systems are being studied in cooperation with the OECD Halden Project. The basic factors affecting plant operator activities and the development of their competence are being investigated. A comprehensive system for the control of plant operational safety is being developed by combining living PSA and safety indicators

  1. Waste management and final storage in Germany - failed for lack of content and a technical basis?

    International Nuclear Information System (INIS)

    The assertion by the political parties at present in government in Germany, SPD and Alliance 90/The Greens, that 'the previous waste management concept for radioactive waste had failed in terms of content and no longer had any technical basis', is a purely ideological statement utterly devoid of any realistic reason. In actual fact, the waste management concept so far pursued in Germany has been transferred into industrial practice in many areas: Transports of radioactive waste and spent fuel elements can be carried out safely at any time; spent fuel has been reprocessed on an industrial scale for many years. The central interim stores of Ahaus, Gorleben, and Lubmin, all of which are in operation, actually represent sufficient capacity for the interim storage of spent fuel elements. The successful exploration of the Gorleben salt dome has advanced far. No result so far would detract from its suitability. Consequently, the federal government should not try 'to elaborate a (new) national waste management plan for the inherited burden of radioactive waste', but rather invest all its power to make functional as quickly as possible the missing building blocks in the existing waste management concept. In doing so, it would make an important contribution to domestic peace and to the international recognition of Germany as a high-tech country. Part 1 of the article covers reprocessing and interim storage, while part 2, which will be published in atw 8/9, will be about problems of final storage. (orig.)

  2. Thermal hydraulic analysis of LANL/IPPE/EDO-GP 1 MW LBE target

    International Nuclear Information System (INIS)

    The authors have carried out numerical simulations of the thermal hydraulic behavior of the LANL/IPPE/EDO-GP neutron spallation to be tested in LANSCE beam in year 2,000. The LANL/IPPE/EDO-GP target design, along with energy deposition provided by the Los Alamos meson Physics Facility (LAMPF) accelerator beam, has been analyzed with two commercial computational fluid dynamic (CFD) computer codes at Los Alamos. The LAMPF proton beam deposits energy in the target window diaphragm as well as the liquid metal coolant target. The computer codes were used to determine the maximum temperature in the spallation target diaphragm. The computational results from the two CFD codes are in general agreement and are consistent with the preliminary IPPE analysis. Limited studies were also performed to investigate ways to enhance cooling of the target by directing more flow through the center of the target. The goal of this latter part of the study is to provide guidance for the future design of a spallation target for the ATW project

  3. Reel-to-reel continuous deposition of CexZr1-xO2 single buffer layer for YBCO coated conductors

    International Nuclear Information System (INIS)

    In this paper, a study regarding the epitaxial growth of CexZr1-xO2 film on biaxially textured Ni-5at.%W substrate and its use as a single buffer layer of a YBCO coated conductors was reported. Films of Ce-Zr mixed oxide were prepared by direct-current (d.c.) reactive magnetron sputtering with the two sputtering guns arranged symmetrically with respect to the substrate. In sputtering process, d.c. power of Zr was fixed in 200 W while that of Ce was varied with 30 W, 50 W, 75 W, and 100 W, respectively. It was confirmed that the composition of the films could be controlled with modulating power of Ce target. All samples exhibited good epitaxial growth with c-axis perpendicular to the substrate surface. Atomic force microscope revealed a continuous, dense, and crack-free surface morphology for Ce0.32Zr0.68O2 thin films, which provided themselves as the good single buffer to the YBa2Cu3O7-δ (YBCO) coated conductors. High quality Ce0.32Zr0.68O2 buffer layers up to 100-m length could be fabricated with production speed of about 1.2m/h. X-ray scans have been performed as a function of length to determine the crystallographic consistency of the epitaxial Ce0.32Zr0.68O2 over length.

  4. The mechanism of the nano-CeO2 films deposition by electrochemistry method as coated conductor buffer layers

    International Nuclear Information System (INIS)

    Highlights: • Crack-free CeO2 film thicker than 200 nm was prepared on NiW substrate by ED method. • Different electrochemical processes as hydroxide/metal mechanisms were identified. • The CeO2 precursor films deposited by ED method were in nano-scales. - Abstract: Comparing with conventional physical vapor deposition methods, electrochemistry deposition technique shows a crack suppression effect by which the thickness of CeO2 films on Ni–5 at.%W substrate can reach a high value up to 200 nm without any cracks, make it a potential single buffer layer for coated conductor. In the present work, the processes of CeO2 film deposited by electrochemistry method are detailed investigated. A hydroxide reactive mechanism and an oxide reactive mechanism are distinguished for dimethyl sulfoxide and aqueous solution, respectively. Before heat treatment to achieve the required bi-axial texture performance of buffer layers, the precursor CeO2 films are identified in nanometer scales. The crack suppression for electrochemistry deposited CeO2 films is believed to be attributed to the nano-effects of the precursors

  5. YBa2Cu3O7-x films prepared by TFA-MOD method for coated conductor application

    International Nuclear Information System (INIS)

    The epitaxial growth of YBCO films both on (001) SrTiO3 (STO) and Ni-W biaxially textured metallic substrates prepared by metal-organic deposition (MOD) using a trifluoroacetic acid (TFA) solution is reported. The degree of epitaxy of the YBCO films was investigated by X-ray diffraction and scanning electron microscopy (SEM). The as deposited films exhibit good morphological and structural properties. The ω-scan of the YBCO films grown on (001) SrTiO3 single crystal substrate and on Pd/CeO2/YSZ/CeO2 buffered biaxially textured Ni-5at%W (Ni-W) tapes has a full width at half maximum (FWHM) of 0.120 and 3.40, respectively. The φ-scan of (113) peak of YBCO film grown on Ni-W substrate has FWHM of 6.10. The YBCO/STO film has a zero resistance critical temperature of Tc(R = 0) = 92 K and a critical current density Jc > 2 MA/cm2 at 77 K and in zero magnetic field

  6. Annual conference on nuclear technology. Nuclear power 2001: option for the future

    International Nuclear Information System (INIS)

    The Dresden Palace for Culture was the venue of the ANNUAL MEETING ON NUCLEAR TECHNOLOGY on May 15-17, 2001, the first to be held in Dresden and the first also to be held in one of the new German federal states. Although no nuclear plant is in operation in East Germany after the Greifswald Nuclear Power Station was decommissioned, nuclear technology continues to play an important role especially in research and university teaching in this part of Germany. The organizers of the conference, Deutsches Atomforum e.V. (DAtF) and Kerntechnische Gesellschaft e.V. (KTG), welcomed more than 1000 participants from nineteen countries. The three-day program, with its traditional, proven structure, featured plenary sessions on the first day, and specialized sessions, technical sessions, poster sessions, and other events on the following days. The partner country at the Annual Meeting on Nuclear Technology was Russia, with a session specially devoted to selected topics of the country. The conference was accompanied by a technical exhibition with company meeting points of vendors, suppliers, and service industries. A video film forum was arranged for the interested public which featured contributions about nuclear research, nuclear power plant operation, transport and storage as well as decommissioning. Another major event was a workshop on 'Preserving Competence in Nuclear Technology'. The plenary day is described in this summary report, while the results of the technical sessions as seen by the rapporteurs are printed elsewhere in this issue of atw 8/9, 2001. (orig.)

  7. Electrochemical Treatment of Synthetic and Actual Dyeing Wastewaters Using BDD Anodes

    Directory of Open Access Journals (Sweden)

    Nasr Bensalah

    2010-06-01

    Full Text Available In this work, the treatment of synthetic wastewaters containing methylene blue (MB and rhodamine B (RB and actual textile wastewaters (ATW using boron doped diamond (BDD anodic oxidation was investigated. Voltammetric study has shown that both MB and RB can be oxidized directly at the anode surface in the potential region where the electrolyte salt is stable. Galvanostatic electrolyses of synthetic and actual industrial wastewaters have led to total abatement of COD and TOC at different operating conditions (electrolyte salt and initial pollutant concentration and current density and the efficiency of the electrochemical process was governed only by mass-transfer limitations. The nature of the supporting electrolyte has a great influence on the rate and the efficiency of the electrochemical oxidation of dyes. The treatment in the presence of NaCl appears to be more efficient in the COD removal, while in the presence of Na2SO4 improves the TOC removal. From the experimental results it seems that the primary mechanisms in the oxidation of dyes are the mediated electro-oxidation by hydroxyl radicals and other oxidants electro-generated from supporting electrolyte oxidation.

  8. Review and comment on the advanced spent fuel management process (1): Technical aspects and non-proliferation concerns

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yo Taik

    2001-01-01

    Efforts are made to analyze the project, the Advanced Spent Fuel Management Technology (ASFMT), which is currently carried out at Korea Atomic Energy Research Institute, on the technical feasibility and validity as well as on the nuclear non-proliferation concerns. The project is a part of a program under the 'Long and Midterm Nuclear Development Program'. On the technical analysis, reviewed the papers presented at the national and international meetings on the subject by KAERI staffs, and also participated to various technical discussions on the 'Mock-up Test', currently in progress. On the non-proliferation concerns, the ASFMT project was reviewed and analyzed in reference to various programs currently in progress or in a formulation stages in US, such as the DOE TOPS and ATW. Further reviewed the past JASNEC process and programs for possible application of the ASFMT project for JASNEC project. Provided a few thoughts for effectively carrying out the ASFMT project, and a plan for the next phase is presented.

  9. The mechanism of the nano-CeO{sub 2} films deposition by electrochemistry method as coated conductor buffer layers

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Yuming; Cai, Shuang [Department of Physics, Shanghai University, Shanghai 200444 (China); Shanghai Key Laboratory of High Temperature Superconductors, Shanghai 200444 (China); Liang, Ying, E-mail: yliang@ecust.edu.cn [Institute of Nuclear Technology and Application, School of Science, East China University of Science and Technology, Shanghai 200237 (China); Bai, Chuanyi; Liu, Zhiyong; Guo, Yanqun; Cai, Chuanbing [Department of Physics, Shanghai University, Shanghai 200444 (China); Shanghai Key Laboratory of High Temperature Superconductors, Shanghai 200444 (China)

    2015-05-15

    Highlights: • Crack-free CeO{sub 2} film thicker than 200 nm was prepared on NiW substrate by ED method. • Different electrochemical processes as hydroxide/metal mechanisms were identified. • The CeO{sub 2} precursor films deposited by ED method were in nano-scales. - Abstract: Comparing with conventional physical vapor deposition methods, electrochemistry deposition technique shows a crack suppression effect by which the thickness of CeO{sub 2} films on Ni–5 at.%W substrate can reach a high value up to 200 nm without any cracks, make it a potential single buffer layer for coated conductor. In the present work, the processes of CeO{sub 2} film deposited by electrochemistry method are detailed investigated. A hydroxide reactive mechanism and an oxide reactive mechanism are distinguished for dimethyl sulfoxide and aqueous solution, respectively. Before heat treatment to achieve the required bi-axial texture performance of buffer layers, the precursor CeO{sub 2} films are identified in nanometer scales. The crack suppression for electrochemistry deposited CeO{sub 2} films is believed to be attributed to the nano-effects of the precursors.

  10. Updating the Finite Element Model of the Aerostructures Test Wing using Ground Vibration Test Data

    Science.gov (United States)

    Lung, Shun-fat; Pak, Chan-gi

    2009-01-01

    Improved and/or accelerated decision making is a crucial step during flutter certification processes. Unfortunately, most finite element structural dynamics models have uncertainties associated with model validity. Tuning the finite element model using measured data to minimize the model uncertainties is a challenging task in the area of structural dynamics. The model tuning process requires not only satisfactory correlations between analytical and experimental results, but also the retention of the mass and stiffness properties of the structures. Minimizing the difference between analytical and experimental results is a type of optimization problem. By utilizing the multidisciplinary design, analysis, and optimization (MDAO) tool in order to optimize the objective function and constraints; the mass properties, the natural frequencies, and the mode shapes can be matched to the target data to retain the mass matrix orthogonality. This approach has been applied to minimize the model uncertainties for the structural dynamics model of the Aerostructures Test Wing (ATW), which was designed and tested at the National Aeronautics and Space Administration (NASA) Dryden Flight Research Center (DFRC) (Edwards, California). This study has shown that natural frequencies and corresponding mode shapes from the updated finite element model have excellent agreement with corresponding measured data.

  11. Chemically deposed layer sytems for the realization of YBa2Cu3O7-δ band conductors

    International Nuclear Information System (INIS)

    The aim of this thesis was to produce new buffer-layer systems for biaxially texturated Ni5at%W substrates by means of chemical processes. As very promising materials for the buffer layers CaTiO3 and SrTiO3 were chosen. The production of the single layers pursued from the organometallic prestage by means of dip coating and subsequent head treatment. During the work first the single precursor solutions were to be developed. A main component of the theses forms the understanding of the texture development during the heat treatment of precursor layers on biaxially texturated metallic substrates. Based on this the growth of thick buffer layers is studied and by means of YBCO layers, which were deposed by beans of a pulsed laser, the functionality of the synthesized buffer layers proved. A further component of this thesis formes the influence of nanoscaling precipitations in thew YBCO on its superconducting properties. The YBCO deposition pursued via a variation of the TFA process, as substrate (001)-oriented SrTiO3 monocrystals were applied

  12. High order boron transport scheme in TRAC-BF1

    International Nuclear Information System (INIS)

    In boiling water reactors (BWR), unlike pressurized water reactors (PWR) in which the reactivity control is accomplished through movement of the control rods and boron dilution, the importance of boron transport lies in maintaining the core integrity during ATWS-kind severe accidents in which under certain circumstances a boron injection is required. This is the reason for implementing boron transport models thermal-hydraulic codes as TRAC-BF1, RELAP5 and TRACE, bringing an improvement in the accuracy of the simulations. TRAC-BF1 code provides a best estimate analysis capability for the analysis of the full range of postulated accidents in boiling water reactors systems and related facilities. The boron transport model implemented in TRAC-BF1 code is based on a calculation according to a first order accurate upwind difference scheme. There is a need in reviewing and improving this model. Four numerical schemes that solve the boron transport model have been analyzed and compared with the analytical solution that provides the Burgers equation. The studied numerical schemes are: first order Upwind, second order Godunov, second-order modified Godunov adding physical diffusion term and a third-order QUICKEST using the ULTIMATE universal limiter (UL). The modified Godunov scheme has been implemented in TRAC-BF1 source code. The results using these new schemes are presented in this paper. (author)

  13. Demonstration of control rod holding stability of the self actuated shutdown system in Joyo for enhancement of fast reactor inherent safety

    International Nuclear Information System (INIS)

    Self actuated shutdown system (SASS) with a Curie point electromagnet (CPEM) has been developed for use in a large-scale liquid metal cooled fast breeder reactor (LMFBR) in order to establish the passive shutdown capability against anticipated transient without scram (ATWS) events. The basic characteristics of SASS have already been investigated by various out-of-pile tests for material elements. As the final stage of the development, the stability of SASS needs to be confirmed under the actual reactor-operational environment with high temperature, high neutron flux, and flowing sodium to ensure the high plant availability factor. For this purpose, the demonstration test of holding stability using the reduced-scale experimental equipment of SASS was conducted in the 1st and 2nd operational cycles of the experimental fast reactor Joyo MK-III. The rod-holding stability and the rod-recovering functions of the driving system to re-connect and pull out the separated control rod were fully confirmed. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation. (author)

  14. Radioactive wastes. From where, how much, to where?; Radioaktive Abfaelle. Woher, wieviel, wohin?

    Energy Technology Data Exchange (ETDEWEB)

    Ammann, M

    2008-09-15

    This report helps to the popularization of the Nagra's works accomplished for the management and disposal of the radioactive wastes in Switzerland. The radioactive wastes are partitioned into 3 different types: high level waste (HLW), alpha-toxic waste (ATW) and low- and intermediate-level waste (L/ILW). Most of the radioactive wastes are produced in the nuclear power plants, but also by many applications in medicine, industry and research. They have to be correctly disposed of. Mankind and environment have to be protected against them in the long term. The type and quantity of the wastes are accurately known. At the nuclear power plants as well as in the central storage pool of the Zwilag AG and in the federal interim storage facility in Wuerenlingen, there is enough storage capacity for all radioactive wastes in Switzerland. Radioactive wastes can be safely disposed of in deep geological repositories for a time period long enough that the radioactivity is reduced to natural values. Nagra has proved the feasibility of such repositories and its results were accepted by the Federal Council.

  15. Selection of flowing liquid lead target structural materials for accelerator driven transmutation applications

    International Nuclear Information System (INIS)

    The beam entry window and container for a liquid lead spallation target will be exposed to high fluxes of protons and neutrons that are both higher in magnitude and energy than have been experienced in proton accelerators and fission reactors, as well as in a corrosive environment. The structural material of the target should have a good compatibility with liquid lead, a sufficient mechanical strength at elevated temperatures, a good performance under an intense irradiation environment, and a low neutron absorption cross section; these factors have been used to rank the applicability of a wide range of materials for structural containment Nb-1Zr has been selected for use as the structural container for the LANL ABC/ATW molten lead target. Corrosion and mass transfer behavior for various candidate structural materials in liquid lead are reviewed, together with the beneficial effects of inhibitors and various coatings to protect substrate against liquid lead corrosion. Mechanical properties of some candidate materials at elevated temperatures and the property changes resulting from 800 MeV proton irradiation are also reviewed

  16. Demonstration test of the holding stability of the self actuated shutdown system in the experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Self actuated shutdown system (SASS) with a Curie point electromagnet (CPEM) has been developed for use in a large scale fast breeder reactor (FBR) in order to establish the passive shutdown capability against anticipated transient without scram (ATWS) events. The basic characteristics of SASS have already been investigated by various out-of-pile tests for material elements. As the final stage of the development, the stability of SASS needs to be confirmed under the actual reactor-operational environment with high temperature, high neutron flux, and flowing sodium in order to ensure the high plant availability factor. For this purpose, the demonstration test of holding stability using the reduced-scale experimental equipment of SASS was conducted in the 1st and 2nd operational cycles of the experimental fast reactor JOYO MK-III. As a result of this study, the rod-holding stability and the rod-recovering functions of the driving system to re-connect and pull out the separated control rod were fully confirmed. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation. (author)

  17. Review of the Technical Status on the Debris Bed Cooling Model

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Cho, Chung Ho; Lee, Yong Bum

    2007-09-15

    Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double-fault initiators such as ATWS events without coolant boiling or fuel melting. However, for the future design of sodium cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth due consideration of triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. In this study, review of the technical status on the debris bed cooling model was carried out for in-vessel retention of the core debris0.

  18. Categorization of core-damage sequences by containment event tree analysis for boiling water reactor with Mark-II containment

    International Nuclear Information System (INIS)

    In the present study, containment responses to core damage accidents were analyzed for a large spectrum of core damage sequences, which were defined by front-line system event trees, in a BWR with Mark-11 containment by using the Accident Progression Event Tree (APSET) method and their characteristics were examined in terms of mainly probabilistic aspects such as their respective conditional probabilities of containment failure modes and accident termination. This paper showed that various core damage sequences could be categorized into a small number of groups, each of which consisted of the sequences with similar containment response characteristics, as follows: Interfacing system LOCA; ATWS with high pressure injection available; Loss of long-term containment heat removal; Station blackout; Loss of coolant injection with the reactor not depressurized; Loss of coolant injection with the reactor depressurized; Loss of short-term containment heat removal; and Reactor pressure vessel rupture. The above categorization provides a perspective on the potential containment failure modes and the effectiveness of some accident mitigative measures, which could be useful for studying accident management strategies and as well for assisting the analysts in carrying out future CET analyses. (author)

  19. Recent improvements of reactor physics codes in MHI

    International Nuclear Information System (INIS)

    This paper introduces recent improvements for reactor physics codes in Mitsubishi Heavy Industries, Ltd(MHI). MHI has developed a new neutronics design code system Galaxy/Cosmo-S(GCS) for PWR core analysis. After TEPCO’s Fukushima Daiichi accident, it is required to consider design extended condition which has not been covered explicitly by the former safety licensing analyses. Under these circumstances, MHI made some improvements for GCS code system. A new resonance calculation model of lattice physics code and homogeneous cross section representative model for core simulator have been developed to apply more wide range core conditions corresponding to severe accident status such like anticipated transient without scram (ATWS) analysis and criticality evaluation of dried-up spent fuel pit. As a result of these improvements, GCS code system has very wide calculation applicability with good accuracy for any core conditions as far as fuel is not damaged. In this paper, the outline of GCS code system is described briefly and recent relevant development activities are presented

  20. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

    International Nuclear Information System (INIS)

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs

  1. Successful retrieval of an unexpanded coronary stent from the left main coronary artery during primary percutaneous coronary intervention

    Directory of Open Access Journals (Sweden)

    Šalinger-Martinović Sonja

    2011-01-01

    Full Text Available Introduction. Dislodgement and embolization of the new generation of coronary stents before their deployment are rare but could constitute a very serious complication. Case Outline. We report a case of a stent dislodgement into the left main coronary artery during the primary coronary intervention of infarct related left circumflex artery in a patient with acute myocardial infarction. The dislodged and unexpanded bare-metal stent FlexMaster 3.0x19 mm (Abbot Vascular was stranded and bended in the left main coronary artery (LMCA, probably by the tip of the guiding catheter, but stayed over the guidewire. It was successfully retrieved using a low-profile Ryujin 1.25x15 balloon catheter (Terumo that was passed through the stent, inflated and then pulled back into the guiding catheter. After that, the whole system was withdrawn through the 6 F arterial sheath via the transfemoral approach. After repeated cannulation via the 6F arterial sheath, additional BMW and ATW guidewires were introduced into the posterolateral and obtuse marginal branches and a bare-metal stent Driver (Medtronic Cardiovascular Inc 3.0x18 mm was implanted in the target lesion. Conclusion. Stent dislodgement is a rare but potentially life-threatening complication of the percutaneous coronary intervention. This incident occurring in the LMCA in particular during an acute myocardial infarction requires to be urgently resolved. The avoidance of rough manipulation with the guiding catheter and delivery system may help in preventing this kind of complications.

  2. Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

    International Nuclear Information System (INIS)

    Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author)

  3. Parallel beam dynamics calculations on high performance computers

    International Nuclear Information System (INIS)

    Faced with a backlog of nuclear waste and weapons plutonium, as well as an ever-increasing public concern about safety and environmental issues associated with conventional nuclear reactors, many countries are studying new, accelerator-driven technologies that hold the promise of providing safe and effective solutions to these problems. Proposed projects include accelerator transmutation of waste (ATW), accelerator-based conversion of plutonium (ABC), accelerator-driven energy production (ADEP), and accelerator production of tritium (APT). Also, next-generation spallation neutron sources based on similar technology will play a major role in materials science and biological science research. The design of accelerators for these projects will require a major advance in numerical modeling capability. For example, beam dynamics simulations with approximately 100 million particles will be needed to ensure that extremely stringent beam loss requirements (less than a nanoampere per meter) can be met. Compared with typical present-day modeling using 10,000-100,000 particles, this represents an increase of 3-4 orders of magnitude. High performance computing (HPC) platforms make it possible to perform such large scale simulations, which require 10's of GBytes of memory. They also make it possible to perform smaller simulations in a matter of hours that would require months to run on a single processor workstation. This paper will describe how HPC platforms can be used to perform the numerically intensive beam dynamics simulations required for development of these new accelerator-driven technologies

  4. Sparking plasma sintering method for developing cube textured Ni7W/Ni12W/Ni7W multi-layer substrates used for coated conductors

    International Nuclear Information System (INIS)

    Mechanically strengthened, highly cube textured Ni-7at.%W/Ni-12at.%W multi-layer substrates used for coated conductors have been prepared by advanced spark plasma sintering technique. The key innovation for developing this weakly magnetic and reinforced substrate was to use a new powder metallurgy and sintering route to bond multi-layers of Ni7W-Ni12W-Ni7W together in order to get an initial ingot, then followed by optimized cold working and annealing. Particular efforts were made in view of the optimization of the design, pressing as well as the heat treatment processes of the starting ingots in order to obtain a chemically gradient composite bulk, thus ensuring the subsequent cold deformation of the bulk. The produced composite substrates have a strong {100} texture on the top Ni7W outer layer determined by EBSD and X-ray. The percentage of the biaxially orientated grains within a misorientation angle of 10 deg. is as high as 97.5%, while the length percentage of low angle GBs ranging from 2 deg. to 10 deg. in the composite substrate reaches 87.2%. Moreover, the yield strength σ0.2 of the tape approaches 333 MPa, and the saturation magnetization is substantially reduced by 81.6% at 77K when compared to that of a commercial used Ni5W substrate

  5. Risk Assessment Review Group report to the U.S. Nuclear Regulatory Commission

    International Nuclear Information System (INIS)

    The Risk Assessment Review Group was organized by the U.S. Nuclear Regulatory Commission on July 1, 1977, with four elements to its charter: clarify the achievements and limitations of WASH-1400, the ''Rasmussen Report''; assess the peer comments thereon, and responses to those comments; study the present state of such risk assessment methodology; and recommend to the Commission how (and whether) such methodology can be used in the regulatory and licensing process. Areas of study include: risk assessment methodologies; statistical issues; completeness; the data base; and the WASH-1400 assessment of the damage to human health from radiation after a postulated accident. Specific items discussed include: Browns Ferry; common cause failure; human factors; format and scrutability; the peer review process; earthquakes; risk perception; allegations by UCS concerning WASH-1400 treatment of quality assurance and quality control; current role of probabilistic methods in the regulatory process; acts of violence; ATWS; influence of design defects in quality assurance failures; and calculation of population doses from given releases of radionuclides

  6. The KSNPP risk-effect analysis of the digital safety-critical systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, H. G.; Eom, H. S.; Jang, S. C.; Ha, J. J

    2004-02-01

    The study was performed for evaluating the risk effect of digital systems on the total plant. Based on risk monitor, a fault tree model for the Korean Standard Nuclear Power Plants (KSNPP), we integrate the fault-tree models for Digital Plant Protection System (DPPS) and Digital Engineered SaFety Actuation System (DESFAS) which are the most important safety-critical I and C systems in the KSNPP. In this study, however, three important factors (the probabilities of manual actuation failure, the software failure probability, and the watchdog timer fault coverage) are treated as the variables of the sensitivity study because quantification methodologies for these factors are not developed yet. Not only the unavailability of digital safety-critical system itself, but also the risk effect of digital systems on the total plant should be assessed to prove the safety of digital systems. The result of sensitivity study shows that the Anticipated-Transient-Without-Scram (ATWS) frequency changes from 8.433E-06 (no./yr) to 1.269E-04. The Core-Damage Frequency (CDF) changes from 7.714E-06 to 9.994E-05. The Large-Early-Release Frequency (LERF) changes from 1.243E-05 to 6.744E-05.

  7. Environmental effects on fatigue of steels for structural parts in water-steam-circuits of light water reactors. Considerations concerning the question of transferability of results from laboratory tests to real operating conditions; Der Einfluss des Mediums auf Ermuedungsvorgaenge in Staehlen fuer Strukturbauteile in Wasser-Dampf-Kreislaeufen von Leichtwasserreaktoren. Ueberlegungen zur Frage der Uebertragbarkeit von Ergebnissen aus Laborversuchen auf den realen Anlagenbetrieb

    Energy Technology Data Exchange (ETDEWEB)

    Roth, Armin [AREVA NP GmbH, Erlangen (Germany)

    2008-07-01

    .B. von Staehlen, beeinflussen kann. Schon vor einigen Jahrzehnten wurde experimentell nachgewiesen, dass z.B. Hochtemperaturwasser in Laborversuchen zur Untersuchung des Ermuedungsverhaltens von Staehlen zu deutlichen Effekten fuehrt. Dabei wird je nach Versuchsfuehrung entweder die Zeit bis zur Initiierung von Anrissen verkuerzt oder die Wachstumsgeschwindigkeit vorhandener Risse erhoeht. Dieser zu erwartende Einfluss des Mediums auf den Ermuedungsprozess wurde in den Anfaengen der Regulierung von Konstruktion und Auslegung von Bauteilen und Komponenten fuer Kernkraftwerke in den relevanten Regelwerken (z.B. ASME Boiler and Pressure Vessel Code, Section III) weltweit nicht explizit beruecksichtigt. Pauschal beruecksichtigt werden Umgebungseffekte dagegen in den entsprechenden da/dN-Risswachstumskurven des ASME Code, Section XI zur Bewertung des betrieblichen Verhaltens von Oberflaechenfehlern. Historisch betrachtet erfolgte in den Regelwerken die Festlegung der Vorgehensweise zur Komponentenauslegung allerdings lange vor dem gezielten experimentellen Nachweis der Umgebungseffekte auf Rissinitiierung und Risswachstum durch Ermuedung von Staehlen in Hochtemperaturwasser. Trotz dieser Erkenntnis ist es weltweit nicht zu generischen, systematischen Schaeden in Medium fuehrenden Systemen von Leichtwasserreaktoren (LWR) durch Korrosionsermuedung infolge von Fehlauslegung gekommen. Vereinzelt aufgetretene Schaeden mit deutlichen Merkmalen umgebungsbeeinflusster Ermuedungsvorgaenge konnten immer eindeutig auf das Vorkommen von unerwarteten, nicht im spezifizierten Belastungskollektiv enthaltenen Betriebstransienten zurueckgefuehrt werden. Zu diesen Ursachen zaehlen z.B. das Auftreten von thermischer Schichtung oder lokale, stroemungsinduzierte Vibrationen. In diesem Beitrag werden Ueberlegungen vorgestellt, welche die zu beobachtende Diskrepanz zwischen der diesbezueglich weltweit ueberwiegend positiven Betriebserfahrung und den Ergebnissen aus Laborversuchen mit dem

  8. Environmental effects on fatigue of steels for structural parts in water-steam-circuits of light water reactors. Considerations concerning the question of transferability of results from laboratory tests to real operating conditions; Der Einfluss des Mediums auf Ermuedungsvorgaenge in Staehlen fuer Strukturbauteile in Wasser-Dampf-Kreislaeufen von Leichtwasserreaktoren. Ueberlegungen zur Frage der Uebertragbarkeit von Ergebnissen aus Laborversuchen auf den realen Anlagenbetrieb

    Energy Technology Data Exchange (ETDEWEB)

    Roth, Armin [AREVA NP GmbH, Erlangen (Germany)

    2008-07-01

    .B. von Staehlen, beeinflussen kann. Schon vor einigen Jahrzehnten wurde experimentell nachgewiesen, dass z.B. Hochtemperaturwasser in Laborversuchen zur Untersuchung des Ermuedungsverhaltens von Staehlen zu deutlichen Effekten fuehrt. Dabei wird je nach Versuchsfuehrung entweder die Zeit bis zur Initiierung von Anrissen verkuerzt oder die Wachstumsgeschwindigkeit vorhandener Risse erhoeht. Dieser zu erwartende Einfluss des Mediums auf den Ermuedungsprozess wurde in den Anfaengen der Regulierung von Konstruktion und Auslegung von Bauteilen und Komponenten fuer Kernkraftwerke in den relevanten Regelwerken (z.B. ASME Boiler and Pressure Vessel Code, Section III) weltweit nicht explizit beruecksichtigt. Pauschal beruecksichtigt werden Umgebungseffekte dagegen in den entsprechenden da/dN-Risswachstumskurven des ASME Code, Section XI zur Bewertung des betrieblichen Verhaltens von Oberflaechenfehlern. Historisch betrachtet erfolgte in den Regelwerken die Festlegung der Vorgehensweise zur Komponentenauslegung allerdings lange vor dem gezielten experimentellen Nachweis der Umgebungseffekte auf Rissinitiierung und Risswachstum durch Ermuedung von Staehlen in Hochtemperaturwasser. Trotz dieser Erkenntnis ist es weltweit nicht zu generischen, systematischen Schaeden in Medium fuehrenden Systemen von Leichtwasserreaktoren (LWR) durch Korrosionsermuedung infolge von Fehlauslegung gekommen. Vereinzelt aufgetretene Schaeden mit deutlichen Merkmalen umgebungsbeeinflusster Ermuedungsvorgaenge konnten immer eindeutig auf das Vorkommen von unerwarteten, nicht im spezifizierten Belastungskollektiv enthaltenen Betriebstransienten zurueckgefuehrt werden. Zu diesen Ursachen zaehlen z.B. das Auftreten von thermischer Schichtung oder lokale, stroemungsinduzierte Vibrationen. In diesem Beitrag werden Ueberlegungen vorgestellt, welche die zu beobachtende Diskrepanz zwischen der diesbezueglich weltweit ueberwiegend positiven Betriebserfahrung und den Ergebnissen aus Laborversuchen mit dem

  9. VAK Kahl - decommissioning and demolition continued under new auspices; VAK Kahl - Fortsetzung des Rueckbaus unter neuem Vorzeichen

    Energy Technology Data Exchange (ETDEWEB)

    Hackel, W.; Runge, H. [RWE NUKEM GmbH, Alzenau (Germany)

    2001-11-01

    The Kahl experimental nuclear power station (VAK), the first German nuclear power plant, was decommissioned after 25 years of operation (1961 to 1985). The BWR plant generated approx. 2 million kWh of electricity in 150,000 hours of operation at a gross power of MWe. After the operator, VAK GmbH, had filed an application for decommissioning, the first of four decommissioning permits was issued in 1988. The plant is to be demolished completely so that the site will no longer be within the scope of the Atomic Energy Act. By 2001, demolition work covered by the first decommissioning permit had been finished, also the 2nd and 3rd decommissioning permits had largely been completed, and work under the 4th decommissioning permit had been begun. To acquire technical and organization experience and know-how, the decommissioning and demolition phases are accompanied by research and development work carried out by the operators and by VAK shareholders RWE and E.ON. After the bulk of the work had been completed, the radioactive inventory had been removed from the plant, and the end of the project was in sight, RWE NUKEM GmbH was commissioned to carry on. The main objectives now are speedy completion of the jobs still to be finished, further development for other projects of the know-how acquired, and job protection. The main work still to be carried out includes dismantling of systems no longer needed and of the biological shield as well as decontamination of building structures accompanied by the clearance of buildings and open areas for subsequent conventional demolition. The waste arising will be packaged in accordance with its classification, and will be removed into interim storage or managed in the conventional way. The project is to be completed in the 3rd quarter of 2006. (orig.) [German] Das Versuchsatomkraftwerk Kahl (VAK) mit Siedewasserreaktor, das erste deutsche Kernkraftwerk, wurde nach 25 Betriebsjahren (1961 bis 1985) stillgelegt. Die Siedewasserreaktoranlage

  10. Causes, consequences, and therapy of the Radiophobia syndrome; Ursachen, Folgen und Therapie des Radiophobie-Syndroms

    Energy Technology Data Exchange (ETDEWEB)

    Becker, K.

    2004-03-01

    The final storage of high-level radioactive waste, which is said to be still open while, in fact, it was solved technically a long time ago and is only being blocked for political reasons, as well as alleged technical risks of German nuclear power plants which have never been demonstrated or proven, are listed again and again as grounds for opting out of the use of nuclear power. There is hardly any doubt that one of the main causes underlying also these arguments, and thus the main reason for the insufficient public acceptance of nuclear power in Germany at the present time as a safe, inexpensive, and non-polluting source of primary energy, is the widespread fear of radiation (radiophobia). Consequently, solutions proposed for successfully managing this radiophobia must be examined. Continued scientific studies of the subject do not seem to be promising, as funds are available at present only for continuing the search for negative biological effects. Important preconditions for a change in attitude are the appropriate initiatives to be taken by the relatively small number of sufficiently independent experts of proven scientific repute. Initiatives of this kind can now be observed in numerous countries and regions in the world. It must be pointed out in this connection, as is underlined again and again by experienced experts, that risk acceptance is not a matter of factual arguments, but of emotions. Psychological and pedagogic sensitivity certainly are important elements in changing public opinion in the interest of a more realistic assessment of the radiation risk and the acceptance of nuclear power. (orig.) [German] Die angeblich noch offene, tatsaechlich aber laengst technisch geloeste und nur politisch blockierte Frage der Endlagerung hochradioaktiver Abfaelle, ebenso wie vorgebliche, tatsaechlich aber nie nachgewiesene technische Risiken der deutschen Kernkraftwerke werden immer wieder als Ausstiegsgruende fuer die Kernenergie genannt. Es bestehen kaum

  11. TRAC Real Time: A high fidelity solution for NSSS modelling. Application to Lungmen and Grafenrheinfeld NPP simulators

    International Nuclear Information System (INIS)

    Nuclear Island (NSSS) modelling represents an essential part of a simulator software, as the accuracy and scope used is essential when representing appropriately the systems behaviour in operational transients, where the transitions of phase water-vapour are dominant such as: ATWS, LOCA, Feed and Bleed, Mid Loop Operation, etc. Tecnatom has been using, since the early 90's, its real-time simulation technology, the binomial TRAC-RT and NEMO, a 6-equations thermalhydraulic code and three-dimensional neutronic code for high fidelity modelling of the Primary System of several full scope simulators. Two latest projects which have been faced are the object of this paper. The first of them refers to Lungmen NPP Full scope simulator, an ABWR type being built by GE for Taiwan Power Company. The NSSS generated model is connected with the rest of BOP conventional simulation. The validation process has been carried out according to the methodology defined in ANSI 3.5 standard, taking like reference the engineering model that GE possesses for this Power Station. The second project describes the NSSS upgrading of D3 simulator, owned by KSG, having Grafenrheinfeld (KKG), a German PWRKWU NPP as reference unit. The development platform is Digital UNIX, connected by reflective memory (RMS) to the existing ENCORE simulator platform. Real-time requirements being fulfilled. In both projects, the model generated with TRAC-RT and NEMO represents, not only the primary circuit, but also the steam lines, given their complexity and importance. Once more, these two project show the trend of training simulators in incorporating more and more accurate models, using engineering grade models

  12. Development of Basic Key Technologies for Gen IV SFR Safety Evaluation

    International Nuclear Information System (INIS)

    Safety issues and design requirements on control rod worth were identified through the evaluation of safety design characteristics and the preliminary safety evaluation. This results will be taken into account for the conceptual design studies of the demonstration reactor in the next stage. The Level-1 Pasa has been performed and a quantitative Cdf value was produced for the selected design from the several candidates. The inherent safety characteristics of the selected design were evaluated through the DBE and ATWS analyses. A surrogate material for Tru has been selected which is applicable to the study of liquidus/solidus temperature test for the metallic fuel containing Tru. A methodology for the regression analysis with surrogate material has been developed and valuable data on metal fuel liquidus/solidus temperature have been measured. A simple mechanistic model describing a bending of subassemblies has been formulated based on the foreign test data and existing models. Its applicability has been evaluated for the Phenix design. New criteria of the core damage for the SFR PSA were identified. The list of initiating events, system response event tree, and core response event tree, which constitute a PSA methodology for an SFR, have been introduced. By developing the SFR PIRT, phenomenological model features, which have to be satisfied in a safety code, were defined and the PIRT results were applied to the design of the PDRC test facility. Bases for a safety evaluation methodology for the SFR DBEs have been also prepared. A draft version of the topical report on the code for local fault analysis has been completed. Since 2007, the MARS-LMR code has been developed and assessments for model validation with the test data from EBR-II and Phenix reactor have been continued. The code has been applied to the evaluation of passive safety of a conceptual design of Gen IV SFR

  13. Textured YBCO films grown on wires: application to superconducting cables

    International Nuclear Information System (INIS)

    Efforts to fabricate superconducting wires made of YBa2Cu3O7 (YBCO) on La2Zr2O7 (LZO) buffered and biaxially textured Ni-5 at.%W (NiW) are described. Wires were manually shaped from LZO buffered NiW tapes. Different diameters were produced: 1.5, 2 and 3 mm. The wires were further covered with YBCO grown by metal organic chemical vapor deposition (MOCVD). We developed an original device in which the round substrate undergoes an alternated rotation of 180° around its axis in addition to a reel-to-reel translation. This new approach allows covering the whole circumference of the wire with a YBCO layer. This was confirmed by energy dispersive x-ray spectroscopy (EDX) analysis coupled to a scanning electron microscope (SEM). For all wire diameters, the YBCO layer thickness varied from 300 to 450 nm, and the cationic composition was respected. Electron backscattering diffraction (EBSD) measurements were performed directly on an as-deposited wire without surface preparation allowing the investigation of the crystalline quality of the film surface. Combining EBSD with XRD results we show that YBCO grows epitaxially on the LZO buffered NiW wires. For the first time, superconductive behaviors have been detected on round substrates in both the rolling and circular direction. Jc reached 0.3 MA cm−2 as measured at 77 K by transport and third-harmonic detection. Those preliminary results confirm the effectiveness of the MOCVD for complex geometries, especially for YBCO deposition on small diameter wires. This approach opens huge perspectives for the elaboration of a new generation of YBCO-based round conductors. (paper)

  14. DCS Cabinet Power Loss Analysis for CPR 1000 Nuclear Power Plant%CPR1000核电厂DCS机柜失电分析

    Institute of Scientific and Technical Information of China (English)

    周亮; 赵岩峰; 孙永滨

    2014-01-01

    The DCS overall structure of CRP1000 nuclear power plant was introduced . Based on the RPC ,the signal interface character and signal processing mechanism on the key root were analyzed .By the power loss analyzing of RPC ,the RPC loss power may lead reactor trip signal from anticipated transient without scram (ATWS) system .The results indicate that it is necessary to search DCS cabinet power loss analysis .Optimi‐zing and assigning the main waterflow signals can avoid trigger reactor trip signal by mistake .The DCS cabinet power loss analysis can optimize the I&C (instrumentation and control) design and increase the nuclear plant’s reliability .%介绍了CPR1000核电厂数字化控制系统(DCS)的总体结构,以反应堆保护机柜(RPC)为基础,分析RPC的信号接口特性和信号关键路径节点的信号处理机制。结合RPC Ⅳ保护通道失电造成未能停堆的预期瞬变(ATWS)系统误发停堆信号的原因进行分析及优化,结果表明:对DCS机柜失电分析的研究是必要的,通过对RPC Ⅳ的给水流量信息进行优化和合理分配,可避免误发停堆信号。失电分析可优化仪控的设计,提高核电厂的可靠性。

  15. Modeling and analysis of a 2,4-MW CW magnicon

    International Nuclear Information System (INIS)

    This paper compares the results of small-signal theory and three-dimensional computer modeling of the magnicon, a new type of deflection-modulated microwave amplifier that has great potential for high-power, high-efficiency microwave generation. The selection of operating parameters and the theory of operation of the magnicon are also presented. The magnicon uses two circular cavity assemblies which support rotating RF fields. The input cavity assembly deflection-modulates an electron beam into an expanding spiral path, and the output cavity extracts the kinetic energy from the modulated beam. Static magnetic fields in the input cavity assembly confine the beam and establish the loaded Q of the input cavities. Static magnetic fields in the output cavity produce cyclotron motion at frequencies that are multiples of the microwave frequency. The interaction between the cyclotron motion and the rotating RF fields allows for a distributed, rather than concentrated, extraction of the energy in the electron beam. To date, most experimental work on the magnicon has been performed at the USSR Academy of Sciences, where a 915-MHz magnicon has developed a power output of 2.6 MW, a gain of 30 dB, and an efficiency of 73%. Initial modeling of the magnicon has demonstrated the basic physics of the device and indicates that even higher efficiencies may be achievable. As the accelerator community considers RF intensive projects like accelerator transmutation of nuclear waste (ATW) and accelerator production of tritium (APT), which require hundreds of megawatts of continuous-wave RF energy, the high efficiency and high average power of the magnicon make it an attractive candidate for these applications. A modeling effort is currently under way at Los Alamos to predict the efficiency of a 2.4-MW CW magnicon at 700 MHz. The effort includes a small-signal analysis of the input structure and a three-dimensional computer simulation of the entire device

  16. Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results

    International Nuclear Information System (INIS)

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the Internal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

  17. Morality and ethics in high technology; Moral und Ethik in der Spitzentechnologie

    Energy Technology Data Exchange (ETDEWEB)

    Schroeter, K.U.

    2003-06-01

    The ethical debate about what is feasible culminates, for one side, in the indignant moral question whether man is allowed to do all he is able to do and, for the other side, in the very obligation to keep redefining the limits of creation, and to act accordingly. Consequently, the Young Generation, at their meeting in Gronau, Westphalia (about which we reported), discussed about ''High Technology - Responsible on Ethical and Moral Grounds?'' The paper presented to the participants by pastor Kai Uwe Schroeter reflects this dichotomy, but also takes a clear position in favor of the expansion of nuclear power. This issue of atw contains a revised version of the paper. It is published in the hope that it will furnish arguments for the philosophical and ethical debates about high technology. (orig.) [German] Die ethische Diskussion ueber das Machbare gipfelt fuer die einen in der moralischen Entruestung: ''Darf der Mensch alles, was er kann?'', fuer die anderen geradezu in der Verpflichtung, die Grenzen der Schoepfung immer neu zu definieren - und entsprechend zu handeln. ''Spitzentechnologie - ethisch und moralisch verantwortbar'' lautete demgemaess das Thema der Jungen Generation bei ihrer vergangenen Tagung im westfaelischen Gronau (die afw berichtete). Der von Pfarrer Kai Uwe Schroeter vor den Teilnehmern gehaltene Vortrag spiegelt dieses Spannungsfeld wider, bezieht aber auch eindeutig Position fuer den Ausbau der Kernenergie. Wir veroeffentlichen an dieser Stelle den Vortrag in ueberarbeiteter Fassung und hoffen damit zur philosophisch-ethischen Diskussion der Spitzentechnologie Argumentationshilfen zu liefern. (orig.)

  18. Waste management and final storage in Germany - failed for lack of content and a technical basis? Pt. 2

    International Nuclear Information System (INIS)

    The assertion by the political parties at present in government in Germany, SPD and Alliance 90/The Greens, that ''the previous waste management concept for radioactive waste had failed in terms of contents and no longer had any technical basis'', is a purely ideological statement utterly devoid of any realistic reason. In actual fact, the waste management concept so far pursued in Germany has been transferred into industrial practice in many areas: transports of radioactive waste and spent fuel elements can be carried out safely at any time; spent fuel has been reprocessed on an industrial scale for many years. The central interim stores of Ahaus, Gorleben, and Lubmin, all of which are in operation, actually represent sufficient capacity for the interim storage of spent fuel elements. The successful exploration of the Gorleben salt dome has advanced far. No result so far would detract from its suitability. Consequently, the federal government should not try ''to elaborate a (new) national waste management plan for the inherited burden of radioactive waste,'' but rather invest all its power to make functional as quickly as possible the missing building blocks in the existing waste management concept. In doing so, it would make an important contribution to domestic peace and to the international recognition of Germany as a high-tech country. Part 1 of the article, which was published in atw 7 (2000) pp. 453-456, covers repro cessing and direct final storage of spent fuel elements with interim storage in special casks while part 2 in this issue contains a survey of the final storage options and the final storage projects in Germany (orig.)

  19. Development of Safety Analysis Technology for LMR

    International Nuclear Information System (INIS)

    In the safety analysis code system development area, the development of an analysis code for a flow blockage could be brought to completion throughout an integrated validation of MATRA-LMR-FB. The safety analysis code of SSC-K has been evolved by building detailed reactivity models and a core 3 dimensional T/H model into it, and developing its window version. A basic analysis module for SFR features also have been developed incorporating a numerical method, best estimated correlations, and a code structure module. For the analysis of the HCDA initiating phase, a sodium boiling model to be linked to SSC-K and a fuel transient performance/cladding failure model have been developed with a state-of-the-art study on the molten fuel movement models. Besides, scoping analysis models for the post-accident heat removal phase have been developed as well. In safety analysis area, the safety criteria for the KALIMER-600 have been set up, and an internal flow channel blockage and local faults have been analyzed for the assembly safety evaluation, while key safety concepts of the KALIMER-600 has been investigated getting through the analyses of ATWS as well as design basis accidents like TOP and LOF, from which the inherent safety due to a core reactivity feedback has been assessed. The HCDA analysis for the initiating phase and an estimation of the core energy release, subsequently, have been followed with setup of the safety criteria as well as T/H analysis for the core catcher. The thermal-hydraulic behaviors, and released radioactivity sources and dose rates in the containment have been analyzed for its performance evaluation in this area. The display of a data base for research products on the KALIMER Website and the detailed process planning with its status analysis, have become feasible from achievements in the area of the integrated technology development and establishment

  20. Fuel performance under transients, and accident management using Geno-Fuzzy concept for nuclear reactors

    International Nuclear Information System (INIS)

    Simulation of Pressurized Water Reactor Power Plant (PWR) has been investigated by simulating all components installed in the power plant namely: the reactor core, steam generator, pressurizer, reactor coolant pumps, and turbine. All plant components have been introduced. This simulator is useful for transient analysis studies, engineering designs, safety analysis, and accident management. Accidents in Pressurized Water Reactor Nuclear Power Plant (PWR NPP) may be occurred either due to component failures or human error during maintenance or operation. The main target of accident management is to mitigate accidents if it occurs. The Geno-Fuzzy concept is the way to select some important plant state variables as a gene for the overall plant state chromosome. The selected genes are: reactor power, primary coolant pressure, steam generator water level, and onset boiling on clad surface which has direct impact on fuel behavior. Each of these genes has associated fuzzy level. The main objective of Geno-Fuzzy is turning the plant gene from abnormal states to the normal state by associated control variable using the inference wise fuzzy technique. The Pressurized Water Reactor Nuclear Power Plant simulator has been tested for a typical PWR, for normal transients, Anticipated Transient Without Scram (ATWS), and using the proposed Geno-Fuzzy concept for accident management, which gives very good results in reactor accident mitigation. Some of these tested accidents are; reactor control rod ejection, change in turbine steam load, and loss of coolant flow, which have direct effects on fuel safety and performance. The parameters affecting the behavior of the reactor fuel integrity are analyzed to be considered in future reactor designs. (author)

  1. Simulator experiments: effects of experience of senior reactor operators and of presence of a shift technical advisor on performance in a boiling water reactor control room

    International Nuclear Information System (INIS)

    This report describes the first experiment in a Nuclear Regulatory Commission-sponsored program of training simulator experiments and field data collection to evaluate the effects of selected performance shaping factors on the performance of nuclear power plant control room operators. The factors investigated were the experience level of the Senior Reactor Operator (SRO) and the presence of a Shift Technical Advisor (STA). Data were collected from 16 two-man crews of licensed operators (one SRO and one RO). The crews were split into high and low SRO-experience groups on the basis of the years of experience of the SROs as SROs. One half (4 of the 8 crews in each group) of the high- and low-SRO experience groups were assisted by an STA or an SRO acting as an STA. The crews responded to four simulated plant casualties which ranged in severity from an uncomplicated turbine trip to an anticipated transient without scram (ATWS). No significant differences in overall performance were found between groups led by high (25 to 114 months licensed as an SRO) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have shorter task performance times than crews led by high experience SROs. Although a tendency for the STA-assisted groups to score higher on four of the five measures was observed, the presence of the STA had no statistically significant effect on overall team performance. The correlation between individual performance, as measured by four of the task performance measures, and experience, measured by months as a licensed operator, was not statistically significant, nor was the correlation between task performance and recency of simulator training. 18 references, 5 figures, 13 tables

  2. Ramona-4B development for SBWR safety studies

    International Nuclear Information System (INIS)

    The Simplified Boiling Water Reactor (SBWR) is a revolutionary design of a boiling-water reactor. The reactor is based on passive safety systems such as natural circulation, gravity flow, pressurized gas, and condensation. SBWR has no active systems, and the flow in the vessel is by natural circulation. There is a large chimney section above the core to provide a buoyancy head for natural circulation. The reactor can be shut down by either of four systems; namely, scram, Fine Motion Control Rod Drive (FMCRD), Alternate Rod Insertion (ADI), and Standby Liquid Control System (SLCS). The safety injection is by gravity drain from the Gravity Driven Cooling System (GDCS) and Suppression Pool (SP). The heat sink is through two types of heat exchangers submerged in the tank of water. These heat exchangers are the Isolation Condenser (IC) and the Passive Containment Cooling System (PCCS). The unique design of SBWR imposes new requirements on the analytic methods for modeling its behavior. The close coupling between the power and flow, and also flow distribution among the parallel channels require a multidimensional power-prediction capability. The startup of the reactor has vapor generation and condensation taking place in the core requiring a model with a non-homogeneous, nonequilibrium, two-phase formulation. The instability at low flow/high power conditions requires modeling of the control systems and balance of plant, which has significant impact on the amplitude of the instability-induced power and flow oscillations. The RAMONA-4B code has been developed to simulate the normal operation, reactivity transients, and to address the instability issues for SBWR. The objective of this project is develop a Sun SPARC and IBM RISC 6000 based RAMONA-4B code for applications to SBWR safety analyses, in particular for stability and ATWS studies

  3. Comparative Performance with Different Versions of Low Heat Rejection Combustion Chambers with Crude Rice Bran Oil

    Directory of Open Access Journals (Sweden)

    Krishna M.V.S. Murali

    2014-12-01

    Full Text Available Jak wiadomo, oleje roslinne sa obiecujacym substytutem paliw ropopochodnych, poniewaz ich własciwosci sa podobne do oleju dieslowskiego, sa odnawialne i łatwe do wyprodukowania. Niemniej, surowe oleje roslinne wykazuja wady, takie jak wysoka lepkosc i mała lotnosc, co wymaga komory spalania o małych stratach ciepła, której istotnymi cechami sa m.in. wyzsza temperatura robocza, maksymalne wydzielanie ciepła i zdolnosc do wykorzystania paliwa o mniejszej wartosci kalorycznej (CV. Przeprowadzono eksperymenty majace na celu ocene osiagów silnika z róznymi komorami spalania o małych stratach ciepła (LHR, takich jak głowica cylindra o pokryciu ceramicznym (LHR-1, tłok izolowany szczelina powietrzna z denkiem ze stopu Superni (superstop niklu i tuleja cylindra z wkładka z Superni izolowana szczelina powietrzna (LHR-2 oraz głowica cylindra z pokryciem ceramicznym, tłok i tuleja cylindra izolowane szczelinami powietrznymi (LHR-3. Badania prowadzono przy normalnej temperaturze oleju roslinnego (surowy olej z otrab ryzowych, CRBO i zmiennym cisnieniu w otworze wtryskiwacza. Parametry osiagów silnika (uzyteczna sprawnosc termiczna, uzyteczny współczynnik zuzycia energii, temperatura gazu wydychanego, obciazenie obiegiem chłodziwa i współczynnik napełnienia oraz emisje wydechowe [poziomy dymu i tlenków azotu, NOx] zostały wyznaczone przy róznych wartosciach sredniego uzytecznego cisnienia w silniku. Charakterystyki spalania [cisnienie szczytowe, czas wystepowania cisnienia szczytowego, maksymalna szybkosc wzrostu cisnienia] zostały wyznaczone w warunkach pracy silnika z pełnym obciazeniem.

  4. Aspiration tests in aqueous foam using a breathing simulator

    Energy Technology Data Exchange (ETDEWEB)

    Archuleta, M.M.

    1995-12-01

    Non-toxic aqueous foams are being developed by Sandia National Laboratories (SNL) for the National Institute of Justice (NIJ) for use in crowd control, cell extractions, and group disturbances in the criminal justice prison systems. The potential for aspiration of aqueous foam during its use and the resulting adverse effects associated with complete immersion in aqueous foam is of major concern to the NIJ when examining the effectiveness and safety of using this technology as a Less-Than-Lethal weapon. This preliminary study was designed to evaluate the maximum quantity of foam that might be aspirated by an individual following total immersion in an SNL-developed aqueous foam. A.T.W. Reed Breathing simulator equipped with a 622 Silverman cam was used to simulate the aspiration of an ammonium laureth sulfate aqueous foam developed by SNL and generated at expansion ratios in the range of 500:1 to 1000:1. Although the natural instinct of an individual immersed in foam is to cover their nose and mouth with a hand or cloth, thus breaking the bubbles and decreasing the potential for aspiration, this study was performed to examine a worst case scenario where mouth breathing only was examined, and no attempt was made to block foam entry into the breathing port. Two breathing rates were examined: one that simulated a sedentary individual with a mean breathing rate of 6.27 breaths/minute, and one that simulated an agitated or heavily breathing individual with a mean breathing rate of 23.7 breaths/minute. The results of this study indicate that, if breathing in aqueous foam without movement, an air pocket forms around the nose and mouth within one minute of immersion.

  5. Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results; Analisis de eventos internos para la Unidad 1 de la Central Nucleoelectrica de Laguna Verde. Cuantificacion de secuencias de accidente y resultados

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1994-07-01

    The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the {sup I}nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

  6. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station; Analisis de eventos internos para la Unidad 1 de la Central Nucleolelectrica de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 03000 Mexico D.F. (Mexico)

    1993-07-01

    This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  7. Simplified containment event tree analysis for the Sequoyah Ice Condenser containment

    International Nuclear Information System (INIS)

    An evaluation of a Pressurized Water Reactor (PER) ice condenser containment was performed. In this evaluation, simplified containment event trees (SCETs) were developed that utilized the vast storehouse of information generated by the NRC's Draft NUREG-1150 effort. Specifically, the computer programs and data files produced by the NUREG-1150 analysis of Sequoyah were used to electronically generate SCETs, as opposed to the NUREG-1150 accident progression event trees (APETs). This simplification was performed to allow graphic depiction of the SCETs in typical event tree format, which facilitates their understanding and use. SCETs were developed for five of the seven plant damage state groups (PDSGs) identified by the NUREG-1150 analyses, which includes: both short- and long-term station blackout sequences (SBOs), transients, loss-of-coolant accidents (LOCAs), and anticipated transient without scram (ATWS). Steam generator tube rupture (SGTR) and event-V PDSGs were not analyzed because of their containment bypass nature. After being benchmarked with the APETs, in terms of containment failure mode and risk, the SCETs were used to evaluate a number of potential containment modifications. The modifications were examined for their potential to mitigate or prevent containment failure from hydrogen burns or direct impingement on the containment by the core, (both factors identified as significant contributors to risk in the NUREG-1150 Sequoyah analysis). However, because of the relatively low baseline risk postulated for Sequoyah (i.e., 12 person-rems per reactor year), none of the potential modifications appear to be cost effective. 15 refs., 10 figs. , 17 tabs

  8. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    International Nuclear Information System (INIS)

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule (section 50.62); (2) station blackout (section 50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term (section 50.34(f) and section 100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements (section 50.55a, etc); (6) ECCS acceptance criteria (section 50.46)(b); (7) combustible gas control (section 50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle (section 51.51); and (11) (standards section 50.55a)

  9. Analysis of internal events for the Unit 1 of the Laguna Verde nuclear power station

    International Nuclear Information System (INIS)

    This volume presents the results of the starter event analysis and the event tree analysis for the Unit 1 of the Laguna Verde nuclear power station. The starter event analysis includes the identification of all those internal events which cause a disturbance to the normal operation of the power station and require mitigation. Those called external events stay beyond the reach of this study. For the analysis of the Laguna Verde power station eight transient categories were identified, three categories of loss of coolant accidents (LOCA) inside the container, a LOCA out of the primary container, as well as the vessel break. The event trees analysis involves the development of the possible accident sequences for each category of starter events. Events trees by systems for the different types of LOCA and for all the transients were constructed. It was constructed the event tree for the total loss of alternating current, which represents an extension of the event tree for the loss of external power transient. Also the event tree by systems for the anticipated transients without scram was developed (ATWS). The events trees for the accident sequences includes the sequences evaluation with vulnerable nucleus, that is to say those sequences in which it is had an adequate cooling of nucleus but the remoting systems of residual heat had failed. In order to model adequately the previous, headings were added to the event tree for developing the sequences until the point where be solved the nucleus state. This process includes: the determination of the failure pressure of the primary container, the evaluation of the environment generated in the reactor building as result of the container failure or cracked of itself, the determination of the localization of the components in the reactor building and the construction of boolean expressions to estimate the failure of the subordinated components to an severe environment. (Author)

  10. Reel-to-reel continuous simultaneous double-sided deposition of highly textured CeO2 templates for YBa2Cu3O7-δ coated conductors

    International Nuclear Information System (INIS)

    A reel-to-reel system which allows simultaneous two-sided deposition of epitaxial CeO2 buffer layers on long length biaxially textured Ni-5 at.%W tape with direct current (dc) reactive magnetron sputtering is described. Deposition is accomplished through two opposite symmetrical sputtering guns with a radiation heater. Meter-long double-sided epitaxial CeO2 buffer layers have been produced for the first time on textured metal substrates in a run using a reel-to-reel process with a speed of about 1.2 m h-1. The CeO2 films were characterized by means of x-ray diffraction (XRD) and atomic force microscopy (AFM). The samples exhibited good epitaxial growth with the c-axis perpendicular to the substrate surface for both sides. Full width at half maximum (FWHM) values of the out-of-plane and in-plane orientation for both sides were 3.20 and 3.10, 5.30 and 5.10, respectively. AFM observations revealed a smooth, dense and crack-free surface morphology. In addition, x-ray scans have been performed as a function of length to determine the crystallographic consistency of the epitaxial CeO2 over the length. Subsequently anyttria-stabilized zirconia (YSZ) barrier and CeO2 cap layers were deposited to complete the CeO2/YSZ/CeO2 structure via the same process. Epitaxial YBa2Cu3O7-δ (YBCO) films grown by dc sputtering on the short prototype CeO2/YSZ/CeO2/NiW conductors yielded self-field critical current densities (Jc) as high as 1.3 MA cm-2 at 77 K. An Ic value of 113 A cm-1 was obtained for double-sided YBCO coated conductors

  11. Fabrication of the cube textured NiO buffer layer by line-focused infrared heating for coated conductor application

    International Nuclear Information System (INIS)

    Epitaxial growth of NiO on the bi-axially textured Ni-3 at.%W (Ni-3W) substrate as seed layer for coated conductor were studied. The bi-axially textured NiO was formed on the Ni-3W tapes using a line-focused infrared heater by oxidizing the surface of the substrate at 800-950 deg. C for 15-120 s in oxygen atmosphere. The thickness of the NiO layer could be controlled by changing heat-treatment, which was estimated as approximately 200-500 nm in the cross-sectional SEM micrographs of the NiO/Ni template. This thickness is enough to block the diffusion of the Ni in the substrate to the superconducting layer. The samples showed strong texture development of NiO layer. The sample oxidized at 900 deg. C with the tape transferring speed of 30 mm/h exhibited ω-scan full width at half maximum (FWHM) values for Ni-3W(2 0 0) and NiO(2 0 0) were 3.97 deg., and 3.67 deg., and φ-scan FWHM values for Ni-3W(1 1 1) and NiO(1 1 1) were 9.58 deg., and 8.79 deg., respectively. Also, the (1 1 1) pole-figure of the NiO buffer layer showed the good symmetry of the four peaks, securing the epitaxial growth of the buffer layers on the NiO layer. Also NiO layer exhibited root-mean-square roughness value of 39 nm by AFM (10 x 10 μm) investigation

  12. YKAe Research programme on nuclear power plant systems behaviour and operational aspects of safety 1990-1994, Final report

    International Nuclear Information System (INIS)

    The research programme on Nuclear Power Plant Systems Behaviour and Operational Aspects of Safety was carried out between 1990 and 1994. In the field of Safe operational margins of nuclear fuel and reactor core, an up-to-date steady-state fuel performance model was validated for higher burn-ups and well-characterized VVER fuel experiments were carried out. A comprehensive reactor analysis code system was extended and validated for complex 3-D phenomena, such as ATWS and boron dilution transients. Advanced hydraulics methods were added to the dynamics codes. Experiments were carried out with PACTEL, the most comprehensive thermal-hydraulic test facility for VVER-440-type reactors worldwide. For example, a series of natural circulation tests were provided for the first VVER-related international standard problem of the OECD/NEA. Advanced foreign computer codes for severe accidents were evaluated and modified for the needs of Finnish power plants. Specific progress was made in modelling the reflooding of an overheated core and in the structural analysis of a pressure vessel under corium load, as well as in experimental and theoretical studies of aerosol and hydrogen behaviour. Fire modelling was improved by implementing advanced calculation methods and by validating them against our own experiments and international test data. Techniques in living probabilistic safety assessment and risk decision-making were refined in pilot applications for continuous monitoring, follow-up and management of risks of an operating power plant. In the area of human reliability and organizational performance, factors important for the continuous development of the competence of control room operator teams and plant maintenance staff were identified. (237 refs., 75 figs., 13 tabs.)

  13. Numerical simulation of boron injection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tinoco, Hernan, E-mail: htb@forsmark.vattenfall.s [Forsmarks Kraftgrupp AB, SE-742 03 Osthammar (Sweden); Buchwald, Przemyslaw [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Frid, Wiktor, E-mail: wiktor@reactor.sci.kth.s [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

    2010-02-15

    The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of

  14. An Advanced Option for Sodium Cooled TRU Burner Loaded with Uranium-Free Fuels

    Energy Technology Data Exchange (ETDEWEB)

    You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    The sodium cooled fast reactors of this kind that are called burners are designed to have low conversion ratio by reducing fuel volume fraction or reducing neutron leakage or increasing neutron absorption. However, the typical SFR burners have a limited ability of TRU burning rate due to the fact that they use metallic or oxide fuels containing fertile nuclides such as {sup 238}U and {sup 232}Th and these fertile nuclides generate fissile nuclides through neutron capture even if they are designed to have low conversion ratio (e.g., 0.6). To further enhance the TRU burning rate, the removal of the fertile nuclides from the initial fuels is required and it will accelerate the reduction of TRUs that are accumulated in storages of LWR spent fuels. However, it has been well-known 4 that the removals of the fertile nuclides from the fuel degrade the inherent safety of the SFR burner cores through the significant decrease of the fuel Doppler effect, the increase of sodium void reactivity worth, and reduction of delayed neutron fraction. In this work, new option for the sodium cooled fast TRU burner cores loaded with fertile-free metallic fuels was proposed and the new cores were designed by using the suggested option. The cores were designed to enhance the inherent safety characteristics by using axially central absorber region and 6 or 12 ZrH1.8 moderator rods per fuel assembly. For each option, we considered two different types of fertile-free ternary metallic fuel (i.e., TRU-W-10Zr and TRU-Ni-10Zr). Also, we performed the BOR (Balance of Reactivity) analyses to show the self-controllability under ATWS as a measure of inherent safety. The core performance analysis showed that the new cores using axially central absorber region substantially improve the core performance parameters such as burnup reactivity swing and sodium void reactivity worth.

  15. TASS code topical report. V.1 TASS code technical manual

    International Nuclear Information System (INIS)

    TASS 1.0 code has been developed at KAERI for the initial and reload non-LOCA safety analysis for the operating PWRs as well as the PWRs under construction in Korea. TASS code will replace various vendor's non-LOCA safety analysis codes currently used for the Westinghouse and ABB-CE type PWRs in Korea. This can be achieved through TASS code input modifications specific to each reactor type. The TASS code can be run interactively through the keyboard operation. A simimodular configuration used in developing the TASS code enables the user easily implement new models. TASS code has been programmed using FORTRAN77 which makes it easy to install and port for different computer environments. The TASS code can be utilized for the steady state simulation as well as the non-LOCA transient simulations such as power excursions, reactor coolant pump trips, load rejections, loss of feedwater, steam line breaks, steam generator tube ruptures, rod withdrawal and drop, and anticipated transients without scram (ATWS). The malfunctions of the control systems, components, operator actions and the transients caused by the malfunctions can be easily simulated using the TASS code. This technical report describes the TASS 1.0 code models including reactor thermal hydraulic, reactor core and control models. This TASS code models including reactor thermal hydraulic, reactor core and control models. This TASS code technical manual has been prepared as a part of the TASS code manual which includes TASS code user's manual and TASS code validation report, and will be submitted to the regulatory body as a TASS code topical report for a licensing non-LOCA safety analysis for the Westinghouse and ABB-CE type PWRs operating and under construction in Korea. (author). 42 refs., 29 tabs., 32 figs

  16. Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 2: Accident and Thermal Fluids Analysis PIRTs

    Energy Technology Data Exchange (ETDEWEB)

    Ball, Sydney J [ORNL; Corradini, M. [University of Wisconsin; Fisher, Stephen Eugene [ORNL; Gauntt, R. [Sandia National Laboratories (SNL); Geffraye, G. [CEA, France; Gehin, Jess C [ORNL; Hassan, Y. [Texas A& M University; Moses, David Lewis [ORNL; Renier, John-Paul [ORNL; Schultz, R. [Idaho National Laboratory (INL); Wei, T. [Argonne National Laboratory (ANL)

    2008-03-01

    An accident, thermal fluids, and reactor physics phenomena identification and ranking process was conducted by a panel of experts on the next generation nuclear plant (NGNP) design (consideration given to both pebble-bed and prismatic gas-cooled reactor configurations). Safety-relevant phenomena, importance, and knowledge base were assessed for the following event classes: (1) normal operation (including some reactor physics aspects), (2) general loss of forced circulation (G-LOFC), (3) pressurized loss-of-forced circulation (P-LOFC), (4) depressurized loss-of-forced circulation (D-LOFC), (5) air ingress (following D-LOFC), (6) reactivity transients - including anticipated transients without scram (ATWS), (7) processes coupled via intermediate heat exchanger (IHX) (IHX failure with molten salt), and (8) steam/water ingress. The panel's judgment of the importance ranking of a given phenomenon (or process) was based on the effect it had on one or more figures of merit or evaluation criteria. These included public and worker dose, fuel failure, and primary (and other safety) system integrity. The major phenomena of concern that were identified and categorized as high importance combined with medium to low knowledge follow: (1) core coolant bypass flows (normal operation), (2) power/flux profiles (normal operation), (3) outlet plenum flows (normal operation), (4) reactivity-temperature feedback coefficients for high-plutonium-content cores (normal operation and accidents), (5) fission product release related to the transport of silver (normal operation), (6)emissivity aspects for the vessel and reactor cavity cooling system (G-LOFC), (7) reactor vessel cavity air circulation and heat transfer (G-LOFC), and (8)convection/radiation heating of upper vessel area (P-LOFC).

  17. The significance of biometric parameters in determining anterior teeth width

    Directory of Open Access Journals (Sweden)

    Strajnić Ljiljana

    2013-01-01

    Full Text Available Background/Aim. An important element of prosthetic treatment of edentulous patients is selecting the size of anterior artificial teeth that will restore the natural harmony of one’s dentolabial structure as well as the whole face. The main objective of this study was to determine the correlation between the inner canthal distance (ICD and interalar width (IAW on one side and the width of both central incisors (CIW, the width of central and lateral incisors (CLIW, the width of anterior teeth (ATW, the width between the canine cusps (CCW, which may be useful in clinical practice. Methods. A total of 89 subjects comprising 23 male and 66 female were studied. Their age ranged from 19 to 34 years with the mean of 25 years. Only the subjects with the preserved natural dentition were included in the sample. All facial and intraoral tooth measurements were made with a Boley Gauge (Buffalo Dental Manufacturing Co., Brooklyn NY, USA having a resolution of 0.1mm. Results. A moderate correlation was established between the interalar width and combined width of anterior teeth and canine cusp width (r = 0.439, r = 0.374. A low correlation was established between the inner canthal distance and the width of anterior teeth and canine cusp width (r = 0.335, r = 0.303. The differences between the two genders were highly significant for all the parameters (p < 0.01. The measured facial distances and width of anterior teeth were higher in men than in women. Conclusion. The results of this study suggest that the examined interalar width and inner canthal distance cannot be considered reliable guidelines in the selection of artificial upper anterior teeth. However, they may be used as a useful additional factor combined with other methods for objective tooth selection. The final decision should be made while working on dentures fitting models with the patient’s consent.

  18. ESBWR - Robust design for natural circulation and stability performance effectiveness

    International Nuclear Information System (INIS)

    ESBWR is a 4500 MWt Generation III+ natural circulation reactor with an array of robust design features and passive safety systems to deliver highly effective plant performance during normal operation and to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US design certification process. This paper focuses on the natural circulation-driven plant performance aspects during normal operation, and stability evaluation of the robust ESBWR design. The TRACG computer code is used for the analysis of ESBWR plant performance, safety analysis, and stability margins. The paper describes the evaluation of ESBWR stability performance during normal power operation including operation in the Core Power-Feed Water Temperature Operating Domain. For ESBWR the normal power operation condition has the highest power/flow ratio and is limiting from the perspective of stability. The paper includes results from detailed evaluation of the most limiting decay ratio for out-of-phase regional oscillations calculated by perturbing the core inlet flow rate in this out-of-phase mode about the line of symmetry for the azimuthal harmonic mode. The paper also summarizes the ESBWR regional mode stability evaluations during a limiting transient (Loss of Feedwater Heating), and during ATWS (Anticipated Transient without Scram). Nominal decay ratios of limiting Channel oscillation, Core wide oscillation and Regional oscillation are within the maximum acceptance criterion of 0.8, at 95% content and 95% confidence. These stability evaluation results indicate decay ratio is within design limits. The paper also describes the evaluation of ESBWR stability performance during plant startup, and summarizes the defense-in-depth stability solution for ESBWR. (authors)

  19. Development of the Joyo MK-II core bowing reactivity calculation code

    Energy Technology Data Exchange (ETDEWEB)

    Tabuchi, Shiro; Torimaru, Tadahiko; Yoshida, Akihiro; Aoyama, Takafumi [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1999-09-01

    The study on the passive safety test by using the Experimental Fast Reactor Joyo has been performed to demonstrate the inherent safety of fast breeder reactors. In this study, emphasis was placed on the improvement on the accuracy of the feedback reactivity analysis. As a bowing reactivity might play a significant roll in ATWS analysis because of its effectively short time constant and relatively large magnitude, an emphasis was placed upon the evaluation of the analysis precision of bowing reactivity. Taking into account of the refueling and irradiation history of the individual core component, the core bowing behavior in Joyo has been analyzed by using the MK-II core management code system MAGI, the interface code TETRAS which interpolate neutron flux and coolant temperature at the position of wrapper tube, and the core bowing calculation code BEACON. Calculation accuracy of above mentioned system was evaluated through the comparison of calculated and measured permanent distortion of subassemblies. In 1996, core bowing reactivity was calculated by AURORA code using the above calculated bowing behavior of individual core component as input. But because an approximate two dimensional material reactivity worth map was utilized in AURORA, it was made clear that some amount of error caused by extrapolation could not be neglected. Therefore calculation code ARCHCOM (Analysis of Reactivity Change due to Core Mechanics) which utilize three dimensional material reactivity worth map as input was developed for the Joyo MK-II core bowing reactivity calculation. This code reduces above mentioned extrapolation error that used to be occurred at isolated core component, such as control rod or irradiation rig and at the interface region between fuel and reflector which had sharp bowing reactivity worth gradient. (author)

  20. Parallel beam dynamics calculations on high performance computers

    International Nuclear Information System (INIS)

    Faced with a backlog of nuclear waste and weapons plutonium, as well as an ever-increasing public concern about safety and environmental issues associated with conventional nuclear reactors, many countries are studying new, accelerator-driven technologies that hold the promise of providing safe and effective solutions to these problems. Proposed projects include accelerator transmutation of waste (ATW), accelerator-based conversion of plutonium (ABC), accelerator-driven energy production (ADEP), and accelerator production of tritium (APT). Also, next-generation spallation neutron sources based on similar technology will play a major role in materials science and biological science research. The design of accelerators for these projects will require a major advance in numerical modeling capability. For example, beam dynamics simulations with approximately 100 million particles will be needed to ensure that extremely stringent beam loss requirements (less than a nanoampere per meter) can be met. Compared with typical present-day modeling using 10,000 endash 100,000 particles, this represents an increase of 3 endash 4 orders of magnitude. High performance computing (HPC) platforms make it possible to perform such large scale simulations, which require 10 close-quote s of GBytes of memory. They also make it possible to perform smaller simulations in a matter of hours that would require months to run on a single processor workstation. This paper will describe how HPC platforms can be used to perform the numerically intensive beam dynamics simulations required for development of these new accelerator-driven technologies. copyright 1997 American Institute of Physics

  1. Turkey's way to nuclear energy. An example for a newcomer's new build

    International Nuclear Information System (INIS)

    The government of the Republic of Turkey acted very determined for several years to put the first nuclear power plant in Turkey to full operation by 2020. The economic growth of Turkey, which is far higher than the EU's average, requires a modern and reliable energy supply for the population and businesses. The Turkish government's energy policy and energy economics decisions for the realization of the necessary steps to achieve the energy targets are implemented quickly. In this process, the reduction of the dependence on energy imports plays a significant role. Hereunto, the build-up of nuclear power in Turkey is to be used. Lower ecological disadvantages than fossil forms of energy production and higher production reliability than thermal or hydroelectric power plants are attributed to nuclear power plants. The site for the first nuclear power plant in Mersin-Akkuyu has been determined and the site is being scientifically and systematically explored, so that the government-selected Russian partner can construct and operate the facility. The tender for the second planned nuclear power plant in Sinop-Inceburun is being prepared. The energy economics legislative and especially the nuclear and radiation protection regulation systems, including the so-called sub-legal nuclear regulations for commercial nuclear power plants, are being developed. In particular, the required safety standards for the construction and operation of nuclear power plants need to be further elaborated. For these procedures, the required personnel for the authorities and experts have to be intensively trained and prepared for their practical tasks. Considering this background and subsequent to the report in atw 2007, 15 et seq. above all the safety aspects of nuclear power plants in terms of their planning, site selection, construction and operation will be examined. Without a reliable legal framework and sound technical regulations rules the licensing process for construction and operation

  2. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  3. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    International Nuclear Information System (INIS)

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing facilities. - 3. Advances in Water

  4. Life extension of German nuclear power plants only with the consent of the Federal Council? The importance and extent of the need for consent to an amendment to the German Atomic Energy Act

    International Nuclear Information System (INIS)

    In its coalition agreement of October 26, 2009, the new German federal government plans ''to extend the service life of German nuclear power plants while, at the same time, complying with the strict German and international safety standards.'' This has triggered a debate not only about (nuclear) energy, as in the past election campaign in the summer of 2009, but also about the constitutional law issue whether an amendment to the Atomic Energy Act resulting in longer operating life of nuclear power plants required the consent of the Federal Council (the ''Bundesrat,'' the second chamber of parliament). After the election to the state parliament in North Rhine-Westphalia on May 9, 2010, majority in the Federal Council changed. As a consequence, no consent to an amendment to the Atomic Energy Act must be expected. In view of the large number of recent statements about constitutional law in opinions for various federal and ministerial accounts as well as firms and associations, the outline by R. Scholz in the May issue of atw 2010 will be followed in this issue by the key points of examination of the need for consent, under aspects of constitutional law, and an attempt will be made to explain the evaluations underlying the generation of a legal concept about these items. The decision by the German Federal Constitutional Court of May 4, 2010, published on June 11, 2010, plays a major role in this respect because it established clarity in some important aspects of a legal subject matter in the field of state admini-stration on behalf of the federation, albeit in the field of air traffic law, not nuclear law. However, the structures of the norms in the German Basic Law (Art. 87c and Art. 87d, para.2) to be applied are almost identical. The energy policy and energy economy aspects of a plant life extension are considered along with the option of an appeal to the Federal Constitutional Court against any plant life extension. Finally, the key findings are summarized briefly

  5. The basic research on the CDA initiation phase for a metallic fuel FBR

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Go; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan); Kawada, Ken-ichi; Niwa, Hazime

    1998-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku University and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled. (1) Target and Results of analysis: The accident initiator considered is a LOF accident with ATWS. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transient phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large material movement to in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics. The results of a sample case, that is a metallic fueled core at the beginning of cycle, show this improvement is appropriate. (3) Conclusion: The behavior at CDA of a metallic fueled core of a fast reactor was analyzed using the CDA initiation phase analysis code and the knowledge of the important characteristics at the CDA initiation phase was obtained

  6. CFD predictions of standby liquid control system mixing in lower plenum of a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, Christopher, E-mail: christopher.boyd@nrc.gov; Skarda, Raymond, E-mail: Raymond.skarda@nrc.gov

    2014-11-15

    Highlights: • Computational fluid dynamics analysis of BWR lower plenum. • Mixing and stratification of the standby liquid control system injection. • Scoping study highlights the expected flow paths and limitations of experiments. - Abstract: During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC). An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density. A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The

  7. Results from Accelerator Driven TRIGA Reactor Experiments at The University of Texas at Austin

    International Nuclear Information System (INIS)

    Accelerator Driven Transmutation of High-Level Waste (ATW) is one possible solution to the fuel reprocessing back-end problem for the disposal of high level waste such as minor actinides (Am, Np or Cm) and long-lived fission products. International programs continue to support research towards the eventual construction and operation of a proton accelerator driven spallation neutron source coupled to a subcritical 'neutron amplifier' for more efficient HLW transmutation. This project was performed under DOE AFCI Reactor-Accelerator Coupling Experiments (RACE). A 20 MeV Electron Linac was installed in the BP no 5 cave placing neutron source adjacent to an offset reactor core to maximize neutron coupling with available systems. Asymmetric neutron injection 'wasted' neutrons due to high leakage but sufficient neutrons were available to raise reactor power to ∼100 watts. The Linac provided approximately 100 mA but only 50% reached target. The Linac cooling system could not prevent overheating at frequencies over 200 Hz. The Linac electron beam had harmonics of primary frequency and periodic low frequency pulse intensity changes. Neutron detection using fission chambers in current mode eliminated saturation dead time and produced better sensitivity. The Operation of 'dual shielded' fission chambers reduced electron noise from linac. Benchmark criticality calculation using start-up data showed that the MCNPX model overestimates reactivity. TRIGA core was loaded to just slightly supercritical by adding graphite elements and measuring reactivity of $0.037. MCNPX modeled TRIGA core with and without graphite to arrive at 'true' measured subcritical multiplication of 0.998733± 0.00069. Thus, Alpha for the UT-RACE TRIGA core was approximately 155.99 s-1. The Stochastic Feynman-Alpha Method (SFM) accuracy was evaluated during transients and reactivity changes. SFM was shown to be a potential real-time method of reactivity determination in future ADSS but requires stable

  8. SIMULATE-3K linkage with reactor systems codes

    International Nuclear Information System (INIS)

    SIMULATE-3K is Studsvik Scandpower's best-estimate three-dimensional core kinetics code. SIMULATE-3K has been coupled to several best-estimate reactor systems codes including, RELAP5-3D, RELAP5-3.3, TRACE V5.0, and RETRAN-3D. The coupled codes can be applied to existing reactors and to advanced reactor designs. The S3K linkage to each of the systems codes is a direct, explicit coupling of the two codes on a synchronous time-step basis. The coupling provides an execution method for the S3K three-dimensional neutronic model using the Nuclear Steam Supply System (NSSS) boundary conditions calculated by the systems code. Also, it allows the S3K calculated total core power and core power distributions to drive the system model core. Detailed calculations from the component codes result in a methodology for analyzing limiting transients such as steam line breaks, rod drops/ejections, and ATWS scenarios. These transient events require detailed three- dimensional core data and information about the behavior of NSSS components. A coupled analysis of these transients is important because the core behavior is closely tied to the NSSS system. For example, to capture the timing and characteristics of the important thermal-hydraulic phenomena and/or operations events, such as valve closures, safety injection, or control system interactions, requires a detailed plant model. The Peach Bottom 2 turbine trip transient is used to assess the accuracy of the coupled code calculations. Comparisons of the important plant parameters to results from RELAP5-3D, RELAP5-3.3, and TRACE V5.0 calculations are shown and discussed. The MSLB benchmark is also used to demonstrate the capabilities of the coupled code systems. Comparisons of the calculated reactor power to the reference data are shown can discussed. The comparisons demonstrate the applicability of S3K, either standalone or coupled with a system analysis code, to properly model system response during accident scenarios. (author)

  9. Expert report of ENSI on the request of EKKB AG for a general license - Project 'New nuclear power plant to replace the Beznau plant'

    International Nuclear Information System (INIS)

    The 'Ersatz Kernkraftwerk Beznau AG' (EKKB) Company submitted to the Swiss Federal Inspectorate of Nuclear Safety (ENSI) a request for a general license for a new power plant to be built near to the Beznau power plants. According to the law, all damage risks with a probability higher than 10-4/a must be taken into account through protection measures. The considered risks concern the power plant itself as well as the population in the neighbourhood and the environment. The purpose of the general license is to demonstrate that the site chosen for the foreseen power plant is acceptable and that the risks can be counteracted through adequate measures. The buildings of the power plant and their partition on the Beznau Island in the Aare River are briefly described. The reactor is a Light Water Reactor of third generation with a maximum electrical power of 1450 MWel ±20%. The main cooling is provided by a hybrid system of water evaporation and air heating, what reduces the plume at the exit of the cooling tower. First, it is demonstrated that, in the case of a very unlikely severe accident in the power plant, the people in the neighbourhood can be evacuated quickly. Then, numerous types of possible accidents in the neighbourhood of the power plant are analyzed in order to settle their possible negative influence on the operation of the power plant: bursting of gas containers on the neighbouring roads and railways, fires of all types of hydrocarbons, air pollution through chloride gas, etc. The check by ENSI of the EKKB studies on the potential danger for the power plant through neighbouring industrial plants, roads or railways demonstrated that none of the considered accidents presents an unacceptable risk for the power plant: on the one hand, these plants are located too far from the power plant, so that a sensible injury to the power plant safety can be excluded; on the other, the protection of the power plant can be guaranteed through appropriate technical measures

  10. Level-2 PSA for the prototype fast breeder reactor MONJU applied to the accident management review

    International Nuclear Information System (INIS)

    An accident management guideline (AMG) of the prototype fast breeder reactor MONJU has been presented to Nuclear and Industry Safety Agency (NISA) of METI by Japan Atomic Energy Agency (JAEA) with an evaluation result of an effectiveness of the AMG by employing Level-1 and Level-2 PSAs. Japan Nuclear Energy Safety Organization (JNES) evaluated the three events - PLOHS, LORL and ATWS events - and scrutinized the results of the Level-2 PSA carried out by JAEA from the view point of an accident management (AM) review. Regarding ATWS events, we have carried out a qualitative evaluation of the results of JAEA's evaluation and carried out a quantitative evaluation of the containment failure frequency (CFF) in relation to Protected-Loss-of-Heat-Sink (PLOHS) and Loss-of-Reactor-Level (LORL) events. Evaluation of the containment failure probability CFF has been conducted based on the results of the Level-1 PSA by employing the code system developed by JNES. We conducted a close examination of the procedure that JAEA followed to evaluate CFFs in PLOHS and LORL events. It was confirmed that JAEA's Level-2 PSA quantified the phenomenal event trees was expanded in the three processes - the plant response process, the core damage process and the containment vessel response process - based on various analytical and experimental evidence and otherwise followed much the same basic evaluation procedures employed by JNES. As for PLOHS and LORL, quantitative evaluation of CFF was conducted according to the following procedures: Development of an event flow diagram, Development of a phenomenal event tree, Quantification of the phenomenal event tree, Evaluation of containment failure frequencies, and Evaluation of the effectiveness of the AM measures. In the evaluation of the PLOHS and LORL events, the following analytical codes were used; Plant dynamic characteristic analytical code (NALAP-II), Nuclear characteristics analytical system (ARCADIAN-FBR/MVP), Nuclear dynamics analysis code

  11. Status of Accelerator Driven Systems Research and Technology Development

    International Nuclear Information System (INIS)

    One of the greatest challenges for nuclear energy is how to properly manage the highly radioactive waste generated during irradiation in nuclear reactors. In order for nuclear power to exploit its full potential as a major sustainable energy source, there needs to be a safe and effective way to deal with this waste. Since 1995, several scenario studies have been conducted on different advanced nuclear fuel cycle and waste management options in various countries. Examples include the collaborative projects under “Global sustainable nuclear energy scenarios for long term development and deployment of nuclear energy” of the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative, and the scenario studies conducted under the auspices of the OECD Nuclear Energy Agency and the Euratom research project PATEROS — Partitioning and Transmutation European Roadmap for Sustainable Nuclear Energy. Some of the proposed long term nuclear fuel cycles include an innovative concept of a hybrid system for the transmutation of long lived radioisotopes. This is usually the called accelerator driven system (ADS) — or accelerator driven transmutation of waste (ATW) — and consists of a high power proton accelerator, a heavy metal spallation target that produces neutrons when bombarded by the high power beam, and a subcritical core that is neutronically coupled to the spallation target. The ADS, which has been developed in different countries for more than 40 years, is claimed to offer new prospects and advantages for the transmutation of high level radioactive waste. The ADS would convert highly radioactive material to non-radioactive material or material with a much shorter half-life. In addition, these hybrid systems can generate electricity during the conversion of transuranic waste. In 1997, under the guidance of its Technical Working Group on Fast Reactors (TWG-FR), the IAEA published IAEA-TECDOC-985, Accelerator Driven Systems: Energy

  12. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    Science.gov (United States)

    Scarlat, Raluca Olga

    approach to the PB-FHR annular pebble bed core cooled by fluoride salt mixtures generated a model that is called Pod. Pod. was used to show the resilience of the PB-FHR core to generation of hot spots or cold spots, due to the effect of buoyancy on the flow and temperature distribution in the packed bed. Pod. was used to investigate the PB-FHR response to ATWS transients. Based on the functional requirements for the core, Pod. was used to generate an optimized design of the flow distribution in the core. An analysis of natural circulation loops cooled by single-phase Boussinesq fluids is presented here, in the context of reactor design that relies on natural circulation decay heat removal, and design of scaled experiments. The scaling arguments are established for a transient natural circulation loop, for loops that have long fluid residence time, and negligible contribution of fluid inertia to the momentum equation. The design of integral effects tests for the loss of forced circulation (LOFC) for PB-FHR is discussed. The special case of natural circulation decay heat removal from a pebble bed reactor was analyzed. A way to define the Reynolds number in a multi-dimensional pebble bed was identified. The scaling methodology for replicating pebble bed friction losses using an electrically resistance heated annular pipe and a needle valve was developed. The thermophysical properties of liquid fluoride salts lead to design of systems with low flow velocities, and hence long fluid residence times. A comparison among liquid coolants for the performance of steady state natural circulation heat removal from a pebble bed was performed. Transient natural circulation experimental data with simulant fluids for fluoride salts is given here. The low flow velocity and the relatively high viscosity of the fluoride salts lead to low Reynolds number flows, and a low Reynolds number in conjunction with a sufficiently high coefficient of thermal expansion makes the system susceptible to

  13. Research about Automatic Adjustment Solution of the Advance Force at the Perffusion Drills Using Fluid Elements / Badanie Systemu Automatycznej Regulacji SIŁY Posuwu W Wiertnicach Udarowych Z Wykorzystaniem ELEMENTÓW PŁYNOWYCH

    Science.gov (United States)

    Cotetiu, Adriana; Cotetiu, Radu; Ungureanu, Nicolae

    2013-12-01

    This paper presents the actual solution used by Secoma Company and part of research regarding a personal solution concerning the implementation of the digital devices in the pressing strength's control of a pneumatic rotating drill, which is included in the structure of the drilling installation. The monostable fluidic element, which was proposed to be used, is a special device, with an incompressible fluid as supply jet and compressible fluid as command jet. The fluidic command proposed solution presents superior advantages given the existing variants and the automation solutions with electronic components. This is due to the higher security in hostile work environments (moist environment, with high methane gas contents, with fire danger, with high temperature) of their high feasibility and maintenance. For the practical achievement of the automated regulation with fluidic elements, of the type tested in the experimental plan, it is necessary to choose a monostabile fluidic amplifier for the prototype device, which respects several clear conditions regarding wall attachment angle and geometrical parameters. W pracy przedstawiono rozwiązanie stosowane przez firmę Secoma oraz omówiono część badań dotyczących rozwiązań w dziedzinie implementacji urządzeń cyfrowych do regulacji siły naporu w obrotowych wiertnicach pneumatycznych będących częścią urządzenia wiertniczego. Zaproponowano użycie mono-stabilnego elementu płynowego, będącego specjalnym urządzeniem zawierającym płyn nieściśliwy jako strugę zasilająca i płyn ściśliwy jako strugę sterującą. Rozwiązanie z wykorzystaniem elementu płynowego daje dodatkowe korzyści w odniesieniu do obecnie stosowanych rozwiązań zawierających komponenty elektryczne, przyczyniając się do poprawy bezpieczeństwa pracy w środowisku niebezpiecznym (w warunkach wysokiej wilgotności, wysokich stężeń metanu, zagrożenia pożarowego, wysokich temperatur), ponadto są one łatwe w użyciu i

  14. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    studied to assess the possibilities for using three-dimensional cores in training simulators. The core model results have been compared with the Loviisa WWER-type plant measurement data in steady state and in some transients. Hypothetical control rod withdrawal, ejection and boron dilution transients have been calculated with various three-dimensional core models for the Loviisa WWER-440 core. Several ATWS analyses for the WWER-1000/91 plant have been performed using the three-dimensional core model. In this context, the results of APROS have been compared in detail with the results of the HEXTRAN code. The three-dimensional Olkiluoto BWR-type core model has been used for transient calculation and for severe accident re-criticality studies. The one-dimensional core model is at present used in several plant analyser and training simulator applications and it has been used extensively for safety analyses in the Loviisa WWER-440 plant modernisation project. (orig.) 75 refs. The thesis includes also eight previous publications by author

  15. Thermal Power Of The TS-300B Refrigerator in the Aspects of Statistical Research / Moc Cieplna Chłodziarki TS-300B W Aspekcie Badań Statystycznych

    Science.gov (United States)

    Nowak, Bernard; Łuczak, Rafał

    2015-09-01

    refrigerator, both on the test stand in the manufacturer's laboratory and in the workings of underground mines. The evaluation of the measurement data distributions, as well as an analysis of the basic descriptive statistics of the mentioned variables were carried out, determining their measures of central tendency, location, dispersion and asymmetry. Artykuł dotyczy poprawy cieplnych warunków pracy w wyrobiskach górniczych kopalń podziemnych stosujących lokalne systemy chłodnicze. Rozważa się w nim skuteczność schładzania powietrza chłodziarką sprężarkową bezpośredniego działania typu TS-300B. Bardzo często, w wyniku niedotrzymania wymaganych warunków pracy wymienionego systemu chłodzenia powietrza, występują rozbieżności między prognozowanymi, a więc oczekiwanymi efektami jego pracy a rzeczywistością. Dlatego, dla poprawy skuteczności pracy tego systemu, opracowano, pod kątem efektywnego wykorzystania mocy chłodniczej parownika takiej chłodziarki, łatwe w zastosowaniu praktycznym kryteria jakości. Otrzymano je w postaci modeli statystycznych określających wpływ zmiennych niezależnych, tj. parametrów powietrza wlotowego do parownika (temperatury, wilgotności i wydatku objętościowego) oraz parametrów wody chłodzącej skraplacz (temperatury i wydatku objętościowego) na moc cieplną chłodnicy powietrza traktowaną jako zmienna zależna. Równania statystyczne opisujące pracę rozważanego systemu chłodzenia powietrza wyznaczono na podstawie wielorakiej regresji liniowej i nieliniowej. Utworzone funkcje zmodyfikowano poprzez zmianę wartości współczynników w przypadku regresji liniowej oraz współczynników i wykładników w przypadku regresji nieliniowej, przy zmiennych niezależnych. Otrzymano w ten sposób funkcje dogodniejsze w praktycznych wykorzystaniach. Korzystając z metod statystyki klasycznej oceniono jakość dopasowania funkcji regresji do danych eksperymentalnych. Porównano także wartości mocy cieplnych

  16. Development of integrated analytical tools for level-2 PSA of LMFBR

    International Nuclear Information System (INIS)

    analysis codes, ABAQUS and FINAS, were used to analyze the large shape deformation. Due to the high temperature of the coolant, the fuel pin cladding also suffers the high temperature creep. Once the cladding fails, the fission products (FPs) kept inside of the fuel pin will release into sodium. Then, all noble gases and a part of volatile FPs will transfer to the cover gas region. The remaining FPs in sodium will circulate in the primary cooling system. During the circulation, some part of FPs will deposit on the inner surface of the components. Such transfer behavior of FPs is analyzed by the ACTOR code. The progress of (ii) the core disruption phase is analyzed by the AZORES code. In the phase the core will melt and the molten core materials will fall on the bottom of the reactor vessel and cause the melt-through of the vessel wall. High temperature and high pressure in the containment vessel are caused by the sodium-concrete and the hydrogen burning. On the other hand, the neutronics of the various geometries of the disrupted core is solved by the ARCADIRN-FBR code. Using the neutronics data, the APK code performs the analyses of the super-prompt criticality and the prompt criticality phenomena. The derived information, such as the debris temperature, is used in the AZORES calculation of the core debris-concrete reaction. For ATWS (anticipated transient without scrum), the progress of the core disruption phase is quite different from the initial events with reactor scrum, such as PLOHS, particularly in the early time period. Development of an analysis code named ASTERIA was initiated to analyze ULOF (unprotected loss-of-flow). In (iii) the containment vessel (CV) response process, the timing of containment failure and the paths of FPs to the environment is important for the evaluation of the amount of the released FPs into the environment. These items are also analyzed by the AZORES code using the aerosol behavior module. PRD provides an analytical perspective that

  17. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    Science.gov (United States)

    Scarlat, Raluca Olga

    approach to the PB-FHR annular pebble bed core cooled by fluoride salt mixtures generated a model that is called Pod. Pod. was used to show the resilience of the PB-FHR core to generation of hot spots or cold spots, due to the effect of buoyancy on the flow and temperature distribution in the packed bed. Pod. was used to investigate the PB-FHR response to ATWS transients. Based on the functional requirements for the core, Pod. was used to generate an optimized design of the flow distribution in the core. An analysis of natural circulation loops cooled by single-phase Boussinesq fluids is presented here, in the context of reactor design that relies on natural circulation decay heat removal, and design of scaled experiments. The scaling arguments are established for a transient natural circulation loop, for loops that have long fluid residence time, and negligible contribution of fluid inertia to the momentum equation. The design of integral effects tests for the loss of forced circulation (LOFC) for PB-FHR is discussed. The special case of natural circulation decay heat removal from a pebble bed reactor was analyzed. A way to define the Reynolds number in a multi-dimensional pebble bed was identified. The scaling methodology for replicating pebble bed friction losses using an electrically resistance heated annular pipe and a needle valve was developed. The thermophysical properties of liquid fluoride salts lead to design of systems with low flow velocities, and hence long fluid residence times. A comparison among liquid coolants for the performance of steady state natural circulation heat removal from a pebble bed was performed. Transient natural circulation experimental data with simulant fluids for fluoride salts is given here. The low flow velocity and the relatively high viscosity of the fluoride salts lead to low Reynolds number flows, and a low Reynolds number in conjunction with a sufficiently high coefficient of thermal expansion makes the system susceptible to