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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. ATW economics

    International Nuclear Information System (INIS)

    A parametric systems model of the ATW [Accelerator Transmutation of (Nuclear) Waste] has been used to examine key system tradeoffs and design drivers on the basis of unit costs. This model has been applied primarily to the aqueous-slurry blanket concept for an ATW that generates net-electric power from the fissioning of spent reactor fuel. An important goal of this study is the development of essential parametric tradeoff studies to aid in any eventual engineering design of an ATW that would burn and generate net- electric power from spent reactor fuel

  3. ATW economics

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A.

    1993-07-01

    A parametric systems model of the ATW [Accelerator Transmutation of (Nuclear) Waste] has been used to examine key system tradeoffs and design drivers on the basis of unit costs. This model has been applied primarily to the aqueous-slurry blanket concept for an ATW that generates net-electric power from the fissioning of spent reactor fuel. An important goal of this study is the development of essential parametric tradeoff studies to aid in any eventual engineering design of an ATW that would burn and generate net- electric power from spent reactor fuel.

  4. ATW neutronics design studies

    International Nuclear Information System (INIS)

    The Accelerator Transmutation of Waste (ATW) concept has been proposed as a transuranics (TRU) (and long-lived fission product) incinerator for processing the 87,000 metric tonnes of Light Water Reactor used fuel which will have been generated by the time the currently deployed fleet of commercial reactors in the US reach the end of their licensed lifetime. The ATW is proposed to separate the uranium from the transuranics and fission products in the LWR used fuel, to fission the transuranics, to send the LWR and ATW generated fission products to the geologic repository and to send the uranium to either a low level waste disposal site or to save it for future use. The heat liberated in fissioning the transuranics would be converted to electricity and sold to partially offset the cost of ATW construction and operations. Options for incineration of long-lived fission products are under evaluation. A six-year science-based program of ATW trade and system studies was initiated in the US FY 2000 to achieve two main purposes: (1) ''to evaluate ATW within the framework of nonproliferation, waste management, and economic considerations,'' and (2) ''to evaluate the efficacy of the numerous technical options for ATW system configuration.'' This paper summarizes the results from neutronics and thermal/hydraulics trade studies which were completed at Argonne National Laboratory during the first year of the program. Core designs were developed for Pb-Bi cooled and Na cooled 840 MWth fast spectrum transmuter designs employing recycle. Additionally, neutronics analyses were performed at Argonne for a He cooled 600 MWth hybrid thermal and fast core design proposed by General Atomics Co. which runs critical for 3/4 and subcritical for 1/4 of its four year once-thin burn cycle. The mass flows and the ultimate loss of transuranic isotopes to the waste stream per unit of heat generated during transmutation have been calculated on a consistent basis and are compared. (Long-lived fission

  5. Operation of Kernkraftwerk Brunsbuettel stopped again

    International Nuclear Information System (INIS)

    With its decision of 28 August 1980, by which a rejection of the Schleswig Administrative Court was rejected, the Lueneburg Higher Administrative Court has re-established the suspensive power of an action against the license for the new start-up of Kernkraftwerk Brunsbuettel until 31 October 1980. According to the Court, the necessary modification of the emergency cooling system is not absolutely ensured; therefore, a formal licensing procedure is required in which the plaintiff can raise his objections already in the appeal proceedings. This is now made possible by the two-month delay in the enforcement. (HSCH)

  6. Building stop for Kernkraftwerk Brokdorf extended

    International Nuclear Information System (INIS)

    The Higher Administrative Court at Lueneburg has confirmed in its decision of Oct. 17th, 1977 - VII OVG B 22, 42/77 - the building stop for the Kernkraftwerk Brokdorf with the time limit 'until an application, ready for examination, concerning an intermediate storage site for depositing spent fuel elements has been filed, and geological investigations for proving the suitability of a certain site for the final storage of radioactive wastes have been initiated. Thus, the court heads for a kind of a compromise, the justification and the legal grounds of which one can have different views on: On the one hand, the nuclear hazard potential, which only takes effect at a much later stage, has here already been anticipated in the interest of future safety in a preliminary legal protection procedure concerning a 1st partial erection licence, thus, without need, intensifying the supply crisis to be expected in meeting the energy demand during the next decade, as well as causing the future operator heavy financial losses. On the other hand, the project as such is not dismissed, but its execution is probably only delayed. (orig.)

  7. 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

  8. 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.

  9. 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

  10. Kernkraftwerke Lippe-Ems (KLE). 1997 annual report

    International Nuclear Information System (INIS)

    The Kernkraftwerke Lippe-Ems GmbH (KLE) is the operator of the Emsland reactor station at Lingen (Ems), equipped with a 1300 MW PWR. The partners of KLE are VEW ENERGIE, PreussenElektra, and RWE Energie. The nuclear power station was used over the reporting period as a base-load power plant operating under full-load conditions. The total gross electricity output was 11235 million kWh, which is the highest annual total output ever since start-up. The net electricity generation over the reporting amounted to 10650 million kWh. The personnel employed with KLE at the end of the reporting year amounts to 287. The reduction of the nominal capital of the GmbH to 900 000 TDM, decided by the partners in 1996, became legally effective in April 1997. The dominant features of the structure of expenses of KLE are expenses for the fuel inventory and spent fuel management, for the legally required reserves for decommissioning, and writeoffs. The contractual electric power rates negotiated with the partners cover all load and capacity-dependent expenses; in addition, the partners receive their contractual shares of the profit resulting from interests paid on the nominal capital. (orig./CB)

  11. Simulation of ATWS conditions in pressurized water reactors; Simulation von ATWS-Transienten in Druckwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Mittag, Siegfried; Rohde, Ulrich; Grundmann, Ulrich; Weiss, Frank-Peter [Forschungszentrum Dresden-Rossendorf, Dresden (Germany). Inst. fuer Sicherheitsforschung

    2009-02-15

    Safety analyses of nuclear power plants use ATWS (Anticipated Transients without Scram) as a term covering also events involving assumed failure of the reactor scram system. In this type of failure of the reactor scram system, power development in the reactor core is determined only by neutron kinetics feedback via the fuel temperature, moderator temperature and moderator density. If borated coolant is supplied, there is the additional feedback from the boron concentration. For nuclear power plants, coupled code complexes are developed and used which are made up of a thermohydraulic plant code and a 3D neutron kinetics model of the reactor core. These lend themselves to the use in analyses of ATWS states. The work presented here about the ATWS problem was performed in the interest of a consistent application of the DYN3D 3D neutron kinetics code in combination with the ATHLET thermohydraulic system code, and the quantification of differences resulting from variations of initial and boundary conditions. The DYN3D/ATHLET coupled code complex was validated by recalculation of various operating transients and by solving benchmark problems. The article contains results computed taking into account the influence of systems engineering and neutron kinetics boundary conditions. The calculations shown are methodological studies in no way demonstrating proof. (orig.)

  12. An analysis on a highly degraded ATWS scenario

    International Nuclear Information System (INIS)

    This paper presents the result of the analysis using the MAAP code to model a highly degraded ATWS scenario. The analysis was performed as a part of the overall Accident Management Program evaluation under development by Taiwan Power Company for its Chinshan Nuclear Power Station (CSNPS). It is concluded that the CSNPS has the capability to withstand this highly degraded ATWS scenario if its current Emergency Operating Procedures are executed during the event

  13. Scaling and cost tradeoffs for ATW: preliminary considerations

    International Nuclear Information System (INIS)

    The use of accelerator-produced spallation neutrons to transmute actinide and long-lived fission-products portends a means to alleviate requirements for deep-geological disposal. The accelerator performance, target-blanket physics, chemical-processing requirements, and overall systems engineering are closely coupled in determining the economic incentives for the Accelerator Transmutation of Waste (ATW). Preliminary estimates of and insight into the economics of a net-power-producing ATW are provided by a simplified (analytic) cost-based systems model. Even for large-capacity systems, the accelerator dominates the economics and technology for the ATW cases examined. Since the accelerator represents an important (∼50%) add-on cost to an ATW-based power plant, reducing the accelerator requirement by increasing the blanket neutron multiplication, increasing the thermal-conversion efficiency, and reducing neutron leakage and parasitic absorption in the target-blanket assembly are main avenues for improving the economic prospects of an ATW that would be fuelled with low-reactivity, actinide ''waste'' generated by light-water reactors (LWRs). This route to improved systems has strong implications for the thermalhydraulic, neutronic, and chemical-processing design of the ATW. (author) 1 tab., 4 refs

  14. 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

  15. Separations technologies supporting the development of a deployable ATW 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 first several years of the program will be directed toward an elucidation of related technical issues and to the establishment, by means of comprehensive trade studies, of an optimum configuration of the elements of the chemical processing infrastructure required for support of the total ATW system. By adopting this sort of disciplined systems engineering approach, it is expected that development and demonstration costs can be minimized and that it will be possible to deploy an ATW system that is an environmentally sound and economically viable venture

  16. 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

  17. Document management. Interleaf administrates ``handbooks`` at the Gundremmingen nuclear power plant; Dokumenten-Management. Interleaf verwaltet ``Handbuecher`` im Kernkraftwerk Gundremmingen

    Energy Technology Data Exchange (ETDEWEB)

    Rademeier, E. [DV Consult, Heiligkreuzsteinach (Germany)

    1996-08-01

    The Gundremmingen nuclear power plant with its two 1344 MW units (B and C) is the largest boiling water reactor site in Germany. It is operated by Kernkraftwerke Gundremmingen Betriebsgesellschaft (KGB), a daughter of RWE Energie, Essen, and BAYERNWERK, Munich. For two years, KGB has been using Interleaf software to prepare and distribute its handbook. (orig.) [Deutsch] Das Kernkraftwerk (KKW) Gundremmingen ist mit seinen beiden 1 344-MW-Bloecken (B und C) der groesste Siedewasser-Standort in Deutschland. Betreiber ist die Kernkraftwerke Grundremmingen Betriebsgesellschaft mbH (KGB), eine Tochter von RWE Energie AG, Essen, und der Bayernwerk AG, Muenchen. Seit zwei Jahren setzt die KGB bei der Erstellung und Verteilung eines Handbuchs auf die Dokumenten-Management-Software von Interleaf. (orig.)

  18. 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.

  19. 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

  20. 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.

  1. Strategy and Economic Prospect of Back-end Cycle through ATW

    International Nuclear Information System (INIS)

    Strategy and economic prospect of back-end cycle through ATW has been studied. Nuclear fuel cycle through ATW is a single stratum of back-end cycle. By ATW, volume of spent fuel which should be disposed in long term can be reduced from 70,000 MHTM to 3,000 MHTM and half-life of spent fuel can be reduced from 15,700,000 years to 300 years. Strategic values of the ATW cycle are to prevent proliferation risk and to reduce the uncertainty of long term dispose. Economic prospect of the ATW cycle will give some advantages on reducing of spent fuel volume and its disposal period, and producing electricity. (author)

  2. 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.

  3. 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

  4. 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

  5. MSIV closure ATWS mitigation of SBWR with the standby liquid control system

    Energy Technology Data Exchange (ETDEWEB)

    Khan, H.J.; Cheng, Hsiang S.; Rohatgi, U.S.

    1994-09-01

    An Anticipated Transient Without Scram (ATWS) initiated by inadvertent closure of the Main Stream Isolation Valve has been analyzed using the RAMONA-4B code of Brookhaven Laboratory (1994). The Simplified Boiling Water Reactor (SBWR) operating in natural circulation is designed with many passive safety features. This analysis demonstrates the effectiveness of the heat-removal system during an ATWS, followed by shut down of the reactor through injection of boron into the reactor core from the Standby Liquid Control System (SLCS).

  6. 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

  7. 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)

  8. 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)

  9. Kernkraftwerk Sued - the Wyhl decision of March 30, 1982/July 6, 1982

    International Nuclear Information System (INIS)

    This volume contains the complete Wyhl decision of the Higher Administrative Court of Baden-Wuerttemberg of March 30, 1982-X575/77, X578/77, X583/77 which cores 548 pages. According to the press release the complete decision has been delivered to the counsels of the parties to the lawsuit on July 7, 1983; on the appeal of the defendant Land and the attending Kernkraftwerk Sued GmbH the Higher Administrative Court has amended the decisions of the Administrative Court of Freiburg of March 14, 1977 and has rejected the actions of nine citizens against the first part construction permit. Moreover, the senate has sent to the parties to the lawsuit the decision of March 30, 1982, by which the value in dispute for these proceedings on appeal is fixed to DM 180,000. The time for the lodging of an appeal, which has been admitted by the senate in this process, begins with the delivery of the completely well-founded decision. Moreover, the volume contains a 10 pages summary of contents of the decision and a table of contents of the reasons for the decision. (orig./HSCH)

  10. 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.

  11. 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

  12. 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)

  13. Modernization of kernkraftwerk Beznau's core monitoring system with Studsvik's Gardel system

    International Nuclear Information System (INIS)

    The Kernkraftwerk Beznau (KKB), is located in Switzerland. The selection of a new core monitoring system to replace the existing one at KKR was driven by the following criteria: 1) Improve operational performance by reducing uncertainties of calculated parameters associated with thermal margin (e.g., radial pin power peaking factors, etc.); 2) Improve operational performance by implementing a pin-based PCI model; 3) Improve operational performance by implementing on-line DNBR calculations based on plant-specific DNB correlations; 4) Provide the reactor operators and reactor engineers a reliable, intuitive tool to assist them in evaluating unexpected operational conditions; 5) Eliminate tedious and expensive cycle initialization procedures; 6) Reduce costs for hardware and software support for the core monitoring system; 7) Minimize impact on plant process computer. After a one-year evaluation period, KKB chose GARDEL core monitoring system. GARDEL is based on Studsvik's well-established in-core fuel management code system, consisting primarily of the lattice code, CASMO, and the core physics model, SIMULATE. design in the world. In addition to the Studsvik physics model, GARDEL includes a highly automated core physics model update throughout the cycle, triggered by monitoring changes of plant process computer signals. All data is archived in a flexible, highly efficient database system. GARDEL also includes a flexible, graphical user interface as well as international language support. All data - both calculated parameters, as well as collected plants signals - are easily accessed via the user interface for analysis or via included automatic reporting functions. Multiple users can use the system simultaneously. GARDEL is unique in modularity and flexibility, as it runs on a variety of hardware systems with no proprietary hardware required. It is important to note that GARDEL does not run on the plant computer. No modifications to the plant computer are required for

  14. 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

  15. 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

  16. Summary: 'A roadmap for developing Accelerator Transmutation of Waste (ATW) technology'. A report to Congress

    International Nuclear Information System (INIS)

    The U.S. Congressional Conference Report accompanying the Fiscal Year 1999 Energy and Water Development Appropriation Act directed the U.S. Department of Energy, through its Office of Civilian Radioactive Waste Management, to conduct a study of accelerator transmutation of waste (ATW). It was transmitted to the U.S. Congress on November 1, 1999. The Report to Congress made it clear that the U.S. Administration, in transmitting the report, was not taking a position either way on those recommendations. If an ATW program were to be undertaken in the U.S., the pace and funding would have to be evaluated and planned in light of the currently unproven technologies involved, the potential benefits, and overall Government budget priorities. (author)

  17. 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

  18. 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.

  19. 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

  20. 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.

  1. A roadmap for the development ATW technology: Systems scenarios and integration

    International Nuclear Information System (INIS)

    As requested by the US Congress, a roadmap has been established for development of ATW Technology. The roadmap defines a reference system along with preferred technologies which require further development to reduce technical risk, associated deployment scenarios, and a detailed plan of necessary R and D to support implementation of this technology. Also, the potential for international collaboration is discussed which has the potential to reduce the cost of the program. In addition, institutional issues are described that must be addressed in order to successfully pursue this technology, and the benefits resulting from full implementation are discussed. This report uses as its reference a fast spectrum liquid metal cooled system. Although Lead-Bismuth Eutectic is the preferred option, sodium coolant is chosen as the reference (backup) technology because it represents the lowest technical risk and an excellent basis for estimating the life cycle cost of the systems exists in the work carried out under DOE's ALMR (PRISM) program. Metal fuel and associated pyrochemical treatment is assumed. Similarly a linear accelerator has been adopted as the reference. A reference ATW plant was established to ensure consistent discussion of technical and life cycle cost issues. Over 60 years of operation, the reference ATW plant would process about 10,000 tn of spent nuclear reactor fuel. This is in comparison to the current inventory of about 40,000 tn of spent fuel and the projected inventory of about 86,000 tn of spent fuel if all currently licensed nuclear power plants run until their license expire. The reference ATW plant was used together with an assumed scenario of no new nuclear plant orders in the US to generate the deployment scenario for ATW. In the R and D roadmap, key technical issues are identified and timescales proposed for the resolution of these issues. For the accelerator the main issue is the achievement of the necessary reliability in operation. To avoid

  2. 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

  3. Texture optimization of Ni-5at.%W for coated conductor applications

    International Nuclear Information System (INIS)

    For the application of Ni-5at.%W as substrate for high-temperature superconductors the development of a very intense Cube texture is essential. Therefore, the texture development in Ni-5at.%W was studied in detail by experiments and computer simulations. Cold rolling of the material resulted in a pronounced β-fiber texture which transformed by recrystallization into a strong Cube texture and then sharpened further during subsequent grain growth. The cold rolling process was found to be limited by the occurrence of deformation inhomogeneities to strains of ε = 3-4 and led to the formation of Cube deformation bands. These bands formed very effective nucleation sites during recrystallization if ε > 2.5. The deformation texture evolution could be successfully simulated with the grain interaction model (GIA) when a certain amount of random shear deformation was allowed. During annealing of cold rolled Ni-5at.%W recrystallization nuclei emerged from the deformed Cube bands earlier and with higher frequency than for any other orientation and the Cube volume fraction increased from 0.1% to 50%. Furthermore, the Cube nuclei had the highest mean misorientation of all orientations and thus they had the highest fraction of mobile high angle boundaries (HAGB). Together with the nucleation advantage this led to a bimodal size distribution of large Cube grains and smaller grains of other orientations after recrystallization was completed. If the annealing temperature for recrystallization was too high this nucleation and growth advantage of the Cube grains was reduced. Thus, a two-step annealing process with a lower temperature for recrystallization and a higher temperature for the grain growth stage was found to be beneficial. With the statistical recrystallization texture model (StaRT) the development of the recrystallization texture was simulated based on results of a previous GIA simulation and an experimentally determined nucleation spectrum. A delay function was

  4. Polonium release from an ATW burner system with liquid lead-bismuth coolant

    International Nuclear Information System (INIS)

    The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods

  5. 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.

  6. 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

  7. 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.

  8. Safety assessment of the Indonesian multipurpose reactor RSG-GAS against ATWS and hypothetical accidents

    International Nuclear Information System (INIS)

    Investigation on ATWS and hypothetical accidents for the Indonesian Multipurpose Reactor RSG-GAS have been undertaken by computer simulation technique. Two computer codes, namely RELAP5 and PARET-ANL, were used as the main tools. The RELAP5 was utilized to perform system analysis while the PARET-ANL code was used to perform the reactor core analysis in more detail. Two different models have been applied as a basis of the simulation: Typical Working Core model (IWC-model) consisting of four regions with different radial power factors; and the hot-channel model consisting of two regions with different radial power factors. Both RELAP5 ad PARET-ANL results showed that in the occurrence of ATWS, failure on fuel element or fuel plate was limited to the region with the most highest power factor. The results also indicated that no high pressure development occurs in that region, so that mechanical damage on the fuel element or other core components due to pressure shock did not happen.(author)

  9. 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.)

  10. ``Ring monitoring system`` - a new concept for gamma dose rate monitoring shown on the example of the nuclear power plant Pilgrim Station; ``Ring Monitoring System`` - ein neues ODL-Ueberwachungskonzept, dargestellt am Beispiel des Kernkraftwerks Pilgrim Station

    Energy Technology Data Exchange (ETDEWEB)

    Holzheimer, C. [Hoermann Systemtechnik GmbH, Kirchseeon/Muenchen (Germany); Kristl, H. [Hoermann Systemtechnik GmbH, Kirchseeon/Muenchen (Germany)

    1994-12-31

    The `Ring Monitoring System` represents a measurement and alerting system for the supervision of gamma dose rate in the vicinity of a nuclear power plant. This system especially meets the customer demand for direct and continuous access upon the monitoring stations from different central computer systems. (orig.) [Deutsch] Mit dem `Ring Monitoring System` wurde ein Gammadosisleistungs-Messsystem zur Ueberwachung eines Kernkraftwerkes realisiert, welches vor allem der Kundenforderung nach direktem und kontinuierlichem Zugriff auf die Messstationen von verschiedenen Auswertezentralen aus gerecht wird. (orig.)

  11. 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)

  12. Thermal fatigue due to beam interruptions for an ALMR-type ATW

    International Nuclear Information System (INIS)

    Accelerator-driven subcritical reactors have been proposed for tasks such as accelerator transmutation of waste (ATW). In such a device, long-lived radioactive fission products and transuranic elements from spent fuel would be converted into short-lived radioactive isotopes and stable isotopes. One concern about such devices is that current proton accelerators typically experience beam interruptions many times per day. The beam interruptions last from seconds to hours. A beam interruption leads to a temperature transient similar to, but faster than, that from a reactor scram. These temperature transients cause thermal fatigue that can eventually lead to structural failures in various reactor components. The objective of the work reported here was to determine the design implications of this thermal fatigue

  13. Estimates of thermal fatigue due to beam interruptions for an ALMR-type ATW

    International Nuclear Information System (INIS)

    Thermal fatigue due to beam interruptions has been investigated in a sodium-cooled ATW (accelerator transmutation of waste) using the Advanced Liquid Metal mod B design as a basis for the subcritical source driven reactor. A keff of 0.975 was used for the reactor. Temperature response in the primary coolant system was calculated, using the SASSYS-1 code, for a drop in beam current from full power to zero in 1 microsecond. Temperature differences were used to calculate thermal stresses. Fatigue curves from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code were used to determine the number of cycles various components should be designed for, based on these thermal stresses. (authors)

  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. Design criteria and mitigation options for thermal fatigue effects in ATW blankets

    International Nuclear Information System (INIS)

    Thermal fatigue due to beam interruptions is an issue that must be addressed in the design of an ATW blanket. Two different approaches can be taken to address this issue. One approach is to analyze current ATW blanket designs in order to set interrupt frequency design limits for the accelerator. The other approach is to assume that accelerator reliability can not be guaranteed before design and construction of the blanket. In this approach the blanket must be designed so as to accommodate an accelerator with a beam interruption frequency significantly higher than current high power accelerators in order to provide a margin of error. Both approaches are considered in this paper. Both a sodium cooled blanket design and a lead-bismuth cooled blanket design are considered. Thermal hydraulic analysis of the blanket for beam interruption transients is carried out with the SASSYS-1 systems analysis code to obtain the time histories of the coolant temperatures in contact with structural components. These coolant temperatures are then used in a detailed structure temperature calculation to obtain structure surface and structure average temperatures. The difference between the average temperature and the surface temperature is used to obtain thermal strains. Low cycle fatigue curves from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code are used to determine the number of cycles that the structural components can endure, based on these strains. Calculations are made for base case designs and for a number of mitigation options. The mitigation options include using two separate accelerators to provide the beam, reducing the thickness of the above core load pads in the subassemblies, increasing the coolant flow rate or reducing power in order to reduce the core temperature rise, and reducing the superheat in the once-through steam generator. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Cobo, J.L., E-mail: jlcobos@iqn.upv.es [Universitat Politècnica de València, Thermal-Hydraulics and Nuclear Engineering Group (TIN), Institute for Energy Engineering (IEE), Valencia (Spain); Escrivá, A., E-mail: aescriva@iqn.upv.es [Universitat Politècnica de València, Thermal-Hydraulics and Nuclear Engineering Group (TIN), Institute for Energy Engineering (IEE), Valencia (Spain); Mendizabal, R., E-mail: rmsanz@csn.es [Consejo de Seguridad Nuclear, 28040 Madrid (Spain); Pelayo, F., E-mail: fpl@csn.es [Consejo de Seguridad Nuclear, 28040 Madrid (Spain); Melara, J., E-mail: jls@iberdrola.es [IBERINCO, IBERDROLA Ingeniería y Construcción, Madrid (Spain)

    2014-10-15

    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.

  18. 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

  19. 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

  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. Performance comparison of liquid metal and gas cooled ATW system point designs

    International Nuclear Information System (INIS)

    As part of the Advanced Accelerator Application (AAA) program in the U.S., preliminary design studies have been performed at Argonne National Laboratory (ANL) and Los Alamos National Laboratory (LANL) to define and compare candidate Accelerator Transmutation of Waste (ATW) systems. The studies at ANL have focused primarily on the transmutation blanket component of the overall system. Lead-bismuth eutectic (LBE), sodium, and gas cooled systems are among the blanket technology options currently under consideration. This paper summarizes the results from neutronics trade studies performed at ANL. Core designs have been developed for LBE and sodium cooled 840 MWt fast spectrum accelerator driven systems employing re-cycle. Additionally, neutronics analyses have been performed for a helium-cooled 600 MWt hybrid thermal and fast spectrum system proposed by General Atomics (GA), which is operated in the critical mode for three cycles and in a subcritical accelerator driven mode for a subsequent single cycle. For these three point designs, isotopic inventories, consumption rates, and annual burnup rates are compared. The mass flows and the ultimate loss of transuranic (TRU) isotopes to the waste stream per unit of heat generated during transmutation are also compared on a consistent basis. (author)

  2. 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

  3. Comparison of different variants of ATWS type accident calculations by means of the ATHLET and RELAP codes

    International Nuclear Information System (INIS)

    The results of thermal hydraulic analyses of anticipated transients without scram (ATWS) served as the basis for the new Emergency Operating Procedures for WWER-440/V-213 reactors. Because of the differences in the behavior of parameters in the calculations by the ATHLET code (for the Dukovany NPP) and by the RELAP code (for the Bohunice V2 plant), the major parameters in selected calculations were compared and the differences were explained on graphs. The starting calculations, in which no operator intervention was taken into account, were used for the comparison. (P.A.)

  4. 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

  5. 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

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

    International Nuclear Information System (INIS)

    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 the 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 systems) 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. The authors 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

  7. 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.

  8. 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

  9. 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)

  10. 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...

  11. 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.

  12. Core thermal hydraulic response of using HPCI for MSIV closure event at ATWS condition for a BWR

    International Nuclear Information System (INIS)

    The High Pressure Core Injection (HPCI) system is part of a BWR Emergence Core Cooling System (ECCS). HPCI is actuated to insure the adequacy of the water inventory in plant transient or accident conditions. Different HPCI configurations are used in BWR plant designs. One type is that the HPCI is injected outside the core shroud, the other type is that the HPCI is partly injected outside the core shroud and partly into the core through the core spray loop. This paper presents the results of using different HPCI flow paths during MSIV event at ATWS condition. A typical BWR-4 plant is modeled and the thermal hydraulic response is obtained. The simulation is performed using the TRAC/BF1 code. The results of using divided HPCI path are compared with those using normal HPCI path and those injecting water directly into the core. (author)

  13. Analysis of a high pressure ATWS [anticipated transient without scram] with very low make-up flow

    International Nuclear Information System (INIS)

    A series of calculations were performed to analyze the response of General Electric Company's (GE) advanced boiling water reactor (ABWR) during an anticipated transient without scram (ATWS). This work investigated the early plant response with an assumed failure or manual inhibit of the high pressure core flooder (HPCF). Consequently, the reactor core isolation cooling (RCIC) and control rod drive (CRD) systems are the only sources of high pressure injection available to maintain core cooling. Steam leaving the reactor pressure vessel was diverted to the pressure suppression pool (PSP) via the steam line and the safety relief valves. The combination of an unscrammed core and the CRD and RCIC injection sources make this a particularly challenging transient. System energy balance calculations were performed to predict the core power and PSP heat-up rate. The amount of vessel vapor superheat and the PSP temperature were found to significantly affect the resultant core power. Consequently, detailed thermal-hydraulic calculations were performed to simulate the system response during the postulated transient. 15 refs., 15 figs., 4 tabs

  14. 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

  15. ATWS events revisited. The impact of varying initial and boundary conditions analysed with the GRS coupled code system ATHLET-QUABOX/CUBBOX

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, K.D.; Velkov, K.; Pautz, A. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany). Forschungsinstitute

    2010-05-15

    According to the German RSK guidelines (RSK-Leitlinien, /RSK96/) and the update of the German body of nuclear regulatory rules (Aktualisierung des Kerntechnischen Regelwerks, /REG08/), the assessment of ATWS (Anticipated Transients without Scram) transients is required as an essential part of certain licensing procedures for German nuclear power plants. In the RSK comment /RSK05/ of 7/7/2005 these rules have been stated precisely: the assessments must aim to demonstrate that for anticipated transients with postulated failure of the fast shut down system, the criteria for the limitation of the primary pressure and the coolability of the reactor core are obeyed. Additionally, the long term subcriticality of the core has to be demonstrated. Since ATWS are considered beyond-design basis accidents (BDBA, or according to /REG08/, so called class 4a events), best-estimate assumptions on system availability and initial/boundary conditions can be used for the analyses. In German nuclear regulatory practice, point kinetics is typically used to capture the inherent feedback mechanisms of the reactor core (employing the so called void curve to describe the dependency of core reactivity on coolant density), and the reactor core is assumed to operate at 100% nominal power and at xenon equilibrium. In this contribution, we take a closer look at this approach, considering also non-xenon-equilibrium cases at different load factors, and using 3D-neutronics methods instead of the simple point kinetics method. (orig.)

  16. 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)

  17. 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.)

  18. 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)

  19. Impact of modeling effects, initial and boundary conditions on performing ATWS analysis with the coupled system code ATHLET/BIPR-VVER

    International Nuclear Information System (INIS)

    The work demonstrates the successful application of coupled thermal-hydraulics/neutron-kinetics system codes by performing analyses of complex transients. Two simulation cases (Case no.1 and Case no.2) are compared for a NPP with VVER-1000 reactor (type V-320). The two cases differ in core layout, the initial and boundary conditions and in the nodalization schemas of the reactor pressure vessel. The main objective is to identify the importance of modelling differences on main NPP parameter histories for an ATWS case with loss of main feed water. This comparison can contribute to further developments and optimizations by performing safety analyses with coupled codes. The analyses have been carried out with the coupled system code ATHLET/BIPR-VVER, developed to perform best estimate simulations of three-dimensional neutron-kinetics and thermal-hydraulics processes in VVER reactors. (author)

  20. Analysis of the Loss of Feedwater scenario for RBMK-1500 by the coupled code Athlet-Quabox/Cubbox taking into account late reactor trip signals and ATWS conditions

    International Nuclear Information System (INIS)

    For the accident analysis of RBMK-1500 reactors the GRS system code ATHLET and the 3-dimensional-reactor core model QUABOX/CUBBOX have been adapted to the special requirements of this type of reactor and a coupled version of both codes for RBMK-1500 investigations has been developed. The coupled model was applied to analyse the event Loss of Feedwater by best estimate methods to investigate the consequences of failures of reactor trip initiations. Therefore, late trip signals where credited to determine the available time for a safe shutdown, relevant for the demands on the newly developed diverse shut down system for Ignalina NPP Unit 2. These results are also compared to cases assuming ATWS conditions. The results show that a reactor trip at t equals 143 s from the begin of the transient, neglecting two preceding trip initiations, is sufficient for a safe shut down. If the shutdown function fails (ATWS), the transient behaviour is fully determined by the reactivity feedback, which is mainly dominated by the change of fuel temperature. The loss of feedwater provides worse cooling conditions and the fuel temperatures are rising significantly causing a power reduction to about 10 %. This power reduction is not sufficient to avoid that cladding temperature limits of 1200 C degrees and pressure tube temperature limits of 650 C degrees are exceeded. The transient provides an example to demonstrate the advantage of the coupled code system in comparison to point kinetic models since local effects influencing the reactivity behaviour are taken into account and maximum loaded channels are described in an adequate manner. (authors)

  1. 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).

  2. 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.

  3. 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

  4. 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.)

  5. 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.

  6. 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 .

  7. 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

  8. Safety requirements for nuclear power plants; Sicherheitsanforderungen an Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-03-03

    The safety requirements for German nuclear power plants from March 3, 2015 are based on the safety requirements for German nuclear power plants from November 22, 2012 including the changes and corrections approved by the German Ministry for environment and reactor safety and the Federal State authorities.

  9. Transmutation calculations for the accelerator transmutation of waste (ATW) program

    International Nuclear Information System (INIS)

    The disposal of radioactive waste by the transmutation of long-lived radionuclides is being considered; now using neutrons produced with an intense beam of 1.6-GeV protons on a Pb-Bi target. Study teams have been active in the areas of accelerator design, beam transport, radiation transport, transmutation, fluid flow and heat transfer, process chemistry and system analyses. Work is of a preliminary and developmental nature. Here we describe these preliminary efforts in transmutation calculations; the tools developed, status of basic nuclear data, and some early results. These calculations require the description of the intensity and spectrum of neutrons produced by the beam, the distribution of nuclides produced in the medium-energy reactions, the transport of particles produced by the beam, the transmutation of the target materials and transmutation products, and the decay properties of the inventory of radionuclides produced

  10. ATWS performance analysis of MDP and ALMR designs

    International Nuclear Information System (INIS)

    KALIMER development program is now being lead by KAERI with a goal of developing a liquid metal reactor with enhanced safety. The goal of enhanced safety can be accomplished in large part by the adoption of a safety philosophy which emphasize utilization of natural , or inherent thermal, mechanical, hydraulic, and neutronic responses to normal and off-normal operating conditions. Since the KALIMER program is now in its concept development phase, there is a need of assessing the inherent safety potential of ALMR and MDP designs in order to accommodate the lessons learned form these designs and establish a knowledge-base for the development of safety and transient analysis code. This report summarizes the results of transient analyses for ALMR and MDP reference core designs by using ARIES code of Ge. (author). 6 tabs., 41 figs

  11. Controlled energy generation from nuclear fusion. 60th year atw

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, Georg [Pintsch Bamag AG, Frankfurt am Main (Germany)

    2015-02-15

    Prospects increase, that with a controlled process of nuclear fusion one day an additional nuclear energy source will be commercially exploitable. In what follows, scientific principles according to the most recent research will be presented. Since approximately 30 years we are aware of the fact, that energy in form of light and heat provided by the sun and other fixed stars since over four billions years resulted from reactions of atomic nuclei. A series of such reactions became known which are considered for 'thermonuclear' processes, for example the carbon cycle by Bethe, where hydrogen is converted into helium. Most of the reflections and experiments dealt until 1938 with the reaction between nuclei of light elements. The possibility of splitting heavy nuclei was not anticipated. Its discovery by Hahn and Strassmann was a complete surprise - so to speak a rash reaction to release energy at the end of the element row. This 'way out' captured the interest of nuclear physicist for more than a decade. Only today, by starting to construct big nuclear power plants - only today, being able to assess the possibilities and limitations of this technology, the idea of energy generation through nuclear fusion steps into the foreground of nuclear research.

  12. Cross-Validation of Neutronic Tools for ATW System Design

    International Nuclear Information System (INIS)

    A cross validation study has been conducted on several fast reactor design codes to determine if different calculational algorithms lead to modeling inconsistencies. The study utilized the lattice physics codes, MC2 and TRANSX, and diffusion codes, DIF3D and UM2DB. Based on the reliability of Monte Carlo techniques, MCNP served as a benchmark to validate the diffusion codes. The validation study involved comparing the values of keff, fluxes, fission rates, capture rates, net core leakages, and power distributions for two fast reactor designs. The only observable modeling inconsistency was found in the reflector flux spectrum calculation. The diffusion codes that used TRANSX-generated cross sections resulted in a 30% over-prediction in the flux spectrum. A sensitivity study conducted on the UM2DB homogenous model revealed that an error in the TRANSX-generated reflector scattering matrix caused the 30 % over-prediction in the flux. While not proven in this paper, we believe that TRANSX is not responsible for the inconstancy in the reflector scattering matrix. Instead, we believe the inconstancy is a result of an error in the MATXS11 library. (authors)

  13. The Los Alamos accelerator driven transmutation of nuclear waste (ATW) concept development of the ATW target/blanket system

    International Nuclear Information System (INIS)

    The studies carried out in the frame of the Accelerator Driven Transmutation Technology (ADTT) program developed at Los Alamos in order to solve the nuclear waste problem and to build a new generation of safer and non-proliferant nuclear power plants, are presented

  14. Steam generator behaviour and measures taken to assure their function at Kernkraftwerk Beznau

    International Nuclear Information System (INIS)

    The Beznau nuclear power plants each have two U-tube steam generators. Failures in the Inconel-600 tube already occurred after quite a short operating period. When looking for the causes, it was found that apart from chemistry and materials selection, also steam generator design and hydraulic conditions play a major role. For the time being, steam generators need not be replaced as there are various methods of repair. Recently, defective heating tubes are sealed with hydraulically braced journals or even repaired by internal tube welding. (orig.)

  15. Kernkraftwerke Lippe-Ems GmbH (KLE). 1995 annual report

    International Nuclear Information System (INIS)

    The tasks and activities of the operators of the Emsland reactor station (KKE) are reported. The year-end statement of 1995 includes many details about the financial situation, (such as the balance sheet, profit and loss account, etc.). (UA)

  16. Safety provision for nuclear power plants during remaining running time; Sicherheitsgewaehrleistung fuer Kernkraftwerke waehrend der Restlaufzeit

    Energy Technology Data Exchange (ETDEWEB)

    Rossnagel, Alexander [Kassel Univ. (DE). Kompetenzzentrum fuer Klimaschutz und Klimaanpassung (CliMA); Hentschel, Anja [Kassel Univ. (Germany). Forschungsschwerpunkt Umweltrech

    2012-06-15

    With the phasing-out of the industrial use of nuclear energy for the power generation, the risk of the nuclear power plants has not been eliminated in principle, but only for a limited period of time. Therefore, the remaining nine nuclear power plants must also be used for the remaining ten years according to the state of science and technology. Regulatory authorities must substantiate the safety requirements for each nuclear power plant and enforce these requirements by means of various regulatory measures. The consequences of Fukushima must be included in the assessment of the safety level of nuclear power plants in Germany. In this respect, the regulatory authorities have the important tasks to investigate and assess the security risks as well as to develop instructions and orders.

  17. Nuclear power plants in Germany. Performance 2011; Kernkraftwerke in Deutschland. Betriebsergebnisse 2011

    Energy Technology Data Exchange (ETDEWEB)

    Wesselmann, Christopher (comp.)

    2012-07-01

    The report on the performance nuclear power plants in Germany in 2011 includes the operational results, safety analyses, revisions, quality management, environmental management, status of radioactive waste management, and eventual programs (TACIS, WANO) for the nuclear power plants Biblis A, Biblis B, Brokdorf KBR, Brunsbuettel KKB, Emsland KKE, Grafenrheinfeld KKG, Grohnde KWG, Gundremmingen KRB B, Gundremmingen KRB C, Isa KK1, Isar KK2, Kruemmel KKK, Neckarwestheim GKN I, Neckarwestheim GKN II, Philippsburg KKP 1, Philippsburg KKP 2, and Unterweser KKU.

  18. 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.

  19. 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.

  20. 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)

  1. Safety requirements for nuclear power plants. Content, legal validity and execution; Sicherheitsanforderungen an Kernkraftwerke. Inhalt, rechtlicher Geltungsanspruch und Vollzug

    Energy Technology Data Exchange (ETDEWEB)

    Mueller-Dehn, Christian [E.ON Kernkraft GmbH, Hannover (Germany). Nuclear Regulation and Policy

    2014-05-15

    With the approval of the 'Safety Requirements for Nuclear Power Plants' in November 2012 and the key fitting 'Interpretations' in November 2013, the decade-lasting trial on the development and actualisation of the nuclear technical regulations were successfully concluded. In terms of content the safety requirements stipulate the requirements for damage precaution as well as further safety optimisations according to paragraph 7d Atomic Energy Act (AtG). Even thought they are no international law, they tie all responsible authorities related to atomic law in the framework of existing regulations in the borders of the respective approval parameters and existing laws. In its regulations priority is given at the supervisory process to the existing approval situation and the application of safety requirements in the approval process are only acknowledged in the scope of technical modifications. (orig.)

  2. Identification of the vibrational behaviour and system parameter estimation of pressure and core vessel of the Kernkraftwerk Biblis-A

    International Nuclear Information System (INIS)

    Previous investigations of the reactor vibrations have demonstrated that the pressure vessel and its internals perform pendular motions. In this report the identification of the vibrative behaviour by use of modern estimation methods is described. Based on a double-pendulum lumped parameter model, the state vector enlarged by unknown system parameters is estimated from noisy pre-operational measurements of the PWR BIBLIS-A. For this task a Maximum-A-Posteriori (MAP) identification filter algorithm was employed. (orig.)

  3. 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.

  4. 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)

  5. 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.

  6. Advanced handbook for accident analyses of German nuclear power plants; Weiterentwicklung eines Handbuches fuer Stoerfallanalysen deutscher Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Kerner, Alexander; Broecker, Annette; Hartung, Juergen; Mayer, Gerhard; Pallas Moner, Guim

    2014-09-15

    The advanced handbook of safety analyses (HSA) comprises a comprehensive electronic collection of knowledge for the compilation and conduction of safety analyses in the area of reactor, plant and containment behaviour as well as results of existing safety analyses (performed by GRS in the past) with characteristic specifications and further background information. In addition, know-how from the analysis software development and validation process is presented and relevant rules and regulations with regard to safety demonstration are provided. The HSA comprehensively covers the topic thermo-hydraulic safety analyses (except natural hazards, man-made hazards and malicious acts) for German pressurized and boiling water reactors for power and non-power operational states. In principle, the structure of the HSA-content represents the analytical approach utilized by safety analyses and applying the knowledge from safety analyses to technical support services. On the basis of a multilevel preparation of information to the topics ''compilation of safety analyses'', ''compilation of data bases'', ''assessment of safety analyses'', ''performed safety analyses'', ''rules and regulation'' and ''ATHLET-validation'' the HSA addresses users with different background, allowing them to enter the HSA at different levels. Moreover, the HSA serves as a reference book, which is designed future-oriented, freely configurable related to the content, completely integrated into the GRS internal portal and prepared to be used by a growing user group.

  7. 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

  8. Fifty years of Erlangen radiochemistry

    International Nuclear Information System (INIS)

    On June 29, 2006, the Radiochemical Laboratory of AREVA NP GmbH (formerly Siemens AG) in Erlangen celebrated its fiftieth anniversary. The occasion was marked by an event attended by more than 1,000 guests, among them Werner Gebauhr, the 85-year-old founder and first head of the Laboratory; the Managing Directors of AREVA NP GmbH, Ralf Gueldner and Ruediger Steuerlein; representatives of universities, research institutions, power utilities, and public authorities. The present head of the Radiochemical Laboratory, Wilfred Morell, sketched the highlights of the work performed over the past fifty years, which ranged from solid-state and very-high-purity materials technologies to development and service activities for nuclear technology. Manfred Erve, head of the Technical Center of AREVA NP GmbH, of which the Radiochemical Laboratory is a part, emphasized the changes in priorities over the past fifty years, which had always been met successfully by Radiochemistry. In the scientific part of the event, Wolfgang Schwarz (E.ON Kernkraftwerk GmbH, KKW Isar), Ulf Ilg (EnBW Kraftwerk AG, KKW Philippsburg), and Hans-Josef Allelein (Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH) explained 3 major subject areas in which Erlangen Radio-chemistry over many years has contributed basic findings (see other articles in this atw issue). On the occasion of the anniversary, a comprehensive booklet was published under the title of '50 Jahre Radiochemie Erlangen - 1956-2006'. (orig.)

  9. 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...

  10. 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.

  11. 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

  12. 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....

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

    International Nuclear Information System (INIS)

    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. 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

  15. Self-sustaining emergency power supply for the nuclear power plant Beznau. Project AUTANOVE; Autarke Notstrom-Versorgung fuer das Kernkraftwerk Beznau (KKB). Projekt AUTANOVE

    Energy Technology Data Exchange (ETDEWEB)

    Kaeser, Roland [Axpro AG - Kernenergie, Beznau (Switzerland). Kernkraftwerk Beznau

    2010-05-15

    The NPP Beznau is sited close to the Aare with sufficient cooling water supply so that no cooling tower is necessary. The author describes the project AUTANOVE, an self-sustaining emergency power supply for the NPP Beznau, including an evaluation of the reliability for the accidental situations fire and internal flooding, external flooding and low-water, air plane crash and safety earthquake. The new system includes two new seismic qualified, physically separated emergency diesel generators, for each unit. Deterministic and probabilistic safety analyses show further increase of the already high safety level.

  16. Experience in using new safety I and C systems at the Beznau nuclear power station; Erfahrungen mit neuer Sicherheitsleittechnik im Kernkraftwerk Beznau

    Energy Technology Data Exchange (ETDEWEB)

    Farruggio, David; Hangartner, Christian; Schaeuble, Thomas [Nordostschweizerische Kraftwerke AG, Doettingen (Switzerland). Kernkraftwerk Beznau

    2009-01-15

    Beznau Nuclear Power Station is made up of 2 nearly identical units built between 1966 and 1972. Unit 1 started commercial operation in December 1969, unit 2, in March 1972. In an effort to always keep the power plant at the latest state of engineered safeguards, backfitting was started early on, also in the field of electrical engineering and I and C. The equipment originally installed for the reactor protection and control system, due to its age, suffered from a lack of support by the vendor and from bottlenecks in spare parts supplies. Consequently, there had to be a change. Planning initiated replacement in 1994. In a first phase, the concepts almost exclusively based on digital control systems were examined. Two of these concepts were worked out in detail in another phase, finally resulting in the decision to implement backfitting of the reactor protection and control system with TELEPERM XS. The reactor protection and control system was replaced in 2000 and 2001. The experience since accumulated has been mainly positive. The hardware is stable in operation, with hardly any failures. The robust architecture prevents the few failures from impacting plant operation. The software has been implemented in such a way that technical process functions are carried out according to design, both in normal operation and during transients. (orig.)

  17. 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.

  18. 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.)

  19. 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)

  20. Notification of the interpretation concerning the safety requirements for nuclear power plants from November 22, 2012; Bekanntmachung der Interpretation zu den Sicherheitsanforderungen an Kernkraftwerke vom 22. November 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-03-03

    The safety requirements for nuclear power plants from November 22, 2012 include uncertainties concerning the scope of interpretation that could trigger difficulties for design and application. The notification on the interpretation of the safety requirements includes changes and corrections.

  1. 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.

  2. ENSI's technical view on the periodic safety review 2008 of the nuclear power plant Goesgen; Sicherheitstechnische Stellungnahme zur Periodischen Sicherheitsueberpruefung 2008 des Kernkraftwerks Goesgen

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

    The owner of a license for a nuclear power plant operation in Switzerland has to undergo every 10 years a comprehensive safety check called 'periodic safety review' (PSR). The regulatory authority, the Swiss Federal Nuclear Safety Inspectorate (ENSI), reviews the documents supplied by the licensee. The Goesgen power plant (KKG) obtained its operation license and started operation in 1978. A first PSR was performed in the years 1996 to 1998 (PSR 1998) and reported. KKG delivered an analysis of the safety status, an evaluation of subsystems as well as test reports. The new PSR covers the period 1998 to 2007. The basis of the evaluation by ENSI is the new nuclear energy law in force since 1 February 2005. In comparison to PSR 1998, new aspects have to be considered like the description of the safety concept, including the technical safety classification of buildings, systems and components, or consideration of the protection objective 'limitation of the radiation exposure'. The PSR 2008 is focussed on the estimate of the nuclear safety of KKG. Basically, for the operation of a nuclear power plant, a sufficient protection has to be guaranteed against the release of radioactive materials to the environment as well as the irradiation of persons, during normal operation as well as in the case of accidents. The licensee of a nuclear power plant in operation must retrofit his plant according to the experience already gained and the state-of-the-art. The purpose of the PSR is to check the quality of the plant in the domain of safety. A probabilistic safety assessment (PSA) study must prove that the probability of damages to the reactor core is smaller than 10{sup -5}/year. In Switzerland the life time of a nuclear power plant is not limited by a fixed maximum time of operation. On the contrary, the limitation proceeds from safety criteria. Insufficiencies in the plant design are often recognized only through the evolution of the technique or some unexpected events. Ageing damages like radiation induced brittleness, fatigue, erosion and corrosion appear only after some time of operation; such damages are normally foreseen at the beginning of operation and control programs are implemented. In this report, the modifications against the previous PSR concerning the site parameters and the safety concept of KKG are listed and analysed as well as the measures taken to improve them. Further, in addition to the evaluation of the changes in the organisation and in the personnel, ENSI evaluates also the KKG management and its safety culture. The safety relevant components of the plant are evaluated from the point of view of their status and from the experience gained from former incidents. By means of deterministic analyses of incidents, the behaviour of the plant and the fulfilment of the legal protection goals are checked. For the check of the protection against accidents unforeseen in the plant design concept, the safety level of the plant is checked by means of a PSA method that calculates the frequency of damage to the core and the radioactive material release. Another chapter is devoted to the analysis of the plant internal organisation and administrative measures in case of emergency.

  3. 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.

  4. 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.

  5. Public announcement of the planned erection and operation of a nuclear power plant on the village lands of Vahnum near Wesel (Kernkraftwerk Vahnum)

    International Nuclear Information System (INIS)

    The RWE intends to erect and to operate in Vahnum near Wesel a nuclear power plant with two PWRs and two natural-draught cooling towers. The thermal capacity of each block is to be 3,765 MWth, the electrical net capacity 1,232 MWe. (HP)

  6. Experiences concerning the preparation of seismic probabilistic safety analyses for German nuclear power plants; Erfahrung aus der Erstellung von seismischen probabilistischen Analysen fuer deutsche Kernkraftwerke

    Energy Technology Data Exchange (ETDEWEB)

    Bloem, Theodor; Obenland, Ralf; Ulrich, Holger [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2010-05-15

    For several German nuclear power plants the Westinghouse Electric Germany GmbH has performed probabilistic safety analyses. The seismic resistance of safety relevant plant components has to be assessed considering all possible damage scenarios for earthquake levels below and above the design earthquake; the incidence probability has to be evaluated. Components and buildings with earthquake resistant design in a condition comparable to the initial state can withstand earthquakes stronger than the design earthquake. In case of aged components or components without periodic inspection unknown damages could occur. The seismic probabilistic safety assessments have been performed for nuclear power plants with increased seismic hazard. The calculated core damage probability was 10{sup -7} per year.

  7. 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.)

  8. Zirconium and technetium recovery and partitioning in the presence of actinides in modified Purex process for ATW program. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Dzekun, E.G.; Fedorov, Y.S.; Galkin, B.Y.; Lyubtsev, R.I.; Mashkin, A.N.; Mishin, E.N.; Zilberman, B.Y. [Radievyj Inst., Leningrad (Russian Federation)

    1994-12-31

    The modified Purex process flowsheet is based on combination of all irradiated materials, their joint dissolution and reprocessing as a NPP spent fuel solution with abnormal Pu content after addition of recycled depleted U concentrate. Some groups of long-lived radionuclides could be completely recovered and localized at the stage of extraction reprocessing using 30% TBP. Studies were conducted for 10 y to develop the process for recovery, concentration, and localization of U, Pu, Np, Tc, and Zr within 1st extraction cycle. Actinides are recovered from high-level raffinate of this cycle after evaporation and feed adjustment. Results in this report show that combined deep recovery of several elements from highly irradiated materials by TBP extraction, for further transmutation, is possible. Selective stripping of Zr from solvent phase containing U, Pu, Np, and Tc is quite effective. Development of the modified Purex process is not complete; main problem to be solved should be oxide separation from the loop and permissible storage duration before reprocessing and reuse in the loop.

  9. 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

  10. 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).

  11. 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)

  12. 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.

  13. 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).

  14. ENSI's view on technical safety for the long term operation of reactors 1 and 2 in the Beznau nuclear power plant; Sicherheitstechnische Stellungnahme zum Langzeitbetrieb des Kernkraftwerks Beznau Block 1 und Block 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-11-15

    The reactors 1 and 2 of the Beznau nuclear power plant (KKB) are operated since about 40 years. For an operation beyond the design period of 40 years the Swiss Federal Nuclear Safety Inspectorate (ENSI) demands the evidence to be brought that the design limits of the safety relevant components will not be reached during the extended operation period. In 2008 the license holder of KKB delivered the requested documentation on material ageing on the basis of deterministic as well as probabilistic safety analyses and concluded that both reactors can be safely operated beyond 40 years. Thanks to continuous additional outfits, both reactors are in good condition from the point of view of technical safety. With a view to the extension of operation beyond 40 years, KKB already applied the necessary measures regarding technics, finances and personnel in order to keep the present technical level. Since 1991 KKB has analysed and checked components that are difficult to replace. From the evidence presented, ENSI concluded that both reactors are able to be operated up to 60 years long, however with two restrictions for reactor 1 because there the material used for the reactor pressure vessel (RPV) suffered more neutron brittleness than in reactor 2. In addition, reactor 1 is much more affected by ageing phenomena than reactor 2, but, according to neutron fluence calculations, the limiting criteria will not be reached even after 60 years of operation. Some corrosion damages were noted at the lower part of the RPV due to water containing boron acid; they are more pronounced in reactor 1 than in reactor 2. Even though the calculations done by KKB are very conservative, they show that also in the long term the operation limiting criteria about the mechanical resistance of the RPV are never reached. ENSI concludes that the safety design of both KKB reactors ensures safe control of the design basis accidents. Both reactors were continuously fitted with new equipment. With the planed replacement of current emergency electric supply equipment by Diesel engines for both reactors the safety of the plant will still be increased. However, it seems that the earthquake risk in Switzerland was underestimated in the past; it must be proven that the dose limits can also be respected under consideration of the new assumptions on earthquake risk.

  15. Manufacture and assembly of a crane for manipulation of transport and storage casks in the nuclear power plant Beznau (Switzerland); Herstellung und Montage einer Krananlage zur Handhabung von Transport- und Lagerbehaeltern im Kernkraftwerk Beznau (Schweiz)

    Energy Technology Data Exchange (ETDEWEB)

    Koselowski, Eiko [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

    2008-07-01

    In the frame of increased capacity of the interim storage facility for spent fuel casks in the NPP Beznau (Switzerland) an overhead crane system was built. The contributions describes design, construction, manufacture, assembly, commissioning and start-up of the system.

  16. Target: The green meadow. How much knowledge is needed for the dismantling of nuclear power plants?; Ziel: die Gruene Wiese. Wieviel Know-how man braucht, um ein Kernkraftwerk zurueckzubauen

    Energy Technology Data Exchange (ETDEWEB)

    Bach, Friedrich-Wilhelm; Hassel, Thomas [Unterwassertechnikum Hannover (UWTH), Hannover (Germany). Inst. fuer Werkstoffkunde

    2013-07-01

    As from the year 2022, there will no nuclear power plant exist in Germany. In the contribution under consideration two scientists from the Institute of Materials Science (Hanover, Federal Republic of Germany) report on the preparations and the necessary technical knowledge in order to dismantle the highly complex nuclear facilities and to recultivate former nuclear power plant sites.

  17. 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.)

  18. 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.

  19. Study on the possible consequences of a severe accident in a Swiss nuclear power plant on the drinking water supply; Untersuchung moeglicher Folgen eines schweren Unfalls in einem schweizerischen Kernkraftwerk auf die Trinkwasserversorgung

    Energy Technology Data Exchange (ETDEWEB)

    Ustohalova, Veronika; Kueppers, Christian; Claus, Manuel

    2014-06-18

    The study on the possible consequences of a severe accident in a Swiss nuclear power plant on the drinking water supply covers the following issues: estimation of possible source terms and radioactive materials release rates, airborne water contamination, water contamination by direct pollution, consequences for the drinking water supply, emergency measures in case of a drinking water contamination, routine surveillance of surface and ground water and improvement possibilities in nuclear power plants.

  20. Plant specific safety inspection of German nuclear power plants taking into account the Fukushima-I (Japan) events; Anlagenspezifische Sicherheitsueberpruefung deutscher Kernkraftwerke unter Beruecksichtigung der Ereignisse in Fukushima-I (Japan). RSK-Anforderungskatalog-Vorspann

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-03-30

    The German Parliament requested (17 March 2011) a comprehensive inspection of German nuclear power plants. For this purpose independent expert commissions should perform a new risk analysis of all German NPPS and nuclear installations with respect to the lessons learned from the Fukushima (Japan) events and other extraordinary damage scenarios. The Reactor safety commission (RSK) was assigned by the German Bundesamt fuer Strahlenschutz to develop a catalogue of requirements for this safety inspection. The contribution summarizes the required inspection volume (status 30.03.2011) including the following events: natural events like earth quakes, floods, weather-based consequences and possible superposition. Additionally the following assumptions have to be considered: event independent postulated common failures or systematic faults, station blackout larger than 2 hours, long-term failure of the auxiliary cooling water supply; aggravating boundary conditions for the performance of emergency measures (non-availability of power supply), hydrogen generation and detonation hazard, restricted personnel availability, non-accessibility due to high radiation levels, impeded technical support from outside. (orig.)

  1. Catalogue of requirements for a plant-specific safety inspection of German nuclear power plants taking into account the Fukushima-I (Japan) events; Anforderungskatalog fuer anlagenbezogene Ueberpruefungen deutscher Kernkraftwerke unter Beruecksichtigung der Ereignisse in Fukushima-I (Japan)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-03-30

    The catalogue of requirements for a plant-specific safety inspection of German nuclear power plants taking into account the Fukushima-I (Japan) events worked out by the German RSK (reactor safety commission) includes the following inspection topics: natural events like earth quakes, floods, weather-based consequences and possible superposition; civilization-based events like airplane crash, gas release, reactor accident consequences for neighboring units, terroristic impacts, external attacks on computer-based control systems. Further event-independent assumptions have to be considered: station blackout, long-term emergency power supply requirement, failure of auxiliary cooling water supply, efficacy of preventive measures, aggravating boundary conditions for the performance of emergency measures.

  2. Expertise about the request of the nuclear power plant Leibstadt for increasing the power to 3600 MW{sub th}; Gutachten zum Gesuch des Kernkraftwerks Leibstadt um Leistungserhoehung auf 3600 MW{sub th}

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-15

    On July 31, 1992, the nuclear power plant Leibstadt AG (KKL) submitted a request for a modification of its operation license for increasing the reactor thermal power to 3600 MW. In its examination, the Federal Agency for the Safety of Nuclear Installations (HSK) investigated the effects of the power increase on reactor safety, especially on the safety criteria which can limit the power. In doing this, a distinction has to be made between normal operation, design incidents and out-of-design accidents. KKL is a boiling water reactor (BWR) with a current maximum thermal power of 3,138 MW with 648 fuel assemblies in the core. Since the start of operation in February 1984, the reactor has been continuously improved and now almost fulfils the present state-of-the-art of science and technology for BWRs. After some incidents during the early years, the plant shows a high level of availability. During the past 6 years some fuel assembly damage has cast a shadow on good operational experience, but until now the collective irradiation dose of the plant staff and the environment has remained mostly below legal limits, as well as for the release of radioactive materials to the atmosphere and to the Rhine River. Calculations of core loading with the fuel assemblies presently used at KKL have shown that the operation and safety limits of the reactor core can still be preserved with a thermal power of 3600 MW. For normal operation, no objection can be raised against the power increase. This increase, however, has to be carried out step-by-step in order to gain experience concerning plant behaviour. With the higher power rating, increased dose rates are expected on systems and components, in plant rooms and in the plant area, which also leads to increased dose rates to the staff and environment. This increase has to be estimated and, possibly, correction measures will have to be taken in order to reduce them. Especially to be monitored is the dose rate increase in the machine hall. In the case of design incidents too, all safety-relevant limits and the maximal tolerable dose rates in the environment must be respected. In the context of design incidents, the 'transitory' group also constitutes the limiting case with the higher power. For the complete judgement of the safety of a nuclear power plant it is not sufficient to estimate the effects of a design incident through deterministic methods. The evaluation of the effects of out-of-design accidents needs a probabilistic safety analysis which determines the frequency as well as the consequences of an accident. The results show that KKL represents a very small risk for the environment. In KKL the measures necessary for safe operation and protection of mankind and environment at a thermal power of 3600 MW have already been taken or will be taken shortly. According to its examination, HSK concludes that there are no safety-relevant reasons speaking against an operational license for the increased thermal power. The increase will, however, have to be carried out in 4 steps of 1 year each in order to gain operational experience

  3. Analysis of the EU stress test results for the NPP Fessenheim and Beznau. Pt. 2. Beznau; Analyse der Ergebnisse des EU-Stresstest der Kernkraftwerke Fessenheim und Beznau. T. 2. Beznau

    Energy Technology Data Exchange (ETDEWEB)

    Brettner, Mathias [Physikerbuero Bremen (Germany); Pistner, Christoph; Kurth, Stephan [Oeko-Institut e.V. - Institut fuer Angewandte Oekologie, Darmstadt (Germany)

    2012-10-11

    As a consequence of the reactor accidents in Fukushima Daiichi the safety status of nuclear power plants was performed by national and international surveillance processes. In Germany the safety status of the NPP was testes by the reactor safety commission, an expert commission of Baden-Wuerttemberg, an expert commission of Bavaria and the EU stress test. The evaluation criteria based on national and international surveillance processes were focused on earthquakes, flooding and postulated failures of the electricity supply - station blackout and long-lasting failure of the emergency power supply. In Germany extended requirements included the electricity supply and the emergency cooling water supply. The authors identify essential safety relevant systems in the NPP Beznau, including technical systems, energy supply, plant-internal emergency measures, and discuss specific Swiss requirements. The evaluation of the EU stress test for the NPP Fessenheim covers the issues earthquake, flooding, spent fuel element pool, electricity supply, cooling water supply and identification of further safety relevant deficiencies.

  4. Analysis of the EU stress test results for the NPP Fessenheim and Beznau. Pt. 1. Fessenheim; Analyse der Ergebnisse des EU-Stresstest der Kernkraftwerke Fessenheim und Beznau. T. 1. Fessenheim

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Kueppers, Christian; Kurth, Stephan; Mohr, Simone [Oeko-Institut e.V. - Institut fuer Angewandte Oekologie, Darmstadt (Germany); Brettner, Mathias [Physikerbuero Bremen (Germany)

    2012-10-11

    As a consequence of the reactor accidents in Fukushima Daiichi the safety status of nuclear power plants was performed by national and international surveillance processes. In Germany the safety status of the NPP was testes by the reactor safety commission, an expert commission of Baden-Wuerttemberg, an expert commission of Bavaria and the EU stress test. The evaluation criteria based on national and international surveillance processes were focused on earthquakes, flooding and postulated failures of the electricity supply - station blackout and long-lasting failure of the emergency power supply. In Germany extended requirements included the electricity supply and the emergency cooling water supply. The authors identify essential safety relevant systems in the NPP Fessenheim. The evaluation of the EU stress test for the NPP Fessenheim covers the issues earthquake, flooding, spent fuel element pool, electricity supply, cooling water supply and identification of further safety relevant deficiencies.

  5. 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)

  6. 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.)

  7. 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.

  8. 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

    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...

  9. 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; Suo, Hongli

    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...... ordered crystal structure along the film thickness observed by a transmission electron microscope. On the basis of the enhanced understanding of the crystallization processes, we demonstrate a possibility of engineering the surface morphology and texture in the film deposited on technical substrates using...... backscatter diffraction. A strong dependence of the morphology and texture on the film thickness is observed, mainly due to (i) the transition of growth mode associated with the critical film thickness, i.e., increasing the film thickness leads to the grain morphology changing from 2-dimensional discs (highly...

  10. 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....... and morphology are investigated in details. It is found that a rotated cube-on-cube epitaxy of Gd2Zr2O7//NiW in-plane texture forms as soon as the (004) out-plane texture appears, implying that epitaxial growth dominates the crystallization processes. Thermal energy plays an important role in minimizing...... and body according to surface or cross-sectional observation and Rutherford Backscattering Spectrometry analysis, pointing to inhomogeneous structure through film thickness, i.e., dense in the surface layer but porous in the body. This is attributed to trapped gas generated during either decomposition...

  11. Document management. Interleaf administrates ''handbooks'' at the Gundremmingen nuclear power plant

    International Nuclear Information System (INIS)

    The Gundremmingen nuclear power plant with its two 1344 MW units (B and C) is the largest boiling water reactor site in Germany. It is operated by Kernkraftwerke Gundremmingen Betriebsgesellschaft (KGB), a daughter of RWE Energie, Essen, and BAYERNWERK, Munich. For two years, KGB has been using Interleaf software to prepare and distribute its handbook. (orig.)

  12. 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.)

  13. The development of engineered safeguards for nuclear power plants in the political and technical environment in the Federal Republic of Germany since 1955; Die Entwicklung der Sicherheitstechnik fuer Kernkraftwerke im politischen und technischen Umfeld der Bundesrepublik Deutschland seit dem Jahr 1955

    Energy Technology Data Exchange (ETDEWEB)

    Laufs, P. [Stuttgart Univ. (Germany). Philosophische-Historische Fakultaet

    2007-01-15

    The safety of nuclear power plants is determined largely by the integrity of the internally pressurized coolant containment system. The highly radioactive materials (fission products) generated within this pressurized containment (primary system) in the reactor core during nuclear power plant operation constitute an extremely great potential hazard. Catastrophic failure of the primary system, and the release into the environment of the radioactive inventory, must be avoided at all costs. Because of the high coolant pressure and the high power density, pressurized water reactors (PWR) impose particularly strict requirements with respect to reactor safety. German nuclear power plants equipped with light water reactors enjoy the reputation of being among the safest plants in the world. This frequent statement is justified in the light of the research and development work performed jointly by industry, government agencies, science, and expert bodies between the 1960s and the 1990s. The research projects, which implied considerable financial expenditures, their internationally acknowledged results, and the resultant additional backfitting measures conducted in German nuclear power plants at many billions of expenditures, were hardly noticed by the German body politic. (orig.)

  14. 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.

  15. 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)

  16. 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.

  17. 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.

  18. 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.

  19. Anticipated transients without scram for WWER reactors. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    Anticipated transients without scram (ATWS) are anticipated operational occurrences followed by the failure of one reactor scram function. Current international practice requires that the capability of pressurized water reactors (PWRs) to cope with ATWS be demonstrated following a systematic evaluation of plants' defence in depth. Countries operating PWRs require design consideration of ATWS events on a deterministic basis. The regulatory requirements may concern either specific mitigating systems or acceptable plant performance during these events. The prevailing international practice for performing transient analysis of ATWS for licensing is the best estimate approach. Available transient analyses of ATWS events indicate that WWER reactors, like PWRs, have the tendency to shut themselves down if the inherent nuclear feedback is sufficiently negative. Various control and limitation functions of the WWER plants also provide a degree of defence against ATWS. However, for most WWER plants, complete and systematic ATWS analyses have yet to be submitted for rigorous review by the regulatory authorities and preventive or mitigative measures have not been established. In addition, it has also been recognized that plant behaviour in case of ATWS also relies on certain system functions (use of pressurizer safety valves for liquid discharge, availability of steam dump valve to both the condenser (BRU-K) and the atmosphere (BRU-A) for secondary side pressure control, and others) which have been identified as safety issues and need to be qualified for accident conditions. In all countries operating WWERs, the need for ATWS investigations is recognized and reflected in the safety improvement programmes. ATWS analysis for WWERs is not required for the licensing process in Bulgaria, the Czech Republic (with the exception of the Temelin nuclear power plant) and Russia. Design consideration of ATWS is required if expert assessments of probabilistic safety assessment (PSA) results

  20. MIDAS/PK code development using point kinetics model

    International Nuclear Information System (INIS)

    In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

  1. MIDAS/PK code development using point kinetics model

    Energy Technology Data Exchange (ETDEWEB)

    Song, Y. M.; Park, S. H. [KAERI, Taejon (Korea, Republic of)

    1999-05-01

    In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation.

  2. EP1000 anticipated transient without scram analyses

    International Nuclear Information System (INIS)

    The present paper summarizes the main results of the Anticipated Transient Without Scram (ATWS) analysis activity, performed for the European Passive Plant Program (EPP). The behavior of the EP1000 plant following an ATWS has been analyzed by means of the RELAP5/Mod3.2 code. An ATWS is defined as an Anticipated Transient accompanied by a common mode failure in the reactor protection system, such that the control rods do not scram as required to mitigate the consequences of the transient. According to the experience gained in PWR design, the limiting ATWS events, in a PWR, have been found to be the heatup transients caused by a reduction of heat removal capability by the secondary side of the plant. For this reason, the Loss of Normal Feedwater initiating event, to which the failure of the reactor scram is associated, has been analyzed. The purpose of the study is to verify the performance requirements set for the core feedback characteristics (that is to evaluate the effect of the low boron core neutron kinetic parameters), the overpressure protection system, and boration systems to cope with the EUR Acceptance Criteria for ATWS. Another purpose of this analysis was to support development of revised PSA success criteria that would reduce the contribution of ATWS to the large release frequency (LRF). The low boron core improved the basic EP1000 response to an ATWS event. In particular, the peak pressure was significantly lower than that which would result from a standard core configuration. The improved ATWS analysis results also permitted improved ATWS PSA success criteria. For example, the reduced peak pressure allows the use of other plant features to mitigate the event, including manual initiation of feed-bleed cooling in the event of PRHR HX failure. As a result, the core melt frequency and especially the LRF are significantly reduced. (author)

  3. Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

    International Nuclear Information System (INIS)

    An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented

  4. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Schriber, S.O.

    1997-12-31

    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.

  6. 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...

  7. 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.

  8. 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.)

  9. 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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  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 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.

  16. 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.

  17. 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.

  18. 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)

  19. 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...

  20. 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

  1. Planning the research and development necessary for accelerator transmutation of waste, leading to integrated proof of performance testing

    International Nuclear Information System (INIS)

    The Research and Development (R and D) Plan for the Accelerator Transmutation of Waste (ATW) Program has been developed for the Department of Energy, Office of Nuclear Energy (DOE/NE) to serve as a focus and progressional guide in developing critical transmutation technologies. It is intended that the Plan will serve as a logical reference considering all elements of an integrated accelerator-driven transmutation system, and will maximize the use of resources by identifying and prioritizing research, design, development and trade activities. The R and D Plan provides a structured framework for identifying and prioritizing activities leading to technically-justifiable integrated Proof of Performance testing within ten years and ultimate demonstration of Accelerator Transmutation of Waste (ATW). The Plan builds from the decision objectives specified for ATW, utilizes informational input from the ATW Roadmap and programmatic System Point Design efforts, and employs the knowledge and expertise provided by professionals familiar with ATW technologies. With the firm intent of understanding what, why and when information is needed, including critical interfaces, the Plan then develops a progressional strategy for developing ATW technologies with the use of a Technology Readiness Level (TRL) scale. The TRL approach is first used to develop a comprehensive, yet generic, listing of experimental, analytical and trade study activities critical to developing ATW technologies. Technology-specific and concept-specific aspects are then laid over the generic mapping to gage readiness levels. Prioritization criteria for reducing technical uncertainty, providing information to decision points, and levering off of international collaborations are then applied to focus analytical, experimental and trade activities. (author)

  2. Comparison of accelerator-based with reactor-based nuclear waste transmutation schemes

    International Nuclear Information System (INIS)

    An overview of the most significant studies in the last 35 years of partitioning and transmutation of commercial light water reactor spent fuel is given. Recent Accelerator-based Transmutation of Waste (ATW) systems are compared with liquid-fuel thermal reactor systems that accomplish the same objectives. If no long-lived fission products (e.g. 99Tc and 129I) are to be burned, under ideal circumstances the neutron balance in an ATW systems becomes identical to that for a thermal reactor system. However, such a reactor would need extraordinarily rapid removal of internally-generated fission products to remain critical at equilibrium without enriched feed. The accelerator beam thus has two main purposes (1) the burning of long-lived fission products that could not be burned in a comparable reactor's margin (2) a relaxing of on-line chemical processing requirements without which a reactor-based system cannot maintain criticality. Fast systems would require a parallel, thermal ATW system for long-lived fission product transmutation. The actinide-burning part of a thermal ATW system is compared with the Advanced Liquid Metal Reactor (ALMR) using the well-known Pigford-Choi model. It is shown that the ATW produces superior inventory reduction factors for any near-term time scale. (author)

  3. The U.S. accelerator transmutation of waste program

    International Nuclear Information System (INIS)

    A national project to develop a future capability to separate actinides and long-lived fission products from spent fuel, to transmute them, and to dispose off the remaining waste in optimal waste forms has begun in the United States. This project is based on the Accelerator-driven Transmutation of Waste (ATW) program developed during the 1990s at Los Alamos National Laboratory, and has its technological roots in several technologies that have been developed by the multi-mission laboratories of the U.S. Department of Energy (DOE). In the Fiscal Year 1999 Energy and Water Appropriation Act, the U.S. Congress directed the DOE to study ATW and by the end of FY99 to prepare a 'roadmap' for developing this technology. DOE convened a steering committee, assembled four technical working groups consisting of members from many national laboratories, and consulted with several individual international and national experts. The finished product, 'A Roadmap for Developing ATW Technology - A Report to Congress', recommends a five-year, $281 M, science-based, technical-risk-reduction program. This paper provides an overview of the U.S. Roadmap for developing ATW technology, the organization of the national ATW Project, the critical issues in subsystems and technological options, deployment scenarios, institutional challenges, and academic and international collaboration

  4. 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.

  5. 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.

  6. Disposition of Nuclear Waste Using Subcritical Accelerator-Driven Systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-27

    ATW destroys virtually all the plutonium and higher actinides without reprocessing the spent fuel in a way that could lead to weapons material diversion. An ATW facility consists of three major elements: (1) a high-power proton linear accelerator; (2) a pyrochemical spent fuel treatment i waste cleanup system; (3) a liquid lead-bismuth cooled burner that produces and utilizes an intense source-driven neutron flux for transmutation in a heterogeneous (solid fuel) core. The concept is the result of many years of development at LANL as well as other major international research centers. Once demonstrated and developed, ATW could be an essential part of a global non-proliferation strategy for countries that could build up large quantities of plutonium from their commercial reactor waste. ATW technology, initially proposed in the US, has received wide and rapidly increasing attention abroad, especially in Europe and the Far East with major programs now being planned, organized and tided. Substantial convergence presently exists on the technology choices among the programs, opening the possibility of a strong and effective international collaboration on the phased development of the ATW technology.

  7. Automatic reactor power control device

    International Nuclear Information System (INIS)

    Anticipated transient without scram (ATWS) of a BWR type reactor is judged to generate a signal based on a reactor power signal and a scram actuation demand signal. The ATWS signal and a predetermined water level signal to be generated upon occurrence of ATWS are inputted, and an injection water flow rate signal exhibiting injection water flow rate optimum to reactor flooding and power suppression is outputted. In addition, a reactor pressure setting signal is outputted based on injection performance of a high pressure water injection system or a lower pressure water injection system upon occurrence of ATWS. Further, the reactor pressure setting signal is inputted to calculate opening/closing setting pressure of a main steam relief valve and output an opening setting pressure signal and a closure setting pressure signal for the main steam relief valve. As a result, the reactor power and the reactor water level can be automatically controlled even upon occurrence of ATWS due to failure of insertion of all of the control rods thereby enabling to maintain integrity and safety of the reactor, the reactor pressure vessel and the reactor container. (N.H.)

  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. 10 years of chemistry at Beznau power station

    International Nuclear Information System (INIS)

    Kernkraftwerk Beznau, with a PWR-type reactor, consists of two separate, identical units each of which has a power of 350 MW. The servicing system, e.g. chemistry and radiation protection, is the same for both units. In spite of high availability, there have been great chemical problems in the secondary loop. The Beznau power plant, unfortunately, had to play an important role in the development of chemical conditioning of the secondary loop, in particular steam generator chemistry. For this reason, the problems are presented in a historical review. (GL)

  10. 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.)

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. Planning and reporting of Russian transmutation research projects within ISTC. Phase 1

    International Nuclear Information System (INIS)

    The International Scientific and Technical Center (ISTC) in Moscow funds research of civil interest to counteract the risk of nuclear weapon proliferation. Recently, new technical concepts, 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 Russian experts are knowledgeable and well equipped for doing research in the different technical fields of relevance for the transmutation concepts. Thus, a number of ISTC projects 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. The present report describes the back ground, the status and near term activities of a few ISTC projects of relevance for the ATW concept, which are planned with the participation of a Swedish reference group. 4 refs

  16. Anticipated transient without scram analysis of the simplified boiling water reactor following main steam isolation valve closure with boron injection

    International Nuclear Information System (INIS)

    The simplified boiling water reactor (SBWR) operating in natural circulation is designed with many passive safety features. An anticipated transient without scram (ATWS) initiated by inadvertent closure of the main steam isolation valve (MSIV) in an SBWR has been analyzed using the RAMONA-4B code of Brookhaven National Laboratory. This analysis demonstrates the predicted performance of the SBWR during an MSIV closure ATWS, followed by shutdown of the reactor through injection of boron into the reactor core from the standby liquid control system

  17. Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

    International Nuclear Information System (INIS)

    The uncertainty in predicting the effectiveness of inherent shutdown (ISD) in innovative designs results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require ISD; (2) the approximation of representing all such ATWS events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. In this summary, the methodology and associated results of work used to establish probabilities of failure of inherent shutdown of innovative LMRs to the unprotected loss-of-flow (LOF) accident are discussed

  18. Nuclear power plants 1994 - a world survey

    International Nuclear Information System (INIS)

    The World Survey 1994 of nuclear power plants in operation presents seven tables specially compiled by atw, showing the progress in 1994 and the year-end status. The tables list the data compiled under the following criteria: (1) Gross nuclear electricity produced from 1970-1994. (2) Gross electricity produced in 1994 by the NPP selected for atw survey. (3) NPP that commenced operation in 1994. (4) NPP of the world, aranged by country. (5). Reactor vendors by country. (6) NPP units by reactor type. (7) Ranking of NPP according to gross electricity produced. (orig.)

  19. 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.

  20. 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.

  1. 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.

  2. D1+ Simulator: A cost and risk optimized approach to nuclear power plant simulator modernization

    International Nuclear Information System (INIS)

    D1-Simulator is operated by Kraftwerks-Simulator-Gesellschaft (KSG) and Gesellschaft f?r Simulatorschulung (GfS) at the Simulator Centre in Essen since 1977. The full-scope control room training simulator, used for Kernkraftwerk Biblis (KWB) is based on a PDP-11 hardware platform and is mainly programmed in ASSEMBLER language. The Simulator has reached a continuous high availability of operation throughout the years due to specialized hardware and software support from KSG maintenance team. Nevertheless, D1-Simulator largely reveals limitations with respect to computer capacity and spares and suffers progressively from the non-availability of hardware replacement materials. In order to ensure long term maintainability within the framework of the consensus on nuclear energy, a 2-years refurbishing program has been launched by KWB focusing on quality and budgetary aspects. The so-called D1+ Simulator project is based on the re-use of validated data from existing simulators. Allowing for flexible project management methods, the project outlines a cost and risk optimized approach to Nuclear Power Plant (NPP) Simulator modernization. D1+ Simulator is being built by KSG/GfS in close collaboration with KWB and the simulator vendor THALES by re-using a modern hardware and software development environment from D56-Simulator, used by Kernkraftwerk Obrigheim (KWO) before its decommissioning in 2005. The Simulator project, launched in 2004, is expected to be completed by end of 2006. (author)

  3. 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.)

  4. District heat from large-scale power plants: Analysis of economic efficiency exemplified by the Nuclear Power Station Philippsburg-2 (KKW Philippsburg-2)

    International Nuclear Information System (INIS)

    Meanwhile there are some examples of the uncoupling of district heat from large-scale nuclear power stations primarily built for electricity generation. Thus the 350-MW Nuclear Power Stations Beznau I and II (Kernkraftwerke Beznau I und II der Nordostschweizerischen Kraftwerke AG) will uncouple, when completed, 70 MW for the supply of eight communities. In the Federal Republic of Germany the Nuclear Power Station Stade (Kernkraftwerk Stade) has been supplying steam for the production and the heating of operational buildings of a salt works since the beginning of 1984. 40 MW thermal output are at their disposal for that end. The essay under discussion investigates the question of how the economic efficiency of the uncoupling of district heat from the secondary cycle of the Nuclear Power Station Philippsburg-2, which is shortly before its completion, is to be evaluated from the point of view of a district heat supply enterprise. As district heat consumers private and public buildings as well as industrial works in the municipality of Philippsburg are in question. (orig.)

  5. 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.

  6. Use of scaled BWR lower plenum boron mixing tests to qualify the boron transport model used in TRACG

    International Nuclear Information System (INIS)

    In 2001 GEH applied best estimate methods combined with a statistical methodology to determine upper bound limits for key licensing parameters for anticipated operation occurrence (AOO) transient and anticipated transients without scram (ATWS) overpressure analyses for operating Boiling Water Reactors (BWRs). The methodology was subsequently extended for ESBWR AOO, ATWS, loss of coolant, and stability analyses. GEH is extending the methodology to long-term ATWS analyses for the operating BWRs. A long-term ATWS scenario uses injection of borated water to achieve reactor shutdown. Predicting the mixing and transport of boron is important for calculating the impact on the key licensing parameters. For the many operating BWRs where the denser boron solution is injected into the lower plenum, stratification may occur, delaying boron transport to the core region. CFD modeling can be used to model the stratification and mixing of the boron solution, but such calculations are extremely computer intensive and not cost effective; therefore, a more-empirical approach supported by a theoretical scaling of the dominant phenomena and backed by test data and benchmark calculations is used. The paper presents the TRACG lower plenum boron transport model qualification effort. The scaling basis used to implement the TRACG boron transport model for BWR applications is discussed. (authors)

  7. 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.

  8. 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...

  9. 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.)

  10. 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

  11. Leibstadt nuclear power station. A survey of seven years of operation

    International Nuclear Information System (INIS)

    After some twenty years of planning and construction, Leibstadt was commissioned. In 1965, the application for a site permit was filed for a 600 MW nuclear power plant to be cooled by river water; it was granted in 1969. In 1971, cooling by means of river water was forbidden, and the project was changed to cooling tower operation. In 1973, the site permit was confirmed for a plant of 940 MW power to be equipped with a cooling tower. In the same year, the Bau- und Betriebsgesellschaft Kernkraftwerk Leibstadt AG was founded. Ground excavation as the first step of plant construction was begun in 1974. The boiling water reactor unit was accepted into commercial operation on December 15, 1984. (orig.)

  12. 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

  13. 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

  14. 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.)

  15. 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.)

  16. 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

  17. 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

  18. 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

  19. BWR stability analysis

    International Nuclear Information System (INIS)

    The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

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

    International Nuclear Information System (INIS)

    The proposed ATWS acceptance criteria for LWRs (as specified in NUREG-0460) are in principle, applicable to LMFBRs. For LMFBRs, the major criterion will be the assurance that the plant protection system (shutdown or scram) has a sufficiently high reliability (low failure rate) so that core disruptive accidents (as currently defined) will lie outside the design basis. For early plants, however, mitigating systems may also be required. Alternative accident scenarios for LMFBRs, which are initiated from the shutdown state or may lead to potential core disruptive accidents even following scram, need to be examined in greater detail. The proposed LWR-ATWS criteria do not appear to present any new or unforeseen design and/or safety questions for LMFBRs. They do, however, specify design goals for mitigating systems which may insure conformance with NRC policy. Preliminary recommendations are made for future research and evaluation

  1. 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

  2. New nuclear power plants in Europe 1985. Pt. 2

    International Nuclear Information System (INIS)

    The report is subdivided into sections separately reviewing the various countries and their projects. The atw report on new nuclear power plants in Europe contains both a survey of the Federal Republic of Germany, which was published in the April 1985 issue, and an overview of the nuclear power plant situation in 26 European countries including the Soviet Union and six other CMEA countries. Also this year's review includes specific status reports, complete with technical information, about all nuclear generating units under construction, in the project and concrete planning phases. The nineteen nuclear power plants newly commissioned in Europe since last year's atw report was published are covered in a similar way. Moreover, introductory summaries describe the plants in operation in each country and their 1984 electricity generation. (orig./UA)

  3. 2011 annual meeting on nuclear technology. Topical sessions. Pt. 5

    International Nuclear Information System (INIS)

    Summary report on the Topical Session of the Annual Conference on Nuclear Technology held in Berlin, 17 to 19 May 2011: - Sodium Cooled Fast Reactors. The reports on the Topical Sessions: - CFD-Simulations for Safety Relevant Tasks, - Final Disposal: From Scientific Basis to Application, - Characteristics of a High Reliability Organization (HRO) Considering Experience Gained from Events at Nuclear Power Stations, and - Nuclear Competence in Germany and Europe have been covered in atw 7, 8/9, 10 and 11 (2011). (orig.)

  4. 2011 annual meeting on nuclear technology. Topical sessions. Pt. 5; Jahrestagung Kerntechnik 2011. Fachsitzungsberichte. T. 5

    Energy Technology Data Exchange (ETDEWEB)

    Fazio, Concetta [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany). Nuclear Safety Research Programme

    2011-12-15

    Summary report on the Topical Session of the Annual Conference on Nuclear Technology held in Berlin, 17 to 19 May 2011: - Sodium Cooled Fast Reactors. The reports on the Topical Sessions: - CFD-Simulations for Safety Relevant Tasks, - Final Disposal: From Scientific Basis to Application, - Characteristics of a High Reliability Organization (HRO) Considering Experience Gained from Events at Nuclear Power Stations, and - Nuclear Competence in Germany and Europe have been covered in atw 7, 8/9, 10 and 11 (2011). (orig.)

  5. 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

  6. 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

  7. New nuclear power plants in Europe 1982. Pt. 2

    International Nuclear Information System (INIS)

    European nuclear power plants now under construction or in the projecting stage are listed in tables by countries. The general situation in all countries is briefly outlined at the beginning of the reports; plants already in operation are also considered. This review is the update of the reports published every year in atw. The West German situation has already been discussed in the first part of the review. (UA)

  8. 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)

  9. 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....

  10. Generation of consistent conservative cross section data for the coupled system code ATHLET-QUABOX/CUBBOX

    International Nuclear Information System (INIS)

    The paper gives an overview on first experiences in generation and application of conservative two group assembly homogenized cross section data for the couple system code ATHLET-QUABOX/CUBBOX. A methodology is worked out based on adapting the macroscopic scattering cross sections, which allows to convert the integral reactor void reactivity curve in a predefined manner e.g. from a steep to more flat dependence. The studies will be applied to ATWS analyses. (authors)

  11. Experience from bi-lateral exercise hydra between Switzerland and Federal Republic of Germany

    International Nuclear Information System (INIS)

    In November 1987, a bi-lateral exercise between Switzerland and the Federal Republic of Germany took place. The emergency organizations on both sides were alerted. An anticipated transient without scram (ATWS) at the BWR Leibstadt (Mark III Type) situated directly on the border was assumed. The paper describes the problems which occurred because of the different emergency counter measure concepts in both countries. The most important lessons learned are discussed

  12. Design calculations of the thermal-spectrum accelerator-driven system for LWR waste destruction

    International Nuclear Information System (INIS)

    A number of nuclear physics design issues concerning Accelerator Driven-salt Reactor based on the so called ATW concept proposed by Los Alamos are discussed. General description of concept using internal moderation with graphite block is presented. Burn-up, salt processing and safety criteria (reactivity temperature coefficients and kinetics parameters) are presented for different spectra (graphite to salt ratio) and an optimal variant of the blanket with non-positive temperature reactivity coefficients is provided and results are discussed. (Authors)

  13. New nuclear power plants in Europe 1980. Pt. 2

    International Nuclear Information System (INIS)

    The atw report including Part I published in the April issue which deals with the Federal Republic of Germany, covers the situation in the nuclear power plant sector in 26 European countries, among them seven countries of CMEA. The total number covered of nuclear generating units in operation, under construction, projected or planned in those countries amounts to 338 with an aggregate 245,212 MWe. A striking feature of the statistical survey is the very slight change over the previous year. The number of plants in operation has risen to 131 with 57,962 MWe (in the 1979 atw report: 123 with 49,049 MWe). 108 plants with 93,557 MWe are under construction (last year: 107 with 92,576 MWe), 63 plants with 59,618 MWe are projected (last year: 63 plants with 59,346 MWe). In the course of 1980, a total of 22 nuclear generating units with an aggregate 16,300 MWe are to be newly commissioned in eight European countries, 14 of them with 12,060 MWe in four countries of Western Europe, eight with 4240 MWe in four countries of Eastern Europe. At the time of this writing, five of these plants with 4000 MWe had already been commissioned, four of them with 3400 MWe in countries of Western Europe, one with 600 MWe in a country of Eastern Europe. As always, the atw report contains detailed status reports about the nuclear generating units under construction and projected in all countries and reviews the commissioning of plants newly started up since the time of the last atw-report. This information is supplemented by summary introductions to the situation in each country and by tabulated surveys. (orig.)

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

    International Nuclear Information System (INIS)

    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

  15. Particle accelerator requirements for burning radioactive wastes

    International Nuclear Information System (INIS)

    One of the four subprograms of the Accelerator Driven Transmutation Technology (ADTT) program is the Accelerator Transmutation of Waste (ATW) subprogram which in term is a program. The capacity to burn waste is given by the enhanced neutron economy made possible by the presence of the extra accelerator-produced neutrons. The better neutron economy increases the capacity of burning (transmute) anything that absorbs neutrons including long-lived fission products and actinides. By these transmuted or 'burning' systems, the bulk of long-lived radionuclides in disposed radioactive wastes can be reduced by factors of 1000 and the storage time of residual products can be greatly reduced from hundreds of thousands of years to several centuries. This paper presents particle beam requirements for ADTT applications. These are most often specified with a nominal 1 GeV energy and an average beam current for ATW ranging from 100 to 250 m A depending on different applications and different system concepts. It is sketched the reference RF-Linac Accelerator Design for ATW systems. (author)

  16. Improving CT scan capabilities with a new trauma workflow concept: Simulation of hospital logistics using different CT scanner scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Fung Kon Jin, P.H.P., E-mail: p.fungkonjin@amc.uva.nl [Trauma Unit Department of Surgery, Academic Medical Center, Amsterdam (Netherlands); Dijkgraaf, M.G.W., E-mail: m.dijkgraaf@amc.uva.nl [Department of Clinical Epidemiology, Biostatistics, and Bioinformatics, Academic Medical Center, Amsterdam (Netherlands); Alons, C.L., E-mail: clalons@few.vu.nl [Department of Mathematics, VU University Amsterdam, Amsterdam (Netherlands); Kuijk, C. van, E-mail: c.vankuijk@vumc.nl [Department of Radiology, VU Medical Center, Amsterdam (Netherlands); Beenen, L.F.M., E-mail: l.beenen@amc.uva.nl [Department of Radiology, Academic Medical Center, Amsterdam (Netherlands); Koole, G.M., E-mail: koole@few.vu.nl [Department of Mathematics, VU University Amsterdam, Amsterdam (Netherlands); Goslings, J.C., E-mail: j.c.goslings@amc.uva.nl [Trauma Unit Department of Surgery, Academic Medical Center, Amsterdam (Netherlands)

    2011-11-15

    Introduction: The Amsterdam Trauma Workflow (ATW) concept includes a sliding gantry CT scanner serving two mirrored (trauma) rooms. In this study, several predefined scenarios with a varying number of CT scanners and CT locations are analyzed to identify the best performing patient flow management strategy from an institutional perspective on process quality. Materials and methods: A total of six clinically relevant scenarios with variables that included the number of CT scanners, CT scanner location, and different patient categories (regular, urgent, and trauma patients) were evaluated using computer simulation. Each scenario was simulated using institutional data and was assessed for patient waiting times, idle time of CT scanners, and overtime due to scheduling. The best 2- and 3-scanner scenarios were additionally evaluated with the ATW-concept. Results: Based on institutional data, the best 2-scanner scenario distributes all 3 patient categories over both scanners and plans 4 urgent patients per hour while locating both scanners outside of the trauma room. The best 3-scanner scenario distributes urgent and regular patients over all 3 scanners and trauma patients on only 1 scanner and locates all CT scanners outside of the trauma room. The ATW concept reduces waiting times and overtime, while increasing idle time. Conclusion: Choosing the optimal planning and distribution strategies depends on the number and location of available CT scanners, along with number of trauma, urgent and regular patients. The Amsterdam Trauma Workflow concept could provide institutions with the ability of early CT scanning in trauma patients without influencing regular and urgent CT scanning.

  17. 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.

  18. Monte Carlo simulations of models for accelerator transmutation of waste

    International Nuclear Information System (INIS)

    The Los Alamos Accelerator Transmutation of Waste (ATW) program is directed toward the dual goals of alleviating the problems associated with existing high-level radioactive defense wastes, and of developing systems for the generation of fission energy with minimal production of high-level, long-lived nuclear wastes. In the Los Alamos ATW concept, a high-current, high-energy proton accelerator creates and intense flux of neutrons through spallation in heavy metal targets. The high neutron flux levels available in such systems allow the rapid burning even of nuclides with small cross sections, the design of systems with dilute inventories, and the operation of systems far from criticality. A crucial tool for ATW simulations is the LAHET Code System (LCS), which consists of the Los Alamos version of the HETC Monte Carlo code, a special version of the MCNP code, and several tallying and postprocessing utilities. Here we present results for a baseline system designed to transmute technetium. 16 refs

  19. 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.

  20. 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.

  1. 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

  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. Retran kinetics studies for loss of feedwater anticipated transients without scram

    International Nuclear Information System (INIS)

    Results are presented for postulated loss of feedwater (LOFW) anticipated transient without scram (ATWS) analyses performed with the RETRAN-02 MOD003 computer code for a four-loop Westinghouse pressurized water reactor in order to demonstrate the capabilities of the RETRAN code to predict system transient responses similar to the Westinghouse LOFTRAN code and the Brookhaven National Laboratory RELAP code results. Differences between the RETRAN, LOFTRAN and RELAP code predictions can be attributed to differences in methodology and modelling techniques. These LOFW ATWS studies provide additional insight into the kind of detailed parametric analyses required for using RETRAN for ATWS studies or for the evaluation of any reactor transient dominated by moderator density feedback phenomena. This paper gives the results of point kinetics sensitivity studies and steam generator relief valve modelling effects. The effect of using a realistic steam generator model with multiple banked relief valve setpoints versus an ideal single plateau value is presented. The modelling of actual multiple banked opening and closing pressure setpoints result in higher primary system pressure due to the fast steam generator inventory depletion

  4. Improving CT scan capabilities with a new trauma workflow concept: Simulation of hospital logistics using different CT scanner scenarios

    International Nuclear Information System (INIS)

    Introduction: The Amsterdam Trauma Workflow (ATW) concept includes a sliding gantry CT scanner serving two mirrored (trauma) rooms. In this study, several predefined scenarios with a varying number of CT scanners and CT locations are analyzed to identify the best performing patient flow management strategy from an institutional perspective on process quality. Materials and methods: A total of six clinically relevant scenarios with variables that included the number of CT scanners, CT scanner location, and different patient categories (regular, urgent, and trauma patients) were evaluated using computer simulation. Each scenario was simulated using institutional data and was assessed for patient waiting times, idle time of CT scanners, and overtime due to scheduling. The best 2- and 3-scanner scenarios were additionally evaluated with the ATW-concept. Results: Based on institutional data, the best 2-scanner scenario distributes all 3 patient categories over both scanners and plans 4 urgent patients per hour while locating both scanners outside of the trauma room. The best 3-scanner scenario distributes urgent and regular patients over all 3 scanners and trauma patients on only 1 scanner and locates all CT scanners outside of the trauma room. The ATW concept reduces waiting times and overtime, while increasing idle time. Conclusion: Choosing the optimal planning and distribution strategies depends on the number and location of available CT scanners, along with number of trauma, urgent and regular patients. The Amsterdam Trauma Workflow concept could provide institutions with the ability of early CT scanning in trauma patients without influencing regular and urgent CT scanning.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-01

    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.

  6. 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

  7. Automatic control device for feedwater flow rate into reactor

    International Nuclear Information System (INIS)

    In automatic control for a water injection flow rate, an anticipated transient without screw (ATWS) signal is outputted upon judgement of the occurrence of ATWS event based on a reactor power signal and a scram demand signal, and a high pressure water injection system inactivation signal is outputted upon detection for the inactivation of a high pressure water injection system. An ATWS/high pressure water injection system inactivation judging section outputs a high pressure water injection system inactivation signal. A reactor pressure capable of water injection and a pressure change signal for setting opening/closing of a main steam relief valve corresponding thereto are calculated to output the same to a pressure control section for setting opening/closing of the main steam relief valve. Even if insertion of the entire control rods should fail upon scram by the loss of reactor water to disable the scram, and high pressure water injection system is not operated, the reactor pressure and the water level of the reactor are automatically controlled, and water is injected from a low pressure water injection system with no trouble, to suppress the reactor power. Then, the integrity of the reactor pressure vessel and the reactor container can be maintained. (N.H.)

  8. 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

  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 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.)

  11. Nuclear energy and reactor safety: national and international perspectives; Kernenergie und Reaktorsicherheit: Nationale und internationale Perspektiven

    Energy Technology Data Exchange (ETDEWEB)

    Birkhofer, A. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany)

    2000-06-01

    At the end of 1998, 433 nuclear power stations were operating around the world while a further 36 were under construction. The experience gained worldwide in the operation of such power stations amounts to some 8,000 years of operation, with pressurised-water and water-boiler reactors constituting about 70% of this figure. A safety evaluation of western nuclear power stations produced good results. Although a serious accident occurred at Harrisburg in 1979, no damage was caused outside the power station. The efforts during recent decades to continue to develop safety levels have resulted in improvements to many safety indicators. However, major safety deficiencies were discovered in Eastern Europe. The disaster which occurred at the Chernobyl power station is dramatic evidence of this. One of the reasons why reactor safety has become a cross-border task is partly due to this catastrophe. Germany has had a considerable influence on the development of safety levels in other countries to date. But there is a serious risk of this influence being lost due to Germany's intention to stop using nuclear power, especially because this will endanger the preservation of technical competence in Germany. (orig.) [German] Ende 1998 waren weltweit 433 Kernkraftwerke am Netz, weitere 36 Reaktoren in Bau. Die weiltweite Betriebserfahrung liegt inzwischen bei rund 8000 Anlagenbetriebsjahren, wovon etwa 70% auf westliche Druck- und Siedewasserreaktoren entfallen. Die Sicherheitsbilanz der wesentlichen Kernkraftwerke ist gut. Zwar kam es 1979 zu einem schweren Unfall in Harrisburg, es traten aber keine Schaeden ausserhalb des Kraftwerks auf. Die Anstrengungen der letzten Jahrzehnte zur Weiterentwicklung der Sicherheit schlagen sich in Verbesserungen vieler Sicherheitsindikatoren nieder. Zu erheblichen Sicherheitsdefiziten kam es dagegen in Osteuropa. Die Reaktorkatastrophe von Tschernobyl hat dies drastisch gezeigt. Nicht zuletzt dadurch ist die Reaktorsicherheit heute zu einer

  12. 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

  13. 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

  14. Anticipated transient without scram events at Salem. Another lesson in operating experience

    International Nuclear Information System (INIS)

    The two anticipated transient without scram (ATWS) events at the Salem nuclear power plant have been called the most significant operating events in terms of reactor safety since the accident at Three Mile Island. Simultaneous equipment failures at Salem in February 1983 resulted in the first time a U.S. commercial nuclear power plant failed to scram automatically on a valid reactor protection signal. The plant was operating at low power levels on both occasions and was promptly scrammed manually so there was no plant damage and no direct danger to public health and safety. However, the implications of the Salem events are of wide significance in terms of reactor trip system reliability in particular and utility management controls in general. The Salem event, by itself, was a mild transient, but the potential existed for a much more serious event. Together, the two Salem events provide a number of specific lessons for improving and assuring the safety, reliability and economics of commercial nuclear power plants. The paper gives the details of the ATWS events. The reactor trip system is described with emphasis on the reasons why the trip breakers failed to open on demand. The operating history of the circuit breakers used in the reactor trip system at Salem and other PWRs is analysed and evaluated. The implications of the Salem events were assessed by the United States Nuclear Regulatory Commission (NRC) and a number of proposed actions have been identified to ensure proper operation of all reactor trip breakers in the future. Further, as the result of the Salem events, the NRC continued to study the proposed ATWS regulation with regard to additional requirements such as whether a diverse scram system for Westinghouse-designed plants is necessary

  15. 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)

  16. 2009 Annual meeting on nuclear technology. Pt. 2. Topical sessions

    International Nuclear Information System (INIS)

    Summary report on the Topical Sessions of the Annual Conference on Nuclear Technology held in Dresden, May 12 to 14, 2009: Flexible Concept for Sustainable Use of Nuclear Energy and Waste Minimization; Thermohydraulic Experiments for Reactors of the Second and Third Generation. The reports on the Topical Sessions: Advances in the Development of Integrated Management Systems for the Optimisation of Safety and Operational Availability of Nuclear Power Stations (Dr. Markus Nie and Dipl.-Ing. Karl Ramler), and; Fuel Elements: Zero Failure - Road to the Target (Dipl.-Ing. Andreas Huettmann) have been covered in atw 8/9 (2009). (orig.)

  17. 2009 Annual meeting on nuclear technology. Pt. 2. Topical sessions; Jahrestagung Kerntechnik 2009. T.2. Fachsitzungsberichte

    Energy Technology Data Exchange (ETDEWEB)

    Fazio, Concetta [Forschungszentrum Karlsruhe/KIT, Eggenstein-Leopoldshafen (Germany); Delpech, Marc [Centre CEA de Saclay (Essonne), Gif-sur-Yvette (France); Schaffrath, Andreas [TUeV NORD SysTec GmbH und Co. KG, Abt. Sicherheitsanalyse und Systemtechnik (ETB), Hamburg (Germany)

    2009-10-15

    Summary report on the Topical Sessions of the Annual Conference on Nuclear Technology held in Dresden, May 12 to 14, 2009: Flexible Concept for Sustainable Use of Nuclear Energy and Waste Minimization; Thermohydraulic Experiments for Reactors of the Second and Third Generation. The reports on the Topical Sessions: Advances in the Development of Integrated Management Systems for the Optimisation of Safety and Operational Availability of Nuclear Power Stations (Dr. Markus Nie and Dipl.-Ing. Karl Ramler), and; Fuel Elements: Zero Failure - Road to the Target (Dipl.-Ing. Andreas Huettmann) have been covered in atw 8/9 (2009). (orig.)

  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

    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.

  20. 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.

  1. 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

  2. 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.

  3. Inherent controllability in modular ALMRs

    International Nuclear Information System (INIS)

    As part of recent development efforts on advanced reactor designs ANL has proposed the IFR (Integral Fast Reactor) concept. The IFR concept is currently being applied to modular sized reactors which would be built in multiple power paks together with an integrated fuel cycle facility. It has been amply demonstrated that the concept as applied to the modular designs has significant advantages in regard to ATWS transients. Attention is now being focussed on determining whether or not those advantages deriving from the traits of the IFR can be translated to the operational/DBA (design basis accident) class of transients. 5 refs., 3 figs., 3 tabs

  4. 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

  5. List of the world's nuclear power plants. Tabulated statistical survey as of September 1994, covering 31 countries, listing nuclear power plants in operation, under construction, in the order book

    International Nuclear Information System (INIS)

    The list of the world's nuclear power plants published annually by atw in its 1994 edition (featuring the state as of September) contains a total of 483 nuclear power plants with an aggregate 413 939 MWe, which are in operation and under construction in 33 states. Of these, 422 with an aggregate 357 923 MWe are in operation in 30 countries and 61 plants with 56 106 MWe are under construction in 17 countries. The list contains all nuclear power plants with their designations, locations, operators, types, manufactures, net powers, and the actual and planned dates, respectively, of the beginning of commercial operation. (orig.)

  6. Extension of the Paks full-scale simulator to severe accidents

    International Nuclear Information System (INIS)

    In the present form the Paks full-scale simulator can be used only for th e initial phase of LOCA transients. This simulator is currently being upgraded to simulate severe accidental situations in the primary circuit, such as core dryout, reflooding, fuel element rupture, ATWS events and control rod ejection. The upgrading is due to be finished in the first half of 1993. The main characteristics of the new models are described, a detailed nodalization scheme of the primary circuit is given, and the verification process of the simulator is presented. (author) 6 refs.; 7 figs

  7. 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...

  8. 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.

  9. Human error probability evaluation as part of reliability analysis of digital protection system of advanced pressurized water reactor - APR 1400

    International Nuclear Information System (INIS)

    A case of study on human reliability analysis has been performed as part of reliability analysis of digital protection system of the reactor automatically actuates the shutdown system of the reactor when demanded. However, the safety analysis takes credit for operator action as a diverse mean for tripping the reactor for, though a low probability, ATWS scenario. Based on the available information two cases, viz., human error in tripping the reactor and calibration error for instrumentations in protection system, have been analyzed. Wherever applicable a parametric study has also been performed

  10. Heat Flux and Entropy Produced by Thermal Fluctuations

    DEFF Research Database (Denmark)

    Ciliberto, S.; Imparato, Alberto; Naert, A.;

    2013-01-01

    conservation law for the fluctuating entropy, which we justify theoretically. The system is ruled by the same equations as two Brownian particles kept at different temperatures and coupled by an elastic force. Our results set strong constraints on the energy exchanged between coupled nanosystems held at......We report an experimental and theoretical analysis of the energy exchanged between two conductors kept at different temperature and coupled by the electric thermal noise. Experimentally we determine, as functions of the temperature difference, the heat flux, the out-of-equilibrium variance, and a...... different temperatures....

  11. 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)

  12. 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)

  13. New nuclear power plants in Europe 1987. Pt. 2

    International Nuclear Information System (INIS)

    The atw report on new nuclear power plants in Europe contains both a survey of the Federal Republic of Germany, which was published in the April 1987 issue, and an overview of the nuclear power plant situation in 20 European countries including the Soviet Union and six other CMEA countries. Also this year's review includes specific status reports, complete with technical information, about all nuclear generating units under construction, in the project and concrete planning phases. Introductory summaries describe the plants in operation in each country and their 1986 electricity generation. A general introduction provides an outlook on developments in Western and Eastern Europe. (orig./HP)

  14. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities

  15. On-line fuel and control rod integrity surveillance in BWRs

    International Nuclear Information System (INIS)

    Surveillance of fuel and control rod integrity in a BWR core is essential to maintain a safe and reliable operation of the nuclear power plant. Any actions to be taken in the event of a fuel failure during reactor operation should be based on the best available information regarding the failure and expected consequences. The detection of fuel and control rod failures in BWRs is usually performed by analyzing samples of off-gases and coolant taken with a certain time intervals, e.g. once a week or once a month. This procedure can, however, leave the failure undetected in the core for quite some time. Therefore, a sufficient improvement of the surveillance of fuel and control rods can be achieved by simultaneous measurements of He and gamma emitting noble gases on-line in the off gas system. In this paper, experiences of such measurements performed at Kernkraftwerk Leibstadt (KKL) in Switzerland and Forsmark nuclear power plant (NPP) in Sweden will be presented. (author)

  16. 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.)

  17. 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.

  18. Erection and construction progress of the THTR 300 prototype nuclear power station

    International Nuclear Information System (INIS)

    Despite the new aspects that have arisen in the commercialization of the high temperature reactor line, the construction of the THTH 300 MW Nuclear Power Station represents an important and necessary milestone in the development of the HTR line, no matter which development variants will ultimately be implemented. The builder and owner of the plant is Hochtemperatur-Kernkraftwerk GmbH (HKG), the contractor is the Konsortium THTR, a group comprised of Brown, Boveri and Cie. AG (BBC), Hochtemperatur-Reaktorbau GmbH (HRB), and NUKEM GmbH. The primary system is characterized by the pebble bed core consisting of some 675,000 fuel elements and designed for a thermal core power of 750 MW, by the containment of the core in graphite internals, helium as the coolant, and six cooling gas circulators and six steam generators, all of which are housed in the pressure vessel, which is designed as a prestressed concrete pressure vessel. In processing the seven partial construction permits granted so far, the licensing authorities imposed stringent conditions which, in addition to requiring major changes to be made in the buildings, resulted in a drastic modification of the overall concept which was implemented only at considerable expense, both in terms of funds and time. Orders for more than 90% of the components and systems of the primary circuit have been placed; most of the major components have been completed. (orig.)

  19. 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.)

  20. The SNR-300 prior to commissioning

    International Nuclear Information System (INIS)

    Construction of the SNR-300, a nuclear power plant equipped with a sodium cooled fast breeder reactor (327 MWe), was begun in Kalkar in 1973. Now, after twelve years, the reactor vessel and all other sodium carrying parts are filled with sodium. This more or less means completion of the nuclear power plant. The governments of the Federal Republic of Germany, Belgium and the Netherlands support the project financially and commissioned the Schnell-Brueter-Kernkraftwerksgesellschaft mbH (SBK) of Essen to manage it in the light of industrial principles and with a view to the later commercial application of fast breeder reactors. RWE holds a 68.85% interest in SBK, the Belgian Electronucleaire (EN) and the Netherlands Samenwerkende Elektriciteits-Productiebedrijven (SEP) each hold 14.75%, and the British Central Electricity Generating Board (CEGB) holds 1.65%. The contract for the turnkey power plant was awarded to a consortium of vendors consisting of the Internationale Natrium-Brutreaktor-Baugesellschaft (INB), in which the German Interatom holds 70% and the Netherlands and Belgian firms of Neratoom and Belgonucleaire each hold 15%, and the Bauarbeitsgemeinschaft Kernkraftwerk Kalkar under the leadership of the German Hochtief AG. The decision in 1972 to build the plant had been preceded by many years of research and development work at the Karlsruhe Nuclear Research Center and research laboratories abroad, which was continued in a supporting program during the construction phase. (orig.)

  1. 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)

  2. 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

  3. 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

  4. New Development of Accelerator Methods for Energy Production in the 21st Century - Alternative to Nuclear reactors

    International Nuclear Information System (INIS)

    In recent years, and particularly in light of a growing influence of ecological and political factors in the field of energy development, new concepts of nuclear technology based on the use of nuclear particle accelerators are being considered. Subcritical source-driven nuclear systems (SSDNS) are given close attention, whereby the application of a high-intensity external neutronic source can cause major fission power by affecting the subcritical nucleus. Recent research within the project of the accelerator driven transmutation technology (ADTT) in Los Alamos shows that now it is possible to use this concept, specially owing to a considerable improvement of the high-power accelerators predominantly realised in the military project for the development of space weapons. Accelerator driven neutronic spallation is considered a particularly promising source in achieving the subcritical fission. The project also reviews the development of new technologies related to the problem of nuclear waste by means of accelerator driven transmutation of waste (ATW). This method could result in the demolition of plutonium, trans-uranium elements and long-lasting fission products and thereby in their removal from the environment- Possible consequences of ADTT and ATW on the energy sector development in the 21st century are being considered. (author)

  5. Nuclear power plants 1995 - a world survey

    International Nuclear Information System (INIS)

    The atw Statistics Report compiled by atw lists 428 nuclear power plants with 363 397 gross MWe in operation in 30 countries in late 1995. Another 62 units with 55 180 gross MWe were under construction in 18 countries. This adds up to a total of 490 units with an aggregate 418 577 MWe. In the course of 1995 four units in four countries started commercial operation. In the survey of electricity generation in 1995 for which no information was made available from China and Kasachstan, a total of 417 nuclear power plants were covered. In the year under review they generated an aggregate 2 282 614 GWH, which is 3.4% more than in the previous year. The highest nuclear generation again was recorded in the USA with 705 771 GWh, followed by France with 377 021 GWh. The Grohnde power station in Germany attained the maximum annual production figure of 11 359 GWh. The survey includes nine tables indicating the generating performance of each nuclear power plant, the development of electricity generation in nuclear plants, and status of nuclear power plants at the end of 1995 arranged by countries, types of reactors, and reactor manufacturers. (orig.)

  6. 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.

  7. 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)

  8. Investigations of anticipated transients without reactor scram and other selected safety devices

    International Nuclear Information System (INIS)

    A detailed study of anticipated transients without scram (ATWS) has been carried out in the Federal Republic of Germany based on a boiling water reactor (BWR) and a pressurized water reactor (PWR) reference plant. The study includes transient calculations as well as reliability analyses of the entire scram system (sensors, logic, actuating system). In addition, the influence of other safety related systems (pressure relief system, pump control system in a BWR) has been evaluated. During all ATWS, system pressure does not exceed 110% of design pressure. Only for short periods (several seconds) and only in small areas of the core might film boiling occur. The availability of the scram systems for both a BWR and a PWR is on the order of 10-5 per demand. From these results, it is concluded that no independent second scram system is necessary. However, the detailed analysis has griven an indication of where hardware measures could be taken to mitigate the transients (e.g., increase of valve capacity) or further improve the availability of the scram system

  9. 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

  10. 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.

  11. Important objectives for the performance of an individual plant evaluation

    International Nuclear Information System (INIS)

    The Pennsylvania Power ampersand Light Company (PP ampersand L) operates the Susquehanna Steam Electric Station, a two-unit 1050-MW(electric) boiling water reactor (BWR) 4, Mark II plant. Unit 1 has been in operation since 1982 and Unit 2 since 1984. The PP ampersand L views on probabilistic risk assessment (PRA) have been shaped by early in-house experience during the period 1981 to 1985 in deterministic studies of station black-out and anticipated transient without scram (ATWS) and a conventional level 3 PRA for Susquehanna performed by a contractor. In the performance of studies of station black-out and ATWS, the objective was to assure that procedures would utilize all of the plant's capabilities to avoid or minimize damage regardless of the extent of additional equipment failures accompanying or resulting from these initiators. It was found that when the emergency operating procedures (EOPs) were properly structured to accommodate the potential for additional equipment failures, damage to the plant could be avoided or greatly reduced in severity below what could be expected in the absence of optimized procedures (even for very severely degraded conditions of the plant). Accomplishment of such an improvement required that the modifications be incorporated into the EOPs explicitly and that operators be trained in their use. In the formulation of the Susquehanna EOPs and the associated program of operator training, it was attempted to assure that the optimized procedures and associated training would be in place and effective

  12. 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)

  13. 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.

  14. 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

  15. 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

    International Nuclear Information System (INIS)

    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 10-10 m/s and 10-9 m/s respectively. The volume of waste allocated to the L/ILW repository is somewhat smaller for the 10-9 m/s variant than for the 10-10 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 to safety of the different

  16. 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

  17. 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

  18. Transmutation of Isotopes - Ecological and Energy Production Aspects

    International Nuclear Information System (INIS)

    This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions - after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors - are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional

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

    International Nuclear Information System (INIS)

    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

  20. German (GRS) approach to accident analysis (part I). German licensing basis for accident analyses. Applicants accident analyses in second part license for Konvoi-plants. Appendix 1. Assessor accident analyses in second part license for Konvoi-plants. Appendix 2. Reference list of DBA to be considered in the safety status analysis of a PSR. Appendix 3a. Reference list of special very rare and BDB plant conditions to be considered in the safety status analysis of a PSE. Appendix 3b

    International Nuclear Information System (INIS)

    Appendix 1: The Safety Analysis Report (S.A.R.) is presented from 3 Handbooks - ECC Handbook (LOCA), Plant Dynamics Handbook (Transients incl. ATWS), and Core Design Handbook. The first one Conceived as Living handbook, Basis for design, catalogue of transients, specifications and licensing. Handbook contains LOCA in primary system, it contains also core damage analysis, and description of codes, description of essential plant data and code input data. The second one consists of Basis for design, commissioning, operation, and catalogue of transients, specifications and licensing, as well as specified operation, disturbed operation, incidents, non-LOCA, SS-procedures and Code description. The third book consists of Reactivity balance and reactivity coefficients, efficiency of shutdown systems. Calculation of burn up cycle, power density distribution, and critical boron concentration. Also Codes used, as SAV79A standard analysis methodology including FASER for nuclear data generation, MEDIUM and PANBOX for static and transient core calculations. Appendix 2: The three TUEV (Technical Inspection Agencies) responsible for the three individual plants of type KONVOI: TUEV Bayern for ISAR-2, TUV-Hanover for KKE, TUEV-Stuttgart for GKN-2 and GRS performed the safety assessment. TUV-Bayern for disturbance and failure of secondary heat sink without loss of coolant (failure of main heat sink, erroneous operation of valves in MS and in FW system, failure of MFW supply), long term LONOP, performance of selected SBLOCA analyses. TUV Hanover for disturbances due to failure of MCPs, short term LONOP, damages of SG tubes incl. SGTR, performance of selected LOCA analyses (blowdown phase of LBLOCA). TUV-Stuttgart for breaks and leaks in MS and FW system with and without leaks in SG tubes. GRS for ATWS, sub-cooling transients due to disturbances on secondary side, initial and boundary conditions for transients with opening of pressurizer valves with and without stuck-open, most of the

  1. 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

  2. Core supervision methods and future improvements of the core master/presto system at KKB

    International Nuclear Information System (INIS)

    Kernkraftwerk Brunsbuettel (KKB) is a KWU 806 MWe BWR located at the lower river Elbe, in Germany. The reactor has been in operation since 1976 and is now operating in its 14. cycle. The core supervision at KKB is performed with the ABB CORE MASTER system. This system mainly contains the 3-D simulator PRESTO supplied by Studsvik Scandpower A/S. The core supervision is performed by periodic PRESTO 3-D evaluations of the reactor operation state. The power distribution calculated by PRESTO is adapted with the ABB UPDAT program using the on-line LPRM readings. The thermal margins are based on this adapted power distribution. Related to core supervision, the function of the PRESTO/UPDAT codes is presented. The UPDAT method is working well and is capable of reproducing the true core power distribution. The quality of the 3-D calculation is, however, an important ingredient of the quality of the adapted power distribution. The adaptation method as such is also important for this quality. The data quality of this system during steady state and off-rate states (reactor manoeuvres) are discussed by presenting comparisons between PRESTO and UPDAT thermal margin utilisation from Cycle 13. Recently analysed asymmetries in the UPDAT evaluated MCPR values are also presented and discussed. Improvements in the core supervision such as the introduction of advanced modern nodal methods (PRESTO-2) are presented and an alternative core supervision philosophy is discussed. An ongoing project with the goal to update the data and result presentation interface (GUI) is also presented. (authors)

  3. 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

  4. 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

  5. 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.)

  6. 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

  7. 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

  8. 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.)

  9. 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.)

  10. 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

  11. Morality and ethics in high technology

    International Nuclear Information System (INIS)

    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.)

  12. 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

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. Pipe loads in the case of a safety valve blowdown with phase change

    International Nuclear Information System (INIS)

    For design of piping systems the dynamical response of the structure due to flow induced forces under normal and accident conditions has to be known. As an example in the present paper we discuss the safety- and relief valve systems of a KWU pressurized water reactor that functionally protects the primary reactorsystem against overpressure. This facility consists essentially of the pressurizer and the main blowdown pipes to the safety- and relief valves. The transient loadings for this piping system emerge from flow induced forces during valve operation under normal conditions (steam discharge) and under accident conditions (subcooled water and 2-phase mixture discharge). The latter situation may occur in a few cases of anticipated transients without reactor shutdown (ATWS). (orig./GL)

  18. 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.

  19. 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.)

  20. 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.

  1. The heating operational summarization in three winters of a 5 MW test heating reactor

    International Nuclear Information System (INIS)

    The 5 MW THR (5 MW test heating reactor) is a new type reactor with inherent safety developed by INET (Institute of Nuclear Energy Technology). It is the first 'pressure vessel type' heating reactor in operation in the world. It was put into operation in November, 1989. Since then it has operated for three winter seasons. The total operation time has reached to 8174 hours and its availability of heating has reached to 99%. The advanced technology of this reactor has been proved in the past three years operation. The characteristics of power regulating, load following, reactivity disturbance and the variation of parameters under the condition of ATWS (anticipated transients without scram) were studied with experiments in 5 MW THR. The 5 MW THR is an ideal heating reactor and has outstanding performances

  2. 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

  3. Calculation of a measured WWER-1000 pump trip transient using coupled 3D kinetics, power controller and thermohydraulic system models

    International Nuclear Information System (INIS)

    The HEXART 3D kinetics code, coupled to a Power Controller model and NPP simulator has been used to calculate a start-up transient of Unit 3 of the South Ukrainian NPP. The objectives are to test the 3D neutron kinetics and power controller ex-vessel detection models in a complex ATWS. The initiating event is simultaneous trip of 2 adjacent MCP out of 4 at 97 % of the nominal rated power. This transient is a serious validation test of core dynamics and plant component interactions with the reactor core. Modeling of the Ex-vessel Detection System and non-uniform coolant mixing is of particular importance. The computed are in satisfactory agreement with the plant measurements. (Authors)

  4. 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).

  5. 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.

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

    DEFF Research Database (Denmark)

    Wulff, Anders Christian

    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 eect 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 Gd2Zr2O7 buer layer was investigated taking grain boundaries into account. 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...

  7. 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.)

  8. An overview of accelerator-driven transmutation technology

    International Nuclear Information System (INIS)

    Accelerator-Driven Transmutation Technology, or ADT2, is a collection of programs that share a common theme - they each have at their heart an intense source of neutrons generated by a high-energy proton beam striking a heavy metal target. The beam energy, typically 1000 MeV, is enough for a single proton to smash a target atom into atomic fragments. This so-called spallation process generates large numbers of neutrons (around 20 to 30 per proton) amid the atomic debris. These neutrons are of high value because they can be used to transmute neighboring atoms by neutron capture. Three distinct ADT2 program elements will be described. These are ADEP - accelerator-driven energy production, ABC - accelerator based conversion (of plutonium) and ATW - accelerator transmutation of waste

  9. Current and Future Applications of Thermal-Hydraulic and Neutronic Best- Estimate Methods in Support of the Swiss Licensing Process

    International Nuclear Information System (INIS)

    Best-estimate methods are employed within the STARS project at PSI. The project provides independent technical expertise primarily to HSK in support of the Swiss licensing process in the area of deterministic safety analysis, and covers applications ranging from operational occurrences as well as beyond-design-basis transients. Some examples are presented: an assessment of the coolability of a PWR for reduced availability of ECCS, timing studies for an MSIV-ATWS event in a BWR in order to quantify the effect of delayed operator actions, predictions of planned start-up tests and research on BWR-stability to deepen the understanding of the phenomenon. Future work will include an extensive use of 3D-kinetics in transient analysis and the development of a comprehensive uncertainty methodology. Calculations on high burnup fuel behavior for RIA- and LOCA-scenarios are planned; also CFD applications will be developed to assess mixing problems in reactor geometries on a rigorous basis. (authors)

  10. 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.)

  11. Determining critical flow valve characteristics using extrapolation techniques

    International Nuclear Information System (INIS)

    This report presents the methodology and documentation of the calibration of the Loss-of-Fluid Test (LOFT) power-operated relief and safety relief valve (PORV + SRV) for the L9-3 anticipated transient without scram (ATWS) experiment. A multiposition globe valve was calibrated to produce scaled high-pressure flow rates using a low-pressure calibration facility and a simple RELAP5 critical flow model to extrapolate the calibration data to expected operating pressures. It was demonstrated that an accurate high-pressure, multiphase flow calibration can be performed without the necessity of actual high-pressure testing. This technique, when applied to large pressurized water reactor (LPWR) safety and relief valves, represents a potentially large savings in the capacity qualification procedure of full-scale pressure reduction valves

  12. Development of a RBMK-1500 Reactor Model Based on a Coupled Version of The Thermal - Hydraulic Code ATHLET and the 3D Neutronics Code QUABOX/CUBBOX

    International Nuclear Information System (INIS)

    The codes ATHLET and QUABOX/CUBBOX were developed by German company GRS for light water reactors. For RBMK-1500 NPP a model for the simulation of transients was developed, using a coupled version of the thermal - hydraulic system code ATHLET and the 3D core model QUABOX/CUBBOX. The coupled code system was applied to the analysis of an ATWS event with 'loss of feedwater'. There are local differences, which are influencing the transient behaviour and can be taken into account only by the 3D core model. The LAC system can only compensate the strong reactivity feedback by the fuel temperature rise for a short time. The results of 3D -kinetics and point kinetics models were compared. (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. 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)

  15. Concept of the new generation high safety liquid metal reactor (LMFR)

    International Nuclear Information System (INIS)

    The comparative analysis of the inner stability of the liquid metal reactors to severe accidents was made using the asymptotic reactivity balance. The group of the BN-reactors, Superphenix, IFR, LMFR were considered. This paper lists the characteristics of the reactors, used in the self-protectiveness analysis. The authors present the maximum coolant temperatures in post-accident asymptotic state for IFRs as on of the possible designs of a high safety fast reactor with metal fuel, U-Pu-Zr and LMFR. As is known, these values are very important for assessment of the ATWS accidence consequences. The authors consider the following situations and their combinations: loss of reactor coolant flow-LOFWS, loss of heat sink-LOHSWS, uncontrolled reactor sodium overcooling (down to the freezing point)-OVCWS, uncontrolled excess reactivity insertion-TOPWS. The calculation results demonstrate a high stability of the IFR and LMFR reactors to the most severe accidence sequences

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

    International Nuclear Information System (INIS)

    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.)

  17. Nuclear power 1996: potential for further development. Session reports of the annual meeting on nuclear technology, May 21-23, 1996 in Mannheim

    International Nuclear Information System (INIS)

    For the third time, the 'Rosengarten', the congress center of the city of Mannheim, was the venue of the Annual Nuclear Conference, this year on May 21-23, 1996. Attendance showed a slight increase, and the organizers, Deutsches Atomforum (DAtF) and the Kerntechnische Gesellschaft (KTG), welcomed more than 1000 participants at the world's biggest event of this kind. The program was arranged in the traditional, proven format, with plenary sessions on the first day, and technical sessions, poster sessions, special events, and technical excursions on the other two days. These proceedings were accompanied by an exhibition arranged by vendors, suppliers, and service companies. Following the summary of the plenary day published on pp. 385-95 in atw 6/96, the survey in this issue covers the technical sessions as seen by the rapporteurs. (orig.)

  18. Nuclear power 1997: Assured know-how. Session reports of the annual meeting on nuclear technology, May 13-15, 1997, in Aachen

    International Nuclear Information System (INIS)

    After 1986 the 'Eurogress' of Aachen, was the venue of the Annual Nuclear Conference, this year on May 15, 1997. Attendance showed a slight increase, and the organizers. Deutsches Atomforum (DAtF) and the Kerntechnische Gesellschaft (KTG), welcomed more than 1000 participants at the world's biggest event of this kind. The program was arranged in the traditional proven format, with plenary sessions on the first day, and technical sessions, poster sessions, special events, and technical excursions on the other two days. These proceedings were accompanied by an exhibition arranged by vendors, suppliers, and service companies. Following the summary of the plenary day published on pp. 375 to 386 in atw 7/97, the survey in this issue covers the technical sessions as seen by the rapporteurs. (orig.)

  19. Flight Research and Validation Formerly Experimental Capabilities Supersonic Project

    Science.gov (United States)

    Banks, Daniel

    2009-01-01

    This slide presentation reviews the work of the Experimental Capabilities Supersonic project, that is being reorganized into Flight Research and Validation. The work of Experimental Capabilities Project in FY '09 is reviewed, and the specific centers that is assigned to do the work is given. The portfolio of the newly formed Flight Research and Validation (FRV) group is also reviewed. The various projects for FY '10 for the FRV are detailed. These projects include: Eagle Probe, Channeled Centerbody Inlet Experiment (CCIE), Supersonic Boundary layer Transition test (SBLT), Aero-elastic Test Wing-2 (ATW-2), G-V External Vision Systems (G5 XVS), Air-to-Air Schlieren (A2A), In Flight Background Oriented Schlieren (BOS), Dynamic Inertia Measurement Technique (DIM), and Advanced In-Flight IR Thermography (AIR-T).

  20. AcEST: DK963417 [AcEST

    Lifescience Database Archive (English)

    Full Text Available TST39A01NGRL0016_H19 487 Adiantum capillus-veneris mRNA. clone: TST39A01NGRL0016_H19. 5' end seq ... 5.4 tr|A0Y2K3|A0Y2K3_9GAMM ATP-dependent helicase HepA ... OS=Alteromona... 33 9.3 >tr|A6QD08|A6QD08_SULNB Pu ... 442 >tr|A0Y2K3|A0Y2K3_9GAMM ATP-dependent helicase HepA ... OS=Alteromonadales bacterium TW-7 GN=ATW7_12171 PE ...

  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. ALMOS-2 - computer program for accident analyses of BWR-type reactor plants. Pt. 1

    International Nuclear Information System (INIS)

    ALMOS-2 is a non-linear, one-dimensional digital computer program for accident analyses of boiling water reactor plants. The program is capable of calculating transients with large amplitude deviations from steady state conditions, such as pump failures, loss of heat sink, feedwater system failures, control rod malfunction and safety system failures. The program is a suitable tool for the analysis of Anticipated Transients Without Scram (ATWS). The model includes the reactor core, the main coolant system and the safety system. Boundary conditions to these systems can be defined as functions of time by input data. The program description includes a short introduction into the physical model, a presentation of the program structure and program components and a detailed input data description. For test purposes the results of a transient calculation are added. (orig.)

  3. 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.

  4. 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

  5. 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.)

  6. New nuclear power plants in Europe 1983. Pt. 2

    International Nuclear Information System (INIS)

    The atw-report informs in part 1 which appeared in the April volume (4/83) and also includes the Federal Republic of Germany of the situation on the nuclear power plants section in 26 European countries, including 7 countries of the Eastern hemisphere. In these countries, 364 nuclear power plant blocks with a total gross performance of 282313 MWe are being operation, built, projected, and planned. Presently operating are 165 blocks with 89841 MWe gross, i.e. 10 blocks and 8450 MWe more than last year. Presently in the construction phase are 114 blocks with 105908 MWe (previous year: 118 with 109009 MWe). 67 blocks with 70272 MWe are projected (previous year: 77 with 78111 MWe). (orig./UA)

  7. Nuclear power plants in Europe 1991

    International Nuclear Information System (INIS)

    The 'Nuclear Power Plants in Europe 1991' report compiled by atw covers the situation in the nuclear power plant sector in seventeen European countries, among them six former CMEA countries. (The number of countries included in the review has decreased as a result of the unification of the two German states.) As per July 1991, these countries show a total of 222 (1990: 230) nuclear generating units in operation with a cumulated installed gross power of 174,535 MWe (174,675 MWe), and 36 (53) units with 34,615 MWe (49,205 MWe) under construction. This adds up to a total of 258 (283) nuclear generating units with an aggregate 209,150 MWe (223,880 MWe). (orig.)

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

    International Nuclear Information System (INIS)

    KAERI is developing the design concept 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 resistance, and sustainability. 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. The design specifications improved during the first year in the 4th design phase of the KALIMER project were reflected in the analysis. 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 design concept 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 version 1.3 code to investigate the plant responses and safety margins of KALIMER-600. 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 to evaluate the safety of KALIMER-600 subassemblies

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

    International Nuclear Information System (INIS)

    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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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.

  15. 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.)

  16. Human factors review for Severe Accident Sequence Analysis (SASA)

    International Nuclear Information System (INIS)

    The paper will discuss work being conducted during this human factors review including: (1) support of the Severe Accident Sequence Analysis (SASA) Program based on an assessment of operator actions, and (2) development of a descriptive model of operator severe accident management. Research by SASA analysts on the Browns Ferry Unit One (BF1) anticipated transient without scram (ATWS) was supported through a concurrent assessment of operator performance to demonstrate contributions to SASA analyses from human factors data and methods. A descriptive model was developed called the Function Oriented Accident Management (FOAM) model, which serves as a structure for bridging human factors, operations, and engineering expertise and which is useful for identifying needs/deficiencies in the area of accident management. The assessment of human factors issues related to ATWS required extensive coordination with SASA analysts. The analysis was consolidated primarily to six operator actions identified in the Emergency Procedure Guidelines (EPGs) as being the most critical to the accident sequence. These actions were assessed through simulator exercises, qualitative reviews, and quantitative human reliability analyses. The FOAM descriptive model assumes as a starting point that multiple operator/system failures exceed the scope of procedures and necessitates a knowledge-based emergency response by the operators. The FOAM model provides a functionally-oriented structure for assembling human factors, operations, and engineering data and expertise into operator guidance for unconventional emergency responses to mitigate severe accident progression and avoid/minimize core degradation. Operators must also respond to potential radiological release beyond plant protective barriers. Research needs in accident management and potential uses of the FOAM model are described. 11 references, 1 figure

  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. New nuclear power plants in Europe 1984. Pt 2

    International Nuclear Information System (INIS)

    The atw report on new nuclear power plants in Europe contains both a survey of the Federal Republic of Germany, which was pubslished in the April 1984 issue, and an overview of the nuclear power plant situation in 26 European countries including the Soviet Union and six other CMEA countries. Also this year's review includes specific status reports, complete with technical information, about all nuclear generating units under construction, in the project and concrete planning phases. The fifteen nuclear power plants newly commissioned in Europe since last year's atw report was published are covered in a similar way. Moreover, introductory summaries describe the plants in operation in each country and their 1983 electricity generation. A general introduction provides an outlook on developments in Western and Eastern Europe. The total number of nuclear power plants now in operation and under construction in Europe is 300 units with an aggregate gross 219, 320 MWe. Of these, 185 units are located in Western Europe, 115 in Eastern Europe. The units currently in operation of this total are 180 with 103,978 MWe in sixteen European countries; of these, 126 units with 74,869 MWe are run in eleven West European countries, 54 units with an aggregate 29,109 MWe in five East European countries. Of the 120 nuclear generating units at present under construction with an aggregate 115,342 MWe in fifteen European countries, 59 units with 63,442 MWe are located in eight West European, 61 units with 51,900 MWe in seven East European countries. (orig./UA)

  19. 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

  20. 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.

  1. 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)

  2. 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

  3. Assessment of General Atomics accelerator transmutation of waste concept based on gas-turbine-modular helium cooled reactor technology

    International Nuclear Information System (INIS)

    An assessment has been performed for an Accelerator Transmutation of Waste (ATW) concept based on the use of the high temperature gas reactor technology. The concept has been proposed by General Atomics for the ATW system. The assessment was jointly conducted at Argonne National Laboratory (ANL) and Los Alamos national laboratory to assess and to define the potential candidates for the ATW system. This report represents the assessment work performed at ANL. The concept uses recycled light water reactor (LWR)-discharge-transuranic extracted from irradiated oxide fuel in a critical and sub-critical accelerator driven gas-cooled transmuter. In this concept, the transmuter operates at 600 MWt first in the critical mode for three cycles and then operates in a subcritical accelerator-driven mode for a single cycle. The transmuter contains both thermal and fast spectrum transmutation zones. The thermal zone is fueled with the TRU oxide material in the form of coated particles, which are mixed with graphite powder, packed into cylindrical compacts, and loaded in hexagonal graphite blocks with cylindrical channels; the fast zone is fueled with TRU-oxide material in the form of coated particles without the graphite powder and the graphite blocks that has been burned in the thermal region for three critical cycles and one additional accelerator-driven cycle. The fuel loaded into the fast zone is irradiated for four additional cycles. This fuel management scheme is intended to achieve a high Pu isotopes consumption in the thermal spectrum zone, and to consume the minor actinides in the fast-spectrum zone. Monte Carlo and deterministic codes have been used to assess the system performance and to determine the feasibility of achieving high TRU consumption levels. The studies revealed the potential for high consumption of Pu-239 (97%), total Pu (71%) and total TRU (64%) in the system. The analyses confirmed the need for burnable absorber for both suppressing the initial excess

  4. A concept of recriticality-free fast reactor under core degradation accidents and the feasibility of associated practical measure

    International Nuclear Information System (INIS)

    We have analyzed melting and relocation behavior of metallic fuel pin, which has comparatively thin fuel meat in terms of the smeared density in a fuel pin, under accidental condition leading to core degradation. It is for exploring the possibility of eliminating neutronic recriticality in a metallic-fueled fast reactor, where the special fuel pins with the thin meat are partially embedded in the core for realizing controlled materials relocation (CMR) concept. A metallic-fueled fast reactor potentially has advantageous features compared to a MOX-fueled core concerned with burnup reactivity, breeding ratio, and maximum neutron irradiation flux. However, the thermo-hydraulic behavior coupled with neutronics for metallic-fueled fast reactors has not been well examined so far, especially, concerning the recriticality features of metallic-fueled core and its prevention/protection measures. In this study, we have analytically examined the degradation characteristics of metallic-fueled pin, focusing on in-pin fuel melting and relocation, which should contribute to the elimination of recriticality. The analyzed reactor core is a 1000 MWe-class metallic-fueled core. In this core, we embed so-called leading-channel-type fuel subassemblies (FAs), which are about 20-30% to the total fuel assemblies. In a leading FA, fuel pins have comparatively thin fuel pin of which in-pin fuel smear density is about 60%. With this fuel pin structure, the fuel meat can easily relocate on melting. The assumed major driving force is gravity. In this study, we have analyzed ATWS events and discussed the feasibility of realizing CMR in a metallic-fueled core with in-pin fuel relocation. At first, we used a point-kinetics system analysis code, i.e., ARGO, to analyze UTOP (unprotected transient overpower accident) events and determined the location and number of leading FAs to be embedded in the core, which replaced regular FAs. Then, a single pin in a leading FA was modeled to be analyzed for its

  5. Core Design Studies for TRU Transmutation in a Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The objectives of this research project is (1) to develop the conceptual core designs for TRU transmutation covering a large variation in power level and conversion ratio and (2) to perform relevant verification and validation analyses through the analyses of fast critical experimental assemblies. An homogeneous and detailed heterogeneous models of metal fueled critical assemblies, BFS-73-1, BFS-75-1, and BFS-55-1, were produced from this study through a review of the critical experiments. Based on these models, BFS critical assemblies were analyzed by a fast reactor analysis code system (TRANSX/ TWODANT/DIF3D) with different evaluated nuclear data files including ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, JENDL-AC2008 in addition to ENDF/B-VI.6. A study of library difference on computational results by both a conventional diffusion method and a Monte-Carlo transport method has been carried out with those models. In addition to the analysis by the design code, Monte-Carlo high fidelity simulation was carried out to support the diffusion solution, mainly an effect of unit fuel cell heterogeneity. BFS and ZPPR critical assemblies were analyzed by both KAERI and ANL systems and the results of the analyses were reviewed by the other side. This improve the reliability of the results of both institutes. For the effective TRU transmutation, the conceptual core design was performed under core power ranged from 1,500MWt to 4,500MWt and found that there is no appreciable degradation in performance or reactivity coefficients for the core power level up to 1,800 MWe and confirmed the possibility of the large scaled transmutation reactor. Even at each pre-determined power level, performance parameters, reactivity coefficients and its implication on the safety analysis can be different when a target TRU conversion ratio changes. In order to address this aspect of design, a variation study of TRU conversion ratio change was covered. Three ATWS events such as UTOP, ULOF and ULOHS are

  6. 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。

  7. 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

  8. Evaluation of Nodal Reactor Physics Methods for Quasi-Static and Time-Dependent Coupled Neutronic Thermal - Analysis of Pressurized Water Reactor Transients

    Science.gov (United States)

    Feltus, Madeline Anne

    1990-01-01

    This thesis examines coupled time-dependent thermal -hydraulic (T/H) and neutronics solution methods for Pressurized Water Reactor (PWR) transient analysis. The degree of equivalence is evaluated between the typical quasi-static approach and a newly-developed iterative tandem method. Four specific PWR transients that exhibit a wide range of Reactor Coolant System (RCS) T/H response were investigated: (1) a Station Blackout Anticipated Transient Without Scram (ATWS), (2) a Loss of Feedwater ATWS, (3) a Total Loss of RCS Flow with Scram, and (4) a Main Steam Line Break (MSLB). Rather than using simplified RCS and core models, the theory and method in this thesis were applied practically by using realistic models for an actual four-loop Westinghouse PWR plant. The time-dependent STAR kinetics code, based on the QUANDRY Analytic Nodal Method, and the RETRAN and MCPWR T/H systems codes were used to develop a new, fully coupled, tandem STAR/MCPWRQ methodology that runs tandemly on an enhanced 386/387 IBM PC architecture. MCPWRQ uses externally calculated power input rather than point kinetics power level results. The tandem method was compared to quasi -static STAR and time-dependent STAR 2-D and 3-D kinetics results. The new STAR/MCPWRQ method uses RETRAN time-dependent T/H and point kinetics power input as a first estimate. STAR and MCPWRQ are used tandemly to couple STAR 3-D, time-dependent core power results with the MCPWRQ RCS T/H phenomena. This thesis shows that: (a) quasi-static and point kinetics methods are not able to describe severe PWR transient phenomena adequately; and (b) fully coupled, 3-D, time -dependent, tandem (or possibly parallel) analysis methods should be used for PWR reactor transients instead. By tandemly coupling the RCS response in terms of updated core inlet conditions with 3-D time-dependent core kinetics response, the core power response and T/H conditions are forced to be self-consistent during the entire transient. The transient analyses

  9. System Studies for the ADTF: Target and Materials Test Station

    International Nuclear Information System (INIS)

    To meet the objectives of the Advanced Accelerator Applications (AAA) program, the Accelerator-Driven Test Facility (ADTF) provides a world-class accelerator-driven test facility to: - Provide the capability to assess technology options for the transmutation of spent nuclear fuel and waste through a proof-of-performance. - Provide a user facility that allows testing of advanced nuclear technologies and applications, material science and research, experimental physics, and conventional nuclear engineering science applications. - Provide the capability, through upgrades or additions to the ADTF accelerator, to produce tritium for defense purposes, if required. - Provide the capability, through upgrades or additions, to produce radioisotopes for medical and commercial purposes. These missions are diverse and demand a facility with significant flexibility. In order to meet them, it is envisioned that we construct two target stations: the Target and Materials Test (TMT) station and the Subcritical Multiplier (SCM) test station. The two test stations share common hot-cell facilities for post-irradiation examination. It is expected the TMT will come online first, closely followed by the SCM. The TMT will provide the capability to: - Irradiate small samples of proposed ATW (accelerator-driven transmutation of waste) fuels and materials at prototypic flux, temperature, and coolant conditions (requires intense source of neutrons). - Perform transient testing. - Test liquid (lead-bismuth) and solid spallation targets with water, sodium, or helium coolant. - Test generation-IV fuels for advance nuclear systems (requires high-intensity thermal flux). - Irradiate fission product transmutation targets. - Test advanced fuel and coolant combinations, including helium, water, sodium, and lead-bismuth. - Produce isotopes for commercial and medical applications. - Perform neutron physics experiments. The SCM will provide the capability to: - Irradiate large samples of proposed ATW

  10. 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

  11. Development of Integrated Evaluation System for Severe Accident Management

    International Nuclear Information System (INIS)

    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

  12. 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

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

    International Nuclear Information System (INIS)

    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

  14. 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.)

  15. 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.

  16. 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.

  17. Benchmark Analyses of the Shutdown Heat Removal Tests Performed in the EBR-II Reactor

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) established in 2012 a coordinated research project (CRP) on EBR-II Shutdown Heat Removal Tests (SHRT). The CRP aims to improve design and simulation capabilities in fast reactor neutronics, thermal hydraulics, plant dynamics and safety analyses through benchmark analysis of a protected and an unprotected loss-of-flow test from the EBR-II SHRT program. Activities will include core physics studies and thermal-hydraulics/safety assessments. Benchmark specifications provided by Argonne National Laboratory will be used by the CRP participants to develop a neutronic model of the EBR-II core to assess reactor power distribution, decay heat parameters, and reactivity feedback coefficients for subsequent transient scenario safety analysis. Investigations of thermal hydraulics characteristics and plant behaviour will focus on predicting natural convection cooling accurately by evaluating the reactor core flow and temperatures in comparison to experimental data. The ultimate evaluation will be in the prediction of the fission power history during an Anticipated Transient Without Scram (ATWS) by the coupled neutronic/thermal-hydraulic/structural models. This paper outlines the CRP and discusses the SHRT tests and the benchmark specifications. (author)

  18. 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.

  19. 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

  20. 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.)

  1. 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

  2. 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. On a...... microlevel, the CeO2 film epitaxially grows well on top of the Ni5W tape. A continuous, smooth, and crack-free morphology was observed on the CeO2 film and the fraction of low-angle grain boundaries (≤ 10°) is about 98 %. This process is a potential possibility for producing long-length textured CeO2/Ni5W...

  3. 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

  4. 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)

  5. 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

  6. Chemical separations schemes for partitioning and transmutation systems

    International Nuclear Information System (INIS)

    In the initial phase of the U.S. Accelerator Transmutation of Waste (ATW) program, a single-tier system was foreseen in which the transuranics and long-lived fission products (specifically, 99Tc and 129I) recovered from spent LWR oxide fuel would be sent directly to an accelerator-driven transmuter reactor [1]. Because the quantity of fuel to be processed annually was so large (almost 1,500 tons per year), an aqueous solvent extraction process was chosen for LWR fuel processing. Without the need to separate transuranics from one another for feed to the transmuter, it became appropriate to develop an advanced aqueous separations method that became known as UREX. The UREX process employs an added reagent (acetohydroxamic acid) that suppresses the extraction of plutonium and promotes the extraction of technetium together with uranium. Technetium can then be efficiently removed from the uranium; the recovered uranium, being highly decontaminated, can be disposed of as a low-level waste or stored in an unshielded facility for future use. Plutonium and the other transuranic elements, plus the remaining fission products, are directed to the liquid waste stream. This stream is calcined, converting the transuranics and fission products to their oxides. The resulting oxide powder, now representing only about four percent of the original mass of the spent fuel, is reduced to metallic form by means of a pyrometallurgical process. Subsequently, the transuranics are separated from the fission products in another pyro-metallurgical step involving molten salt electrorefining

  7. 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)

  8. 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.).

  9. Implementation of digital control and protection systems of China advanced research reactor

    International Nuclear Information System (INIS)

    China Advanced Research Reactor (CARR), a reactor of the 21st century with high performance is being constructed in China. The requirements of reliability and stability on the control and protection (c and p) system are the main points raised. Especially, with the development of digital technology, the c and p system of CARR is demanded to match the trend of digitization in the field of reactor control. The c and p system, including reactor protection system, reactor monitoring and control system, reactor power regulating system, and the mitigation system for ATWS (Anticipate Transient Without Scram), adopts digital technology, and the digital display screen will replace the analog panels in the main control room. The c and p system of CARR adopts redundant technology with 2 or 3 redundant channels to improve the system reliability. The 10/100 Mbps self-adaptive redundant optic fiber industry Ethernet ring network is used to interlink operator workstations, supervisor workstation, and I/O control stations. Commercial grade equipment with mature experience in industrial application are applied to the c and p system of CARR, which have high reliability, good interchangeability, and is easily purchased, the software-developing tools fully match the international industry standards. The realization of digital c and p system of CARR will promote the progress of digital control technology for reactors in China, and certainly become a technical basic platform for developing informational and intelligent reactors in China. (authors)

  10. 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

  11. 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

  12. 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.

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

    International Nuclear Information System (INIS)

    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. kinf, peaking factor, MTC, and control rod worth of new BAs were compared with those of the conventional BA

  14. Nuclear power plants 1985 - a world survey

    International Nuclear Information System (INIS)

    The Quick Statistics annually compiled by atw lists 355 nuclear power plants with 263,027 gross MWe in operation in 26 countries in late 1985. Another 163 units with 157801 MWe were under construction in 25 countries, and 75 units with 77,328 MWe were ordered in 18 countries. This adds up to a total of 593 units with an aggregate 498,156 MWe. In the course of 1985 30 nuclear generating units with 28,375 MWe were newly commissioned in 17 countries and 33 units in 14 countries started commercial operation. In the survey of electricity generation in 1985 for which no information was made available from the Eastern block countries, a total of 289 nuclear power plants were covered. In the year under review they generated an aggregate 1,270,854 GWh, which is 18.4% less than in the previous year. The highest nuclear generation again was recorded in the USA with 404,515 GWh, followed by France with 24,644 GWh. The KWG Grohnde Nuclear Power Station in the Federal Republic of Germany attained the maximum annual production figure of 1,477 GWh. The survey includes 10 tables indicating the generating performance of each nuclear power plant, the development of electricity generation in nuclear plants, and status of, nuclear power plants at the end of 1985 arranged by countries, types of reactors, and reactor manufacturers. (orig.)

  15. 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)

  16. The Influence of W Addition on Cube Textured Ni Substrates for YBCO Coated Conductor

    International Nuclear Information System (INIS)

    We fabricated cube-textured Ni and Ni-W alloy substrates for coated conductors and characterized the effects of W addition on microstructure, mechanical strength, and magnetic properties of the substrate. Pure Ni and Ni-(2, 3, 5at.%)W alloys were prepared by plasma arc melting, heavily cold rolled and then annealed at various temperatures of 600-1300 degree C. The texture was evaluated by pole-figure and orientation distribution function (ODF) analysis. Mechanical properties were investigated by micro Vickers hardness and tension test. Ferromagnetism of the substrate was measured by physical property measurement system (PPMS). It was observed that Ni-W substrates had sharp cube texture, and the full-width at half-maximums (FWHMs) of in-plane texture was 4.42 degree - 5.57 degree, which is better than that of pure Ni substrate. In addition cube texture of Ni-W substrates was retained at higher temperature up to 1300 degree C. Microstructural observation showed that the Ni-W substrates had fine grain size and higher mechanical properties than the pure Ni substrate. These improvements are probably due to strengthening mechanisms such as solid solution hardening and/or grain size strengthening. PPMS analysis showed that addition of W effectively reduced saturation magnetization in applied magnetic field and Curie temperature.

  17. Modeling a nuclear reactor for experimental purposes

    International Nuclear Information System (INIS)

    The Loss-of-Fluid Test (LOFT) Facility is a scale model of a commercial PWR and is as fully functional and operational as the generic commercial counterpart. LOFT was designed and built for experimental purposes as part of the overall NRC reactor safety research program. The purpose of LOFT is to assess the capability of reactor safety systems to perform their intended functions during occurrences of off-normal conditions in a commercial nuclear reactor. Off-normal conditions arising from large and small break loss-of-coolant accidents (LOCA), operational transients, and anticipated transients without scram (ATWS) were to be investigated. This paper describes the LOFT model of the generic PWR and summarizes the experiments that have been conducted in the context of the significant findings involving the complex transient thermal-hydraulics and the consequent effects on the commercial reactor analytical licensing techniques. Through these techniques the validity of the LOFT model as a scaled counterpart of the generic PWR is shown

  18. 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.

  19. A fast lead-cooled incinerator for economical actinide burning

    International Nuclear Information System (INIS)

    A fast lead-cooled modular reactor is proposed as an efficient incinerator of plutonium and minor actinides (MAs) for application to advanced fuel cycles devoted to transmutation. This actinide burner reactor (ABR) is loaded only with transuranics (TRU) in a fertile-free Zr-based metallic fuel to maximize the incineration rates and features (a) streaming fuel assemblies that enhance neutron leakage to achieve favorable neutronic feedbacks and (b) a double-entry control rod system that reduces reactivity perturbations during seismic events and flattens the axial power profile. A detailed neutronic analysis shows that the delayed neutron fraction is comparable to that of fast reactors and that negative reactivity feedbacks from lead voiding, Doppler, fuel thermal expansion and core radial expansion lead to safety characteristics similar to those of the Integral Fast Reactor. The ABR TRU destruction rate is the same as that of the ATW and fuel cycle cost analysis shows potential for economical accomplishment of the transmutation mission compared to other proposed actinide burning options. (author)

  20. Feedback of experience as a contribution to safety - an operator's duty

    International Nuclear Information System (INIS)

    Exchanges of information about events and reliability problems make important contributions to nuclear power plant safety. In an article in the last but one issue of atw, journalist Timm Kraegenow proposed that the operators of nuclear power plants model their attitude towards this subject, and their flows of information, on the example of civil aviation. In particular, vendors should be made the nodes of information exchange. However, in-depth comparison shows that the two areas, civil aviation and nuclear power, have fundamentally different structures. The safety of aircraft designs continues to be the vendor's responsibility throughout the service life of that aircraft, and it is the vendor who holds the type certification. When a deficit becomes apparent, the vendor is the partner to be contacted by the competent authority, and it is the vendor's duty to elaborate solutions. In the case of nuclear power plants, however, responsibility for safety after plant commissioning rests with the operator, i.e. the licensee. The licensee initiates improvements in safety and is also the addressee of instructions by the authorities. This fundamental difference in responsibility for safety is in the nature of things. It is bound to affect also exchanges of information as a module of safety. Although comparison with civil aviation is interesting, it becomes apparent in the end that the way in which flows of information are designed cannot be transferred to the nuclear sector. (orig.)

  1. 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)

  2. 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.

  3. No consensus for consensus's sake. Interview with DAtF-President Dr. Otto Majewski

    International Nuclear Information System (INIS)

    The German power industry is not prepared to make special sacrifices at the expense of its shareholders, staff and customers merely to cure the apparent internal problems of the German government coalition and help the Greens to regain lost votes. Unless the Federal Minister for the Environment rescinds his current cascade of conditions and requirements impeding the efforts on the part of operators to resume fuel transports, everything 'boils down to a court case'. That the French government, contrary to general expectations, had not given the green light this year for the construction of a first EPR plant was due primarily to internal political reasons, which had to be respected. It did not really matter whether that decision was taken in 2000 or 2001. The important thing was that the EPR was built at all, which was beyond any doubt. This is the gist of the statements by Dr. Otto Majewski, President of the German Atomic Forum and Chief Executive Officer of Bayernwerk AG, in the interview granted atw Editor-in-Chief Gerhard Kuebler and printed in this issue of 'Atomwirtschaft'. (orig.)

  4. 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

  5. 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.

  6. ECN contributions to ADTT '96

    International Nuclear Information System (INIS)

    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.)

  7. Failure analysis of the standby liquid control system for a boiling water reactor with fuzzy cognitive maps

    International Nuclear Information System (INIS)

    Highlights: → FCMs are proposed in order to determine failure modes in systems and equipment in BWRs. → A simplified model is compared with the fault tree analysis technique. → Five case scenarios are studied in order to test the performance of the method. → The proposed method shows consistency with the traditional fault tree technique. - Abstract: A fuzzy cognitive maps (FCM) application is proposed as a simple method to determine failure modes and effects of the standby liquid control system (SLC) during anticipated transient without scram (ATWS) in a boiling water reactor (BWR). The SLC has an important contribution to the total core damage frequency in a BWR. This is the first step in the development of an expert system that could involve many emergency systems of a BWR to simulate accident sequences, through the knowledge representation and reasoning with FCM designs in order to automate the decision making process. A simplified model of the SLC is analyzed with the fault tree analysis technique in order to compare this results with those obtained with the FCM and show consistency with the results, in order to see that both techniques show similar results even if the approaches are different.

  8. 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.

  9. 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.

  10. Analysis of the LaSalle Unit 2 Nuclear Power Plant: Risk Methods Integration and Evaluation Program (RMIEP)

    International Nuclear Information System (INIS)

    This volume presents the results of the initiating event and accident sequence delineation analyses of the LaSalle Unit II nuclear power plant performed as part of the Level III PRA being performed by Sandia National Laboratories for the Nuclear Regulatory Commission. The initiating event identification included a thorough review of extant data and a detailed plant specific search for special initiators. For the LaSalle analysis, the following initiating events were defined: eight general transients, ten special initiators, four LOCAs inside containment, one LOCA outside containment, and two interfacing LOCAs. Three accident sequence event trees were constructed: LOCA, transient, and ATWS. These trees were general in nature so that a tree represented all initiators of a particular type (i.e., the LOCA tree was constructed for evaluating small, medium, and large LOCAs simultaneously). The effects of the specific initiators on the systems and the different success criteria were handled by including the initiating events directly in the system fault trees. The accident sequence event trees were extended to include the evaluation of containment vulnerable sequences. These internal event accident sequence event trees were also used for the evaluation of the seismic, fire, and flood analyses

  11. Dependability Evaluation of Advanced Diverse Protection System

    International Nuclear Information System (INIS)

    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

  12. Quantitative modeling of digital reactor protection system using Markov state-transition model

    International Nuclear Information System (INIS)

    Recently, digital instrumentation and control systems have been increasingly installed for important safety functions in nuclear power plants such as the reactor protection system (RPS) and the actuation system of the engineered safety features. Since digital devices consist of not only electronic hardware but also software that can control microprocessors, the functions specific to digital equipment such as self-diagnostic functions have been becoming available. These functions were not realized with conventional electric components. On the other hand, it has been found that it is difficult to model the digital equipment reliability in probabilistic risk assessment (PRA) using conventional fault tree analysis technique. OECD/NEA CSNI Working Group of Risk Assessment (WGRisk) set up the task group DIGREL to develop the basis of reliability analysis method of the digital safety system and is now discussing about several issues including quantitative dynamic modeling. This paper shows that, taking account of the relationship among the RPS failures, demand after the initiating event, detection of the RPS fault by self-diagnostic or surveillance tests, repair of the RPS components and plant shutdown operation by the plant operators as a stochastic process, the anticipated transient without scram (ATWS) event can be modeled by the event logic fault tree and Markov state-transition diagrams assuming the hypothetical 1-out-of-2 digital RPS. (author)

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. 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)

  18. The coupled code TRAB-3D-SMABRE for 3D transient and accident analyses

    International Nuclear Information System (INIS)

    The three-dimensional TRAB-3D core dynamics code is being internally coupled to the thermal hydraulics system code SMABRE. The codes have previously been coupled with a parallel coupling scheme. VTT's reactor dynamics codes have performed well in all the situations that they have originally been designed for. The most important limitation of the present code models is their inability to handle coolant flow reversal in the core channel, a phenomenon that can be encountered in e.g. BWR ATWS cases or VVER power excursions. The new coupling of the two codes is realized on the level of each node of each channel in the core, with each fuel bundle described with its own channel. Necessary interfaces have been created, an improved version of SMABRE's thermal hydraulics solution method developed, and a steady state procedure developed. A satisfactorily working steady state solution has been achieved. The next step in the development will be testing of the transient calculation. Besides solving the flow reversal limitation of the present dynamics models, a successful coupling will allow expanding into more realistic modelling of an open core. (orig.)

  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. 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.

  1. The safety of the Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    The hallmarks of the approach to safety embodied in the Integral Fast Reactor (IFR) are: Large margins between the operating conditions and physical safety limits; reliance on passive processes to hold power production in balance with heat removal; and totally passive removal of decay heat, independent of the equipment and structures in the balance-of-plant. Should equipment in the balance-of-plant or control system fail, IFRs will passively regulate their own power so as to remain undamaged for all such initiators, even in the anticipated transient without scram (ATWS) scenarios. Decay heat is removed through a heat-transport path that operates at ambient pressure, is contained along with the reactor core in a double-walled top-entry tank of coolant, has large thermal inertia, is driven by natural convection, is completely independent of the balance-of-plant equipment, and is always in operation. The resulting level of safety exceeds the already acceptable levels attained in current-generation reactors now licensed and operating safety throughout the world. (author)

  2. Employing of RELAP-5 code for LBB (leak-before-break) deterministic analysis in the Ignalina NPP

    International Nuclear Information System (INIS)

    For coolant leak rate calculations through possible cracks in Ignalina NPP pipes, SQUIRT and RELAP5 thermal-hydraulic codes were used. SQUIRT is well known as a computer program that predicts the leakage rate for cracked pipes in NPP. As this program calculates only water leak rate, RELAP5 code model, that calculates water and steam leak rate, was created. The RELAP5 (reactor excursion and leak analysis program) can model transients in light water reactors (LWR) systems, such as loss of coolant, operational transients, anticipated transients without scram (ATWS), e.g. loss of feed-water, loss of offsite power, station blackout, and turbine trip. The RELAP5 code employs hydrodynamic, heat structure and reactor kinetics models with control and trip systems and time step control. If crack opening area (COA) is sufficiently big, then crack can be modeled as narrow pipe. The pipe cross-section shape can be assumed as round, rectangular and elliptical. This paper shows, that RELAP5 code could be employed for calculations of coolant discharge through cracks. For model verification a comparison of SQUIRT, RELAP5 and experimental results was performed. Analysis shows that calculated RELAP5 and SQUIRT results compare favourably with experimental data. It means, that RELAP5 model is suitable for calculations of leak through through-wall cracks in pipes. (A.C.)

  3. 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

  4. 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

  5. Description and calibration of a CT simulator for use in radiotherapy

    International Nuclear Information System (INIS)

    The characteristics and performance of a simulator-CT are studied. Image quality obtained with this system allows to use it for precise dosimetric calculation. Its geometric characteristics produce a tunnel diameter which is bigger than those of diagnostic CT. This allows to obtain images with the patient set in treatment position (Specially when inclined plane and other accessories are used), ensuring a correct treatment simulation. A Mecasero simulator with a Kermath tomographic system attached is studied. Image quality is evaluated from slices taken from an AAPM phantom. Obtained images are processed with ATW. a 16 bit image processing software developed in cooperation with the Optics Dept. University of Valencia. The obtained contrast allows to assure the correct localisation of internal structures such as lungs, aerial pathways, spinal chord or bones. Its spatial resolution allows to extract external contours accurately. Noise level and its origin is also discussed. The work shows that images obtained with a simulator CT properly calibrated have enough quality to be used for dosimetric calculations with reliability. Image quality loss is compensated by the geometric properties of the system, which allows taking images with the patient set in any treatment position. (Author) 9 refs

  6. Reliability enhancement of APR + diverse protection system regarding common cause failures

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Y. G.; Kim, Y. M.; Yim, H. S. [KEPCO Engineering and Construction Company, Inc., 1045 Daedeok-Daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Lee, S. J. [Chungnam National Univ., 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of)

    2012-07-01

    The Advanced Power Reactor Plus (APR +) nuclear power plant design has been developed on the basis of the APR1400 (Advanced Power Reactor 1400 MWe) to further enhance safety and economics. For the mitigation of Anticipated Transients Without Scram (ATWS) as well as Common Cause Failures (CCF) within the Plant Protection System (PPS) and the Emergency Safety Feature - Component Control System (ESF-CCS), several design improvement features have been implemented for the Diverse Protection System (DPS) of the APR + plant. As compared to the APR1400 DPS design, the APR + DPS has been designed to provide the Safety Injection Actuation Signal (SIAS) considering a large break LOCA accident concurrent with the CCF. Additionally several design improvement features, such as channel structure with redundant processing modules, and changes of system communication methods and auto-system test methods, are introduced to enhance the functional reliability of the DPS. Therefore, it is expected that the APR + DPS can provide an enhanced safety and reliability regarding possible CCF in the safety-grade I and C systems as well as the DPS itself. (authors)

  7. Reliability enhancement of APR + diverse protection system regarding common cause failures

    International Nuclear Information System (INIS)

    The Advanced Power Reactor Plus (APR +) nuclear power plant design has been developed on the basis of the APR1400 (Advanced Power Reactor 1400 MWe) to further enhance safety and economics. For the mitigation of Anticipated Transients Without Scram (ATWS) as well as Common Cause Failures (CCF) within the Plant Protection System (PPS) and the Emergency Safety Feature - Component Control System (ESF-CCS), several design improvement features have been implemented for the Diverse Protection System (DPS) of the APR + plant. As compared to the APR1400 DPS design, the APR + DPS has been designed to provide the Safety Injection Actuation Signal (SIAS) considering a large break LOCA accident concurrent with the CCF. Additionally several design improvement features, such as channel structure with redundant processing modules, and changes of system communication methods and auto-system test methods, are introduced to enhance the functional reliability of the DPS. Therefore, it is expected that the APR + DPS can provide an enhanced safety and reliability regarding possible CCF in the safety-grade I and C systems as well as the DPS itself. (authors)

  8. YBa{sub 2}Cu{sub 3}O{sub 7-x} films prepared by TFA-MOD method for coated conductor application

    Energy Technology Data Exchange (ETDEWEB)

    Rufoloni, A [ENEA, Frascati Research Center, Frascati, Rome (Italy); Augieri, A [ENEA, Frascati Research Center, Frascati, Rome (Italy); Celentano, G [ENEA, Frascati Research Center, Frascati, Rome (Italy); Galluzzi, V [ENEA, Frascati Research Center, Frascati, Rome (Italy); Mancini, A [ENEA, Frascati Research Center, Frascati, Rome (Italy); Vannozzi, A [ENEA, Frascati Research Center, Frascati, Rome (Italy); Petrisor, T [Technical University of Cluj Napoca (Romania); Ciontea, L [Technical University of Cluj Napoca (Romania); Boffa, V [Pirelli Labs, Milan (Italy); Gambardella, U [INFN-LFN, Frascati, Rome (Italy)

    2006-06-01

    The epitaxial growth of YBCO films both on (001) SrTiO{sub 3} (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 {omega}-scan of the YBCO films grown on (001) SrTiO{sub 3} single crystal substrate and on Pd/CeO{sub 2}/YSZ/CeO{sub 2} buffered biaxially textured Ni-5at%W (Ni-W) tapes has a full width at half maximum (FWHM) of 0.12{sup 0} and 3.4{sup 0}, respectively. The {phi}-scan of (113) peak of YBCO film grown on Ni-W substrate has FWHM of 6.1{sup 0}. The YBCO/STO film has a zero resistance critical temperature of Tc(R = 0) = 92 K and a critical current density Jc > 2 MA/cm{sup 2} at 77 K and in zero magnetic field.

  9. Metal propionate synthesis of epitaxial YBa{sub 2}Cu{sub 3}O{sub 7-x} films

    Energy Technology Data Exchange (ETDEWEB)

    Ciontea, L; Petrisor, T Jr; Petrisor, T [Technical University of Cluj, Str. C. Daicoviciu Nr. 15, 400020 Cluj-Napoca (Romania); Angrisani, A; Celentano, G; Rufoloni, A; Vannozzi, A; Augieri, A; Galuzzi, V; Mancini, A [ENEA Frascati, Via Enrico Fermi 45, 00044, Frascati, Roma (Italy)], E-mail: Lelia.Ciontea@chem.utcluj.ro

    2008-02-15

    A modified TFA-MOD method, using only barium trifluoroacetate, is presented. The yttrium and copper triflouroacetates were replaced by the alcoholic solutions of Cu and Y acetates dispersed in propionic acid. Fourier transformed infrared spectroscopy (FT-IR), thermal analyses (DTA/TG) coupled with mass spectrometry (MS) and X-ray diffraction analyses were used to study the decomposition of the precursor. The method permits the shortening of the pyrolysis time by a factor 4, with respect to conventional TFA-MOD method, due to the smaller amount of evolved hydrofluoric acid. Using this method 600 nm thick YBCO films were grown both on (100)SrTiO{sub 3} and on CeO{sub 2}/YSZ/CeO{sub 2}/Pd buffered Ni-5at.%W substrates. The as obtained films exhibit good morphological, structural and superconducting properties with T{sub c} (R=0) greater than 91K and with an out-of-plain texture of 0.24 deg. and 1.9 deg., respectively.

  10. Metal propionate synthesis of epitaxial YBa2Cu3O7-x films

    International Nuclear Information System (INIS)

    A modified TFA-MOD method, using only barium trifluoroacetate, is presented. The yttrium and copper triflouroacetates were replaced by the alcoholic solutions of Cu and Y acetates dispersed in propionic acid. Fourier transformed infrared spectroscopy (FT-IR), thermal analyses (DTA/TG) coupled with mass spectrometry (MS) and X-ray diffraction analyses were used to study the decomposition of the precursor. The method permits the shortening of the pyrolysis time by a factor 4, with respect to conventional TFA-MOD method, due to the smaller amount of evolved hydrofluoric acid. Using this method 600 nm thick YBCO films were grown both on (100)SrTiO3 and on CeO2/YSZ/CeO2/Pd buffered Ni-5at.%W substrates. The as obtained films exhibit good morphological, structural and superconducting properties with Tc (R=0) greater than 91K and with an out-of-plain texture of 0.24 deg. and 1.9 deg., respectively

  11. 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.

  12. 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)

  13. 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

  14. 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

  15. 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)

  16. Beta project: Some preliminary results of PSA level 2 for Unit 1 of Kalinin NPP

    International Nuclear Information System (INIS)

    The goals of this project are: transferring methodology of PSA level 1 and 2 performance to GAN and support of Russian organizations and development of probabilistic model of the Kalinin NPP-1 WWER-1000. SAPHIRE 7.19 and MELCOR 1.8.5 software tools are used. The main PSA level 1 results, for example: transients - 33.8%, primary LOCAs inside containment (sump clogging and no sump clogging) - 41.2% and 19.0%, total CDF (Core Damage Frequency) for all initial event groups - 2.43x10-5 (including of ATWS (Anticipated Transient Without Scram) 3.24x10-5) ect. are presented. An example of PSA level 1 event tree and Plant Damage State (PDS) matrix are given. Specific design features for severe accident progress and used nodalization for MELCORE like in-vessel modeling, MCL modeling, secondary side modeling, containment modeling are listed and illustrated. The main results of MELCORE analysis (fraction of I release and fraction of Cs release) are also presented

  17. 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)

  18. Markovian analysis of limiting conditions of operation for the reactor protection system

    International Nuclear Information System (INIS)

    The main conclusions of the point value calculations are supported by the uncertainty analysis. An uncertainty analysis was performed by the Monte Carlo sampling technique with the Markov model to assess the effect of the uncertainty in the data base on the results of point value calculations. A modified version of the SAMPLE program in WASH-1400 that can receive multiple outputs from MARELA was used. Comparison between two alternatives characterized by uncertainties usually requires a preference assessment (under certainty). In some special cases, namely, in the case of stochastic dominance, the comparison is straightforward. Stochastic dominance means that the likelihood that the attribute of interest will be less than a specific value is always larger for one alternative than for the other. Policy 1 stochastically dominates (is better than) Policy 2 on the ATWS core damage probability while Policy 2 stochastically dominates Policy 1 on the spurious scram core damage probability. As far as the total core damage probability is concerned, Policy 1 stochastically dominates Policy 2. However, Policy 2 stochastically dominates Policy 1 on the average reactor downtime

  19. The 50th anniversary of the N.S. Otto Hahn. When nuclear power said 'Ahoy'

    International Nuclear Information System (INIS)

    In September 1963 the construction of the only German nuclear ship to date, NS Otto Hahn, started in Kiel. For the 50th 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. 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

  1. Pulsed electron deposition (PED) of single buffer layer for 'low-cost' YBCO coated conductors

    International Nuclear Information System (INIS)

    The challenge for the commercialization of YBCO Coated Conductors (CC) is the development of a low cost manufacturing process to allow for a cheap, fast and continuous deposition of superconducting coatings with high electrical performance. We are currently investigating 2 ways to reduce the CC production costs: i) reducing the complexity of the CC architecture, by growing a single buffer layer based on doped CeO2, and ii) utilizing a new reel-to-reel apparatus for long length CC processing, equipped with a cheap and reliable deposition system (PED, Pulsed Electron Deposition). In this work we report on the successful continuous deposition of very thick (up to 700 nm) doped-CeO2 single buffer layers on biaxially textured Ni-5at%W substrates by PED. XRD patterns display complete orientation and very good texture quality of our samples (FWHM out-of-plane values of ∼ 6 deg.), over 20 cm length. Optical and electron microscopy show a dense and crack-free film surface and dielectric strength measurement confirms excellent insulating properties. Preliminary results indicate that the simplified single buffer layer structure could be a reliable solution for the reduction of HTS CC production costs

  2. 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.

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. Transients

    International Nuclear Information System (INIS)

    The lecture includes typical transients to be analyzed, the requirements put on computer codes and a description of the computer codes as well as results obtained with these codes. Transients analysis is necessary within the licensing of reactors, in risk evaluation and in other basic studies (e.g. on ATWS). The development of transient codes has been influenced by new requirements due to the extension in applications mentioned above. As examples the BWR model ALMOS and the PWR model ALMOD are described. These codes include a one-dimensional simulation of the neutron kinetics and the thermohydraulics in the coolant system. Also included in the simulation are all components of the control and safety systems, which are influencing the dynamic behaviour of the plant. Special emphasis is put on the problems of model verification (comparison with measurements). The transients behaviour of plants under extreme conditions, such as transients with a failure of the scram system, is described in detail. Examples are the loss of heat sink and the station black out for both a BWR and a PWR. (orig.)

  8. Application of Markov state-transition model to reliability analysis of 2-out-of-4 reactor trip system

    International Nuclear Information System (INIS)

    Recently, digital instrumentation and control systems have been increasingly installed to the important safety features in nuclear power plants such as the reactor trip system and the actuation system of the engineered safety features. On the other hand, it has been found that it is difficult to model the digital equipment reliability in probabilistic risk assessment (PRA) by the conventional Fault Tree Analysis technique. The OECD/NEA CSNI Working Group of Risk Assessment (WGRisk) set up the task group DIGREL to discuss several issues including quantitative dynamic modeling. This paper shows that, taking account of the relationship among the reactor trip system failures, demand after the initiating event, detection of the reactor trip system fault by self-diagnostic or surveillance tests, repair of the failed reactor trip system components and plant shutdown operation by the plant operators as a stochastic process, the anticipated transient without scram (ATWS) event can be modeled by the event logic fault tree and Markov state-transition diagrams in the case of the 2-out-of-4 digital reactor trip system. (author)

  9. Using proliferation risk as a design metric in the development of nuclear systems

    International Nuclear Information System (INIS)

    The necessity has arisen for newly proposed nuclear systems to be evaluated with regard to their potential aid to any proliferation. Thus, a mechanism is needed to introduce nonproliferation as a measure in the design phase of a new nuclear system. To accomplish this, a methodology for quantifying and measuring the proliferation risk of proposed system options is required. Such quantification has its difficulties due to inherent uncertainty, e.g. what is the probability that a quantity of material will be stolen in a given situation? Also, the lack of data on such occurrences makes the task of quantification nearly insurmountable. A systematic approach is necessary to estimate the proliferation risk. Currently, an advanced nuclear power system, the Accelerator Transmutation of Waste (ATW) program has been initiated to develop a system that will concurrently generate electricity while destroying long-lived radioactive isotopes. Therefore, because of the issues noted above, an effort to introduce proliferation risk into the design phase has been started. The purpose of this paper is to review previous work in quantification of proliferation risk in an effort to develop the proper basis for the current work. It should be noted that while proliferation on a national level has been studied extensively, efforts to quantify proliferation risk of individual nuclear systems or processes have been limited. Consequently, the available literature base is relatively sparse. (author)

  10. 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.

  11. 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.

  12. Nuclear power plants in Europe 1995. Report about operation, construction, and planning in 18 European countries

    International Nuclear Information System (INIS)

    Report about Operation, Construction, and Planning in 18 European Countries Eighteen European countries operate and build, respectively, nuclear power plants in 1995. The ''Nuclear Power Plants in Europe 1995'' atw report singles out the main events and lines of development. As per August 1995, 214 (1994: 215) nuclear generating units (which means power reactors for the purposes of this report) with an aggregate 177,010 (176,322) MWe installed gross capacity are in operation in seventeen countries, and 26 (30) units with 24,786 (28,086) MWe are under construction in seven countries. This adds up to a total of 240 (245) nuclear generating units with an aggregate 201,796 (204,408) MWe. In the nuclear power plants in Europe, some 1048 TWh of nuclear power was converted into electric power in 1994; 792 TWh of this aggregate was converted in 137 units in the European Union (EU). In the EU the share of nuclear power in the public supply of electricity was 36%. Lithuania, with 77%, has the highest share of nuclear power in Europe, followed by France with 75% and Belgium with 56%. The lowest percentage, only 5%, is recorded in the Netherlands. As a consequence of electricity imports, nuclear power holds considerable shares in the public electricity supply also of countries in which no nuclear power plants are operated, such as Italy or Austria. (orig.)

  13. New nuclear power plants in Europe 1986. Pt. 2

    International Nuclear Information System (INIS)

    The atw report on new nuclear power plants in Europe contains both a survey of the Federal Republic of Germany, which was published in the April 1986 issue, and an overview of the nuclear power plant situation in 26 European countries including the Soviet Union and six other CMEA countries. Also this year's review includes specific status reports, complete with technical information, about all nuclear generating units under construction, in the project and concrete planning phases. Introductory summaries describe the plants in operation in each country and their 1985 electricity generation. A general introduction provides an outlook on developments in Western and Eastern Europe. The total number of nuclear power plants now in operation and under construction in Europe is 302 units with an aggregate gross 224,846 MWe. Of these, 183 units are located in Western Europe, 119 in Eastern Europe. The units currently in operation of this total are 211 with 138,265 MWe in sixteen European countries; of these, 145 units with 99,396 MWe are run in eleven West European countries, 66 units with an aggregate 38,869 MWe in five East European countries. Of the 91 nuclear generating units at present under construction with an aggregate 86,581 MWe in fourteen European countries, 38 units with 40,551 MWe are located in six West European, 53 units with 46,030 MWe in seven East European countries. (orig.)

  14. 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

  15. 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

  16. Reports within the area of nuclear power plant instrumentation: Part 1: Laboratory test of analogue and digital instrument components. Part 2: Dynamic deviations in reactor pressure water level signals caused by sensing lines

    Energy Technology Data Exchange (ETDEWEB)

    Bergdahl, Bengt-Goeran [GSE Power Systems AB, Nykoeping (Sweden)

    2004-11-01

    type TDE220. The transmitters exhibited deviating dynamics during ordinary sensor tests. The laboratory test confirms the observed deviation in comparison with transmitters of other types. The construction with Bourdon tube is judged to be the reason for the deviations. The report also presents results from trouble shooting with steam pressure transmitters at KKM (Kernkraftwerk Muehleberg m Switzerland). It was possible to identify the intermittent sensor error with the aid of controlled pressure changes. Service of the transmitter pointed out a crack on the electronic filter unit. This was judged to be the reason for the intermittent signal interrupts. Finally, two possibilities used at KKM to investigate the dynamics of temperature sensors are described. Both methods are based on artificial cooling of the sensor. One of them is applied during power operation of the plant and the other during outage. (The present report is a translation of the Swedish language report SKI-R--03-07, published Dec 2002)

  17. Reports within the area of nuclear power plant instrumentation: Part 1: Laboratory test of analogue and digital instrument components. Part 2: Dynamic deviations in reactor pressure water level signals caused by sensing lines

    International Nuclear Information System (INIS)

    type TDE220. The transmitters exhibited deviating dynamics during ordinary sensor tests. The laboratory test confirms the observed deviation in comparison with transmitters of other types. The construction with Bourdon tube is judged to be the reason for the deviations. The report also presents results from trouble shooting with steam pressure transmitters at KKM (Kernkraftwerk Muehleberg m Switzerland). It was possible to identify the intermittent sensor error with the aid of controlled pressure changes. Service of the transmitter pointed out a crack on the electronic filter unit. This was judged to be the reason for the intermittent signal interrupts. Finally, two possibilities used at KKM to investigate the dynamics of temperature sensors are described. Both methods are based on artificial cooling of the sensor. One of them is applied during power operation of the plant and the other during outage. (The present report is a translation of the Swedish language report SKI-R--03-07, published Dec 2002)

  18. 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

  19. 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)

  20. 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

  1. Systems behaviour and operational aspects of safety, research programme in 1990-1994

    International Nuclear Information System (INIS)

    Studies in the field of 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. Assessment of complex transient and accident situations provides up-to-date methods to evaluate the whole spectrum of transients and accidents relevant for licensing, plant-specific PSA and accident management. PACTEL is a new facility (scaled 1:305) for the Loviisa reactors) for thermal hydraulic studies. Advanced computer codes from the US Nuclear Regulatory Commission's severe accident research program are implemented, modified as needed and applied to level-2 PSAs and 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 new APROS process simulation environment. New technology in nuclear power plant automation is particularly relevant for future plants but new technology is also applied at the present plants, particularly in case of life extension. Distributed digital automation systems will be assessed and validated. Computerized operator support systems are studied in cooperation with the OECD Halden Project. PRA/PSA methods are increasingly useful in management of nuclear power plant safety and performance. Prototype of a comprehensive plant safety information system is developed by combining living PSA, plant status monitoring and safety indicators into a kind of 'dynamic PSA'. The use of PSA methods during new plant design is also studied

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 238U and 232Th 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

  4. Development of regulatory technology for thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

  5. 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)

  6. 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 (ARI), 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 RAMONA-4B code has been developed to simulate the normal operation, reactivity transients, and to address the instability issues for SBWR. The code has a three-dimensional neutron kinetics coupled to multiple parallel-channel thermal-hydraulics. The two-phase thermal hydraulics is based on a nonhomogeneous nonequilibrium drift-flux formulation. It employs an explicit integration to solve all state equations (except for neutron kinetics) in order to predict the instability without numerical damping. The objective of this project is to 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

  7. Uncertainty analysis for hot channel

    International Nuclear Information System (INIS)

    The fulfillment of the safety analysis acceptance criteria is usually evaluated by separate hot channel calculations using the results of neutronic or/and thermo hydraulic system calculations. In case of an ATWS event (inadvertent withdrawal of control assembly), according to the analysis, a number of fuel rods are experiencing DNB for a longer time and must be regarded as failed. Their number must be determined for a further evaluation of the radiological consequences. In the deterministic approach, the global power history must be multiplied by different hot channel factors (kx) taking into account the radial power peaking factors for each fuel pin. If DNB occurs it is necessary to perform a few number of hot channel calculations to determine the limiting kx leading just to DNB and fuel failure (the conservative DNBR limit is 1.33). Knowing the pin power distribution from the core design calculation, the number of failed fuel pins can be calculated. The above procedure can be performed by conservative assumptions (e.g. conservative input parameters in the hot channel calculations), as well. In case of hot channel uncertainty analysis, the relevant input parameters (k x, mass flow, inlet temperature of the coolant, pin average burnup, initial gap size, selection of power history influencing the gap conductance value) of hot channel calculations and the DNBR limit are varied considering the respective uncertainties. An uncertainty analysis methodology was elaborated combining the response surface method with the one sided tolerance limit method of Wilks. The results of deterministic and uncertainty hot channel calculations are compared regarding to the number of failed fuel rods, max. temperature of the clad surface and max. temperature of the fuel (Authors)

  8. 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.

  9. 2002 annual conference on nuclear technology. Nuclear power reassessed

    International Nuclear Information System (INIS)

    The Liederhalle conference center in Stuttgart was the venue of this year's ANNUAL MEETING ON NUCLEAR TECHNOLOGY on May 14-16, 2002. This location in a German federal state with an approximately 60% share of nuclear power in electricity generation, and situated close to this year's partner country, France, whose more than 70% of electricity from nuclear power are unparalleled in the European Union, marked the importance assigned to nuclear power in the future scenario. Accordingly, many presentations made it clear that an unbiased assessment of nuclear power is indispensable in view of the urgent problems to be managed in the future. The organizers, the Deutsches Atomforum (DAtF) e.V. and the Kerntechnische Gesellschaft (KTG) e.V., welcomed approximately 1 000 participants from nearly twenty countries. The proven, traditional structure of the program of the three-day meeting again featured plenary sessions on the first day, and specialized sessions and technical sessions, poster sessions and special events on the other two days. France, the partner country, was represented in contributions on the plenary day and in the technical program. The conference was accompanied by a specialized exhibition with meeting points of the manufacturing, supplies and service industries which impressively underscored the importance of this industry for Germany as an industrialized high-tech country with promising prospects for the future, and for Europe as a whole. The plenary day and the specialized sessions are covered in this summary article. The results of the technical sessions as well as detailed reports about specialized sessions can be found elsewhere in this issue and in the next issue of atw, 8-9/2002. (orig.)

  10. 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.)

  11. 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.)

  12. 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

  13. Summary of session W2

    International Nuclear Information System (INIS)

    This paper reports on the development and validation of the SASSYS-1 code which is being used to analyze the transient behavior of the U.S. innovative designs, SAFR and PRISM, as well as the future metal fueled core for FFTF. The description emphasized the high degree of flexibility of SASSYS-1 and its capability to do detailed core calculations needed to accurately model reactivity feedbacks. Validation work based on calculations of EBR-II SHRT (shutdown neat removal) tests and FFTF transient experiments was presented. The paper presents a discussion of the modifications to the SSC code required to model the ATWS events in SAFR and PRISM, along with results of calculations done with the modified SSC. The probability that passive shutdown could be defeated and severe core damage sustained has been investigated, and some results are presented in this paper. It was stated that simple design choices can keep risk almost arbitrarily low, and that the leading uncertainties will be reduced through future R ampersand D and demonstration testing. Uncertainties exist due to uncertainties in feedback coefficients and in the models and codes used to do the calculations. Large margin and relatively small uncertainties lead to low probabilities (less than 10-3) of defeat of passive shutdown by exceeding temperature limits. The given value is considered to be conservative to allow for uncertainties in knowledge of all relevant accident sequences at this stage in design and analysis. This paper also deals with experimental work on decay heat removal in PRISM and SAFR. Experiments on air-side heat transfer in the RVACS/RACS passive decay heat removal path, was reported. A data correlation for use in predicting heat transfer has been discussed for the RVACS conditions

  14. The combined thermohydraulics-neutronics code TRAB-SMABRE for 3D plant transient and accident analyses

    International Nuclear Information System (INIS)

    TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable part in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but the rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. (authors)

  15. The trab-smabre for 3D plant transient and accident analyses

    International Nuclear Information System (INIS)

    VTT's three-dimensional TRAB-3D core dynamics code and SMABRE thermal hydraulic system code have been coupled together using an internal coupling scheme in order to increase flexibility in the thermal hydraulics modeling of the core calculation. VTT's reactor dynamics codes have performed well in all the situations that they have originally been designed for. The most important limitation of the present code models is their inability to handle coolant flow reversal in the core channel, a phenomenon that can be encountered in e.g. BWR ATWS cases or VVER power excursions. The new coupling of the two codes is realized on the level of each node of each channel in the core, with each fuel bundle described with its own channel. TRAB-3D performs only the neutronics calculation, SMABRE takes care of the hydraulics calculation of the whole cooling circuit including the reactor core, while heat transfer calculation can be carried out optionally by either code. The codes have earlier been coupled using a parallel coupling scheme. Several modifications were necessary in SMABRE, concerning modeling of hydraulics, heat transfer, geometry and the matrix solution. The accuracy of the steady state calculation in the coupled code has been improved to a level suitable for both PWR and BWR calculations, as compared against the SIMULATE and reference TRAB-3D codes. The BWR dynamics calculations are being tested with single disturbances, such as control rod movements, pump coast-down etc. Besides allowing modeling of reversed flow in the core, the internally coupled code will make future modeling of in-core cross-flows or even 3D flow in a PWR (such as EPR) open core geometry possible, e.g. by using the porous medium approach. (authors)

  16. Uncertainty assessment for accelerator-driven systems

    International Nuclear Information System (INIS)

    The concept of a subcritical system driven by an external source of neutrons provided by an accelerator ADS (Accelerator Driver System) has been recently revived and is becoming more popular in the world technical community with active programs in Europe, Russia, Japan, and the U.S. A general consensus has been reached in adopting for the subcritical component a fast spectrum liquid metal cooled configuration. Both a lead-bismuth eutectic, sodium and gas are being considered as a coolant; each has advantages and disadvantages. The major expected advantage is that subcriticality avoids reactivity induced transients. The potentially large subcriticality margin also should allow for the introduction of very significant quantities of waste products (minor Actinides and Fission Products) which negatively impact the safety characteristics of standard cores. In the U.S. these arguments are the basis for the development of the Accelerator Transmutation of Waste (ATW), which has significant potential in reducing nuclear waste levels. Up to now, neutronic calculations have not attached uncertainties on the values of the main nuclear integral parameters that characterize the system. Many of these parameters (e.g., degree of subcriticality) are crucial to demonstrate the validity and feasibility of this concept. In this paper we will consider uncertainties related to nuclear data only. The present knowledge of the cross sections of many isotopes that are not usually utilized in existing reactors (like Bi, Pb-207, Pb-208, and also Minor Actinides and Fission Products) suggests that uncertainties in the integral parameters will be significantly larger than for conventional reactor systems, and this raises concerns on the neutronic performance of those systems

  17. 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.)

  18. 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

  19. University Programs of the U.S. Department of Energy Advanced Accelerator Applications Program

    International Nuclear Information System (INIS)

    The Advanced Accelerator Applications (AAA) Program was initiated in fiscal year 2001 (FY-01) by the U.S. Congress, the U.S. Department of Energy (DOE), and the Los Alamos National Laboratory (LANL) in partnership with other national laboratories. The primary goal of this program is to investigate the feasibility of transmutation of nuclear waste. An Accelerator-Driven Test Facility (ADTF), which may be built during the first decade of the 21. Century, is a major component of this effort. The ADTF would include a large, state-of-the-art charged-particle accelerator, proton-neutron target systems, and accelerator-driven R and D systems. This new facility and its underlying science and technology will require a large cadre of educated scientists and trained technicians. In addition, other applications of nuclear science and engineering (e.g., proliferation monitoring and defense, nuclear medicine, safety regulation, industrial processes, and many others) require increased academic and national infrastructure and student populations. Thus, the AAA Program Office has begun a multi-year program to involve university faculty and students in various phases of the Project to support the infrastructure requirements of nuclear energy, science and technology fields as well as the special needs of the DOE transmutation program. In this paper we describe university programs that have supported, are supporting, and will support the R and D necessary for the AAA Project. Previous work included research for the Accelerator Transmutation of Waste (ATW) project, current (FY-01) programs include graduate fellowships and research for the AAA Project, and it is expected that future programs will expand and add to the existing programs. (authors)

  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. 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

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. 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).

  7. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    Energy Technology Data Exchange (ETDEWEB)

    Papukchiev, Angel [GRS mbH Forschungsinstitute, Garching (Germany); Schaefer, Anselm [ISaR GmbH, Garching (Germany)

    2008-07-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  8. Thermal-hydraulic study of a high conversion light water reactor

    International Nuclear Information System (INIS)

    A high conversion light water reactor (HCLWR) has been developed at JAERI to improve fuel utilization. The fuel assemblies of the HCLWR have a triangular tight rod lattice in order to attain a high conversion ratio by reducing the moderator-to-fuel volume ratio. The tight lattice core leads to some thermal-hydraulic problems peculiar to the HCLWR. Therefore, experiments on steady-state and transient critical heat flux (CHF), turbulent mixing, bundle pressure drop, fuel vibration induced by fluid flow and reflood cooling have been performed to obtain data base and develop evaluation methods for the HCLWR. The evaluation methods were applied to a JAERI-proposed double-flat-core type HCLWR. Under the operational condition, the minimum allowable DNBR criterion is satisfied and the bundle pressure drop and flow-induced vibration and displacement of the fuel rod are within the design limit of a current LWR. Safety analyses under vital conditions such as the large break loss-of-coolant accident (LOCA), small break LOCA, pump trip accident, locked rotor accident, anticipated transient without scram (ATWS) induced by station blackout and control rod cluster ejection accident have been performed with a best-estimate REFLA/TRAC code. The results showed that the present HCLWR meets the current safety criteria for a LWR. The design features such as the large water inventory in the upper plenum, short core length, low axial peaking factor and negative void and coolant temperature reactivity coefficients contribute to the thermal-hydraulic feasibility of the double-flat-core type HCLWR. (author)

  9. 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)

  10. 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

  11. 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.

  12. 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.

  13. 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

  14. 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

  15. 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)

  16. 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

  17. 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)

  18. 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.

  19. 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

  20. 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