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  • 1.
    Anglart, Henryk
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Laurien, Eckart
    Schulenberg, Thomas
    International students workshop on innovative light water reactors2008In: ATW. Internationale Zeitschrift für Kernenergie, ISSN 1431-5254, Vol. 53, no 6, p. 380-386Article in journal (Refereed)
    Abstract [en]

    Nuclear reactor design is still one of the most fascinating subjects of mechanical engineering. Thirty students from 10 worldwide nations demonstrated this impressively in a recent workshop on supercritical water cooled reactors of the 4(th) generation, held from March 31 to April 3, 2008, in Karlsruhe, Germany, hosted by the Karlsruhe Institute of Technology. Bachelor and master students as well as young scientists working on their doctorate presented their own particular contribution to design and analyses of innovative reactor components, including its safety systems and other plant design. Their presentations were accompanied by lectures of leading scientists working in the European project of the "High Performance Light Water Reactor" which is sponsored by the if European Commission as part of its 6(th) Framework Programme. The workshop is an initiative of the Generation IV International Forum.

  • 2. Hein, Hieronymus
    et al.
    Keim, Elisabeth
    Bechler, Eduoard
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Ganswind, Jens
    Knobel, René
    König, Günter
    Barriero, Pablo
    Widera, Martin
    de Jong, André
    CARINA: A program for experimental investigation of the irradiation behaviour of German Reactor Pressure Vessel materials2013In: ATW. Internationale Zeitschrift für Kernenergie, ISSN 1431-5254, Vol. 58, no 5Article in journal (Refereed)
    Abstract [en]

    The proof of a sufficient safety margin against brittle fracture of the reactor pressure vessel (RPV) is an important part of the operational safety of nuclear power plants. The RPV safety assessment procedure applicable in Germany is described in KTA 3201.2 of the Nuclear Safety Standard Commission (KTA). This deterministic assessment concept is based on the comparison of load curves with the material resistance curve in terms of fracture toughness. The fracture toughness curve can be determined either indirectly according to the RT-(NDT) concept based on Charpy tests or directly according to the more appropriate RTT0 approach based on Master Curve analysis of fracture toughness tests, respectively. In the recently completed research project CARINA the data base for pre-irradiated original RPV steels of German PWR construction lines was extended by comprehensive fracture toughness testing. The data obtained up to neutron fluences of 7.67 x 10(19) n/cm(2) (E > 1 MeV) are analysed and discussed particularly in terms of Master Curve applications. The experimental results show that optimized RPV manufacturing specifications and reactor designs are advantageous for a long-term plant operation in comparison to less optimized materials with lower toughness and to reactor designs with substantial higher neutron irradiation. With the obtained data, experiences and insights an essential contribution was also made to the integration of the Master Curve concept in German safety standards.

  • 3.
    Reisch, Frigyes
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    High Pressure Boiling Water Reactor: HP-BWR2010In: ATW. Internationale Zeitschrift für Kernenergie, ISSN 1431-5254, Vol. 55, no 2, p. 107-+Article in journal (Refereed)
    Abstract [en]

    Some 400 Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are;

    1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used

    2. Environment friendly; -Improved thermal efficiency, feeding the turbine with similar to 340 degrees C (15 MPa) steam instead of similar to 285 degrees C (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced

    3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment Well proved simple dry containment or wet containment can be used.

  • 4.
    Reisch, Frigyes
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Neutron kinetics of the Chernobyl accident2006In: ATW. Internationale Zeitschrift für Kernenergie, ISSN 1431-5254, Vol. 51, no 4, p. 254-255Article in journal (Refereed)
    Abstract [en]

    The classical reactor kinetic equations with six groups of delayed neutrons are not solved analytically. Here they are solved numerically with MATLAB and applied to the Chernobyl accident. The results are presented graphically. Now, 20 years after the accident it is important for today's and tomorrow's generations of nuclear engineers to learn not to design reactors with runaway characteristics which can cause an avalanche like power excursion . The Chernobyl type of reactor has a positive void coefficient, which means that when a part of the water is replaced by steam the power will increase. At the Chernobyl experiment the steam content in the coolant channels increased suddenly causing a catastrophic power excursion. The presented analyses gives details about the importance of the magnitude of the void coefficient. Also the delayed neutrons behaviour is described.

  • 5.
    Reisch, Frigyes
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Smart grids at a nuclear power plant: ensuring power supply for priority consumers with islanding controlled by symmetrical components2010In: ATW. Internationale Zeitschrift für Kernenergie, ISSN 1431-5254, Vol. 55, no 11, p. 694-696Article in journal (Refereed)
    Abstract [en]

    In a 3-phase system, when a short circuit between 2 lines or a single or 2 lines to earth occurs, 3 types of currents are immediately produced: positive, negative and 0 sequence. Using these signals, less priority consumers can be disconnected from the grid and the power of the nuclear reactor, which supplies all consumers, can be diverted to supply priority consumers, although with only a couple of cycle times due to dwindling amplitude. This is an example of an island operation with a nuclear power reactor and prioritised consumers connected via a smart grid.

  • 6. Roudén, Jenny
    et al.
    Hein, Hieronymus
    AREVA Gmbh Germany.
    May, Johannes
    AREVA Gmbh Germany.
    Planman, Tapio
    VTT Finland.
    Todeshini, Patrick
    EdF France.
    Brumowski, Milan
    UJV Czech Republic.
    Ballesteros, Antonio
    JRC Petten, Holland.
    Gillemot, Ferenc
    MTA Hungary.
    Chaouadi, Rachid
    SCK-CEN, Belgium.
    Efsing, Pål
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Altstadt, Eberhard
    Forschung Center Rossendorff, Germany.
    Towards Safe Long-Term Operation of Reactor Pressure Vessels2015In: ATW. Internationale Zeitschrift für Kernenergie, ISSN 1431-5254, Vol. 60, no 5, p. 287-293Article in journal (Refereed)
    Abstract [en]

    This publication summarizes the long term operation (LTO) conditions on European NPPs and provides recommendations on reactor pressure vessel (RPV) irradiation surveillance based on the work preformed in the work package 7 "Surveillance guidelines" of the LONGLIFE international project. The LONGLIFE project "treatment of long term irradiation embrittlement effects in RPV safety assessment" was 50% funded by the Euratom 7th framework programme of the European commision. The project coordinated by the Helmholtz-centrum Dresden Rossendorf successfully finalized in 2014.

  • 7. Willschutz, H G
    et al.
    Altstadt, E
    Weiss, F P
    Sehgal, Balraj
    KTH, Superseded Departments, Physics.
    Insights from the FOREVER-Programme and the accompanying finite element calculations2004In: ATW. Internationale Zeitschrift für Kernenergie, ISSN 1431-5254, Vol. 49, no 5, p. 345-+Article in journal (Refereed)
    Abstract [en]

    To obtain an improved understanding and knowledge of the melt pool convection thermal loads, the vessel creep, and vessel failure modes occurring during the late phase of a core melt down accident the FOREVER-experiments have been performed at the Royal Institute of Technology, Stockholm. These experiments simulated the behaviour of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurised vessel scenario. A Finite Element model was developed simulating melt pool convection and calculating the temperature field within the melt pool and within the vessel wall. After performing successful pre- and post-test calculations, a discussion about the lessons learned from the experiments and the analyses led to the idea of providing a vessel support and an external water-flooding device.

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