Untersuchung zu Primaerfreisetzungen in das Reaktorschutzgebaeude von modularen Hochtemperaturreaktoren. Teilvorhaben: Rechnungen zu spezifischen Belastungspfaden Ergebnisbericht zur Basisrechnung 'Bruch der Frischdampfleitung'

SIGLECopy held by FIZ Karlsruhe; available from UB/TIB Hannover / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekDEGerman

Untersuchung zu Primaerfreisetzungen in das Reaktorschutzgebaeude von modularen Hochtemperaturreaktoren. Teilvorhaben: Rechnungen zu spezifischen Belastungspfaden Ergebnisbericht zur Basisrechnung 'Bruch aeussere DAL'

SIGLECopy held by FIZ Karlsruhe; available from UB/TIB Hannover / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekDEGerman

Belastung und Stoerfallverhalten eines 'Innencontainments' an einem modifizierten HTR-Modul. Fachbericht 2.1

SIGLEAvailable from TIB Hannover: FR 3407(2.1) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekDEGerman

Moeglichkeiten zur Wasserstoffbeseitigung Phase III: Ausgewaehlte Massnahmen zur Wasserstoffbeherrschung bei hypothetischen Unfaellen im Druckwasserreaktor

SIGLEAvailable from GRM Werbeberatung - Werbemittlung - PR, Eggenstein-Leopoldshafen (DE) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekDEGerman

Ermittlung der Temperaturen und Waermestroeme im Reaktorgebaeude bei Ausfall des Zellenkuehlsystems

SIGLETIB Hannover: FR 3407(Schl) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekDEGerman

Quantitative probabilistic safety criteria for licensing and operation of nuclear power plants Comparison of the international status and development

Internationally, probabilistic safety analyses (PSA) represent the state of the art not only in the licensing process for new nuclear power plants (NPPs), but increasingly also for evaluating the safety level of older NPP, e.g. as part of periodic safety reviews (PSR). Primarily, the qualitative results of PSA are of interest for assessing whether the safety features of NPPs are designed in a well-balanced way and for identifying plant vulnerabilities to severe accidents and effective measures to improve the plant safety level through refurbishment. Quantitative probabilistic safety criteria (QPS) have not yet reached the same level of acceptance. In this report, available quantitative probabilistic safety criteria for licensing and supervision of nuclear facilities have been compiled for a number of countries, primarily for OECD member states. This report is to identify the emerging state of the art and to point out contentious aspects in introducing QPS. A general trend can be identified: to adopt the recommendations from IAEA Safety Series No. 75 (INSAG-3) and INSAG-12 respectively, either directly or after adaptation to some specific national format: older NPPs should not have expected core damage frequencies (CDF) higher than 10"-"4/a; for licensing of new NPP, their CDF should be below 10"-"5/a according to general consensus; and the expected frequency of large early releases should be below 10"-"6/a for new plants. A large number of countries surveyed have taken these INSAG values directly, some have decided to opt for more demanding criteria; there is no country in the sample which has decided for less stringent criteria than the INSAG recommendations. Most of the countries consider those criteria as safety targets rather than as sharply defined boundary values, again following the INSAG recommendations. The Netherlands and the United Kingdom are exceptions, they require demonstration of compliance with legally binding QPS in the licensing procedure. (orig.)Available from TIB Hannover: RR 1571(03/03) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekSIGLEDEGerman

Untersuchung zu Primaerfreisetzungen in das Reaktorschutzgebaeude von modularen Hochtemperaturreaktoren. Teilvorhaben: Rechnungen zu spezifischen Belastungspfaden Abschlussbericht

SIGLEAvailable from TIB Hannover: RR2035(40.050-1) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekDEGerman

Untersuchungen der Spaltprodukt-Rueckhaltung durch HTR-Reaktorschutzgebaeude nach dem Vented-Confinement-Konzept. Fachbericht 2.2

SIGLEAvailable from TIB Hannover: FR 3407(2.2) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekDEGerman

Systemstudie zur Entwicklung eines Core-Catcher-Konzepts fuer Leichtwasserreaktoren Abschlussbericht/Hauptband

SIGLEDEGerman

Untersuchungen der Spaltprodukt-Rueckhaltung durch HTR-Reaktorschutzgebaeude nach dem Vented-Confinement-Konzept. Fachbericht 1.4 Thermodynamik und Stroemungsvorgaenge im RSG

SIGLETIB: FR 3407(1.4) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekDEGerman