Qucosa – Hemholtz-Zentrum Dresden-Rossendorf
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The Code DYN3DR for Steady-State and Transient Analyses of Light Water Reactor Cores with Rectangular Geometry
Get PDFThe code DYN3D/M2 was developed for steady-state and transient analysis of reactor cores with hexagonal fuel assemblies. The new version DYN3DR contains a method for neutron kinetics solving the two group neutron diffusion equation by a nodal method for cartesian geometry. The thermal-hydraulic model FLOCAL simulating the two phase flow of coolant and the fuel rod hehaviour is used in the two versions. The fundamentals of the neutron kinetics are described. The accuracy of the code is demonstated by comparisons with the results of rod ejection benchmarks for PWR with rectangular fuel assemblies. The developed algorithm of neutron kinetics are suitable for using parallel processing. The speedup of neutronic calculation is demonstrated for a steady state solution of diffusion equation
Dilepton production in a chemically equilibrating, expanding and hadronizing quark-gluon plasma
Get PDFErmittlung der Neutronendosis von bestrahlten WWER-Reaktordruckbehältermaterialien
Get PDFIn this report is described the theoretical and experimental determination of all needed parameters for the neutron exposure of the different specimens that were irradiated in the Rheinsberg reactor within the period from 1984 until 1988 to obtain data for neutron embrittlement studies. The methodical approach is represented, possible sources of errors are discussed and all the needed results are collected. The work can be divided into a pure calculation part, into the description of the extraction of the monitors and their gamma spectrometric analysis and into the discussion of the spectrum adjustment procedure which combines experimental and theoretical results. Therefore, this representation deals not only with this special task but gives also a survey about the methodical approach and the state of art in Rossendorf for the general problem of determination of neutron fluences, because the developed methods are generally applicable and not limited to the special problem of neutron embrittlement. Different problems with respect of neutron fluences exist also in Germany. The calculation of fluences is based on an accurate description of the history of the reactor for all irradiation periods, i.e. the time and space dependence of burnup, power and fisson sources. This information was given for a time and space grid of all fuel elements. For the calculation of the fluences were not used time dependent fluxes, but at first integral sources of fission neutrons are calculated for different fissionable isotopes. Using these given integral source distributions the fluences were calculated and the results from the different fissionable source were composed. For the improvement of the comparison between experimental and theoretical results a special method was used calculating special integral sources for the given detector, because each detector according to its decay notes another integral source resp. flux distribution, which has to be theoretically corrected. The basis of all transport calculations was the Monte Carlo method in a special problem adjusted kind. Special procedures were developed and successfully applicated for the reduction of statistical errors. Therefore, also for single specimens results with small statistical errors were obtained. It is remarkable that this accurate method which allows a realistic 3-dimensional description of the system reactor-surroundings could be applied with reasonable calculation times for the great number of needed calculations
Bestimmung bruchmechanischer Kennwerte an Reaktordruckbehälterstählen mit der Einproben-Compliance-Methode und Korrelationen zur Kerbschlagarbeit
Get PDFInstitute of Safety Research, Annual Report 1994
Get PDFThe report gives an overview on the scientific work of the Institute of Safety Research in 1994
Berechnung von Neutronenflußdichteschwankungen in WWER-Druckwasserreaktoren infolge strömungsinduzierter Schwingungen
Get PDFZusammenfassung Dargestellt wird ein in sich geschlossenes 3D-Modell zur Berechnung der ortsabhängigen Übertragungsfunktionen für stochastische Schwingungen der Regel- und Brennelemente in Reaktoren vom Typ WWER-440 unter Berücksichtigung der heterogenen Brennstoffbeladung realer Spaltzonenkonfigurationen. Basis sind die nodalen Bilanz-Gleichungen zur Berechnung der stationären Neutronediußverteilung in der Eingruppen-Diffusionsnäherung für 3D-Nodes in Hexagonal-Z-Geometrie. Die Absorberteiie eines WWER- 440-Regelelements werden durch entsprechende Randbedingungen für die Diffusionsgleichun im neutronenphysikalischen Reaktormodell erfaßt. Die Modellierung der Rauschquellen erfolgt auf der Grundlage einer Sweigruppen-Diffusionsnäherung. Die Neutronenflußdichteschwankungen werden dabei durch Störungen bei der Kickstreuung thermischer Neuttonen in "Absorber-Nodes" bzw. durch stochastische Schwankungen der Zweigruppen-Wirkungsquerschiiitte an den radialen Nodegrenzen der 'Brennstoff-Nodes" beschrieben. Zur Berechnung der Ubertragungsfunktionen für die Schwankungen der Neutronenflußdichte werden die nodalen Bilanz-Gleichungen in der Prompt-Response-Approximation verwendet. Die Kopplung der Zweigruppen-Quellmodeiie für Brennstoff- und Absorberteil eines Regelelements an das nodale Eingruppen-Übertragungsmodel erfolgt im Sinne der modifizierten Eingruppen-Diffussionsnäherung. Anhand numerischer Berechnungen werden wesentliche Charakteristika der Übertragungsfunktionen für Regelelementschwingungen in Dwckwasserreaktoren vom Typ WWER-440 diskutiert
Two-phase flow behaviour during a medium size cold leg LOCA test on PMK-2 (IAEA - SPE-4)
Get PDFThe experiment to the IAEA standard problem exercise No. 4 was carried out in April 1993 on the integral test facility PMK-2 in Budapest. It was a 3.2 mm break on the downcomer head. The high pressure injection cooling was assumed to be not available. As an accident management measure bleed and feed on the secondary side of the steam generator was applied. Research Center Rossendorf contributed to the experiment of SPE-4 by supplying needle shaped conductivity probes for the measurement of local void fractions in the primary circuit of the PMK-II test facility. In the course of the standard problem exercise No. 4 RCR contributed with posttest calculations using the thermalhydraulic code ATHLET. The report comprises a description of the initial and boundary conditions of the test and a phenomenological description of the thermalhydraulic events during the transient. The void fraction probe signals yields valuable information for deeper understanding of the thermalhydraulic occurrences and for code validation. This was the case particularly for correction of the level measurement. A description of the thermalhydraulic occurrences from the viewpoint of code verification is given and the results of RELAP5 (KFKI-AEKI) and of ATHLET-calculations (RCR) are compared. Referring to this description, the sensitivity of the result to main influences is investigated using the ATHLET-code
