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AbstractAbstract
[en] The study investigates the accident of rupture of the sphere discharge pipe and determines the reactivity data for two different configurations of equal volume, comparing the surface area. The data are obtained by two-dimensional diffusion calculations using the ASTERIX-II code. The results show a reactivity decrease of 15.3% and 42.4%, respectively. (DG)
[de]
Zur Ermittlung des Reaktivitaetswerts beim Abriss des Kugelabzugsrohres wurden zwei verschiedene Konfigurationen bei gleichem Volumen gewaehlt und in Bezug auf die Oberflaeche miteinander verglichen. Hierzu dienten zweidimensionale Diffusionsrechnungen mit dem Rechenprogramm ASTERIX-II. Die Ergebnisse zeigen einen Reaktivitaetsrueckgang um 15,3% bzw. 42,4%. (DG)Original Title
Reaktivitaetsverhalten von Kugelschuettungen im Kugelabzugsraum des HTR-Modul
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Secondary Subject
Source
Mertens, J. (comp.); Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Nukleare Sicherheitsforschung; 212 p; Nov 1985; p. 181-186
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Report
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AbstractAbstract
[en] The systematic classification of the ATWS has been undertaken by analogy to the considerations made for LWR. The initiating events of ATWS and protection actions of safety systems resulting from monitoring of the system variables have been described. The main emphasis of this work is the analysis of the core dynamic consequences of scram failure during the anticipated transients. The investigation has shown that because of the temperature feedback mechanisms a temperature rise during the ATWS results in a self-shutdown of the reactor. Further inherent safety features of the HTR - conditioned by the high heat capacity of the core and by the compressibility of the coolant - do effectively counteract an undesirable increase of temperature and pressure in the primary circuit. In case of the long-term failure of the forced cooling and following core heatup, neutron physical phenomena appear which determine the reactivity behaviour of the HTR. They are, for instance, the decay of Xenon 135, release of the fission products and subsiding of the top reflector. The results of the computer simulations show that a recriticality has to be excluded during the first 2 days if the reactor is shutdown by the reflector rods at the beginning of the accident. (orig./HP)
[de]
Zunaechst wird die systematische Klassifizierung der ATWS-relevanten Ereignisablaeufe (Anticipated Transients Without Scram) bei HTR in Anlehnung an die Ueberlegungen fuer LWR vorgenommen und die Schutzaktionen der Sicherheitssysteme durch Ueberwachung von Systemvariablen erlaeutert. Der Schwerpunkt der Arbeit besteht in der Analyse der coredynamischen Auswirkungen des Ausfalls der Reaktorschnellabschaltsysteme bei derartigen Ereignisablaeufen. Die Untersuchung hat gezeigt, dass infolge von Temperaturrueckkopplungsmechanismen ein Temperaturanstieg im Zuge von ATWS zur Selbstabschaltung des Reaktors fuehrt. Hinzu kommt, dass die weiteren inhaerenten Sicherheitseigenschaften des HTR, bedingt durch die hohe Waermekapazitaet sowie durch die Kompressibilitaet des Kuehlmittels, einer unguenstigen Temperaturentwicklung und Druckerhoehung im Primaerkreis effektiv entgegenwirken. Bei Stoeranfaellen mit langdauerndem Ausfall der erzwungenen Corekuehlung und nachfolgender Coreaufheizung treten neutronenphysikalische Phaenomene auf, die das Reaktivitaetsverhalten des HTR bestimmen. Dazu zaehlen z.B. Zerfall von Xenon-135, Freisetzung neutronenabsorbierender Spaltprodukte und der Absturz des Deckenreflektors. Am Beispiel des HTR-500 zeigen die Ergebnisse der Simulationsrechnungen, dass eine Rekritikaltitaet innerhalb der ersten 2 Tage auszuschliessen ist, wenn der Reaktor mit dem Beginn des Stoerfalls nur durch die Reflektorstaebe abgeschaltet wird. (orig./HP)Original Title
Coredynamik von HTR unter ATWS- und Stoerfallbedingungen
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May 1988; 122 p
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Report
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ACCIDENTS, BETA DECAY RADIOISOTOPES, BETA-MINUS DECAY RADIOISOTOPES, DESIGN BASIS ACCIDENTS, EVEN-ODD NUCLEI, GAS COOLED REACTORS, GRAPHITE MODERATED REACTORS, HOMOGENEOUS REACTORS, HOURS LIVING RADIOISOTOPES, INTERMEDIATE MASS NUCLEI, ISOMERIC TRANSITION ISOTOPES, ISOTOPES, MINUTES LIVING RADIOISOTOPES, NUCLEI, POISONING, RADIOISOTOPES, REACTOR ACCIDENTS, REACTOR COMPONENTS, REACTOR SHUTDOWN, REACTORS, SAFETY, SHUTDOWN, SOLID HOMOGENEOUS REACTORS, XENON ISOTOPES
Reference NumberReference Number
INIS VolumeINIS Volume
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Khan, L.A.; Nabbi, R.
Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Zentralabteilung Forschungsreaktoren1987
Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Zentralabteilung Forschungsreaktoren1987
AbstractAbstract
[en] In connection with conversion of Pakistan Research Reactor (PARR) from the use of Highly Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel, steady-state thermal hydraulic analysis of both existing HEU and proposed LEU cores has been carried out. Keeping in mind the possibility of power upgrading, the performance of proposed LEU core, under 10 MW operating conditions, has also been evaluated. Computer code HEATHYD has been used for this purpose. In order to verify the reliability of the code, IAEA benchmark 2 MW reactor was analyzed. The cooling parameters evaluated include: coolant velocity, critical velocity, pressure drop, temperature distribution in the core, heat fluxes at onset of nucleate boiling, flow instability and burnout and corresponding safety margins. From the results of the study it can be concluded that the conversion of the core to LEU fuel will result in higher safety margins, as compared to existing HEU core, mainly because the increased number of fuel plates in the proposed design will reduce the average heat flux significantly. Anyhow upgrading of the reactor power to 10 MW will need the flow rate to be adjusted between 850 to 900 m3/hr, to achieve reasonable safety margins, at least, comparable with the existing HEU core. (orig.)
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Feb 1987; 70 p
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Report
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ACTINIDES, BOILING, ELEMENTS, ENERGY TRANSFER, ENRICHED URANIUM, ENRICHED URANIUM REACTORS, FUEL ELEMENTS, ISOTOPE ENRICHED MATERIALS, MATERIALS, METALS, NUCLEATE BOILING, PHASE TRANSFORMATIONS, POOL TYPE REACTORS, REACTOR COMPONENTS, REACTORS, RESEARCH AND TEST REACTORS, RESEARCH REACTORS, SAFETY, URANIUM, VELOCITY, WATER COOLED REACTORS, WATER MODERATED REACTORS
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AbstractAbstract
[en] Benchmark problem is solved using personal computer Leopard - UM2DB codes in order to validate reactor physics group methodology as well as code versions. Results are compared with those of various research centers. Differences around 1% were found
Original Title
Leopard - UM2DB (pc) en la solucion del problema Benchmark
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Journal Article
Journal
Nucleares; ISSN 0120-7067; 
; v. 6(no.11-14); p. 3-8
; v. 6(no.11-14); p. 3-8
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Kalker, K.J.; Nabbi, R.; Bormann, H.J.
Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Zentralabteilung Forschungsreaktoren1986
Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Zentralabteilung Forschungsreaktoren1986
AbstractAbstract
[en] Four small computer codes developed by ZFR are presented, which have been used for several years during operation of the research reactors FRJ-1, FRJ-2, AVR (all in Juelich) and DR-2 (Riso, Denmark). Because of interest coming from the other reactor stations the codes are documented within the frame work of the IAEA Research Contract No. 3634/FG. The zero-dimensional burnup program CREMAT is used for reactor cores in which flux measurements at each individual fuel element are carried out during operation. The program yields burnup data for each fuel element and for the whole core. On the basis of these data, fuel reloading is prepared for the next operational period under consideration of the permitted minimum shut down reactivity of the system. The program BURNY calculates burnup for fuel elements inaccessible for flux measurements, but for which 'position weighting factors' have been measured/calculated during zero power operation of the core, and which are assumed to be constant in all operational situations. The code CURIAX calculates post-irradiation data for discharged fuel elements needed in their manipulation and transport. These three programs have been written for highly enriched fuel and take into account U-235 only. The modification of CREMAT for LEU Cores and its combiantion with ORIGEN is in preparation. KINIK is an inverse kinetic code and widely used for absorber rod calibration at the abovementioned research reactors. It includes a special polynomial subroutine which can easily be used in other codes. (orig.)
[de]
Vier kleinere, in der ZFR entwickelte Rechenprogramme werden vorgestellt, die seit vielen Jahren an den Forschungsreaktoren FRJ-1, FRJ-2, AVR (alle in Juelich) und am DR-2 in Riso (Daenemark) benutzt werden. Das Interesse anderer Reaktorstationen an diesen Programmen fuehrte im Rahmen des IAEA-Projekts Nr. 3634/FG zu vorliegender Dokumentation. Der nulldimensionale Abbrandcode CREMAT setzt die Moeglichkeit voraus, in jedem einzelnen Brennelement eines Cores Flussmessungen waehrend des Reaktorbetriebs durchfuehren zu koennen. Die damit fuer jedes Brennelement und das Core berechneten Abbranddaten ermoeglichen die Planung der Brennelement-Konfiguration fuer das naechste Betriebscore unter Beruecksichtigung der Mindestreaktivitaet fuer die Reaktorabschaltung. Das Programm BURNY leistet das Gleiche fuer Brennelemente, deren Konstruktion keine Flussmessungen zulaesst, fuer die aber in der Anfahrphase des Reaktors bei Null-Leistung 'Positions-Gewichts- Faktoren' bestimmt werden, die dann als konstant fuer alle Betriebszustaende angenommen werden. Der Code CURIAX berechnet Werte fuer Restleistung und -Aktivitaet abgebrannter Brennelemente, die bei deren Manipulation und Transport bekannt sein muessen. Diese drei Programme sind fuer hoch angereicherten Brennstoff geschrieben und beruecksichtigen daher nur U-235. Eine Erweiterung von CREMAT fuer LEU-Brennstoff sowie dessen Kopplung mit ORIGEN ist in Arbeit. Der inverskinetische Code KINIK wird haeufig fuer die Kalibrierung von Steuerabsorbern an den oben genannten Reaktoren benutzt; er enthaelt ein spezielles Unterprogramm fuer Polynom- Approximation, das als Modul leicht in andere Programme uebernommen werden kann. (orig.)Original Title
CREMAT, BURNY, CURIAX; KINIK
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Source
Jan 1986; 142 p; CONTRACT 3634/FG
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Report
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Software
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ACTINIDE NUCLEI, ALPHA DECAY RADIOISOTOPES, COMPUTER CODES, DIAGRAMS, EVEN-ODD NUCLEI, HEAVY NUCLEI, INFORMATION, ISOMERIC TRANSITION ISOTOPES, ISOTOPES, KINETICS, MINUTES LIVING RADIOISOTOPES, NUCLEI, RADIOISOTOPES, REACTORS, RESEARCH AND TEST REACTORS, SIMULATION, URANIUM ISOTOPES, YEARS LIVING RADIOISOTOPES
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AbstractAbstract
[en] The core dynamic analysis of an anticipated heat removal transient without scram in a high-temperature gas-cooled reactor has indicated that in case of a failure of core cooling, the reactor undergoes a selfshutdown after 1 min because of its negative temperature coefficients of reactivity. If the decay heat removal system operates according to plant specification, recriticality, and thus nuclear power generation, occurs. However, the maximum rise in fuel elements temperature is limited to 500C due to the high heat capacity of the core. Without taking into consideration the effect of xenon feedback on the neutron kinetics, a new steady core state is established after 2 h in which the fuel temperature and gas outlet temperature at the lower core edge are 1950C higher than in normal operation. Due to transient xenon poisoning, a rise in gas outlet temperature only occurs during the first 70 min and amounts to 700C. For this reason undesirable transient strains on the components connected behind the core are not expected. A slow xenon buildup during the first hour ensures a long-term subcriticality of the reactor. Without any contribution from the shutdown system, this leads to a decrease in nuclear power and thus to core cooling with functioning decay heat removal
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Journal Article
Journal
Nuclear Technology; ISSN 0029-5450; ; v. 64(1); p. 5-13
; v. 64(1); p. 5-13
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AbstractAbstract
[en] The HEATHYD code used for modelling of fluid flow and heat transfer in reactor cores is described. 9 refs, 4 figs, 1 tab
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Source
Cullen, D.E. (Lawrence Livermore National Lab., CA (United States)); Muranaka, R.; Schmidt, J. (International Atomic Energy Agency, Vienna (Austria)) (eds.); International Centre for Theoretical Physics, Trieste (Italy); International Atomic Energy Agency, Vienna (Austria). Nuclear Data Section; 752 p; ISBN 981-02-0517-1; ; 1991; p. 597-623; World Scientific; Singapore (Singapore); Workshop on reactor physics calculations for applications in nuclear technology; Trieste (Italy); 12 Feb - 16 Mar 1990
; 1991; p. 597-623; World Scientific; Singapore (Singapore); Workshop on reactor physics calculations for applications in nuclear technology; Trieste (Italy); 12 Feb - 16 Mar 1990
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Book
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Conference
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Nabbi, R.; Wolters, J.
IGORR 8: 8. meeting of the International Group on Research Reactors. Proceedings2001
IGORR 8: 8. meeting of the International Group on Research Reactors. Proceedings2001
AbstractAbstract
[en] Radiation damage in the structures of the aluminum tank resulting from the reaction of fast and thermal neutrons with aluminum was simulated using the 3D MCNP Monte-Carlo code. Comparison of the local neutron flux with measurement showed a deviation less than 5%. Radiation damage in term of concentration of silicon was also determined for the front plate of the 4H-2 channel by the calculation of the total rate of Al(n,β)Si reaction. The results are in very good agreement with the measured Si-concentration obtained by chemical analysis. The results of calculation show that the tensile elongation of aluminum in the most irradiated segment of the 2TAN beam tube is reduced by 62% of the initial value. The content of Si is considerably lower than the end-of-life limit by 46% so that from the radiation damage point of view, the power operation of the reactor could be continued for additional 2.82E+6 MWh corresponding to approx. 20 years. (orig.)
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Technische Univ. Muenchen (Germany); Framatome ANP GmbH, Erlangen (Germany); 324 p; 2001; p. 237-243; IGORR 8: 8. meeting of the International Group on Research Reactors (IGORR); Munich (Germany); 17-20 Apr 2001; Available from TIB Hannover
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Miscellaneous
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Conference
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BARYON REACTIONS, BARYONS, CALCULATION METHODS, CONTAINERS, ELEMENTARY PARTICLES, ELEMENTS, ENRICHED URANIUM REACTORS, FERMIONS, HADRON REACTIONS, HADRONS, HEAVY WATER COOLED REACTORS, HEAVY WATER MODERATED REACTORS, IRRADIATION REACTORS, ISOTOPE PRODUCTION REACTORS, MATERIALS TESTING REACTORS, MECHANICAL PROPERTIES, METALS, NEUTRONS, NUCLEAR REACTIONS, NUCLEON REACTIONS, NUCLEONS, RADIATION EFFECTS, RADIATION FLUX, REACTORS, RESEARCH AND TEST REACTORS, RESEARCH REACTORS, SIMULATION, TANK TYPE REACTORS, TENSILE PROPERTIES
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Wolters, J.; Nabbi, R.
Proceedings of the international topical meeting on advanced reactors safety: Volume 21997
Proceedings of the international topical meeting on advanced reactors safety: Volume 21997
AbstractAbstract
[en] Feed and bleed cooling of the FRJ-2 research reactor can reduce the risk of core damage considerably as a probabilistic safety analysis has revealed. The question whether water circulation via the core would be maintained when the water in the tank has reached saturation point has been answered positively by an investigation with the thermohydraulic code CATHENA. A siphon with a water column and the special feature of self-acting restoration of the column after depressurization proved well during tests and will be installed as the relief equipment required to blow off the steam produced by the residual heat of the core during bleed cooling. 4 refs., 9 tabs
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Source
American Nuclear Society, La Grange Park, IL (United States); 715 p; 1997; p. 1320-1327; American Nuclear Society, Inc; La Grange Park, IL (United States); ARS '97: American Nuclear Society (ANS) international meeting on advanced reactors safety; Orlando, FL (United States); 1-5 Jun 1997; American Nuclear Society, Inc., 555 N. Kensington Ave., La Grange Park, IL 60526 (United States)
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Book
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Conference
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AbstractAbstract
[en] The studies and preparations to convert the FRJ-2 DIDO reactor at KFA Juelich from HEU to LEU were being carried out in a joint effort with industry (Interatom, Nukem) in the framework of the German AF-Program and started in the early eighties. It turned out very soon that the fuel element design has to be modified from the present HEU-EB element (electron beam welded fuel tubes). The basic data for LEU elements and cores and the essential requirements for the conversion process were determined mid of the eighties. The conversion procedure comprising several steps was agreed upon by the licensing authority nearly at the same time. Reactorphysics and thermohydraulic calculations proved that the RS design is superior to the EB design and will provide greater safety margins as well in HEU as in LEU operation. Due to significant delays in performing the first steps of the conversion procedure and because of the necessity to repeat test irradiations with the new RS type elements in HEU and LEU manufactured by CERCA it is expected that the HEU-LEU transition phase will not begin before the 3rd quarter of 1995. Since FRJ-2 at that time will have reached a considerably high age it could not be excluded that it is closed rather than converted to LEU operation mid of the nineties
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Argonne National Lab., IL (United States); 447 p; Jul 1993; p. 348-359; International meeting on reduced enrichment for research and test reactors; Newport, RI (United States); 23-27 Sep 1990; Also available from OSTI as DE94006497; NTIS
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Report
Literature Type
Conference
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ACTINIDES, ELEMENTS, ENRICHED URANIUM, ENRICHED URANIUM REACTORS, HEAVY WATER COOLED REACTORS, HEAVY WATER MODERATED REACTORS, IRRADIATION REACTORS, ISOTOPE ENRICHED MATERIALS, ISOTOPE PRODUCTION REACTORS, MATERIALS, MATERIALS TESTING REACTORS, METALS, REACTOR COMPONENTS, REACTORS, RESEARCH AND TEST REACTORS, RESEARCH REACTORS, TANK TYPE REACTORS, URANIUM
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