Zur selbsttätig sicheren Begrenzung von nuklearer Leistung und Brennstofftemperatur in innovativen Kernreaktoren

Nuclear energy probably will not contribute significantly to the future worldwide energysupply until it can be made catastrophefree. Therefore it has to be shown, that theconsequences of even largest accidents will have no major impact to the environmentof a power plant.In this paper one of the basic conditions for such a nuclear technology is discussed.Using mainly the modular pebble-bed high-temperature reactor as an example, thedesign principles, analytical methods and the level of knowledge as given today incontrolling reactivity accidents by inherent safety features of innovative nuclear reactorsare described. Complementary possibilities are shown to reach this goal with systems ofdifferent types of construction . Questions open today and resulting requirements forfuture activities are discussed .Today's knowledge credibly supports the possibility of a catastrophefree nucleartechnology with respect to reactivity event

Das zweidimensionale Reaktordynamikprogramm TINTE Teil 2: Anwendungsbeispiele

The TINTE code system deals with the nuclear and the thermal transient behaviour of the primary circuit of an HTGR taking into consideration the mutual feedback effects in two-dimensional r-z geometry. In Part One of this report (JÜI-2167, Nov. 1987) the initial equations were compiled and methods of solution discussed. In an appendix to this second part they are completed by some supplementary points. The TINTE code construction and a detailed input description will be discussed inPart Three. The present Part Two shows examples of application, especially a comparrative calculation of dynamic experiments performed at the AVR. A good agreament between calculational and experimental results is found. Further examples show the flexibility of TINTE : first of all, individual moduli of TINTE are used to find a solution to a thermofluid problem. In addition TINTE is used to demonstrate the mutual feedback between nuclear and thermal processes in heating reactors, including those with natural convective conditions, without any control rod movement

Das zweidimensionale Reaktordynamikprogramm Tinte Teil 1 : Grundlagen und Lösungsverfahren

The TINTE code deals with the nuclear and the thermal transient behaviour of an HTR taking into consideration the mutual feedback effects in two-dimensional r-z-geometry. Initial equations, approximations and solution procedures are compiled in this first part of the description. This involves the following subproblems : $\cdot$ Time-dependent neutron flux calculation. $\cdot$ Time-dependent heat source distribution (local and non-local fractions). $\cdot$ Time-dependent heat transport from the fuel to the fuel element surface. $\cdot$ Time-dependent global temperature distribution. $\cdot$ Gas-flow even under natural circulation conditions for both a given total mass flow and a given pressure difference. $\cdot$ Convection and its feedback to the circulation. The iterations of subproblem solutions, necessary because of the separate treatment,are discussed for both the transient case and of the determination of the steady initial state

Berechnung von Streumatrizen für die Materialien des MOSEL- Reaktors

Für schnelle und epithermische Reaktoren ist die genaue Kenntnis der Streudaten von großer Bedeutung. Es wurden deshalb neue Streudaten für die insbesondere im MOSEL-Reaktor benutzten Elemente und Isotope zusammengetragen und ein Programm zu ihrer Auswertung, zur Berechnung neuer Streumatrizen geschrieben, bei dem der Mittelungsfluß in den einzelnen Gruppen beliebig vorgegeben werden kann. Außerdem wurde der Einfluß der neuen Streudaten und der verschiedenen Mittelungspektren auf die Kritikalität und auf $\bar{\eta}$, die mittlere Zahl der Neutronen pro Absorption im U-233, bei einer harten und einer weichen MOSEL-Anordnung untersucht

Die Behandlung der Nachzerfallswärme im Reaktordynamikprogramm TINTE

The TINTE code system deals with the nuclear and the thermal transient behaviour of the primary circuit of an 1-lTGR tatting into consideration the mutual feedback effects in two-dimensional r-z-geometry. In this paper an update of the treatment of delayed heat production is presented. It is based on the german norm DIN 15485, the rules ofwhich had to be adjusted for use in a dynamics code. For the description of the fuel element power history a substitutehistogram has been constructed from local burnup and optionally from information about shuffling of the fuel bails. As an example the depressurisation accident of a MODUL-1-ITR is calculated. The results obtained are very similar to others previously reported

Analysis of the gas diffusion process during a hypothetical air ingress accident in a modular high temperature gas cooled reactor

In order to simulate the diffusion process during a hypothetical air ingress accident in a modular high temperature gas cooled reactor, a one-dimensional coupled diffusion-convection model has been established. In this analysis it is shown first, that experiments performed at the Japan Atomic Energy Research Institute (JAERI) have been recalculated successfully, thus validating the new model. Applying this model to the NACOK facility, now under construction at the Institute for Safety Research and Reactor Technology (ISR) of the KFA Research Center Jiilich, a delay time until the onset of natural circulation was found to be 14 hours. For the 200MW HTR-MODULE designed by the SIEMENS company about 39 hours are predicted. It was found that pressure disturbances, e.g. caused by environmental noise could significantly shorten the time delay until the onset of natural circulation. The onset time delay remains constant below 40 dB noise, reaches half of its value at 80 dB and tends to zero at 120 dB. Thus, the relevance of the delay effect with respect to safety related questions is reduced

Studie zur Untersuchung der Wirksamkeit einer zentralen Neutronenquelle für das 300-MWe-THTR-Kernkraftwerk Schmehausen

The loading and starting phase of the THTR-3oo is eontrolled by an external neutron source. The possible positions of this source result from the requirement to fulfil certain criteria concerning the intensity and the interpretation of the detector-Signal. In this study it is investigated, how far a source centrally inserted into the reactor can be equal to this tack and whether it is possible to go over to a non-central position at a later point of time. lt is shown that the central source can fulfil all discussed demands. A source position lying in the vicinity of the side reflector demands a special choice of the detector position, which can be exactly determined only by further three-dimensional caIculations. Beyond that this study shows that during the loading phase in the subcritical range the spatial neutron flux distribution is expected as to be completely different from that of the critical reactor

Sozialistisches Wirtschaftsrecht und sozialistisches Rechtsbewusstsein der Werktaetigen zu einigen Ergebnissen von Untersuchungen im VEB Kombinat Tiefbau Berlin und im VEB Werk fuer Fernsehelektronik Berlin

HUB(11) - 82 HB 3355 / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekSIGLEDEGerman

V. S. O. P. ('94) Computer Code System for Reactor Physics and Fuel Cycle Simulation

V. S. O. P. ('Very Superior Old Programs) is a system of codes lurked together for the simulationof reactor life histories and temporary in-depth research. In comprises neutron cross sectionlibraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculationwith depletion and shut-down features, in-core and out-of--pile fuel management, fuel cyclecost analysis, and thermal hydraulics (at present restricted to 's). Various techniques havebeen employed to accelerate the iterative processes and to optimize the internal data transfer.The storage requirement is confined to 17 M-Bytes.The code system has extensively been used for comparison studies of reactors, their fuel cycles,simulation of safety features, developmental research, and reactor assessments

Zur chemischen Stabilität bei innovativen Kernreaktoren : Korrosion von graphitischen Strukturen durch Wasser und Luft, Konzepte zum inhärenten Schutz vor schweren Schäden

Nuclear energy probably will not contribute significantly to the future worldwide energy supply until it can be made $\mathbf{catastrophe-free}$. Therefore it has to be shown, that the consequences of even largest accidents will have no major impact an the environment of a power plant. In this paper one of the basic conditions for such a nuclear technology is discussed. Using mainly the pebble-bed high-temperature reactor as an example, experimental and theoretical results and the level of knowledge as given today in graphite corrosion processes and their consequences are described. Questions open today and resulting requirements for future activities are discussed. Today's knowledge allows the conclusion, that a $\mathbf{catastrophe-free}$ nuclear technology with respect to chemical stability can be realized