Volume 75, Issue 3

This paper provides information about the possibility to apply the FEMAXI-6 code for RBMK-1500 reactors. New thermal properties and correlations responsible for thermal analysis (thermal conductivity and heat capacity) according to RBMK-1500 specification were included into the FEMAXI-6 code. The RBMK-1500 fuel rod model using the adapted FEMAXI-6 code was developed and tested by employing uncertainty and sensitivity analysis. The processes in fuel rods during normal plant operation were modelled. The results received were compared with calculations performed by specialists from the Kurchatov Institute (designers of RBMK). The reasonable agreement of both calculation results shows that the adapted FEMAXI-6 code and the developed model are suitable for future analysis of processes in fuel rods of RBMK-1500 reactors. Also this work demonstrates what sensitivity and uncertainty analysis can be used for the model development.

The thermal response of the experimental fuel elements is examined by thermocouples placed at the centre of the fuel. Such measurements allow the determination of thermal feedback effects induced by simultaneous liberation of fission gases and pellet thermal expansion densification, swelling and relocation of fuel fragments. This paper describes briefly measuring techniques developed and currently used in INR Pitesti, and presents and discusses selected experimental results obtained during in-pile experiments performed in the TRIGA SS reactor of INR Pitesti in order to measure the fuel centerline temperature variation during power change.

Fire resistance of structures, systems and components (SSC) is widely identified as a fundamental issue for fire risk assessment at nuclear power plants (NPP). It is currently supported by several national or international ambitious experimental programs. Fire risk assessment at NPPs is performed under a series of hypothesis, e.g. the fire load. A problem will arise if tests used for conformity assessment do not appropriately represent the hypothesis selected during the fire risk analysis. In this case, SSCs may possibly be not meeting the expectations related to their fire resistance capacity supposed by the fire risk analysis. Ways of tackling this issue involve a regulatory approach and adapted requirements for components. While the widespread regulatory approach consists in seeking enveloping scenarios and conservative assumptions, a better knowledge of SSCs behaviour under for NPPs typical situations may be also desirable. The Institute for Energy of the European Commission's Joint Research Centre (JRC) at Petten (NL) launched a questionnaire with the aim of gathering opinions of regulatory bodies and their technical support organisations and manufacturers in Europe on this topic. Major goal was to estimate the relevance of the issue according to the perception of the specialists. The present paper collects the answers given by seven entities (AREVA, GRS, iBMB, IRSN, SJUB, SUEZ-TRACTEBEL, and VTT) and makes suggestions for addressing the issues identified by the participants.

A new method for analyzing the problem of the thermal-hydrodynamic instability of OTSG (Once-Through Steam Generator) of movable nuclear power devices is presented in this paper. The method is based on multivariable frequency domain theory and is more efficient for analyzing the instability of OTSG with coupling interactions and complicated boundary conditions than the usually used single variable method. A mathematical model for the system is derived from the fundamental equations by using perturbation, Laplace-transform and the nodalization techniques. The stable boundary and parametric effects on the stability of the system are evaluated with a computer code. Numerical examples are given and the predictions of the model are in good agreement with the experimental results.

A computation code THERMOHYD has been developed to perform the steady state thermal-hydraulic calculations of the MTR type research reactors fueled with plate-type-fuel, in the ranges of pressures and temperatures typical for MTR reactors. The code calculates the coolant and clad surface temperatures along the cooling channel, as well as the critical heat flux and provides values for the critical phenomena parameters (RDR, ONBR and DNBR ratios) for reactor cores. The code takes into account the variation of the physical properties of the coolant with temperature. It also accounts for the uncertainties due to fuel fabrication tolerances, possible deviations in the operational measurements, the errors in the experimental correlations used in the analysis, and the errors due to simplifications made in the calculations which have been considered in the analysis. The THERMOHYD code was tested against the well known PARET code and proved to be accurate and more conservative in some aspects.

In this study, production routes of medical 110,111 In radionuclides produced by 110 Cd(p, n) 110 In, 111 Cd(p, 2n) 110 In, 112 Cd(p, 3n) 110 In, 107 Ag(α, n) 110 In, 109 Ag(α, 3n) 110 In, 111 Cd(p, n) 111 In, 112 Cd(p, 2n) 111 In, 113 Cd(p, 3n) 111 In, 114 Cd(p, 4n) 111 In, 109 Ag(α, 2n) 111 In, 107 Ag(α, γ) 111 In and 113 In(n, 3n) 111 In reactions have been investigated up to 70 MeV incident particle energy. In these calculations, the pre-equilibrium and equilibrium effects have been investigated. The pre-equilibrium calculations involve the geometry dependent hybrid model and the cascade exciton model. The full exciton model has also been used to investigate the pre-equilibrium direct effects. Equilibrium effects are calculated according to the Weisskopf–Ewing model. The calculated results are compared with the experimental data taken from the literature.

In this study, the nuclear energy level density based on nuclear collective excitation mechanism has been identified in terms of low-lying collective level bands near the neutron binding energy. Nuclear level density parameters of some large deformed nuclei of radionuclides widely used in medical applications have been calculated by using different collective excitation modes of observed nuclear spectra. The calculated parameters have been used successfully in estimation of the neutron-capture cross section basic data for the production of new medical radionuclides. The investigated radionuclides have been considered in the region of mass number 150 < A < 200. The method used in the present work assumes equidistance spacing of the collective coupled state bands of the interest radionuclides. The calculated results have been compared with the compiled values from the literatures for s-wave neutron resonance data.

In this article the transport model and inductive approach reported in a recent work have been extended, and a discrete ordinates operator for the diffuse transmission of a mixed neutron beam through a multilayered medium is derived. At first, a mathematical description of the mixed beam transport problem that we are concerned with is given. The resulting mathematical problem is, according to Chandrasekhar, decomposed into uncollided and diffuse problems. After treating the uncollided problem analytically, a discrete ordinates formulation of the diffuse problem is considered. A set of spatially discretised equations that are free from spatial truncation error is then introduced. Finally, a discrete ordinates operator that replaces the entire multilayered medium in diffuse neutron transmission computations is derived inductively. To the best of our knowledge, such an operator has never been reported previously in neutron transport literature. Numerical results for a four-layer medium are presented to illustrate the computational merit of a computer code based on our diffuse transmission operator.

Kurzfassung In dem nachfolgenden Beitrag werden die Grundsätze der Kritikalitätssicherheitsauslegung eines Endlagers ausgedienter Kernbrennstoffe erläutert. Diese Erläuterungen umfassen: Sicherheitsprinzipien der Kritikalitätssicherheitsauslegung eines Endlagers Konsistenz der Sicherheitsprinzipien beim Übergang von der Betriebsphase in die Nachbetriebsphase des Endlagers Die prinzipielle Abhängigkeit der Länge erforderlicher Nachweiszeiträume von den eingelagerten Nuklidinventaren und deren zukünftigen Entwicklungen, die durch den radioaktiven Zerfall sowie durch mögliche langzeitliche Entwicklungen des Endlagers bedingte Einwirkungen bestimmt werden Wahrscheinlichkeit des Eintretens von Kritikalität hinsichtlich bisher erwogener Einlagerungskonzepte und die Notwendigkeit des Nachweises der Sicherheit des Endlagers gegen potentielle Kritikalität in der Nachbetriebsphase Die sich zur Beantwortung der Frage der Sicherheit des Endlagers gegen potentielle Kritikalität für die Kritikalitätsanalyse ergebenden Aufgaben werden skizziert. Die Bearbeitung dieser Aufgaben macht die Modellierung komplexer Mechanismen (Szenarien), die zu Kritikalität führen können, die Modellierung der komplexen und verschiedenartigen Kritikalitätsexkursionen und die Modellierung der Auswirkungen solcher Exkursionen auf die Sicherheit des Endlagers erforderlich. Die Entwicklung entsprechender Modelle erfordert die konzertierte Zusammenarbeit vieler Fachdisziplinen. Es wird angeregt, die bereits bestehenden Kooperationen unter Einbeziehung der Kritikalitätsanalyse in einer nationalen Initiative zum Projekt „Endlagerung“ noch näher miteinander zu vernetzen. Aufgabe einer solchen Kooperation wird es auch sein, die Risiken aller Tätigkeiten, Einrichtungen und Randbedingungen, die mit der Abfallkonditionierung sowie der Beladung, dem Betrieb und der Nachbetriebsphase eines Endlagers verbunden sind, zu bewerten und so optimal auszubilanzieren, dass das Gesamtrisiko dem Minimierungsgebot genügt. Dies schließt eine Bilanzierung des Risikos einer potentiellen Kritikalität in der Nachbetriebsphase gegen die Risiken ein, die mit der Durchführung von Maßnahmen verbunden sind, die zur physikalisch sicheren Verhinderung des Eintretens von Kritikalität in der Nachbetriebsphase vor und während der Einladung in das Endlages erforderlich wären.

In both, industry and scientific research, there is a demand for the determination of material thicknesses. In this work, two methods are described which are very simple, fast, effective and low-cost to determine the exactly thickness of thin foils (μm). Both methods are based on the energy loss of alpha particles when passing through the foil and the detection by high energy resolution alpha spectroscopy. The strengths and weakness of the two methods are discussed.