Results 1 - 10 of 1938
Results 1 - 10 of 1938. Search took: 0.022 seconds
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[en] Owing to growing interest in the study of emitted antineutrinos from nuclear reactors to test the Atomic Energy Agency safeguards, antineutrino flux was studied in the Tehran Research Reactor (TRR) using ORIGEN code. According to our prediction, antineutrino rate was obtained 2.6 x 1017 (ve/sec) in the core No. 57F of the TRR. Calculations indicated that evolution of antineutrino flux was very slow with time and the performed refueling had not an observable effect on antineutrino flux curve for a 5 MW reactor with the conventional refueling program. It is seen that for non-proliferation applications the measurement of the contribution of 239Pu to the fission using an antineutrino detector is not viable in the TRR.
[en] The nuclear reactor remote monitoring system of the state of Baden-Wuerttemberg (Kernreaktor-Fernueberwachung Baden-Wuerttemberg - KFUeBW) is implemented according to the recently renewed ''recommendations for remote monitoring of nuclear power plants''. In Baden-Wuerttemberg, the application area of the system covers both, the surveillance of internal procedures on one hand, and the handling of incidents or accidents on the other. The following paper shows the role of the KFUeregarding the determination and evaluation of the radiological situation in the range of off-site emergency response. Progress is reported on the measurement conception and the technical possibilities for the investigation of the radiological situation after the end of the deposition of radio nuclides (ground phase). (orig.)
[en] Gadolinium is mixed with nuclear fuel for extended burnup, increased reactor cycle Lengths, and in-out core refueling schemes. The purpose of this work is to investigate the effect of Gadolinium on neutronic parameters. A comprehensive analysis of different fuel assemblies bearing gadolinia with different concentration and different number of gadolinia rods was carried out. The EPR fuel assemblies were chosen in this study. The fuel assemblies are of seven different types, each assembly has unique design characteristics in terms of enrichment, number of gadolinia rods and gadolinium concentration. The calculations were achieved using MCNP6 code and with ENDF/B-VII library. A fine mesh tally was superimposed on the geometry to illustrate flux distribution in the whole assembly and its change with burnup. The variation of reactivity and isotopic composition for each region of the assembly with burnup and its influence on flux distribution was studied. The results showed that the presence of gadolinium greatly affects the power distribution. It causes some skewness in axial power shape towards the top of the assembly, the assembly which has largest number of gadolinia rods and the lowest enrichment has the largest skewness of power distribution at BOC. At EOC the flux and axial power have a double hump shape due to the buildup of gadolinium isotopes and the large decrease in reactivity with burnup.
[en] Higher order approximations of the Chebyshev polynomials of first kind (T) are used for the first time in calculation of the diffusion lengths of monoenergetic neutrons in a homogeneous slab. In the method, the diffusion lengths of the neutrons are calculated using various values of the c, the number of secondary neutrons per collision. First, the traditional Legendre polynomials (P) approximation and then the present T method are used separately. The numerical results for the diffusion lengths are tabulated in the tables up to an order of N = 9. A brief comparison is also done between the results obtained from the present method and the ones in literature. The advantages of the present method can easily be observed from the good accordance between results given in the tables for comparison and its easily executable equations. For many of the c values, the results obtained from T method are better than the results obtained from P method.
[en] Exposure of radioactivity applications should be handled reliably in repositories, radiotherapy rooms, and research centers built with cement-based composites which is generally used as an engineering barrier. The design of certain materials for radioactive exposure requires special handling considering the degradation mechanism of host composite environment and barrier capability. In this study, celestite (SrSO) minerals having favoring properties for shielding ability was used as aggregates in barrier composites. Strontium mineral-based aggregates were partially replaced with conventional concrete aggregates at different ratios. The high rate X-ray shielding ability and mechanical performance of developed composites were holistically investigated in the presence of real-case radiation. The use of celestite mineral resulted in higher performance both in mechanical and shielding capability of X-rays at a certain level. Microstructural findings also revealed that interface properties of composite paste and celestite minerals were compatible up to 30% of celestite aggregate replacement.
[en] The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V220.127.116.11 code "Accident Source Term Evaluation Code" covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.
[en] The mission of the Comprehensive Nuclear-Test-Ban Treaty International Data Centre (IDC) is to: (a) acquire data over a Global Communications Infrastructure from a global network of 337 facilities of the International Monitoring Systems (IMS), (b) to process and analyze these data, and (c) to provide the IMS data, IDC products and services to Member States. In effect, the IDC symbolizes a new brand of arms control for the information age, leveraging Internet communications, knowledge-based data fusion, graphical decision support systems and Web-based user interfaces to achieve its mission. During 2000, the IDC was disseminating products based on data from about 90 seismic, hydroacoustic, infrasound and radionuclide stations of the future network. The number of events in the reviewed seismo-acoustic bulletins ranged from 40 to 360 each day. On average, some 200 radionuclide spectra were processed and analysed each month. Users from 45 Member States received an average of close to 18,000 data and product deliveries per month from the IDC. As the IDC continues to prepare for entry-into-force of the CTBT, it will continue to integrate the state-of-the-art in science and technology in order to meet the demands of the increasing volume of new types of IMS data, expanded IDC services, and a growing base of users. (orig.)
[de]Die Aufgaben des Internationalen Datenzentrums (IDC) fuer das Umfassende Verbot fuer Nuklearversuche (UVNV) sind die folgenden : (a) Sammeln der Daten vom globalen Netzwerk mit 337 Einrichtungen des Internationalen Ueberwachungssystems (IMS) ueber eine globale Kommunikationsinfrastruktur, (b) Verarbeitung und Analyse dieser Daten, und (c) Versorgung der Mitgliedsstaaten mit diesen IMS Daten sowie mit IDC Produkten und Diensten. Das IDC symbolisiert damit eine neuen Typ der Ruestungskontrolle im Informationszeitalter und stuetzt sich dabei auf Internet Kommunikation, wissensbasierte Datenfusion, graphische Systeme zur Entscheidungshilfe sowie Web-gestuetzte Schnittstellen um diese Aufgabe zu erfuellen. Im Laufe des Jahres 2000 verbreitete das IDC Produkte, die auf Daten von etwa 90 bereits existierenden seismischen, hydroakustischen, Infraschall- und Radionukliddetektoren des zukuenftigen Netzwerkes basieren. Die Anzahl der Ereignisse in den analysierten seismoakustischen Bulletins reichte von 40 bis 360 pro Tag. Im Durchschnitt wurden rund 200 Radionuklidspektren pro Monat verarbeitet und analysiert. Nutzer aus 45 Mitgliedsstaaten haben im Durchschnitt nahezu 18,000 Daten- und Produktauslieferungen pro Monat vom IDC erhalten. Im Zuge seiner fortlaufenden Vorbereitungen auf das Inkrafttreten des UVNV wird das IDC seine Methoden weiterhin dem Stand von Wissenschaft und Technik anpassen, um den Anforderungen des wachsenden Umfangs von neuen Arten der IMS Daten, erweiterten IDC Diensten und einer wachsenden Nutzerschaft gerecht zu werden. (orig.)
[en] Unauthorized dilution of boric acid in the reactor coolant due to improper operation of the chemical and volume control system (KBA) is a project mode with the violation of normal operation. The operator is obliged to identify unauthorized dilution within half an hour and stop feeding with pure condensate. The power unit has automation, which stops feeding on EP (emergency protection) and ATWS (Anticipated transient without scram) signals. The situation without the intervention of the operator and the failure of automation to stop dilution can be considered as a beyond design basis accident. The article presents an analysis of unauthorized boric acid dilution process in the primary circuit coolant of the NPP-2006 power unit up to the controlled state, either considering the operator's intervention after a 30-minute period of process, i. e. with a forced cessation of dilution, or without the operator's intervention until the tanks with pure condensate are completely emptied.
[en] Turbulent mixing rate between adjacent subchannels in a two-phase flow has been known to be strongly dependent on the flow pattern. The most important aspect of turbulent motion is that the velocity and pressure at a fixed point do not remain constant with time even in steady state but go through very irregular high frequency fluctuations. These fluctuations influence the diffusion of scalar and vector quantities. The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type, heavy water moderated and boiling light water cooled natural circulation based reactor. The fuel bundle of AHWR contains 54 fuel rods set in three concentric rings of 12, 18 and 24 fuel rods. This fuel bundle is divided into number of imaginary interacting flow channel called subchannels. Alteration from single phase to two phase flow situation occurs in reactor rod bundle with raise in power. The two phase flow regimes like bubbly, slug-churn, and annular flow are generally encountered in reactor rod bundle. Prediction of thermal margin of the reactor has necessitated the investigation of turbulent mixing rate of coolant between these subchannels under these flow regimes. Thus, it is fundamental to estimate the effect of spacer grids on turbulent mixing between subchannels of AHWR rod bundle.
[en] Since estimating the minimum departure from nucleate boiling ratio (MDNBR) requires complex calculations, an alternative method has always been considered. One of these methods is neural network. In this study, the Back Propagation Neural network (BPN) and Radial Basis Function Neural network (RBFN) are introduced and compared in order to estimate MDNBR of the VVER-1000 light water reactor. In these networks, the MDNBR were predicted with the inputs including core mass flux, core inlet temperature, pressure, reactor power level and position of the control rods. To obtain the data required to design these neural networks, an externally coupled-code was developed and its ability to estimate the thermo-hydraulic parameters of the VVER-1000 reactor was compared with other numerical solutions of this benchmark and the Final Safety Analysis Report (FSAR). After ensuring the accuracy of this coupled-code, MDNBR was calculated for 272 different conditions of reactor operating, and it was used to design BPN and RBFN. Comparison of these two neural networks revealed that when the output SMEs of the two systems were approximately the same, the training process in RBFN was much faster than in BPN and the maximum network error in RBFN was less than in BPN.