Temat: VVER-1000 - Katalog OPAC zbiorów

Skocz do pozycji: 1.

Tytuł:: Neutronic analysis of nanofluids as a coolant in the Bushehr VVER - 1000 reactor
Autorzy:: Zarifi, E.
Jahanfarnia, G.
Veisy, F.
Powiązania:: https://bibliotekanauki.pl/articles/146618.pdf
Data publikacji:: 2012
Wydawca:: Instytut Chemii i Techniki Jądrowej
Tematy:: nanofluids
neutronic analysis
VVER-1000
WIMS-D5
CITATION
Opis:: The main goal of this study was to perform the neutronic analysis of nanofluids as a coolant in reactor simulation. The variation of multiplication factor and thermal power have been investigated in the Bushehr VVER-1000 reactor core with using different nanofluids as coolant. In the applied analysis, water-based nanofluids containing various volume fractions of Al2O3, TiO2, CuO and Cu nanoparticles were used. The addition of different types and volume fractions of nanoparticles were found to have various effects on reactor neutronic characteristics. By using WIMS-D5 and CITATION code, the appropriate nanofluid with optimum volume percentage of nanoparticles was achieved. The results show that at low concentration (0.1% volume fraction) alumina is the optimum nanoparticle for normal reactor operation.
Źródło:: Nukleonika; 2012, 57, 3; 375-381
0029-5922
1508-5791
Pojawia się w:: Nukleonika
Dostawca treści:: Biblioteka Nauki

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Skocz do pozycji: 2.

Tytuł:: Estimation of control rod worth in a VVER-1000 reactor using DRAGON4 and DONJON4
Autorzy:: Saadatian-derakhshandeh, F.
Safarzadeh, O.
Shirani, A. F.
Powiązania:: https://bibliotekanauki.pl/articles/146900.pdf
Data publikacji:: 2014
Wydawca:: Instytut Chemii i Techniki Jądrowej
Tematy:: control rod worth
DRAGON4
DONJON4
VVER-1000 reactor
Opis:: One of the main issues in safety and control systems design of power and research reactors is to prevent accidents or reduce the imposed hazard. Control rod worth plays an important role in safety and control of reactors. In this paper, we developed a justifiable approach called D4D4 to estimate the control rod worth of a VVER-1000 reactor that enables to perform the best estimate analysis and reduce the conservatism that utilize DRAGON4 and DONJON4. The results are compared with WIMS-D4/CITATION to show the effectiveness and superiority of the developed package in predicting reactivity worth of the rod and also other reactor physics parameters of the VVER-1000 reactor. The results of this study are in good agreement with the plant’s FSAR.
Źródło:: Nukleonika; 2014, 59, 2; 67-72
0029-5922
1508-5791
Pojawia się w:: Nukleonika
Dostawca treści:: Biblioteka Nauki

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Skocz do pozycji: 3.

Tytuł:: Study of boron dilution phenomenon in the core and fuel assemblies of Bushehr VVER-1000 reactor in normal operating conditions
Autorzy:: Rahmani, Y.
Zarifi, E.
Pazirandeh, A.
Powiązania:: https://bibliotekanauki.pl/articles/148568.pdf
Data publikacji:: 2010
Wydawca:: Instytut Chemii i Techniki Jądrowej
Tematy:: VVER-1000
boron dilution
WIMS D5
CITATION
COBRA-EN
core and sub-channel analysis
Opis:: The spatial temperature distributions in fuel and coolant, results in appearing local changes in those elements densities in the reactor core, and also due to the complete solubility of boric acid in the coolant, there will be a direct correlation between the changes in the boron concentration and the coolant density. Because of the gradual reduction of boron concentration, first a local positive reactivity will be inserted into the core which will cause slight thermo-neutronic fluctuations in the reactor core. Of course, the trend of this process in the case of excessive reduction of the density of the coolant and evaporation of water (accident scenarios) will be reversed and subsequently the negative reactivity will be given to the system. With regard to the importance of this phenomenon, the spatial changes of boron concentration in the core and fuel assemblies of Bushehr VVER-1000 reactor have been examined. In line with this, by designing a complete thermo-neutronic cycle and by using CITATION, WIMS D-5 and COBRAN-EN codes, coolant temperature distribution and boron concentration will be calculated through this procedure, which first by using the output results of WIMS and CITATION codes, the thermal power of each fuel assembly will be calculated and finally, by linking these data to COBRA-EN code and using core and sub-channel analysis methods, the three-dimensional (3D) calculations of boron dilution will be obtained in the core as well as the fuel assemblies of the reactor.
Źródło:: Nukleonika; 2010, 55, 3; 323-330
0029-5922
1508-5791
Pojawia się w:: Nukleonika
Dostawca treści:: Biblioteka Nauki

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Skocz do pozycji: 4.

Tytuł:: A study of the effects of changing burn-up and gap gaseous compound on the gap convection coefficient (in a hot fuel pin) in VVER-1000 reactor
Autorzy:: Rahgoshay, M.
Rahmani, Y.
Powiązania:: https://bibliotekanauki.pl/articles/147623.pdf
Data publikacji:: 2007
Wydawca:: Instytut Chemii i Techniki Jądrowej
Tematy:: VVER-1000
nuclear reactor
burn-up
Ross and Stoute model
gap convection
hot fuel pin
thermal expansion
Opis:: In this article we worked on the result and process of calculation of the gap heat transfer coefficient for a hot fuel pin in accordance with burn-up changes in the VVER-1000 reactor at the Bushehr nuclear power plant (Iran). With regard to the fact that in calculating the fuel gap heat transfer coefficient, various parameters are effective and the need for designing a model is being felt, therefore, in this article we used Ross and Stoute gap model to study impacts of different effective parameters such as thermal expansion and gaseous fission products on the hgap change rate. Over time and with changes in fuel burn-up some gaseous fission products such as xenon, argon and krypton gases are released to the gas mixture in the gap, which originally contained helium. In this study, the composition of gaseous elements in the gap volume during different times of reactor operation was found using ORIGEN code [3]. Considering that the thermal conduction of these gases is lower than that of helium, and by using the Ross and Stoute gap model, we find first that the changes in gaseous compounds in the gap reduce the values of gap thermal conductivity coefficient, but considering thermal expansion (due to burn-up alterations) of fuel and clad resulting in the reduction of gap thickness we find that the gap heat transfer coefficient will augment in a broad range of burn-up changes. These changes result in a higher rate of gap thickness reduction than the low rate of decrease of heat conduction coefficient of the gas in the gap during burn-up. Once these changes have been defined, we can proceed with the analysis of the results of calculations based on the Ross and Stoute model and compare the results obtained with the experimental results for a hot fuel pin as presented in the final safety analysis report of the VVER-1000 reactor at Bushehr [2]. It is noteworthy that the results of accomplished calculations based on the Ross and Stoute model correspond well with the existing experimental results for this reactor.
Źródło:: Nukleonika; 2007, 52, 3; 93-95
0029-5922
1508-5791
Pojawia się w:: Nukleonika
Dostawca treści:: Biblioteka Nauki

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Skocz do pozycji: 5.

Tytuł:: Study of temperature distribution of fuel, clad and coolant in the VVER-1000 reactor core during group-10 control rod scram by using diffusion and point kinetic methods
Autorzy:: Rahgoshay, M.
Rahmani, Y.
Powiązania:: https://bibliotekanauki.pl/articles/146827.pdf
Data publikacji:: 2007
Wydawca:: Instytut Chemii i Techniki Jądrowej
Tematy:: point kinetic
diffusion
COSTANZA-R,Z
VVER-1000
control rod
RELAP5
Opis:: In this paper, through the application of two different methods (point kinetic and diffusion), the temperature distribution of fuel, clad and coolant has been studied and calculated during group-10 control rod scram, in the Bushehr Nuclear Power Plant (Iran) with a VVER-1000 reactor core. In the reactor core of Bushehr NPP, 10 groups of control rods are used of which, group-10 control rods contain the highest amount of injected negative reactivity in terms of quantity as compared to other groups of control rods. In this paper we explain impacts of negative reactivity, caused by a complete or minor scram of group-10 control rods, on thermoneutronic parameters of the VVER-1000 nuclear reactor core. It should be noted that through these calculations and by using the results, we can develop a sound understanding of impacts of this controlling element in optimum control of the reactor core and, on this basis, with careful attention and by gaining access to a reliable simulation (on the basis of results of calculations made in this survey) we can monitor the VVER-1000 reactor core through a smart control system. In continuation, for a more accurate survey and for comparing results of different calculation systems (point kinetic and diffusion), by using COSTANZA-R,Z calculation code (in which neutronic calculations are based on diffusion model) and using WIMS code at different areas and temperatures (for calculation of constant physical coefficients and temperature coefficients needed in COSTANZAR, Z code) for the VVER-1000 reactor core of Bushehr NPP, calculation of temperature distribution of fuel elements and coolant by using diffusion model is made in the course of group-10 control rods scram and afterwards.
Źródło:: Nukleonika; 2007, 52, 4; 159-165
0029-5922
1508-5791
Pojawia się w:: Nukleonika
Dostawca treści:: Biblioteka Nauki

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