Mahdi Ghaed Rahmati; Mostafa hasanzadeh; Seyed Amir Hossein Feghhi
Abstract
Kinetic and neutronic parameters play an important role in analysis of reactors dynamic behavior. Some of these parameters include: effective multiplication factor (keff), reactivity (ρ), neutron flux as well as power spatial distributions, effective delayed neutron fraction (βeff) and prompt ...
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Kinetic and neutronic parameters play an important role in analysis of reactors dynamic behavior. Some of these parameters include: effective multiplication factor (keff), reactivity (ρ), neutron flux as well as power spatial distributions, effective delayed neutron fraction (βeff) and prompt neutron lifetime (lp ). In this work, Monte Carlo modeling and analysis of Isfahan MNSR is performed for calculation of the kinetic and neutronic parameters of using MCNPX2.6 code, slope fit and perturbation methods. Relative differences between results of the MCNPX2.6 code in calculation of the ρ and βeff and the reference values are about 0.5% and 2.1%, respectively. The relative differences between the results of the slope fit and perturbation methods and MCNPX2.6 code in calculation of the parameter with the reference values are about 17.6%, 4.8% and 29.19%, respectively. Therefore, the results of these research show that the MCNPX2.6 code is suitable for calculation of the reactor kinetic parameters such as the βeff, while the perturbation method is a simple and convenient method for calculating the .
Mohammad Javad Safari
Abstract
It is well-known that response function of organic scintillation detectors does not appear with photopeaks. Instead, their dominant feature is a continuum, usually called the Compton edge that innately exposes the resolution characteristics of detection system. ...
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It is well-known that response function of organic scintillation detectors does not appear with photopeaks. Instead, their dominant feature is a continuum, usually called the Compton edge that innately exposes the resolution characteristics of detection system. While, accurate characterization of Compton edge is crucial for calibration purposes, it is also in charge of elaborating the energy resolution of detector. This paper presents a simple method for accurate characterization of the Compton edge in organic scintillation detectors. The method is based on the fact that differentiating the response function leads to accurate estimation of the constituting functions. The differentiation method, in addition to the location of the Compton edge, gives insights into the parameters of the folded Gaussian function which could lead to depict the energy resolution. Moreover, it is observed that the uncorrelated noise in the measurement of the response function does not impose significant uncertainties in the evaluations, so it could preserve its functionality even in lower-quality measurements. By simulation of the bounded electrons and considering the Doppler effects, we are able to demonstrate -the first ever- estimation for intrinsic Doppler resolution of an organic plastic scintillator. Even though, this possibility is an immediate result of benefiting the presented method for analysis of the Compton continua.
Mohsen Salehi; Mohsen Shayesteh
Abstract
The analysis deals with the assessment of best estimate code RELAP5/SCDAP mod3.4 in the simulation of double-ended loss coolant accident as a LBOCA, 4 in break as a SBLOCA an SBO accident with considering except accumulator water where no core cooling water systems are available. ...
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The analysis deals with the assessment of best estimate code RELAP5/SCDAP mod3.4 in the simulation of double-ended loss coolant accident as a LBOCA, 4 in break as a SBLOCA an SBO accident with considering except accumulator water where no core cooling water systems are available. The reference plant is SURRY nuclear power plant as a Westinghouse three-loop nuclear power plant. In order to mitigation accident, the in-vessel retention strategy was investigated for the prevention of lower plenum failure. It has been concluded that during the SBLOCA, LBLOCA conditions bottom of active fuel is uncovered at 6340 s and 2160 s, respectively. It occurred for two times at 11650 s and 15608 s in SBO. At 6792 s and 57002 s in the LBLOCA and SBO due to reaching melting point and in the SBLOCA at 15215 s due to lower plenum creep rupture, failure of the reactor pressure vessel occurred. The results show that hydrogen production in the SBO is more than the other two cases. For the prevention of the lower plenum failure, the in-vessel molten material retention strategy is investigated as a passive system. The results show that lower plenum heat flux can be kept below the critical heat flux and its integrity is preserved in two cases of this analysis.
Afshin Hedayat
Abstract
Both of small and medium sized reactors and small modular reactors are called SMRs. They are reviewed and discussed in this paper, particularly integral Pressurized Water Reactors (iPWRs). Studies show that PWRs are the most interested, designed and constructed ...
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Both of small and medium sized reactors and small modular reactors are called SMRs. They are reviewed and discussed in this paper, particularly integral Pressurized Water Reactors (iPWRs). Studies show that PWRs are the most interested, designed and constructed nuclear reactor type worldwide. Some innovative small modular PWRs like the MASLWR, NuScale, CAREM-25, SMART and ACP-100 have several outstanding characteristics to be promisingly recognized as near term options of the next generation of small modular PWRs. They have several inherently safety features and improved passive safety system. They require smaller infrastructure and capital costs. They can be also developed rapidly in different and independent modular unites even for remote area or outlands without required infrastructure or electrical grids. It should be noted that new modern economy strategies like the Return of Investment (ROI) issues may advice medium or large reactors rather than small units for developed and industrial countries while small modular plans can be much more interesting and accessible for new comers or even developing countries. Finally, multi-applicability is an appropriate solution to develop expensive nuclear power plants economically as well as multi-purpose research reactors (especially by means of small modular iPWRs).
Nahid Sohebi; Mahdi Eshghi; Majid Hamzavi; Mohsen Bigdeli
Abstract
In the present work, the eigenvalue and eigenvector has been obtained by the Bohr Hamiltonian for even-even nuclei. The competition between γ-stable and γ-rigid collective motions has been created in the presence of the rigidity parameter. The β-part ...
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In the present work, the eigenvalue and eigenvector has been obtained by the Bohr Hamiltonian for even-even nuclei. The competition between γ-stable and γ-rigid collective motions has been created in the presence of the rigidity parameter. The β-part of the collective potential has been chosen to be equal to the generalized Hulthen potential, while the γ-angular part of the problem is associated with Ring-shaped potential around the γ=π/6 and the Harmonic oscillation around the γ=0. In both cases, the effect of rigidity and free parameters on energy spectrum of Os-180, Dy-162, Gd-160, Ru-100, Pd-114, and Xe-124 nuclei have been investigated. Also, the rates of B(E2) transition have been calculated and compared with experimental data. This model has an appropriate description of energy spectra for the mentioned nuclei.
Afshin Hedayat
Abstract
Fast neutron irradiation is one of the most strategic radiation applications of research reactors. Usually, it is performed around the reactor core containing lower neutron flux. In this paper, a hybrid object has been introduced and analyzed to enhance irradiating ...
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Fast neutron irradiation is one of the most strategic radiation applications of research reactors. Usually, it is performed around the reactor core containing lower neutron flux. In this paper, a hybrid object has been introduced and analyzed to enhance irradiating applications of the fast neutrons in the core of a Material Testing Reactor (MTR). The tool includes an old-type low-consumed HEU control fuel element, a dry channel, and a Cd filter. It is supposed to be installed at the internal neutron trap (D4 positions) of TRR core configuration. Calculating results are very promising for using the proposed tool to increase neutron fluxes, reduce thermal and epi-thermal neutron fluxes, and shift the neutron spectrum toward the fast neutron region (hardening effect) at the chosen irradiating location. Primary safety parameters are also checked and passed successfully. Furthermore, there are also some other presented safety items which must be checked carefully and conservatively in order to refabricate and install such a irradiating tool in an in-core location of a MTR.
