Nuclear Science and Technology

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Activity of radioactive isotopes in the offshore seawater at vicinity areas of Phu Quy, Phu Quoc and Bach Long Vy islands
Nuclear Science and Technology - Tập 11 Số 3 - 2021
In this work, activity of radioactive isotopes in some offshore seawater samples at vicinity areas of three islands has been investigated Phu Quy (Binh Thuan province), Phu Quoc (Kien Giang province), and Bach Long Vy (Hai Phong province). The ranges of radioactivity 226Ra, 238U, 137Cs, 90Sr, and 239,240Pu in seawater at three offshore islands are 2.09 ÷ 4.77 mBq/L, 2.33 ÷ 4.95 mBq/L, 0.98 ÷ 1.45 ...... hiện toàn bộ
#Radioactivity in seawater #monitoring #offshore island
Study on gamma-irradiation degradation of chitosan swollen in H₂O₂ solution and its antimicrobial activity for E. coli
Nuclear Science and Technology - Tập 4 Số 4 - 2014
Degradation of chitosan in swollen state with hydrogen peroxide solution (5% w/v) by γ-irradiation was investigated. Molecular weight (Mw) of irradiated chitosan was determined by gel permeation chromatography (GPC). Fourier transform infrared (FT-IR) and ultraviolet-visible (UV-vis) spectrawere analyzed to study the structure changes of degraded chitosan. The results showed that the chitosan of l...... hiện toàn bộ
#Chitosan #degradation #E. coli #gamma-irradiation #hydrogen peroxide
Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)
Nuclear Science and Technology - Tập 6 Số 1 - 2016
A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA).The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solve...... hiện toàn bộ
#coupled neutronics thermal-hydraulics #nodal method #adiabatic method #improved quasi-static method #PWR #reactivity initiated accident #benchmark verification
CALCULATION OF THE MODERATOR TEMPURATURE COEFFICIENT OF REACTIVITY FOR MINIATURE NEUTRON SOURCE REACTORS
Nuclear Science and Technology - Tập 1 Số 4 - 2011
This paper presents calculated results of the moderator temperature coefficient of reactivity through global reactor calculations and the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (MNSR). In this study, the group constants were calculated with the WIMSD code and the global reactor calculation was accomplished by the CITATION code. This...... hiện toàn bộ
#MNSR #the moderator temperature coefficient of reactivity #group constants #lattice cell calculation #global reactor calculation
Recent updates on perturbation analysis of the KALIMER-600 TRU burner
Nuclear Science and Technology - Tập 3 Số 3 - 2013
Sodium void worth and the Doppler coefficient, which are very important safety parameters in safety analysis of the KALIMER-600 TRU burner, should be carefully evaluated. A perturbation analysis for the TRU burner has thus been performed using the perturbation method, not only to make sure the reactivity feedbacks meet the predetermined design targets but also to obtain insight into the actual phy...... hiện toàn bộ
#TRU burner #sodium void worth #Doppler effect #perturbation method #PERT-K #DIF3D #isotope-wise reactivity
Enrichment determination of low – enriched uranium material by gamma spectroscopic method
Nuclear Science and Technology - Tập 4 Số 2 - 2014
In this work the non - destructive gamma spectroscopic method for determination of uranium enrichment is presented. In order to determine the uranium enrichment the activity ratios of 234U/235U and 238U/235Uwere measured. The activity ratios234U/235U and 238U/235U were determined by using intrinsic efficiency calibration. As a test of this method, low - enriched uranium standard was measured, the ...... hiện toàn bộ
#Uranium enrichment #gamma-spectrometry #intrinsic efficiency calibration #MGA method
Evaluation of slip ratio correlations in two-phase flow
Nuclear Science and Technology - Tập 10 Số 1 - 2020
Critical flow is one of the essential parameters in LOCA accident analysis in which pressure difference is very high. Void fraction (α), in another term, slip ratio, s, is the key parameter that could affect critical flow prediction. Henry-Fauske (HF) model is the model for critical flow calculation existing in current computer codes such as MARS, RELAP, TRACE. However, the limitation of this mode...... hiện toàn bộ
#Void fraction #slip ratio #critical flow
Establishment of neutron dose calibration fields based on a ²⁵²Cf source by Monte-Carlo method
Nuclear Science and Technology - Tập 6 Số 2 - 2016
This paper presents calculation results based on Monte-Carlo method to select an appropriate neutron moderator and design four configurations for a 252Cf irradiation system. These configurations provide six neutron spectra with the various average energies (1.04 MeV, 1.38 MeV, 1.69 MeV, 2.05 MeV, 2.46 MeV and 2.93 MeV) suitable for the calibration of neutron survey meters and personal dosimeters.
#²⁵²Cf irradiation system #Monte-Carlo method #neutron moderator #neutron spectrum #neutron dose calibration
Application of correlation pattern recognition technique for neutron– gamma discrimination in the EJ-301 liquid scintillation detector
Nuclear Science and Technology - Tập 8 Số 2 - 2018
The ability to distinguish between neutrons and gamma-rays is important in the fast - neutron detection, especially when using the scintillation detector. A dual correlation pattern recognition (DCPR) method that was based on the correlation pattern recognition technique has been developed for classification of neutron/gamma events from a scintillation detector. In this study, an EJ-301 liquid sci...... hiện toàn bộ
#Correlation pattern recognition method #EJ301 detector #pulse shape discrimination (PSD)
Verification of TVS-2006 fuel rod design of VVER-AES2006 reactor under steady-state operating condition Using FRAPCON-3.5 code
Nuclear Science and Technology - Tập 5 Số 2 - 2015
The purpose of this paper is to discuss the independent verification of TVS-2006 fuel rod design used in VVER-AES2006 reactor (Novovoronezh NPP-2 Power, Unit 1), based on the acceptance criteria and the reference data given in the Preliminary Safety Analysis Report of the State Research, Design, Construction and Survey Institute “Atomenergoproekt” (PSAR) and the operation of VVER-1000 ...... hiện toàn bộ
#Nuclear fuel rod design #fuel behaviour #design verification #acceptance criteria
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