Power calculation of VVER-1000 reactor using a thermal method, applied to primary–secondary circuits

In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were performed using MCNP6 code together with both ENDF/B-VII.1 and ENDF/B-VIII libraries. The effect of thorium introduction on the neutronic parameters of the VVER-1000 reactor was discussed. The reference core was initially filled with enriched uranium oxide fuel and then fueled with uranium-thorium fuel. The calculations determine the delayed neutron fraction βeff, the temperature reactivity coefficients, the fuel consumption, and the production of the transuranic elements during reactor operation. βeff and the Doppler coefficient (DC) are found to be in agreement with the design values. It is found that the core loaded with uranium and thorium has lower delayed neutron fraction than the uranium oxide core. The moderator temperature coefficients of the uranium-thorium core are found to be higher than those of the uranium core. Results indicated that thorium has lower production of minor actinides (MAs) and transuranic elements (mainly plutonium isotopes) compared with the relatively large amounts produced from the uranium-based fuel UO2.

show abstract

“…e brief details of the core FAs were illustrated by Aghaie et al [9]. e reactor core, FA, and fuel rod characteristics are shown in Table 1 [10].…”

Section: The Vver-1000 Reactor Core Descriptionmentioning

confidence: 99%

Neutronic Performance of the VVER-1000 Reactor Using Thorium Fuel with ENDF Library

Reda

Gomaa

Bashter

et al. 2021

Science and Technology of Nuclear Installations

View full text Add to dashboard Cite

show abstract

“…At present the accidents for VVER-1000 reactors are thoroughly investigated where the researchers investigated the basis design improvement for VVER-1000 reactors [5][6][7][8][9][10][11][12][13][14][15]. In reviewing the literature, RELAP5 loop code found a validated tool for thermal hydraulic analysis of nuclear reactors [16,17].…”

Section: Introductionmentioning

confidence: 99%

Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

Tabadar

Jabbari

Khaleghi

et al. 2018

Alexandria Engineering Journal

Self Cite

View full text Add to dashboard Cite

Removing the residual heat from a nuclear reactor is an important safety aspect of thermal hydraulic analysis. In this study, a typical VVER-1000 reactor residual heat removal system has been evaluated using RELAP5 thermal hydraulic loop code during cool-down. Reactor cooling down starts from hot state temperature and then continues to the cool-down stages with 130°C and 70°C, respectively. The second stage of cooling down is the boundary of the reactor repair condition. Main cooling pump head, steam generator (SG) water level, system pressure, the level of coolant in the pressurizer (PRZ), and the temperature of fuel element are examined in a steady state condition. PRZ level, primary circuit and secondary circuit pressure/temperature, and SG water level are evaluated during 32,000 s after cool-down scenario. By comparison, it concluded that the results of RELAP5 code are in agreement with plant experimental data and final safety analysis report (FSAR). Thus, it is proved that the studied reactor is capable to remove the residual heat generated during shutdown. Moreover, RELAP5 is properly recommended for analysis of the VVER-1000 pressurized water reactor during cool-down. Ó 2017 Faculty of Engineering, Alexandria University. Production and hosting by Elsevier B.V. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).

show abstract

The safety improvement of VVER-1000 NPP against SBO accident using portable air-cooled diesel generator

Jabbari

Hadad

Pirouzmand

2020

Annals of Nuclear Energy

View full text Add to dashboard Cite

Power calculation of VVER-1000 reactor using a thermal method, applied to primary–secondary circuits

Cited by 7 publications

References 4 publications

Neutronic Performance of the VVER-1000 Reactor Using Thorium Fuel with ENDF Library

Neutronic Performance of the VVER-1000 Reactor Using Thorium Fuel with ENDF Library

Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

The safety improvement of VVER-1000 NPP against SBO accident using portable air-cooled diesel generator

Contact Info

Product

Resources

About