Calculation and measurement of neutron flux in internal parts of the VVER-1000 mock-up

Košťál, Michal; Švadlenková, Marie; Cvachovec, František; Jánský, Bohumil; Milčák, Ján; Juříček, Vlastimil; Rypar, Vojtěch; Kolros, Antonín; Losa, Evžen

doi:10.1016/j.anucene.2014.07.024

Cited by 7 publications

(2 citation statements)

References 16 publications

Supporting

Mentioning

Contrasting

Unclassified

Order By: Relevance

“…Table 3 show that, the DRAGON transport version 4 the absolute flux, for the present HEU fuel and the potential LEU fuels. Results obtained for existing HEU fuel were benchmarked and had good agreements with the previously published results (Khattab & Sulieman, 2010;CIAE, 1993). As may be seen from these tables.…”

Section: Neutron Flux Spectrumsupporting

confidence: 74%

Deterministic evaluation of safety parameters and neutron flux spectra in the MNSR research reactor using DRAGON-4 code

Zain

Hajjaji

Bardouni

et al. 2018

Journal of Radiation Research and Applied Sciences

View full text Add to dashboard Cite

Comparative studies for conversion of the fuel from high enriched uranium (HEU) to low enriched uranium (LEU) in the reactor (MNSR) were performed using DRAGON code deterministic. In this work the HEU fuel (UAl 4-Al, 90% enriched with Al-clad) and LEU (12.6% UO 2 enriched with zircaloy-4 alloy clad) cores were analyzed. This model was utilized in this work to calculate the neutron energy flux spectrum in the first inner and outer irradiation sites of the Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections has been evaluated from ENDF/B-VII library. The neutron fluxes has been calculated using 69 energy groups. The neutron energy flux for every group has been calculated dividing the neutron flux by the width of every energy group. As well, the neutron flux spectrum per unit lethargy has been calculated by multiplying the neutron energy flux spectra for every energy group by the average energy of every group. The thermal neutron flux was calculated by summing the neutron fluxes from (0.0e0.625) eV, the fast neutron flux was calculated by summing the neutron fluxes from (0.5e10) MeV for the existing HEU and potential LEU fuels. In this work, the effective multiplication factor (k eff), excess reactivity (P ex), ShutDown Margin (SDM), Control Rod Worth (CRW), Safety Reactivity Factor (SRF) and neutron energy flux spectrum for every group was calculated by dividing the neutron flux by the width of every energy group. In final, the results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels have been good agreements with the HEU results, compare with Monte Carlo MCNP4C code measured. Therefore, the LEU fuel has been validated as an appropriate option for fuel conversion of the MNSR in the future.

show abstract

Section: Neutron Flux Spectrumsupporting

confidence: 74%

Deterministic evaluation of safety parameters and neutron flux spectra in the MNSR research reactor using DRAGON-4 code

Zain

Hajjaji

Bardouni

et al. 2018

Journal of Radiation Research and Applied Sciences

View full text Add to dashboard Cite

show abstract

“…Однако более точное значение можно определить из сравнения результатов бенчмарковских экспериментов с их моделированием кодом MCNPX. В настоящее время такие работы выполняются, в качестве бенчмарка используются эксперименты на чешском реакторе нулевой мощности LR-0 [9]. Рис.…”

Section: выводыunclassified

Calculation of Neutron Fluence and Energy Release in WWER-1000 Structural Components Using Monte Carlo Method

Abdullaiev

Soldatov

Hann

et al. 2018

Nucl. and Rad. Safe.

View full text Add to dashboard Cite

The paper describes the methodology for the calculation of neutron fluence and energy release in WWER-1000 reactor cavity and baffle using the Monte Carlo MCNPX code. It formulates an approach to the simulation of a 3-D neutron source and conditions for the transport of neutrons and photons in the core. The research presents preliminary results of neutron fluence and energy release calculation in WWER-1000 reactor cavity and baffle.

show abstract

The effect of local power increase on neutron flux in internal parts of the VVER-1000 Mock-Up in LR-0 reactor

Košťál

Losa

Schulc

et al. 2018

Annals of Nuclear Energy

View full text Add to dashboard Cite

Calculation and measurement of neutron flux in internal parts of the VVER-1000 mock-up

Cited by 7 publications

References 16 publications

Deterministic evaluation of safety parameters and neutron flux spectra in the MNSR research reactor using DRAGON-4 code

Deterministic evaluation of safety parameters and neutron flux spectra in the MNSR research reactor using DRAGON-4 code

Calculation of Neutron Fluence and Energy Release in WWER-1000 Structural Components Using Monte Carlo Method

The effect of local power increase on neutron flux in internal parts of the VVER-1000 Mock-Up in LR-0 reactor

Contact Info

Product

Resources

About