Effect of Pin and Subassembly Heterogenity in Sodium Void Worth Calculations

2010 ◽  
Author(s):  
G. Raghu Kumar ◽  
C. P. Reddy ◽  
V. Sathyamoorthy
Keyword(s):  
Monte Carlo Code ◽  
Reactor Accident ◽  
Fast Reactors ◽  
Neutron Spectra ◽  
Energy Scenario ◽  
Higher Power ◽  
Breeding Ratio ◽  
Short Doubling Time ◽  
Undesirable Feature ◽  
Total Sodium

Metal fuelled sodium cooled fast reactors are known to have high breeding ratio and short doubling time. Due to these features they play a very important role in the energy scenario, where higher power growth is required. Large sodium cooled fast reactors have positive sodium void coefficient, which is considered to be undesirable feature even though reactor safety can be established for all design based accidents like loss of flow and transient over power accidents. These types of fast reactors, which have harder neutron spectra are having higher sodium void coefficient compared to ceramic fuelled fast reactors. In many of the safety analysis the total sodium void is calculated and it is used in the safely evaluation. However the sodium in the metal fuelled reactor has got three parts, namely bonding sodium, coolant sodium and the sodium in the inter space of subassembly hexagonal cans. In the reactor accident scenario the behavior of these three components of sodium will be different and will effect the sequence of the accident. The finer details, of the fuel sub assembly, are modeled in to Monte Carlo code and the sodium void coefficient is calculated for each of the component for the fuel zones. This study will be helpful in improving safety of the reactor and also reducing the conservatism in the safely features.

Breeding Ratio For Fast Reactors

1971 ◽  
Vol 10 (2) ◽  
pp. 101-102
Author(s):  
P. Goldschmidt
Keyword(s):  
Fast Reactors ◽  
Breeding Ratio

scholarly journals Neutron and photon scattering properties of high density concretes used in radiation therapy facilities: A Monte Carlo study

2017 ◽  
Vol 23 (3) ◽  
pp. 61-65 ◽  
Author(s):  
Asghar Mesbahi ◽  
Rezvan Khaldari
Keyword(s):  
Monte Carlo ◽  
Radiation Therapy ◽  
Monte Carlo Study ◽  
High Density ◽  
Photon Beam ◽  
Scattered Photon ◽  
Monte Carlo Code ◽  
Photon Scattering ◽  
Neutron Spectra ◽  
Ordinary Concrete

Abstract In the current study the neutron and photon scattering properties of some newly developed high density concretes (HDCs) were calculated by using MCNPX Monte Carlo code. Five high-density concretes including Steel-Magnetite, Barite, Datolite-Galena, Ilmenite-ilmenite, Magnetite-Lead with the densities ranging from 5.11 g/cm3 and ordinary concrete with density of 2.3 g/cm3 were studied in our simulations. The photon beam spectra of 4 and 18 MV from Varian linac and neutron spectra of clinical 18 MeV photon beam was used for calculations. The fluence of scattered photon and neutron from all studied concretes was calculated in different angles. Overall, the ordinary concrete showed higher scattered photons and Datolite-Galena concrete (4.42 g/cm3) had the lowest scattered photons among all studied concretes. For neutron scattering, fluence at the angle of 180 was higher relative to other angles while for photons scattering fluence was maximum at 90 degree. The scattering fluence for photons and neutrons was dependent on the angle and composition of concrete. The results showed that the fluence of scattered photons and neutrons changes with the composition of high density concrete. Also, for high density concretes, the variation of scattered fluence with angle was very pronounced for neutrons but it changed slightly for photons. The results can be used for design of radiation therapy bunkers.

Breeding Ratio for Fast Reactors

1970 ◽  
Vol 9 (4) ◽  
pp. 450-451 ◽  
Author(s):  
P. Goldschmidt
Keyword(s):  
Fast Reactors ◽  
Breeding Ratio

scholarly journals Thermal Energetic Reactor with High Reproduction of Fission Materials

2012 ◽  
Vol 2012 ◽  
pp. 1-5
Author(s):  
Vladimir M. Kotov
Keyword(s):  
Zirconium Alloys ◽  
Fast Reactors ◽  
Additional Energy ◽  
Production Scale ◽  
Neutron Sources ◽  
Liquid Salt ◽  
Stage Of Development ◽  
Initial Stage ◽  
Higher Power ◽  
Channel Type

Existing thermal reactors are energy production scale limited because of low portion of raw uranium usage. Fast reactors are limited by reprocessing need of huge mass of raw uranium at the initial stage of development. The possibility of development of thermal reactors with high fission materials reproduction, which solves the problem, is discussed here. Neutron losses are decreased, uranium-thorium fuel with artificial fission materials equilibrium regime is used, additional in-core and out-core neutron sources are used for supplying of high fission materials reproduction. Liquid salt reactors can use dynamic loading regime for this purpose. Preferable construction is channel type reactor with heavy water moderator. Good materials for fuel element shells and channel walls are zirconium alloys enriched by90Zr. Water cooled reactors with usage 12% of raw uranium and liquid metal cooled reactors with usage 25% of raw uranium are discussed. Reactors with additional neutron sources obtain full usage of raw uranium with small additional energy expenses. On the base of thermal reactors with high fission materials reproduction world atomic power engineering development supplying higher power and requiring smaller speed of raw uranium mining, than in the variant with fast reactors, is possible.

scholarly journals Simulation of the neutron spectra evolution with GEANT4 Monte Carlo code

2021 ◽  
Vol 25 (2) ◽  
Author(s):  
V. P. Smolyar ◽  
A. O. Mileva ◽  
V. O. Tarasov ◽  
H. H. Neboha ◽  
V. D. Rusov
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Code ◽  
Neutron Spectra

scholarly journals Nuclear Data Uncertainty Quantification and Propagation for Safety Analysis of Lead-Cooled Fast Reactors

2020 ◽  
Vol 2020 ◽  
pp. 1-14
Author(s):  
Ishita Trivedi ◽  
Jason Hou ◽  
Giacomo Grasso ◽  
Kostadin Ivanov ◽  
Fausto Franceschini
Keyword(s):  
Cross Sections ◽  
Nuclear Data ◽  
Safety Analysis ◽  
Data Uncertainty ◽  
Monte Carlo Code ◽  
Capture Cross ◽  
Fast Reactors ◽  
Safety Margins ◽  
The Impact ◽  
Capture Cross Sections

In this study, the Best Estimate Plus Uncertainty (BEPU) approach is developed for the systematic quantification and propagation of uncertainties in the modelling and simulation of lead-cooled fast reactors (LFRs) and applied to the demonstration LFR (DLFR) initially investigated by Westinghouse. The impact of nuclear data uncertainties based on ENDF/B-VII.0 covariances is quantified on lattice level using the generalized perturbation theory implemented with the Monte Carlo code Serpent and the deterministic code PERSENT of the Argonne Reactor Computational (ARC) suite. The quantities of interest are the main eigenvalue and selected reactivity coefficients such as Doppler, radial expansion, and fuel/clad/coolant density coefficients. These uncertainties are then propagated through safety analysis, carried out using the MiniSAS code, following the stochastic sampling approach in DAKOTA. An unprotected transient overpower (UTOP) scenario is considered to assess the effect of input uncertainties on safety parameters such as peak fuel and clad temperatures. It is found that in steady state, the multiplication factor shows the most sensitivity to perturbations in 235U fission, 235U ν, and 238U capture cross sections. The uncertainties of 239Pu and 238U capture cross sections become more significant as the fuel is irradiated. The covariance of various reactivity feedback coefficients is constructed by tracing back to common uncertainty contributors (i.e., nuclide-reaction pairs), including 238U inelastic, 238U capture, and 239Pu capture cross sections. It is also observed that nuclear data uncertainty propagates to uncertainty on peak clad and fuel temperatures of 28.5 K and 70.0 K, respectively. Such uncertainties do not impose per se threat to the integrity of the fuel rod; however, they sum to other sources of uncertainties in verifying the compliance of the assumed safety margins, suggesting the developed BEPU method necessary to provide one of the required insights on the impact of uncertainties on core safety characteristics.

Measurement of neutron spectra in fast reactors

1971 ◽  
Vol 30 (6) ◽  
pp. 678-679
Author(s):  
Yu. A. Kazanskii
Keyword(s):  
Fast Reactors ◽  
Neutron Spectra
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