Breeding Ratio for Fast Reactors

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

Breeding Ratio For Fast Reactors

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

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.

Innovative Core Design for Generation IV Sodium-Cooled Fast Reactors

2007 ◽  
pp. 48-62 ◽  
Author(s):  
L. Buiron ◽  
Ph. Dufour ◽  
G. Rimpault ◽  
G. Prulhiere ◽  
C. Thevenot ◽  
...  
Keyword(s):  
Fast Reactors

Study Ways to Increase KW in Fast Reactors with MOX Fuel Aiming for the Minimum of Sodium Void Reactivity Effect

2019 ◽  
Vol 12 (4) ◽  
pp. 50-61
Author(s):  
А. Harutyunyan ◽  
S.B. Vygovskiy ◽  
A. Khachatryan
Keyword(s):  
Fast Reactors ◽  
Reactivity Effect ◽  
Mox Fuel

Nuclear Non-Proliferation Security on Exportation of Fast Reactors with a Closed Fuel Cycle

Atomic Energy ◽  
2021 ◽  
Author(s):  
N. V. Gorin ◽  
N. P. Voloshin ◽  
Yu. I. Churikov ◽  
A. N. Chebeskov ◽  
V. P. Kuchinov ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Fast Reactors ◽  
Closed Fuel Cycle

scholarly journals Improved FAST Device for Inherent Safety of Oxide-Fueled Sodium-Cooled Fast Reactors

Energies ◽  
2021 ◽  
Vol 14 (15) ◽  
pp. 4610
Author(s):  
Ahmed Amin E. Abdelhameed ◽  
Chihyung Kim ◽  
Yonghee Kim
Keyword(s):  
Operating Temperature ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Inherent Safety ◽  
Oxide Fuel ◽  
Fast Reactors ◽  
Power Oscillation ◽  
Negative Reactivity ◽  
Oscillation Problem ◽  
Reactivity Coefficient

The floating absorber for safety at transient (FAST) was proposed as a solution for the positive coolant temperature coefficient in sodium-cooled fast reactors (SFRs). It is designed to insert negative reactivity in the case of coolant temperature rise or coolant voiding in an inherently passive way. The use of the original FAST design showed effectiveness in protecting the reactor core during some anticipated transients without scram (ATWS) events. However, oscillation behaviors of power due to refloating of the absorber module in FAST were observed during other ATWS events. In this paper, we propose an improved FAST device (iFAST), in which a constraint is imposed on the sinking (insertion) limit of the absorber module in FAST. This provides a simple and effective solution to the power oscillation problem. Here, we focus on an oxide fuel-loaded SFR that is characterized by a more negative Doppler reactivity coefficient and higher operating temperature than the metallic-loaded SFR cores. The study is carried out for the 1000 MWth advanced burner reactor with an oxide fuel-loaded core during postulated ATWS events that are unprotected transient over power, unprotected loss of flow, and unprotected loss of the heat sink. It was found that the iFAST device has promising potentials for protecting the oxide SFR core during the various studied ATWS events.

scholarly journals Statistical Analysis of Tritium Breeding Ratio Deviations in the DEMO Due to Nuclear Data Uncertainties

2021 ◽  
Vol 11 (11) ◽  
pp. 5234
Author(s):  
Jin Hun Park ◽  
Pavel Pereslavtsev ◽  
Alexandre Konobeev ◽  
Christian Wegmann
Keyword(s):  
Monte Carlo ◽  
Fusion Reactor ◽  
Nuclear Data ◽  
Data File ◽  
Nuclear Model ◽  
Model Parameters ◽  
Tritium Breeding ◽  
Tritium Breeding Ratio ◽  
Nuclear Data Library ◽  
Breeding Ratio

For the stable and self-sufficient functioning of the DEMO fusion reactor, one of the most important parameters that must be demonstrated is the Tritium Breeding Ratio (TBR). The reliable assessment of the TBR with safety margins is a matter of fusion reactor viability. The uncertainty of the TBR in the neutronic simulations includes many different aspects such as the uncertainty due to the simplification of the geometry models used, the uncertainty of the reactor layout and the uncertainty introduced due to neutronic calculations. The last one can be reduced by applying high fidelity Monte Carlo simulations for TBR estimations. Nevertheless, these calculations have inherent statistical errors controlled by the number of neutron histories, straightforward for a quantity such as that of TBR underlying errors due to nuclear data uncertainties. In fact, every evaluated nuclear data file involved in the MCNP calculations can be replaced with the set of the random data files representing the particular deviation of the nuclear model parameters, each of them being correct and valid for applications. To account for the uncertainty of the nuclear model parameters introduced in the evaluated data file, a total Monte Carlo (TMC) method can be used to analyze the uncertainty of TBR owing to the nuclear data used for calculations. To this end, two 3D fully heterogeneous geometry models of the helium cooled pebble bed (HCPB) and water cooled lithium lead (WCLL) European DEMOs were utilized for the calculations of the TBR. The TMC calculations were performed, making use of the TENDL-2017 nuclear data library random files with high enough statistics providing a well-resolved Gaussian distribution of the TBR value. The assessment was done for the estimation of the TBR uncertainty due to the nuclear data for entire material compositions and for separate materials: structural, breeder and neutron multipliers. The overall TBR uncertainty for the nuclear data was estimated to be 3~4% for the HCPB and WCLL DEMOs, respectively.

Prospects for fast reactors

1987 ◽  
Vol 92 (1-2) ◽  
pp. 57-71
Author(s):  
C. W. Blumfield
Keyword(s):  
Fast Reactor ◽  
High Energy ◽  
High Energy Density ◽  
Fast Reactors ◽  
Advantages And Disadvantages ◽  
Reactor Technology ◽  
European Collaboration ◽  
Fast Reactor Fuel ◽  
Fuel Reprocessing ◽  
Planned Activities

SynopsisThe background to recent major advances in the construction and operation of fast reactors is outlined with particular reference to the Dounreay Prototype Fast Reactor. The advantages and disadvantages of sodium as a coolant of the high energy density assembly are discussed and an account given of the consequences of a leak and the precautions taken against this eventuality. Attention is drawn to the safety aspects of the system. The economics of the plans for fuel reprocessing are explained and an account given of the progress in the fabrication of fast reactor fuel pins. Finally the environmental impact of present and planned activities on the Dounreay site is presented in the context of participation in the European Collaboration on Fast Reactor Technology and attention drawn to the importance of the planning inquiry in progress at Dounreay.

Possibility and Desirability of Using Electromagnetic Pumps in the Emergency Heat Removal Systems of Fast Reactors

Atomic Energy ◽  
2017 ◽  
Vol 122 (3) ◽  
pp. 178-184 ◽  
Author(s):  
E. V. Bogdanova ◽  
V. G. Zotov ◽  
S. A. Malkin ◽  
I. V. Vitkovskii ◽  
I. R. Kirillov ◽  
...  
Keyword(s):  
Heat Removal ◽  
Fast Reactors
Sign in / Sign up

Export Citation Format

Share Document