Development and Application of the Neutronics/Thermal-Hydraulics Coupling Code for Safety Analysis of EBR-II Loss of Heat Sink Tests Without Scram

2016 ◽  
Author(s):  
Daogang Lu ◽  
Chao Guo ◽  
Danting Sui
Keyword(s):  
Fast Reactor ◽  
Heat Sink ◽  
Fuel Cycle ◽  
Transient Behavior ◽  
Safety Analysis ◽  
Cold Pool ◽  
Thermal Hydraulics ◽  
Closed Fuel Cycle ◽  
Zone Method ◽  
Gen Iv

In the GEN IV technology evaluations, the LMFBR (Liquid Metal Fast Breeder Reactor) system which includes SFR (Sodium-cooled Fast Reactor) and LFR (Lead-cooled Fast Reactor) was top-ranked in sustainability due to its closed fuel cycle and it is top-ranked in proliferation resistance and physical protection because it employs a long-life core. It is necessary to conduct the coupled neutronics and thermal-hydraulics simulation when the feedback effects are significant in the safety analysis of Anticipated Transients Without Scram (ATWS) in LMFBR. Thus, a neutronics-thermalhydraulics coupling code for safety analysis of LMFBR was developed and used to analyze whole-plant transient behavior of the Experimental Breeder Reactor II (EBR-II) under Loss of Heat Sink Without Scram (LOHSWS) tests in this paper. The two mixing zone method for cold pool coupled with SAC-CFR was used and the predicted results agree well with measurements which are taken from EBR-II LOHSWS test data.

scholarly journals Gas Cooled Fast Reactor Research and Development in the European Union

2009 ◽  
Vol 2009 ◽  
pp. 1-7 ◽  
Author(s):  
Richard Stainsby ◽  
Karen Peers ◽  
Colin Mitchell ◽  
Christian Poette ◽  
Konstantin Mikityuk ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Nuclear Energy ◽  
Fuel Cycle ◽  
Energy System ◽  
Outlet Temperature ◽  
Framework Programme ◽  
The European Union ◽  
Closed Fuel Cycle ◽  
Gen Iv ◽  
Actinide Transmutation

Gas-cooled fast reactor (GFR) research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV), that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5) GCFR project in 2000, through FP6 (2005 to 2009) and looking ahead to the proposed activities within the 7th Framework Programme (FP7).

scholarly journals MELT-III: a neutronics, thermal-hydraulics computer program for fast reactor safety analysis. Volume II

1974 ◽  
Author(s):  
A.E. Waltar ◽  
W.L. Partain ◽  
D.C. Kolesar ◽  
L.D. O'Dell ◽  
A. Jr. Padilla ◽  
...  
Keyword(s):  
Computer Program ◽  
Fast Reactor ◽  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Reactor Safety

Physics Studies of Weapons Plutonium Disposition in the Integral Fast Reactor Closed Fuel Cycle

1995 ◽  
Vol 121 (1) ◽  
pp. 17-31 ◽  
Author(s):  
R. N. Hill ◽  
D. C. Wade ◽  
J. R. Liaw ◽  
E. K. Fujita
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Closed Fuel Cycle

scholarly journals A Simplified Supercritical Fast Reactor with Thorium Fuel

2014 ◽  
Vol 2014 ◽  
pp. 1-9
Author(s):  
Peng Zhang ◽  
Kan Wang ◽  
Ganglin Yu
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Mechanical Design ◽  
Inherent Safety ◽  
Fuel Conversion ◽  
Reactor Physics ◽  
Mox Fuel ◽  
Feasible Option ◽  
Gen Iv ◽  
Simulation Scheme

Super-Critical water-cooled Fast Reactor (SCFR) is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.

Definition of Breeding Gain for the Closed Fuel Cycle and Application to a Gas-Cooled Fast Reactor

2007 ◽  
Vol 157 (2) ◽  
pp. 185-199 ◽  
Author(s):  
W. F. G. van Rooijen ◽  
J. L. Kloosterman ◽  
T. H. J. J. van der Hagen ◽  
H. van Dam
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Closed Fuel Cycle ◽  
Definition Of

scholarly journals Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
A. Rais ◽  
D. Siefman ◽  
G. Girardin ◽  
M. Hursin ◽  
A. Pautz
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Transient Behavior ◽  
Nuclear Data ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Nuclear Data Library ◽  
And Control

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

scholarly journals Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

2009 ◽  
Vol 2009 ◽  
pp. 1-9 ◽  
Author(s):  
W. F. G. van Rooijen ◽  
J. L. Kloosterman
Keyword(s):  
Nuclear Fuel ◽  
Fast Reactor ◽  
Fuel Cycle ◽  
Fission Products ◽  
Nuclear Fuel Cycle ◽  
Future Research ◽  
Minor Actinide ◽  
Closed Fuel Cycle ◽  
Decay Heat ◽  
Closed Nuclear Fuel Cycle

The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all Heavy Metal isotopes are to be recycled in the reactor. In this paper, an overview is presented of recent results obtained in the study of the closed fuel cycle and the influence of the addition of extra Minor Actinide (MA) isotopes from existing LWR stockpiles. In the presented work, up to 10% of the fuel was homogeneously replaced by an MA-mixture. The results are that addition of MA increases the potential of obtaining a closed fuel cycle. Reactivity coefficients generally decrease with increasing MA content. Addition of MA reduces the reactivity swing and allows very long irradiation intervals up to 10% FIMA with a small reactivity swing. Multirecycling studies show that a 600 MWth GCFR can transmute the MA from several PWRs. By a careful choice of the MA-fraction in the fuel, the reactivity of the fuel can be tuned to obtain a preset multiplication factor at end of cycle. Preliminary decay heat calculations show that the presence of MA in the fuel significantly increases the decay heat for time periods relevant to accidents (104–105s after shutdown). The paper ends with some recommendations for future research in this promising area of the nuclear fuel cycle.

scholarly journals A Preliminary Safety Analysis for the Prototype Gen IV Sodium-Cooled Fast Reactor

2016 ◽  
Vol 48 (5) ◽  
pp. 1071-1082 ◽  
Author(s):  
Kwi Lim Lee ◽  
Kwi-Seok Ha ◽  
Jae-Ho Jeong ◽  
Chi-Woong Choi ◽  
Taekyeong Jeong ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Safety Analysis ◽  
Gen Iv

scholarly journals Determination of equilibrium fuel composition for fast reactor in closed fuel cycle

2017 ◽  
Vol 153 ◽  
pp. 07034
Author(s):  
Mikhail Ternovykha ◽  
Georgy Tikhomirov ◽  
Yury Khomyakov ◽  
Igor Suslov
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Fuel Composition ◽  
Closed Fuel Cycle

Dynamic Simulation Research on Primary Coolant System of Sodium Cooled Fast Reactor

2012 ◽  
Author(s):  
Yang Yu ◽  
Yun Guo
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Safety Analysis ◽  
Operating Conditions ◽  
Cold Pool ◽  
Primary Coolant ◽  
Simulation Research ◽  
Flow And Heat Transfer ◽  
Comparison Of The Results ◽  
Pool Type

The Chinese Experimental Fast Reactor (CEFR) is a 65MWt/20MWe sodium cooled fast reactor. It is a pool-type reactor where the reactor and other internals such as pumps and intermediate heat exchangers (IHX) are immersed in a sodium pool. In this paper a one-dimensional dynamic code was developed to model the primary sodium circuit which included the reactor core, IHX, pumps, hot and cold pool etc. Moreover, the model of the property of sodium flow and heat transfer correlations was collected and compiled. This paper also discusses the mathematical models of various components of the primary sodium circuit, the numerical techniques to solve the models, the thermal-hydraulic studies of some design basis events such as the loss of primary pump or secondary pump accident etc, the comparison of the results of the code with that of the safety analysis report. Studies were conducted simulating both full and low power operating conditions. The dynamic code has been validated, and the results show that it has a benign response to some typical accidents. Finally, the model and code derived in this paper could be used in the safety analysis of pool-type sodium cooled fast reactor, and adopted in the development of CEFR simulation platform.

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