scholarly journals Characteristics of a gas-cooled fast reactor with minor actinide loading

2021 ◽  
Vol 8 (2) ◽  
pp. 1-9
Author(s):  
Hoai Nam Tran ◽  
Yasuyoshi Kato ◽  
Van Khanh Hoang ◽  
Sy Minh Tuan Hoang
Keyword(s):  
Fast Reactor ◽  
Supercritical Co2 ◽  
Cycle Length ◽  
Minor Actinide ◽  
Co2 Gas ◽  
Control Rod ◽  
The Core ◽  
Thermal Output ◽  
Long Life ◽  
Gas Turbine Cycle

This paper presents the neutronics characteristics of a prototype gas-cooled (supercritical CO2-cooled) fast reactor (GCFR) with minor actinide (MA) loading in the fuel. The GCFR core is designed with a thermal output of 600 MWt as a part of a direct supercritical CO2 (S-CO2) gas turbine cycle. Transmutation of MAs in the GCFR has been investigated for attaining low burnup reactivity swing and reducing long-life radioactive waste. Minor actinides are loaded uniformly in the fuel regions of the core. The burnup reactivity swing is minimized to 0.11% ∆k/kk’ over the cycle length of 10 years when the MA content is 6.0 wt%. The low burnup reactivity swing enables minimization of control rod operation during burnup. The MA transmutation rate is 42.2 kg/yr, which is equivalent to the production rates in 7 LWRs of the same electrical output.

scholarly journals Impact of the Core Minor Actinide Content on Fast Reactor Reactivity Coefficients

2011 ◽  
Vol 48 (4) ◽  
pp. 628-634 ◽  
Author(s):  
Giuseppe PALMIOTTI ◽  
Massimo SALVATORES ◽  
Monchai ASSAWAROONGRUENGCHOT
Keyword(s):  
Fast Reactor ◽  
Minor Actinide ◽  
The Core ◽  
Reactivity Coefficients

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Massimo Sarotto ◽  
Gabriele Firpo ◽  
Anatoly Kochetkov ◽  
Antonin Krása ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Control Rod ◽  
The Core ◽  
Major Interest ◽  
Reactor Spectrum ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Sensitivity And Uncertainty Analysis

Abstract During the EURATOM FP7 project FREYA, a number of experiments were performed in a critical core assembled in the VENUS-F zero-power reactor able to reproduce the ALFRED lead-cooled fast reactor spectrum in a dedicated island. The experiments dealt with the measurements of integral and local neutronic parameters, such as the core criticality, the control rod and the lead void reactivity worth, the axial distributions of fission rates for the nuclides of major interest in a fast spectrum, the spectral indices of important actinides (238U, 239Pu, 237 Np) with respect to 235U. With the main aim to validate the neutronic codes adopted for the ALFRED core design, the VENUS-F core and its characterization measurements were simulated with both deterministic (ERANOS) and stochastic (MCNP, SERPENT) codes, by adopting different nuclear data libraries (JEFF, ENDF/B, JENDL, TENDL). This paper summarizes the main results obtained by highlighting a general agreement between measurements and simulations, with few discrepancies for some parameters that are discussed here. Additionally, a sensitivity and uncertainty analysis was performed with deterministic methods for the core reactivity: it clearly indicates that the small over-criticality estimated by the different codes/libraries resulted to be lower than the uncertainties due to nuclear data.

scholarly journals Calculation of excore detector weighting functions for a sodium-cooled TRU burner mockup using MCNP5

2015 ◽  
Vol 5 (2) ◽  
pp. 15-25
Author(s):  
Viet Ha Pham Nhu ◽  
Min Jae Lee ◽  
Sunghwan Yun ◽  
Sang Ji Kim
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Detector Response ◽  
Weighting Functions ◽  
Control Rod ◽  
The Core ◽  
Power Regulation ◽  
Fuel Assemblies ◽  
Control Rods ◽  
Horizontal Layers

Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A.

Modeling of Reactivity Effects and Transient Behaviour of Large Sodium Fast Reactor

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
Core Model ◽  
Primary System ◽  
Control Rod ◽  
Transient Behaviour ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Reactivity Coefficients ◽  
Core Conditions ◽  
Point Kinetics

Abstract In the paper the reactivity characteristics of the core of the large sodium fast reactor Superphenix (SPX) were evaluated and compared with available experimental data. The analysis was performed using the TRACE system code modified for the fast reactor applications. The simplified core model was developed aiming to overcome the lack of detailed information on design and realistic core conditions. Point Kinetics neutronic model with all relevant reactivity feedbacks was used to calculate transient power. The paper focuses on challenging issue of modelling of the transient thermal responses of primary system structural elements resulting in reactivity feedbacks specific to such large fast reactor which cannot be neglected. For these effects, the model was equipped with dedicated heat structures to reproduce important feedbacks due to vessel wall, diagrid, strongback, control rod drive lines thermal expansion. Peculiarly, application of the model was considered for a whole range of core conditions from zero power to 100% nominal. The developed core model allowed reproducing satisfactorily the core reactivity balance between zero power at 180?C and full power conditions. Additionally, the reactivity coefficients k, g, h at three power levels were calculated and satisfactory agreement with experimental measurements was also observed. The study demonstrated feasibility of application of relatively simple model with adjusted parameters for analysis of different conditions of very complex system.

Development of Numerical Estimation Method Using Spatial Connection Methodology for Thermal Striping in Upper Plenum of Reactor Vessel of an Advanced Loop-Type Sodium-Cooled Fast Reactor in Japan

2017 ◽  
Author(s):  
Masaaki Tanaka ◽  
Satoshi Murakami
Keyword(s):  
Numerical Simulation ◽  
Fast Reactor ◽  
Estimation Method ◽  
Local Area ◽  
Outer Side ◽  
Numerical Estimation ◽  
Analysis Model ◽  
Control Rod ◽  
The Core ◽  
Thermal Striping

Thermal striping on the core instrumentation plate (CIP) at the bottom of the upper internal structure (UIS) of an advanced loop-type sodium-cooled fast reactor in Japan (Advanced-SFR) has been numerically investigated. At the top of the core below the CIP, the sodium at high temperature flows out from the fuel subassemblies (FSs) and the sodium at low temperature flows out from the primary control rod (PCR) and backup control rod (BCR) channels, and also the radial blanket fuel subassemblies (RBFSs) at the outer side of the core. In order to predict the thermal striping on the CIP caused by mixing fluids at different temperatures from the FSs, the PCR and the BCR channels, and the RBFSs, a numerical estimation method using a spatial connection methodology between the upper plenum analysis and the local area analysis for the target area has been developed. By using the connection methodology, the numerical simulation considering the influence of the transversal flow in the UIS and the external flow around the UIS in the upper plenum can be performed to improve the accuracy of the estimation results. In this paper, the outline of the spatial connection methodology including data transfer technique from the upper plenum analysis to the local area analysis was described. As a validation process, numerical simulation of the water experiment using the test apparatus named TAFUT which was 1/3-scaled 1/6 partial model of the upper plenum of the Advanced-SFR was performed to confirm applicability of the spatial connection methodology to a practical thermal striping problem. The numerical result of temperature distribution was compared with the measured result in TAFUT experiment. Additionally, mesh sensitivity of the local area analysis model to the numerical results was indicated by using a small and a large area models in order to suggest an appropriate local area analysis model.

Preliminary Core Design of the Long Life S-CO2 Cooled Fast Reactor With 300WMth

2014 ◽  
Author(s):  
Baolin Liu ◽  
Hongchun Wu ◽  
Youqi Zheng ◽  
Liangzhi Cao ◽  
Xianbao Yuan
Keyword(s):  
Fast Reactor ◽  
Nuclear Power ◽  
Power Plants ◽  
Single Channel ◽  
Reactor Core ◽  
Fuel Temperature ◽  
The Core ◽  
Long Life ◽  
High Conversion Ratio ◽  
Thermal Hydraulic Analysis

Gas cooled fast reactors are one of the Generation 4 nuclear power plants with hard neutron spectrum and high conversion ratio. In the study a long life Supercritical CO2 (S-CO2) cooled fast reactor core design with 300 MWth is presented. Physical calculation was carried out based on Dragon and CITATION, and thermal hydraulic analysis was performed based on the single channel code. The MOX fuel was utilized in the core design, and the tube-in-duct (TID) assembly was chosen for its excellent characteristics. According to the physical and thermal hydraulic coupling calculation, the reactor in the study can be operated with 300MWth for 20Ys without shuffling or refueling. Through the core life power peaking was kept relatively low, and the fuel temperature was kept below the 1800 degree centigrade.

Modeling the Molten Salt Reactor Experiment With the GOTHIC Code

2020 ◽  
Author(s):  
Rodney Harvill ◽  
Jeff Lane ◽  
John Link ◽  
Anita Gates ◽  
Tom Kindred
Keyword(s):  
Measured Data ◽  
Reactor Power ◽  
Molten Salt Reactor ◽  
Delayed Neutron ◽  
Control Rod ◽  
Decay Heat ◽  
The Core ◽  
Thermal Output ◽  
Fuel Salt ◽  
Good Agreement

Abstract GOTHIC 8.3(QA) includes capabilities for modeling advanced, non-light water cooled reactors. Important capabilities introduced in GOTHIC 8.3(QA) include fluid property tables for various molten salts, an enhancement to the tracer tracking module to allow radioactive decay energy to be released locally in the carrier fluid and other improvements to the neutron kinetics module. With these new capabilities in place, GOTHIC is used to benchmark steady-state and transient conditions in the Molten Salt Reactor Experiment (MSRE), which operated at Oak Ridge National Laboratory from 1965 to 1969. In this experimental reactor, UF4 fuel was dissolved in molten fluoride salt, and criticality could be achieved only in the graphite moderated core. An air-cooled radiator transferred fission and decay heat to the environment. The design thermal output of the MSRE was 10 MWt, but the radiator design limited the output to 8 MWt. The original design parameters neglected the impact of decay heat on system temperatures. GOTHIC is used to benchmark system operating parameters at both the 10 MWt design condition and the 8 MWt operating condition, both with and without decay heat. The cases that include decay heat apply 7% of the nominal thermal output using the eleven decay heat precursors from ASB 9-2 as tracers. The results of the benchmark exhibit good agreement with design and operating data and demonstrate heat-up due to decay heat in the fuel salt outside the core. In the MSRE, delayed neutron precursors are not confined to the core because the fuel and fission products flow through the system. As a result, there are different values for (effective) delayed neutron fraction with and without flow, and the decay of delayed neutron precursors outside the core under full-flow conditions reduces reactivity by 0.212 % δk/k. Zero power physics testing included fuel salt pump start-up and coast-down transients with a control rod automatically moving to maintain criticality. The control rod motion calculated by GOTHIC is a reasonable match to measured data from these transients. Low power testing included a natural convection transient with no control rod motion such that reactor power was responding to heat load demand from the radiator. The reactor power and fuel salt and coolant salt temperatures calculated by GOTHIC exhibit good agreement with measured data.

Study on Ultra-Long Life, Small U-ZR Metallic Fuelled Core With Burnable Poison

2002 ◽  
Author(s):  
Kenji Tsuji ◽  
Hiromitsu Inagaki ◽  
Akira Nishikawa ◽  
Hisato Matsumiya ◽  
Yoshiaki Sakashita ◽  
...  
Keyword(s):  
Conceptual Design ◽  
Fast Reactor ◽  
Major Part ◽  
Reactor Core ◽  
Core Concept ◽  
Safety Requirements ◽  
Burnable Poison ◽  
The Core ◽  
Long Lifetime ◽  
Long Life

A conceptual design for a 50MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled core has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up.

Study on Thermal Striping at UIS of Advanced Loop Type Fast Reactor: Water Experiment Using a 1/3 Scale 60 Degree Sector Model

2009 ◽  
Author(s):  
Jun Kobayashi ◽  
Nobuyuki Kimura ◽  
Akira Tobita ◽  
Hideki Kamide ◽  
Osamu Watanabe ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Thermal Fatigue ◽  
Technology Development ◽  
Development Project ◽  
Control Rod ◽  
Sector Model ◽  
The Core ◽  
Technology Development Project ◽  
Frame Work ◽  
Thermal Striping

An advanced loop type sodium cooled fast reactor, JSFR, has been investigated in the frame work of Fast Reactor Cycle Technology Development Project (FaCT). As the temperatures difference between the control rod channels and the core fuel subassemblies is around 100 °C, temperature fluctuation due to the fluid mixing at the core outlet may cause high cycle thermal fatigue at the bottom of Upper Internal Structure (UIS). Then, a water experiment was conducted using an 1/3 scale 60 degree sector model. Temperature and its fluctuation intensity distributions around the control rod were measured and an effect of the improved structure against the thermal fatigue was examined.

Alternative Absorber Materials for Control Rods in ALFRED

2021 ◽  
Author(s):  
Hui Guo ◽  
Xin Jin ◽  
Kuaiyuan Feng ◽  
Hanyang Gu
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Operating Life ◽  
Insertion Depth ◽  
Control Rod ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Reference Design ◽  
Core Characteristics ◽  
Control Rods

Abstract The next-generation reactors require improved safety performance and longer cycle length, which initiate the research on alternative absorber materials. In this context, potential absorber materials including borides (B4C, HfB2, and ZrB2), rare earth oxides (Eu2O3, Gd2O3, Sm2O3, and Dy2TiO5), metals/alloys (Hf and AIC), and metal hydride (HfHx) were compared in a large sodium fast reactor. The design of control rods for Generation-IV fast reactors strongly depends on the core characteristics. In this paper, some alternative absorbers are assessed in a lead fast reactor ALFRED using depletion capability in the Monte-Carlo particle transport code OpenMC. Results show that the ALFRED reference control rod design with B4C largely satisfies the shutdown and operation requirements. 60% 10B enriched HfB2 and HfH1.18 can replace the operation part of the reference design. In the future, the safe operating life of B4C and HfB2 should be assessed taking into account the irradiation-induced swelling, temperature margin, and gas release. HfH1.18 has a limited and local influence on the core power distribution. Eu2O3 has little loss on the absorption ability after 5 cycle irradiation. This oxide absorber satisfies the shutdown function even with only half control rod insertion, while its critical insertion depth at beginning of the cycle should be increased to realize reactivity compensation function.

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