Conceptual Design of CANDU-SCWR With Thermal-Hydraulics Coupling

2010 ◽  
Author(s):  
Ping Yang ◽  
Liangzhi Cao ◽  
Hongchun Wu ◽  
Changhui Wang
Keyword(s):  
Power Distribution ◽  
Average Power ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Loading Pattern ◽  
Burnable Poison ◽  
Management Scheme ◽  
Fuel Bundle ◽  
Hydraulic Coupling ◽  
Operation Pressure

A CANDU-SCWR core is designed by using a 3D neutronics/thermal-hydraulic coupling method. In the fuel channel design, a typical 43-element fuel bundle is used, the coolant and the moderator are supercritical water and heavy water respectively. The thickness of the moderator is optimized to ensure the negative coolant coefficient during operation. With 1220 MW electric power, the reactor core is designed with a diameter of 4.8m and length of 4.95m, and there are totally 300 fuel channels, each of which consists of 10 fuel bundles. The inlet coolant temperature is set to be 350 °C °C and the operation pressure is 25 MPa. In order to flatten the radial power distribution, the loading pattern of the equilibrium cycle is optimized, and an optimized fuel management scheme is used with three batches refueling, burnable poison Dy2O3 is used to flatten the power peaking. The numerical results show that the average power density is 42.75 W/cm3, while the maximum linear element rate (LER) is 575W/cm. The average discharged burnup of the equilibrium is 48.3GWD/tU, and a high average outlet coolant temperature of 625 °C is achieved with a maximum cladding surface temperature less than 850 °C. Besides, the coolant temperature coefficient is negative throughout the cycle.

scholarly journals Study on the effect of various burnable poisons on pebble bed reactor with OTTO fuelling schemes using Serpent 2 Monte Carlo code

2021 ◽  
Vol 927 (1) ◽  
pp. 012018
Author(s):  
Nicholas Sidharta ◽  
Almanzo Arjuna
Keyword(s):  
Power Density ◽  
Power Distribution ◽  
Reactor Core ◽  
Monte Carlo Code ◽  
Volumetric Fraction ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Burnable Poison ◽  
Discrete Element Method Simulation ◽  
Upper Level

Abstract Pebble bed reactor with a once-through-then-out fuelling scheme has the advantage of simplifying the refueling system. However, the core upper-level power density is relatively higher than the bottom, producing an asymmetric core axial power distribution. Several burnable poison (BP) configurations are used to flatten the peak power density and improve power distribution while suppressing the excess core reactivity at the beginning of the burnup cycle. This study uses HTR-PM, China’s pebble bed reactor core, to simulate several burnable poison (BP) configurations. Serpent 2 coupled with Octave and a discrete element method simulation is used to model and simulate the pebble bed reactor core. It is found that erbium needs a large volumetric fraction in either QUADRISO or distributed BP to perform well. On the other hand, gadolinium and boron need a smaller volumetric fraction but perform worse in radial power distribution criteria in the fuel sphere. This study aims to verify the effect of BP added fuel pebbles on an OTTO refueling scheme HTR-PM core axial power distribution and excess reactivity.

scholarly journals Preliminary Study on Small Long-Life PWR Reactor Design Fueled Thorium with Protactinium as Burnable Poiso

2017 ◽  
Vol 4 (3) ◽  
Author(s):  
Jubaidah Jubaidah
Keyword(s):  
Power Distribution ◽  
Axial Direction ◽  
Coolant Temperature ◽  
Fissile Material ◽  
Reactor Kinetics ◽  
Burnable Poison ◽  
Long Life ◽  
Kinetic Calculations ◽  
In The Beginning ◽  
Data Library

Thorium as fissile material has overweigh advantages compared to Uranium in order to have a long-life PWR reactor without replacement fuel. Additional of Protactinium as a burnable poison is effective to reduce excess reactivity in the beginning of the reactor operation. This study aims to analyze the neutronic, burnup fuel, thermal-hydraulic and kinetic calculations in regard to reactor safety. The calculation is performed with SRAC program using JENDL-32 as data library. The results show that the value of keff for 33 years are still in the range 1.0154 <keff<1.0171 and the burn-up level has increased, due to the additional fissile material as  conversion results of Th232 and Pa231. Axial linear power distribution indicates normal distribution pattern. The coolant temperature distribution in the axial direction is quite flat. Key words: PWR, Th-232, U-233, Pa-231, neutronic, thermal-hydraulic, burn-up level, reactor kinetics

Conceptual Core Design of HAPPY200 Reactor

2018 ◽  
Author(s):  
Xiaosheng Li ◽  
Linsen Li ◽  
Lianghui Peng ◽  
Xiaosong Chen ◽  
Zhaocan Meng ◽  
...  
Keyword(s):  
Temperature Coefficient ◽  
Power Distribution ◽  
Reactor Core ◽  
Design Parameters ◽  
Coolant Temperature ◽  
Control Rod ◽  
The Core ◽  
Passive Safety System ◽  
Reactivity Coefficients ◽  
Full Power

The pressure and coolant temperature of Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY200) is significantly lower than PWR of the NPP, the core design and analysis were completed according to the design parameters and features of HAPPY200. The fuel assembly and its feature was firstly designed and studied based on the investigation of different types of fuel assemblies. Then the core configuration was studied and optimized according to the design parameters of HAPPY200; Eventually, neutronics calculation of the core was performed and key parameters were obtained including cycle length, power distribution, control rod worth, reactivity coefficients and etc. The study shows that with the core design HAPPY200 can be operated for 18 months in full power and reactivity control system can maintain criticality of the core in the full cycle. Due to the non-soluble boron design of the reactivity control scheme, moderator temperature coefficient and isothermal temperature coefficient are both negative, the Doppler temperature coefficients and power coefficients in different phase of the lifetime and in different power levels are also negative, therefore, the reactivity safety of the reactor core can be ensured.

scholarly journals Numerical Analysis of the Dynamic Interaction between Two Closely Spaced Vertical-Axis Wind Turbines

Energies ◽  
2021 ◽  
Vol 14 (8) ◽  
pp. 2286
Author(s):  
Yutaka Hara ◽  
Yoshifumi Jodai ◽  
Tomoyuki Okinaga ◽  
Masaru Furukawa
Keyword(s):  
Wind Turbines ◽  
Power Distribution ◽  
Phase Synchronization ◽  
Average Power ◽  
Vertical Axis ◽  
Turbine Rotor ◽  
Mean Power ◽  
Vertical Axis Wind Turbines ◽  
Fluid Body ◽  
First Time

To investigate the optimum layouts of small vertical-axis wind turbines, a two-dimensional analysis of dynamic fluid body interaction is performed via computational fluid dynamics for a rotor pair in various configurations. The rotational speed of each turbine rotor (diameter: D = 50 mm) varies based on the equation of motion. First, the dependence of rotor performance on the gap distance (gap) between two rotors is investigated. For parallel layouts, counter-down (CD) layouts with blades moving downwind in the gap region yield a higher mean power than counter-up (CU) layouts with blades moving upwind in the gap region. CD layouts with gap/D = 0.5–1.0 yield a maximum average power that is 23% higher than that of an isolated single rotor. Assuming isotropic bidirectional wind speed, co-rotating (CO) layouts with the same rotational direction are superior to the combination of CD and CU layouts regardless of the gap distance. For tandem layouts, the inverse-rotation (IR) configuration shows an earlier wake recovery than the CO configuration. For 16-wind-direction layouts, both the IR and CO configurations indicate similar power distribution at gap/D = 2.0. For the first time, this study demonstrates the phase synchronization of two rotors via numerical simulation.

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.

Improvement of Temperature Evaluation Model of Biological Shielding Concrete for HTTR Test Simulating LOFC With VCS Inactive

2013 ◽  
Author(s):  
Shoji Takada ◽  
Shunki Yanagi ◽  
Kazuhiko Iigaki ◽  
Masanori Shinohara ◽  
Daisuke Tochio ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Thermal Radiation ◽  
Cooling System ◽  
Evaluation Model ◽  
Reactor Core ◽  
Reactor Power ◽  
Coolant Temperature ◽  
Water Cooling ◽  
Reactor Outlet

HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.

ACR-1000™ Fuel Design Verification

2010 ◽  
Author(s):  
David J. Wren ◽  
Patrick Reid ◽  
Len L. Wright
Keyword(s):  
High Performance ◽  
Reactor Core ◽  
Cold Test ◽  
Performance Criteria ◽  
Extensive Experience ◽  
Fuel Bundle ◽  
Test Loop ◽  
Fuel Design ◽  
And Performance ◽  
The Impact

The ACR-1000™ design is an evolutionary advancement of the proven CANDU® reactor design that delivers enhanced economic performance, safety, operability and maintainability. The fuel for the ACR-1000 design is based on the well established CANDU fuel bundle design that has over 40 years of demonstrated high performance. Building on its extensive experience in fuel design and analysis, and fuel testing, AECL has designed a CANFLEX-ACR™ fuel bundle that incorporates the latest improvements in CANDU fuel bundle design. The ACR-1000 fuel bundle also includes features that enable the ACR-1000 to achieve higher fuel burn-up and improved reactor core physics characteristics. To verify that the CANFLEX-ACR fuel bundle design will meet and exceed all design requirements, an extensive program of design analysis and testing is being carried out. This program rigorously evaluates the ability of the fuel design to meet all design and performance criteria and particularly those related to fuel failure limits. The design analyses address all of the phenomena that affect the fuel during its residence in the reactor core. Analysis is performed using a suite of computer codes that are used to evaluate the temperatures, deformations, stresses and strains experienced by the fuel bundle during its residence in the reactor core. These analyses take into account the impact of fuel power history and core residence time. Complementing the analyses, testing is performed to demonstrate the compatibility of the fuel with the reactor heat transport system and fuel handling systems, and to demonstrate the ability of the fuel to withstand the mechanical forces that it will experience during its residence in the core. The testing program includes direct measurement of prototype fuel element and fuel bundle properties and performance limits. A number of different test facilities are used including a cold test loop and a hot test loop with a full-scale ACR-1000 fuel channel that operates at reactor coolant temperatures, pressures and flows. This paper summarizes the out-reactor test program and related analysis that provide the basis for verifying that the ACR-1000 fuel design meets its requirements.

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

Towards Optimal In-Core Fuel Management of Thorium-Plutonium-Fuelled PWR Cores

2013 ◽  
Author(s):  
Nurjuanis Z. Zainuddin ◽  
Benjamin A. Lindley ◽  
Geoffrey T. Parks
Keyword(s):  
Nuclear Waste ◽  
Cycle Length ◽  
Feasibility Studies ◽  
Fuel Management ◽  
Loading Pattern ◽  
Burnable Poison ◽  
The Core ◽  
Mox Fuel ◽  
Operational Characteristics

Plutonium is a significant proliferation concern as well as a major contributor to the long-term toxicity of nuclear waste. Partial incineration in PWRs with uranium-MOX fuel is often considered to mitigate these concerns. Thorium-MOX is an alternative fuel with superior material properties and higher plutonium destruction rates, as shown in multiple feasibility studies. However, the core performance and operational characteristics (e.g. discharge burn-up, feasibility of controlling the core) are ultimately dependent on the core loading pattern (LP) and burnable poison (BP) design. In this paper, the LP for Th-Pu fuel of various compositions is optimized for (1) discharge burn-up, (2) radial form factor (RFF), (3) cycle length, (4) moderator temperature coefficient (MTC), and (5) reactivity swing over cycle. Maximizing the cycle length makes the discharge burn-up and reactivity swing worse due to placement of once- and twice-burnt fuel near the core periphery. It also makes the MTC less negative. The harder neutron spectrum of Th-Pu fuel compared to conventional U fuel favours the use of distributed integral burnable poisons to control the reactivity swing over the cycle. This leads to a significant amount of dissimilarity between LPs with relatively similar performance measures, and between optimal LPs for different Pu loadings in the fuel. The RFF can vary throughout the cycle but a careful placement of the assemblies can mitigate this. The cycle reactivity swing is controlled using enriched soluble boron, which makes the MTC worse, and this constrains feasibility for high Pu loading in the fuel.

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