scholarly journals Validation of Doppler Temperature Coefficients and Component Power Distribution for the Advanced Neutronics Component Program KYLIN V2.0

2021 ◽  
Vol 9 ◽  
Author(s):  
Lei Jichong ◽  
Xie Jinsen ◽  
Chen Zhenping ◽  
Yu Tao ◽  
Yang Chao ◽  
...  
Keyword(s):  
Engineering Design ◽  
Power Distribution ◽  
Nuclear Power ◽  
Nuclear Engineering ◽  
Temperature Coefficients ◽  
The Core ◽  
Fuel Assemblies ◽  
Design Requirements ◽  
Design Software ◽  
Good Agreement

This work is interested in verifying and analyzing the advanced neutronics assembly program KYLIN V2.0. Assembly calculations are an integral part of the two-step calculation for core design, and their accuracy directly affects the results of the core physics calculations. In this paper, we use the Doppler coefficient numerical benchmark problem and CPR1000 AFA-3G fuel assemblies to verify and analyze the advanced neutronics assembly program KYLIN V2.0 developed by the Nuclear Power Institute of China. The analysis results show that the Doppler coefficients calculated by KYLIN V2.0 are in good agreement with the results of other well-known nuclear engineering design software in the world; the power distributions of AFA-3G fuel assemblies are in good agreement with the results of the RMC calculations, it’s error distribution is in accordance with the normal distribution. It shows that KYLIN V2.0 has high calculation accuracy and meets the engineering design requirements.

scholarly journals Application of Evolutionary Simulated Annealing Method to Design a Small 200 MWt Reactor Core

2021 ◽  
Vol 10 (4) ◽  
pp. 16-23
Author(s):  
Tran Viet Phu ◽  
Tran Hoai Nam ◽  
Hoang Van Khanh
Keyword(s):  
Simulated Annealing ◽  
Power Plant ◽  
Nuclear Power ◽  
Cycle Length ◽  
Reactor Core ◽  
Loading Pattern ◽  
The Core ◽  
Fuel Assemblies ◽  
Full Power ◽  
Annealing Method

This paper presents the application of an evolutionary simulated annealing (ESA) method to design a small 200 MWt reactor core. The core design is based on a reference ACPR50 reactor deployed in a floating nuclear power plant. The core consists of 37 typical 17x17 PWR fuel assemblies with three different U-235 enrichments of 4.45, 3.40 and 2.35 wt%. Core loading pattern (LP) has been optimized for obtaining the cycle length of 900 effective full power days, while minimizing the average U-235 enrichment and the radial power peaking factor. The optimization process was performed by coupling the ESA method with the COREBN module of the SRAC2006 system code.

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.

scholarly journals A study on the core loading pattern of the VVER-1200/V491

2021 ◽  
Vol 7 (1) ◽  
pp. 21-27
Author(s):  
Vinh Thanh Tran ◽  
Viet Phu Tran ◽  
Thi Dung Nguyen
Keyword(s):  
Power Distribution ◽  
Nuclear Power ◽  
Safety Standards ◽  
Loading Pattern ◽  
The Core ◽  
Fuel Burn ◽  
Loading Patterns ◽  
Safety Parameters ◽  
Safety Features ◽  
Effective Multiplication

The VVER-1200/V491 was a selected candidate for the Ninh Thuan I Nuclear Power Plant.However, in the Feasibility Study Safety Analysis Report (FS-SAR) of the VVER-1200/V491, the core loading pattern of this reactor was not provided. To assess the safety features of the VVER- 1200/V491, finding the core loading patterns and verifying their safety characteristics are necessary. In this study, two core loading patterns of the VVER-1200/V491 were suggested. The first loading pattern was applied from the VVER-1000/V446 and the second was searched by core loading optimization program LPO-V. The calculations for power distribution, the effective multiplication factor (k-eff), and fuel burn-up were then calculated by SRAC code. To verify several safety parameters of loading patterns of the VVER-1200/V491, the neutron delayed fraction (DNF), fuel andmoderator temperature feedbacks (FTC and MTC) were investigated and compared with the safety standards in the VVER-1200/V491 FS-SAR or the VVER-1000/V392 ISAR.

Representative neutronic characteristics calculations for the VVER-1000 reactors using SRAC and MCNP5

2016 ◽  
Vol 6 (2) ◽  
pp. 21-30
Author(s):  
Huu Tiep Nguyen ◽  
Viet Phu Tran ◽  
Tuan Khai Nguyen ◽  
Vinh Thanh Tran ◽  
Minh Tuan Nguyen
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Multiplication Factor ◽  
Benchmark Problem ◽  
Fuel Assemblies ◽  
Effective Multiplication Factor ◽  
Neutronic Characteristics ◽  
Effective Multiplication ◽  
Good Agreement

This paper presents the results of neutronic calculations using the deterministic and Monte-Carlo methods (the SRAC and MCNP5codes) for the VVER MOX Core Computational Benchmark Specification and the VVER-1000/V392 reactor core. The power distribution and keff value have been calculated for a benchmark problem of VVER core. The results show a good agreement between the SRAC and MCNP5 calculations. Then, neutronic characteristics of VVER-1000/V392 such as power distribution, infinite multiplication factor (k-inf) of the fuel assemblies, effective multiplication factor keff, peaking factor and Doppler coefficient were calculated using the two codes.

Uncertainty Analysis of Power Distribution for NESTOR Based on the Double Latin Hypercube Sampling Method

2018 ◽  
Author(s):  
Hongkuan Liao ◽  
Qing Li ◽  
Yingrui Yu ◽  
Yuying Hu ◽  
Lei Wu ◽  
...  
Keyword(s):  
Uncertainty Analysis ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Sampling Method ◽  
Latin Hypercube Sampling ◽  
Physical Models ◽  
Latin Hypercube ◽  
Analysis Process ◽  
Design Software

Due to the complexity of the reactor system, many approximations are used in the nuclear design and calculations inevitably. The accuracy of the nuclear design software is closely related to the safety of the reactor design and operation. Besides, improving the accuracy is an effective way to excavating the economy of nuclear power plants. To analyze the uncertainty of power distribution for NESTOR nuclear design software, we suggested an uncertainty analysis method based on the double Latin Hypercube Sampling (LHS) method and the random sampling statistical analysis (RSSA) method, and built an uncertainty analysis process based on LHS. The uncertainty of physical model and the uncertainty of the change of parameters were both taken into consideration with the double samplings, and 3481 core states were generated by the double samplings. Therefore, the uncertainty of power distribution could be directly analyzed through modeling computation, and the uncertainty of radial power distribution was achieved as ±3.856% under the condition of 95% confidence coefficient and 95% probability. Meanwhile, according to deviation transmission idea, we obtained the uncertainty of power distribution from physical models and the change of parameters based on the measured method. The result shows that the accuracy using the double sampling method is nearly the same to which achieved by the deviation transmission idea, and more conservative.

Validation DualSPHysics Code for Liquid Sloshing Phenomena

2014 ◽  
Author(s):  
Sergei K. Buruchenko ◽  
Alejandro J. C. Crespo
Keyword(s):  
Nuclear Power ◽  
Liquid Sloshing ◽  
Accident Analysis ◽  
The Core ◽  
Particle Hydrodynamics ◽  
In Series ◽  
Core Meltdown ◽  
Smoothed Particle ◽  
Smoothed Particle Hydrodynamics Method ◽  
Good Agreement

The DualSPHysics code is proposed as a numerical tool for the simulation of liquid sloshing phenomena. A particular type of sloshing motion can occur during the core meltdown of a liquid metal cooled reactor (LMR) and can lead to a compaction of the fuel in the center of the core possibly resulting in energetic nuclear power excursions. This phenomenon was studied in series of “centralized sloshing” experiments with a central water column collapsing inside the surrounding cylindrical tank. These experiments provide data for a benchmark exercise for accident analysis codes. To simulate “centralized sloshing” phenomena, a numerical method should be capable to predict the motion of the free surface of a liquid, wave propagation and reflection from the walls. The DualSPHysics code based on the smoothed particle hydrodynamics method was applied to the simulation of “centralized sloshing” experiments. Simulation results are compared with the experimental results. In a series of numerical calculations it is shown that overall motion of the liquid is in a good agreement with experimental observations. Dependence on the initial and geometrical symmetry is studied and compared with experimental data.

scholarly journals Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

2013 ◽  
Vol 2013 ◽  
pp. 1-7 ◽  
Author(s):  
Toshio Wakabayashi
Keyword(s):  
Power Distribution ◽  
Outer Core ◽  
Monte Carlo Code ◽  
The Core ◽  
Fuel Burn ◽  
Fuel Assemblies ◽  
Core Characteristics ◽  
Power Flattening ◽  
Reactivity Insertion ◽  
Operational Time

An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

Research and Engineering Application of Repaired Fuel Assembly Reusing Technique

2017 ◽  
Author(s):  
Lihua Wang ◽  
Qingxiang Yang ◽  
Ping Yang ◽  
Jiazheng Liu ◽  
Libing Zhu ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Engineering Application ◽  
Evaluation Technique ◽  
Fuel Rods ◽  
The Core ◽  
Fuel Assemblies

Due to debris in the coolant against clad, fuel clad wear, fuel handling fault and so on, fuel rods maybe be damaged during the operation of nuclear power plants, in order that the fuel assemblies with damaged fuel rods are discharged before scheduled. If the damaged fuel assemblies are not reloaded into the core of the nuclear power plant, the fuel utilization decreases and the economy of the nuclear power plant is partly lost. For retrieving the loss of the economy, the damaged fuel assemblies can be repaired by replacing damaged fuel rods with dummy rods which don’t include fissile nuclides. Then, the repaired fuel assemblies can be reloaded into the core. As the repaired fuel assemblies are different with the normal fuel assemblies, especially the number of the damaged fuel rods is considerable, a whole quantitative analysis is very necessary to evaluate the effects from the reuse of the repaired fuel assemblies. In this paper, a full scope evaluation of reload design are performed including nuclear design, fuel design, thermal hydraulic design and safety evaluation, and some necessary improvements are done for the software system, design methods and progress which have been used in the normal reload design. As results, an integrated evaluation technique is developed to evaluate the feasibility and safety of reusing the repaired fuel assemblies, and the key effects due to the reuse of the repaired fuel assemblies are extracted, and the different effects are studied for the different materials of the dummy rods which can be used to conduct how to choose the proper material of dummy rods. In addition, this technique has been successfully applied in the engineering and the loss of economy due to the damage of fuel assemblies was retrieved partly. Therefore, the integrated evaluation technique has also important directive to other nuclear power plants if the repaired fuel assemblies are planned to reuse.

scholarly journals Determination of 18-month fuel cycle parameters for the purpose of fuel costs minimization at the basis of use constructions of fuel assemblies in VVER-1200 reactors

2019 ◽  
Vol 5 (1) ◽  
pp. 9-15
Author(s):  
Taha M. Hashlamoun ◽  
Sergey B. Vygovsky ◽  
Sergey T. Leskin ◽  
A. Safa Duman
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fuel Cycle ◽  
Life Time ◽  
Reactor Vessel ◽  
Reactor Fuel ◽  
The Core ◽  
Fuel Assemblies ◽  
Fuel Costs

This article presents the results of research, that were focused on determining the optimal parameters of the extension of (reactor life-time) reactor fuel cycle in order to reduce the total operating costs of nuclear power plants during the transition from 12-month reactor fuel cycle to 18-month fuel cycle. The relevance of the research is related to the fact that, in recent years, there is a transition at all operating nuclear power plants VVER-1000 (1200) from 12-month reactor fuel cycle to extended 18-month fuel cycle. At the same time, represent the interests to solve the problem of conservation the extension of reactor life-time while reducing the number of loaded fuel assemblies with fresh fuel assemblies, which would reduce the total operating, and fuel costs. Search for solutions of this problem is associated with mandatory implementation of all requirements for the safe operation of the reactor and the reduction of the maximum fast neutron fluence on the reactor vessel in comparison with its value at the operating nuclear power plants. In the present work, with using the program PROSTOR software complex researched the neutron-physical characteristics of the core at the nominal parameters of the VVER-1200 reactor through the implementation of various fuel cycle strategies. The article developed various schemes of fuel-reloading for an 18-month fuel cycle with a different number of fuel assemblies. The article carries out a comparative analysis of the main parameters in the core for fuel-reloading schemes options of an 18- and 12-month fuel cycle with each other. Determine the minimum amount of fuel assemblies and provide the necessary duration of the reactor life-time for 18-month fuel cycle with using the extension of reactor life-time by reducing the power at the end of the reactor cycle to 70% of the nominal power. In the article, the arrangements of fuel assemblies were developed to provide limitations of local power by volume of the core, which reduce the fluence of fast neutrons on the reactor vessel in comparison with the projected value of the fluence. This article shows that the 18-month fuel cycle for the VVER-1200 reactor is more economical than the 12-month fuel cycle. These studies were carried out for the VVER-1200 reactor at the power of 100% of the nominal.

scholarly journals Application of Electrical Impedance Measurements’ Method for Studies of Bubble Flows

2021 ◽  
Vol 12 (3) ◽  
pp. 183-193
Author(s):  
I. A. Konovalov ◽  
A. A. Chesnokov ◽  
A. A. Barinov ◽  
S. M. Dmitriev ◽  
A. E. Khrobostov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Electrical Impedance ◽  
Wire Mesh ◽  
Measuring System ◽  
Gas Content ◽  
The Core ◽  
Multicomponent Flows ◽  
Liquid Flows ◽  
Good Agreement

One of the important tasks in carrying out a computational justification of the reliability and safety of equipment that is part of the projected nuclear power plants today is the modeling of the bubbly regime of the coolant flow. In this regard the aim of this work is the use of extended methods of using matrix conductometric systems which are widespread in research practice for study of gas-liquid flows.The work uses a method of primary processing of experimental data aimed at eliminating of excess conductivity in the cells of the developed wire mesh sensor which makes it possible to obtain the values of the true volumetric gas content in the investigated area.Subsequent analysis of the possibilities to estimate the volumes of registered gas bubbles by the gradient method as well as the size of the interface in the sensor cells which plays a key role in modeling the interfacial heat and mass transfer.Comparison of readings values with the control instruments cues showed a good agreement. The presented work is an adaptation of the use of a conductometric measuring system for the study of multicomponent flows with the aim of further application for the study of two-component flows in the channels of the core simulator using wire mesh sensors.

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