scholarly journals Calculations of research reactor thermal hydraulics based on VVER-440 fuel assamblies

2019 ◽  
Vol 5 (4) ◽  
pp. 317-321
Author(s):  
Thi Zieu Chang Doan ◽  
Georgy E. Lazarenko ◽  
Denis G. Lazarenko
Keyword(s):  
Flow Regime ◽  
Research Reactor ◽  
Reactor Design ◽  
Coolant Flow ◽  
Normal Operation ◽  
Coolant Temperature ◽  
Saturation Curve ◽  
Thermal Hydraulics ◽  
Pressurized Water ◽  
Wide Range

Having thoroughly analyzed the design features of VVER-type pressurized water reactors and VVR-type research reactors, the authors propose a design of a research reactor with low-enriched fuel based on deeply updated VVER-440 fuel assemblies. The research reactor is intended to solve a wide range of applied problems in nuclear physics, radiation chemistry, materials science, biology, and medicine. The calculated thermal hydraulics confirms the correctness of the fundamental approaches laid down in the reactor design. An equivalent reactor core model in the form of a thick-walled cylinder was considered, and the radial power density distribution was obtained. According to the heat power level, five groups of FAs were identified. For each group, the coolant mass flow rate was calculated, which ensures alignment with the outlet coolant temperature. The coolant flow regime was also estimated. It turned out that for the first row of FAs, the flow regime is in the transition region, while for the other rows the flow regime is laminar. A test by the Gr.Pr≥1.105 criterion showed its conformity (the calculated value was 1.96.106), indicating the transition to a viscous-gravitational regime. The FE surface overheating was calculated relative to the mixed coolant average temperature. The axial coolant flow temperature distribution is the same in all the FAs, the change in power is compensated by the corresponding change in the coolant flow. The maximum coolant overheating on the FE wall relative to the flow core is observed in the central FAs, reaching 31 °C, the boiling margin is about 15 °C. The estimates showed a significant dynamic pressure margin during natural thermal-convective circulation. By calculation, the values of the FE surface overheating during the reactor normal operation were obtained. An approximately 15-degree surface overheating margin relative to the saturation curve is shown, which guarantees the absence of cavitation wear of the FE claddings. In general, the performed calculations confirmed the correctness of the approaches laid down in the reactor design and made it possible to specify the core thermal hydraulics necessary for further developing the concept.

The Principle and Preliminary Validation of a New Subchannel Code

2013 ◽  
Author(s):  
Ning Bai ◽  
Yuanbing Zhu ◽  
Zhihao Ren ◽  
Haibo He ◽  
Haoliang Lu ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Constitutive Relations ◽  
Reactor Design ◽  
Safety Analysis ◽  
Industrial Applications ◽  
Transfer Model ◽  
Design Codes ◽  
Thermal Hydraulics ◽  
Two Phase ◽  
Pressurized Water

Following China’s road map of nuclear technology development, the development of self-reliant nuclear design codes becomes one of the most significant steps in the plan. Among the nuclear design codes, thermal-hydraulic analysis code is indispensable because it is the foundation of reactor safety analysis and reactor design. Recently, China Guangdong Nuclear Power Group has launched a series of R&D projects of reactor design code development. The sub-channel analysis code-LINDEN becomes one of the key subprojects. Since the sub-channel code is developed for thermal-hydraulic design and safety analysis of pressurized water reactors (PWRs), the basic requirements for the LINDEN code are reliability and stability. Therefore, the mathematical model and numerical method developed in the code are based on the matured approaches that have been used in various industrial applications. These models and methods includes: four-equation drift framework model of two-phase flow; the classical heat transfer model and fuel rod model (Poisson equation) as well as the constitutive relations; explicit formulation and stepping algorithms for equation solutions. The solver module of the code is programmed using object-oriented C/C++ language with the structural design.. With all these features, the code was developed to be stable, scalable and compatible. The code’s applicability has been further improved after model improvement and design optimization according to characteristics of China’s proprietary type of reactor. COBRA-IV and LINDEN have been used to conduct the thermal-hydraulics analysis for the Daya bay unit 1 and 2 nuclear plants at the steady-state conditions. The results demonstrate that the two codes agree well with each other. The preliminary tests show that the LINDEN code should be suitable for thermal-hydraulics analysis of large PWRs.

Investigation on Multi-Objective Optimal Design of a Two Pass Condenser

2013 ◽  
Author(s):  
Lei Chen ◽  
Chang-qi Yan ◽  
Jian-jun Wang
Keyword(s):  
Flow Rate ◽  
Optimization Design ◽  
Optimization Techniques ◽  
Coolant Flow ◽  
Geometric Constraints ◽  
Coolant Flow Rate ◽  
Coolant Temperature ◽  
Outer Diameter ◽  
Pressurized Water ◽  
Multi Objective

Condenser is one of the key components in nuclear power plant with pressurized water reactor. It is important to control the dimension and weight in the design of condenser through optimization techniques. In this paper, a mathematic model of a two pass condenser is set up for Qinshan I condenser. Some modifications are made based on the original multi-objective algorithm, and the comparison between modified algorithm and the original one is conducted. Furthermore, the multi-objective optimization design of the condenser, taking minimization of the coolant flow-rate and net weight as objectives, is carried out considering thermohydraulic and geometric constraints through hybrid Pareto-sorting multi-objective genetic algorithm (HPSMOGA). The sensitivities of some parameters, which may influence the coolant flow-rate and the net weight of condenser, are also analyzed. The results show that the mathematical model is agreeable for the condenser. it is also shown that the proposed multi-objective optimal method is more effective in searching non-dominated solutions. the sensitivity analysis show that the tube outer diameter, tube pitch, coolant velocity and coolant temperature rising influence the coolant flow-rate and net weight of the condenser more than other variables. The corresponding results would provide guidance in the engineering design of this type of condenser.

Application of the Failure Assessment Diagram to the Evaluation of Pressure-Temperature Limits for a Pressurized Water Reactor

1985 ◽  
Vol 107 (2) ◽  
pp. 192-196 ◽  
Author(s):  
K. K. Yoon ◽  
J. M. Bloom ◽  
W. A. Pavinich ◽  
H. W. Slager
Keyword(s):  
Fracture Mechanics ◽  
Reactor Vessel ◽  
Normal Operation ◽  
Coolant Temperature ◽  
Operating Pressure ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Failure Assessment Diagram ◽  
Failure Assessment

The failure assessment diagram approach, an elastic-plastic fracture mechanics procedure based on the J-integral concept, was used in the evaluation of pressure-temperature (P-T) limits for the beltline region of the vessel of a pressurized water reactor. The main objective of this paper is to illustrate the application of an alternate fracture mechanics method for the evaluation of pressure-temperature limits, as allowed by the Code of the Federal Regulation 10 CFR 50, Appendix G. The evaluation of P-T limits for the beltline region of a pressurized water reactor was based on the following assumptions: • ASME Pressure Vessel and Piping Code, Section III, Appendix G reference flaw • End-of-life fluence level in the beltline region • Longitudinal flaw in the beltline weld • J-resistance material toughness curves obtained from the U.S. Nuclear Regulatory Commission’s heavy section steel technology (HSST) program • Other material properties obtained from the Babcock & Wilcox Integrated Reactor Vessel Material Surveillance Program The maximum allowable pressure levels were calculated at 33 time points along the given reactor bulk coolant temperature history representing the normal operation of a pressurized water reactor. The results of the calculations showed that adequate margins of safety on operating pressure for the critical weld in the beltline of the pressurized water reactor vessel are assured.

scholarly journals Design Analyses for Full Core Conversion of The Dalat Nuclear Research Reactor

2014 ◽  
Vol 4 (1) ◽  
pp. 10-25
Author(s):  
Ba Vien Luong ◽  
Vinh Vinh Le ◽  
Ton Nghiem Huynh ◽  
Kien Cuong Nguyen
Keyword(s):  
Steady State ◽  
Power Distribution ◽  
Research Reactor ◽  
Nuclear Research ◽  
Coolant Temperature ◽  
Thermal Hydraulics ◽  
Kinetics Parameters ◽  
Nuclear Research Reactor ◽  
Enriched Uranium ◽  
Full Core

The paper presents calculated results of neutronics, steady state thermal hydraulics and transient/accidents analyses for full core conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the Dalat Nuclear Research Reactor (DNRR). In this work, the characteristics of working core using 92 LEU fuel assemblies and 12 beryllium rods were investigated by using many computer codes including MCNP, REBUS, VARI3D for neutronics, PLTEMP3.8 for steady state thermal hydraulics, RELAP/MOD3.2 for transient analyses and ORIGEN, MACCS2 for maximum  hypothetical accident (MHA). Moreover, in neutronics calculation, neutron flux, power distribution, peaking factor, burn up distribution, feedback reactivity coefficients and kinetics parameters of the working core were calculated. In addition, cladding temperature, coolant temperature and ONB margin were estimated in steady state thermal hydraulics investigation. The working core was also analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and MHA. Obtained results show that DNRR loaded with LEU fuel has all safety features as HEU and mixed HEU-LEU fuel cores and meets requirements in utilization as well.

Analysis for the Case of Channel Flow and Reactivity Changes for Advanced Heavy Water Reactor

2004 ◽  
Author(s):  
A. Srivastava ◽  
P. Majumdar ◽  
D. Mukhopadhyay ◽  
H. G. Lele ◽  
S. K. Gupta
Keyword(s):  
Heavy Water ◽  
Natural Circulation ◽  
Heat Removal ◽  
Coolant Flow ◽  
Thermal Hydraulics ◽  
Primary Coolant ◽  
Control Rod ◽  
Water Reactor ◽  
Wide Range ◽  
Heavy Water Reactor

The proposed Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level with no primary coolant pumps. Decrease in coolant flow or control rod malfunction can lead to undesirable rise in clad surface temperature depending upon severity and characteristics and response of the reactor and associated systems. In this paper safety assessment of the AHWR is made due to above events of different severity. Cause for events under category of decrease in coolant flow is mainly channel blockage of different severity at different locations. There is no other reason as it is natural circulation based reactor. Effect of flow decrease can be different in different channels and at different axial locations. In this paper channel blockages of different sizes are analysed at core inlet and using slave channel approach. Changes in reactivities can occur due to inadvertent withdrawal of one or more control rods from reactor core. In this analysis one control rod assembly is assumed to be removed from core. The event is simulated by addition of 5 mk reactivity in 120 seconds depending on the speed of withdrawal of assembly. The analysis for the above events are complex due to various complex and wide range of phenomena involved during different PIEs under this category. It involves single and two phase natural circulation at different power levels, inventories and pressures, coupled neutronics and thermal hydraulics behaviour, and coupled controller and thermal hydraulics. In this paper summary of analysis for each event is presented. In this paper, various modeling complexities are brought out; evaluation of acceptance criteria is made and design implications of each event are discussed.

scholarly journals Gas Damping in Capacitive MEMS Transducers in the Free Molecular Flow Regime

Sensors ◽  
2021 ◽  
Vol 21 (7) ◽  
pp. 2566
Author(s):  
Boris A. Boom ◽  
Alessandro Bertolini ◽  
Eric Hennes ◽  
Johannes F. J. van den Brand
Keyword(s):  
Flow Regime ◽  
Gas Diffusion ◽  
Diffusion Time ◽  
Molecular Flow ◽  
Free Molecular Flow ◽  
Simple Analytical Model ◽  
Gas Damping ◽  
Wide Range ◽  
Capacitive Mems ◽  
Insight Into

We present a novel analysis of gas damping in capacitive MEMS transducers that is based on a simple analytical model, assisted by Monte-Carlo simulations performed in Molflow+ to obtain an estimate for the geometry dependent gas diffusion time. This combination provides results with minimal computational expense and through freely available software, as well as insight into how the gas damping depends on the transducer geometry in the molecular flow regime. The results can be used to predict damping for arbitrary gas mixtures. The analysis was verified by experimental results for both air and helium atmospheres and matches these data to within 15% over a wide range of pressures.

Component and Integration Test of an FPGA-Based PWR Protection Sub-System Using UVM

2017 ◽  
Author(s):  
Xuanxuan Shui ◽  
Yichun Wu ◽  
Junyi Zhou ◽  
Yuanfeng Cai
Keyword(s):  
Nuclear Power ◽  
Power Industry ◽  
Nuclear Power Industry ◽  
Normal Operation ◽  
Static Timing Analysis ◽  
Specific Test ◽  
Pressurized Water ◽  
Safety Margins ◽  
Integration Test ◽  
Test Process

Field programmable gate arrays (FPGAs) have drawn wide attention from nuclear power industry for digital instrument and control applications (DI&C), because it’s much easier and simpler than microprocessor-based applications, which makes it more reliable. FPGAs can also enhance safety margins of the plant with potential possibility for power upgrading at normal operation. For these reasons, more and more nuclear power corporations and research institutes are treating FPGA-based protection system as a technical alternative. As nuclear power industry requires high reliability and safety for DI&C Systems, the development method and process should be fully verified and validated. For this reason, to improve the application of FPGA in NPP I&C system, the specific test methods are critical for the developers and regulators. However, current international standards and research reports, like IEC 62566 and NUREG/CR-7006, which have demonstrated the life circle of the development of FPGA-based safety critical DI&C in NPPs, but the specific test requirements and methods which are significant to the developers are not provided. In this paper, the whole test process of a pressurized water reactor (PWR) protection sub-system (Primary Coolant Flow Low Protection, Over Temperature Delta T Protection, Over Power Delta T Protection) is described, including detail component and integration tests. The Universal Verification Methodology (UVM) based on System Verilog class libraries is applied to establish the verification test platform. All these tests are conducted in a simulation environment. The test process is driven by the test coverage which includes code coverages (i.e., Statement, Branch, Condition and Expression, Toggle, Finite State Machine) and function coverage. Specifically, Register Transaction Level (RTL) simulation is conducted for Component tests, while RTL simulation, Gate Level simulation, Timing simulation and Static timing analysis are conducted for the integration test. The issues (e.g., the floating point calculation, FPGA resource allocation and optimization) arose in the test process are also analyzed and discussed, which can be references for the developers in this area. The component and integration tests are part of the Verification and Validation (V&V) work, which should be done by the V&V team separated from the development team. The testing method could assure the test results reliable and authentic. It is practical and useful for the development and V&V of FPGA-based safety DI&C systems.

Construction of an Elementary Model for the Dynamic Analysis of a Pressurized Water Reactor

2013 ◽  
Vol 284-287 ◽  
pp. 652-656 ◽  
Author(s):  
Chiung Wen Tsai ◽  
Chun Kuan Shih ◽  
Jong Rong Wang
Keyword(s):  
Transfer Functions ◽  
Feedback Loops ◽  
Dynamic Responses ◽  
Coolant Temperature ◽  
Inlet Temperature ◽  
Pressurized Water Reactor ◽  
Elementary Model ◽  
Pressurized Water ◽  
Water Reactor ◽  
Neutron Kinetics

A lumped-parameter numerical model was constructed based on the conservation laws of mass and energy and the point neutron kinetics with 6 groups of delayed neutron to represent the dynamics of primary loop of a pressurized water reactor (PWR) core. On the viewpoint of control theory, the coupled phenomenon of neutron kinetics and thermohydraulics can be recognized as a dynamic system with feedback loops which is caused by the Doppler effect and the coolant temperature difference. Scilab was implemented to representing the equivalent transfer functions and associated feedback loops of a PWR core. The dynamic responses were performed by the perturbations of coolant inlet flow, coolant inlet temperature, and reactivity insertion.

scholarly journals Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

2015 ◽  
Vol 19 (3) ◽  
pp. 989-1004 ◽  
Author(s):  
Ezddin Hutli ◽  
Valer Gottlasz ◽  
Dániel Tar ◽  
Gyorgy Ezsol ◽  
Gabor Baranyai
Keyword(s):  
Heat Transfer ◽  
Structural Integrity ◽  
Complex Geometry ◽  
Measurement Techniques ◽  
Coolant Flow ◽  
Temperature Fields ◽  
Head Part ◽  
Pressurized Water ◽  
Area Of Interest ◽  
Water Reactors

The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs). Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively) are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel). The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

scholarly journals Criticality safety analysis of spent fuel pool for a PWR using UO2, MOX, (Th-U)O2 and (TRU-Th)O2 fuels

2019 ◽  
Vol 7 (3A) ◽  
Author(s):  
Claubia Pereira ◽  
Jéssica P. Achilles ◽  
Fabiano Cardoso ◽  
Victor F. Castro ◽  
Maria Auxiliadora F. Veloso
Keyword(s):  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Operation Conditions ◽  
Transport Code ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Criticality Safety

A spent fuel pool of a typical Pressurized Water Reactor (PWR) was evaluated for criticality studies when it uses reprocessed fuels. PWR nuclear fuel assemblies with four types of fuels were considered: standard PWR fuel, MOX fuel, thorium-uranium fuel and reprocessed transuranic fuel spiked with thorium. The MOX and UO2 benchmark model was evaluated using SCALE 6.0 code with KENO-V transport code and then, adopted as a reference for other fuels compositions. The four fuel assemblies were submitted to irradiation at normal operation conditions. The burnup calculations were obtained using the TRITON sequence in the SCALE 6.0 code package. The fuel assemblies modeled use a benchmark 17x17 PWR fuel assembly dimensions. After irradiation, the fuels were inserted in the pool. The criticality safety limits were performed using the KENO-V transport code in the CSAS5 sequence. It was shown that mixing a quarter of reprocessed fuel withUO2 fuel in the pool, it would not need to be resized 

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