Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

2006 ◽  
Author(s):  
A. Gorzel
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressure Increase ◽  
Thermal Hydraulics ◽  
Short Term ◽  
Speed Reduction ◽  
Pump Speed ◽  
Transient Behaviour ◽  
Limit Value

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and impermissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second — much smaller — maximum that would occur around one second after the first one in the absence of a SCRAM.

Effect of Ocean Conditions on Neutronic/Thermal-Hydraulic Coupling of IPWR

2018 ◽  
Author(s):  
Genglei Xia
Keyword(s):  
Natural Circulation ◽  
Oscillation Amplitude ◽  
Three Dimensional ◽  
Simulation Models ◽  
Reactor Power ◽  
Coolant Flow ◽  
Uneven Distribution ◽  
Thermal Hydraulics ◽  
Primary Coolant ◽  
Ocean Conditions

In this paper, a thermal-hydraulics analysis code under non-inertial system was developed by establishing dynamic simulation models of typical ocean conditions and modifying RELAP5 code. Based on the modified code, the effects of the inclination, fluctuation and rolling conditions on the natural circulation characteristics of an integrated PWR were studied. The dynamic input and output interface model of RELAP5 was further established, and RELAP5 was coupled with a two-group three-dimensional neutron kinetics code to realize the improvement of nuclear feedback module in thermal hydraulic code. And the distribution of core flow and power under different motions was analyzed. The results indicate that, the uneven distribution of coolant flow increases with the increasing inclination angle, this trend leads to an uneven distribution of primary coolant flow. The flow fluctuation has a 180° phase difference under rolling conditions, and the reactor power and coolant flow oscillation increases with the increasing rolling period and amplitude. In the case of heaving motion, the peaks of the oscillation amplitude of the flow and power lying in the hottest channel as the additional forces on the fluid of each channel are spatially uniform. The code developed in this paper has the functions of modelling ocean conditions and three-dimensional coupled neutronics/thermal-hydraulics, and can be used as a simulation tool for Floating PWR.

scholarly journals RAST-K v2—Three-Dimensional Nodal Diffusion Code for Pressurized Water Reactor Core Analysis

Energies ◽  
2020 ◽  
Vol 13 (23) ◽  
pp. 6324
Author(s):  
Jinsu Park ◽  
Jaerim Jang ◽  
Hanjoo Kim ◽  
Jiwon Choe ◽  
Dongmin Yun ◽  
...  
Keyword(s):  
Power Plants ◽  
Single Channel ◽  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Power ◽  
Hydraulic Calculation ◽  
Pressurized Water ◽  
Computational Performance ◽  
Nodal Diffusion ◽  
Calculation Models

The RAST-K v2, a novel nodal diffusion code, was developed at the Ulsan National Institute of Science and Technology (UNIST) for designing the cores of pressurized water reactors (PWR) and performing analyses with high accuracy and computational performance by adopting state-of-the-art calculation models and various engineering features. It is a three-dimensional multi-group nodal diffusion code developed for the steady and transient states using microscopic cross-sections generated by the STREAM code for 37 isotopes. A depletion chain containing 22 actinides and 15 fission products and burnable absorbers was solved using the Chebyshev rational approximation method. A simplified one-dimensional single-channel thermal-hydraulic calculation was performed with various values for the thermal conductivity. Advanced features such as burnup adaptation and CRUD modeling capabilities are implemented for the multi-cycle analysis of commercial reactor power plants. The performance of RAST-K v2 has been validated with the measured data of PWRs operating in Korea. Furthermore, RAST-K v2 has been coupled with a sub-channel code (CTF), fuel performance code (FRAPCON), and water chemistry code for multiphysics analyses. In this paper, the calculation models and engineering features implemented in RAST-K v2 are described, and then the application status of RAST-K v2 is presented.

scholarly journals Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

Nukleonika ◽  
2015 ◽  
Vol 60 (3) ◽  
pp. 531-536 ◽  
Author(s):  
Igor P. Królikowski ◽  
Jerzy Cetnar
Keyword(s):  
Cross Sections ◽  
Nuclear Reactions ◽  
Nuclear Reactors ◽  
Input Parameter ◽  
Three Dimensional ◽  
Reactor Core ◽  
Coupled Analysis ◽  
Thermal Hydraulics ◽  
Power Profile ◽  
The Impact

Abstract Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD) code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

A Simplified Two-Group Multipoint Kinetics Model

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Santosh K. Pradhan ◽  
K. Obaidurrahman ◽  
Kannan N. Iyer
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Reactor Core ◽  
Coupled System ◽  
Kinetics Model ◽  
Primary Circuit ◽  
Thermal Hydraulics ◽  
Time Step ◽  
Large Power

Abstract Detailed multiphysics modeling of nuclear power plants has become a necessity in the era of best-estimate analysis. For a number of transients with strong coupling between the neutronics in the reactor core and the fluid-dynamics in the primary circuit and overall heat transfer, it is required to carry out coupled system thermal hydraulics and core three-dimensional (3D) neutronics analysis. Point kinetics approach in the system thermal-hydraulics (TH) code RELAP5 limits its use for many reactivity-induced transients, which involve asymmetric core behavior. In a recent development, a simplified multipoint kinetics model has been coupled with system TH code RELAP5 to circumvent its inadequacy for the analysis of reactivity-induced transients involving asymmetric core behavior. The objective of this paper is to validate the simplified multipoint kinetics model against an asymmetric fast transient benchmark problem in a large power reactor. Time-step and nodalization sensitivity studies have been performed. It is demonstrated that the multipoint kinetics model results are in good agreement with the benchmark, advocating its applicability.

Preliminary Neutronics and Thermal-Hydraulics Study on Thorium-Based HTR-PM With Outer Breeding Zone

2018 ◽  
Author(s):  
Qiudong Wang ◽  
Bing Xia ◽  
Jiong Guo ◽  
Ding She ◽  
Lei Shi ◽  
...  
Keyword(s):  
Heavy Metal ◽  
Reactor Core ◽  
Volume Ratio ◽  
Reactor Power ◽  
Normal Operation ◽  
Design Parameters ◽  
Fuel Particle ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Breeding Zone

In this work, a two-zone reactor core, which contains an inner driving zone and an outer ThO2 breeding zone, is designed under the framework of the HTR-PM. The main aim of this work is to investigate the feasibility of thorium utilization in the mature design of the HTR-PM with the inherent safety features. The neutronics and thermal-hydraulics characteristics are investigated to optimize the design parameters by using VSOP. The aim of optimization is to maximize the conversion of thorium to 233U in the breeding zone. The preliminary results indicate that the volume ratio of the breeding zone to the driving zone has significant influence on the power peaking factor and the maximum fuel temperature in normal operation and accidental conditions. On the other hand, the increase of reactor power will lead to increase of maximum fuel temperature after DLOFC accident. More heavy metal loading in the breeding zone will raise 233U yield, while the influence of fuel particle radius on the conversion ratio is negligible. An optimized 200 MWt two-zone reactor design is obtained with volume ratio of the driving zone to the breeding zone of 4:1, and 7 g and 30 g heavy metal per fuel sphere in the driving zone and the breeding zone, respectively.

Study on Structural Integrity Assessment of Reactor Pressure Vessel Based on Three-Dimensional Thermal-Hydraulics and Structural Analyses

2014 ◽  
Author(s):  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
Tadashi Watanabe ◽  
Yutaka Nishiyama
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Thermal Hydraulics ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Coolant Water ◽  
Reactor Pressure

In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during pressurized thermal shock (PTS) events, the thermal history of the coolant water and the heat transfer coefficient between the coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events and Jackson-Fewster correlation. Subsequently, using these values, structural integrity assessments of RPV are performed by structural analysis; e.g., loading that affects crack propagation is evaluated. Three-dimensional TH and structural analyses are recommended for precise assessments of the structural integrity of RPV. In this study, we performed TH and structural analyses simulating typical PTS events using three-dimensional models of cold-leg, downcomer and RPV to more accurately assess the structural integrity of RPV. From these analyses, we obtained loading histories from the reactor core region of RPV in which a crack is postulated in the structural integrity assessment. We discuss the conservativeness of current analysis methods on the structural integrity assessment of RPV through the comparison of loading conditions due to PTS events.

Validation of TRACG05 for Application to BWR AOO

2012 ◽  
Author(s):  
Asuka Matsui ◽  
Masashi Tamitani ◽  
Yoshiro Kudo ◽  
Sho Takano ◽  
Tatsuya Iwamoto ◽  
...  
Keyword(s):  
Experimental Data ◽  
Three Dimensional ◽  
Reactor Core ◽  
Expansion Method ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Neutron Kinetics ◽  
Safety Issues ◽  
Best Estimate ◽  
Assessment Matrix

TRACG code, coupling a three-dimensional neutron kinetics model for the reactor core with thermal-hydraulics based on two-fluid conservation equations, is a best-estimate (BE) code for BWRs to realistically simulate their transient and accidental behaviors. TRACG05 is the latest version and was originally developed to analyze Reactivity Initiated Accident (RIA). TRACG05 incorporates the same neutronics model of the latest core simulator with a three-group analytic-polynomial nodal expansion method. In addition to application to RIA safety analyses, TRACG05 has been planned to apply to safety analyses for Anticipated Operational Occurrences (AOOs) in BWRs by using a Best Estimate Plus Uncertainty (BEPU) methodology. To apply BEPU with TRACG05 to BWR AOOs, validations must be performed to evaluate the uncertainties of models relevant to important phenomena by comparing with appropriate test results for BWR AOOs. At first, a PIRT (Phenomena Identification and Ranking Table) was developed for each event scenario in AOOs to identify relevant physical processes and to determine their relative importance. According to the PIRT, an assessment matrix was established for separate effects tests (SETs), component effects tests (CETs), integral effects tests (IETs), and integral BWR plant start-up tests. The assessment matrix related the important phenomena to the test database, which was confirmed that all the important phenomena were covered by all tests specified in the matrix. According to the assessment matrix, comparison analyses have been specified to perform systematic and comprehensive validations of TRACG05 applicability to AOOs. The comparison analyses were done as the integrated code system with the up-stream reactor core design codes, therefore higher quality was enabled to evaluate the safety parameters. As the result, the uncertainties of important models in TRACG05 were determined so as to enable BEPU approaches for AOO safety issues. Here, as a SET, comparisons between TRACG05 and experimental data of void fraction in a bundle simulating an actual fuel bundle, which is one of the most important models in the application of TRACG05 to AOO analyses are shown. In addition, as pressurization event in AOOs, comparisons between TRACG05 and experimental data of Peach Bottom 2 Turbine Trip Test, which is one of integral tests for a BWR plant, are shown. This is the only test showing large neutron flux increase and strong coupling of neutron kinetics and thermal-hydraulics in the core due to void and Doppler feedbacks. Furthermore, a sensitivity analysis regarding a delay time of control rod (CR) insertion initiation which was the most sensitive uncertainty to the results is also shown.

Coupled Neutron Dynamic and Thermal Hydraulic Models of the HTR Simulator

2013 ◽  
Author(s):  
Zhe Sui ◽  
Jun Sun ◽  
Chunlin Wei ◽  
Yuanle Ma ◽  
Wenzhi Shan
Keyword(s):  
Nuclear Power ◽  
Dimensional Space ◽  
Three Dimensional ◽  
Reactor Core ◽  
Hydraulic Modeling ◽  
Thermal Hydraulics ◽  
Dynamic Simulations ◽  
Hydraulic Models ◽  
Coupling Techniques ◽  
Dimensional Space Time

Along with the design and construction of the high temperature reactor pebble bed module (HTR-PM), the engineering simulator system (ESS) has been accomplished to deal with the key techniques of the HTR simulator. In our previous papers, the three dimensional space-time neutron dynamics and the thermal hydraulic modeling of the HTR reactor core were introduced. In this paper, we concentrated on the detailed coupling techniques of the neutron dynamic and thermal hydraulic models, which was one of the most important assurances to the dynamic simulations of the HTR-PM. The mechanism of minus temperature feedback effect of the HTR was introduced by the materials participated in the coupling calculations, as well as those parameters being transferred. In parallel calculations, the neutron dynamics and thermal hydraulics were coupled by exchanging the power density and the temperature in fuels, moderates, reflectors and so on. With complete reactor core model, the coupling calculations of the neutron dynamics and thermal hydraulics were tested in many static and dynamic cases to show good performances. Based on that model ability, the ESS can simulate the start-up, shut-down and several accidents for the whole nuclear power plant.

Operational Model and its Validation with Drift Tests in Water Areas Around the Baltic Sea

1994 ◽  
Vol 29 (2-3) ◽  
pp. 293-308
Author(s):  
J. Koponen ◽  
M. Virtanen ◽  
H. Vepsä ◽  
E. Alasaarela
Keyword(s):  
Three Dimensional ◽  
Fast Response ◽  
Velocity Measurements ◽  
Short Term ◽  
Large Lakes ◽  
Operational Model ◽  
Water Currents ◽  
Emergency Situations ◽  
Sensitivity Tests ◽  
The Baltic

Abstract Three-dimensional (3-D) mathematical models of water currents, transport, mixing, reaction kinetic, and interactions with bottom and air have been used in Finland regularly since 1982 and applied to about 40 cases in large lakes, inland seas and their coastal waters. In each case, model validity has been carefully tested with available flow velocity measurements, tracer studies and water quality observations. For operational use, i.e., for spill combatting and sea rescue, the models need fast response, proven validity and illustrative visualization. In 1987-90, validated models were implemented for operational use at five sea areas along the Finnish coast. Further validation was obtained in model applications from nine documented or arranged cases and from seven emergency situations. Sensitivity tests supplement short-term validation. In the Bothnian Sea, it was nescessary to start the calculation of water currents three days prior to the start of the experiment to reduce initial inaccuracies and to make the coastal transport estimates meaningful.

scholarly journals Long Short-Term Memory Networks for Pattern Recognition of Synthetical Complete Earthquake Catalog

2021 ◽  
Vol 13 (9) ◽  
pp. 4905
Author(s):  
Chen Cao ◽  
Xiangbin Wu ◽  
Lizhi Yang ◽  
Qian Zhang ◽  
Xianying Wang ◽  
...  
Keyword(s):  
Pattern Recognition ◽  
Short Term Memory ◽  
Three Dimensional ◽  
Earthquake Risk ◽  
Earthquake Activity ◽  
Earthquake Catalog ◽  
Short Term ◽  
Term Memory ◽  
Earthquake Migration ◽  
Long Short Term Memory

Exploring the spatiotemporal distribution of earthquake activity, especially earthquake migration of fault systems, can greatly to understand the basic mechanics of earthquakes and the assessment of earthquake risk. By establishing a three-dimensional strike-slip fault model, to derive the stress response and fault slip along the fault under regional stress conditions. Our study helps to create a long-term, complete earthquake catalog. We modelled Long-Short Term Memory (LSTM) networks for pattern recognition of the synthetical earthquake catalog. The performance of the models was compared using the mean-square error (MSE). Our results showed clearly the application of LSTM showed a meaningful result of 0.08% in the MSE values. Our best model can predict the time and magnitude of the earthquakes with a magnitude greater than Mw = 6.5 with a similar clustering period. These results showed conclusively that applying LSTM in a spatiotemporal series prediction provides a potential application in the study of earthquake mechanics and forecasting of major earthquake events.

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