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.

Optimization design for mechanical seals in reactor coolant pumps based on a fluid–solid strong-interaction model

2011 ◽  
Vol 225 (8) ◽  
pp. 1851-1862 ◽  
Author(s):  
C-J Liao ◽  
W-F Huang ◽  
Y-M Wang ◽  
S-F Suo ◽  
X-F Liu
Keyword(s):  
Strong Interaction ◽  
Optimization Design ◽  
Power Plants ◽  
Interaction Model ◽  
Reactor Power ◽  
Mechanical Seals ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
The Operating Mechanism ◽  
And Performance

The study on the mechanism and performance of the mechanical seals in reactor coolant pumps (RCPs) is very important for the safe operations of pressurized water reactor power plants. By exploring the operating mechanism of the first seal of the hydrostatic mechanical seal in RCPs, an analytical fluid–solid strong-interaction model of the seal is proposed in this article. The model holds that the mechanical deformations of the seal assembly are dominated by the deflections of the seal rings, and this idea is demonstrated by the numerical simulation result of a fluid–solid interaction (FSI) model. Using the analytical FSI model, the regularity that the leakage rate of the first seal varies with the differential pressure in a RCP is obtained, and compared with the operational data, which is used to verify the model. Based on the understanding of the reliability of the seal, a dimensionless parameter Λ that acts as an attribute to the reliability is proposed in this article. Using the analytical FSI model and Λ as the optimization algorithm and optimization object, respectively, the optimum designs about the seal faceplateconfigurations are performed. Also, the specific optimization conclusions are given simultaneously.

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.

Steam Generator Replacement Project at TVA’s Sequoyah Unit 1 Nuclear Power Plant

2004 ◽  
Author(s):  
Myron R. Anderson
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Power ◽  
Original Design ◽  
Pressurized Water ◽  
Repair Work ◽  
Building Roof ◽  
Tennessee Valley ◽  
Generator Replacement

Pressurized Water Reactor Power Plants have at times required that large components be replaced (steam generators weighing 750,000 lbs) which have necessitated performing first time modifications to the plant that were unintended during the original design. The steam generator replacement project at Tennessee Valley Authority (TVA’s) Sequoyah Nuclear Power Station necessitated (1) two large temporary openings (21’×45’) in the plant’s Shield Building roof (2’ thick concrete) by hydro-blasting to allow the removal of the old generators and installation of the new, (2) removal and repair of the concrete steam generator enclosure roofs (20’ diameter, 3’ thick) which were removed by wire saw cutting and (3) the seismic qualification of; the design and construction of an extensive ring foundation for; the use of one of the world largest cranes to remove these components through the roof. This removal and replacement process had to be performed in an expeditious manner to minimize the amount of time the plant is shutdown so the plant could return to providing power to the grid. This paper will address some of the many technical and construction considerations required to perform this demolition and repair work safely, efficiently and in a short as possible duration.

Power Maximization of Fully Passive Lead-Bismuth Eutectic Cooled Small Modular Reactor

2014 ◽  
Author(s):  
Jaehyun Cho ◽  
Yong-Hoon Shin ◽  
Il Soon Hwang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Reactor Core ◽  
Design Stage ◽  
Pressurized Water ◽  
Small Modular Reactor ◽  
Power Maximization ◽  
Lead Bismuth Eutectic ◽  
Water Reactors

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.

Coordinated Control of a Small Pressurized Water Reactor

2018 ◽  
Author(s):  
Peiwei Sun ◽  
Chong Wang
Keyword(s):  
Control System ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressurized Water Reactors ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Average Temperature ◽  
Water Reactors

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

Stability of Cement-Glass Packages Containing Sodium Borate Salt Generated from Pressurized Water Reactor Power Plants

1987 ◽  
Vol 78 (2) ◽  
pp. 185-190 ◽  
Author(s):  
Tatsuo Izumida ◽  
Fumio Kawamura ◽  
Koichi Chino ◽  
Makoto Kikuchi
Keyword(s):  
Power Plants ◽  
Reactor Power ◽  
Sodium Borate ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

scholarly journals Investigation Of Rod Control System Reliability Of Pwr Reactors

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Syaiful Bakhri
Keyword(s):  
Control System ◽  
System Reliability ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Fault Tree ◽  
Reactor Core ◽  
Pressurized Water ◽  
Control Rod ◽  
Control Rods

<p class="NoSpacing1"><span lang="IN">The Rod Control System is </span>employed<span lang="IN"> to adjust the position of the control rods in the reactor core </span>which corresponds with <span lang="IN">the thermal power generated in the core </span>as well as <span lang="IN">the electric power generated in the turbine. In a Pressurized Water Reactor (PWR) type nuclear power plants, the control-rod drive </span>employs <span lang="IN">magnetic stepping-type mechanism. This </span>type of <span lang="IN">mechanism consists of a pair of circular coils and latch-style jack with the armature. When the </span>electric <span lang="IN">current </span>is <span lang="IN">supplied to the coils sequentially, the control-rods</span>, which <span lang="IN">are held on the drive shaft</span>, can be driven<span lang="IN"> up</span>ward<span lang="IN"> or down</span>ward<span lang="IN"> in increments. </span>This <span lang="IN">sequential current </span>c<span lang="IN">ontrol</span> drive<span lang="IN"> system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or </span>known also as <span lang="IN">the Rod Control System (RCS). The p</span>urpose of this paper is to investigate the RCS reliability <span lang="IN">of APWR </span>using <span lang="IN">the Fault Tree Analysis (FTA)</span> method<span lang="IN"> since </span>the analysis of reliability which considers<span lang="IN"> the FTA</span> for common CRDM <span lang="IN">can </span>not <span lang="IN">be found</span> in <span lang="IN">any </span>public references. <span lang="IN">The FTA method is used to model the system reliability by developing the fault tree diagram of the system. </span>The<span lang="IN"> results show that the failure of the system is very dependent on the failure of most of the individual systems. However, the failure of the system does not affect the safety of the reactor, since the reactor trips immediately if the system fails. The evaluation results also indicate that the Distribution Panel is the most critical component in the system.</span></p>

scholarly journals INTELLIGENT TECHNIQUES FOR ANOMALY DETECTION IN NUCLEAR REACTORS

2021 ◽  
Vol 247 ◽  
pp. 21011
Author(s):  
George Ioannou ◽  
Thanos Tagaris ◽  
Georgios Alexandridis ◽  
Andreas Stafylopatis
Keyword(s):  
Signal Processing ◽  
Neutron Flux ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Machine Learning Techniques ◽  
Full Potential ◽  
Pressurized Water ◽  
Learning Techniques ◽  
Water Reactors

The safe operation of nuclear power plants is highly dependent on the ability of quickly and accurately identifying possible anomalies and perturbations in the reactor. Operational defects are primarily diagnosed by detectors that capture changes in the neutron flux, placed at various points inside and outside of the core. Neutron flux signals are subsequently analyzed with signal processing techniques in an effort to be better described (have their higher-order characteristics uncovered, locate transient events, etc). To this end, the application of intelligent techniques may be extremely beneficial, as it may assist and extend the current level of analysis. Besides, the combination of signal processing methodologies and machine learning techniques in the framework of nuclear power plant data is an emerging topic that has yet to show its full potential. In this context, the current contribution attempts at introducing intelligent approaches and more specifically, deep learning techniques, in neutron flux signal analysis for the identification of perturbations and other anomalies in the reactor core that may affect its operational capabilities. The obtained results of an initial stage of analysis on neutron flux signals captured at pressurized water reactors are encouraging, underlying the robustness and the potential of the proposed approach.

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