scholarly journals Electromagnetic Testing of Rod Cluster Control Assemblies in Pressured-Water Reactor Power Plants

2019 ◽  
Vol 9 (19) ◽  
pp. 4013
Author(s):  
Minhhuy Le ◽  
Sunbo Sim ◽  
Jinyi Lee
Keyword(s):  
Power Plants ◽  
Estimation Error ◽  
Depth Estimation ◽  
Reactor Power ◽  
Hall Sensor ◽  
Pressurized Water ◽  
Control Rod ◽  
Small Width ◽  
Electromagnetic Testing ◽  
Control Rods

This paper presents an electromagnetic testing system for rod cluster control assemblies used in pressurized-water reactors. The system uses several encircling-type magnetic cameras equivalent to a number of the control rods; each sensor probe composes of an encircling Hall sensor array (EHaS) and a bobbin coil. The EHaS has 16 Hall sensor elements that measure the electromagnetic field distribution in the radial direction of the control rod induced by the bobbin coil for defects. Experiments are performed on artificial defects on the cladding tube of real control rods to simulate short-circumferential grooves (SCGs), sliding wears (SWs), and circumferential cracks (CCs). The system can inspect the artificial SCGs, SWs, and CCs with depths up to 20%, 30%, and 40% of the cladding tube thickness (0.47 mm), respectively. Furthermore, the shape and depth of the defects could be estimated. The standard deviations of depth estimation are 18%, 5.8%, and 6.0% for CCs, SCGs, and SWs. The SCGs and SWs have a small and similar estimation error, but the CCs have the highest error of estimation, and have a small width of 0.2 mm.

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>

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.

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.

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 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.

Measurements of the 107Ag neutron capture cross sections with PHWT at the CSNS Back-n facility

2022 ◽  
Author(s):  
X. X. Li ◽  
L. X. Liu ◽  
W. Jiang ◽  
J. Ren ◽  
H. W. Wang ◽  
...  
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Astrophysics ◽  
Reaction Cross ◽  
Pressurized Water ◽  
Resonance Parameters ◽  
Control Rod ◽  
The Cross

Abstract Silver indium cadmium (Ag-In-Cd) control rod is widely used in pressurized water reactor nuclear power plants, and which is continuously consumed in a high neutron flux environment. The mass ratio of 107Ag in Ag-In-Cd control rod is 41.44%. To accurately calculate the consumption value of the control rod, a reliable neutron reaction cross section of the 107Ag is required. Meanwhile, 107Ag is also an important weak r nuclei. Thus, the cross sections for neutron induced interactions with 107Ag are very important both in nuclear energy and nuclear astrophysics. The (n, γ) cross section of 107Ag has been measured in the energy range of 1-60 eV using a back streaming white neutron beam line at China spallation neutron source. The resonance parameters are extracted by an R-matrix code. All the cross section of 107Ag and resonance parameters are given in this paper as datasets. The datasets are openly available at https://www.scidb.cn/s/aaUJbu.

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