scholarly journals EXAMINATION OF PEACH BOTTOM HTGR CONTROL ROD, CONTROL ROD GUIDE TUBE, AND THERMALLY RELEASED SHUTDOWN ROD CO8-01 AFTER 300 EFFECTIVE FULL POWER DAYS OF OPERATION.

1970 ◽  
Author(s):  
O.M. Stansfield
Keyword(s):  
Guide Tube ◽  
Control Rod ◽  
Full Power

Predicting Thermal Mixing and Fatigue Inside Control Rod Guide Tubes

2013 ◽  
Author(s):  
Eric Lillberg
Keyword(s):  
Thermal Fatigue ◽  
Turbulent Mixing ◽  
Low Frequency ◽  
Experimental Measurements ◽  
Guide Tube ◽  
Control Rod ◽  
Thermal Mixing ◽  
Purge Flow ◽  
Control Rods ◽  
Good Agreement

The cracked control rods shafts found in two Swedish NPPs were subjected to thermal fatigue due to mixing of cold purge flow with hot bypass water in the upper part of the top tube on which the control rod guide tubes rests. The interaction between the jets formed at the bypass water inlets is the main source of oscillation resulting in low frequency downward motion of hot bypass water into the cold purge flow. This ultimately causes thermal fatigue in the control rod shaft in the region below the four lower bypass water inlets. The transient analyses shown in this report were done to further investigate this oscillating phenomenon and compare to experimental measurements of water temperatures inside the control rod guide tube. The simulated results show good agreement with experimental data regarding all important variables for the estimation of thermal fatigue such as peak-to-peak temperature range, frequency of oscillation and duration of the temperature peaks. The results presented in this report show that CFD using LES methodology and the open source toolbox OpenFOAM is a viable tool for predicting complex turbulent mixing flows and thermal loads.

Numerical Analysis of the Corium Behavior Within the Fuel Support Piece by MPS

2016 ◽  
Author(s):  
Ronghua Chen ◽  
Lie Chen ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Eutectic Reaction ◽  
Flow Behavior ◽  
Three Dimensional ◽  
Severe Accident ◽  
Guide Tube ◽  
Melt Flow ◽  
Control Rod ◽  
Mps Method ◽  
Moving Particle ◽  
Accident Conditions

In the typical boiling water reactor (BWR), each control rod guide tube supports four fuel assemblies via an orificed fuel support piece in which a channel is designed to be a potential corium relocation path from the core region to the lower head under severe accident conditions. In this study, the improved Moving Particle Semi-implicit (MPS) method was adopted to analyze the melt flow and ablation behavior in this region during a severe accident of BWR. A three-dimensional particle configuration was constructed for analyzing the melt flow behavior within the fuel support piece. Considering the symmetry of the fuel support piece, only one fourth of the fuel support was simulated. The eutectic reaction between Zr (the material of the corium) and stainless steel (the material of the fuel support piece) was taken into consideration. The typical melt flow and freezing behaviors within the fuel support piece were successfully reproduced by MPS method. In all the simulation cases, the melt discharged from the hole of the fuel support piece instead of plugging the fuel support piece. The results indicate that MPS method has the capacity to analyze the melt flow and solidification behavior in the fuel support piece.

Estimation of control rod worth, xenon effect on reactivity and power defect of BAEC TRIGA Mark-II research reactor

2018 ◽  
Vol 33 (39) ◽  
pp. 1850233
Author(s):  
Md. Mehedi Hassan ◽  
K. M. Jalal Uddin Rumi ◽  
Md. Nazrul Islam Khan ◽  
Rajib Goswami
Keyword(s):  
Theoretical Analysis ◽  
Low Power ◽  
Temperature Effects ◽  
Research Reactor ◽  
Power Level ◽  
Fuel Temperature ◽  
Control Rod ◽  
Full Power ◽  
Triga Mark Ii ◽  
Control Rods

In this work, control rod worth, xenon (Xe) effect on reactivity and power defect have been measured by doing experiments in the BAEC TRIGA Mark-II research reactor (BTRR) and through established theoretical analysis. Firstly, to study the xenon-135 effect on reactivity, reactor is critical at 2.4 MW for several hours. Next, experiments have been performed at very low power (50 W) to avoid temperature effects. Moreover, for the power defect experiment, different increasing power level has been tested by withdrawing the control rods. Finally, it is concluded that the total control rods worth of the BAEC TRIGA Mark-II research reactor, as determined through this study, is enough to run the reactor at full power (3 MW) considering the xenon-135 and fuel temperature effects.

Experimental Analyses by SIMMER-III on Duct-Wall Failure and Fuel Discharge/Relocation Behavior

2013 ◽  
Author(s):  
Hidemasa Yamano ◽  
Yoshiharu Tobita
Keyword(s):  
Three Dimensional ◽  
Computer Code ◽  
Failure Behavior ◽  
Guide Tube ◽  
Control Rod ◽  
Duct Wall ◽  
Experimental Analyses ◽  
Multi Phase ◽  
Pool Formation ◽  
Good Agreement

This paper describes experimental analyses using SIMMER-III/IV, which are two/three-dimensional multi-component multi-phase Eulerian fluid-dynamics codes, for the purpose of the code validation. Two topics of key phenomena in core disruptive accidents were presented in this paper: duct-wall failure and fuel discharge/relocation behavior. To analyze the duct-wall failure behavior, the SCARABEE BE+3 in-pile experiments were selected. The SIMMER-III calculation was in good agreement with the overall event progression; which was characterized by coolant boiling, clad melting, fuel failure, molten pool formation, duct-wall failure, etc.; observed in the experiment. The CAMEL C6 experiment investigated the fuel discharge and relocation behavior through a simulated control rod guide tube, which is important in evaluating the neutronic reactivity. SIMMER-IV well simulated fuel-coolant interaction, sodium voiding, fuel relocation behavior observed in the experiment. These experimental analyses indicated the validity of the SIMMER-III/IV computer code for the duct wall failure and fuel discharge/relocation behavior.

Effect of CRGT Cooling on Modes of Global Vessel Failure of a BWR Lower Head

2012 ◽  
Author(s):  
Walter Villanueva ◽  
Chi-Thanh Tran ◽  
Pavel Kudinov
Keyword(s):  
Heat Transfer ◽  
Vessel Wall ◽  
Reactor Vessel ◽  
Creep Model ◽  
Guide Tube ◽  
Mechanical Loads ◽  
Control Rod ◽  
Melt Pool ◽  
Creep Analysis ◽  
Lower Head

An in-vessel stage of a severe core melt accident in a Nordic type Boiling Water Reactor (BWR) is considered wherein a decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. This process induces creep leading to a mechanical failure of the reactor vessel wall. The focus of this study is to investigate the effect of Control Rod Guide Tube (CRGT) and top cooling on the modes of global vessel failure of the lower head. A coupled thermo-mechanical creep analysis of the lower head is performed and cases with and without CRGT and top cooling are compared. The debris bed heat-up, re-melting, melt pool formation, and heat transfer are calculated using the Phase-change Effective Convectivity Model and transient heat transfer characteristics are provided for thermo-mechanical strength calculations. The creep analysis is performed with the modified time hardening creep model and both thermal and integral mechanical loads on the reactor vessel wall are taken into account. Known material properties of the reactor vessel as a function of temperature, including the creep curves, are used as an input data for the creep analysis. It is found that a global vessel failure is imminent regardless of activation of CRGT and top cooling. However, if CRGT and top cooling is activated, the mode and timing of failure is different compared to the case with no CRGT and top cooling. More specifically, with CRGT and top cooling, there are two modes of global vessel failure depending on the size of the melt pool: (a) ‘ballooning’ of the vessel bottom for smaller pools, and (b) ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool for larger pools. Without CRGT and top cooling, only a ballooning mode of global vessel failure is observed. Furthermore, a considerable delay (about 1.4 h) on the global vessel failure is observed for the roughly 30-ton debris case if CRGT and top cooling is implemented. For a much larger pool (roughly 200-ton debris), no significant delay on the global vessel failure is observed when CRGT and top cooling is implemented, however, the liquid melt fraction and melt superheat are considerably higher in non-cooling case.

Flow mixing inside a control-rod guide tube – Experimental tests and CFD simulations

2011 ◽  
Vol 241 (12) ◽  
pp. 4803-4812 ◽  
Author(s):  
Kristian Angele ◽  
Ylva Odemark ◽  
Mathias Cehlin ◽  
Bengt Hemström ◽  
Carl-Maikel Högström ◽  
...  
Keyword(s):  
Experimental Tests ◽  
Guide Tube ◽  
Cfd Simulations ◽  
Control Rod ◽  
Flow Mixing

Influencing Factors of the HTR-10 Core Dynamics in Reactivity Insertion ATWS Tests

2016 ◽  
Author(s):  
Feng Gou ◽  
Fubing Chen ◽  
Yujie Dong
Keyword(s):  
Influencing Factors ◽  
Reactor Power ◽  
Sensitivity Analyses ◽  
Simulation Accuracy ◽  
Control Rod ◽  
Poisoning Effect ◽  
Core Dynamics ◽  
New Energy ◽  
Full Power ◽  
Reactivity Insertion

After the full power operation of the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10), several safety demonstration tests, representing the anticipated transient without scram (ATWS) conditions, were successfully performed on this reactor. Among these tests, two reactivity insertion ATWS tests were conducted by withdrawing a single control rod without reactor scram at 30% rated power. In the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, these two tests have been reanalyzed using the THERMIX code, and the code itself was strictly checked through the test data. According to the previous code benchmark activities utilizing the HTR-10 tests, the temperature coefficient of reactivity (TCR), the residual heat level (RHL) and the xenon poisoning effect (XPE) could be considered the most important influencing factors of the THERMIX simulation accuracy for the core dynamics. In this study, sensitivity analyses are performed on the basis of the assumed variations of TCR, RHL and XPE. The impacts of these concerned parameters on the reactor power transient are qualitatively identified.

Numerical Simulation of Flow Field Inside Reactor Upper Plenum for PWR With Code_Saturne

2017 ◽  
Author(s):  
Jiawei Liu ◽  
Puzhen Gao ◽  
Tingting Xu ◽  
Jiesheng Min ◽  
Guofei Chen
Keyword(s):  
Stokes Equations ◽  
Fluid Dynamic ◽  
Structure Design ◽  
Navier Stokes ◽  
Guide Tube ◽  
Flow Induced Vibration ◽  
Local Pressure ◽  
Pressurized Water ◽  
Local Flow ◽  
Control Rod

Flow characteristic in upper plenum has a strong influence on reactor functional margin and rod cluster control assembly (RCCA) guide tube wear. Upper plenum flow governs loops flow rate measurement via hot leg temperature which has also an influence on the reactor protection system. For RCCA guide tube wear, it appears in operation with RCCA flow-induced vibration, leading to its replacement. It is important to know the flow condition in the upper plenum, and in particular the outlet. Existing Generation III reactors have their own specialties on the design. Comparison between current technologies is a good way for better understanding on the key structure design for the upper plenum. In this paper, simplified models based on upper plenum structure of Korean advanced pressurized water reactor (PWR) and Westinghouse design AP1000 are constructed and meshed with a volume around 6 million cells to obtain a 3-dimensional global and local flow distributions inside the upper plenum and to characterize the vital flow features for reactor safety. The Navier-Stokes equations are solved with standard k-ε turbulence model by using EDF in-house open source computational fluid dynamic (CFD) software: Code_Saturne. Through calculations, pressure and velocity distributions are obtained, axial and lateral variations have been analyzed. Compared with APR1400, it can be observed that for the design of AP1000, the rotational flow entrained in the edge of upper plenum and high velocity area due to the hot leg suction effect contribute to the relatively lower local pressure, and may have an impact on the drop velocity of control rod.

scholarly journals PRELIMINARY POWER TRANSIENT ANALYSIS OF THE SUPER FR WITH AXIALLY HETEROGENEOUS CORE

2021 ◽  
Vol 247 ◽  
pp. 07002
Author(s):  
Tsutomu Okui ◽  
Akifumi Yamaji
Keyword(s):  
Temperature Gradient ◽  
Transient Analysis ◽  
Large Temperature ◽  
Control Rod ◽  
Mox Fuel ◽  
Fuel Layer ◽  
Axial Temperature ◽  
Short Period ◽  
Full Power ◽  
Supercritical Water Cooled Reactor

The Super FR is one of the SuperCritical Water cooled Reactor (SCWR) concepts with once-through direct cycle plant system. Recently, new design concept of axially heterogeneous core has been proposed, which consists of multiple layers of MOX and blanket fuels. To clarify the safety performance during power transient, safety analyses have been conducted for uncontrolled control rod (CR) withdrawal and CR ejection at full power. RELAP/SCDAPSIM code was used for the safety analysis. The results show that the peak cladding surface temperature (PCST) is high in the upper MOX fuel layer. It is also shown that axial temperature gradient of cladding greatly increases in a short period. Suppressing such large temperature gradient may be a design issue for the axially heterogeneous core from the viewpoint of ensuring fuel integrity.

Validation of Analysis Models on Relocation Behavior of Molten Core Materials in Sodium-Cooled Fast Reactors Based on the Melt Discharge Experiment

2020 ◽  
Author(s):  
Kai Igarashi ◽  
Ryoji Onuki ◽  
Takaaki Sakai ◽  
Shinya Kato ◽  
Ken-ichi Matsuba ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Heat Removal ◽  
Early Discharge ◽  
Core Material ◽  
Bottom Plate ◽  
Guide Tube ◽  
Control Rod ◽  
Analysis Models

Abstract In order to improve the safety of nuclear power plants, it is necessary to make sure measures against their severe accidents. Especially, in the case of a sodium-cooled fast reactor (SFR), there is a possibility of significant energy release due to formation of a large-scale molten fuel pool accompanied by re-criticality in the event of a core disruptive accident (CDA). It is important to ensure in-vessel retention that keeps and confines damaged core material in the reactor vessel even if the CDA occurs. CDA scenario initiated by Unprotected Loss Of Flow (ULOF), which is a typical cause of core damage, is generally categorized into four phases according to the progression of core-disruptive status, which are the initiating, early-discharge, material-relocation and heat-removal phases for the latest design in Japan. During the material-relocation phase, the molten core material flows down mainly through the control rod guide tube and is discharged into the inlet coolant plenum below the bottom of the core. The discharged molten core material collides with the bottom plate of the inlet plenum. Clarification of the accumulation behavior of molten core material with such a collision on the bottom plate is important to reduce uncertainties in the safety assessment of CDA. In present study, in order to make clear behavior of core melt materials during the CDAs of SFRs, analysis was conducted using the SIMMER-III code for a melt discharge simulation experiment [1] in which low-melting-point alloy was discharged into a shallow water pool. This report shows the validation results for the melt behavior by comparing with the experimental data.

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