scholarly journals Potential for low fracture toughness and lamellar tearing on PWR steam generator and reactor coolant pump supports. Resolution of generic technical activity A-12

1979 ◽  
Author(s):  
R.P. Snaider ◽  
J.M. Hodge ◽  
H.A. Levin ◽  
J.J. Zudans
Keyword(s):  
Fracture Toughness ◽  
Steam Generator ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Lamellar Tearing

Steam Generator and Reactor Coolant Pump Coupled Flow Field Analysis

2010 ◽  
Author(s):  
Shuan Xia ◽  
Minghui Weng ◽  
Jian Qiu ◽  
Xinxin Pan
Keyword(s):  
Flow Field ◽  
Steam Generator ◽  
Field Analysis ◽  
Normal Operation ◽  
Plant Design ◽  
Coupled Flow ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Flow Field Analysis ◽  
Test Result

In the design of some passive PWRs, Reactor Coolant Pump (RCP) is welded directly to the Steam Generator (SG) channel head. This design cancels the support of RCP and simplifies the layout of Reactor coolant system. What’s more, this design also do good to the mitigation of SBLOCA. But this design makes the flow field in the SG channel head and RCP inlet complex and there may be vortex in this flow field for which reason the SG outlet resistance will increase and affect the long-term steady operation. For this issue, some company made tests on it. But the cost of test is high and the applicability of the test result is limited. If the parameters or components size changed a little, the test result will be no longer applicable. To solve this problem, this article considers using 3-D CFD flow field analysis software to analyse the SG and RCP coupled flow field. Through steps of 3D model establishing – meshing in Gambit – analyzing in Fluent, this paper obtains the flow filed condition of SG-RCP coupled part during normal operation and so as to support plant design.

Structural Integrity Evaluation of High Strength Anchor Bolt for Supporting Steam Generator and Reactor Coolant Pump in Nuclear Power Plants

2013 ◽  
Author(s):  
Eun-Mo Lim ◽  
Nam-Su Huh ◽  
Hee-Jin Shim ◽  
Chang-Kyun Oh ◽  
Hyun-Su Kim
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
High Strength ◽  
Nominal Stress ◽  
Coolant Pump ◽  
Anchor Bolt ◽  
Reactor Coolant ◽  
Anchor Bolts

In Korea, a fitness-for service evaluation for assuring structural integrity of high strength anchor bolts which support nuclear components such as steam generator and reactor coolant pump, has been one of the important issues in nuclear industry. The main failure mechanism of high strength anchor bolts supporting nuclear components might be degradation due to stress corrosion cracking and brittle fracture. In the present work, the structural integrity of high strength anchor bolts which are used to support steam generator and reactor coolant pump of one of the Korean older vintage nuclear power plants is evaluated by adopting a procedure proposed by Electric Power Research Institute (EPRI) based on an elastic fracture mechanics concept. In this EPRI’s procedure, an accurate estimation of nominal stress acting on the cross section of the bolt is a crucial element since a structural integrity of an anchor bolt is evaluated in the EPRI’s procedure using this nominal stress incorporating reference flaw factors reflecting effects of stress concentration due to bolt thread and reference sized surface crack. In this context, detailed elastic finite element stress analyses are firstly performed on the anchor bolt assemblies to come up with nominal stress in the cross-section of anchor bolt. As for loading condition, bolt pretention as well as normal and faulted loads of the anchor bolts were considered. In addition, the structural integrity of the anchor bolts is demonstrated by comparing nominal stresses of anchor bolts with the maximum allowable stresses obtained by using the EPRI’s reference flaw factors and critical fracture toughness. Furthermore, the accuracy of EPRI’s reference flaw factors which are derived on the assumption that reference sized surface crack is existed on the thread roots is investigated using 3-dimensional elastic finite element fracture mechanics analyses.

Effect of Support Friction in Thermal Stress of VVER-1000 RCS for NOP Condition

2014 ◽  
Author(s):  
Truong Quang Nguyen ◽  
Ihn Namgung
Keyword(s):  
Thermal Stress ◽  
Yield Stress ◽  
Steam Generator ◽  
Reactor Vessel ◽  
Inside Surface ◽  
Steam Generators ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Hexahedral Elements ◽  
Effect Of Support

The main purpose of this research is to investigate the effect of friction in the thermal stress of Reactor Coolant System (RCS) of VVER-1000. RCS is a large system connecting reactor vessel, steam generators and RC Pumps. During the heat-up of reactor, the RCS expand and during cool-down of reactor, it contracts. Because of the heavy weight of reactor and steam generator, the friction at the support of RCS affects the thermal stress of RCS. In this paper how much support friction contributes to the development of thermal stress is assessed in order to investigate the thermal stress and effect of support friction. A quarter-symmetry model of VVER-1000 RCS is developed in ANSYS and meshed with hexahedral elements to ensure better solution accuracies. The model includes reactor vessel, steam generator and reactor coolant pump. Internals of reactor vessel, steam generators and RCPs are represented by point mass to simplify the model. Temperature of inside surface of hot-leg side of reactor vessel to inlet side of steam generator is assumed same uniform hot-leg temperature, and the temperature of inside surface of outlet side of steam generator to reactor vessel is uniform cold-leg temperature. All outside surface are assumed insulated. The analysis includes neither transient thermal loading nor dynamic loadings. The analysis results show that friction at support brings little effect on the peak thermal stress. The peak thermal stress occurs at hot-leg nozzle of reactor pressure vessel and it approached near yield stress. If load combination is included the localized total stress at hot-leg nozzle could go over the yield stress. This peak stress could affect fatigue life in a long run. A recommendation is made that a detailed fatigue analysis of VVER-1000 RCS is necessary.

Numerical and Experimental Research on the Fluid-Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump

2019 ◽  
Vol 141 (6) ◽  
Author(s):  
Rui Xu ◽  
Yun Long ◽  
Yaoyu Hu ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Large Mass ◽  
Stiffness Coefficient ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Various Boundary Conditions ◽  
Clearance Flow ◽  
Mass Coefficient ◽  
Cfd Method ◽  
Canned Motor

Reactor coolant pump (RCP) is one of the most important equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor RCP, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a long clearance flow. The fluid-induced forces of the clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally analyzed in this work. A transient computational fluid dynamics (CFD) method has been used to investigate the fluid-induced force of the clearance. A vertical experiment rig has also been established for the purpose of measuring the fluid-induced forces. Fluid-induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the CFD method and the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor RCP does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid-induced forces of the clearance flow.

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