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.

The IRIS Spool-Type Reactor Coolant Pump

2002 ◽  
Author(s):  
J. M. Kujawski ◽  
D. M. Kitch ◽  
L. E. Conway
Keyword(s):  
Reactor Core ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Steam Generators ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
The Us ◽  
Loss Of Coolant ◽  
The Individual ◽  
Integral Reactor

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called “spool” pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper.

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.

Steam Generator of the International Reactor Innovative and Secure

2002 ◽  
Author(s):  
L. Cinotti ◽  
M. Bruzzone ◽  
N. Meda ◽  
G. Corsini ◽  
C. V. Lombardi ◽  
...  
Keyword(s):  
Steam Generator ◽  
Reactor Core ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Design Activity ◽  
Steam Generators ◽  
The Us ◽  
The Subject ◽  
The Individual ◽  
Integral Reactor

IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the main reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long-life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. The design of the steam generators, which are internally contained within the reactor vessel, is a major design effort in the development of the integral IRIS concept. The ongoing design activity about the steam generator is the subject of this paper.

The Application of Concurrent Installation of the RCS Cross Over Leg and the Reactor Vessel Internals

2010 ◽  
Author(s):  
K.-W. Park ◽  
J.-H. Bae ◽  
S.-H. Park
Keyword(s):  
Feasibility Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Study Results ◽  
Under Construction

The reactor vessel internals (RVI) of a pressurized water reactor (PWR) must be installed precisely in the reactor vessel (RV) according to the requirements for levelness, orientation and vertical alignments for its proper functions and structural integrity. For the precise installation, deformation of the RV should be controlled during the RVI installation. Traditionally, the RVI has been installed in the RV after the completion of welding work for large bore pipings in the reactor coolant system (RCS). To reduce installation time, the concurrent installation of the RVI and RCS pipings is investigated. This paper describes the feasibility study on the concurrent installation including the Finite Element Method (FEM) analyses of the RV deformation due to the welding and heat treatment of the pipings. Based on the feasibility study results, the optimum schedule of the RVI installation in parallel with the installation of the cross-over leg pipings (reactor coolant pump inlet pipings) and confirmation measurement locations are developed. Thereby the concurrent installation will be applied to the nuclear power plants under construction in Korea, and it is expected to reduce installation period of 2 months compared to the traditional sequential installation method.

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.

Effects of Tube Rupture Modeling Methods on MSSV Lift Time Following a MSGTR Event in PWR

2002 ◽  
Author(s):  
Ji Hwan Jeong ◽  
Ki Yong Choi ◽  
Keun Sun Chang
Keyword(s):  
Steam Generator ◽  
Flow Rate ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Best Estimate ◽  
Main Steam ◽  
Safety Systems ◽  
Estimate System

A multiple steam generator tube rupture (MSGTR) event in APR1400, an advanced pressurized water reactor, is investigated using the best estimate thermal hydraulic system code, MARS1.4. The effects of parameters such as the number of ruptured tubes, rupture location, affected steam generator on analysis of the MSGTR event in APR1400 are taken into account. In particular, the effects of tube rupture modeling are compared. In the present study, single tube (STM) and double tube modeling (DTM) are examined for assessment on the main steam safety valve (MSSV) lift time. Nuclear steam supply system (NSSS) and several safety systems that are relevant to the APR1400 are modeled. Automatic safety systems are assumed to mitigate the MSGTR events including the reactor protection trip, reactor coolant pump trip, the pressurizer heaters, high-pressure safety injection (HPSI) pumps, and the valves for atmospheric dump, main steam safety, main steam isolation, and turbine stop and bypass. When five tubes are ruptured, the STM permits the operator response time of 2085 seconds before lifting of MSSVs. The effects of rupture location on the MSSV lift time is not significant in case of STM, while the MSSV lift time for tube-top rupture is found to be 25.3% larger than that for rupture at hog-leg side tube sheet in case of DTM. The MSSV lift time for the cases that both steam generators are affected (4C5x, 4C23x) are found to be larger than that for the single steam generator cases (4A5x, 4B5x) due to a bifurcation of the primary leak flow. The discharge coefficient of Cd is found to affect the MSSV lift time only for smaller value of Cd below 0.5. For larger values of Cd than 0.5, its effect on the leak flow rates as well as the MSSV lift time become negligible. It is found that the most dominant parameter governing the MSSV lift time is the leak flow rate. Whichever modeling method is used, it gives the similar MSSV lift time if the leak flow rate is similar, except the case of both steam generators are affected. Therefore, the system performance and the MSSV lift time of the APR1400 are strongly dependent on the break flow model used in the best estimate system code.

Characteristics of Transition Boiling in Sodium-Heated Steam Generator Tubes

1979 ◽  
Vol 101 (2) ◽  
pp. 270-275 ◽  
Author(s):  
D. M. France ◽  
R. D. Carlson ◽  
T. Chiang ◽  
R. Priemer
Keyword(s):  
Thermal Stress ◽  
Steam Generator ◽  
Tube Wall ◽  
Water Pressure ◽  
Thermal Fluctuations ◽  
Straight Tube ◽  
Steam Generators ◽  
Steam Generator Tubes ◽  
Heated Length ◽  
Boiling Zone

Thermal fluctuations were measured in the tube wall in the transition boiling zone of a full-scale LMFBR sodium-heated steam generator tube. The tube had an inside diameter = 10 mm, wall thickness = 2.90 mm, heated length = 13.1 m, and material = 2 1/4 Cr-1 Mo steel. Water flowed vertically upwards inside the straight tube, and sodium flowed counter-currently in a surrounding annulus. Results of thermal, spectral, and thermal stress analyses are presented for a test within the normal operating range of LMFBR steam generators. Results of other tests are presented that show the effects and sensitivity of sodium temperature and water pressure on the severity of the thermal fluctuations.

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