Piping System Structural Integrity Simulation for Post LOCA Water Hammer Loads

2002 ◽  
Author(s):  
B. W. Manning ◽  
T. Stevens ◽  
G. Morandin ◽  
R. G. Sauve´ ◽  
R. Richards ◽  
...  
Keyword(s):  
Structural Damage ◽  
Structural Integrity ◽  
Water Hammer ◽  
Piping System ◽  
Linear Modeling ◽  
Linear Elastic ◽  
Beam Elements ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Non Linear

The Canadian Nuclear Safety Commission (CNSC) required as part of the operating license for Ontario Power Generation’s Darlington Nuclear Generating Station, that the structural integrity of the piping following a loss of coolant accident (LOCA) be demonstrated. This is necessary to ensure that no subsequent pressure boundary failures will impede the ability to maintain fuel cooling. The injection of cold emergency coolant following a LOCA creates the potential for the occurrence of condensation-induced water hammers (CIWH) in the primary heat transport (PHT) system piping. Classical linear elastic piping analysis using the class 1 NB-3656 rules of the ASME Boiler & Pressure Vessel Code failed to demonstrate the adequacy of the piping and/or its supports that were designed using the linear elastic rules of subsection NF for nine of the twelve piping models that comprise the PHT system. A decision was made to undertake a state-of-the-art non-linear explicit analysis in order to qualify the piping. Strain rather than stress limits would be applied similar to those being developed by ASME for nuclear packaging undergoing accidental impact during transportation. In order to address the feasibility of this approach, a non-linear analysis was performed on a portion of one of the piping systems. The piping was modeled as shells and again as beam elements with and without detailed modeling of the supports. After these initial simulations, it was determined that the piping could be modeled with simplified beam elements, however, the supports would require a more detailed modeling in order to determine the extent of support damage and the effect the supports have on the integrity of the piping system itself. This paper addresses the non-linear modeling of the piping models and discusses the modeling details, assumptions and analysis results. This approach is shown to be a useful alternative for predicting the extent of structural damage that can be expected by a Level D event such as a condensation induced water hammer following a loss of coolant accident.

Development of a Method to Assess Impact Hazards

2015 ◽  
Author(s):  
Peter Gill ◽  
Adam Toft
Keyword(s):  
Structural Integrity ◽  
Nuclear Reactors ◽  
Primary Circuit ◽  
Plant Safety ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Particle Hydrodynamics ◽  
Smoothed Particle ◽  
Circuit Components ◽  
Safety Systems

The structural integrity of pressure retaining primary circuit components and of the containment boundary is of great significance for the safety justification of all nuclear reactors. In this regard, it is important to understand the vulnerability of safety related components to potential accidents. Whilst the direct consequences of pipe or pressure vessel failure, for example in terms of the extent of loss of coolant, may be tolerable, the indirect consequences of failure may not. A Loss of Coolant Accident (LOCA) may indirectly damage plant safety systems, including the containment boundary, due to the effect of missiles generated by the LOCA. This paper describes a study to develop a modern numerical modelling technique to estimate damage by missiles. Smoothed Particle Hydrodynamics (SPH) has been applied to simulate the acceleration of the missile due to the fluid jet.

Calculation of Forces on Reactor Containment Fan Cooler Piping

2004 ◽  
Author(s):  
Joseph S. Miller ◽  
Kevin Ramsden
Keyword(s):  
Water Flow ◽  
Piping System ◽  
Pressurized Water ◽  
Hydraulic Loading ◽  
Hydraulic Behavior ◽  
Loss Of Coolant Accident ◽  
Service Water ◽  
Loss Of Coolant ◽  
Hydraulic Load ◽  
Water Pumps

The purpose of this paper is to present the results of the Reactor Containment Fan Cooler (RCFC) system piping load calculations. These calculations are based on piping loads calculated using the EPRI methodology (Refs. 1 & 2) and RELAP5 (Ref. 3) to simulate the hydraulic behavior of the system. The RELAP5 generated loads were compared to loads calculated using the EPRI GL 96-06 methodology. This evaluation was based on a pressurized water reactor’s RCFC coils thermal hydraulic behavior during a Loss of Offsite Power (LOOP) and a loss of coolant accident (LOCA). The RCFC consist of two banks of service water and chill water coils. There are 5 SX and 5 chill water coils per bank. Therefore, there are 4 RCFC units in the containment with 2 banks of coils per RCFC. Two Service water pumps provide coolant for the 4 RCFC units (8 banks total, 2 banks per RCFC unit and 2 RCFC units per pump). Following a LOOP/LOCA condition, the RCFC fans would coast down and upon being reenergized, would shift to low-speed operation. The fan coast down is anticipated to occur very rapidly due to the closure of the exhaust damper as a result of LOCA pressurization effects. The service water flow would also coast down and be restarted in approximately 43 seconds after the initiation of the event. The service water would drain from the RCFC coils during the pump shutdown and once the pumps restart, water is quickly forced into the RCFC coils causing hydraulic loading on the piping. Because of this scenario and the potential for over stressing the piping, an evaluation was performed by the utility using RELAP5 to assess the piping loads. Subsequent to the hydraulic loads being analyzed using RELAP5, EPRI through GL 96-06 provided another methodology to assess loads on the RCFC piping system. This paper presents the results of using the EPRI methodology and RELAP5 to perform thermal hydraulic load calculations and compares them.

The Effect of Piping System Non-Linearity on the Unstable Growth of a Circumferential Through-Wall Crack in a Pipe

2003 ◽  
Author(s):  
E. Smith
Keyword(s):  
Cross Section ◽  
Material Properties ◽  
Structural Integrity ◽  
Crack Size ◽  
Instability Criterion ◽  
Piping System ◽  
Linear Elastic ◽  
Piping Systems ◽  
Simple Model System ◽  
Linear Manner

During the last twenty-five years, considerable attention has been given to the structural integrity of steel piping systems, and in particular to the effect of circumferential cracks on their integrity. From a safety perspective, it is important that any crack, say for example a stress corrosion crack or fatigue crack, will not develop into a through-wall crack which will then propagate unstably, thus leading to a guillotine rupture and possibly a pipe whip scenario. One way of guaranteeing that this does not happen is to ensure that unstable growth of a circumferential through-wall crack is unable to occur. An appropriate methodology is based on tearing modulus concepts with the instability criterion being expressed in the form TAPP > TMAT where TAPP is the applied tearing modulus, a measure of the crack driving force, and TMAT is the material tearing modulus, a measure of the material’s crack growth resistance. With a piping system that behaves in a linear elastic manner, TAPP involves only the system’s geometry parameters and the crack size but not the magnitudes of the applied loadings or the material properties of the cracked cross-section; the behaviours of the cracked cross-section and the remainder of the piping system are therefore decoupled. If, however, the system behaves in a non-linear manner say, for example, as a result of excessive deformation arising as a consequence of large deformations, then TAPP also involves the material properties of the cracked cross-section; material and piping system geometry parameters are then not decoupled in the instability criterion. The paper illustrates this point by analysing a simple model system where the non-linearity arises from excessive deformation at a connection.

Effects of Residual Stress by Weld Overlay Cladding and PWHT on the Structural Integrity of RPV During PTS

2007 ◽  
Author(s):  
Makoto Udagawa ◽  
Jinya Katsuyama ◽  
Kunio Onizawa
Keyword(s):  
Residual Stress ◽  
Structural Integrity ◽  
Crack Size ◽  
Operating Conditions ◽  
Residual Stress Field ◽  
Weld Overlay ◽  
Loss Of Coolant Accident ◽  
System Pressure ◽  
Loss Of Coolant ◽  
Main Steam

In order to assess the structural integrity of a reactor pressure vessel (RPV), it is assumed that a surface crack resides through the cladding at the inner surface of the vessel. It is, therefore, important to precisely evaluate stress intensity factor (SIF) under the residual stress field due to weld overlay cladding and post-weld heat treatment (PWHT). In this work, numerical simulation based on thermal-elastic-plastic-creep analysis using finite element method was performed to evaluate residual stress distribution near the cladding layer produced by weld overlay cladding and PWHT. The tensile residual stress of about 400 MPa occurs in the cladding at room temperature after the PWHT. The residual stress distributions under the normal operating conditions (system pressure and temperature) of RPV were also evaluated. The effect of residual stress and evaluation methods on SIF behavior for various crack size were studied under typical pressurized thermal shock (PTS) conditions such as small break loss of coolant accident (SBLOCA), main steam line break (MSLB) and large break loss of coolant accident (LBLOCA). It is clarified from comparison of this weld simulation with the other simple methods that SIF is affected by residual stress by weld overlay cladding and PWHT.

Seismic Design Margin of the Piping and Support System: Part 3—Evaluation of Seismic Margin of the Piping System

2010 ◽  
Author(s):  
Eiji Shirai ◽  
Takanori Yamada ◽  
Kazutoyo Ikeda ◽  
Toshiaki Yoshii ◽  
Masami Kondo ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Support System ◽  
Earthquake Engineering ◽  
Nuclear Power ◽  
System Response ◽  
Piping System ◽  
Linear Modeling ◽  
Safety Margins ◽  
Non Linear

Seismic safety is one of the major key issues of nuclear power plant safety in Japan. It is demonstrated that nuclear piping possesses large safety margins through the piping and support system test, which consisted of three dimensional piping, supports, U-bolts, and concrete anchorages, using the E-defense vibration table of National Research Institute for Earth Science and Disaster Prevention, Hyogo Earthquake Engineering Research Center, on extremely high seismic excitation level [1,2,3]. In the above test, the non-linear hysteretic behaviors of the support are quite complicated, but the dissipated energies introduce large damping effects on the piping system response. In order to evaluate the inelastic behavior of the support with respect to the whole piping system response, the following simulation methodology for the support re-evaluation is proposed. 1) Non-linear modeling of the support: • Failure mode and failure capacity of each support. • Simplified non-linear modeling of each support. 2) Simulation Analysis of the piping and support system: • Considering the non-linearity both of the supports and elbows in the piping system. 3) Evaluation of seismic margin: • Focused on the failure level for the support system, and the fatigue damage for the strain range of the piping. The limit state analysis of the typical piping system of a nuclear power plant is presented in this paper, and it is demonstrated that these evaluations of the seismic margins would give important insight into the support reinforcement program on the seismic re-evaluation work.

Re-Evaluation of the Structural Integrity of the Chinese Qinshan 300-MWe Reactor Pressure Vessel Under Pressurized Thermal Shock Using the Master Curve Method

2015 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Hu Hui ◽  
Hui Li ◽  
Fuzhen Xuan
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Master Curve ◽  
Structural Integrity ◽  
Master Curve Method ◽  
Loss Of Coolant Accident ◽  
Pressurized Thermal Shock ◽  
Loss Of Coolant ◽  
Curve Method ◽  
Reactor Pressure

Pressurized thermal shock (PTS) is a potential major threat to the structural integrity of the reactor pressure vessel (RPV) in a nuclear power plant. A comprehensive structural integrity analysis of the Chinese Qinshan 300-MWe RPV subjected to PTS events including the small break loss-of-coolant accident (SB-LOCA) and large break loss-of-coolant accident (LB-LOCA) transients was performed by Shanghai nuclear engineering and design institute (SNERDI). The J-integral values at the deepest and the near cladding-base interface points of the crack were calculated with the linear elastic material model. And the RTPTS values were determined by the tangent approach. In the case that the RTNDT at or beyond the RPV design life may exceed the RTPTS according to the previous analysis procedure, the objective of this paper is to apply the Master Curve method to the re-evaluation of the integrity of this RPV, taking account of constraint and crack length effects. The over-conservatism in the previous evaluation is identified by comparing the new calculation with the previous one. The new RTPTS values are increased to varied extents for the different loading transients.

scholarly journals Assessment on Structural Integrity of In-service Machine Using De-noised Vibrational Modal Data and Artificial Neural Network

2018 ◽  
Vol 237 ◽  
pp. 03002 ◽  
Author(s):  
Zhi Chao Ong ◽  
Ee Teng Yap ◽  
Zubaidah Ismail ◽  
Shin Yee Khoo
Keyword(s):  
Modal Analysis ◽  
Structural Damage ◽  
Structural Integrity ◽  
Oil And Gas ◽  
Oil And Gas Industry ◽  
Industrial Applications ◽  
Piping System ◽  
Daily Production ◽  
Integrity Assessment ◽  
Modal Data

The recent oil price drop creates a demand for swift action within oil and gas industry to shift focus from increasing daily production rates, to optimizing existing assets in achieving growth. Industrial machinery, one of the industry’s key asset many times failed due to high amplitude vibration that contributes to accelerated wear and tear and subsequently results in high cycle fatigue failure. As such there is a need to develop a structural integrity assessment for in–service machinery for continuous and safe operation. Vibration–based method such as Experimental Modal Analysis (EMA) is widely used for damage detection on civil and piping system under stationary environment. However, in industrial applications, system shutdown is very costly. EMA is also undesirable in this case due to the dominant ambient and system disturbances on the in–service system. An alternative method called Impact-Synchronous Modal Analysis (ISMA) is developed to perform modal analysis under noisy environment. Applying the ISMA technique in de-noising the non–synchronous disturbances at upstream could generate a cleaner and static–like modal data downstream for analysis. Artificial Neuron Networks (ANN) is then applied extensively in structural damage identification purposes based on changes in modal data due to its excellent pattern recognition ability. By leveraging on the latest technologies, i.e. ISMA and ANN as proposed, it allows real–time monitoring of assets, in this case, the machines, as well as the ability to transform continuous streams of data into useful information to predict damages.

scholarly journals Application of Inelastic Method and Its Comparison with Elastic Method for the Assessment of In-Box LOCA Event on EU DEMO HCPB Breeding Blanket Cap Region

2021 ◽  
Vol 11 (19) ◽  
pp. 9104
Author(s):  
Anoop Retheesh ◽  
Francisco A. Hernández ◽  
Guangming Zhou
Keyword(s):  
Failure Modes ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Inelastic Analysis ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Institute Of Technology ◽  
Elastic Design ◽  
The Given ◽  
Karlsruhe Institute

The Helium Cooled Pebble Bed (HCPB) breeding blanket, being developed by the Karlsruhe Institute of Technology (KIT) and its partners is one of the two driver blanket candidates to be selected for the European demonstration fusion power plant (EU DEMO). The in-box Loss of Coolant Accident (LOCA) is a postulated initiating event of the breeding blanket (BB) that must be accounted within the design basis. In this paper, the BB cap region is analyzed for its ability to withstand an in-box LOCA event. Initially, an assessment is performed using conventional elastic design codes for nuclear pressure vessels. However, it is thought that the elastic rules are not ‘equipped’ to assess the material damage modes which are essentially inelastic. Therefore, a non-linear inelastic analysis is further performed to better understand the damage in the material. Two predominant inelastic failure modes are thought to be relevant and addressed: exhaustion of ductility and plastic flow localization. While the design of HCPB BB has been predominantly based on the elastic design-by-analysis studies, results from the present study show that the elastic rules may be overly conservative for the given material and loading and could lead to inefficient designs. To our knowledge, this study is the first attempt to investigate the structural integrity of the European DEMO blankets under in-box LOCA conditions using the inelastic methods.

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