Integrity Assessment of a German PWR RPV Considering Loss of Constraint

2010 ◽  
Author(s):  
Dieter Siegele ◽  
Igor Varfolomeev ◽  
Jo¨rg Hohe ◽  
Volker Hardenacke ◽  
Gerhard Nagel
Keyword(s):  
Fracture Toughness ◽  
Brittle Failure ◽  
Power Plants ◽  
Pressure Vessels ◽  
Significant Loss ◽  
Flange Joint ◽  
Local Approach ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Failure Assessment

The brittle failure assessment for five pressurized water reactor pressure vessels (RPV) of German nuclear power plants (NPP) has been revisited according to an advanced state of the art. Besides of recent innovation in fracture toughness curves and reference temperatures being already in the codes, also the effect of loss of constraint had to be considered when fracture toughness values determined from deep cracks in fracture toughness specimen with high multi-axial state of stress were transferred to crack configurations in the component. Thus, the available concepts were compared for their fitness for purpose, i.e. for their ability to give a fracture toughness representative to the crack configuration or flaw postulate in the component. The results of the investigation reveal a significant lower constraint in the component resulting in increased fracture toughness and showing that the brittle failure assessment based on the high constraint fracture toughness from the standard specimens can be very conservative. For consideration of the constraint conditions in the component besides the deterministic T-stress parameter also probabilistic local approach concepts based on the Weibull model were used which have the advantage of considering both the local stress strain field and the material volume under high loading. The loss of constraint was determined for several flaw postulates in the leading situations on the RPV being the coolant inlet nozzle corner and the flange joint. A considerable loss of constraint was demonstrated for flaw postulates with broken clad in the ferritic nozzle corner. Also in the flange joint the loss of constraint is evident for small flaws. In addition, for flaw postulates under the intact cladding the loss of constraint is remarkably higher than with broken postulated cladding. In summary, with the measured material toughness and the significant loss of constraint a considerable inherent margin against brittle failure can be demonstrated for the investigated load cases.

Brittle Failure Assessment of a PWR-RPV for Normal-Anomalous Operation and Loss of Coolant Accident Loads

2005 ◽  
Author(s):  
Dieter Siegele ◽  
Igor Varfolomeyev ◽  
Gerhard Nagel
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Brittle Failure ◽  
Thermal Loading ◽  
Crack Arrest ◽  
Flange Joint ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Failure Assessment ◽  
Loss Of Coolant

The brittle failure assessment for the reactor pressure vessel (RPV) of a 1300 MW pressurized water reactor (PWR) was revised according to the advanced state of the art. The RPV steel is 22 NiMoCr 37 (A 508 Cl.2). The expected neutron fluence at End of License (EOL) after 32 years of full operation is Φ < 2.3·1018 cm−2. The assessment followed a multi-barrier concept to prove independently exclusion of crack initiation, crack arrest and exclusion of the load necessary to advance the arrested cracks through the RPV wall. For operational conditions, a combination of shut-down and anomalous operation was proven as the leading transient. This transient follows a load path decreasing with temperature and thus producing a warm pre-stressing for flaws in the RPV. Therefore, a warm pre-stress (WPS) effect is to be considered. For loss of coolant accident (LOCA) conditions, a leak size screening was performed to find the leading transients. Enveloping transients were determined and used for the brittle failure assessment. As covering crack postulates circumferential cracks in the flange joint to the vessel and axial cracks in the nozzle corner were investigated both under the assumption of a broken clad. For the LOCA loads, the leading leak size was determined by finite element analyses and fracture mechanics based assessment of different combinations of cold/hot leg leak position and cold/hot leg injection of emergency core cooling (ECC). Leading situations are in the flange joint for the cylinder and in the nozzle corner for the flange. In the nozzle corner, high thermal loading from cross corner cooling produces large plastic deformation and loss of constraint. In this situation, the fracture toughness KIC as determined from deeply cracked fracture mechanics specimens is not representative for the component. Thus, loss of constraint in the component was considered by the application of a constraint modified master curve to determine the constraint representative fracture toughness. For the investigated load cases the relevant part of the load paths is at upper shelf temperatures, with the exception of the anomalous operation starting at 50°C to lower temperatures. The results of the investigation demonstrate preclusion of initiation for normal and anomalous operation as well as for LOCA loads. For LOCA, two further barriers were proven with crack arrest for postulated crack extension well within allowable depth in the RPV wall and with the preclusion of the load necessary to advance the arrested crack through the wall.

Constraint Based Assessment of Postulated Nozzle Corner Cracks

2002 ◽  
Author(s):  
Dieter Siegele ◽  
Igor Varfolomeyev ◽  
Gerhard Nagel
Keyword(s):  
Fracture Toughness ◽  
Brittle Failure ◽  
Thermal Stresses ◽  
Crack Tip ◽  
Pressure Vessels ◽  
Significant Loss ◽  
Large Shift ◽  
Corner Crack ◽  
Crack Tip Constraint ◽  
Corner Cracks

Brittle failure for reactor pressure vessels (RPV) under loss of coolant accidents (LOCA) considering strip and plume cooling show the nozzle corner as leading region in load level (e.g. Siegele et al., 1999). For such transients postulated nozzle corner cracks are also leading in crack driving force compared to the fracture toughness of the material. On the other hand, the crack tip constraint and consequently the failure probability is reduced by yielding caused by high thermal stresses and in addition by the crack size being small relative to the bulk of the flange material. Under these conditions, the flaw assessment in the nozzle corner using fracture toughness data obtained on standard specimens with high constraint level is over-conservative. In this situation it is suitable to introduce the loss of crack tip constraint into the brittle failure analysis of the nozzle corner. This investigation focuses on the assessment of surface cracks in the nozzle corner of an RPV. Using the finite element method temperature and stress calculations are carried out for a postulated LOCA event. For crack postulates in the nozzle corner of various size and geometry in the nozzle corner, crack driving parameters (such as the J-integral and the stress intensity factor) are determined as functions of the crack tip temperature. To account for the crack tip constraint, the T-stress is then evaluated and used along with the master curve approach as suggested by Wallin (2001). Significant loss of constraint is found for the nozzle corner resulting in a large shift of the fracture toughness curve to lower temperatures, thus excluding initiation of the crack postulates.

The Fracture Toughness Properties of China Manufactured Reactor Pressure Vessel Steels in Transition Temperature Range

2018 ◽  
Author(s):  
Kai Sun ◽  
Xiaoyong Wu ◽  
Guoyun Li ◽  
Bang Wen
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Temperature Range ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Over the past decade, many generation III pressurized water reactor power plants have been under construction in China. Most reactor pressure vessel steels for these plants construction are homemade. Historically, Charpy V-notch specimens are predominantly used to monitor the toughness of RPV steels. However, fracture toughness provides the quantitative predictions of the critical crack size and the allowable stress in structural integrity assessment. This paper evaluates the fracture toughness properties of China manufactured RPV steels directly measured in transition temperature range by using master curve method. Some specimens were irradiated in the High Flux Engineering Test Reactor. The influences of loading rate, test temperature, specimen configuration and neutron irradiation on T0 were also investigated. The experimental results show that China manufactured RPV steels exhibit good fracture toughness properties.

Application of Probabilistic Fracture Mechanics to Reactor Pressure Vessel Using PASCAL4 Code

2020 ◽  
Vol 143 (2) ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Thermal Transients ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Probabilistic fracture mechanics (PFM) is considered to be a promising methodology in structural integrity assessments of pressure-boundary components in nuclear power plants since it can rationally represent the inherent probabilistic distributions for influence parameters without over-conservativeness. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 which enables the failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor (PWR) was conducted using PASCAL4, and the effects of nondestructive examination (NDE) and neutron flux reduction on failure frequencies of the RPV were quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for probabilistic integrity assessments of embrittled RPVs and can enhance the applicability of PFM methodology.

Neutronics Studies on a Generic Pressure-Channel Reactor With Nuclear Steam Superheat

2014 ◽  
Author(s):  
Matthew Baldock ◽  
Wargha Peiman ◽  
Andrei Vincze ◽  
Rand Abdullah ◽  
Khalil Sidawi ◽  
...  
Keyword(s):  
Power Distribution ◽  
Power Plants ◽  
Saturation Pressure ◽  
Pressure Vessels ◽  
Limiting Factor ◽  
Outlet Temperature ◽  
Pressurized Water ◽  
Water Reactors ◽  
1960S And 1970S ◽  
Pressure Channel

In order to increase the thermal efficiency of steam-cycle power plants it is necessary to achieve steam temperatures as high as possible. Current limiting factor for Nuclear Power Plants (NPPs) in achieving higher operating temperatures and, therefore, thermal efficiencies is pressures at which they can operate. From basic thermodynamics it is known that to increase further an outlet temperature in water-cooled reactors a pressure must also be increased. Current level of pressures in Pressurized Water Reactors (PWRs) is about 15–16 MPa. Therefore, next stage should be supercritical pressures, at least 23.5–25 MPa. However, such supercritical-water reactors with pressure vessels of 45–50 cm thickness don’t exist yet. One way around larger pressure vessels as well as the limit of temperature of the coolant on the saturation pressure is to employ a Pressure Channel (PCh) design with Superheated Steam channels (SHS). PCh reactors allow for different coolants and bundle configurations in one reactor core, in this case, steam would be a secondary coolant. In the 1960s and 1970s the USA and Soviet Union tested reactors using pressure channels to super-heat steam in-core to achieve outlet temperatures greater than what is currently possible with convention reactors. Nuclear materials are carefully chosen based on their neutron interaction properties in addition to their strength and resistance to corrosion. Introducing steam channels will not only change the neutronics behavior of the coolant, but require different fuel cladding and pressure-channel materials, specifically, stainless steels or Inconels, to withstand high-temperature steam. This paper will investigate the affect that steam, SS-304 and Inconel will have on neutron economy when introduced into a reactor design as well as required changes to fuel enrichment. It will also be necessary to investigate the effects of these material changes on power distribution inside a reactor. Pressure-channel design requires methods of fine control to maintain a balanced core-power distribution, the introduction of non-uniform coolant and reactor materials will further complicate maintaining uniform reactor power. The degree to which SHS channels will affect the power distribution is investigated in this paper.

Alternative RTNDT for Linde 80 Weld Materials

2002 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Screening Criteria ◽  
Master Curve Method ◽  
Asme Code ◽  
Curve Method

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.

Estimation of Through-Wall Cracking Frequency of RPV Under PTS Events Using PFM Analysis Method for Identifying Conservatism Included in Current Japanese Code

2014 ◽  
Author(s):  
Kazuya Osakabe ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Structural Integrity ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Assessment Procedure ◽  
Analysis Method ◽  
Integrity Assessment ◽  
The U.S

To assess the structural integrity of reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events, the deterministic fracture mechanics approach prescribed in Japanese code JEAC 4206-2007 [1] has been used in Japan. The structural integrity is judged to be maintained if the stress intensity factor (SIF) at the crack tip during PTS events is smaller than fracture toughness KIc. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of pressure components has become attractive recently because uncertainties related to influence parameters can be incorporated rationally. A probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated taking frequencies of event occurrence and crack arrest after crack initiation into consideration. In this study, in order to identify the conservatism in the current RPV integrity assessment procedure in the code, probabilistic analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007, that a semi-elliptic axial crack is postulated on the inside surface of RPV wall, is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Evaluation of the Crack Interaction and Failure Behavior of Components With Crack Fields Using a Damage Mechanical Approach

2019 ◽  
Author(s):  
Christian Swacek ◽  
Patrick Gauder ◽  
Michael Seidenfuss
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Base Material ◽  
Pressure Vessels ◽  
Failure Behavior ◽  
Multiple Cracks ◽  
Critical Crack ◽  
Ongoing Research ◽  
Integrity Assessment ◽  
Crack Tips

Abstract In 2012 non-destructive testing measurements (NDT) of the reactor pressure vessels (RPV) in the Belgian Nuclear Power Plants Doel 3 and Tihange 2 revealed a high quantity of indications in the upper and lower core shells. The most likely explanation is that the measured indications are hydrogen flakes positioned in segregated zones in the base material of the pressure vessels. These hydrogen flakes have a laminar and quasi-laminar orientation with an inclination up to 15° to the pressure retaining surface. Under internal pressure, the crack tips undergo predominantly mixed mode loading conditions, where the induced stress and strain fields of the single crack tips influence each other. The safety assessment of crack afflicted pressurized components is performed by fracture mechanical approaches. For the evaluation of multiple cracks in crack fields, state of the art codes and standards apply interaction criteria and grouping methods, to determine a representative crack, which has to be evaluated. In this paper, the interaction of cracks in crack fields is numerically and experimentally evaluated. Damage mechanical models based on the Rousselier- and the Beremin model are used to investigate numerically the interaction of cracks in crack fields. Experimental data from ferritic flat tensile specimens afflicted with cracks are used to verify the numerical results. The damage mechanical calculations reveal critical crack arrangements due to coalescence behavior and cleavage fracture probability. These results and ongoing research intends the derivation of interaction criteria for cracks in crack fields. The interaction criteria will be used for the definition of a representative flaw for a conservative integrity assessment of crack afflicted components.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
Author(s):  
M. Niffenegger ◽  
O. Costa Garrido ◽  
D. F. Mora ◽  
G. Qian ◽  
R. Mukin ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

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