Pressurized Thermal Shock Loading Research Articles

Elastic and elastoplastic fracture analysis of a reactor pressure vessel under pressurized thermal shock loading

Abstract Due to the limitation of the manufacturing process and improper operation, there may be defects near the inlet nozzles of reactor pressure vessel (RPV). Under pressurized thermal shock (PTS), the crack region is subjected to a high tensile stress, and the material toughness is gradually decreased in the cooling process. So it is necessary to evaluate the integrity of RPV in the thermo-mechanical coupling field. The 3-D finite element model is established for the beltline region around the inlet nozzles of a RPV. According to the calculated nil-ductility reference temperature, the elastic constitutive relation is firstly established for the fracture analysis. The stress intensity factors along the crack tips are compared with the material toughness. Using the XFEM, the crack propagation path is obtained to verify the predicted results. At relatively low reference temperatures, the ductility characteristics of the nuclear material can be exhibited. In this case, the ultimate bearing capacity of the structure is demonstrated by the elastoplastic fracture analysis. Finally, the safe range of reference temperature is summed up to prevent the surface crack from running through the whole wall thickness of the vessel during the PTS transient.

Coupled RELAP5, 3D CFD and FEM analysis of postulated cracks in RPVs subjected to PTS loading

The fracture mechanic analysis of a reactor pressure vessel subjected to pressurized thermal shock loading is one of the most important issues for the assessment of life time extension of a nuclear power plant. The most severe scenario occurs during cold water injection in the cold leg due to a Loss-Of-Coolant Accident (LOCA). In the present study a comprehensive fracture mechanics analysis is performed. Two hypothetical LOCAs are assumed for an adopted reference design of a two-loop Pressurized Water Reactor. Boundary conditions obtained from the RELAP5 code are used as input for Computational Fluid Dynamics (CFD) simulations. For the structural integrity analysis, submodeling technique and the eXtended Finite Element Method (XFEM) based on temperatures calculated by CFD are applied. The results from the 3D FEM calculations are compared to those from a simplified axisymmetric model based on axisymmetric thermal hydraulic model results. The analysis identifies the worst crack orientation and location. It also proves that a complete model is needed for a correct analysis as the simplified model is not conservative and fails to describe accurately the local plume effect.

The probabilistic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

The pressurized thermal shock (PTS) event poses a potentially significant challenge to the structural integrity of the reactor pressure vessel (RPV) during the long time operation (LTO). In the USA, the “screening criteria” for maximum allowable embrittlement of RPV material, which forms part of the USA regulations, is based on the probabilistic fracture mechanics (PFM). The FAVOR software developed by Oak Ridge National Laboratory (ORNL) is used to establish the regulation. As the technical basis of FAVOR is not the most widely-used and codified methodologies, such as the ASME and RCC-M codes, in most countries (with exception of the USA), proving RPV integrity under the PTS load is still based on the deterministic fracture mechanics (DFM). As the maximum nil-ductility-transition temperature (RTNDT) of the beltline material for the 54 French RPVs after 40 years operation is higher than the critical values in the IAEA-TECDOC-1627 and European NEA/CSNI/R(99)3 reports (while still obviously lower than the “screening criteria” of the USA), it may conclude that the RPV will not be able to run in the LTO based on the DFM. In the FAVOR, the newest developments of fracture mechanics are applied, such as the warm pre-stress (WPS) effect, more accurate estimation of the flaw information and less conservation of the toughness (such as the three-parameter Weibull distribution of the fracture toughness). In this paper, the FAVOR software is first applied to show both the methodology and the results of the PFM, and then the limits in the current FAVOR software (Version 6.1, which represents the baseline for re-assessing the regulation of 10 CFR 50.61), lack of the impact of the constraint effect, ductile–tearing initiation failure analysis, elastic–plastic fracture mechanics analysis, safety assessment of the crack front interface points and interaction among multiple cracks, are also discussed.

The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

The deterministic structural integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. While the nil-ductility-transition temperature (RTNDT) parameter is widely used, the influence of fluence and temperature distributions along the thickness of the base metal wall cannot be reflected in the comparative analysis. This paper introduces the method using a structure safety margin (SM) parameter which is based on a comparison between the material toughness (the fracture initiation toughness KIC or fracture arrest toughness KIa) and the stress intensity factor (SIF) along the crack front for the integrity analysis of a RPV subjected to PTS transients. A 3-D finite element model is used to perform fracture mechanics analyses considering both crack initiation assessment and arrest assessment. The results show that the critical part along the crack front is always the clad-base metal interface point (IP) rather than the deepest point (DP) for either crack initiation assessment or crack arrest assessment under the thermal load. It is shown that the requirement in Regulatory Guide 1.154 that ‘axial flaws with depths less than 20% of the wall thickness and all circumferential flaws should be modeled in infinite length’ may be non-conservative. As the assessment result is often poor universal for a given material, crack and transient, caution is recommended in the safety assessment, especially for the IP. The SIF reduces under the thermal or pressure load if the map cracking (MC) effect is considered. Therefore, the assumption in the ASME and RCCM codes that the cladding should be taken into account in determining the stress fields is conservative when there are small cracks in the cladding (considering MC effect).

Structural integrity assessment of the reactor pressure vessel under the pressurized thermal shock loading

Fracture mechanics analysis of pressurized thermal shock (PTS) is the key element of the integrity evaluation of the nuclear reactor pressure vessel (RPV). While the regulation of 10 CFR 50.61 and the ASME Code provide the guidance for the structural integrity, the guidance has been prepared under conservative assumptions. In this paper, the effects of conservative assumptions involved in the PTS analysis were investigated. The influence of different parameters, such as crack size, cladding effect and neutron fluence, were reviewed based on 3-D finite element analyses. Also, the sensitivity study of elastic–plastic approach, crack type and cladding thickness were reviewed. It was shown that crack depth, crack type, plastic effect and cladding thickness change the safety margin (SM) significantly, and the SM at the deepest point of the crack is not always smaller than that of the surface point, indicating that both the deepest and surface points of the crack front should be considered. For the reference transient, deeper cracks always give more conservative prediction. So compared to the prescribed analyses of a set of postulated defects with varying depths in the ASME code, it only needs to assess the crack with maximum depth in the code for the reference transient according to the conclusions.

Recent Japanese research activities on probabilistic fracture mechanics for pressure vessel and piping of nuclear power plant

This paper describes a review of recent Japanese activities on probabilistic fracture mechanics (PFM) analyses. Japan Atomic Energy Agency (JAEA: previously JAERI) had sponsored research committees on PFM organized by Japan Society of Mechanical Engineers (JSME) and Japan Welding Engineering Society (JWES) for more than a decade. This work still continues with the same members in JWES. The purpose of the continuous activity is to provide probabilistic approaches in several fields of integrity problems of nuclear power plant. This paper shows some of the newest results of the JWES research committee. First topic is evaluation of the new JSME code case with rules of Fitness-For-Service from the view of PFM, including reactor pressure vessel subject to pressurized thermal shock loading, piping with a crack of the allowable size and effect of sizing accuracy for piping integrity. The next one is development of new PFM techniques including reliability assessment of piping with domestic (Japanese) SCC data and maintenance optimization of LWRs based on risk and economic models. The last topic is the international round robin program just starting from 2008.

Approximation method for the calculation of stress intensity factors for the semi-elliptical surface flaws on thin-walled cylinder

A simple approximation method for the stress intensity factor at the tip of the axial semi-elliptical cracks on the cylindrical vessel is developed. The approximation methods, incorporated in VINTIN (Vessel INTegrity analysis-INner flaws), utilizes the influence coefficients to calculate the stress intensity factor at the crack tip. This method has been compared with other solution methods including 3-D finite element analysis for internal pressure, cooldown, and pressurized thermal shock loading conditions. For these, 3-D finite-element analyses are performed to obtain the stress intensity factors for various surface cracks witht/R=0.1. The approximation solutions are within ±2.5% of the those of finite element analysis using symmetric model of one-forth of a vessel under pressure loading, and 1–3% higher under pressurized thermal shock condition. The analysis results confirm that the approximation method provides sufficiently accurate stress intensity factor values for the axial semi-elliptical flaws on the surface of the reactor pressure vessel.

Life management of reactor pressure vessels under pressurized thermal shock loading: deterministic procedure and application to Western and Eastern type of reactors

The integrity of the reactor pressure vessel (RPV) has to be maintained throughout the plant life and it is also one of the main considerations regarding the plant life extension. Adequate approaches to the RPV integrity assessment provides a basis for plant-safe operation and for timely implementation of preventive and corrective measures if necessary. A substantial part of the RPV integrity assessment is related to the pressurized thermal shock (PTS) analysis. The safety of the RPV during such a severe loading condition has to be proven in the design stage and during life of the component. Irradiation programs allow for an adequate description of the material response and by means of analytical and numerical methods the structural analysis can be performed. The PTS-analysis of the RPV is a multidisciplinary effort and involves among others detailed thermal hydraulic analysis and structural analysis including fracture mechanic assessment. Operational load situations during accident conditions are dominated by pressure and temperature transients during an emergency core cooling event following a loss of coolant accident relevant for PTS. Cold water is injected into the cold legs and falling into the downcomer of the RPV. Complex thermal hydraulic situations occur and the fluid temperatures in the downcomer as well as the RPV wall-to-fluid heat transfer coefficient are highly non-linear. To determine the spatial fluid temperature and wall-to-fluid heat transfer coefficient distribution in the downcomer and the respective thermal stresses with adequate accuracy, fluid–fluid mixing scenarios are taken into account. Plume cooling during a loss of coolant accident of the RPV leads to a locally increased axial stress distribution in the plume region compared to the axial symmetric cooling. The higher axial stresses make the circumferential crack more relevant for the safety assessment. The transient temperatures and stresses in the RPV-wall are calculated by means of 3D finite element (FE) calculations and are performed for given transients. The safety assessment deals with a safety margin against the material toughness curve or the determination of the allowable brittle to ductile transition temperature. Using the global thermal hydraulic parameters calculated with adequate system codes, the fluid temperature and RPV-wall-to-fluid heat transfer coefficient distribution in the downcomer has been calculated with the Siemens code KWU-MIX. In the following fracture mechanics calculation, two flaw types have been considered, axial and circumferential cracks. By means of 3D FE calculations the sub-surface flaws were investigated and from the results of the fracture parameters safety margins could be derived. In this report the Siemens procedure is applied as an example to an Eastern plant WWER-1000 type, but it has been successfully carried out for different types of RPV also: Western 3 and 4-loop plants (German and French type) where the results are reported in [Proc. ASME PVP Conf., (1999), Boston] [7] and [NUREG/CR-6651 (ORNL/TM-1999/231), Oak Ridge National Laboratory, (1999)] [5]. Although all considered pressure vessels are different in their construction: the French type is constructed with a thermal shield between RPV wall and internals, the Eastern type has two rings of injection nozzles, which leads to a complex thermal hydraulic situation the general procedure could be applied. Following results have been obtained: –plume cooling leads to increased axial stresses in the plume region –therefore circumferential flaws are more severe than axial flaws –safety margins for axial flaws are higher than for circumferential flaws.

Consistency of fracture assessment criteria for the NESC-1 thermal shock test

In this report, a best estimate analysis of the NESC Spinning Cylinder test is performed and the consistency of the state of the art application of the different fracture parameters and assessments criteria is examined. Direct fracture toughness measurements best describes the fracture behavior of the NESC Spinning Cylinder in the simulated pressurized thermal shock loading. The concept based on the ASME RT NDT reference temperature (either Pellini or Cv test) does not provide consistent description of the NESC1 fracture assessment. In general the fracture mode, i.e. ductile initiation and tearing followed by cleavage event, was successfully estimated by 3D FE based fracture assessment, combined with Master Curve description of fracture toughness.

Fracture mechanics analyses including constraint investigations on reactor pressure vessels under pressurized thermal shock loading

In the integrity assessment of reactor pressure vessels the emergency core cooling system injection is one of the main load cases to be analysed. For a more generic assessment, parametric analyses are done concerning the load pattern, the postulated crack geometry and the material properties. With detailed axisymmetric and 3D-finite element (FE) models of reactor pressure vessels, temperature distributions in the structures and the deformations are calculated by elastic-plastic FE analyses for axisymmetric and asymmetric strip-like cooling assumptions. With it, boundary conditions for FE detail models with partly and 360° circumferential cracks are determined. Different material properties for the circumferential weld near the core mid-height, the cladding and the base metal with estimated influence of neutron irradiation are considered. Crack loading based on the J-integral is calculated for different crack geometries, and different parameters describing the constraint on the crack ligament with proposals for a fracture assessment concept are discussed. Furthermore, simplified methods based on analytical formulas to calculate the stress distribution in the wall and the stress intensity factor are applied and compared with FE results.

Crack initiation and arrest in a pressurized thermal shock test for a model pressure vessel made of VVER-440 reactor pressure vessel steel

A joint pressure vessel integrity research programme involving three partners is being carried out during 1990–1995. The partners are the Central Research Institute of Structural Materials “Prometey” from Russia, IVO International Ltd (IVO) from Finland, and the Technical Research Centre of Finland (VTT). The main objective of the research programme is to increase the reliability of the VVER-440 reactor pressure vessel safety analysis. This is achieved by providing material property data for the VVER-440 pressure vessel steel, and by producing experimental understanding of the crack behaviour in pressurized thermal shock loading for the validation of different fracture assessment methods. The programme is divided into four parts: pressure vessel tests, material characterization, computational fracture analyses, and evaluation of the analysis methods. The testing programme comprises tests on two model pressure vessels with artificial axial outer surface flaws. The first model vessel had circumferential weld seam at the mid-length of the vessel. A special embrittling heat treatment is applied to the vessels before tests to simulate the fracture toughness at the end-of-life condition of a real reactor pressure vessel. The sixth test on the first model led to crack initiation followed by arrest. After the testing phase, material characterization was performed. Comparison of calculated and experimental data generally led to a good correlation, although the work is being continued to resolve the discrepancies between the measured initiation and arrest properties of the material.

Crack growth evaluation of a multimaterial CS-type specimen

During the German HDR Safety Program the integrity of a reactor pressure vessel was studied intensively during a pressurized thermal shock loading. As a part of the work this paper presents a stable crack growth evaluation for a side-grooved, multimaterial CS-type specimen, which was detached from the inner surface of the HDR reactor pressure vessel. The specimen was composed of ferritic base material and two austenitic cladding material layers. The specimen was tested at the Fraunhofer-Institut für Werkstoffmechanik, Freiburg, Germany, and revealed an extremely nonuniform stable crack growth distribution along the crack front. Elastic-plastic finite element analyses were carried out using the 85-version of the ADINA code. For crack growth simulation a separate subroutine package IWM-CRACK linked with ADINA was used. Additional J-integral calculations were carried out based on the ADINA results, using the VTTVIRT code in order to study the path dependence of the J-integral. The crack growth was controlled by J r curves, while actual measured data were available only for the base material, and hypothetical J r curves were applied for the two cladding layers. The calculated maximum load was slightly higher than the maximum load measured in the test. The crack growth was relatively well simulated in quality. The slight numerical discrepancies at some locations along the crack front were consistent with extreme local material brittleness, which was clearly underestimated by the assumed material properties.

Prediction of the first spinning cylinder test using continuum damage mechanics

For many years large-scale experiments have been performed world-wide to validate aspects of fracture mechanics methodology. Special emphasis has been given to correlations between small- and large-scale specimen behaviour in quantifying the structural behaviour of pressure vessels, piping and closures. Within this context, the first three spinning cylinder tests, performed by AEA Technology at its Risley Laboratory, addressed the phenomenon of stable crack growth by ductile tearing in contained yield and conditions simulating pressurized thermal shock loading in a PWR reactor pressure vessel. A notable feature of the test data was that the effective resistance to crack growth, as measured in terms of the J R-curve, was appreciably greater than that anticipated from small-scale testing, both at initiation and after small amounts (a few millimetres) of tearing. In the present paper, two independent finite element analyses of the first-spinning cylinder test (SC 1) are presented and compared. Both involved application of the Rousselier ductile damage theory in an attempt to understand better the transferability of test data from small specimens to structural validation tests. In each instance, the parameters associated with the theory's constitutive equation were calibrated in terms of data from notched-tensile and (or) fracture mechanics tests, metallographic observations and (or) chemical composition. The evolution of ductile damage local to the crack tip during SC 1 was thereby calculated and, together with a crack growth criterion based on the maximisation of opening-mode stress, used as the basis for predicting cylinder R-curves (angular velocity vs. Δa, J integral vs. Δa). Except in the initiation region, the results show the Rousselier model to be capable of predicting correctly the enhancement of tearing toughness of the cylinder relative to that of conventional test specimens, given an appropriate choice of finite element cell size in the region representing the crack tip. As such, they represent a positive step towards achieving the goal to establish continuum damage mechanics as a reliable predictive engineering tool.

Fracture mechanics investigations on cylindrical large scale specimens under thermal shock loading

To investigate the crack growth and crack arrest behaviour of primary circuit materials large scale experiments were conducted on component-like specimens under pressurized thermal shock loading at MPA Stuttgart. The material characteristics varied from high tough material to low tough material with higher nil ductility transition temperature to simulate EOL or beyond EOL-state. All tests started from in-service conditions and were cooled down to room temperature. The specimens showed both stable and unstable crack growth and partly crack arrest. The crack growth behaviour was verified by post test calculations and could be explained with the help of the multiaxiality of the stress state.

Cyclic and transient thermal loading of the HDR reactor pressure vessel with respect to crack initiation and crack propagation

Abstract In a series of thermal loading tests at the HDR reactor pressure vessel – thermal stratification, cyclic thermal shock and pressurized thermal shock – the methods applied in safety analysis had to become qualified by a continuous intercomparison of calculated results and experimental data. Above all the complex boundary conditions of the HDR-tests offer a close approximation to the original components, so that they provide a real assessment of the transferability. The results of the thermal mixing tests indicated that during cold water inflow into the RPV longitudinal strains build up in the cylindrical wall which dominate over that in circumferential direction. During the cyclic thermal fatigue tests incipient crack formation in the cladding as well as the behaviour of crack propagation in the cladding and in the base material was analyzed. In the pressurized thermal shock tests, the nozzle region and the cylinder wall in the incipient crack condition were loaded by long cooling streaks. Even in the aggravated loading condition as the result of a routed cold water streak no remarkable indications of crack growth were noticed. In both cases, cyclic and pressurized thermal shock loading, the expected crack propagation was overpredicted by the fracture mechanical methods used. The non-destructive examination methods used were able to locate all of the cracks but they mostly overpredicted the actual crack depth.

Fracture evaluation of surface cracks in light water reactor pressure vessels under pressurized thermal shock loading

AbstractThe evaluation of pressure vessel integrity affected by pressurized thermal shock (PTS) events caused by emergency cooling due to a small leak break is being addressed. Comprehensive numerical and experimental investigations were performed at MPA Stuttgart and the HDR plant.During cold water feeding necessary to maintain core cooling under accident conditions, high temperature gradients arise which cause thermal stresses in the walls near the inner surface of the reactor pressure vessel (RPV). Supposing there are flaws in the inner wall the thermal stresses can increase the stress intensity considerably compared with the stresses caused by internal pressuse. st the same time, the fract/fre toughness of the ferritic reactor pressure vessel matenalls reduced through the coolmg process. Therefore, the possibility of stable or unstable crack growth cannot be ruled out.This paper describes numerical and experimental investigations with large-scale heavy section specimens with circumferential and semi-e...

A reproducible fatigue test method for the femoral component of total hip prostheses

During the German HDR Safety Program the integrity of a reactor pressure vessel was studied intensively during a pressurized thermal shock loading. As a part of the work this paper presents a stable crack growth evaluation for a side-grooved, multimaterial CS-type specimen, which was detached from the inner surface of the HDR reactor pressure vessel. The specimen was composed of ferritic base material and two austenitic cladding material layers. The specimen was tested at the Fraunhofer-Institut für Werkstoffmechanik, Freiburg, Germany, and revealed an extremely nonuniform stable crack growth distribution along the crack front.Elastic-plastic finite element analyses were carried out using the 85-version of the ADINA code. For crack growth simulation a separate subroutine package IWM-CRACK linked with ADINA was used. Additional J-integral calculations were carried out based on the ADINA results, using the VTTVIRT code in order to study the path dependence of the J-integral.The crack growth was controlled by Jr curves, while actual measured data were available only for the base material, and hypothetical Jr curves were applied for the two cladding layers. The calculated maximum load was slightly higher than the maximum load measured in the test.The crack growth was relatively well simulated in quality. The slight numerical discrepancies at some locations along the crack front were consistent with extreme local material brittleness, which was clearly underestimated by the assumed material properties.