Coordinated Research Projects at IAEA Concerning RPV and Structural Integrity

In order to develop the integrity evaluation technology for aged major components of nuclear power plants, the Japan Power Engineering and Inspection Corporation (JAPEIC) has been carrying out the project named “Nuclear Power Plant Integrated Management Technology (PLIM)” entrusted by Japanese Ministry of Economy, Trade and Industry (METI) since 1996Fiscal Year (FY). One of the objectives of this project is to establish the method for integrity evaluation of aged domestic reactor pressure vessels (RPV) by developing the prediction equations of reduction of Charpy V-notch Upper Shelf Energy (USE) due to neutron irradiation embrittlement and the correlation equations between USE and fracture toughness. Because the tests are now in progress, this paper presents the following preliminary results as of the end of 2000FY, using irradiated Charpy and Compact Tension specimens of RPV materials. • Study on the effect of accelerated irradiation for test specimens. • Comparison of USE values between test results and U.S. surveillance data. • Comparison of USE values between test results and published predictions. • Correlation between USE value and fracture toughness.

Integrity and lifetime of reactor pressure vessels are practically determined by their behavior during “pressurized thermal shock” (PTS) emergency regimes as the most severe regimes during reactor operation. Assessment of these potential regimes is carried out mostly in deterministic way but used procedures are different in different countries. Proper and reliable evaluation of these PTS regimes depends on many parameters and approaches used during computations. During the period 2005 – 2008, the Coordinated Research Project 9 (CRP 9) “Review and Benchmark of Calculation Methods for Structural Integrity Assessment of RPVs During PTS” was organised by the IAEA. The overall objective of this Coordinated Research Project was to perform benchmark deterministic calculations of a typical pressurised thermal shock (PTS) regime and finally to recommend the best practice for PTS assessment. This paper describes main results and collected experience within this project that were bases for the preparation of the “Good Practice Handbook for Deterministic Evaluation of the Integrity of a Reactor Pressure Vessel during a Pressurised Thermal Shock” that will be issued as an IAEA TECDOC. Main parameters discussed in this handbook are: - selection of overcooling sequences; - thermal-hydraulics analyses; - temperature and stress field calculations; - crack tip loading incl.K estimations; - integrity assessment; - analyses of nozzles; - national practices; - results from sensitivity studies. Finally, recommendations for reliable and correct PTS evaluation are given.

Reactor Pressure Vessel (RPV) is one of the most important components in nuclear power plant (NPP). The aging mechanism of primary concern for RPV is irradiation embrittlement, which can result in a decrease of fracture toughness of RPV steel. Prediction of irradiation embrittlement for a certain Chinese domestic manufactured A508-3 steel is performed. The calculation results given by the US safety standards, the French RCC-M standards and other international safety standards are compared with each other and compared to the data from commercial operation NPP surveillance program. The effect of neutron fluence is also investigated. Furthermore, the property of the steel against irradiation embrittlement is evaluated with the regulatory requirements in the relevant standards. It can be predicted that the Chinese domestic steel satisfies the requirements in these standards.

Probabilistic fracture mechanics (PFM) has been recognized as a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency (JAEA) to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against non-ductile fracture. On the other hand, unlike deterministic analysis codes, the verification of PFM analysis codes is not easy. A series of activities has been performed to verify the applicability of PASCAL. In this study, as a part of the verification activities, a working group was established in Japan, with seven organizations from industry, universities and institutes voluntarily participating as members. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group including the verification plan, approaches and results.

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead a RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, 3D-CFD and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a more accurate spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of fracture probabilities on the position inside the RPV is obtained. Using the spatial distribution of fracture probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

The structural components that comprise nuclear reactors and their supporting structures are subjected to harsh operating environments that can challenge their integrity, especially after exposure for extended durations or under accident condition. As one of the most significant components of a Reactor, the Reactor Pressure Vessel (RPV) is exposed to an aggressive environment during the operation time (e.g. more than 40 years). Ageing degradation mechanisms (e.g. thermos-fatigue) could grow initial defects up to a critical size, increasing the susceptibility to failure in the RPV. The conventional methods are mostly based on simple crack and structure geometries. Very limited studies consider the real conditions of the RPV subjected to a thermal shock due to a Loss of Coolant Accident (LOCA). During a LOCA event, the most severe conditions take place when the emergency core cooling (ECC) water is injected inside the cold legs filled initially with hotter water and/or steam. The rapid cooling of the down-comer and the internal RPV surface followed probably by re-pressurization of the RPV causes large temperature gradients and variation of pressure which induces thermal-mechanical stresses. In order to develop the model for integrity assessment of a reactor pressure vessel (RPV) subjected to pressurized thermal shock (PTS), a multi-physics simulation, which includes the thermo-hydraulic, thermo-mechanical and fracture mechanics analyses is necessary. The multi-physics simulations are performed using Cardinal, a wrapping of the GPU-oriented spectral element Computational Fluid Dynamics (CFD) code NekRS and other multi-physics sub-modules within the MOOSE framework. Cardinal now fully supports MOOSE stochastic perturbations of NekRS models with varying boundary conditions, initial conditions, material properties, and any other quantity which is defined by a kernel (such as coefficients in a momentum source model). The implementation is designed in a flexible manner so that scalar values are sent from MOOSE into a user scratch space in NekRS, which can then be applied for any purpose within the NekRS case files (both on the host and device). When modeling PTS, several factors can impact the results significantly. In this report, the impacts of the geometry of the model, Reynolds number and buoyancy effect are investigated. Two geometry, i.e., a simplified model and a realistic RPV model, with both laminar and turbulent flow condition are adopted for the PTS simulation with and without buoyancy effect. The purpose of the investigation is to understand the impact of these factors on the prediction of temperature history of RPV. The accurate prediction on the temperature evolution, which will be exported to Grizzly code for further analyses on the progression of aging mechanisms and their effects on the integrity of RPV structures, is very crucial. Based on the understanding of these factors, a more sophisticated model is built to analysis the PTS under SBLOCA scenario. A literature survey is conducted to pick the SBLOCA scenario for the multi-physics simulation. The analysis helps to explain the form and the transformation of the cold plum when the ECC is activated under SBLOCA. This model can be can be applied to study the PTS effect for different RPV configurations. The results can help to assess structural component degradation for advanced reactors.

When the structural integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) events is assessed, an underclad crack is postulated at the inner surface of RPV and the stress intensity factor (SIF) corresponding to the driving force of non-ductile crack propagation, is evaluated for this crack. On the inner surface of RPV, cladding of the stainless steel is overlay-welded as a means for corrosion protection. Because the cladding is a ductile material, it is important to evaluate the SIF for postulated underclad crack considering the plasticity of cladding. A SIF evaluation method, which takes the effect of plasticity into account using a plastic correction method, has been established in France. In this study, we examined the SIF evaluation method established in France for underclad cracks during PTS transients. The elastic and elastic-plastic analyses based on the finite element method considering PTS events and inner pressure were performed using three-dimensional models including an underclad semi-elliptical crack with different geometry. We discussed the conservativeness of plastic correction method based on the analysis results.

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative. During the past decade, the NRC conducted the interdisciplinary Pressurized Thermal Shock (PTS) Re-evaluation Project that established a technical basis to support a risk-informed revision to current PTS regulations (10CFR Part 50.61). Once the results of the PTS reevaluation are incorporated into a revision of the 10 CFR 50 guidance on PTS, it is anticipated that the regulatory requirements for the fracture toughness of the RPV required to withstand a PTS event (accidental loading) will in some cases be less restrictive than the current requirements of Appendix G to 10 CFR Part 50, which apply to normal operating conditions. This logical inconsistency occurs because the new PTS guidelines will be based on realistic models and inputs whereas existing Appendix G requirements contain known and substantial conservatisms. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements of Appendix G to 10 CFR Part 50 in a manner that is consistent with that used to develop the risk-informed revision to the PTS regulations. Scoping probabilistic fracture mechanics (PFM) analyses have been performed for several hundred parameterized cool-down transients to (1) obtain insights regarding the interaction of operating temperature and pressure parameters on the conditional probability of crack initiation and vessel failure and (2) determine the limits on the permissible combinations of operating temperature and pressure within which the reactor may be brought into or out of an operational condition that remains below the acceptance criteria adopted for PTS of 1 × 10−6 failed RPVs per reactor operating year. This paper discusses the modeling assumptions, results, and implications of these scoping analyses.

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify ‘Incredibility of Failure’ and how these are incorporated in assessments of the integrity of existing plant and proposed ‘new build’ PWR pressure vessels.

The integrity of reactor pressure vessel (RPV) is greatly affected by pressurized thermal shock (PTS). Once crack appears in the nozzle region, the stress concentration around the crack tips may lead to crack propagation, and finally cause a serious security problem. When the transient temperature is above the nil-ductility reference temperature, elastic–plastic constitutive relations are considered in the fracture mechanics analysis. The temperature-related properties of the materials are introduced into a 3-D finite element model to establish the temperature field and stress field of a real RPV. Since the test and safety inspection for RPV with defects under PTS loads are quite difficult and dangerous, the process of the ductile crack propagation is simulated by the extended finite element method (XFEM), and the critical crack sizes for different base wall thicknesses are determined. Then, the quantitative analysis of the effect of the crack position on the ultimate bearing capacity is carried out. For the crack tips with different shapes, the crack propagation law and the shape effect on the ultimate bearing capacity of the whole structure are also analyzed. According to their crack propagation paths and damage degrees, a good agreement is obtained.

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

The Reactor Pressure Vessel (RPV) is an essential component liable to limit the lifetime of PWR plants. The assessment of defects in RPV subjected to pressurized thermal shock (PTS) transients made at a European level does not necessarily take into account the effect of load history (warm pre-stress WPS) on the resistance of RPV material regarding to the risk of brittle failure. Numerous experimental, analytical and numerical results are available, which confirm the beneficial effect of warm pre-stress on RPV steels with an effective and significant increase of the material resistance regarding to the risk of brittle failure. NESC VII, a new project dealing with WPS, launched in 2008 with the participation of numerous international organizations (involving R&D, Utilities and Manufacturers, Technical supports, Regulatory), is now finished. Based on experimental, analytical and numerical tasks, the project focused on topics generally not covered by past experience on WPS: biaxiality of loading on small and large-scale specimens, effect of irradiation, modeling (including analytical and numerical models) … Among these tasks, new original WPS experiments have been conducted on small and large scale specimens to study the influence of biaxial loading on WPS effect, using a fully representative RPV steel. All WPS experiments have been interpreted with a wide panel of methods, including analytical, engineering and numerical analyses. A description of NESC VII project is presented in this paper including a synthesis of the available experimental, analytical and numerical results.

The paper discusses materials and results of the most representative world forum on the problems of flow-accelerated metal corrosion in power engineering—Flow-Accelerated Corrosion (FAC) 2016, the international conference, which was held in Lille (France) from May 23 through May 27, 2016, sponsored by EdF-DTG with the support of the International Atomic Energy Agency (IAEA) and the World Association of Nuclear Operators (WANO). The information on major themes of reports and materials of the exhibition arranged within the framework of the congress is presented. The statistics on operation time and intensity of FAC wall thinning of NPP pipelines and equipment in the world is set out. The paper describes typical examples of flow-accelerated corrosion damage of condensate-feed and wet-steam pipeline components of nuclear and thermal power plants that caused forced shutdowns or accidents. The importance of research projects on the problem of flow-accelerated metal corrosion of nuclear power units coordinated by the IAEA with the participation of leading experts in this field from around the world is considered. The reports presented at the conference considered issues of implementation of an FAC mechanism in single- and two-phase flows, the impact of hydrodynamic and water-chemical factors, the chemical composition of the metal, and other parameters on the intensity and location of FAC wall thinning localized areas in pipeline components and power equipment. Features and patterns of local and general FAC leading to local metal thinning and contamination of the working environment with ferriferous compounds are considered. Main trends of modern practices preventing FAC wear of NPP pipelines and equipment are defined. An increasing role of computer codes for the assessment and prediction of FAC rate, as well as software systems of support of the NPP personnel for the inspection planning and prevention of FAC wall thinning of equipment operating in singleand two-phase flows, is accepted. Different lines of attack on the problem of FAC of pipelines and equipment components of existing and future nuclear power units are reviewed. Promising methods of nondestructive inspection of pipelines and equipment are presented.