F. H. E. de Haan - de Wilde’s research while affiliated with NRG, Nuclear Research & consultancy Group and other places

Nuclear Power Plants (NPP) and research reactors built during the mid-20th century often have incomprehensive material characteristics for their concrete structures. This lack of quality records frequently leads to challenges when attempting to demonstrate safety during life extension projects or when specific modelling is necessary for plants SSCs when design regulations are updated with new or revised requirements. In the framework of long-term operation (LTO), there is limited knowledge about ageing and the structural integrity of concrete structures. In order to increase the knowledge in the field of civil structures, this paper focuses on investigating the ageing mechanisms of civil structures at NRG, Petten, and using the previously calibrated chloride ingress model (PVP2023-105650) to determine expected internal pressure generation due to reinforcement bar (rebar) corrosion within a FEA model to predict concrete cracking. In the most recent paper, PVP2023-105650, the modelling of ageing effects has been investigated by reducing material strength properties and developing a chloride ingress model calibrated to the HFR chimney. The model allows the predictions of rebar diameter reduction and internal pressure generation due to rust production. The FEA model developed under PVP2022-84008 was altered considering the reduced material properties and rebar diameter to gain a better insight into how it affected the determined concrete damage under a heavy drop impact. The purpose of this research paper is to better understand the effect of internal pressure generation due to rust production and to allow the simulation of crack generation due to this degradation mechanism. The model has been calibrated to the results obtained from the HFR chimney. It is seen through the FEA simulation generated in Ansys that there is a close correlation to reality when modelling the calculated internal pressures using the previously developed chloride ingress model. The model is limited in its scope to the HFR chimney where the concrete information was obtained. Furthermore, the model must be recalibrated for each time period as the contact stiffness varies over time. The modelling method is also sensitive to a reduction of mesh size which is contrary to the majority of FEA models. In the future, core samples of relevant HFR areas should be taken to determine the chloride concentration and rebar corrosion. This can allow for more data points for model calibration when simulating concrete cracking due to rebar corrosion.

This paper presents a leakage prediction study for Leak Before Break (LBB) assessments. The LBB concept has emerged as an essential tool for ensuring the safety of nuclear installations and pressurized piping systems. The approach is backed by accurate estimation of leakage rates from postulated cracks. In this study the methodology is based on the UK procedure for assessing structures containing defects (R6). The research is part of the ongoing in-house development of a software for LBB analyses and it aims to better understand and improve leak rate prediction methodologies for LBB assessments. The software is comprised of structural integrity modules and thermal hydraulic modules. This study presents the implementations of different coolant leak rates flow models, the Henry-Fauske model for flashing subcooled fluids and as single-phase Bernoulli model for lower temperatures. The thermal-hydraulics Henry Fauske model is evaluated against available experimental data and the differences are discussed. The Bernoulli based model is compared with another software. The implemented procedure is compared with two different software results for the structural modules. The structural calculations show coherence between the implemented procedure and the benchmark software that is based on the same methodology. In contrast the results are considerably different respect to another available software. The data show the significance and limitations of each model in various cases.

This paper presents the probabilistic study for Leak Before Break (LBB) assessments. The research has been carried out with an in-house developed software that allows to perform both probabilistic and deterministic leak before break analyses for pressurized pipes. The study is based on the UK procedure for the assessment of the integrity of structures containing defects (R6). The procedure applied is the Detectable Leak Before Break (DLBB) that relies on the Failure Assessment Diagram (FAD) Option 1 assessment procedure. The calculations merge the plastic collapse assessment and the brittle failure mode of the pipe in the FAD. The first computed parameter for the LBB assessment is the Critical Crack Length (CCL) distribution density of the postulated through wall defect. The structural integrity assessment is then coupled with the Henry-Fauske two-phase critical flow model for the evaluation of the Leakage Rate (LR). With this coupling the Minimum Detectable Crack Length (MDCL) distribution density is calculated. This step sets the boundary conditions for the fracture mechanics assessment of the postulated defect. The main probabilistic outputs from the assessment are the probability of the structural failure and the probability that the defect does not leak a detectable amount of coolant. The probabilistic method used is the Latin Hypercube Sampling; the Monte Carlo method is applied for verification. These results are then compared with the deterministic output from the LBB procedure.

Pressurized thermal shock (PTS) in a reactor pressure vessel could lead to a sudden brittle and catastrophic cleavage fracture. The combination of radiation embrittlement, the thermal stresses, and low temperatures can cause severe conditions for the structures. Particularly thick-walled reactor pressure vessels, which contain weak spots such as welds and cracks. In order to assess the probability for the initiation and propagation of a cleavage crack, a detailed image of the stress intensity and the temperature is needed. The critical stress intensity for brittle cleavage fracture depends on the ductile to brittle transition temperature. This complex combination of stresses, absolute temperatures and temperature gradients in combination with radiation damage requires an integral approach for the evaluation of the probability for the occurrence of cleavage fracture. In order to get the most accurate image as possible of this problem, in previous work by the authors, simulations were performed with a combined CFD and FEM approach. Where a CFD model simulates the thermal mixing of the fluid and its effects on the reactor pressure vessel wall. The temperature profile on the reactor pressure wall is then used as input of a static structural model using FEM. Over the last few years the complexity of the models increased and different types of transients were investigated [1, 2]. Reducing the amount of modelling simplifications and assumptions should lead to the most complete picture of the risks of the accident scenario. However in order to increase the speed of the calculations some simplifications are needed. In the coming years the simplifications will be added stepwise and their results will be checked against more complicated models. With a focus on verifying the stress intensity around a pre-existing crack, which leads directly to an increase on the probability of cleavage. In order to correctly predict thermal mixing in the fluid, a computationally expensive 3D model is needed. However the full temperature distribution in the reactor pressure vessel at all times is not necessarily needed to determine the stress intensity. A finite element analysis has been performed on a small section of the reactor pressure vessel [3], speeding up the simulations significantly. The largest amount of simulation time is spend on modelling the fluid using CFD. Other approaches to do this involve Thermal Hydraulic models, such as applied in the FAVOR code. The drawback of those codes is that they do not provide fluctuations which we can observe using CFD. A first approach of a comprehensive model which features the heat transition during a transient and the resulting stress intensity. A comparison with the computationally more expensive methods is made. A significant calculation time reduction can be achieved, but more work is needed in order to perform sufficient simulations to account for a full probabilistic analysis.

For many nuclear power plants worldwide the operation period will be extended to 60 or 80 years. As the operation period increases, the importance of knowledge of ageing mechanisms increases. In the framework of LTO there is limited knowledge about ageing and structural integrity of concrete structures. Recent developments have shown that ageing of civil structures receive more attention internationally. In order to increase the knowledge in the field of civil structures, this paper focusses on investigation of ageing of civil structures and determining an ageing management strategy. Knowledge of the ageing mechanisms of civil structures and especially concrete, will lead to improvement of ageing management and assessment methods of concrete. The presented work is part of ongoing research. In the past several steps have been made. Gathering of international information on civil structures ageing issues and management thereof (see PVP2019-93029 [3]), testing of a proposed assessment methodology by application to a nuclear reactor and comparison with another PWR with a steel containment (see PVP2020-21838 [4]) and creating a list of plant specific AMPs dealing with the relevant mechanisms at the various locations in a practical manner (see PVP2021-61499 [5]). The result was a general list with possible relevant locations and what type of measures could be taken. The current steps in the research focus on material behavior of structural concrete and practical assessment in finite element modelling techniques. In a this step, the concrete material and behavior will be investigated and explained. The characteristic properties of the concrete will be summarized and the degradation mechanisms will be identified. The assessment criteria for concrete finite element modelling are defined. The overall goal of the project is to obtain more knowledge on ageing management of civil structures and especially concrete. In the current research steps the aim is to create a predictive tool with the FE modelling technique for the damage quantification in concrete due to its ageing mechanism. The results of this ongoing work are presented in this report. The characteristic properties, degradation mechanisms including modelling options are identified together with an investigation on the assessment criteria for concrete finite element modelling. The behavior of structural concrete will be explained. The conclusion represents input parameters, modelling options and assessment criteria for finite element modelling of concrete. In the future steps a practical assessment in finite modelling techniques will be introduced in order to take into consideration the degradation of the concrete. The material properties will be analyzed and adapted in an FE Model.

For many nuclear power plants worldwide the operation period will be extended to 60 or 80 years in the coming years. As the operation period increases, the importance of knowledge of ageing mechanisms increases. In the framework of LTO there is limited knowledge about ageing and structural integrity of concrete structures. Recent developments have shown that ageing of civil structures receive more attention internationally. In order to increase the knowledge in the field of civil structures, this paper focusses on investigation of ageing of civil structures and determining an ageing management strategy. Knowledge of the ageing mechanisms of civil structures and especially concrete, will lead to improvement of ageing management and assessment methods of concrete. As a first step international information was gathered on civil structures ageing issues and management thereof (see PVP2019-93029). In addition a highlevel assessment methodology was proposed. In the next step the initially proposed assessment methodology has been tested by application to a nuclear reactor. The resulting list of relevant AMPs has been verified with the outcome for another PWR with a steel containment. With this experience the assessment methodology is tested, compared and improved (see PVP2020-21838). The results indicated that the method can be used to obtain a list of plant specific AMPs. What was added to the assessment method is the link to the TLAAs for civil structures. In this follow up step the transition is made from a high level of IGALL AMPs to a practical AMPs that will deal with the right mechanism at the right location. The detailing to a level of practical work instructions for the maintenance of the plant has to be made in order to make real life implementation possible. In this step studying of relevant degradation mechanisms, relevant AMPs (like IAEA AMP305 and AMP306 ) and applicable literature in combination with the practical knowledge from operation of a reactor, has taken place. The international developments on ageing management of concrete will be included. The goal of the project is to obtain more knowledge on ageing management of civil structures and especially concrete. It will lead to an assessment method for civil ageing management and ageing management programs dealing with the relevant mechanisms at the various locations in a practical manner. The results of this ongoing work are presented in this report. For the research reactor all SSCs in scope of the Continued Safe Operation could be linked to the relevant AMP(s) and a resulting set of plant specific AMPs for civil ageing management was obtained. Including the international developments, literature and guidelines, a more general applicable list was created (Table 5 through Table 13). The conclusion is that Figure 2 represents a practical method for obtaining a set of plant specific civil AMPs ready for implementation. For representation in this paper the final outcome is given in as a generic list of actions for a generic reactor (Table 5 through Table 13). In these tables the relevant SSCs, ageing mechanisms and actions are listed. The tables represent an generic list of actions for civil ageing management that might others help develop their ageing management program. Future steps are shifting the focus from the general but practical assessment methodology to finite element modelling techniques for concrete. The assessment criteria for concrete (e.g. in ASME III, ASME XIII or Eurocode) will be investigated and investigation on the modelling of the concrete for ageing are planned.

In 2020, the CFD analyses have been performed on a whole reactor vessel model. The action of accumulators during the PTS transient was considered. Using these data, an attempt to the efficient use of sub-modelling is carried out in order to obtain a strong reduction in the computational costs compared to a full 3D analysis. The analyses were given a probabilistic view by using the Master Curve approach for determining the material(s) fracture toughness. In parallel with these activities, a literature review work was carried out at NRG. The single temperature dependence of the Master Curve was incorporated into characterizations of fracture toughness for all RPV steels of interest such as SA 508 Grade 3 and Grade 4N for both forging and weld material. The literature review helps to prove that the Master Curve approach models the temperature dependence of fracture toughness for a generic pressure vessel before and after irradiation. This is because all of these steels have a BCC matrix phase lattice structure. Based on ASTM E1921, the types of microstructure falling under a BBC matrix, such as bainite, tempered bainite, tempered martensite, ferrite and pearlite, could also be evaluated using the Master Curve model. Furthermore, it is found that the chemical composition is one of parameter to look for as driving force in embrittlement RPV due to irradiation. For the very high nickel steels examined (SA508 Grade 4N), when not combined with copper and moderate manganese, irradiation is not a serious embrittling agent. This paper describes the work performed at NRG in the years 2017–2020 in investigting the Pressurized Thermal Shock (PTS) phenomenon, summarizes the achievements and gives a general judgement of the lessons learned. Moreover, this paper aims to illustrate the scope of planned research on PTS and its role in the new NRG research program PIONIER 2021–2024. An overview of NRG’s effort to align itself with the international community is given. Particular attention is given to the probabilistic problematic related to PTS. In order to better understand this problematic and improve the current state of knowledge NRG will create a PFM tool. The tool aims to use the best practices from existing PFM software to try to answer to questions requiring attention (e.g. thermal-hydraulic uncertainty). The experience accumulated during the previous activities will be included in the tool.

Fatigue is an important ageing mechanism for long-term operation (LTO) of nuclear power plants. The effect of the reactor coolant environment on fatigue is one of the factors to take into account. Application of environmental fatigue codes, generally, leads to large margins for actual fatigue failure of components. In combination with longer operation times, these margins can make it challenging to meet the criteria defined in the standards. The purpose of the work in this paper is, therefore, to gain insight into the conservatism in typical fatigue analyses using design fatigue curves which is done by analyzing the fatigue process up to the actual component failure using crack growth analyses. Prediction of fatigue failure in cylindrical specimen is investigated through numerical simulations of fracture mechanics. A finite element analysis model is implemented for a crack in a cylindrical specimen typically used for fatigue life testing and generating design fatigue curves. The specimen is uniaxially displacement-loaded into the plastic regime, and the reaction force is evaluated as a function of crack depth and shape. Cyclic loading leads to formation of a crack with the depth such that a failure point in the S-N curve corresponds to the 25% load drop. Stress intensity factors are calculated, and number of cycles to failure are determined based on Paris’ law and a two-stage crack growth relationship. Simulation results are compared to experimental fatigue life data and show good agreement. The outcome of the investigation can be extended to fatigue life of geometrically complex thermo-mechanical components, such as nozzles, under transient thermal loading occurring during operation of a nuclear power plant, in order to assess conservatism of fatigue failure criteria based on experimentally obtained S-N design curves.

The High Flux Reactor (HFR) is a multipurpose nuclear reactor located in Petten, the Netherlands. With its 45 MWth it is one of the most powerful and versatile research reactors in the world. Its main roles are material irradiation and medical isotopes production. The output of the reactor in terms of medical isotopes is important at a global level (60% of European demand). Every day in the Netherlands alone 30.000 patients are treated using isotopes produced in the HFR. The importance of the HFR dates back in time. The HFR has been in service since 1961. Due to the long life (58 years to date) of the reactor an efficient integrated ageing management program (AMP) is envisaged as it is foreseen that the HFR will continue to operate for a prolonged period of time. The development of the AMP has begun in 2018 (CSO project) and will be completed in view of the IAEA CSO mission. The HFR is the second reactor in the world to undergo this type of IAEA review and one of the objectives of this project is to set a state of the art when it comes to research reactors long term operation. The CSO project foresees four major sections: scoping and screening, development and improvement of plant programs, (re) validation of time limited ageing analyses (TLAAs) and realization of the ageing management program. In this paper the focus will lie on the TLAAs. The applicable TLAAs were scoped starting from the IGALLs TLAAs list. The TLAAs relevant for the HFR are: TLAA fatigue, TLAA reactor vessel, TLAA leak before break, TLAA manufacturing flaws TLAA beryllium and TLAA equipment qualification. The latter was developed in the framework of the equipment qualification plant program and does not figure as an independent TLAA in the CSO project. For each TLAA the principal problematics will be highlighted and the possible solutions illustrated.

For many nuclear power plants worldwide the operation period will be extended to 60 or 80 years in the coming years. As the operation period increases, the importance of knowledge of ageing mechanisms increases. In the framework of LTO there is limited knowledge about ageing and structural integrity of concrete structures. Recent developments have shown that ageing of civil structures receive more attention internationally (for example concrete degradation in bunker building Doel and buried piping attention in topical peer review EU). In order to increase the knowledge in the field of civil structures, this work focusses on investigation of ageing of civil structures and determining an ageing management strategy. Knowledge of the ageing mechanisms of civil structures and especially concrete, will lead to improvement of ageing management and assessment methods of concrete. As a first step international information was gathered on civil structures ageing issues and management thereof (see PVP2019-93029). In addition a very high level assessment methodology was proposed. The goal of the project is to obtain more knowledge on ageing management of civil structures and especially concrete. It will lead to an assessment method for civil ageing management and ageing management programs dealing with the relevant mechanisms at the various locations in a practical manner. The results of this ongoing work are presented in this report. The initial proposed assessment methodology has been tested by application to the HFR research reactor. The resulting list of relevant AMPs has been verified with the outcome for another PWR with a steel containment. With this experience the assessment methodology is improved. In addition each civil SSC in the scope of the Continued Safe Operation program is linked to the relevant AMP(s). The improved, but not finalized assessment method of ageing management for civil structures can be seen in figure 2. The proposed assessment method for ageing of civil structures has been tested, compared and improved. The results indicated that the method can be used to obtain a list of plant specific AMPs. The comparison of the list of relevant AMPs for a steel containment PWR, showed similar results. What is added to the assessment method is the link to the TLAAs for civil structures. The detailing to a level of practical work instructions for the maintenance of the plant has to be made in the near future. In the near future the step will be made from a high level of IGALL AMP to a practical AMP that will deal with the relevant mechanisms at the various locations. Therefore further steps are in studying of relevant degradation mechanisms, relevant AMPs (like AMP305 [9],AMP306 [10]) and applicable literature (e.g.[21]) in combination with the practical knowledge from operation of a reactor. The international developments on ageing management of concrete will be included. It is foreseen that the future report will contain information on concrete degradation mechanisms relevant for nuclear reactors. If findings requires calculations the assessment method will be verified with the finite element modelling techniques.