M. H. C. Hannink’s research while affiliated with NRG, Nuclear Research & consultancy Group and other places

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.

Turbulent mixing of hot and cold flows can lead to thermal fatigue of piping systems. Especially in primary pipework of nuclear power plants this is an important, safety related issue. Because the frequencies of the involved temperature fluctuations are generally too high to be detected well by common plant instrumentation, accurate numerical simulations are indispensable for a proper fatigue assessment. In this paper, fatigue crack growth analyses are presented for turbulent mixing in a T-junction. The temperatures in the fluid and the pipe were calculated by large eddy simulations with conjugate heat transfer, and the resulting stresses in the pipe by the finite element method. The stresses in the pipe wall have subsequently been used as input for crack growth analyses, which were performed according to ASME Boiler and Pressure Vessel Code, Section XI (2017). The time to a through-wall crack has been calculated for different temperature differences between the hot and cold flows. The analyses presented in this paper have been performed in preparation for the analyses of the EU project ATLAS+. In that project similar analyses will be performed on a different configuration.

Fatigue is an important ageing effect to manage for long term operation (LTO) of nuclear power plants (NPPs). One of the steps in the process of LTO assessment is the revalidation of TLAAs (time limited ageing analyses). The safety margins of NPP Borssele with respect to fatigue were demonstrated by projecting the fatigue analyses to the end of the intended period of LTO. Besides this, it has to be ensured that the analyses remain valid during the entire period of LTO. Periodic verification of the load assumptions that are made in the analyses is therefore an important aspect of adequate fatigue management. In this paper, an integrated fatigue management approach is presented, coupling load monitoring, transient counting and fatigue assessment. The approach for periodic verification of the load assumptions in the fatigue analyses consists of two parts. First, it is verified whether the occurred numbers of cycles of the different load cases remain smaller than the numbers of cycles assumed in the fatigue analyses. Secondly, it is verified whether the thermal transients defined for the load cases in the fatigue analyses conservatively represent the occurred thermal transients. For the verification, the application LEAF (Load Evaluation Application for Fatigue) was developed. At NPP Borssele, thermal transients occurring during different load cases (e.g. start-up, shutdown, reactor trip) are registered by a temperature monitoring system. The load evaluation application processes the measurement data and verifies the conservatism of the load conditions in the fatigue analysis of the fatigue relevant components in the system. This paper explains the steps that are followed for the load evaluation and gives a demonstration of the results. The presented procedure is an essential part of adequate fatigue management during LTO.

Long Term Operation (LTO) of nuclear power plants (NPPs) requires an ageing management review and a revalidation of Time Limited Ageing Analyses (TLAAs) of structures and components important for nuclear safety. An important ageing effect to manage is fatigue. Generally, the basis for this is formed by the fatigue analyses of the safety relevant components. In this paper, the methodology for the revalidation of fatigue TLAAs is demonstrated for LTO of NPP Borssele in the Netherlands. The LTO demonstration starts with a scoping survey to determine the components and locations having relevant fatigue loadings. The scope was defined by assessment against international practice and guidelines and engineering judgment. Next, a methodical review was performed of all existing fatigue TLAAs. This also includes the latest international developments regarding environmental effects. In order to reduce conservatism, a comparison was made between the number of cycles in the analyses and the number of cycles projected to the end of the intended LTO period. The projected number of cycles is based on transient counting. The loading conditions used in the analyses were assessed by means of temperature measurements by the fatigue monitoring system (FAMOS). As a result of the review, further fatigue assessment or assessment of environmental effects was necessary for certain locations. New analyses were performed using state-of-the-art calculation and assessment methods. The methodology is demonstrated by means of an example of the surge line. The model includes the piping, as well as the nozzles on the pressurizer and the main coolant line. The thermal loadings for the fatigue analysis are based on temperature measurements. Fatigue management of the NPP is ensured by means of the fatigue concept where load monitoring, transient counting and fatigue assessment are coupled through an integrated approach during the entire period of LTO.

For many nuclear power plants worldwide the operation period will be extended from 40 to 60 years in the coming years. As the operation period increases the importance of knowledge of ageing mechanisms like fatigue increases. Knowledge of the influence of the environment is crucial, since environmental fatigue is a relatively new development which is a modification to the existing assessment method and has to be projected to 60 years as well. This paper is a follow up of the ASME PVP2013-97695 paper: overview of international implementation of environmental fatigue. A quantitative comparison of the resulting cumulative usage factors including environmental fatigue is made for the most commonly used and well defined methods. The comparison of the environmental fatigue codes is made on a spray nozzle of the pressurizer. This is a known fatigue relevant location with high stresses due to thermal loading. The high thermal loading is due to the spraying of relative cold water into the warm pressurizer. The comparison is made for 11 methods, sets of fatigue curves and environmental fatigue correction factors (Fen factor), and 4 types of material. The 4 materials are: low alloy, carbon, nickel alloy and austenitic stainless steel. The fatigue curves of ASME 2007, ASME 2010, KTA 1996, KTA 2013, NUREG/CR-6909 and Code Case N-792 are compared. The Fen factors are compared for the following methods: NUREG/CR-6583, NUREG/CR-5704, NUREG/CR-6909, Code Case N-792, JNES SS-0503, JNES SS-1005 and NUREG/CR-6909 rev1. Code Case N-761 is included for the final comparison of the cumulative usage factors including environmental fatigue. The differences in percentages are considerable between the different methods. For this specific case, the difference in cumulative usage factor including environmental fatigue for austenitic steels is 70 %. For nickel alloy materials the difference is 115%. For low alloy materials the difference is the highest: 267%. For carbon steels the difference in cumulative usage factor is 146%. The most conservative cumulative usage factors including environmental fatigue are ASME 2007 or KTA 1996 fatigue curves combined with the NUREG/CR-5704 (austenitic steel and nickel alloy) or NUREG/CR-6583 (low alloy and carbon steel). The next highest results are found by the Japanese methods (JNES-SS-0503 and JNES-SS-1005). The common factor for these methods, is the fatigue curve for austenitic steels as used before 2010. The lowest cumulative usage factors are obtained by implementing NUREG/CR-6909. Using the latest revision of NUREG/CR-6909 the cumulative usage factors increase slightly (about 7%). The paper shows the considerable differences of usage factors when different codes are applied to the same problem. Copyright © 2015 by ASME Country-Specific Mortality and Growth Failure in Infancy and Yound Children and Association With Material Stature Use interactive graphics and maps to view and sort country-specific infant and early dhildhood mortality and growth failure data and their association with maternal

Irradiation Assisted Stress Corrosion Cracking (IASCC) is considered a primary ageing issue for the Reactor Pressure Vessel (RPV) internals of Pressurized Water Reactors (PWR). In particular, this complex phenomenon which develops in an environment featuring thermal and mechanical stresses, interaction with corrosive compounds and irradiation, is affecting the bolts connecting the baffles and the formers in the Nuclear Power Plants’ (NPP) RPVs. The baffle-former assembly is the structure that borders the fuel assemblies region, contributing to keep them in position and separating in the radial direction, the core region from the downcomer region. An evaluation of the stresses and temperatures reached in the baffle-former bolts during normal operation was performed by means of a coupled thermo-mechanical study which uses reactor physics calculations to obtain the fluence in the reactor core and as a consequence the heat deposition in the RPV internals. The heat deposition data are coupled with a finite element model of the bolts and the RPV internals in order to perform a complete analysis taking in account thermal, mechanical and radiation loadings. The study is first carried out focusing on a section of the RPV internals, showing a single row of baffle-former bolts. Then the work is extended to the full core height. The model set up in this work, includes an in-depth study of the behavior of the core internals, in particular baffle-former bolts. The model has the capability of understanding the mechanical and thermal behavior of essential internal components in a PWR.

The MARIOS irradiation experiment is the latest in a series of experiments on americium transmutation, and has been carried out in the framework of EURATOM’s 7th Framework Programme (FP7) project FAIRFUELS, which started the 1st January 2009 and is still ongoing. The Post Irradiation Examination (PIE) of MARIOS samples will be performed under the PELGRIMM project. The transmutation of Minor Actinides (MA) is a fundamental step in order to be able to close the nuclear fuel cycle. One of the attractive possibilities to burn MA, is represented by the Minor Actinides Bearing Blanket (MABB) concept. In this option, MA are diluted in a UO2 matrix and irradiated for a long time (from 2000 to 4100 days) in radial blankets at the periphery of the outer core of a Sodium Fast Reactor (SFR). Past experimental activities in the field of transmutation and testing of innovative nuclear fuel containing Am has proved that the release or trapping of helium as well as the studies on the swelling of such kind of fuel is a key issue for its safety design. Therefore, the main objective of the MARIOS experiment is the study of the in-pile behaviour of UO2 (with natural uranium) containing Am as minor actinide in order to gain knowledge on the role of the microstructure and of the temperature on the gas release and on fuel swelling for the MABB concept.

For many nuclear power plants worldwide the operation period will be extended from 40 to 60 years in the coming years. As the operation period increases the importance of knowledge of ageing mechanisms like fatigue increases. Knowlegde of the influence of the environment is crucial, since environmental fatigue is a relatively new development which is a modification to the existing assessment method and has to be projected to 60 years as well. This paper contains the results of a literature survey of environmentally assisted fatigue in nuclear power plants. It describes the current status and developments in the world. The main regulatory rules, guidelines and methods from the US, Germany, Japan, Finland and France are presented. At this moment different approaches for incorporating the effect of the coolant water environment exist, although the general trend is towards a more uniform approach worldwide. The most common approach is the incorporation of an environmental fatigue correction factor (Fen) in the fatigue derivation of the cumulative usage factor. The Fen formulas and the S-N fatigue curves differ but the general equations are: Display Formula Fen = N air / N water and Display Formula CUF = Σ U partial * Fen partial Alternatives like using fatigue curves including the environmental effects, using threshold criteria and calculation of an allowable Fen based on testing, are described. Research and material tests are still on-going and subject of international development. An overview of the current international state-of-the-art is presented. Copyright © 2013 by ASME Country-Specific Mortality and Growth Failure in Infancy and Yound Children and Association With Material Stature Use interactive graphics and maps to view and sort country-specific infant and early dhildhood mortality and growth failure data and their association with maternal

The PYCASSO (PYrocarbon irradiation for Creep And Swelling/Shrinkage of Objects) irradiations were a part of RAPHAEL (ReActor for Process Heat And Electricity), a 4-year Integrated Project on Very High Temperature Reactors, which started in April 2005 as part of the European 6th Framework Program. In the Fuel Technology Sub-Project (SP-FT), the PYCASSO experiments were performed to determine the effect of high temperature (900–1100°C) irradiation on thermo-mechanical properties of various coating materials for TRISO coated particle fuel. Effects due to the presence of fuel, such as pressurization or chemical attack by fission products, have been excluded by irradiating surrogate (ZrO2 and Al2O3) kernels. The partners involved in this irradiation are CEA (France), JAEA (Japan) and KAERI (Republic of South Korea). The PYCASSO irradiations have taken place in the High Flux Reactor (HFR) Petten and were coordinated by NRG (The Netherlands). In this paper we will discuss the performance of PYCASSO-I by a direct comparison between the in-pile behavior of the experiment and the behavior of the corresponding finite element model. The experience of PYCASSO-I led to recommendations for the follow-up irradiation PYCASSO-II irradiation, of which the performance is also presented. The PYCASSO-I experiment has been successfully dismantled, which was a complex procedure of retrieving samples from the highly activated sample holders. The dismantling was successful, as demonstrated by relatively low sample activity, which will most likely allow exporting the samples from hot cells and enable PIE in a glovebox. Also the measured fluence is reported, which shows that the experiment fluence target has been perfectly met. Finally, the proposed PIE for the PYCASSO-I and PYCASSO-II samples in the ARCHER FP7 proposal is presented.

Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like 241Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. The MARIOS irradiation experiment is the latest of a series of experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS). MARIOS experiment is carried out in the framework of the 4-year project FAIRFUELS of the EURATOM 7th Framework Programme (FP7). During the past years of experimental work in the field of transmutation and tests of innovative nuclear fuel containing americium, the release or trapping of helium as well as swelling has shown to be the key issue for the design of such kinds of target. Therefore, the main objective of the MARIOS experiment is to study the in-pile behaviour of UO2 containing minor actinides (MAs) in order to gain knowledge on the role of the microstructure and of the temperature on the gas release and on fuel swelling.The MARIOS experiment will be conducted in the HFR (high flux reactor) in Petten (The Netherlands) and will start in the beginning of 2011. It has been planned that the experiment will last 11 cycles, corresponding to 11 months.This paper covers the description of the objective of the experiment, as well as a general description of the design of the experiment.