Fatigue of NPP Components Simulated by Non-Uniformly Strained Stainless Steel Specimens

2019 ◽  
Author(s):  
Jussi Solin ◽  
Jouni Alhainen ◽  
Esko Arilahti ◽  
Tommi Seppänen ◽  
Wolfgang Mayinger
Keyword(s):  
Strain Localization ◽  
Endurance Limit ◽  
Cyclic Softening ◽  
Normal Operation ◽  
Primary Circuit ◽  
Material Microstructure ◽  
Beneficial Effects ◽  
Normal Frequency ◽  
Testing Approach ◽  
Exact Quantification

Abstract Comprehensive experimental research on fatigue performance of niobium stabilized (type 347) steel has revealed beneficial effects of hot holds aimed to simulate normal operation of NPP between the fatigue relevant transients. Reduction of plastic strain, extension of life and increase of endurance limit has been demonstrated in strain controlled HCF tests. Our latest results indicate moderate, but still measurable ‘hold effects’ even without any stop of straining and loading, when blocks of low rate cycles are applied between normal frequency straining at constant 325°C. A new lab testing approach was developed to simulate the ‘component behavior’ in moderate strain concentrations within the NPP primary circuit. Strain concentrations in range of 1.5 ≤ Kε ≤ 2 are simulated through displacement controlled straining of standard and modified geometry LCF specimens. New results confirm the earlier results and introduce another consequence of holds. Cyclic softening promotes localization of strain, but hold hardening reverses this trend. The holds retard strain localization not only within the material microstructure, but also in geometric strain concentrations. We conclude that the geometric delocalization of strain can amplify beneficial hold effects for components. The local strains may reduce below the endurance limit resulting to run-out tests beyond millions of cycles, even though notable values of fatigue usage (CUF) had been accumulated during earlier phases of the tests. Applicability of the transferability factor introduced in 2013 to the German KTA standard No. 3201.2 is supported. Exact quantification of the factor is not easy, but in all considered cases Fhold ≤ 1. This means that the fatigue usages are overestimated without this factor.

Cyclic Mechanical Properties of 16MnR Welded Joint under Constant

2011 ◽  
Vol 197-198 ◽  
pp. 1658-1661
Author(s):  
Ying Xiong ◽  
Han Ying Zheng
Keyword(s):  
Mechanical Properties ◽  
Weld Metal ◽  
Endurance Limit ◽  
Welded Joint ◽  
Cyclic Softening ◽  
Fatigue Tests ◽  
Test Results ◽  
Control Test ◽  
Strain Control ◽  
Fatigue Endurance

Fatigue tests are carried out for 16MnR welded joint under constant strain control. Test results reveal that 16MnR weld metal exhibits characteristic of cyclic softening and non-masing obviously. The strain–life curve can be best described by the three-parameter equation. It shows the fatigue endurance limit in the heat-affecting zone (HAZ) of welded joint is lower than that in the weld metal.

scholarly journals A Safety Re-Evaluation of the AVR Pebble Bed Reactor Operation and Its Consequences for Future HTR Concepts

2008 ◽  
Author(s):  
Rainer Moormann
Keyword(s):  
Fission Product ◽  
End Of Life ◽  
Fission Products ◽  
Normal Operation ◽  
Primary Circuit ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Hot Gas ◽  
Design Basis ◽  
Gas Tight

The AVR pebble bed reactor (46 MWth) was operated 1967–1988 at coolant outlet temperatures up to 990°C. Also because of a lack of other experience the AVR operation is a basis for future HTRs. This paper deals with insufficiently published unresolved safety problems of AVR and of pebble bed HTRs. The AVR primary circuit is heavily contaminated with dust bound and mobile metallic fission products (Sr-90, Cs-137) which create problems in current dismantling. The evaluation of fission product deposition experiments indicates that the end of life contamination reached several percent of a single core inventory. A re-evaluation of the AVR contamination is performed in order to quantify consequences for future HTRs: The AVR contamination was mainly caused by inadmissible high core temperatures, and not — as presumed in the past — by inadequate fuel quality only. The high AVR core temperatures were detected not earlier than one year before final AVR shut-down, because a pebble bed core cannot be equipped with instruments. The maximum core temperatures were more than 200 K higher than precalculated. Further, azimuthal temperature differences at the active core margin were observed, as unpredictable hot gas currents with temperatures > 1100°C. Despite of remarkable effort these problems are not yet understood. Having the black box character of the AVR core in mind it remains uncertain whether convincing explanations can be found without major experimental R&D. After detection of the inadmissible core temperatures, the AVR hot gas temperatures were strongly reduced for safety reasons. Metallic fission products diffuse in fuel kernel, coatings and graphite and their break through takes place in long term normal operation, if fission product specific temperature limits are exceeded. This is an unresolved weak point of HTRs in contrast to other reactors and is particularly problematic in pebble bed systems with their large dust content. Another disadvantage, responsible for the pronounced AVR contamination, lies in the fact that activity released from fuel elements is distributed in HTRs all over the coolant circuit surfaces and on graphitic dust and accumulates there. Consequences of AVR experience on future reactors are discussed. As long as pebble bed intrinsic reasons for the high AVR temperatures cannot be excluded they have to be conservatively considered in operation and design basis accidents. For an HTR of 400 MWth, 900°C hot gas temperature, modern fuel and 32 fpy the contaminations are expected to approach at least the same order as in AVR end of life. This creates major problems in design basis accidents, for maintenance and dismantling. Application of German dose criteria on advanced pebble bed reactors leads to the conclusion that a pebble bed HTR needs a gas tight containment even if inadmissible high temperatures as observed in AVR are not considered. However, a gas tight containment does not diminish the consequences of the primary circuit contamination on maintenance and dismantling. Thus complementary measures are discussed. A reduction of demands on future reactors (hot gas temperatures, fuel burn-up) is one option; another one is an elaborate R&D program for solution of unresolved problems related to operation and design basis accidents. These problems are listed in the paper.

Numerical Approach for the Consideration of Hold Time Effects

2014 ◽  
Author(s):  
Sven H. Reese ◽  
Johannes Seichter ◽  
Dietmar Klucke ◽  
H. Ertugrul Karabaki ◽  
Wolfgang Mayinger
Keyword(s):  
Hold Time ◽  
Austenitic Steels ◽  
Normal Operation ◽  
Primary Circuit ◽  
Published Data ◽  
Fatigue Lifetime ◽  
Time Effects ◽  
Positive Effects ◽  
Surge Line ◽  
Circuit Components

In recent years the Environmentally Assisted Fatigue (EAF) became an item, which has to be considered additionally in terms of ensuring a conservative determination of the actual component’s health status resp. the CUF. For practical application, the consideration of the so called Fen-factor leads to the reduction of the admissible cycles in fatigue calculations. Beyond that the influence of elevated temperatures has been identified as one parameter having a negative influence on the admissible cycles as well. For example the German KTA 3201.2 defines for austenitic steels separate fatigue curves for temperatures above 80°C and for temperatures below 80°C. In summary on the one hand parameters influencing component’s lifetime negatively have to be considered in terms of conservative calculations. On the other hand, there are other parameters which influence the component’s fatigue lifetime in a positive manner. As such positive effects are neglected so far, CUF allowing for EAF tend to become over conservative leading to oversized components. Therefore, positive effects should be considered as well in the framework of a comprehensive and detailed analysis making sure not to overdesign components. When taking a closer look on the operational behavior of primary circuit components, fatigue loading is mainly defined by long steady-state periods with no significant changes in the loadings and by normally short outage periods with no thermal loading. For example fatigue of a PWR surge-line is mostly caused by short in-surge and out-surge events during start-up or shut-down of the plant. Normal operation transients mostly not cause fatigue relevant events in the surge-line. Fatigue of PWR spray-lines is primarily generated by very few spray-events during a one-year period of operation. Spray events are mainly caused by significant load ramps. Subsequently the fatigue status of primary circuit components is controlled by long periods with no fatigue relevant loading at operating temperature and few additional loading patterns in between. Experimental investigations have shown that hold time effects have a positive influence on fatigue lifetime of austenitic stainless steel materials. Anyhow, no quantification of these effects has been published in recent years. Within this publication an engineering based approach will be developed to quantify the hold time effect based on literature and published data. On the basis of a practical example the influence of hold time effects will be quantified and a direct comparison to lifetime reducing effect of EAF and temperature will be drawn.

Cyclic, Monotonic and Fatigue Performance of Stabilized Stainless Steel in PWR Water and Research Laboratory

2018 ◽  
Author(s):  
Jussi Solin ◽  
Jouni Alhainen ◽  
Tommi Seppänen ◽  
H. Ertugrul Karabaki ◽  
Wolfgang Mayinger
Keyword(s):  
Stainless Steel ◽  
Endurance Limit ◽  
Elevated Temperatures ◽  
Laboratory Data ◽  
Test Method ◽  
Fatigue Performance ◽  
Austenitic Steels ◽  
Cyclic Behavior ◽  
Normal Operation ◽  
Environment Temperature

Strain controlled LCF testing extended to 10 million cycles revealed an abrupt endurance limit enforced by secondary hardening. In elevated temperatures the ε-N curve is rotated and endurance limit is lowered, but not vanished. When very low strain rates are applied at 325°C in simulated PWR environment, fatigue life is reduced, but far less than predicted according to NUREG/CR-6909. It is possible, but not probable that the difference is due to different stainless grades studied. We assume that the test method plays a more important role. We have repeatedly demonstrated in different tests campaigns that interruptions of straining with holds aiming to simulate steady state normal operation between fatigue relevant cycles can notably extend the fatigue endurance. Further proof is again presented in this paper. The suspected explanation is prevention of strain localization within the material microstructure and also in geometric strain concentrations. This actually suggests, that hold effects should be even more pronounced in real components. Cyclic behavior of austenitic steels is very complex. Transferability of laboratory data to NPP operational conditions depends on test environment, temperature, strain rate and holds in many ways not considered in current fatigue assessment procedures. In addition to penalty factors, also bonus factors are needed to improve transferability. Furthermore, it seems that the load carrying capacity of fatigued stainless steel is not compromised before the crack growth phase. Tensile tests performed after fatigue tests interrupted shortly before end-of-life condition in 325°C (N ≈ 0.85 × N25) showed strength and ductility almost identical to virgin material. This paper provides new experimental results and discusses previous observations aiming to sum up a state of the art in fatigue performance of German NPP primary loop materials.

scholarly journals Chemical Forms of Important Fission Products in Primary Circuit of HTR-PM under Conditions of Normal Operation and Overpressure and Water Ingress Accidents: A Study with a Chemical Thermodynamics Approach

2019 ◽  
Vol 2019 ◽  
pp. 1-12
Author(s):  
Chuan Li ◽  
Wenqian Li ◽  
Lifeng Sun ◽  
Haoyu Xing ◽  
Chao Fang
Keyword(s):  
High Temperature ◽  
Fission Products ◽  
Point Of View ◽  
Normal Operation ◽  
Primary Circuit ◽  
Chemical Thermodynamics ◽  
Chemical Forms ◽  
Operational Conditions ◽  
Water Ingress ◽  
High Temperature Gas

The chemical forms of important fission products (FPs) in the primary circuit are essential to the source term analysis of high-temperature gas-cooled reactors because the volatility, transfer, and diffusion of these radionuclides are significantly influenced by their chemical forms. Through chemical reactions with gaseous impurities in the primary circuit, these FPs exist in diverse chemical forms, which vary under different operational conditions. In this paper, the chemical forms of cesium (Cs), strontium (Sr), silver (Ag), iodine (I), and tritium in the primary circuit of the Chinese pebble-bed modular high-temperature gas-cooled reactor (HTR-PM) under normal conditions and accident conditions (overpressure and water ingress accident) are studied with chemical thermodynamics. The results under normal conditions show that Cs exists mainly in the form of Cs2CO3 at 250°C and gaseous form at 750°C, and for I and Ag, Ag3I3 and Ag convert to gaseous CsI and AgO, respectively, with increasing temperature, while SrCO3 is the only main kind of compound for Sr. It is also observed that new compounds are generated under accidents: I exists in HI form when a water ingress accident occurs. Regarding tritium, the chemical forms of FPs change little, but compounds need higher temperature to convert. Furthermore, hazard of some FPs in different chemical forms is also discussed comprehensively in this paper. This study is significant for understanding the chemical reaction mechanisms of FPs in an HTR-PM, and furthermore it may provide a new point of view to analyze the interaction between FPs and structural materials in reactor as well as their hazards.

scholarly journals Influence of the dissolved hydrogen concentration on the radioactive contamination of the primary loops of DOEL-4 PWR using the OSCAR code

2020 ◽  
Vol 6 ◽  
pp. 7
Author(s):  
Mehdi Gherrab ◽  
Frédéric Dacquait ◽  
Dominique You ◽  
Etienne Tevissen ◽  
Raphaël Lecocq ◽  
...  
Keyword(s):  
Radioactive Contamination ◽  
Corrosion Products ◽  
Normal Operation ◽  
Primary Circuit ◽  
Contamination Assessment ◽  
Primary Coolant ◽  
Surface Activities ◽  
Primary Fluid ◽  
Dissolved Hydrogen ◽  
Chemical Conditions

Corrosion products are generated in the primary circuit during normal operation and are activated in the core. Those activated corrosion products, mainly 58Co and 60Co (coming respectively from the activation of 58Ni and 59Co), are then transported by the primary fluid and deposited on the out-of-flux surfaces (steam generators, primary coolant pipes…). To minimize this radioactive contamination, one needs to understand the behavior of corrosion products by carrying out measurements in PWRs and test loops combined with a reactor contamination assessment code named OSCAR. The aim of this article is to evaluate the influence of the change in the Dissolved Hydrogen (DH) concentration on the contamination of the primary loops of DOEL-4 PWR, a Belgian unit. After the description of the principle of the OSCAR V1.3 code, its use is illustrated with the simulation of DOEL-4. Finally, those calculations are compared to autoclave experiments called DUPLEX with thermodynamic and chemical conditions closed to those observed in PWRs. OSCAR V1.3 calculations show that an increase in the DH concentration results in a decrease in 58Co surface activities. These results are consistent with those from the DUPLEX experiments. Finally, an increase of the DH concentration is then recommended in operating PWRs to reduce the 58Co surface contamination.

Development of a PID-Fuzzy controller in the water level control of a pressurizer of a nuclear reactor

2021 ◽  
Vol 8 (3B) ◽  
Author(s):  
Thiago Souza Pereira de Brito ◽  
Carlos Alberto Brayner de Oliveira Lira ◽  
Wagner Eustáquio de Vasconcelos
Keyword(s):  
Fuzzy Logic ◽  
Pid Controller ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Fuzzy Controller ◽  
Normal Operation ◽  
Primary Circuit ◽  
Level Control ◽  
Pressurized Water ◽  
Safety Margins

It is well known that safety in the operation of nuclear power plants is a primary requirement because a failure of this system can result in serious problems to the environment. A nuclear reactor has several systems that help keep it in normal operation, within safety margins. Many of these systems operate in the control of variable quantities in the primary circuit of a reactor. However, nuclear reactors are nonlinear physical systems, and this introduces a complexity in the control strategies. Among several mechanisms in the thermal-hydraulic system of a reactor that actuate as a controller, the pressurizer is the component responsible for absorbing  pressure variations that occur in the primary circuit. This work aims at the development of a PID controller (Proportional Integral Derivative) based on fuzzy logic to operate in a pressurizer of a nuclear Pressurized Water Reactor. A Fuzzy Controller was developed using the process of fuzzification, inference, and defuzzification of the variables of interest to a pressurizer, then this controller was coupled to a PID Controller building a PID Controller, but oriented by Fuzzy logic. Subsequently, the PID-Fuzzy Controller was experimentally validated in a Simulation Plant in which transients like those in a PWR were conducted. The PID parameters were analyzed and adjusted for better responses and results. The results of the validation were also compared to simple controllers (on / off).

Analysis of Radioactive Source Term for Modular HTGR During Normal Operation

2010 ◽  
Author(s):  
Jianzhu Cao ◽  
Tao Liu ◽  
Yuanyu Wu ◽  
Hong Li ◽  
Yuanzhong Liu
Keyword(s):  
Source Term ◽  
Noble Gas ◽  
Fission Products ◽  
Radioactive Source ◽  
Normal Operation ◽  
Design Parameters ◽  
Primary Circuit ◽  
Secondary Circuit ◽  
Activation Products ◽  
Basic Prediction

The methods of radioactive source term analysis are introduced in detail for the modular high temperature gas cooled reactor in China. Radioactive fission products and activation products produced in the reactor are described. For fission products, the emphasis is on the process from production through release to the environment for noble gas, iodine and long-lived metallic nuclides. For activation products, it mainly introduces the behaviors of H-3 and C-14. Especially the permeation process from primary circuit to secondary circuit is described for H-3. Using the preliminary design parameters of demonstration HTGR in China, basic prediction of radioactive source term is done and the results are given.

Fully Reversed Uniaxial Tension-Compression High Cycle Fatigue Behaviour of Shot-Peened Steels

2011 ◽  
Vol 488-489 ◽  
pp. 21-24
Author(s):  
Alan Plumtree ◽  
M. M. Mirzazadeh
Keyword(s):  
Fatigue Limit ◽  
Shot Peening ◽  
Surface Hardness ◽  
Fatigue Behaviour ◽  
Cyclic Softening ◽  
Beneficial Effects ◽  
Heat Treated ◽  
Quenched And Tempered Steel ◽  
Hot Rolled ◽  
Compressive Residual Stresses

The effect of shot-peening on the uniaxial fatigue behaviour of four engineering steels, heat treated to a similar final hardness was investigated. Forged 0.39%C and 0.72%C steels, a quenched and tempered 0.51%C steel and a 0.50%C powder forged (PF) steel were fatigue tested under fully reversed (R=-1) push-pull loading conditions. Following long life (107) cycling, shot-peening had little effect on the fatigue limit of the 0.39%C and 0.72%C steels whereas the fatigue limit of the PF steel increased 10.4%. Conversely, the fatigue limit of the quenched and tempered steel decreased 12.0% after shot-peening. The results showed that the beneficial effects of shot-peening, such as compressive residual stresses and work hardening, balanced the effects of surface roughness since crack initiation tended to occur below the surface. Microhardness profiles showed that the greatest amount of cyclic softening in the shot-peened regions occurred in the hot rolled steels. Softening was accompanied by a decrease in the depth of surface hardness.

Development of Modeling Transport of Graphite Dust in HTGR During Normal Operation Condition

2012 ◽  
Author(s):  
Zhipeng Chen ◽  
Fei Xie ◽  
Yanhua Zheng ◽  
Lei Shi ◽  
Fu Li
Keyword(s):  
Fission Products ◽  
Life Time ◽  
Normal Operation ◽  
Primary Circuit ◽  
Potential Hazard ◽  
Operation Condition ◽  
Modeling Transport ◽  
Full Power ◽  
Accident Conditions ◽  
Helium Coolant

High temperature gas-cooled reactor (HTGR), especially the pebble-bed core type reactor, will inevitably cause the wear the graphite components and generate graphite dust in the core. The graphite dust is taken away by helium coolant and deposited on the surface of the primary circuit, and the fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. The objective of this paper is to develop a code for calculating the behaviour of graphite dust in the primary circuit of HTGR. The paper is focused on development of models for predicting the deposition rates of the dust. The purpose of the work is to estimate the amount and distribution of deposited dust during plant life time, which was assumed to be 40 full-power years. The result will lay the foundation for further studies of fission products releasing and interaction with dust under accident conditions.

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