The method for calculating the probability of brittle destruction of NPP equipment in different operating modes with postulated defectiveness

2021 ◽  
Vol 14 (1) ◽  
pp. 34-39
Author(s):  
D. A. Kuzmin ◽  
A. Yu. Kuz’michevskiy
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Practical Significance ◽  
Reactor Unit ◽  
Operating Modes ◽  
Brittle Destruction ◽  
Reactor Pressure

The destruction of equipment metal by a brittle fracture mechanism is a probabilistic event at nuclear power plants (NPP). The calculation for resistance to brittle destruction is performed for NPP equipment exposed to neutron irradiation; for example, for a reactor plant such as a water-water energetic reactor (WWER), this is a reactor pressure vessel. The destruction of the reactor pressure vessel leads to a beyond design-basis accident, therefore, the determination of the probability of brittle destruction is an important task. The research method is probabilistic analysis of brittle destruction, which takes into account statistical data on residual defectiveness of equipment, experimental results of equipment fracture toughness and load for the main operating modes of NPP equipment. Residual defectiveness (a set of remaining defects in the equipment material that were not detected by non-destructive testing methods after manufacturing (operation), control and repair of the detected defects) is the most important characteristic of the equipment material that affects its strength and service life. A missed defect of a considerable size admitted into operation can reduce the bearing capacity and reduce the time of safe operation from the nominal design value down to zero; therefore, any forecast of the structure reliability without taking into account residual defectiveness will be incorrect. The application of the developed method is demonstrated on the example of an NPP reactor pressure vessel with a WWER-1000 reactor unit when using the maximum allowable operating loads, in the absence of load dispersion in different operating modes, and taking into account the actual values of the distributions of fracture toughness and residual defectiveness. The practical significance of the developed method lies in the possibility of obtaining values of the actual probability of destruction of NPP equipment in order to determine the reliability of equipment operation, as well as possible reliability margins for their subsequent optimization.

scholarly journals Occupational ALARA Planning for Reactor Pressure Vessel Dismantling at Kori Unit 1

2020 ◽  
Vol 17 (15) ◽  
pp. 5346
Author(s):  
Juyoul Kim ◽  
Batbuyan Tseren
Keyword(s):  
Radiation Exposure ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Occupational Safety ◽  
Health Management ◽  
Primary Component ◽  
Reactor Pressure Vessel ◽  
Safety And Health ◽  
Reactor Pressure

Assessing workers’ safety and health during the decommissioning of nuclear power plants (NPPs) is an important procedure in terms of occupational radiation exposure (ORE). Optimizing the radiation exposure through the “As Low As Reasonably Achievable (ALARA)” principle is a very important procedure in the phase of nuclear decommissioning. Using the VISIPLAN 3D ALARA planning tool, this study aimed at assessing the radiological doses to workers during the dismantling of the reactor pressure vessel (RPV) at Kori NPP unit 1. Fragmentation and segmentation cutting processes were applied to cut the primary component. Using a simulation function in VISIPLAN, the external exposure doses were calculated for each work operation. Fragmentation involved 18 operations, whereas segmentation comprised 32 operations for each fragment. Six operations were additionally performed for both hot and cold legs of the RPV. The operations were conducted based on the radioactive waste drum’s dimensions. The results in this study indicated that the collective doses decreased as the components were cut into smaller segments. The fragmentation process showed a relatively higher collective dose compared to the segmentation operation. The active part of the RPV significantly contributed to the exposure dose and thus the shielding of workers and reduced working hours need to be considered. It was found that 60Co contained in the stainless steel of the reactor vessel greatly contributed to the dose as an activation material. The sensitivity analysis, which was conducted for different cutting methods, showed that laser cutting took a much longer time than plasma cutting and contributed higher doses to the workers. This study will be helpful in carrying out the occupational safety and health management of decommissioning workers at Kori NPP unit 1 in the near future.

Upgrading the Operational Safety of Nuclear Power Plants Through the TACIS Nuclear Safety Assistance Programme

2004 ◽  
Author(s):  
M. Bie`th ◽  
R. Ahlstrand ◽  
C. Rieg ◽  
P. Trampus
Keyword(s):  
European Commission ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Nuclear Safeguards ◽  
Integrity Assessment ◽  
Reactor Pressure

The European Union’ TACIS programme was established for the New Independent States since 1991. One priority for TACIS funding is nuclear safety. The European Commission has made available a total of € 944 million for nuclear safety programmes covering the period 1991–2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. The Joint Research Center (JRC) of the European Commission is carrying out works in the following areas: • On-Site Assistance for TACIS Nuclear Power Plants; • Design Safety and Dissemination of TACIS results; • Reactor Pressure Vessel Embrittlement for VVER in Russia and Ukraine; • Regulatory Assistance; • Industrial Waste Management and Nuclear Safeguards. This paper gives an overview of the Scientific and Technical support that JRC is providing for the programming and the implementation of the TACIS nuclear safety programmes. In particular, two new projects are being implemented to get an extensive understanding of the VVER reactor pressure vessel embritttlement and integrity assessment.

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

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.

Master Curve Approach for Some Japanese Reactor Pressure Vessel Steels

2008 ◽  
Author(s):  
Minoru Tomimatsu ◽  
Takashi Hirano ◽  
Seiji Asada ◽  
Ryoichi Saeki ◽  
Naoki Miura ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steels ◽  
The World ◽  
Reactor Pressure ◽  
Plant Components

The Master Curve Approach for assessing fracture toughness of reactor pressure vessel (RPV) steels has been accepted throughout the world. The Master Curve Approach using fracture toughness data obtained from RPV steels in Japan has been investigated in order to incorporate this approach into the Japanese Electric Association (JEA) Code 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components”. This paper presents the applicability of the Master Curve Approach for Japanese RPV steels.

Evaluation of P-T Limit Curves According to Alternative Fracture Toughness Requirements

2014 ◽  
Vol 1051 ◽  
pp. 896-901
Author(s):  
Sin Ae Lee ◽  
Sung Jun Lee ◽  
Sang Hwan Lee ◽  
Yoon Suk Chang
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Ductile Failure ◽  
Evaluation Procedure ◽  
Limit Curve ◽  
Asme Code ◽  
Temperature And Pressure ◽  
Reactor Pressure

During the heat-up and cool-down processes of nuclear power plants, temperature and pressure histories are to be maintained below the P-T limit curve to prevent the non-ductile failure of the RPV(Reactor Pressure Vessel). The ASME Code Sec. XI, App. G describe the detailed procedure for generating the P-T limit curve. The evaluation procedure is containing the evaluation methods of RTNDT using 10CFR50.61. However, recently, Alternative fracture toughness requirements were released 10CFR50.61a. Therefore, in this study, RTNDT of RPV according to the 10CFR50.61a was calculated and used for evaluation of P-T limit curve of a typical RPV under cool-down condition. As a result, it was proven that the P-T curve obtained from 10CFR50.61 is conservative because RTNDT value obtained from the alternative fracture toughness requirements are significantly low.

Warm Pre-Stress Tests of the Irradiated RPV Base Material of Beznau Unit 1

2016 ◽  
Author(s):  
Florian Obermeier ◽  
Stefan Heußner ◽  
Heinz Hägeli ◽  
Herbert Schendzielorz ◽  
Marco Kaiser ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Base Material ◽  
Reactor Pressure Vessel ◽  
Stress Effect ◽  
Loading Conditions ◽  
Stress Tests ◽  
Reactor Pressure

According to the pertinent regulations, the integrity of a reactor pressure vessel (RPV) of a nuclear power plant is to be assessed by fracture mechanics for postulated flaws under most severe loading conditions. In such an analysis usually loss of coolant accidents are assumed to cause highest possible loading of the structural material of the RPV. This is due to the fact that such a pressurized thermal shock (PTS) event during which cold emergency coolant is injected into the primary system generates additional thermal stresses in the RPV wall. Based on the applicable regulation, the initiation of postulated flaws is to be excluded by the comparison of the calculated crack tip loading and the fracture toughness of the particular material. This kind of assessment was motivation of various research projects in the last decades addressing both evaluation approaches and experimental testing. A crucial result in this context is the existence of the so-called warm pre-stress effect (WPS) on the resulting fracture toughness. Generally, it is known as the increase of the apparent fracture toughness of a flaw in a specimen or structure after loading at high temperatures, generally in the upper shelf region, followed by a reloading at a lower temperature. This represents the typical loading scenario postulated for the assessment of a RPV during a PTS event. Experiments were performed to quantify this effect in the case of the irradiated reactor pressure vessel base material of the nuclear power plant Beznau unit 1. This paper presents the results of the Master Curve tests to determine the reference temperature (according to ASTM E 1921) and the design and testing of the warm pre-stress experiments using irradiated 10×10 mm reconstituted single edge notch bend (SE(B)) specimens. The design of these warm pre-stress tests was based on the loading transients for postulated surface and sub-surface flaws investigated within the scope of the assessment of the Beznau unit 1 reactor pressure vessel against brittle failure. Finite element simulations were performed to transfer the loading conditions at the crack tip of the RPV determined during the brittle fracture safety assessment onto the SE(B) specimen. The simulation results were used to control the loading conditions as a function of time and temperature during the experimental tests. The fracture toughness values of the warm pre-stress specimens were finally compared with the original fracture toughness values determined in the absence of a warm pre-stress effect to demonstrate the increase of the safety margin when the warm pre-stress effect is taken into account.

French PWR Nuclear Power Plants Steam Generators and Reactor Pressure Vessel Heads Aging Management Criteria and Replacement Strategy

2004 ◽  
Author(s):  
Georges Bezdikian
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Time ◽  
Reactor Pressure Vessel ◽  
Steam Generators ◽  
Life Management ◽  
Power Stations ◽  
Reactor Pressure

The approach used by the French utility, concerning the Aging Management system of the Steam Generators (SG) and Reactor Pressure Vessel Heads, applied on 58 PWR NPPs, involves the verification of the integrity of the component and the Life Management of each plant to guarantee in the first step the design life management and in the second step to prepare long term life time in operation, taking into account the degradation of Alloy 600 material and the replacement of these materials by components made with Alloy 690. The financial stakes associated with maintaining the lifetime of nuclear power stations are very high; thus, if their lifetime is shortened by about ten years, dismantling and renewal would be brought forward which would increase their costs by several tens of billions of Euros. The main objectives are: • to maintain current operating performances (safety, availability, costs, security, environment) in the long term, and possibly improve on some aspects; • wherever possible, to operate the units throughout their design lifetime, 40 years, and even more if possible. This paper shows the program to follow the aging evaluation with application of specific criteria for SG and for Vessel Heads, and the replacement of the Steam Generators and Vessel Heads at the best period. The strategy of Steam Generators Replacement are developed and Vessel Head program of monitoring and replacement are detailed.

The Fatigue and Aging Effects on the Materials of the Nuclear Power Plant Components and Systems

2010 ◽  
Author(s):  
Sajjad Akbar ◽  
M. Shahid Khalil
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Service Condition ◽  
Reactor Pressure Vessel ◽  
Aging Effects ◽  
Operational Life ◽  
The Future ◽  
Aging Phenomena ◽  
Reactor Pressure

The first generation of nuclear power plants (NPP) parts, components, equipment and systems have completed their predicted trouble free operational life. Now their materials are posing fatigue and aging effects for the future useful applications in the NPP. For the reliable future operations, this study is made for the management of the fatigue and aging effects on the components/systems. The neutron irradiation and normal aging phenomena will cause deterioration in the properties of materials by increasing brittleness-ductile transition temperature. This would affect the reactor pressure vessel as well. For the design purpose as well as estimating the permissible service life of a reactor pressure vessel, it is essential to have reliable data regarding this effect. Considerable efforts have been made and data search carried out on this work. The routine and schedule maintenance programs are the prerequisite means of managing fatigue and aging effects. Therefore, these are monitored through collecting data on plant service condition, inspection, surveillance, testing and by regularly monitoring such programs. This paper presents the future course of actions, which are necessary in the enhancement of the useful and safe NPP operations throughout the world.

On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld

2017 ◽  
Vol 23 (2) ◽  
pp. 376-384 ◽  
Author(s):  
Kristina Lindgren ◽  
Krystyna Stiller ◽  
Pål Efsing ◽  
Mattias Thuvander
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Number Density ◽  
Cluster Number ◽  
Vessel Steel ◽  
Time Operation ◽  
Reactor Pressure

AbstractRadiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×1023 n/m2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.

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