Research Progress on Irradiation Effects of RPV Materials

2013 ◽  
Author(s):  
Mingquan Feng ◽  
Huajun Mo ◽  
Guoyun Li ◽  
Mingyan Tong ◽  
Xiaosong Liu
Keyword(s):  
Fracture Toughness ◽  
Nuclear Power ◽  
Reactor Core ◽  
Mechanical Tests ◽  
Research Progress ◽  
Shelf Energy ◽  
Neutron Damage ◽  
Irradiation Test ◽  
Test Parameters ◽  
Reactor Pressure

Reactor pressure vessel (RPV) is an important safety component which holds reactor core and keep on high temperature and high pressure coolant. The irradiated region of RPV is subjected to neutron damage so that the fracture toughness decrease and RTNDT of materials with neutron fluence increase. According to fracture toughness requirements based on the rule, RPV beltline materials must have Charpy upper-shelf energy of no less than 102J initially and must maintain upper-shelf energy throughout the life of the vessel of no less than 68J. Chinese manufactures (China First Heavy Industries, China National Erzhong Group Co. and Dongfang Electric Corporation, Shanghai Electric Corporation) have supplied A508-3 steel forging and fabricated RPV for many constructing NPP in China. Nuclear Power Institute of China (NPIC) has being put into practice a program about the irradiation tests and irradiation brettling research of Chinese A508-3 steel from these manufactures. This paper introduces the program and progressing briefly. Specimen, irradiation test parameters, irradiation facility and post irradiation mechanical tests are also described.

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.

Fracture Toughness of Highly Irradiated Pressure Vessel Steels in the Upper Shelf Temperature

2006 ◽  
Author(s):  
Hiroshi Matsuzawa ◽  
Toru Osaki
Keyword(s):  
Fracture Toughness ◽  
Chemical Composition ◽  
Reference Material ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Charpy Impact ◽  
Pressure Vessels ◽  
Shelf Energy ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Nine Reactor Pressure Vessel (RPV) Steels and four RPV weld were irradiated up to 1.2 × 1024n/m2 fast neutron fluence (E>1MeV), and their fracture toughness and Charpy impact energy were measured. As chemical compositions, such as Cu, are known to affect the fracture toughness reduction due to neutron exposure, the above steels were fabricated by changing chemical composition widely to cover the chemical composition of the RPV materials of the operating Japanese nuclear power plants. 2.7 mm thick compact specimens were used to measure the upper shelf fracture toughness of highly irradiated materials, and their Charpy upper shelf energy was also measured. By correlating Charpy upper shelf energy to fracture toughness, the upper shelf fracture toughness evaluation formulae for highly irradiated reactor pressure vessel steels were developed. Both compact and V-notched Charpy impact specimens were irradiated in a test reactor. The fast neutron flux above 1MeV was about 5 × 1016n/(m2s). Charpy impact specimens made of Japanese PWR reference material containing 0.09w% Cu were irradiated simultaneously. The upper shelf energy of the reference material up to the medium fluence level showed little difference in the reduction of upper shelf energy to that which had been in the operating plant and which was irradiated to the same fluence. The developed correlation formulae have been adopted in the Japan Electric Association Code as new formulae to predict the fracture toughness in the upper shelf region of reactor pressure vessels. They will be applied to time limited ageing analysis of low upper shelf reactor pressure vessels in Japan, on a concrete technical basis in very high fluence regions.

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.

Analysis of Dynamic Stress in the Reactor Core Barrel due to the Pressure Pulsations

2015 ◽  
Author(s):  
Igor Orynyak ◽  
Iaroslav Dubyk ◽  
Anatolii Batura
Keyword(s):  
Nuclear Power ◽  
Dynamic Stress ◽  
Reactor Core ◽  
Pressure Pulsations ◽  
Power Station ◽  
Coolant Pump ◽  
The Core ◽  
Calculation Results ◽  
Full Power ◽  
Reactor Pressure

This article presents vibrations analysis of the reactor core barrel caused by pressure pulsations induced by the main coolant pump. For this purpose, the calculations of the pressure distribution in the annulus between the core barrel and the reactor pressure vessel, bounded above by a separating ring were performed. Using transfer matrix method is obtained the solution of two-dimensional problem of pressure pulsations in the annulus between reactor core barrel and reactor vessel. The calculation results are compared with the pulsation pressure measurements made at commissioning unit 2 of the South Ukraine Nuclear Power Station. The distribution of pressure over the height of core barrel was obtained, which makes possible to estimate its strength for variant deformation of the core barrel as a beam, and in the case of deformation of the core barrel as a shell. The calculation results are used to assess the reliability of core barrel pre-load, which clamps the core barrel flange in place at the top, at full power operating.

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.

Fracture Toughness of High-Performance Concrete on Three-Point Bending Notched Beams at Elevated Temperature

2010 ◽  
Vol 89-91 ◽  
pp. 159-164 ◽  
Author(s):  
Samira Djaknoun ◽  
Evariste Ouedraogo ◽  
Ali Ahmed Benyahia
Keyword(s):  
Fracture Toughness ◽  
Nuclear Power ◽  
High Performance ◽  
Applied Load ◽  
Thermal Loading ◽  
High Performance Concrete ◽  
Thermal Conditions ◽  
Mechanical Tests ◽  
Bending Test ◽  
Three Point Bending

High-performance concrete (HPC) are advanced materials used in advances applications such as tunnels or nuclear power plant in which they can be accidentally submitted to severe stress or thermal conditions. The present study deals with the material response to thermal loading conditions. The main objective of this research is the characterization of the fracture toughness under Mode I at high temperature of high performance mortars by using notched specimens in three-point bending test in accordance with the RILEM recommendations. The mechanical loading is applied to the specimens while heated at various temperatures ranging from 25 to 900°C in isothermal conditions. The maximum applied load is found to be maximum at 300°C temperature and then to decrease sharply at higher temperatures. Analysis of SEM micrographs undertaken on the heated specimens after mechanical tests helps in the understanding of the material macroscopic behaviour. The evaluation of the material toughness during the hot testing is undertaken through analytical approach based on Fracture Mechanics. Lastly, the stress intensity factor as well as the energy of fracture evolves similarly versus temperature as the maximum applied load.

Evaluation of Mechanical Properties of the Reactor Pressure Vessel Materials Changes by Small Punch Test Application

2013 ◽  
Author(s):  
Jana Petzová ◽  
Martin Březina ◽  
Ľudovít Kupča
Keyword(s):  
Mechanical Properties ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Mechanical Tests ◽  
Reactor Pressure Vessel ◽  
Small Punch Test ◽  
Small Punch ◽  
Punch Test ◽  
Reactor Pressure

The reactor pressure vessel (RPV) is the most important component of nuclear power plants. RPV steel near the reactor core is subject of irradiation degradation due to the fast neutron flux. Irradiation processes are rather complex but after all the damage of the steel crystal lattice lead to the changes of RPV mechanical properties as well as the shift of the transition temperature to higher values. Hence, monitoring of the RPV material irradiation changes must be proved during the all nuclear power plant (NPP) operation. The new surveillance specimen programs (SSP) at all Slovak NPPs reactors included, among the standard mechanical tests, also new types of evaluation mechanical properties due to method Small Punch Test (SPT).

Reconstitution of V-Notched Charpy Impact Specimen Made of Short Insert

2006 ◽  
Author(s):  
Toru Osaki ◽  
Hiroshi Matsuzawa
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Charpy Impact ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Absorption Energy ◽  
Impact Specimen ◽  
Insert Length ◽  
Reactor Pressure

Reconstitution in this paper means to constitute the original size V-notched Charpy impact specimen, which is made of the irradiated insert cut out from broken piece and un-irradiated tabs welded to the insert. It is a promising technique to secure an adequate number of surveillance specimens for long-term operation of nuclear power plants. Every Japanese nuclear power plant has its own surveillance test program, and is operated considering its unique surveillance test results along with the general reduction tendency of fracture toughness. This practice should be continued and enhanced if possible, after the full use of originally installed specimens, because its fracture toughness is lower than before. Reconstitution of V-notched Charpy impact specimens to the original shape by using a short insert was studied. Charpy absorption energy is generally shifted by reconstitution, if the insert length is as short as 10 mm. Reconstitution with a short insert is necessary when the transverse property of the original specimen is required although only the longitudinal surveillance specimen is installed as in some early constructed reactor pressure vessels in Japan. This case is important when the reactor pressure vessel is suspected to be a so-called low upper shelf toughness reactor pressure vessel. The minimum required insert length to avoid affect on the specimen properties depends on the Charpy absorption energy of the insert and reconstitution weld condition. Correlation between Charpy absorption energy and plastic deformation size, and short time annealing properties of irradiated pressure vessel steels were investigated. A method to evaluate the minimum required insert length was proposed, which depends on the expected Charpy absorption energy and thermal transient during reconstitution. It was demonstrated that the reconstituted specimens of 10 mm-long irradiated inserts, whose upper shelf absorption energy was 69J and required insert length was 9.5mm, showed little shift of upper shelf absorption energy.

scholarly journals Research Progress of Reactor Thermal-Hydraulic Characteristics Under Ocean Conditions in China

2020 ◽  
Vol 8 ◽  
Author(s):  
Xin Jie
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Flow Instability ◽  
Reactor Core ◽  
Research Progress ◽  
Future Research ◽  
Hydraulic Characteristics ◽  
Bubble Behavior ◽  
Ocean Conditions

Floating nuclear power plants are affected by sea wind and waves, which will produce various forms of movement, and cause changes in the thermal and hydraulic characteristics of the reactor core and threaten the safety of reactor operation. In response to the R&D and design requirements of floating nuclear power plants, the research progress on the thermal-hydraulic characteristics of reactors under ocean conditions in China are reviewed in this paper. The emphasis is put on flow heat transfer, bubble behavior, flow instability, and critical heat flux under ocean conditions. Research progresses as well as the issues that need to be focused on in the future research are discussed in detail.

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