Neutron Embrittlement Analysis of Qinshan 320MWe RPV on OLE

2017 ◽  
Author(s):  
Tao Hongxin ◽  
He Yinbiao ◽  
Cao Ming ◽  
Shen Rui
Keyword(s):  
Service Life ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Nuclear Safety ◽  
Radioactive Material ◽  
Pressurized Water Reactor ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Key Factor

One of the fundamental requirements on nuclear safety is to prevent the radioactive material from being released. Therefore, it is paramount to maintain the structural integrity of the pressure boundary of the reactor coolant system. The reactor pressure vessel (RPV), under high temperature, high pressure and high radiation in operation, is the most important as well as a Class I nuclear safety equipment. For a pressurized water reactor (PWR), the life of the RPV determines the service life of the entire nuclear power plant. The key factor controlling the life of a RPV is the accumulation of the neutron flux and which induces irradiation embrittlement degrading the anti-fracture capability of the RPV material. Several anti-fracture capability assessments carried out for the Qinshan 320MWe (QS1) RPV, such as (a) the structural integrity assessment against pressurized thermal shocks; (b) the fracture mechanics assessment under irradiation; (c) the P-T limit curves revised; (d) the evaluation of USE. They all demonstrated that the structural integrity of the QS1 RPV would be maintained for the extended service life.

Applicability of Fracture Toughness Curves Developed for Japanese Pressure Vessel Steels to Structural Integrity Evaluation

2015 ◽  
Author(s):  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Charpy Impact ◽  
Compact Tension ◽  
Transition Curve ◽  
Pressurized Water Reactor ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Reactor Pressure

Surveillance tests have been conducted on Japanese Pressurized Water Reactor (PWR) plants for more than 40 years to monitor irradiation embrittlement of reactor pressure vessel (RPV) beltline materials. Fracture toughness specimens are contained as well as tensile and Charpy impact specimens in a surveillance capsule and utilized for structural integrity evaluation. Therefore, a lot of fracture toughness data have been obtained by fracture toughness tests using such as Compact Tension (CT) and Wedge Opening Loading (WOL) specimens. More than one thousand data have been accumulated for both unirradiated and irradiated materials until 2013. Additionally, in terms of fracture toughness, Master Curve (MC) concept has been widely used for fracture toughness transition curve expression of ferritic steels. Considering such a situation, the new fracture toughness curves using Tr30, which denotes Charpy V-notch 30ft-lb transition temperature, as an indexing parameter were developed based on MC concept depending on product form for Japanese RPV steels in 2014. In this study, applicability of the newly developed curves of Japanese RPV steels to structural integrity evaluation is investigated. Especially, this paper focused on conservatism of the curves and the adequate margin to be added in evaluation of RPV integrity employing statistical methodology.

Fracture Assessment for Pressurized Water Reactor Vessel Nozzles Subjected to Pressure and Thermal Loading

2021 ◽  
Author(s):  
Qibao Chu ◽  
Qing Wang ◽  
Yonggang Fang ◽  
Wei Tan
Keyword(s):  
Crack Front ◽  
Thermal Loading ◽  
Structural Integrity ◽  
Safety Margin ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Main Challenge ◽  
Fracture Assessment ◽  
3D Finite Element Method

Abstract To ensure the structure integrity of the RPV, the main challenge is the embrittlement of beltline material. However, the stress of inlet or outlet nozzles of the RPV which are in general reinforced in comparison with the beltline, is more complex especially under the thermal loads. In recently studies, a lot of works have been done to show that the nozzle region may be more challenging under some conditions. In this paper, a fracture assessment for the RPV nozzles subjected to pressure and thermal loading is discussed using the software ABAQUS 6.12 and Zen Crack 7.9-3. It includes: SIF calculation based on 3D finite element method; structural integrity assessment under a typical LOCA transient; and the fatigue crack growth evaluation under cyclic loading situations. The results show that the SIF along the crack front is obviously asymmetric, and only to assess the safety of the deepest point along the crack front in the ASME and RCC-MR codes may be reconsider. If the KIa criteria is applied, under a typical LOCA transient, it is difficult to obtain an effective fracture safety margin for a 1/4 thickness crack, while based on the KIC criteria, the nozzle is shown to be safe in the case study. The shape of the surface elongated crack (which is often easily produced in the nozzle area) tends to be circle under the cyclic pressure loading situation which shows the crack shape assumed in the ASME and RCC-MR codes is reasonable.

Utilization of NEK full scope simulator for plant operation optimization, nuclear education and engineering in 20 years

2021 ◽  
Author(s):  
Matjaž Žvar ◽  
Tomaž Žagar
Keyword(s):  
Emergency Preparedness ◽  
Nuclear Power ◽  
Impact Analysis ◽  
Training Session ◽  
Nuclear Safety ◽  
Severe Accident ◽  
Electricity Production ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water

Abstract This paper gives an impact analysis of utilization of NPP full scope simulator on operation parameters, training and education in nuclear power plant Krško. The Slovenian Nuclear Safety Administration issued their simulator decree to NEK in April 1995. The first training session on the simulator was performed in April 17th 2000 and since then the simulator has been used on daily bases to improve operator knowledges, skills and performances. At the time, this was the first full scope simulator with the capability to simulate Beyond design basis accidents (severe accidents). The ability to simulate core meltdown and containment breach made it very suitable for emergency preparedness drills. After the 2017 simulator upgrade, fuel meltdown in the spent fuel pool can be simulated using the Modular Accident Analysis Program – MAAP5. This capability is still unique for full scope simulators even today. The simulator is also used for pre-testing of plant modifications before their implementation on site or for just-in-time training for infrequent performed evolutions or for procedure development and testing. The Pressurized Water Reactor Owners Group (PWROG) used the NEK simulator in 2018 to develop the new set of the Severe Accident Management Guidelines, incorporated with a completely new usage approach. In all of these years, the simulator has been actively participating in the increased reliability and stability of the electricity production and in achieving NEK's vision to be a worldwide leader in nuclear safety and excellence.

Homogenization of Cr-Ni Austenitic Steel Studying: Liquation and Microhardness Heterogeneity Equalization

2019 ◽  
Vol 822 ◽  
pp. 53-59 ◽  
Author(s):  
Anton Sergeevich Tsvetkov ◽  
Irina Vladimirovna Teplukhina
Keyword(s):  
Stainless Steel ◽  
Service Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Diffusion Experiment ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Exposure Effect ◽  
Alloy Type ◽  
Radiation Swelling

Currently, a new generation of pressurized water reactor for nuclear power plants with an extended service life (up to 60 years) and a guarantee of their complete safety are being designed in Russia. Analysis of the reactor internal elements performance showed, that designed service life cannot be guaranteed if the reactor’s internal parts would be made from currently used stainless steel (18-10 alloy type). Instead of the used steel, to ensure operability, new austenitic stainless steel (16-25 alloy type), with increased resistance to radiation swelling, is being developed for production of forged ring blanks for core baffle. The use of new steel requires revision of the existing metallurgical production technology stages. Therefore in this paper diffusion experiment was carried out to determine the required duration of homogenization. The results are presented in terms of different duration of the high-temperature exposure effect on the liquation heterogeneity equalization. The relation between duration of homogenization and microhardness is also shown.

Functional Information of System Components Influenced by Counteractions on Computer-Based Procedure

2018 ◽  
Vol 4 (4) ◽  
Author(s):  
Tulis Jojok Suryono ◽  
Akio Gofuku
Keyword(s):  
Nuclear Power ◽  
Radioactive Material ◽  
Functional Information ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Behavior ◽  
Computer Based ◽  
The Impact ◽  
Multilevel Flow Modeling ◽  
Emergency Operating

In an emergency condition of nuclear power plant, operators have to mitigate the accident in order to remove the decay heat and to prevent the release of radioactive material to the environment following the emergency operating procedures (EOPs). The action of operators on a component, for example, changing the parameter level of a component, which is described in a procedure step, will impact other components of the plant and the plant behavior. Nowadays, the advanced main control rooms have been equipped with the computer-based procedures (CBPs) which provide some features and benefits which are not available in paper-based procedures. However, most of CBPs do not provide information of the impact of the counteractions on each procedure step (components influenced and future plant behavior) although it is useful for operators to understand the purpose of the procedure steps before making decisions and taking the actions. This paper discusses the functional information and the method to generate the information using multilevel flow modeling (MFM) model of operator actions on some procedure steps of a simplified EOP of pressurized water reactor (PWR) plant, as an example.

Methodology and Warm Prestressing Effect for Pressurized Thermal Shock Analysis in Nuclear Power Plant

2015 ◽  
Vol 750 ◽  
pp. 104-113
Author(s):  
Gui An Qian ◽  
Markus Niffenegger
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Pressurized Water Reactor ◽  
Local Approach ◽  
Pressurized Water ◽  
Warm Prestressing ◽  
Load Paths ◽  
Pressurized Thermal Shock ◽  
Integrity Analysis

One potential challenge to the integrity of the reactor pressure vessel (RPV) in a pressurized water reactor is posed by pressurized thermal shock (PTS). Therefore, the safety of the RPV with regard to neutron embrittlement has to be analyzed. In this paper, the procedure and method for the structural integrity analysis of RPV subjected to PTS is presented. The FAVOR code is applied to calculate the probabilities for crack initiation and failure by considering crack distributions based on cracks observed in the Shoreham and PVRUF RPVs in the U.S. A local approach to fracture, i.e. the σ*-A* model is used to predict the warm prestressing (WPS) effect on the RPV integrity. The results show that the remaining stress contributes to the WPS effect, whereas the increase of fracture toughness is not completely attributed to the remaining stress. The modeled load paths predict a material toughness increase of 30-100%.

Structural Integrity for Nuclear Power Plant Against Excessive Load on Design Extension Condition

2013 ◽  
Author(s):  
Daigo Watanabe ◽  
Kiminobu Hojo
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Factors ◽  
Design Code ◽  
Acceptance Criteria ◽  
Integrity Assessment ◽  
Current Design ◽  
Factor Design ◽  
Excessive Load ◽  
Load And Resistance Factor

This paper introduces an example of structural integrity evaluation for Light Water Reactor (LWR) against excessive loads on the Design Extension Condition (DEC). In order to assess the design acceptance level of DEC, three acceptance criteria which are the stress basis limit of the current design code, the strain basis limit of the current design code and the strain basis limit by using Load and Resistance Factor Design (LRFD) method were applied. As a result the allowable stress was increased by changing the acceptance criteria from the stress basis limit to the strain basis limit. It is shown that the practical margin of the LWR’s components still keeps even on DEC by introducing an appropriate criterion for integrity assessment and safety factors.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

Structural Integrity Assessment of Steam Generator Tubes Using a New EPRI Statistical Approach

2002 ◽  
Author(s):  
Carlos Alexandre de Jesus Miranda ◽  
Miguel Mattar Neto
Keyword(s):  
Steam Generator ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Tube Wall ◽  
New Approach ◽  
Inconel 600 ◽  
Integrity Assessment ◽  
Tube Wall Thickness ◽  
Uniform Tube

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.

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