Evaluation of Nozzle P-T Limit Curves for Korea Optimized Power Reactor

2020 ◽  
Author(s):  
Hyunchul Lee ◽  
Choonsung Yoo ◽  
Youngjae Maeng ◽  
Sunghoon Yoo
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Power Reactor ◽  
Nuclear Safety ◽  
Reactor Vessel ◽  
Limit Curve ◽  
Reactor Coolant ◽  
Regulatory Issue ◽  
Coolant System ◽  
Reactor Pressure

Abstract The purpose of the Pressure – Temperature (P-T) limit curve is to prevent a failure of reactor pressure vessel during operation of reactor coolant system. In Korea, P-T limit curves have to meet 10CFR50 Appendix G [1] according to Nuclear Safety and Security Commission Notification 2017-20. The P-T limit curves have been traditionally evaluated based on the beltline region which is most affected by neutron irradiation. However due to the geometric discontinuity, the inside corner regions of the vessel nozzles are the most highly stressed regions of reactor vessel. These higher stresses can potentially result in more restrictive P-T limits, even if the reference nil ductility transition temperatures (RTNDT) for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries. In 2014, the NRC issued Regulatory Issue Summary (RIS) 2014-11 [2], which require the consideration of reactor vessel nozzles in P-T limits curve generation. In this paper, P-T limit curves for Korea optimized power reactor (OPR-1000) nozzles at the end of license were evaluated. And then these curves were compared to the traditional beltline region P-T limit curves in order to verify which curve is more limiting.

Effects of Neutron Irradiation on the Mechanical and Electromagnetic Properties of Reactor Pressure Vessel Steels

2017 ◽  
Author(s):  
Li Chengliang ◽  
Shu Guogang ◽  
Chen Jun ◽  
Liu Yi ◽  
Liu Wei ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Power Plants ◽  
Hysteresis Loops ◽  
Reactor Pressure Vessel ◽  
Electromagnetic Properties ◽  
Irradiation Damage ◽  
Indentation Hardness ◽  
Rpv Steel ◽  
Reactor Pressure

The effect of neutron irradiation damage of reactor pressure vessel (RPV) steels is a main failure mode. Accelerated neutron irradiation experiments at 292 °C were conducted on RPV steels, followed by testing of the mechanical, electrical and magnetic properties for both the unirradiated and irradiated steels in a hot laboratory. The results showed that a significant increase in the strength, an obvious decrease in toughness, a corresponding increase in resistivity, and the clockwise turn of the hysteresis loops, resulting in a slight decrease in saturation magnetization when the RPV steel irradiation damage reached 0.0409 dpa; at the same time, the variation rate of the resistivity between the irradiated and unirradiated RPV steels shows good agreement with the variation rates of the mechanical properties parameters, such as nano-indentation hardness, ultimate tensile strength, yield strength at 0.2% offset, upper shelf energy and reference nil ductility transition temperature. Thus, as a complement to destructive mechanical testing, the resistivity variation can be used as a potentially non-destructive evaluation technique for the monitoring of the RPV steel irradiation damage of operational nuclear power plants.

Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

2017 ◽  
Vol 488 ◽  
pp. 222-230 ◽  
Author(s):  
Kristina Lindgren ◽  
Magnus Boåsen ◽  
Krystyna Stiller ◽  
Pål Efsing ◽  
Mattias Thuvander
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Reactor Pressure Vessel Steel ◽  
Steel Welds ◽  
Reactor Pressure

Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

2014 ◽  
Vol 452 (1-3) ◽  
pp. 61-68 ◽  
Author(s):  
Tohru Tobita ◽  
Makoto Udagawa ◽  
Yasuhiro Chimi ◽  
Yutaka Nishiyama ◽  
Kunio Onizawa
Keyword(s):  
Mechanical Properties ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Weld Overlay ◽  
Reactor Pressure

Fast Neutron Irradiation Optimization of Thorium-Fueled SCWR Reactor Pressure Vessel

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Petra Pónya ◽  
Gyula Csom ◽  
Sándor Fehér
Keyword(s):  
Neutron Flux ◽  
Fast Neutron ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Flow Direction ◽  
Reactor Pressure Vessel ◽  
Fast Neutron Flux ◽  
Inner Surface ◽  
Fast Neutron Irradiation ◽  
Reactor Pressure

Abstract Fast neutron irradiation causes embrittlement of the reactor pressure vessel (RPV) material; therefore, it may end operation life before design lifetime. Well-known method to recuperate crystal lattice dislocations is annealing. In the current version of thorium fueled supercritical water-cooled reactor (SCWR) design proposed by the Institute of Nuclear Technology at Budapest University of Technology and Economics (BME NTI), the supercritical fluid flows upward between the core barrel and the inner surface of the RPV thereby, the coolant would keep the RPV's temperature at ∼500 °C. This reverse coolant flow direction would decrease the embrittlement of RPV by constant annealing. To minimize the fast neutron flux increase, a relatively thin shielding connected to the inner surface of the barrel could be used. This presents fast neutron irradiation analysis, performed for different settings of the shielding to reduce fast neutron flux reaching the inner surface of RPV.

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.

Inadvertent Blowdown Event in a Nuclear Boiling Water Reactor

2011 ◽  
Author(s):  
Robert Engel
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Service Level ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Point Of View ◽  
Primary Circuit ◽  
Event Scale ◽  
The Impact ◽  
Reactor Pressure

On March 6th 2007, the Leibstadt Nuclear Power Plant in Switzerland experienced an automatic blowdown of eight safety/relief valves installed on the main steam lines caused by a faulty electrical manipulation while performing planned maintenance during full power operation. Due to the temperature measurements inside the reactor recirculation system and the reactor pressure vessel this event, at a first glance, appeared to be Event No. 23 (Automatic Blowdown event) as an Emergency (Service Level C) Condition in accordance with the relevant reactor pressure vessel Thermal Cycle Diagram. According to the ASME Code Section III, Service Level C limits permit large deformations in areas of structural discontinuity which may necessitate the removal of a component from service for inspection or repair. This paper presents a summary of thermal-hydraulic, stress, fatigue, and fracture mechanical evaluations as well as plant inspections performed to demonstrate the impact of the event on the reactor pressure vessel and associated components and to fulfill the requirements of the Swiss Federal Nuclear Safety Inspectorate. It is shown that the primary circuit of the plant was not inadmissibly stressed by the event and that it was acceptable from a safety-related point of view to return the plant to service. Corresponding to the 7-level International Nuclear and Radiological Event Scale this event was rated afterwards as level 1 (anomaly) by the Swiss Federal Nuclear Safety Inspectorate.

Assessment of Structural Integrity on Irradiated Steel Structure: Focusing on Long Column Type Reactor Pressure Vessel Supports

2019 ◽  
Author(s):  
Goeun Han ◽  
Sukru Guzey
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Element Model ◽  
Reactor Pressure Vessel ◽  
Primary Coolant ◽  
Type Reactor ◽  
Column Type ◽  
Unfavorable Condition ◽  
Reactor Pressure

Abstract The structural steel in a nuclear facility experiences significant degradation due to the accumulated neutron irradiation. Particularly, the long-column type reactor pressure vessel supports have been focused since they resist considerable loading to maintain the primary coolant system in their position and experience high neutron irradiation in low temperature, which is an unfavorable condition for the fracture toughness. This study implemented the API 579-1/ASME FFS-1, fitness-for-service (FFS) method to consider both irradiated mechanical properties and multiple loading cases. A three-dimensional (3D) finite element model of long column type reactor pressure vessel support was built for the linear analysis. The metallurgical properties of reactor pressure vessel support for assessment were estimated by empirical equations. This study provides the structural margin of long-column type reactor pressure vessel support by levels of the loads and levels of the neutron fluence.

Susceptibility to Neutron Irradiation Embrittlement of Heat-Affected Zone of Reactor Pressure Vessel Steels

2019 ◽  
Author(s):  
Hisashi Takamizawa ◽  
Jinya Katsuyama ◽  
Yoosung Ha ◽  
Tohru Tobita ◽  
Yutaka Nishiyama ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Heat Affected Zone ◽  
Pressure Vessel ◽  
Microstructural Analysis ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Irradiation Embrittlement ◽  
Element Analysis ◽  
Thermal Histories ◽  
Reactor Pressure

Abstract The heat-affected zone (HAZ) of reactor pressure vessel (RPV) steels is known to show large scatter in Charpy impact properties because it has inhomogeneous microstructure due to thermal histories of multi-pass welding for butt-welded joints. The correlation between mechanical properties and microstructure such as grain size, phase-fraction, martensite-austenite constituent, on the characteristics of HAZ of un-irradiated materials was investigated. Neutron irradiation was conducted at Japanese Research Reactor −3 (JRR-3) operated by JAEA. The neutron irradiation susceptibility was evaluated based on post-irradiation examinations consisting of mechanical testing and microstructural analysis. In the experiments, typical RPV steel plate and their weldment were prepared. Simulated HAZ materials that have representative microstructures such as coarse-grain HAZ (CGHAZ) and fine-grain HAZ (FGHAZ) were also prepared based on the thermal histories calculated by finite element analysis. For un-irradiated materials, a part of simulated HAZ materials showed a higher reference temperature of the master curve method than that of the base metal (BM). The irradiation hardening of HAZ was almost the same or lower than that of the BM, and the shift of reference temperature for HAZ materials was comparable with that of BM.

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