Do All RPV Head Penetration Leaks Have the Potential to Cause Head Wastage?

2004 ◽  
Author(s):  
Chandra M. Roy ◽  
John R. Fessler ◽  
Jude R. Foulds ◽  
Ronald M. Latanision ◽  
David E. Taylor
Keyword(s):  
Water Stress ◽  
Pressure Vessel ◽  
Recent Experience ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Head Penetration ◽  
Failure Risk ◽  
Primary Water ◽  
Reactor Pressure

The identification of the PWSCC (Primary Water Stress Corrosion Cracking) mechanism responsible for leakage from an Alloy 600 nozzle tube of a PWR RPV (pressurized water reactor reactor pressure vessel) head more than a decade ago led to a significant body of research into understanding the phenomenon and to development of bases for safely managing primary pressure boundary integrity. However, the relatively recent experience at Davis-Besse, wherein penetration leakage resulted in significant vessel head material wastage, led to the heretofore unconsidered issue of vessel failure risk due to head rupture. This paper addresses, in preliminary fashion, one key input to determining the risk associated with head material wastage and potential rupture — the local environmental and fluid conditions associated with a range of leak paths. The results indicate a need for rigorous prediction of fluid conditions for a range of leak situations to help establish criteria for addressing penetration leaks.

Application of Nickel Electroplating for PWR Reactor Pressure Vessel Cladding Repair

2020 ◽  
Author(s):  
Seong Sik Hwang ◽  
Dong Jin Kim
Keyword(s):  
Pressure Vessel ◽  
Thermal Effects ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Technical Standards ◽  
Pressurized Water ◽  
Cladding Layer ◽  
Nickel Electroplating ◽  
Primary Water ◽  
Reactor Pressure

Abstract SA508 low-alloy steel for a reactor vessel was exposed to primary water in a pressurized water reactor (PWR) plant because the cladding layer of type 309 stainless steel for a reactor pressure vessel (RPV) was peeled off owing to an accident in which the thermal sleeve was detached. The advantage of electrochemical deposition (ECD) Ni plating techniques is that the RPV can be repaired without significant thermal effects, and Ni has sound corrosion resistance that can withstand exposure in primary water. The corrosion rate of the damaged part was assessed, and its trend was analyzed. The essential variables of Ni plating for repair of the damaged part were derived. These conditions are applicable variables for the repair plating device and have been carefully adjusted using the repair plating device. The process for establishing ASME technical standards called code case N-840 is described. The process for developing Ni-plating devices and the electrodeposition procedure specifications are described as well.

Examinations of Oxidation and Sulfidation of Grain Boundaries in Alloy 600 Exposed to Simulated Pressurized Water Reactor Primary Water

2013 ◽  
Vol 19 (3) ◽  
pp. 676-687 ◽  
Author(s):  
D.K. Schreiber ◽  
M.J. Olszta ◽  
D.W. Saxey ◽  
K. Kruska ◽  
K.L. Moore ◽  
...  
Keyword(s):  
Electron Microscopy ◽  
Grain Boundary ◽  
Grain Boundaries ◽  
Atom Probe Tomography ◽  
Atom Probe ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Intergranular Attack ◽  
Pressurized Water ◽  
Primary Water

AbstractHigh-resolution characterizations of intergranular attack in alloy 600 (Ni-17Cr-9Fe) exposed to 325°C simulated pressurized water reactor primary water have been conducted using a combination of scanning electron microscopy, NanoSIMS, analytical transmission electron microscopy, and atom probe tomography. The intergranular attack exhibited a two-stage microstructure that consisted of continuous corrosion/oxidation to a depth of ~200 nm from the surface followed by discrete Cr-rich sulfides to a further depth of ~500 nm. The continuous oxidation region contained primarily nanocrystalline MO-structure oxide particles and ended at Ni-rich, Cr-depleted grain boundaries with spaced CrS precipitates. Three-dimensional characterization of the sulfidized region using site-specific atom probe tomography revealed extraordinary grain boundary composition changes, including total depletion of Cr across a several nm wide dealloyed zone as a result of grain boundary migration.

Advanced Inlay System for Inlet/Outlet Nozzles of RV for Preventive Maintenance Against Alloy 600 PWSCC in Japanese PWR Plants

2013 ◽  
Author(s):  
Kazuhide Yamamoto ◽  
Masahiko Kizawa ◽  
Hiroki Kawazoe ◽  
Yuki Kobayashi ◽  
Ken Onishi ◽  
...  
Keyword(s):  
Preventive Maintenance ◽  
Heat Treatments ◽  
Residual Tensile Stress ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Nuclear Plants ◽  
Alloy 690 ◽  
Primary Water

Because many nuclear plants have been in operation for ages, the importance of preventive maintenance technologies is getting higher. One conspicuous problem found in pressurized water reactor (PWR) plants is the primary water stress corrosion cracking (PWSCC) observed in Alloy 600 (a kind of high nickel based alloy) parts. Alloy 600 was used for butt welds between low alloy steel and stainless steel of nozzles of Reactor Vessel (RV), Steam Generator (SG), and Pressurizer (Pz). As PWSCC occurred at these parts may cause Loss of Coolant Accident (LOCA), preventive maintenance is necessary. PWSCC is considered to be caused by a mixture of three elements: high residual tensile stress on surface, material (Alloy 600) and environment. PWSCC can be prevented by improving one of the elements. MHI has been developing stress improvement methods, for example, Water Jet Peening (WJP), Shot Peening by Ultrasonic vibration (USP), and Laser Stress Improvement Process (L-SIP). According to the situation, appropriate method is applied for each part. WJP has been applied for RV nozzles of a lot of plants in Japan. However PWSCC was observed in RV nozzles during the inspection before WJP in recent years, MHI developed the Advanced INLAY system to improve the material from Alloy 600 to Alloy 690. Alloy 600 on the inner surface of the nozzles is removed and welding with Alloy 690 is performed. In addition, heat treatments for the nozzles are difficult for its structural situation, so ambient temperature temper bead welding technique for RV nozzles was developed to make the heat treatments unnecessary. This paper describes the specifications of the advanced INLAY system and introduces the maintenance activities which MHI has applied for three plants in Japan by March 2012.

Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant

2012 ◽  
Vol 9 (4) ◽  
pp. 104016 ◽  
Author(s):  
D. A. Thornton ◽  
D. A. Allen ◽  
A. P. Huggon ◽  
D. J. Picton ◽  
A. T. Robinson ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events

2014 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

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.

Sign in / Sign up

Export Citation Format

Share Document