The EPRI PWR Supplemental Surveillance Program (PSSP) Final Design and Implementation

2017 ◽  
Author(s):  
William Server ◽  
Brian Burgos ◽  
Tim Hardin ◽  
J. Brian Hall
Keyword(s):  
Temperature Shift ◽  
Electric Power Research Institute ◽  
Surveillance Program ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
High Fluences ◽  
Final Design ◽  
Commercial Reactor ◽  
Reactor Pressure ◽  
Selection Of

There is a lack of pressurized water reactor (PWR) surveillance program transition temperature shift and upper shelf toughness decrease data due to neutron irradiation exposure especially at high fluences indicative of 60 to 80 years of plant operation. The Electric Power Research Institute (EPRI) has funded the development of a supplemental reactor pressure vessel (RPV) surveillance program to allow testing of additionally irradiated specimens in two new capsules being installed in two different commercial reactor surveillance capsule positions. The previously irradiated materials were strategically selected and will be further irradiated to give final fluence levels equal to or above those for PWRs operating up to 80 years. This paper describes the final design of the capsules and selection of the key previously irradiated RPV materials reconstituted into new Charpy-size specimens being irradiated in the two PWR Supplemental Surveillance Program (PSSP) capsules.

Assessment of U.S. Embrittlement Trend Equations Considering the Latest Available Surveillance Data

2004 ◽  
Author(s):  
William L. Server ◽  
Randy G. Lott ◽  
Stan T. Rosinski
Keyword(s):  
Statistical Test ◽  
Surveillance Data ◽  
Surveillance Program ◽  
Correlation Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Correlation Models ◽  
Pressurized Thermal Shock ◽  
Surveillance Programs

The mechanistically-guided embrittlement correlation model adopted in ASTM E 900-02 was based on a database of U.S. surveillance results current through calendar year 1998. There exists now an extensive amount of new surveillance data that includes a large amount of boiling water reactor (BWR) results from an integrated, supplemental surveillance program designed to augment the plant-specific BWR surveillance programs. These recent data allow a statistical test of the ASTM E 900-02 embrittlement correlation, as well as the NRC correlation model currently being used in the pressurized thermal shock (PTS) re-evaluation effort and the older Regulatory Guide 1.99, Revision 2 correlation. Even though the ASTM E 900-02 embrittlement correlation is a simplified version of the NRC model, a comparison of the two embrittlement correlation models utilizing the new database has proven to be revealing. Based on the new BWR data, both models are inadequate in their ability to predict BWR results; this inadequacy has even more significance for extrapolation outside of the database for BWR heat-up and cool-down curves, as well as some pressurized water reactor (PWR) heat-up curves. Other aspects of the two models, as revealed from this preliminary look at the new data, are presented.

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.

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.

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.

Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel

2016 ◽  
Vol 138 (3) ◽  
Author(s):  
Kuan-Rong Huang ◽  
Chin-Cheng Huang ◽  
Hsiung-Wei Chou
Keyword(s):  
Structural Reliability ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Benchmark Problems ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Reactor Pressure

Cumulative radiation embrittlement is one of the main causes for the degradation of pressurized water reactor (PWR) reactor pressure vessels (RPVs) over their long-term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the U.S. is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Furthermore, two overcooling transients of steam generator tube rupture (SGTR) and pressurized thermal shock (PTS) addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR RPVs and can also be referred as its regulatory basis in Taiwan.

Simulation of PWR Crud

2014 ◽  
Author(s):  
W. A. Byers ◽  
G. Wang ◽  
M. Y. Young ◽  
J. Deshon
Keyword(s):  
Heat Transfer ◽  
Thermal Conductivity ◽  
Electric Power ◽  
Conductivity Measurement ◽  
Electric Power Research Institute ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fuel Rod ◽  
Radiation Fields ◽  
The Impact

Corrosion products deposit on Pressurized Water Reactor (PWR) fuel rod cladding surfaces and can create a number of issues including increased cladding temperature, elevated cladding corrosion, and the precipitation of boron species within the deposits. The deposits can also release and lead to increased radiation fields on system ex-core surfaces. These effects can vary widely from plant-to-plant. The amount of the deposits, commonly known as crud, is an important factor in determining the impact, but other parameters such as crud thickness, porosity, and composition are also thought to be important. The Electric Power Research Institute (EPRI) has sponsored a number of programs to better understand the characteristics of crud and its effects. Crud has been sampled by fuel scraping and by collecting suspended crud during operation and during fuel cleaning. The chemistry and structure of the crud was then characterized. These data were then used to create simulated crud in laboratory heated rod tests. These tests explored how the crud deposits affected heat transfer at the rod surface and the interaction between the crud and the simulated coolant. This paper discusses the nature of PWR crud and some of the practical aspects of crud simulation. Different approaches to laboratory crud creation will be reviewed, and the success in matching plant crud characteristics will be shown, with special emphasis on the production of crud for thermal conductivity measurement.

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