MAI Benchmark Campaign of International Software for Reactor Pressure Vessel Integrity Assessment

Nuclear Reactor Pressure Vessel (RPV) integrity is a major issue concerning plant safety and this component is one of the few within a Pressurized Water Reactor (PWR) whose replacement is not considered as feasible. To ensure that adequate margins against failure are maintained throughout the vessel service life, research engineers have developed and applied computational tools to study and assess the probability of pressure vessel failure during operating and postulated loads. The Materials Ageing Institute (MAI) sponsored a benchmark study to compare the results from software developed in France, Japan and the United States to compute the probability of flaw initiation in reactor pressure vessels. This benchmark study was performed to assess the similarities and differences in the software and to identify the sources of any differences that were found. Participants in this work included researchers from EDF in France, CRIEPI in Japan and EPRI in the United States, with each organization using the probabilistic software tool that had been developed in their country. An incremental approach, beginning with deterministic comparisons and ending by assessing Conditional Probability of crack Initiation (CPI), provided confirmation of the good agreement between the results obtained from the software used in this benchmark study. This conclusion strengthens the confidence in these probabilistic fracture mechanics tools and improves understanding of the fundamental computational procedures and algorithms.

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Reactor Pressure Vessel Integrity in Light of the Evolution of Materials Science and Engineering

Materials Science Forum ◽

10.4028/www.scientific.net/msf.473-474.287 ◽

2005 ◽

Vol 473-474 ◽

pp. 287-292

Author(s):

Péter Trampus

Keyword(s):

Pressure Vessel ◽

Structural Integrity ◽

Materials Science ◽

Reactor Pressure Vessel ◽

Special Focus ◽

Science And Engineering ◽

Pressurized Water ◽

Integrity Assessment ◽

Materials Science And Engineering ◽

Reactor Pressure

Structural integrity of the reactor pressure vessel of pressurized water reactors is one of the key safety issues in nuclear power operation. Integrity may be jeopardized during operational transients. The problem is compounded by radiation damage of the vessel structural materials. Structural integrity assessment as an interdisciplinary field is primarily based on materials science and fracture mechanics. The paper gives an overview on the service induced damage processes and associated changes of mechanical properties, the prediction of degradation and the assessment of the entire component against brittle fracture with a special focus on how the evolution of materials science and engineering has contributed to reactor vessel structural integrity assessment.

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Ways for Extending the Reactor Pressure Vessel Lifetime

Volume 7: Operations, Applications and Components ◽

10.1115/pvp2008-61479 ◽

2008 ◽

Author(s):

Milan Brumovsky

Keyword(s):

Pressure Vessel ◽

Nuclear Power ◽

Pressure Vessels ◽

Reactor Pressure Vessel ◽

Integrity Assessment ◽

Low Leakage ◽

Vessel Integrity ◽

Direct Use ◽

Reactor Pressure ◽

Lifetime Extension

Reactor pressure vessels are components that usually determine the lifetime of the whole nuclear power plant and thus also its efficiency and economy. There are several ways how to ensure conditions for reactor pressure vessel lifetime extension, mainly: - pre-operational, like: • optimal design of the vessel; • proper choice of vessel materials and manufacturing technology; - operational, like: • application of low-leakage core; • increase of water temperature in ECCS; • insertion of dummy elements; • vessel annealing; • decrease of conservatism during reactor pressure vessel integrity assessment e.g. using direct use of fracture mechanics parameters, like “Master Curve” approach. All these ways are discussed in the paper and some qualitative as well as quantitative evaluation is given.

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Weld Material Investigations of a WWER-440 Reactor Pressure Vessel: Results From the First Trepan Taken From the Former Greifswald NPP

Journal of Engineering for Gas Turbines and Power ◽

10.1115/1.3032461 ◽

2009 ◽

Vol 131 (2) ◽

Cited By ~ 2

Author(s):

Udo Rindelhardt ◽

Hans-Werner Viehrig ◽

Joerg Konheiser ◽

Jan Schuhknecht

Keyword(s):

Fracture Toughness ◽

Pressure Vessel ◽

Pressure Vessels ◽

Reactor Pressure Vessel ◽

Reference Temperature ◽

Monte Carlo Code ◽

Data Set ◽

Integrity Assessment ◽

Inner Surface ◽

Reactor Pressure

Between 1973 and 1990 four units of the Russian nuclear power plants type WWER-440/230 were operated in Greifswald (former East Germany). Material probes from the pressure vessels were gained in the frame of the ongoing decommissioning procedure. The investigations of this material started with material from the circumferential core weld of unit 1. First, this paper presents results of the reactor pressure vessel (RPV) fluence calculations depending on different loading schemes and on the axial weld position based on the Monte Carlo code TRAMO. The results show that the use of the dummy assemblies reduces the flux by a factor of 2–5 depending on the azimuthal position. The circumferential core weld (SN0.1.4) received a fluence of 2.4×1019 neutrons/cm2 at the inner surface; it decreases to 0.8×1019 neutrons/cm2 at the outer surface. The material investigations were done using a trepan from the circumferential core weld. The reference temperature T0 was calculated with the measured fracture toughness values, KJc, at brittle failure of the specimen. The KJc values show a remarkable scatter. The highest T0 was about 50°C at a distance of 22 mm from the inner surface of the weld. The Charpy transition temperature TT41J estimated with results of subsized specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The VERLIFE procedure prepared for the integrity assessment of WWER RPV was applied on the measured results. The VERLIFE lower bound curve indexed with the Structural Integrity Assessment Procedures for European Industry (SINTAP) reference temperature, RTT0SINTAP, envelops the KJc values. Therefore for a conservative integrity assessment the fracture toughness curve indexed with a RT representing the brittle fraction of a data set of measured KJc values has to be applied.

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Evaluating the fracture toughness of reactor pressure vessel (RPV) materials subject to embrittlement**Some portions of this chapter have been gleaned from Chapter 3 of: International Atomic Energy Agency, Integrity of Reactor Pressure Vessels in Nuclear Power Plants: Assessment of Irradiation Embrittlement Effects in Reactor Pressure Vessel Steels, IAEA Nuclear Energy Series NP-T-3.11, IAEA, Vienna (2009), a chapter authored by the first author of this chapter (no attribution in the IAEA document)Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the US Department of Energy. The United States Government retains and the publisher by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

Irradiation Embrittlement of Reactor Pressure Vessels (RPVs) in Nuclear Power Plants ◽

10.1533/9780857096470.3.295 ◽

2015 ◽

pp. 295-332

Author(s):

R.K. Nanstad ◽

W.L. Server ◽

M.A. Sokolov ◽

M. Brumovský

Keyword(s):

United States ◽

Pressure Vessel ◽

Power Plants ◽

The United States ◽

Reactor Pressure Vessel ◽

Department Of Energy ◽

United States Government ◽

Irradiation Embrittlement ◽

States Government ◽

Reactor Pressure

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Supplemental Stress and Fracture Mechanics Analyses of Pressurized Water Reactor Pressure Vessel Nozzles

Volume 3: Design and Analysis ◽

10.1115/pvp2012-78119 ◽

2012 ◽

Author(s):

Matthew Walter ◽

Shengjun Yin ◽

Gary L. Stevens ◽

Daniel Sommerville ◽

Nathan Palm ◽

...

Keyword(s):

Fracture Mechanics ◽

Pressure Vessel ◽

Pressure Vessels ◽

Reactor Pressure Vessel ◽

Additional Stress ◽

Corner Crack ◽

Pressurized Water ◽

The Core ◽

Water Reactors ◽

Reactor Pressure

In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: • To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). • To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. • To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, “Fracture Toughness Requirements,” and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP) Conferences. This work is also relevant to the ongoing efforts of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, Working Group on Operating Plant Criteria (WGOPC) efforts to incorporate nozzle fracture mechanics solutions into a revision to ASME B&PV Code, Section XI, Nonmandatory Appendix G.

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The Fracture Toughness Properties of China Manufactured Reactor Pressure Vessel Steels in Transition Temperature Range

Volume 6A: Materials and Fabrication ◽

10.1115/pvp2018-84110 ◽

2018 ◽

Author(s):

Kai Sun ◽

Xiaoyong Wu ◽

Guoyun Li ◽

Bang Wen

Keyword(s):

Fracture Toughness ◽

Transition Temperature ◽

Pressure Vessel ◽

Power Plants ◽

Reactor Pressure Vessel ◽

Pressurized Water ◽

Integrity Assessment ◽

Temperature Range ◽

Pressure Vessel Steels ◽

Reactor Pressure

Over the past decade, many generation III pressurized water reactor power plants have been under construction in China. Most reactor pressure vessel steels for these plants construction are homemade. Historically, Charpy V-notch specimens are predominantly used to monitor the toughness of RPV steels. However, fracture toughness provides the quantitative predictions of the critical crack size and the allowable stress in structural integrity assessment. This paper evaluates the fracture toughness properties of China manufactured RPV steels directly measured in transition temperature range by using master curve method. Some specimens were irradiated in the High Flux Engineering Test Reactor. The influences of loading rate, test temperature, specimen configuration and neutron irradiation on T0 were also investigated. The experimental results show that China manufactured RPV steels exhibit good fracture toughness properties.

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Study on Structural Integrity Assessment of Reactor Pressure Vessel Based on Three-Dimensional Thermal-Hydraulics and Structural Analyses

Volume 1: Codes and Standards ◽

10.1115/pvp2014-28645 ◽

2014 ◽

Author(s):

Jinya Katsuyama ◽

Genshichiro Katsumata ◽

Kunio Onizawa ◽

Tadashi Watanabe ◽

Yutaka Nishiyama

Keyword(s):

Pressure Vessel ◽

Structural Integrity ◽

Three Dimensional ◽

Reactor Core ◽

Reactor Pressure Vessel ◽

Thermal Hydraulics ◽

Pressurized Water ◽

Integrity Assessment ◽

Coolant Water ◽

Reactor Pressure

In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during pressurized thermal shock (PTS) events, the thermal history of the coolant water and the heat transfer coefficient between the coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events and Jackson-Fewster correlation. Subsequently, using these values, structural integrity assessments of RPV are performed by structural analysis; e.g., loading that affects crack propagation is evaluated. Three-dimensional TH and structural analyses are recommended for precise assessments of the structural integrity of RPV. In this study, we performed TH and structural analyses simulating typical PTS events using three-dimensional models of cold-leg, downcomer and RPV to more accurately assess the structural integrity of RPV. From these analyses, we obtained loading histories from the reactor core region of RPV in which a crack is postulated in the structural integrity assessment. We discuss the conservativeness of current analysis methods on the structural integrity assessment of RPV through the comparison of loading conditions due to PTS events.

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French Four-Loop PWR 1300 MWe Reactor Pressure Vessel: First Results for Integrity Assessment and Life Manangement

Aging Management and Component Analysis ◽

10.1115/pvp2003-2152 ◽

2003 ◽

Author(s):

Georges Bezdikian ◽

Franc¸oise Ternon-Morin ◽

Henriette Churier-Bossennec ◽

Dominique Moinereau ◽

Alain Martin

Keyword(s):

Brittle Fracture ◽

Pressure Vessel ◽

Three Dimensional ◽

Pressure Vessels ◽

Reactor Pressure Vessel ◽

Surveillance Program ◽

Integrity Assessment ◽

First Results ◽

Safety Studies ◽

Reactor Pressure

The process used by the French utility, concerning the Reactor Pressure Vessel (RPV) integrity assessment, applied on 34 PWR NPPs 3-loop Reactors, involved the verification of the integrity of the component under the most severe conditions of situation, was engaged several years ago and the result obtained was the justification of the 900 MWe RPV life management for at least 40 years and to prepare the projection for beyond 40 years. Since 2000, in the continuity of this results, the studies was carried out on the 20 PWR NPPs 4-loop 1300 MWe Reactor Pressure Vessels, and the recent results obtained shows the demonstration of the integrity of the RPV, in the most severe conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and considering major parameters particularly the severity of the transient. This approach, is based on specific mechanical safety studies on the 1300 MWe RPV, to demonstrate the absence of risk of failure by brittle fracture. For these mechanical studies the major input data are necessary: 1 - the fluence distribution and the values of 4-loop RPV RTNDT during the lifetime in operation, 2 - the temperature distribution values in the downcomer and the PTS evaluation. The main results must show significant margins against initiation of the brittle fracture. The flaws considered in this approach are shallow flaws beneath the cladding (subclad flaws) or in the first layer of cladding. The major tasks and expertises engaged by EDF are: • better knowledge of the vessel material properties, including the effect of radiation, • more precise assessment of the fluence and neutronic calculations, • the NDE inspection program based on the inspection of the vessel wall, with a special NDE tool. The principal actions conducted during recent years are: • the optimisation of the fuel management and the new development for the fluence evaluation, • the data gathered from radiation specimen capsules, removed from the vessels (4loop reactor), within the framework of the radiation surveillance program, and the thermal-hydraulic-mechanical calculations based on finite element thermal-hydraulic computations and three dimensional elastic-plastic mechanical analyses.

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Comparison of PTS Guides for Reactor Pressure Vessel Integrity Assessment

Volume 3: Design and Analysis ◽

10.1115/pvp2013-97342 ◽

2013 ◽

Author(s):

Milan Brumovsky

Keyword(s):

Fracture Mechanics ◽

Pressure Vessel ◽

Detailed Comparison ◽

Pressure Vessels ◽

Reactor Pressure Vessel ◽

Integrity Assessment ◽

Asme Code ◽

Vessel Integrity ◽

Degradation Of Materials ◽

Reactor Pressure

Integrity of reactor pressure vessels (RPV) are of the most importance for safety of the whole NPP. From all potential regimes, Pressurized Thermal Shock (PTS) regimes during emergency cooling conditions are the most severe and most important. Several nuclear codes are based in similar approaches but their procedures differ and are based on national knowledge and approach to fracture mechanics as well as non-destructive methods of reactor pressure vessel testing. The paper will compare requirements and procedures for PTS evaluation in accordance with RCC-M code in France [2], KTA in Germany [3], Russian original code PNAEG from 1989 [5] and new procedure from 2004 for WWER vessels [4], as well as VERLIFE procedure and IAEA-NULIFE VERLIFE [6] procedure for WWER RPVs and finally ASME Code requirements [1] including US NRC RG approach. Detailed comparison of individual parameters in calculations are compared — material properties, degradation of materials, calculated defects size and form, fracture mechanics approach, warm pre-stressing possibility etc.

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