Master curve techniques to evaluate an irradiation embrittlement of nuclear reactor pressure vessels for a long-term operation

2008 ◽  
Vol 85 (9) ◽  
pp. 593-599 ◽  
Author(s):  
Bong-Sang Lee ◽  
Min-Chul Kim ◽  
Maan-Won Kim ◽  
Ji-Hyun Yoon ◽  
Jun-Hwa Hong
Keyword(s):  
Master Curve ◽  
Nuclear Reactor ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Term Operation ◽  
Reactor Pressure

Industry Practice for the Neutron Irradiation Embrittlement of Reactor Pressure Vessels in Japan

2010 ◽  
Vol 132 (10) ◽  
Author(s):  
Norimichi Yamashita ◽  
Masanobu Iwasaki ◽  
Koji Dozaki ◽  
Naoki Soneda
Keyword(s):  
Neutron Irradiation ◽  
Correlation Method ◽  
Pressure Vessels ◽  
Surveillance Data ◽  
Surveillance Program ◽  
Irradiation Embrittlement ◽  
Water Reactor ◽  
Term Operation ◽  
Reactor Pressure

Neutron irradiation embrittlement of reactor pressure vessel steels (RPVs) is one of the important material aging issues. In Japan, almost 40 years have past since the first plant started its commercial operation, and several plants are expected to become beyond 40 years old in the near future. Thus, the safe operation, based on the appropriate recognition of the neutron irradiation embrittlement, is inevitable to ensure the structural integrity of RPVs. The amount of the neutron irradiation embrittlement of RPV steels has been monitored and predicted by the complemental use of the surveillance program and embrittlement correlation method. Recent surveillance data suggest some discrepancies between the measurements and predictions of the embrittlement in some old boiling water reactor (BWR) RPV steels with high impurity content. Some discrepancies of pressurized water reactor (PWR) RPV surveillance data from the predictions have also been recognized in the embrittlement trend. Although such discrepancies are basically within a scatter band, the increasing necessity of the improvement of the predictive capability of the embrittlement correlation method has been emphasized to be prepared for the future long term operation. Regarding the surveillance program, on the other hand, only one original surveillance capsule, except for the reloaded capsules containing Charpy broken halves, is available in some BWR plants. This situation strongly pushed establishing a new code for a new surveillance program, where the use of the reloading and reconstitution of the tested specimens is specified. The Japan Electric Association Code, JEAC 4201-2007 “Method of Surveillance Tests for Structural Materials of Nuclear Reactors,” was revised in December 2007, in order to address these issues. A new mechanism-guided embrittlement correlation method was adopted. The surveillance program was modified for the long term operation of nuclear plants by introducing the “long term surveillance program,” which is to be applied for the operation beyond 40 years. The use of the reloading, reirradiation, and reconstitution of the tested Charpy/fracture toughness specimens is also specified in the new revision. This paper reports the application and practice of the JEAC4201-2007 in terms of the prediction of embrittlement and the use of tested surveillance specimens in Japan.

Long Term Operation and Surveillance Specimen Program of WWER RPVs

2010 ◽  
Author(s):  
Milan Brumovsky ◽  
Milos Kytka
Keyword(s):  
Czech Republic ◽  
Pressure Vessel ◽  
Laboratory Tests ◽  
Thermal Ageing ◽  
Reliable Information ◽  
Life Extension ◽  
Irradiation Embrittlement ◽  
Term Operation ◽  
Reactor Pressure

Long Term Operation (LTO) to 60 or 80 years of operation also requires a reliable information about the potential irradiation embrittlement (and also thermal ageing) of reactor pressure vessel materials. Such information is usually obtained from testing specimens within the surveillance specimen program that is designed for the design RPV life, regularly for 40 years only. Life extension requires modification of such program (if there is still time to perform it) or a design of a new – extended one. Such program should have to contain RPV archive materials that are not in every case available. Thus, combination of archive materials and possible surrogate materials must be taken into account for this program. Some complication can be expected with thermal ageing data as some laboratory tests at higher temperatures must be realized. The paper describes such program for NPP Dukovany, Czech Republic with WWER-440 type reactors that are now more than 20 years of operation.

Industry Practice for the Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels in Japan

2009 ◽  
Author(s):  
Norimichi Yamashita ◽  
Masanobu Iwasaki ◽  
Koji Dozaki ◽  
Naoki Soneda
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Correlation Method ◽  
Surveillance Data ◽  
Surveillance Program ◽  
Irradiation Embrittlement ◽  
Term Operation ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Neutron irradiation embrittlement of reactor pressure vessel steels (RPVs) is one of the important material ageing issues. In Japan, almost 40 years have past since the first plant started its commercial operation, and several plants are expected to become beyond 40 years old in the near future. Thus, the safe operation based on the appropriate recognition of the neutron irradiation embrittlement is inevitable to ensure the structural integrity of RPVs. The amount of the neutron irradiation embrittlement of RPV steels has been monitored and predicted by the complementally use of surveillance program and embrittlement correlation method. Recent surveillance data suggest some discrepancies between the measurements and predictions of the embrittlement in some old BWR RPV steels with high impurity content. Some discrepancies of PWR RPV surveillance data from the predictions have also been recognized in the embrittlement trend. Although such discrepancies are basically within a scatter band, the increasing necessity of the improvement of the predictive capability of the embrittlement correlation method has been emphasized to be prepared for the future long term operation. Regarding the surveillance program, on the other hand, only one original surveillance capsule, except for the reloaded capsules containing Charpy broken halves, is available in some BWR plants. This situation strongly pushed establishing a new code for a new surveillance program, where the use of the reloading and reconstitution of the tested specimens is specified. The Japan Electric Association Code, JEAC 4201–2007 “Method of Surveillance Tests for Structural Materials of Nuclear Reactors,” was revised in December, 2007, in order to address these issues. A new mechanism-guided embrittlement correlation method was adopted. The surveillance program was modified for the long term operation of nuclear plants by introducing the “long-term surveillance program”, which is to be applied for the operation beyond 40 years. The use of the reloading, re-irradiation and reconstitution of the tested Charpy/fracture toughness specimens is also specified in the new revision. This paper reports the application and practice of the JEAC4201–2007 in terms of the prediction of embrittlement and the use of tested surveillance specimens in Japan.

Assessment of RPV Surveillance Materials Program Data of a BWR

2016 ◽  
Author(s):  
Rogelio Hernández Callejas ◽  
A. Liliana Medina-Almazán ◽  
Fco. Javier Merino Caballero ◽  
Salvador Vázquez Belmont
Keyword(s):  
Energy Use ◽  
Nuclear Reactor ◽  
Safety Margin ◽  
Life Extension ◽  
Irradiation Embrittlement ◽  
Limiting Condition ◽  
Safety Operation ◽  
Program Data ◽  
Reactor Pressure

Irradiation embrittlement is a limiting condition for the long-term safety operation of a nuclear Reactor Pressure Vessel (RPV). When a Boiling Water Reactor (BWR) is approaching its initial licensing, in order to operate the reactor for another 20 years and more, it should be demonstrated that the irradiation embrittlement of the reactor vessel materials will be adequately managed by ensuring that the fracture toughness properties are above a certain level of the required safety margin. In this work the Charpy specimens recovered from two surveillance capsules of two BWRs (fluence 3.58×1017 – 9.03×1017 n/cm2) were impact tested at temperatures selected to establish the toughness transition and upper shelf of the irradiated RPV materials. The measured transition temperature shifts (ΔRTNDT) and the Upper Shelf Energy (USE) for the plate and weld materials were compared to the predictions calculated according to Regulatory Guide 1.99 Rev.2. The credibility of surveillance data were analyzed according with the five criteria established in the Regulatory Guide 1.99, Revision 2. The Master Curve (MC) approach and the instrumented impact tests using pre-cracked Charpy specimens were implemented in order to fully validate this techniques that can be used for embrittlement monitoring during life extension periods.

Heavy Section Steel Plates for Reactor Pressure Vessels Produced by Basic Oxygen Process

1977 ◽  
Vol 99 (2) ◽  
pp. 314-321
Author(s):  
H. Nakao ◽  
R. Yamaba ◽  
S. Takaishi ◽  
H. Kunitake ◽  
S. Kanazawa
Keyword(s):  
Electric Furnace ◽  
Nuclear Reactor ◽  
Temper Embrittlement ◽  
Pressure Vessels ◽  
Steel Plates ◽  
Irradiation Embrittlement ◽  
Basic Oxygen ◽  
Furnace Process ◽  
Oxygen Process ◽  
Reactor Pressure

Low alloy steel plates with heavy sections for pressure vessels of direct desulfurization unit and nuclear reactor were produced by basic oxygen process instead of the conventional electric furnace process. There was a decrease of impurity and residual elements which increase susceptibility to neutron irradiation embrittlement and temper embrittlement. Productivity also increased by this process. Various properties of the plates thus manufactured were evaluated in comparison with those of electric furnace plates. It was found as a result that the basic oxygen process produces an improved notch toughness for use in nuclear reactor pressure vessels, and approximately the same levels of properties for application to desulfurization-purpose pressure vessels, compared with the electric furnace process in which BOP return scrap was also used especially.

Advanced Structural Integrity Assessment Tools for Safe Long Term Operation (ATLAS+)

2018 ◽  
Author(s):  
Sebastian Lindqvist ◽  
Kim Wallin ◽  
Dominique Moinereau ◽  
Mike Smith ◽  
Stéphane Marie ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Structural Integrity ◽  
Assessment Tools ◽  
Experimental Proof ◽  
Term Operation ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Piping Systems ◽  
Reactor Pressure

The main objective and mission of the ATLAS+ project is to develop advanced structural assessment tools to address the remaining technology gaps for the safe and long term operation of nuclear reactor pressure coolant boundary systems. This is achieved by development and validation of: • innovative quantitative methodologies to transfer laboratory material properties to assess the structural integrity of large components, • enhanced treatment of weld residual stresses when subjected to long term operation, • advanced simulation tools based on fracture mechanics methods using physically based mechanistic models, • improved engineering methods to assess components under long term operation taking into account specific operational demands, • integrated probabilistic assessment methods to reveal uncertainties and justify safety margins. Additionally, the objective is to disseminate the findings of the work through special training sessions and links to the NUGENIA association. The project scope of work focuses on piping systems of the reactor coolant pressure boundary components (RCPB) excluding the reactor pressure vessel (RPV). The project is aimed on an experimental proof of concept and validates the developed methodology both at the laboratory scale and the full scale level. The ATLAS+ project contains 4 main technical work packages and one training and dissemination package. These are summarised here.

scholarly journals Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

2021 ◽  
Vol 13 (19) ◽  
pp. 10510
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Ana María Camacho ◽  
Carlos Mendoza ◽  
John Kickhofel ◽  
Guglielmo Lomonaco
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Prediction Models ◽  
Pressure Vessels ◽  
Defect Size ◽  
Historical Evolution ◽  
Nuclear Power Generation ◽  
Irradiation Embrittlement ◽  
Service Inspection ◽  
Reactor Pressure

The cataloguing and revision of reactor pressure vessels (RPV) manufacturing and in-service inspection codes and their standardized material specifications—as a technical heritage—are essential for understanding the historical evolution of criteria and for enabling the comparison of the various national regulations, integrating the most relevant results from the scientific research. The analysis of the development of documents including standardized requirements and the comparison of regulations is crucial to be able to implement learned lessons and comprehend the progress of increasingly stringent safety criteria, contributing to sustainable nuclear power generation in the future. A novel methodology is presented in this work where a thorough review of the regulations and technical codes for the manufacture and in-service inspection of RPVs, considering the implementation of scientific advances, is performed. In addition, an analysis focused on the differences between irradiation embrittlement prediction models and acceptance criteria for detected defects (both during manufacturing and in-service inspection) described by the different technical codes as required by different national regulations such as American, German, French or Russian is performed. The most stringent materials requirements for RPV manufacturing are provided by the American and German codes. The French code is the most stringent with respect to the reference defect size using as a criterion in the in-service inspection.

Surveillance Specimen Program for Lifetime Extension of RPV

2010 ◽  
Author(s):  
Milan Brumovsky ◽  
Milos Kytka
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Thermal Ageing ◽  
Reactor Pressure Vessel ◽  
Life Extension ◽  
Irradiation Embrittlement ◽  
Term Operation ◽  
Reactor Pressure ◽  
Lifetime Extension

Plant life extension (as well as Long Term Operation) to 60 or 80 years of operation also requires a reliable information about the potential irradiation embrittlement (and also thermal ageing) of reactor pressure vessel materials. Such information is usually obtained from testing specimens within the surveillance specimen program that is designed for the design reactor pressure vessel (RPV) life, regularly for 40 years only. Life extension requires modification of such program (if there is still time to perform it) or a design of a new – extended one. Such program should have to contain RPV archive materials that are not in every case available. Thus, combination of archive materials and possible surrogate materials must be taken into account for this program. Some complication can be expected with thermal ageing data as some laboratory tests at higher temperatures must be realized. The paper describes such program for Nuclear Power Plant (NPP) Dukovany, Czech Republic with WWER-440 type reactors.

Modeling Cascade Aging in Dilute Fe-Cu Alloys

2001 ◽  
Vol 701 ◽  
Author(s):  
B. D. Wirth ◽  
G. R. Odette
Keyword(s):  
Cross Sections ◽  
Production Cross ◽  
Three Dimensional ◽  
Continuous Distribution ◽  
Pressure Vessels ◽  
Rate Theory ◽  
Irradiation Embrittlement ◽  
Cu Alloys ◽  
Reactor Pressure

ABSTRACTThe continued safe operation of nuclear reactors and their potential for lifetime extension depends on ensuring reactor pressure vessel integrity. Reactor pressure vessels and structural materials used in nuclear energy applications are exposed to intense neutron fields that create highly non-equilibrium defect concentrations, consisting of a shell of self-interstitial atom and clusters surrounding a vacancy-rich core, over picosecond time scales. This spatially correlated defect production initiates a long chain of events responsible for microstructure evolution and hence irradiation embrittlement. In this paper, we describe the combined use of molecular dynamics (MD) and kinetic lattice Monte Carlo (KMC) to simulate the long-term rearrangement (aging) of displacement cascades in dilute Fe-Cu alloys. The simulations reveal the formation of a continuous distribution of three dimensional cascade vacancy-Cu cluster complexes and demonstrate the critical importance of spatial, as well as short and long-time correlated processes that mediate the effective production of primary defects. Finally, this approach can generate production cross-sections for vacancy-Cu clusters that can then be used in rate theory type models of long term global micro and microstructural evolution.

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