Neutron Embrittlement Aging Management of Nuclear Reactor Pressure Vessels

2003 ◽  
Author(s):  
William L. Server
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Temperature Shift ◽  
Environmental Parameters ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Vessel Material ◽  
Reactor Pressure

The management of neutron embrittlement of nuclear reactor pressure vessels involves monitoring of the changes in the fracture toughness of surveillance capsule specimens that closely approximate the actual reactor vessel material(s). The measurement of fracture toughness is currently performed in an indirect manner using Charpy V-notch impact specimens, although the direct measurement of fracture toughness is possible using the same small Charpy specimens fatigue precracked to produce acceptable fracture toughness three-point bend specimens. This paper first examines the current Charpy-based approach and the development of a recent embrittlement correlation that has been incorporated into ASTM E 900-02, “Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials.” This correlation provides the latest mechanistically-guided approach to assess the changes in transition temperature shift. This same correlation and mechanistic guidance can be used with measured fracture toughness data developed following ASTM E 1921-02 to account for differences in surveillance material versus actual vessel material. Additionally, environmental parameters such as fluence and temperature also can be adjusted between different irradiation facilities using this latest correlation. This paper focuses on the application of the new ASTM E 900-02 correlation to Charpy-based and fracture toughness-based measurements to develop the best predictive approach for assuring structural integrity of reactor vessel materials. Key technical issues important for extended vessel life also are discussed.

Comparative analysis and verification of engineering methods of shallow cracks effect for fracture toughness prediction for reactor pressure vessels

2020 ◽  
pp. 140-165
Author(s):  
V. I. Kostylev ◽  
B. Z. Margolin
Keyword(s):  
Experimental Data ◽  
Fracture Toughness ◽  
Comparative Analysis ◽  
Nuclear Reactor ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Local Methods ◽  
European Method ◽  
Shallow Crack ◽  
Reactor Pressure

The main features of shallow cracks fracture are considered, and a brief analysis of methods allowing to predict the temperature dependence of the fracture toughness KJC (T) for specimens with shallow cracks is given. These methods include DA-method, (JQ)-method, (J-T)-method, “local methods” with its multiparameter probabilistic approach, GP method uses power approach, and also two engineering methods – RMSC (Russian Method for Shallow Crack) and EMSC (European Method for Shallow Crack). On the basis of 13 sets of experimental data for national and foreign steels, a detailed verification and comparative analysis of these two engineering methods were carried out on the materials of the VVER and PWR nuclear reactor vessels considering the effect of shallow cracks.

Development of Probabilistic Evaluation Models of Fracture Toughness KIc and KIa for Japanese RPV Steels

2015 ◽  
Author(s):  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
Kazuya Osakabe ◽  
Kentaro Yoshimoto
Keyword(s):  
Fracture Toughness ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Confidence Limits ◽  
Lognormal Distributions ◽  
Integrity Assessment ◽  
Probabilistic Evaluation ◽  
Evaluation Models ◽  
Reactor Pressure

Probabilistic fracture mechanics (PFM) analysis code PASCAL3 has been developed to apply the PFM analysis to the structural integrity assessment of domestic reactor pressure vessels (RPVs). In this paper, probabilistic evaluation models of fracture toughness KIc and KIa which have the largest scatter among the associated factors based on the database of Japanese RPV steels are presented. We developed probabilistic evaluation models for KIc and KIa based on the Weibull and lognormal distributions, respectively. The models are compared with the existing lower bound of fracture toughness in the Japanese code and probabilistic model in USA. As the results, the 5% confidence limits of the models established in present work corresponded to lower bounds of fracture toughness in the Japanese code. The comparison in the models between present work and USA showed significant differences that may have an influence on fracture probability of RPV.

Material Properties of Reactor Pressure Vessel Shells Affected by Hydrogen Flaking

2016 ◽  
Author(s):  
Robert Gérard ◽  
Michel De Smet ◽  
Rachid Chaouadi
Keyword(s):  
Fracture Toughness ◽  
Material Properties ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Operating Cycle ◽  
Safety Case ◽  
Safety Cases ◽  
Definition Of ◽  
Reactor Pressure ◽  
Trend Curve

During the summer outages of 2012, large numbers of nearly-laminar indications were found in the core shells of the Doel 3 and Tihange 2 reactor pressure vessels (RPV). As a consequence, both units remained in cold shutdown with their core unloaded. A series of examinations, tests and inspections were performed leading to the conclusion that the indications are hydrogen flakes and that they do not affect the structural integrity of the RPV, regardless of the operating mode, transient or accident condition. All this was documented in the Safety Case reports issued in December 2012 and in the Safety Case Addenda issued in April 2013 [1]. Based on those reports, the Belgian Federal Agency for Nuclear Control (FANC) authorized the restart of both units which went back on-line in June 2013. A key input required for this Safety Case was the definition of the appropriate material properties, in particular fracture toughness, for the RPV shells affected by hydrogen flakes. A material testing program on non-irradiated materials evaluated aspects like the possible effects of macro-segregations and local segregations (ghost lines) and of specimen orientation on the fracture toughness. The irradiation embrittlement sensitivity of the zone of macro-segregation in which the flakes are located was evaluated on the basis of the maximum enrichment in Cu, P and Ni in macro-segregations based on literature data. This was the basis of the trend curve of RTNDT evolution vs. fluence used in the Safety Cases submitted in 2012–2013. The restart authorization in 2013 was accompanied by a number of “mid-term” requirements, to be completed during the first operating cycle after the restart. One of these requirements was the mechanical testing of irradiated specimens containing hydrogen flakes, in order to confirm the conservativeness of the RTNDT trend curve used for the structural integrity analyses. After a first irradiation campaign of a material containing hydrogen flakes in the BR2 reactor of the Belgian Nuclear Research Center SCK.CEN, atypical results were obtained and the utility decided to shut down the units in March 2014. Detailed investigations involving three additional irradiation campaigns in BR2 including other reference materials, among which another material affected by hydrogen flakes, were performed in order to characterize this atypical behaviour and to derive a new conservative RTNDT trend curve. The resulting trend curve was accepted by the FANC and was used in the 2015 Safety Cases [1]. An overview of the Doel 3 and Tihange 2 safety cases is given in [6]. The paper summarizes the results of the material investigations on non-irradiated and irradiated materials and the process leading to the definition of this conservative RTNDT trend curve.

scholarly journals A Data Scatter for a Shift of the Ductile to Brittle Transition Temperature for WWER-1000 Reactor Pressure Vessel Materials

2018 ◽  
pp. 27-30
Author(s):  
V. Revka
Keyword(s):  
Transition Temperature ◽  
Standard Deviation ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Safety Margin ◽  
Temperature Shift ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Data Scatter ◽  
Reactor Pressure

In the most countries that operate the nuclear power plants with reactor pressure vessels a safety margin accounting a data scatter is applied for a conservative evaluation of a radiation shift of the ductile to brittle transition temperature for RPV metal. This scatter is to a significant extent due to material inhomogeneity and errors in determining the temperature shift and neutron fluence. In the regulatory practice of Ukraine, the obsolete approaches are used that can lead to an underestimation or overestimation of the transition temperature shift depending on the number of test data points. In order to use the updated regulatory approaches that will be consistent with international practice, it is necessary to know the magnitude of the data scatter on the transition temperature shift which is characterized by a standard deviation. Therefore, the aim of the research work was to estimate the data scatter for WWER reactor pressure vessel materials using statistical methods. The paper presents the results of a statistical analysis for a large array of surveillance test data for WWER-1000 reactor pressure vessels of NPP units which are operated in Ukraine. The data scatter for RPV base and weld metal has been estimated using a statistical treatment for the dependencies of a transition temperature shift, ΔTF, on the fast (Е > 0,5 MeV) neutron fluence. The ΔTF values have been derived from the Charpy impact tests. The Charpy V-notch specimens have been irradiated in the nuclear power reactors within a neutron fluence range of (3,0 ÷ 92,2)·1022 m-2 in the frame of a national surveillance program. The analysis has shown the data scatter relative to the average regression line for RPV materials is characterized by a standard deviation of 5,5 °С. Based on the results obtained, it was suggested to use a double standard deviation of 11 °С as a safety margin to provide a conservative estimate for the radiation shift of the transition temperature of the WWER-1000 reactor pressure vessel materials.

To Determination of the WWER RPV Steels Crack Resistance Characteristics

2013 ◽  
Author(s):  
Igor Orynyak ◽  
Maksym Zarazovskii ◽  
Sergii Radchenko ◽  
Volodymyr Kozlov
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Crack Resistance ◽  
Temperature Shift ◽  
Charpy Impact ◽  
Temperature Range ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure ◽  
15Kh2nmfa Steel

The efficiency of fracture toughness determined by the methodology of normative document PNAE G-7-002-86 has been analyzed. Crack resistance characteristics of WWER-1000 reactor pressure vessel base metal at unirradiated condition are obtained by experimental way. All specimens were made of the RPV support forging (15Kh2NMFA steel) of abandoned Crimean NPP Fracture toughness experiments were carried out on three types of specimens CT 1T, CT 0.5T and SEB over a temperature range from −130°C to −40°C in fully accordance to the ASTM E1921. Charpy impact energy data obtained on twelve specimens over a temperature range from −80°C to 80°C has been used to determine the 47J transition temperature. Comparison of obtained fracture toughness data with normative curve shows that the last one has unreasonably high lower shelf. It has been found that the PNAE G-7-002-86 Code, which uses the ideology of transition temperature shift, is too conservative to estimate WWER-1000 RPVs resistance against brittle fracture for the pressurized thermal shock (over 90 MPa·√m area of stress intensity factor).

Study on Application of PFM Analysis Method to Japanese Code for RPV Integrity Assessment Under PTS Events

2015 ◽  
Author(s):  
Kazuya Osakabe ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Input Data ◽  
Structural Integrity ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Analysis Method ◽  
Integrity Assessment ◽  
Analysis Models ◽  
Reactor Pressure ◽  
The U.S

A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.

Probabilistic Structural Integrity Evaluation on a Pressurized Water Reactor Pressure Vessel Under Pressure–Temperature Limit Operations

2015 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Temperature Limit ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

The normal reactor startup (heat-up) and shut-down (cool-down) operation limits are defined by the ASME Code Section XI-Appendix G, to ensure the structural integrity of the embrittled nuclear reactor pressure vessels (RPVs). In the paper, the failure risks of a Taiwan domestic pressurized water reactor (PWR) pressure vessel under various pressure-temperature limit operations are analyzed. Three types of pressure-temperature limit curves established by different methodologies, which are the current operation limits of the domestic RPV based on the KIa fracture toughness curve in 1998 or earlier editions of ASME Section XI-Appendix G, the recently proposed limits according to the KIC fracture toughness curve after the 2001 edition of ASME Section XI-Appendix G, and the risk-informed revision method proposed in MRP-250 report that provides more operational flexibility, are considered. The ORNL’s probabilistic fracture mechanics code, FAVOR, is employed to perform a series of fracture probability analyses for the RPV at multiple levels of embrittlement under such pressure-temperature limit transients. The analysis results indicate that the pressure-temperature operation limits associated with more operational flexibility will result in higher failure risks to the RPV. The shallow inner surface breaking flaw due to the clad fabrication defect is the most critical factor and dominates the failure risk of the RPV under pressure-temperature limit operations. Present work can provide a risk-informed reference for the safe operation and regulation of PWRs in Taiwan.

Sign in / Sign up

Export Citation Format

Share Document