Current Issues in Operated WWER Type RPVs

2008 ◽  
Author(s):  
Milan Brumovsky´ ◽  
Milos Kytka
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessels ◽  
Point Of View ◽  
Radiation Embrittlement ◽  
Irradiation Effects ◽  
Static Fracture ◽  
High Level ◽  
Near Future ◽  
Surveillance Specimens ◽  
Reactor Pressure

RPVs of WWER type reactors are manufactured from other type of steels (15Kh2MFA of Cr-Mo-V type for WWER-440 and 15Kh2NMFA of Ni-Cr-Mo-V type for WWER-1000) and according to other Codes and standards than PWR ones, thus some specific problems are currently more important for WWER. The principal problem lies in relatively small number of manufactured and operated WWER type NPPs. Even though a high level of unification in RPVs exists — practically only two designs of RPVs exists (WWER-440 and WWER-1000) — total number is still small. All WWER-440 RPV are practically identical, either they were manufactured for V-230 or V-213 model: the only difference is in the purity of used materials and existence/non-existence of the surveillance programmes. (Fact that some V-230 type vessels were not covered by austenitic cladding is not important from irradiation effects point of view.) Regarding WWER-440/V-230 types, it is necessary to take into account, that even though most of them were successfully annealed, only some of them are still in operation but most of them will be closed in near future. Similar situation is with WWER-1000 RPVs, either they were manufactured for V-320 (most frequent), or V-338 or the newest V-428 — differences are practically only in the content of nickel in critical weldments and/or in design of surveillance specimens capsules. But, Large advantage of all WWER surveillance programmes is in loading static fracture toughness specimens in all programmes. The papers tries to summarize and analyze all current issues connected with radiation embrittlement of operated reactor pressure vessels of WWER type.

Lifetime Management of WWER Reactor Pressure Vessels

2004 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Point Of View ◽  
Mitigation Measures ◽  
Economic Point ◽  
Neutron Fluences ◽  
Vessel Integrity ◽  
Surveillance Specimens ◽  
Reactor Pressure

Reactor pressure vessels are the most important components of nuclear power plants from safety and economic point of view. Thus, special interest is given to their lifetime management that requires a precious and periodical evaluation of their conditions. In Czech Republic, special program has been prepared for reactor pressure vessels in both nuclear power stations — Dukovany NPP with 4 units of WWER-440 MW and Temelin with 2 units of WWER-1000 MW reactors. This program is based on the following activities: • detailed calculations of real and potential regimes, especially of pressurized thermal shock events, • determination of vessel properties on the basis of testing surveillance specimens from standard, modified and supplementary programs, • calculations of neutron fluences on vessel wall, • measurements of neutron fluences on surveillance specimens and in vessel cavity, • evaluation of all results, evaluation of fluence and material property trends, • recommendations for future reactor pressure vessel operation (potential mitigation measures). The paper describes in detail all these activities and gives examples of their results and final evaluations. Lifetime assessment is based on a “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs (VERLIFE)” using “Master Curve” approach for vessel integrity evaluation [1].

Issue of Application for the Pre-Cracked Charpy Specimens to Estimate WWER-1000 RPV Metal Embrittlement

2016 ◽  
Author(s):  
Volodymyr M. Revka ◽  
Liudmyla I. Chyrko
Keyword(s):  
Fracture Toughness ◽  
Test Data ◽  
Charpy Impact ◽  
Pressure Vessels ◽  
Reference Temperature ◽  
Fracture Toughness Test ◽  
Toughness Test ◽  
Standard Specimens ◽  
High Level ◽  
Reactor Pressure

In this study the analysis of fracture toughness test data has been performed in terms of estimation of the proper T0 value for several WWER-1000 RPV materials in unirradiated condition. The surveillance test data for the standard and reconstituted specimens were included in the analysis. It was found that a reference temperature T0 for reconstituted specimens is 31°C higher on average in comparison to the standard specimens. The possible reason is a high level of the stress intensity factor Kmax during the cycle at the stage of completion of crack tip sharpening for standard specimens. Furthermore, the Charpy impact and fracture toughness test data for standard and reconstituted specimens have been compared considering the known relationship between the reference temperature T0 and the transition temperature T28J which corresponds to the Charpy energy level of 28 J. Another objective of this study was to compare the RPV metal embrittlement rate for the two reactor pressure vessels using surveillance test data from standard and reconstituted fracture toughness specimens. The analysis has shown that test data for the reconstituted specimens is consistent with the test data for the standard specimens with regard to the embrittlement rate.

Austenitic Cladding and Master Curve

2015 ◽  
Author(s):  
Milan Brumovsky ◽  
Milos Kytka ◽  
Radim Kopriva ◽  
Michal Falcnik
Keyword(s):  
Fracture Toughness ◽  
Master Curve ◽  
Room Temperature ◽  
Pressure Vessels ◽  
Crystallographic Structure ◽  
Delta Ferrite ◽  
Transition Behavior ◽  
Static Fracture ◽  
Cladding Layers ◽  
Reactor Pressure

Reactor pressure vessels of PWR/BWR/WWER type reactors are covered by austenitic cladding made by welding on their inner wall. Austenitic materials usually have no transition temperature behavior as they have fcc crystallographic structure. But, austenitic cladding made by welding contain usually up to 8 % of delta-ferrite that results in some transition behavior of fracture properties. This transition can be observed in temperature region below room temperature. Surprisingly, this transition behavior in static fracture toughness of both cladding layers can be well described by “Master curve” approach. Results from testing austenitic cladding for WWER type reactors will be shown and discussed, ether in unirradiated as well as irradiated conditions — only small changes in fracture toughness properties in this transition region are observed as a result of irradiation.

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.

Fracture Toughness of Highly Irradiated Pressure Vessel Steels in the Upper Shelf Temperature

2006 ◽  
Author(s):  
Hiroshi Matsuzawa ◽  
Toru Osaki
Keyword(s):  
Fracture Toughness ◽  
Chemical Composition ◽  
Reference Material ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Charpy Impact ◽  
Pressure Vessels ◽  
Shelf Energy ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Nine Reactor Pressure Vessel (RPV) Steels and four RPV weld were irradiated up to 1.2 × 1024n/m2 fast neutron fluence (E>1MeV), and their fracture toughness and Charpy impact energy were measured. As chemical compositions, such as Cu, are known to affect the fracture toughness reduction due to neutron exposure, the above steels were fabricated by changing chemical composition widely to cover the chemical composition of the RPV materials of the operating Japanese nuclear power plants. 2.7 mm thick compact specimens were used to measure the upper shelf fracture toughness of highly irradiated materials, and their Charpy upper shelf energy was also measured. By correlating Charpy upper shelf energy to fracture toughness, the upper shelf fracture toughness evaluation formulae for highly irradiated reactor pressure vessel steels were developed. Both compact and V-notched Charpy impact specimens were irradiated in a test reactor. The fast neutron flux above 1MeV was about 5 × 1016n/(m2s). Charpy impact specimens made of Japanese PWR reference material containing 0.09w% Cu were irradiated simultaneously. The upper shelf energy of the reference material up to the medium fluence level showed little difference in the reduction of upper shelf energy to that which had been in the operating plant and which was irradiated to the same fluence. The developed correlation formulae have been adopted in the Japan Electric Association Code as new formulae to predict the fracture toughness in the upper shelf region of reactor pressure vessels. They will be applied to time limited ageing analysis of low upper shelf reactor pressure vessels in Japan, on a concrete technical basis in very high fluence regions.

Assessment of Irradiation Conditions in WWER-440 (213) RPV Surveillance Location

2004 ◽  
Author(s):  
A. Ballesteros ◽  
J. Bros ◽  
L. Debarberis ◽  
F. Sevini ◽  
D. Erak ◽  
...  
Keyword(s):  
Neutron Flux ◽  
Fuel Cycle ◽  
Pressure Vessels ◽  
Irradiation Temperature ◽  
The Arctic ◽  
Charpy Specimen ◽  
Exposure Level ◽  
Reactor Operation ◽  
Surveillance Specimens ◽  
Reactor Pressure

The key component of WWER is the Reactor Pressure Vessel (RPV). The evaluation and prognosis of RPV material embrittlement and the allowable period of its safe operation are performed on the basis of impact test results of irradiated surveillance specimens (SS). The main problem is that the SS irradiation conditions (temperature of SS, neutron flux and neutron spectrum) have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV condition. In particular, the key issue is the possible difference between the irradiation temperature of the SS and the actual RPV temperature. It is recognized that the direct measurement of temperature by thermocouples during reactor operation is the only way for receiving reliable information. In addition, the neutron field’s parameters for surveillance specimens have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level. The COBRA project (http://ie.jrc.cec.eu.int/ames/), which started in August 2000 and had a duration of three years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contain state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during the reactor operation. The selected reactor was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of the capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5±4°C. The results obtained show that there is not need in temperature correction when data of surveillance specimens studies are used for assessment of WWER-440/213 reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals ∼2.7·1012 cm−2s−1 with E>0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.

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.

Use of Small Specimens for Evaluation of Irradiated Materials

2013 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
Small Volume ◽  
Radioactive Material ◽  
Surveillance Programs ◽  
Series Of Experiments ◽  
Point Bend ◽  
Impact Data ◽  
Surveillance Specimens ◽  
Reactor Pressure ◽  
Fracture Toughness Specimen

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. Small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. The PCVN specimen as well as any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of RPVs. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Precracking and testing of Charpy surveillance specimens would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. However, there is a growing number of indications that there might be a bias in the reference fracture toughness transition temperature, To values derived from PCVN and compact specimens. The present paper summarizes data from the series of experiments that use subsize specimens for evaluation of the transition fracture toughness of reactor pressure vessel (RPV) steels. Two types of compact specimens and three types of three-point bend specimens from five RPV materials were used in these subsize experiments. The current results showed that To determined from PCVN specimens with width (W) to thickness (B) ratio W/B = 1, on average, are lower than To determined from compact specimens with W/B = 2. At the same time, three-point bend specimens with W/B = 2 exhibited To values that were very similar to To values derived from compact specimens. Constraint corrections developed by Dodds et al. are applied to assess the bias.

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.

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