Milos Kytka

Further Results on Attenuation of Neutron Embrittlement Effects in a Simulated RPV Wall

A carefully designed irradiation experiment was conducted in which a 190 mm thick reactor pressure vessel (RPV) wall has been simulated using nineteen 10 mm thick slices, 18 of which are made from key RPV steels, and irradiated under test reactor conditions to investigate the through-wall attenuation of neutron embrittlement. Preliminary results for two of the irradiated materials (a low copper content plate and a high copper content Linde 80 flux weld) were reported earlier. The third irradiated RPV steel was the international reference steel designated JRQ, and this paper describes the results for this steel along with updated analyses for the other two steels. Comparisons of predicted attenuation changes in toughness properties with measured fracture toughness and Charpy V-notch results are presented for all three RPV steels. The predictions of through-wall attenuation follow the practice defined in ASTM E900-02 and Regulatory Guide 1.99, Revision 2, in which the attenuation of high energy neutron fluence (E>1 MeV) is projected based upon an approximate displacements per atom (dpa) change through the wall thickness. The resultant degree of material damage using this dpa-based fluence change is estimated using current embrittlement correlation models.

Characterisation of Mechanical Properties by Small Punch Test

The specific desired properties for structures and components working in critical environments (e.g. different structure parts of power plants) require current information about degradation processes coming out in materials. Obtaining of this information by the help of the classical tests of mechanical properties (tensile test, Charpy test, fracture toughness test, creep test etc.) is very limited namely in the case of nuclear power plants pressure vessel. Hence, the new innovative techniques based on miniaturized specimens have been developed for evaluation of mechanical properties and their changes. One of very promising techniques is Small Punch Test. Present paper deals with characterization of three different steels (15Ch2MFA, 10GN2MFA with different heat treatment and steel O8Ch18NT10 with various degree of deformation).

Long Term Operation and Surveillance Specimen Program of WWER RPVs

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.Copyright

Results on Attenuation of Neutron Embrittlement Effects

A carefully designed irradiation experiment was conducted in which a 180-mm thick reactor pressure vessel (RPV) wall has been simulated using eighteen 10-mm slices of some key RPV steels and irradiated under test reactor conditions to investigate the through wall attenuation of neutron embrittlement. Results from two of the irradiated materials (a low copper content plate and a high copper content Linde 80 flux weld) were reported in 2006. Another RPV plate, the international reference steel, JRQ, was also extensively irradiated in the simulated RPV wall. Comparisons of predicted attenuation changes in toughness properties using measured Charpy V-notch results are presented for the JRQ steel and compared to the results from the low copper content plate and the high copper content Linde 80 weld metal. Also, Charpy V-notch and Master Curve fracture toughness test results are compared for the low copper plate and the high copper weld. Predictions are made of through-wall attenuation following the practice defined in ASTM E 900-02 and Regulatory Guide 1.99, Revision 2, in which the attenuation of high energy neutron fluence (E > 1 MeV) is projected based upon an approximate displacements per atom (dpa) change through the wall thickness.. The resultant degree of material damage using this dpa-based fluence change is estimated using the ASTM E 900–02 embrittlement correlation model and compared to the experimental data.Copyright

On the Failure Mechanisms in Reactor Pressure Vessel with Austenitic Cladding

The austenitic cladding of the WWER pressure vessel is made from two different layers with different fracture toughness values. Based on the fractographic analysis of the tested specimens in the initial, as well as in the irradiated conditions, it was found that individual failure micromechanisms take place during the crack propagation. The obtained results were used to find the relationship between the failure micromechanism changes and the fracture toughness values, as well as to assess the effect of neutron irradiation on the failure micromechanisms.

Correlation Between Transition Temperature Shifts of Impact Tests Using Standard and Subsize Specimens

Special weld of 15Kh2MFA (Cr-Mo-V) type steel was investigated in several laboratories within the IAEA Coordinated round robin exercise. Specimens were irradiated by several fluences and impact tests of standard Charpy size and 5×5 and 3×4 mm specimens were then tested to determine appropriate transition temperatures and their shifts.When applied normally used criteria (impact energy, lateral expansion, ductile fracture appearance), shifts in standard specimens and subsize specimens correlate well but not in 1:1 ratio — shifts in subsize specimens were observed to be smaller than in standard ones — the smaller size of specimens, the smaller transition temperature shifts have been obtained.The paper tries to find appropriate criteria for better correlation between shifts in different sizes of impact specimens that can be supported by proper analysis.Copyright

Austenitic Cladding and Master Curve

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.Copyright

Monitoring Degradation in Nuclear Reactors

Nuclear reactor materials are affected by many stressors during their operation that result either from the nuclear reactions in the reactor core or by operation conditions (temperature, pressure) and water environment.Generally, several different methods of monitoring damage in reactor materials are applied, depending on their type, design and conditions:- Destructive, usually represent by surveillance specimens,- Semi-destructive, usually performed by cutting small specimens from component surface, e.g. for small punch tests,- Non-destructive, representing by non-destructive inspection methods applied during in-service inspections, like traditional ultrasonics, optical, dye-penetrant, eddy current etc. and also some new like automated ball indentation, thermoelectric power measurements etc.The paper summarizes the possibility of application of these methods on main reactor components and shows some typical results and problems.Copyright

Comparison of Transition Temperature Shifts From Notch Impact and Fracture Toughness (Master Curve) Testing

Evaluation of integrity and lifetime of reactor pressure vessels is usually based on fracture mechanics approach using empirical correlation between transition temperatures from impact tests and static fracture toughness test results in the form of “design curve”. Moreover, material degradation during operation is also usually monitored by impact surveillance specimen testing under the assumption that shifts in temperature dependencies if impact toughness and static fracture toughness are the same.To verify this assumption, study of the correlation between these two shifts has been performed on WWER steels — 15Kh2MFA (Cr-Mo-V) and 15Kh2NMFA (Ni-Cr-Mo-V) types. Several sources of results have been used : (a) reconstitution of tested remains of Charpy V-notch impact test specimens from irradiated programs was performed to obtain pre-cracked Charpy size specimens for three point bending type fracture toughness testing, (b) comparison of tests results from surveillance programs irradiated by similar fluences, (c) experimental irradiation programs with accelerated irradiation in research reactor.Additionally, some results from the recent study of irradiation embrittlement of high nickel weld are included — its behavior shows to some extraordinary tendency.Thanks to the use of reconstitution, both series of specimen were irradiated under the same conditions — temperature and neutron fluence and comparison is reliable. Results show that transition temperatures from fracture toughness testing are larger than those from Charpy impact tests. Similar results have been obtained also for other two groups of results.Copyright

Characterization of Local Stress-Strain Behavior in WWER 440 Weld and Base Metal by Instrumented Indentation Technique

15Ch2MFA (base metal) as well as 10ChMFT (weld) steels used for WWER 440 nuclear reactor pressure vessel manufacturing present a gradient in mechanical properties through the wall thickness, which can hardly be assessed by conventional testing such as tensile or Charpy tests. Mechanical properties in the weld and base metal were therefore determined by performing a series of instrumented indentations across the weld at room temperature. The results were treated by so-called automated ball indentation technique. Local stress-strain behavior obtained by instrumented indentation was correlated to the tensile test data and microstructure characterized by metallographic analysis.