Hans-Werner Viehrig

Application of different nuclides in retrospective dosimetry

The activities of nuclides produced via the neutron irradiation of reactor pressure vessel (RPV) steel are used to validate respective fluence calculations. Niobium, nickel, and technetium isotopes from RPV trepans of the decommissioned NPP Greifswald (VVER-440) have been analyzed. The activities were determined by TRAMO (Monte-Carlo) fluence calculations, newly applying 640 neutron-energy groups and ENDF/B7 data. Relative to earlier results, fluences up to 20% higher have been computed, leading to somewhat better agreement between measurement and calculation, particularly in the case of Tc-99. (authors)

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

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 (SNO.1.4) received a fluence of 2.4 ×10 19 neutrons/cm 2 at the inner surface; it decreases to 0.8 ×10 19 neutrons /cm 2 at the outer surface. The material investigations were done using a trepan from the circumferential core weld. The reference temperature To was calculated with the measured fracture toughness values, K Jc , at brittle failure of the specimen. The K Jc values show a remarkable scatter. The highest To was about 50°C at a distance of 22 mm from the inner surface of the weld. The Charpy transition temperature TT 41J 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, RTT SINTAP 0 , envelops the K Jc 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 K Jc values has to be applied.

IAEA Coordinated Research Project on Master Curve Approach to Monitor Fracture Toughness of RPV Steels: Final Results of the Experimental Exercise to Support Constraint Effects

The precracked Charpy single-edge notched bend, SE(B), specimen (PCC) is the most likely specimen type to be used for determination of the reference temperature, T0 , with reactor pressure vessel (RPV) surveillance specimens. Unfortunately, for many RPV steels, significant differences have been observed between the T0 temperature for the PCC specimen and that obtained from the 25-mm thick compact specimen [1TC(T)], generally considered the standard reference specimen for T0 . This difference in T0 has often been designated a specimen bias effect, and the primary focus for explaining this effect is loss of constraint in the PCC specimen. The International Atomic Energy Agency (IAEA) has developed a coordinated research project (CRP) to evaluate various issues associated with the fracture toughness Master Curve for application to light-water RPVs. Topic Area 1 of the CRP is focused on the issue of test specimen geometry effects, with emphasis on determination of T0 with the PCC specimen and the bias effect. Topic Area 1 has an experimental part and an analytical part. Participating organizations for the experimental part of the CRP performed fracture toughness testing of various steels, including the reference steel JRQ (A533-B-1) often used for IAEA studies, with various types of specimens under various conditions. Additionally, many of the participants took part in a round robin exercise on finite element modeling of the PCVN specimen, discussed in a separate paper. Results from fracture toughness tests are compared with regard to effects of specimen size and type on the reference temperature T0 . It is apparent from the results presented that the bias observed between the PCC specimen and larger specimens for Plate JRQ is not nearly as large as that obtained for Plate 13B (−11°C vs −37°C) and for some of the results in the literature (bias values as much as −45°C). This observation is consistent with observations in the literature that show significant variations in the bias that are dependent on the specific materials being tested. There are various methods for constraint adjustments and two methods were used that reduced the bias for Plate 13B from −37°C to −13°C in one case and to − 11°C in the second case. Unfortunately, there is not a consensus methodology available that accounts for the differences observed with different materials. Increasing the Mlim value in the ASTM E-1921 to ensure no loss of constraint for the PCC specimen is not a practicable solution because the PCC specimen is derived from CVN specimens in RPV surveillance capsules and larger specimens are normally not available. Resolution of these differences are needed for application of the master curve procedure to operating RPVs, but the research needed for such resolution is beyond the scope of this CRP.Copyright

PERFECT — Prediction of Irradiation Damage Effects in Reactor Components: Overview of RPV Mechanics Sub-Project

The EURATOM 6th Framework Integrated Project PERFECT (Prediction of Irradiation Damage Effects in Reactor Components) addresses irradiation damage in RPV materials and components by multi-scale modeling. This state-of-the-art approach offers many potential advantages over the conventional empirical methods used in current practice of nuclear plant lifetime management. Launched in January 2004, this 48-month project is focusing on two main components of nuclear power plants which are subject to irradiation damage: the ferritic steel reactor pressure vessel, and the austenitic steel internals. It is the purpose of the present paper to provide an overview of work being carried out in the RPV Mechanics Sub-project of PERFECT to predict the fracture behavior of PWR, BWR and WWER systems.Copyright

Investigation of the Beltline Welding Seam and Base Metal of the Greifswald WWER-440 Unit 1 Reactor Pressure Vessel

The investigation of reactor pressure vessel (RPV) materials from decommissioned nuclear power plants (NPPs) offers the unique opportunity to scrutinize the irradiation behavior under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterization. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I); irradiated and recovery annealed (IA); and irradiated, recovery annealed, and re-irradiated (IAI). The working program is focused on the characterization of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T 0 following the American Society for Testing of Materials (ASTM) Test Standard E1921―08 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step, the trepans taken from the RPV Greifswald unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the master curve (MC) approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T 0 varies through the thickness of the welding seam. The lowest T 0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root To shows a wavelike behavior. The highest T 0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on subsize Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy-V transition temperature TT 41J estimated with results of subsize specimens after the recovery annealing was confirmed by the testing of standard Charpy-V-notch specimens. The evaluated TT 41J shows a better accordance with the irradiation fluence along the wall thickness than the master curve reference temperature T 0 . The evaluated T 0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. The K Jc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during two campaign operations can be assumed to be low for the weld and base metal.

A Round Robin Program of Master Curve Evaluation Using Miniature C(T) Specimens: 3rd Report — Comparison of T0 Under Various Selections of Temperature Conditions

The Master Curve approach for the fracture toughness evaluation is expected to be a powerful tool to ensure the reliability of long-term used RPV steels. In order to get sufficient number of data for the Master Curve approach coexistent with the present surveillance program for RPVs, the utilization of miniature specimens, which can be taken from broken halves of surveillance Charpy specimens, is important. CRIEPI developed the test technique for the miniature C(T) specimens (Mini-CT), whose dimensions are 4 × 10 × 10 mm, and verified the basic applicability of Master Curve approach by means of Mini-CT for the determination of fracture toughness of typical Japanese RPV steels. A round robin program is organized in order to assure the robustness of the testing procedure to the difference in testing machines or operators. The first and second round robin tests (PVP2012-78661 [1], PVP2013-97936 [2]) suggested that the reference temperature T0 evaluation technique by Mini-CT specimen potentially is fairly robust in regard to difference in testing machines and operators, and gives similar loading rate dependency to the larger C(T) specimens. As the final year of the round robin program, “blind tests” were carried out. Here, detailed material information such as the type of materials, estimated T0, existing fracture toughness data for the material, were not given with the specimen, and 6 organizations independently selected the test temperature based on Charpy full curve of the tested material. The selection of test temperature has the variation of −120 °C to −150 °C among the organizations. 8 to 20 specimens in a set were subjected to the Master Curve evaluation and all the 6 organizations successfully obtained valid T0. The scatter range in T0 was at most 16 °C, which was within the acceptable scatter range specified in ASTM E1921-10e1. The selection of test temperature seems to give limited effect as like as that in larger specimens.Copyright

PERFECT—Prediction of Irradiation Damage Effects in Reactor Components: Update of Progress in RPV Mechanics Sub-Project

The EURATOM 6th Framework Integrated Project PERFECT (Prediction of Irradiation Damage Effects in Reactor Components) addresses irradiation damage in RPV materials and components by multi-scale modeling. This approach offers many potential advantages over the conventional empirical methods used in current practice of nuclear plant lifetime management. Launched in January 2004, PERFECT is a 48-month project focusing on two main components of nuclear power plants which are subject to irradiation damage: the ferritic steel reactor pressure vessel (RPV), and the austenitic steel internals. It is the purpose of the present paper to provide an update of progress of work being carried out in the Mechanics Sub-project of PERFECT to predict the fracture behavior of RPVs in PWR and WWER systems.Copyright