William Server

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.

Attenuation of Neutron Radiation Damage Through a Simulated RPV Wall

An experiment has been conducted in which a 180-mm thick reactor pressure vessel (RPV) wall has been simulated using eighteen 10-mm slices and irradiated under test reactor conditions to investigate the through wall attenuation of neutron embrittlement. Attenuation of neutron radiation damage through the wall of an RPV is a process that involves a changing neutron flux spectrum. The effect of the changing spectrum has not been fully studied to define the change in fracture toughness properties through the RPV wall. One low copper content base metal and one high copper content Linde 80 weld metal have been irradiated in various positions through the simulated wall to allow quantification of an improved experimentally-based embrittlement attenuation model. Comparisons are made of predicted attenuation changes in toughness properties with measured fracture toughness and Charpy V-notch results for the high copper content weld metal and the low copper content plate. The predictions of through-wall attenuation follow the practice defined in ASTM E 900-02, in which the attenuation of high energy neutron fluence (E >1 MeV) is projected based upon displacements per atom (dpa) change through the wall thickness. The resultant degree of material damage (Charpy V-notch 41 J transition temperature, T41J) using this dpa-based fluence change is estimated also using the ASTM E 900-02 embrittlement model. The irradiation-induced shift in T41J (ΔT41J) is typically assumed to infer the shift in fracture toughness transition temperature to be used for structural integrity assessments for the reactor pressure vessel. This assumption will be checked by measuring Master Curve fracture toughness properties for the high copper content weld metal and the low copper content plate.

Critical Review of Through-Wall Attenuation of Mechanical Properties in RPV Steels

This paper reviews the current state of knowledge on attenuation of damage parameters in reactor pressure vessels (RPVs). There are two methods for evaluating attenuation of properties through a reactor pressure vessel wall. The first is by direct measurement of the change in mechanical properties from decommissioned RPV sections or from simulated RPV wall experiments. It is shown that, although this approach is appealing, issues associated with knowledge of the start-of-life properties throughout the vessel wall sample and material property data scatter have made past measurements non-definitive in establishing attenuation changes. There is a need for further data on the direct measurement of attenuation, and an experiment is described, that is planned in 2002 under IAEA sponsorship. An alternative method for evaluating embrittlement is the use of a neutron damage exposure parameter and attenuation model coupled with an embrittlement correlation developed from surveillance capsule testing. The significant change in neutron flux spectrum when neutrons are attenuated through the RPV wall defines the need for a suitable neutron exposure damage parameter. The best available neutron exposure damage parameter is dpa. It is shown that plant-specific calculation of dpa through the RPV wall is the best method to be used for the neutron exposure. However, it can lead to slightly less attenuated values for damage at ¼-T and ¾-T for the vessel, as compared to using the simple exponential model quoted in Regulatory Guide 1.99, Rev. 2. Finally it is concluded that when using a surveillance correlation model to predict the attenuation of mechanical properties through the RPV wall, the use of a mechanistically guided model appears to be more appropriate than the embrittlement correlation provided in Regulatory Guide 1.99, Revision 2.

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

Application of the Master Curve Approach for Abnormal Material Conditions

The range of applicability of Master Curve testing Standard ASTM E 1921 is limited to macroscopically homogeneous steels with “uniform tensile and toughness properties”. A majority of structural steels appear to satisfy this requirement by exhibiting fracture toughness data which comply with the assumed KJc vs. temperature dependence and scatter within the specified validity area. As indicated in ASTM E 1921 a criterion for material macroscopic inhomogeneity is often applied using the 2% lower bound (possibly also the 98% upper bound). Data falling below this 2% lower-limit curve may be an indication of material inhomogeneity or susceptibility to grain boundary fracture. When this situation occurs, it is recommended to analyze the material with the so-called SINTAP procedure, which is intended for randomly inhomogeneous materials to assure a conservative lower-bound estimate. When a data set distinctly consists of two or more different data populations instead of one (due to variation of irradiation dose or specimen extraction depth, for instance) adoption of a bimodal (or a multimodal) Master Curve model is generally appropriate. These modal models provide information if the deviation of distributions is statistically significant or if different distributions truly exist for values of reference transition temperature, T0 , characteristic of separate data populations. In the case of data sets representing thick-walled structures (i.e., reactor pressure vessels), indications of abnormal fracture toughness data can be encountered such that material inhomogeneity or fracture modes other than pure cleavage should be suspected. A state-of-the-art review for extended, non-standard Master Curve data and techniques highlights limits of applicability in situations where the basic ASTM E 1921 procedure is not appropriate for material homogeneity or different fracture modes.© 2007 ASME

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

The Technical Foundations of a Unified Adjusted Reference Temperature for RPV Fracture Toughness

The use of a reference temperature to define the transition from ductile-to-brittle behavior is a well-established practice in reactor pressure vessel integrity analysis. Current procedures for analysis of irradiated pressure vessel steels use an Adjusted Reference Temperature (ART or RTPTS) that shift the unirradiated RTDNT value based on Charpy data and apply additional margins. The recent adoption of an ASTM Standard Test Method for Determination of Reference Temperature, To, for Ferritic Steels in the Transition Range (E 1921) has made direct determinations of fracture toughness reference temperature in irradiated materials using Master Curve feasible. ASME Code Case N629 recognizes the potential of this new technology as an alternative means of determining an ART value for use in reactor pressure vessel integrity analysis. This alternative approach has been used in recent plant specific submittals to the NRC. ASTM sub-committee E10.02 is currently considering adopting a standard to provide a unified definition of the Adjusted Reference Temperature. This unified definition would provide consistent procedures for using either RTNDT or Master Curve based definitions of ART. This unified definition would provide common margins for the two approaches and assure a consistent level of confidence in the analysis. The statistical basis of the Master Curve makes it possible to establish a more rational basis for the margins applied in both approaches.

Assessment of U.S. Embrittlement Trend Equations Considering the Latest Available Surveillance Data

The mechanistically-guided embrittlement correlation model adopted in ASTM E 900-02 was based on a database of U.S. surveillance results current through calendar year 1998. There exists now an extensive amount of new surveillance data that includes a large amount of boiling water reactor (BWR) results from an integrated, supplemental surveillance program designed to augment the plant-specific BWR surveillance programs. These recent data allow a statistical test of the ASTM E 900-02 embrittlement correlation, as well as the NRC correlation model currently being used in the pressurized thermal shock (PTS) re-evaluation effort and the older Regulatory Guide 1.99, Revision 2 correlation. Even though the ASTM E 900-02 embrittlement correlation is a simplified version of the NRC model, a comparison of the two embrittlement correlation models utilizing the new database has proven to be revealing. Based on the new BWR data, both models are inadequate in their ability to predict BWR results; this inadequacy has even more significance for extrapolation outside of the database for BWR heat-up and cool-down curves, as well as some pressurized water reactor (PWR) heat-up curves. Other aspects of the two models, as revealed from this preliminary look at the new data, are presented.Copyright

Integrity and embrittlement management of reactor pressure vessels (RPVs) in light-water reactors

Abstract: Validation of the current and continued integrity approaches for the reactor pressure vessel (RPV) has been shown through the many years of safe operation of light-water reactor vessels. There have not been any vessel failures, and this fact is primarily due to proper embrittlement management programs and structural integrity assessment methods. Additionally, there have been several large-scale experiments performed to further validate the integrity of RPVs. This chapter is focused on the embrittlement and integrity management approaches used in different countries that are operating nuclear power plants.

Thermal Annealing of Reactor Pressure Vessels

Some of the current fleet of nuclear power plants is poised to reach their end of life and will require an operating life time extension. Therefore, the main structural components, including the reactor pressure vessel (RPV), will be subject to higher neutron exposures than originally planned. These longer operating times raise serious concerns regarding our ability to manage the reliability of RPV steels at such high doses. Thermal annealing is the only option that can, to some degree, recover irradiated beltline region transition temperature shift and recover upper shelf energy properties lost during radiation exposure and extend RPV service life. This paper reviews the experience accumulated internationally with development and implementation of thermal annealing to RPV and potential perspectives for carrying out thermal annealing on US nuclear power plant RPVs.Copyright