Enhanced radiation embrittlement at high fluence, indicative of extended operating life beyond 60 years for current operating pressurized water reactor (PWR) vessels, has been identified as a potential limiting degradation mechanism. Currently, there are limited U.S. power reactor surveillance data available at fluences greater than 4 × 1019 n/cm2 (E > 1 MeV) for comparison with existing embrittlement prediction models. Additional data will be required to support extended operations to 80+ years, where some plants are projected to have peak vessel fluences approaching 1 × 1020 n/cm2. A number of programs are designed to contribute to the high fluence surveillance data to support extended operating life. The U.S programs include the Coordinated PWR Reactor Vessel Surveillance Program (CRVSP), the PWR Supplemental Surveillance Program (PSSP), and the Light Water Reactor Sustainability (LWRS) Program. The LWRS Program involves generation of high fluence test reactor data on many different reactor pressure vessel steels and model alloys, including some of the same PWR vessel materials irradiated to higher fluences in conventional power reactor surveillance programs. This paper surveys the existing high fluence data and the data projected to come from the above listed programs to show when such data will become available. The data will be used to validate or revise embrittlement trend correlations applicable for the high fluence regime. Mechanical property data are being developed, and fine-scale microstructure data are being produced using state-of-the-art methods.
A pressurized water reactor (PWR) supplemental surveillance program (PSSP) is being designed to provide high-fluence reactor vessel material embrittlement data for the operating U.S. PWRs. Peak reactor pressure vessel (RPV) fluence levels as high as 7 × 1019n/cm2 (E > 1.0 MeV) will be attained as PWRs operate to 60 years and potentially beyond. Therefore, a need exists to obtain high-fluence PWR surveillance data to validate or revise embrittlement trend correlations (ETC) applicable for the high-fluence regime. Without the availability of high-fluence PWR surveillance data, it may be necessary to use an overly conservative ETC, or an ETC with a high margin at high fluence, which could constrain plant pressure–temperature operating curves, increasing startup and shutdown times and costs or increasing the potential of exceeding the pressurized thermal shock screening limit. The PSSP is designed to supplement data produced by the existing 10 CFR 50 Appendix H surveillance programs. The two proposed PSSP capsules will contain Charpy specimens reconstituted from tested PWR surveillance capsule materials, carefully selected for material type, chemistry, and fluence to optimize future ETC development. These capsules will be inserted into one or two U.S.-based Westinghouse-designed operating nuclear power plants for continued irradiation. The selected host plant(s) have relatively high capsule irradiation flux locations, enabling production of high-fluence data prior to the U.S. plants reaching 60 years of operation. The PSSP capsule irradiation will increase the fluence levels up to 1 × 1020n/cm2 on select groups of reactor vessel materials. This paper describes the basis for the PSSP, plant selection for irradiation, and material selection.
Regions of higher-than-normal carbon content due to carbon macro-segregation have been found in large, pressure retaining forged ferritic steel components in some nuclear reactors. Higher carbon content in ferritic steel can decrease the resistance to fracture from the presence of flaws in the material. Acceptable margins against failure of pressurized components in nuclear safety systems must be maintained throughout their service life to ensure core integrity for all operational and postulated transient loading events. Should carbon macro-segregation substantially reduce the material resistance to fracture in safety components, then the margins against through-wall flaw propagation may fall below those specified by regulatory requirements to ensure adequate component and reactor core integrity. Probabilistic fracture mechanics (PFM) analyses were performed to assess the risk and structural significance of postulated carbon macro-segregation in large, forged pressure retaining components in pressurized water reactors (PWRs). The risk assessment was performed to evaluate several forged components and two classes of loading events. The forged components include the ring and head forgings in the reactor pressure vessel (RPV), steam generator (S/G) and pressurizer. The loading events used in the risk evaluation include pressurized thermal shock (PTS) transient events and a normal RPV cooldown event. The analyses included a range of component dimensions, surface and embedded flaw distributions, various levels of carbon macro-segregation up to and beyond the maximum measured values for the components, and the effects of neutron irradiation, including the effects of potential copper and phosphorus co-segregation. The PFM analyses were performed using the software, Fracture Analysis of Vessels, Oak Ridge (FAVOR). The results from the risk assessment indicate that: acceptable margins against failure are maintained through an 80-year operating interval even if carbon macro-segregation were to be present in RPV, S/G and pressurizer ring and head forgings in PWRs; and the risk associated with the presence of carbon macro-segregation in PWR ring and head forgings is significantly lower than regulatory risk related acceptance criteria.
The numerous indications recently found by UT-inspection in the reactor pressure vessel shell forgings at two Belgian nuclear power plants have raised some concerns about the effects of such indications on the vessel integrity and fitness for continued service. The UT indications have been attributed to hydrogen flaking, and preliminary estimates give a density of ∼40 indications per liter, with diameter of about 10–14 mm, oriented at a shallow ∼10° angle to the vessel inner surface. This type of high-density indications would not be characterized as geometric flaws with well defined crack-tip geometry that permits high-fidelity application of traditional fracture mechanics methods. An alternative analysis approach, with higher fidelity simulation of this type of “distributed discontinuities”, is proposed, as described in this paper. From a behavioral standpoint, the UT indications at Doel 3 and Tihange 2 represent material discontinuities whose mechanical effect can be evaluated using a damage-mechanics-based constitutive model. Previously, a special multiphase damage model was developed for cladding with zirconium hydrides, of similar morphology to the Doel 3 indications, in which the metal matrix and the hydride platelets are treated as separate material phases interacting at their interfaces with appropriate constraint conditions between them to ensure strain and stress compatibility. The hydride precipitates are represented as a brittle material and the metal matrix is modeled as a ductile elastic-plastic material. This damage model was implemented in a finite element computer program, and was validated using ring-tension and ring-compression tests of cladding specimens with various hydride morphologies. The model was able to predict specimens complete stress-strain curves and failure states with very high accuracy. The above described damage model is adapted to the high-density UT indications, morphology and distribution similar to the conditions of the Doel 3 vessel. The “hydrogen flakes” are characterized in the model as distributed damage of known orientation and volume fraction. A vessel of typical geometry and radiation-dependent mechanical properties is analyzed for various values of volume fraction of hydrogen flakes, and considering a transient loading scenario that conservatively simulates pressurized thermal shock. Interlinking of the “hydrogen flakes” and propagation of damage through the wall under the specified loading condition are part of the model’s capability of directly predicting whether or not vessel failure will occur. Thus, vessel susceptibility to failure and failure margin are judged by the degree of damage propagation through the wall.
Current USA regulations in 10 CFR 50, Appendices G & H ensure adequate fracture toughness and provide for the monitoring of radiation embrittlement of the ferritic components of the reactor pressure vessel (RV). Regulatory Guide (RG) 1.99, Rev. 2 provides guidance on acceptable methods for predicting the effects of neutron irradiation in order to meet the requirements of Appendix G. Specifically, RG 1.99, Rev. 2 provides an embrittlement prediction model for Charpy transition temperature shift (TTS) and a prediction model for decreased Charpy upper shelf energy (USE). The prediction model for USE decrease has remained unchanged since introduction of RG 1.99 in 1975. The objective of this study is to present new USE prediction model(s) developed using an international light water reactor database similar to the effort behind the recently-updated ASTM E900-15 TTS prediction model. A database of ASME and similar specification USE decrease information was developed from USA and select international light water reactor surveillance capsule data, including the latest surveillance capsule fluence, irradiation temperature, material chemistry and other information. The USE database has more than 1,500 USE change measurements of irradiated RV steels. Several best estimation models to predict irradiated USE of materials were developed based on data fitting. Two types of best estimation models were investigated; one model type uses the ASTM E900-15 predicted TTS as a primary input parameter, while the other does not, so that a USE prediction could be made independently of the ASTM E900-15 TTS prediction. By using the ASTM E900-15 TTS as a primary input, the models of the first type implicitly considered the embrittlement mechanisms of matrix damage and copper rich precipitation. In the non-TTS models, the effect of copper was expressed by a hyperbolic tangent curve that has both an upper value and lower value in order to consider the effect of copper saturation. Associated standard deviations as a function of predicted USE were also established so that bounding predictions could be made. Bounding models from each type that conservatively predict irradiated USE by bounding at least 95% of the USE decrease data in the database were identified. These bounding models are estimated to have relatively low impact on the number of USA plants that are projected to have RV steels that drop below 50 ft-lbs (68 J) relative to RG 1.99, Rev. 2. Finally, a non-TTS model was selected as the recommended model, because it does not require calculation of TTS by ASTM E900-15 and thus is simpler to implement.
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