4 Matching Results

Search Results

Advanced search parameters have been applied.

Advanced Models of LWR Pressure Vessel Embrittlement for Low Flux-HighFluence Conditions

Description: Neutron embrittlement of reactor pressure vessels (RPVs) is an unresolved issue for light water reactor life extension, especially since transition temperature shifts (TTS) must be predicted for high 80-year fluence levels up to approximately 1,020 n/cm{sup 2}, far beyond the current surveillance database. Unfortunately, TTS may accelerate at high fluence, and may be further amplified by the formation of late blooming phases that result in severe embrittlement even in low-copper (Cu) steels. Embrittlement by this mechanism is a potentially significant degradation phenomenon that is not predicted by current regulatory models. This project will focus on accurately predicting transition temperature shifts at high fluence using advanced physically based, empirically validated and calibrated models. A major challenge is to develop models that can adjust test reactor data to account for flux effects. Since transition temperature shifts depend on synergistic combinations of many variables, flux-effects cannot be treated in isolation. The best current models systematically and significantly under-predict transition temperature at high fluence, although predominantly for irradiations at much higher flux than actual RPV service. This project will integrate surveillance, test reactor and mechanism data with advanced models to address a number of outstanding RPV embrittlement issues. The effort will include developing new databases and preliminary models of flux effects for irradiation conditions ranging from very low (e.g., boiling water reactor) to high (e.g., accelerated test reactor). The team will also develop a database and physical models to help predict the conditions for the formation of Mn-Ni-Si late blooming phases and to guide future efforts to fully resolve this issue. Researchers will carry out other tasks on a best-effort basis, including prediction of transition temperature shift attenuation through the vessel wall, remediation of embrittlement by annealing, and fracture toughness master curve issues.
Date: June 17, 2013
Creator: Odette, G. Robert & Yamamoto, Takuya
Partner: UNT Libraries Government Documents Department

Development of High-Temperature Ferritic Alloys and Performance Prediction Methods for Advanced Fission Energy Systems

Description: Reports the results of a comprehensive development and analysis of a database on irradiation hardening and embrittlement of tempered martensitic steels (TMS). Alloy specific quantitative semi-empirical models were derived for the dpa dose, irradiation temperature (ti) and test (Tt) temperature of yield stress hardening (or softening) .
Date: August 14, 2009
Creator: Odette, G. RObert & Yamamoto, Takuya
Partner: UNT Libraries Government Documents Department

He Transport and Fate of Tempered Martensitic Steels: Summary of Recent TEM Observations

Description: As an extension of prior work [1-4], we summarize recent observations made on a He-implanted tempered martensitic steel (TMS), F82H mod 3, irradiated in the HFIR, in both as-tempered (AT) and cold-worked (CW) conditions. A novel implantation technique was used to uniformly inject He into 3-mm diameter TEM discs to depths ranging from ≈ 5-8 µm. The He is generated by two-step transmutation reactions in Ni contained in a NiAl coating layer adjacent to paired 3 mm TEM discs. NiAl layers from 1 to 4 μm thick produced He/dpa ratios between 5 and 40 appm/dpa. The irradiations were at temperatures of 300, 400 and 500°C from 3.9 to 9 dpa and 90 to 380 appm He. Electron transparent samples were prepared by a cross-sectional thinning technique that allowed investigating microstructural evolution over a range of implantation depths. Irradiation of the AT alloy to 9 dpa at 500°C and 380 appm He resulted in relatively large, faceted cavities, that are likely voids, along with a much higher density of smaller He bubbles. The cavities were most often aligned in pearl necklace like strings, presumably due to their formation on pre-existing dislocations. A finer distribution of cavities was also present on precipitate interfaces, lath and grain boundaries. Nine dpa irradiations that produced 190 appm He resulted in a somewhat more random distribution and lower density of smaller matrix cavities; but lower He levels had a less noticeable effect on bubbles in the lath and precipitate boundaries. Corresponding irradiations of the CW F82H produced a larger number of smaller cavities. Irradiation of the AT alloy to 3.9 dpa and 90 ppm He at 400°C produced a similar cavity population to that observed at 500°C at 190 appm He, while the corresponding cavities at 500°C are slightly larger and more numerous at 380 appm ...
Date: February 26, 2010
Creator: Edwards, Danny J.; Kurtz, Richard J.; Odette, G Robert & Yamamoto, Takuya
Partner: UNT Libraries Government Documents Department

A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels

Description: The reactor pressure vessels (RPVs) of commercial nuclear power plants are subject to embrittlement due to exposure to high-energy neutrons from the core, which causes changes in material toughness properties that increase with radiation exposure and are affected by many variables. Irradiation embrittlement of RPV beltline materials is currently evaluated using Regulatory Guide 1.99 Revision 2 (RG1.99/2), which presents methods for estimating the shift in Charpy transition temperature at 30 ft-lb (TTS) and the drop in Charpy upper shelf energy (ΔUSE). The purpose of the work reported here is to improve on the TTS correlation model in RG1.99/2 using the broader database now available and current understanding of embrittlement mechanisms. The USE database and models have not been updated since the publication of NUREG/CR-6551 and, therefore, are not discussed in this report. The revised embrittlement shift model is calibrated and validated on a substantially larger, better-balanced database compared to prior models, including over five times the amount of data used to develop RG1.99/2. It also contains about 27% more data than the most recent update to the surveillance shift database, in 2000. The key areas expanded in the current database relative to the database available in 2000 are low-flux, low-copper, and long-time, high-fluence exposures, all areas that were previously relatively sparse. All old and new surveillance data were reviewed for completeness, duplicates, and discrepancies in cooperation with the American Society for Testing and Materials (ASTM) Subcommittee E10.02 on Radiation Effects in Structural Materials. In the present modeling effort, a 10% random sample of data was reserved from the fitting process, and most aspects of the model were validated with that sample as well as other data not used in calibration. The model is a hybrid, incorporating both physically motivated features and empirical calibration to the U.S. power reactor surveillance data. ...
Date: November 1, 2007
Creator: Eason, Ernest D.; Odette, George Robert; Nanstad, Randy K & Yamamoto, Takuya
Partner: UNT Libraries Government Documents Department