Initial Results from the High Fluence UCSB ATR-2 Irradiation
- Material Department University of California, Santa Barbara, CA 93106 (United States)
- Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)
- Rolls-Royce plc (United Kingdom)
A primary focus of the Department of Energy, Office of Nuclear Energy Light Water Reactor Sustainability Program (LWRSP) is the safe operation of nuclear reactors to extended lifetimes of up to 80 years. One of the primary reactor components of concern is the reactor pressure vessel (RPV), which is increasingly embrittled over time due to damage from high energy neutrons created in the reactor core. Radiation damage leads to the formation of nm-scale precipitates and solute-defect clusters, which result in an increase in the steel's ductile to brittle transition temperature. Transition temperature shifts (TTS) are used to quantify and regulate embrittlement. Extended life operation will subject some RPVs to higher neutron fluences than previously experienced by in-service vessels. Thus, robust prediction methods must be developed that ensure the safety of our nation's fleet of reactors. Current models underpredict embrittlement data from high flux test reactor irradiations at extended life fluences. It has been hypothesized that one reason for this underprediction is the formation of so-called 'Late- Blooming Phase' Mn-Ni-Si precipitates, which form and grow only at high fluence and that are currently not included in regulatory models. Recent research has demonstrated the existence of high volume fractions of intermetallic Mn-Ni-Si precipitate (MNSP) phases at high fluence and have generated significant insight into their character. However, a number of questions remain regarding MNSPs, such as: 1) what alloys are most susceptible to MNSP formation; 2) how is MNSP formation affected by flux (φ), fluence (φt) and irradiation temperature (T{sub irr}); and, 3) how to accurately predict TTS under extended life low flux service conditions? The fact that no two RPVs are exactly the same, both in terms of the steel composition and vessel service conditions, makes these questions particularly difficult to answer. Further, understanding the effect of flux differences between test reactor irradiations and in-vessel service is absolutely critical. Hence, physical models must be developed that predict TTS as a function of alloy solute content and vessel service conditions. A large-scale irradiation designed to develop and validate an extended life embrittlement prediction model was recently completed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory as part of the DOE sponsored Nuclear Scientific User Facilities. This experiment, called the UCSB ATR-2 irradiation, is a collaboration between Oak Ridge National Lab (ORNL) and the University of California, Santa Barbara (UCSB) and is now primarily supported by the DOE Light Water Reactor Sustainability Program. The experiment is designed to investigate the single and combined effects of neutron φ, φt, T{sub irr} and alloy chemistry on precipitate formation, hardening and embrittlement. While postirradiation examinations (PIE) is ongoing, and is expected to take several years to complete, the first preliminary results are presented below. It should also be noted that Rolls-Royce has been a collaborator in UCSB ATR-2 from the start, and there are several other national and international institutions supporting the PIE. (authors)
- OSTI ID:
- 22992152
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 114, Issue 1; Conference: Annual Meeting of the American Nuclear Society. Embedded topical meeting 'Nuclear fuels and structural material for the next generation nuclear reactors', New Orleans, LA (United States), 12-16 Jun 2016; Other Information: Country of input: France; 4 refs.; Available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 United States; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
42 ENGINEERING
DUCTILE-BRITTLE TRANSITIONS
EMBRITTLEMENT
INTERMETALLIC COMPOUNDS
IRRADIATION
NEUTRON FLUENCE
POST-IRRADIATION EXAMINATION
PRECIPITATION
PRESSURE VESSELS
REACTOR CORES
SAFETY
STEELS
TEST REACTORS
TRANSITION TEMPERATURE
WATER COOLED REACTORS
WATER MODERATED REACTORS