Swelling and Microstructural Evolution in 316 Stainless Steel Hexagonal Ducts Following Long-Term Irradiation in EBR-II

Cole, J. I.; Allen, Todd R.; Tsai, H.; Ukai, Shigeharu; Mizuta, S; Akasaka, Naoaki; Donomae, Takako; Yoshitake, Tsuyoshi

doi:10.1520/stp10547s

Cited by 6 publications

(2 citation statements)

References 8 publications

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“…The material (SA Type 304 SS) and the irradiation temperature (≈370°C) in this study are similar to those of PWR reentrant corners. Various types of precipitates were reported by Cole et al 21 for the CW 316 SS hex can that exhibited low swelling (density change) after irradiation to 23-51 dpa at 376-430°C in EBR-II at dose rates relevant to PWR reentrant corners (see Fig. 4, Section 3.3).…”

Section: Low Swelling and Dense Precipitates In Sa Type 304 Irradiatementioning

confidence: 87%

Assessment of void swelling in austenitic stainless steel PWR core internals.

Chung¹

2006

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As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385°C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of "progressive compounded multiplication" of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions.Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and hence, high swelling at EOL, especially in internal regions of small volume where irradiation temperature is high. However, it is considered unlikely that void swelling in a reentrant corner will exceed the threshold level of ≈4% beyond which the swelling rate reaches the steady state rate of 1%/dpa. However, this estimation is only preliminary, and a more accurate quantification of maximum temperature of reentrant corners at EOL and life-extension situations would be useful.iii Foreword vThis contractor-prepared NUREG-series report on void swelling is the first in a series of reports addressing irradiation effects on the internal structures of pressurized-water reactor vessels. Forthcoming NUREG-series reports will address irradiation embrittlement in cast and wrought stainless steels (SSs), as well as the development and application of models for predicting the effects of irradiation.Void swelling results from the agglomeration of microscopically small voids (or vacancies) produced during neutron irradiation. The primary variables that control...

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Section: Low Swelling and Dense Precipitates In Sa Type 304 Irradiatementioning

confidence: 87%

Assessment of void swelling in austenitic stainless steel PWR core internals.

Chung¹

2006

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“…In addition to the technical issues related to ensuring operation of experimental equipment, there is the expense. These factors point to the necessity of studying structural materials [1][2][3] after their use in nuclear reactors. Attempts are underway to further develop research methods and techniques to study such materials.…”

Section: Introductionmentioning

confidence: 99%

Change in shape and size of fuel assembly duct in BN-350 reactor

Sarsekeyev

Chumakov

Mendebayev

2013

Journal of Nuclear Science and Technology

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The paper presents analytical data obtained by directly measuring samples taken at different levels of a hexagonal fuel assembly duct used in the BN-350 fast reactor. The difference between the length of the deflection arc and the width of an intact duct face displays the accumulated irradiation-creep deformation during the operation period. The deflection of the duct faces induced by the irradiation creep was analyzed, and irradiation-creep rate versus distance from the core center was analyzed for the assembly duct in the BN-350 fast reactor.

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