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Title: UCSB ATR-­NSUF Irradiation DMC Sample Inspection Report

Abstract

A variety of tensile samples of Ferritic and Oxide Dispersion Strengthened (ODS or nanoferritic) steels were placed the ATR reactor over 2 Years achieving doses of roughly 4-6 dpa at temperatures of roughly 290°C. After irradiation, samples were shipped from the MFC hot cells at Idaho National Laboratory (INL) to the Wing 9 hot cells in the CMR facility at Los Alamos National Laboratory. Samples were cleaned to removed alpha contamination from the MFC hot cells, and then, as needed removed from their irradiation containers, sorted and inspected. This report will summarize the inspection of the Disc Multipurpose Coupon (DMC) inspection from packet 7-1.

Authors:
 [1];  [1];  [1]
  1. Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Publication Date:
Research Org.:
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
1170627
Report Number(s):
LA-UR-15-20825
DOE Contract Number:
AC52-06NA25396
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
Materials Science(36); Energy Sciences

Citation Formats

Saleh, Tarik A., Quintana, Matthew Estevan, and Romero, Tobias J.. UCSB ATR-­NSUF Irradiation DMC Sample Inspection Report. United States: N. p., 2015. Web. doi:10.2172/1170627.
Saleh, Tarik A., Quintana, Matthew Estevan, & Romero, Tobias J.. UCSB ATR-­NSUF Irradiation DMC Sample Inspection Report. United States. doi:10.2172/1170627.
Saleh, Tarik A., Quintana, Matthew Estevan, and Romero, Tobias J.. Mon . "UCSB ATR-­NSUF Irradiation DMC Sample Inspection Report". United States. doi:10.2172/1170627. https://www.osti.gov/servlets/purl/1170627.
@article{osti_1170627,
title = {UCSB ATR-­NSUF Irradiation DMC Sample Inspection Report},
author = {Saleh, Tarik A. and Quintana, Matthew Estevan and Romero, Tobias J.},
abstractNote = {A variety of tensile samples of Ferritic and Oxide Dispersion Strengthened (ODS or nanoferritic) steels were placed the ATR reactor over 2 Years achieving doses of roughly 4-6 dpa at temperatures of roughly 290°C. After irradiation, samples were shipped from the MFC hot cells at Idaho National Laboratory (INL) to the Wing 9 hot cells in the CMR facility at Los Alamos National Laboratory. Samples were cleaned to removed alpha contamination from the MFC hot cells, and then, as needed removed from their irradiation containers, sorted and inspected. This report will summarize the inspection of the Disc Multipurpose Coupon (DMC) inspection from packet 7-1.},
doi = {10.2172/1170627},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Feb 23 00:00:00 EST 2015},
month = {Mon Feb 23 00:00:00 EST 2015}
}

Technical Report:

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  • New and existing databases will be combined to support development of physically based models of transition temperature shifts (TTS) for high fluence-low flux (φ < 10 11n/cm 2-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1×10 20 n/cm 2 (>1 MeV). All references to neutron flux and fluence in this report are for fast neutrons (>1 MeV). The reactor pressure vessel (RPV) task of the Light Water Reactor Sustainability (LWRS) Program is working with various organizations to obtain archival surveillance materials from commercial nuclear power plants to allow for comparisons of the irradiation-induced microstructural featuresmore » from reactor surveillance materials with those from similar materials irradiated under high flux conditions in test reactors« less
  • The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughnessmore » loss dependent on the radiation sensitivity of the materials. As stated in previous progress reports, the available embrittlement predictive models, e.g. [1], and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.« less
  • Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences thanmore » have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.« less
  • A variety of tensile samples of Ferritic and Oxide Dispersion Strengthened (ODS or nanostructured ferritic) steels were placed the ATR reactor over 2 years achieving doses of roughly 4-6 dpa at temperatures of roughly 290°C. Samples were shipped to Wing 9 in the CMR facility at Los Alamos National Laboratory and imaged then tested in tension. This report summarizes the room temperature tensile tests, the elevated temperature tensile tests (300°C) and fractography and reduction of area calculations on those samples. Additionally small samples were cut from the undeformed grip section of these tensile samples and sent to the NSLS synchrotronmore » for high energy X-ray analysis, initial results will be described here.« less
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