Published 2016 | Version v1
Journal article

Potential Impact of Cladding Wettability on LWR Transient Progression

  • 1. Oak Ridge National Laboratory, 1 Bethel Valley Rd, Oak Ridge, TN, 37831 (United States)
  • 2. Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

Description

The US Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is working closely with the nuclear industry to develop fuel and cladding candidates with potentially enhanced accident tolerance. This type of fuel is also known as accident tolerant fuel (ATF). Two promising cladding material candidates with potentially enhanced accident tolerance are silicon carbide (SiC) and an iron-chromium-aluminum (FeCrAl) alloy. The FeCrAl alloy features significantly enhanced oxidation resistance versus the Zr-based alloys currently used in state-of-the-art light water reactors (LWRs). Thermal-fluid characteristics are a vital element of a holistic engineering evaluation of ATF concepts. One characteristic related to boiling heat transfer is the critical heat flux (CHF). At CHF, the efficacy of heat transfer rapidly degrades and components can heat up rapidly. CHF is the heat flux at which the departure from nucleate boiling (DNB) occurs in PWRs. DNB is directly relevant to the departure from nuclear boiling ratio (DNBR), which is the ratio between the critical heat flux and the actual heat flux. Surface wettability is an important factor that influences CHF. CHF plays a significant role in determining safety margins during normal operation and also in the progression of potential transient or accident scenarios. Significant changes in CHF properties and boiling heat transfer characteristics will impact the progression of transient and accident events and reactor safety characteristics. Additionally, the behavior of the candidate cladding materials post-CHF impacts reactor safety. A brief description of the relationship between cladding material properties and CHF is provided, followed by an analysis of two hypothetical transient scenarios: a super-prompt reactivity-initiated accident (RIA) and an anticipated transient without scram with instability (ATWS-I). The analysis illustrates the potential impact of cladding material properties on safety margins via direct alteration of calculated CHF values. (authors)

Additional details

Publishing Information

Journal Title
Transactions of the American Nuclear Society
Journal Volume
115
Journal Page Range
p. 473-477
ISSN
0003-018X

Conference

Title
2016 ANS Winter Meeting and Nuclear Technology Expo
Dates
6-10 Nov 2016
Place
Las Vegas, NV (United States)

Optional Information

Notes
9 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US)