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)
INIS
- Country of Publication
- United States
- Country of Input or Organization
- France
- INIS RN
- 52083014
- Subject category
- S11: NUCLEAR FUEL CYCLE AND FUEL MATERIALS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- ACCIDENT-TOLERANT NUCLEAR FUELS; CHROMIUM; CLADDING; CRITICAL HEAT FLUX; DEPARTURE NUCLEATE BOILING; FLUIDS; HEAT TRANSFER; INSTABILITY; MAGNETIC SURFACES; NUCLEAR ENERGY; NUCLEAR INDUSTRY; OXIDATION; PWR TYPE REACTORS; REACTIVITY-INITIATED ACCIDENTS; REACTOR SAFETY; SAFETY MARGINS; SILICON CARBIDES; WETTABILITY
- Descriptors DEC
- ACCIDENTS; BOILING; CARBIDES; CARBON COMPOUNDS; CHEMICAL REACTIONS; DEPOSITION; ELEMENTS; ENERGY; ENERGY SOURCES; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; FUELS; HEAT FLUX; INDUSTRY; MAGNETIC FIELD CONFIGURATIONS; MATERIALS; METALS; NUCLEAR FUELS; NUCLEATE BOILING; PHASE TRANSFORMATIONS; POWER REACTORS; REACTOR ACCIDENTS; REACTOR MATERIALS; REACTORS; SAFETY; SILICON COMPOUNDS; SURFACE COATING; THERMAL REACTORS; TRANSITION ELEMENTS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 9 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US)