Validation of the ASSERT subchannel code for MAPLE-X10 reactor conditions
- 1. Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.
Description
The ASSERT subchannel analysis code has been developed specifically to model flow and phase distributions within CANDU fuel channels. Recently, ASSERT has been adapted for use in simulating the MAPLE-X10 reactor. ASSERT uses an advanced drift-flux model, which permits the phases to have unequal velocities and unequal temperatures (UVUT), and thus can model non-equilibrium effects such as phase separation tendencies and subcooled boiling. Modelling subcooled boiling accurately is particularly important for MAPLE-X10. This paper briefly summarizes the non-equilibrium model used in the ASSERT code, the equations used to represent these models, and the algorithms used to solve the equations numerically. Very few modifications to the ASSERT models were needed to address MAPLE conditions. These centered on the manner in which finned fuel rods are treated, and they are discussed in the paper. The paper also gives results from validation exercises, in which the ASSERT code predictions of subcooled boiling void-fraction and critical heat flux were compared to experiments using MAPLE-X10 finned fuel elements in annuli and various bundles. 18 refs., 13 figs., 3 tabs
Additional details
Publishing Information
- ISBN
- 0-919784-31-3
- Imprint Title
- Canadian Nuclear Society/American Nuclear Society 4. international conference on simulation methods in nuclear engineering
- Imprint Pagination
- 2 v.
- Journal Page Range
- (v.1) [18 p.].
- Report number
- INIS-mf--14622
Conference
- Title
- 4. international conference on simulation methods in nuclear engineering.
- Dates
- 2-4 Jun 1993.
- Place
- Montreal, PQ (Canada).
INIS
- Country of Publication
- Canada
- Country of Input or Organization
- Canada
- INIS RN
- 26074015
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- A CODES; COMPUTERIZED SIMULATION; CRITICAL HEAT FLUX; FLUID FLOW; FUEL CHANNELS; FUEL ELEMENT CLUSTERS; FUEL ELEMENTS; MAPLE TYPE REACTORS; SUBCOOLED BOILING; VOID FRACTION
- Descriptors DEC
- BOILING; COMPUTER CODES; ENRICHED URANIUM REACTORS; FUEL ASSEMBLIES; HEAT FLUX; HEAVY WATER MODERATED REACTORS; PHASE TRANSFORMATIONS; REACTOR CHANNELS; REACTOR COMPONENTS; REACTORS; RESEARCH AND TEST REACTORS; SIMULATION; WATER COOLED REACTORS; WATER MODERATED REACTORS