ASSERT/NUCIRC commissioning for CANDU 6 fuel channel CCP analysis
Creators
- 1. Atomic Energy of Canada Ltd., Sheridan Park, ON (Canada). CANDU Operations
- 2. Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.
- 3. Hydro-Quebec, Montreal, PQ (Canada)
- 4. New Brunswick Power, Fredericton, NB (Canada)
Description
CANDU PHWR fuel channel pressure tubes will expand or creep under long-term (aging process) influence of temperature, pressure, and neutron flux. This diametral pressure tube creep will influence the critical channel power (CCP), or conditions that lead to dryout. In order to provide safety analysis models to quantify the effect of diametral pressure tube creep on CCP, a COG (AECL/NBP/HQ) project is underway to commission the ASSERT and NUCIRC codes to establish reliable production tools for the assessment of CANDU6 CCP in nominal (uncrept) and crept pressure tube fuel channels. This paper gives an overview of the background and objectives of the project along with a brief introduction into the subchannel analysis code ASSERT and the 1-D thermalhydraulics code NUCIRC. This project is a multistage endeavour, for which the first stage results are presented. A detailed cross-comparison of the 1-D (NUCIRC) and subchannel (ASSERT) models of pressure drop (ΔP) and critical heat flux (CHF) has been undertaken and has led to several enhancements and refinements to the respective models. These results are presented in addition to results of ASSERT commissioning against NUCIRC for a matrix of ΔP and dryout cases in a nominal pressure tube, which are based upon Gentilly 2 and Point Lepreau site area. Additionally, the initial results of an assessment, using ASSERT, of the effects of creep on ΔP are presented. In concluding, the status and future directions for ASSERT/NUCIRC CANDU 6 CCP analysis project are summarized. (author). 2 refs., 12 figs
Additional details
Publishing Information
- Imprint Title
- Proceedings of the 1994 nuclear simulation symposium
- Imprint Pagination
- 290 p.
- Journal Page Range
- p. 129-138.
- Report number
- INIS-mf--14840
Conference
- Title
- 1994 nuclear simulation symposium.
- Dates
- 12-14 Oct 1994.
- Place
- Pembroke, ON (Canada).
INIS
- Country of Publication
- Canada
- Country of Input or Organization
- Canada
- INIS RN
- 27066225
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- A CODES; CANDU TYPE REACTORS; COMPUTERIZED SIMULATION; CREEP; CRITICAL HEAT FLUX; DRYOUT; FLUID FLOW; FUEL CHANNELS; FUEL ELEMENT CLUSTERS; N CODES; PRESSURE TUBES; PRIMARY COOLANT CIRCUITS; REACTOR SAFETY; REACTOR SAFETY EXPERIMENTS
- Descriptors DEC
- COMPUTER CODES; COOLING SYSTEMS; FUEL ASSEMBLIES; HEAT FLUX; HEAVY WATER MODERATED REACTORS; MECHANICAL PROPERTIES; POWER REACTORS; PRESSURE TUBE REACTORS; REACTOR CHANNELS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; SAFETY; SIMULATION; THERMAL REACTORS; TUBES