TAILIEUCHUNG - Safety analysis methodology for aged candu nuclear reactors

This paper deals with the Safety Analysis for CANDU® 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. | http SAFETY ANALYSIS METHODOLOGY FOR AGED CANDU 6 NUCLEAR REACTORS WOLFGANG HARTMANN1 and JONG YEOB JUNG2 1CANTECH Associates Ltd 4525 Lakeshore Road Burlington Ontario L7L 1B3 Canada 2Korea Atomic Energy Research Institute 989-111 Daedeok-daero Yuseong-gu Daejeon Korea Corresponding author. E-mail hartmannw@ Received August 18 2013 This paper deals with the Safety Analysis for CANDU 6 nuclear reactors as affected by main Heat Transport System HTS aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation LOR analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years EFPY aged core was summarized leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred. KEYWORDS Safety Analysis CANDU Aging Accidents Predictions Methodology 1. INTRODUCTION All industrial plants undergo changes over time and nuclear plants are no exception. The CANDU 6 reactor follows earlier .

• Năng lượng
• Safety analysis methodology for aged candu nuclear reactors
• Safety analysis methodology
• Slow Loss of Regulation
• Heat Transport System
• The thermal hydraulic analysis