Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
250 The Use of Probabilistic Safety Techniques for Evaluating the Advanced CANDU Reactor (ACR) (PSAM-0399)
Download citation file:
- Ris (Zotero)
- Reference Manager
The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications for reactor facilities that are significantly advanced beyond the current generation of operating reactors. One proposed reactor design, developed by Atomic Energy of Canada, Limited (AECL), is an Advanced CANada Deuterium Uranium (CANDU) Reactor (ACR), the ACR-700. The ACR is an enhanced version of earlier CANDU designs. However, unlike the CANDU reactors, which are heavy-water cooled and moderated reactors, the ACR-700 is a light-water cooled and heavy-water moderated reactor. In preparation of a possible design certification review, the NRC (with the assistance of ORNL) began examining selected areas of nuclear safety, identifying accidents that could potentially dominate the risk profile of the ACR-700 design, and evaluating other risk-important design and technology issues. This effort supports the NRC's policy that encourages the use of probabilistic risk assessment (PRA) in all regulatory matters. In addition to identifying potential initiating events and systems judged to be “important” to preventing and mitigating possible accident conditions, a prototype risk evaluation model for the ACR-700 was developed using the SAPHIRE computer code.