Facebook Instagram Youtube Twitter

Emergency Core Cooling System – ECCS

The purpose of the Emergency Core Cooling Systems (ECCS) aims to provide core cooling under loss-of-coolant accident (LOCA) conditions to limit fuel cladding damage. The ECCS limits the fuel cladding temperature below the limit so that the core will remain intact and in place, with its essential heat transfer geometry preserved. The Code of Federal Regulations, CFR, requires the ECCS to be designed so that after any LOCA, the reactor core remains in a geometrical configuration amenable to cooling. The basic criteria are limiting fuel cladding temperature and oxidation to minimize clad fragmentation and to minimize the hydrogen generation from clad oxidation to protect the containment.

The ECCS usually consists of redundant high-pressure systems (e.g.,3×100%) and redundant low-pressure systems (e.g.,3×100%).

  • HPCI. The high-pressure systems are the High-Pressure Coolant Injection system (HPCI) and the Automatic Depressurization system (ADS). The HPCI system maintains adequate reactor vessel water inventory for core cooling on small break LOCAs and depressurizes the reactor vessel to allow the low-pressure ECCS to inject on intermediate-break LOCAs.
  • LPCI. The low-pressure systems are the Low-Pressure Coolant Injection (LPCI) made of the Residual Heat Removal (RHR) system and the Core Spray (CS) system. The LPCI is an emergency system that consists of a pump that injects a coolant into the reactor vessel once it has been depressurized. The CS system (typical for BWRs) provides spray cooling to the reactor core to help mitigate the consequences of the large-break LOCAs when reactor pressure is low enough for the system to inject water into the reactor vessel. For low pressures, the accumulator injection system is also available. The accumulators are independent tanks containing borated coolant stored under nitrogen gas at a given pressure.

See also: Decay Heat Removal

 
References:
Nuclear and Reactor Physics:
  1. J. R. Lamarsh, Introduction to Nuclear Reactor Theory, 2nd ed., Addison-Wesley, Reading, MA (1983).
  2. J. R. Lamarsh, A. J. Baratta, Introduction to Nuclear Engineering, 3d ed., Prentice-Hall, 2001, ISBN: 0-201-82498-1.
  3. W. M. Stacey, Nuclear Reactor Physics, John Wiley & Sons, 2001, ISBN: 0- 471-39127-1.
  4. Glasstone, Sesonske. Nuclear Reactor Engineering: Reactor Systems Engineering, Springer; 4th edition, 1994, ISBN: 978-0412985317
  5. W.S.C. Williams. Nuclear and Particle Physics. Clarendon Press; 1 edition, 1991, ISBN: 978-0198520467
  6. G.R.Keepin. Physics of Nuclear Kinetics. Addison-Wesley Pub. Co; 1st edition, 1965
  7. Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.
  8. U.S. Department of Energy, Nuclear Physics and Reactor Theory. DOE Fundamentals Handbook, Volume 1 and 2. January 1993.

Nuclear Safety:

  1. IAEA Safety Standards, Safety of Nuclear Power Plants: Design, SSR-2/1 (Rev. 1). VIENNA, 2016.
  2. IAEA Safety Standards, Safety of Nuclear Power Plants: Commissioning and Operation, SSR-2/2 (Rev. 1). VIENNA, 2016.
  3. IAEA Safety Standards, Deterministic Safety Analysis for Nuclear Power Plants, SSG-2 (Rev. 1). VIENNA, 2019.
  4. IAEA TECDOC SERIES, Considerations on the Application of the IAEA Safety Requirements for the Design of Nuclear Power Plants, IAEA-TECDOC-1791. VIENNA, 2016.
  5. Safety Reports Series, Accident Analysis for Nuclear Power Plants with Pressurized Water Reactors. ISBN 92–0–110603–3. VIENNA, 2003.
  6. Appendix A to 10 CFR Part 50, “General Design Criteria for Nuclear Plants.”
  7. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition.
  8. Nuclear Power Reactor Core Melt Accidents, Science and Technology Series. IRSN – Institute for Radiological Protection and Nuclear Safety. ISBN: 978-2-7598-1835-8
  9. ANSI ANS 51.1: Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, 1983.

See above:

Safety Systems