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Loss-of-coolant Accident – LOCA

Loss-of-coolant accidents (LOCAs) are postulated accidents that result in a loss of reactor coolant at a rate in excess of the capability of the reactor makeup system from breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the reactor coolant system.

The spectrum of postulated leakage sizes within the reactor coolant pressure boundary has been divided in various ways depending on the selection of the acceptance criteria. Namely:

  • large break LOCAs (LB-LOCAs)
  • small break LOCAs (SB-LOCAs)

LOCA conditions are associated with a rapid decrease in system pressure, cladding ballooning, rupture, and high-temperature steam oxidation. In January 1974, the USNRC published 10 CFR 50.46 establishing acceptance criteria for the ECCSs for LWRs, addressing safety limits that must be assured under LOCA conditions:

  1. Maximum zircaloy cladding temperature (Peak Cladding Temperature),
  2. Maximum oxidation of cladding,
  3. Maximum amount of hydrogen generated by chemical reaction of the zirconium alloy with water and/or steam,
  4. Coolable core geometry,
  5. Long-term cooling.

See also: High-Temperature Steam Oxidation of Zirconium Alloys

Loss of coolant resulting in core dry-out leads to loss of coolability of the core in spite of reactor shutdown; fuel rods are heated, cladding mechanical properties are degraded, and the integrity of the cladding can be lost due to internal fission gas overpressure or thermally induced stresses. If left uncontrolled, the build-up of boric acid due to coolant vaporization in a pressurized water reactor (PWR) could reach precipitation limits and block the coolant channels in the core, preventing adequate heat removal for any size break. At high temperatures, the cladding material reacts with the steam in an exothermic reaction, with hydrogen as a by-product. This reaction represents an additional heat source for the cladding. It can use further degradation of the cladding material due to oxidation and the potential for hydrogen burning or explosions inside the containment. High energy coolant outflow into the containment leads to pressurization of the containment. Containment pressurization, together with high radioactivity in the containment atmosphere (due to fuel dehermetisation), leads to leakages into the environment with potential radiological consequences.

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. It depressurizes the reactor vessel, allowing 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

Acceptance Criteria for LOCAs

In January 1974, the USNRC published 10 CFR 50.46 establishing acceptance criteria for the ECCSs for LWRs, addressing safety limits that must be assured under LOCA conditions:

  • The calculated maximum fuel element cladding temperature shall not exceed 2200°F. This criterion ensures the validity of the ECR criterion.
  • The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation. In other words, this criterion limits maximum Equivalent Cladding Reacted (ECR) to 17% during high-temperature steam oxidation to ensure adequate ductility during the Emergency Core Cooling System (ECCS) quench and possible post-LOCA seismic events.
  • The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounded the fuel, excluding the cladding surrounding the plenum volume, were to react.
  • Calculated changes in core geometry are such that the core remains amenable to cooling.
  • After any calculated successful initial operation of the ECCS, the calculated core temperature is maintained at an acceptably low value, and decay heat is removed for the extended period required by the long-lived radioactivity.

The radiological consequences of the most severe LOCA are within the guidelines of 10 CFR 100 or 10 CFR 50.67.

 
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:

Nuclear Safety