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Control Rod Drop – PWR

A control rod drop event is one of the possible control rod malfunction events, and it belongs to reactivity-initiated events usually described in Chapter 15.4. of the Safety Analysis Report (according to the NUREG-0800).

As a consequence of any of these events, there is a distortion in the core power distribution with a potential reduction of DNBR. For a CR withdrawal, there is also a global reactor power increase, which is reduced later by the reactor power control. A potentially relevant safety aspect comes from the case when a rod drops into the core, and the control system is in automatic mode. If the dropped control rod does not result in the actuation of the reactor trip, then the reactor power may be reestablished by the control system. In this case, the rods will be moved out to compensate for the sudden power decrease. Before achieving a new equilibrium power, a transient overshoot on nuclear power can be expected, coincident with a significant distortion in radial power distribution caused by the dropped rod. High local peaking factors and an overshoot in power may violate the limits on fuel power density.

The magnitude of power deviation is primarily a function of the control rod worth, reactivity coefficients, and core characteristics. In this event, it must be shown that the fuel and the fuel-clad integrity are not challenged.

Acceptance Criteria for AOOs

A specific set of rules and acceptance criteria is applied for the evaluation of deterministic safety analyses. Acceptance criteria are used in deterministic safety analysis to assist in judging the acceptability of the results of the analysis as a demonstration of the safety of the nuclear power plant. Typically, these should focus on neutronic, thermohydraulic, radiological, thermomechanical, and structural aspects, which are often analyzed with different computational tools.

The following are the specific criteria necessary to meet the requirements of GDC for AOOs:

  • Pressure in the reactor coolant and main steam systems should be maintained below a specific value (usually below 110% of the design pressure).
  • Fuel cladding integrity shall be maintained by ensuring that the minimum departure from nucleate boiling ratio (DNBR) remains above the 95/95 DNBR limit for PWRs (a 95% probability at a 95% confidence level) and that the critical power ratio (CPR) remains above the minimum critical power ratio (MCPR) safety limit for BWRs. If the minimum DNBR or MCPR does not meet these limits, the fuel is assumed to have failed.
  • According to 10 CFR 50.59, an AOO should not generate a postulated accident without other faults occurring independently or resulting in a consequential loss of function of the RCS or reactor containment barriers.

By meeting these criteria, it can be demonstrated that automatic functions and control systems can return the facility to its normal operating mode as soon as possible, and it can be demonstrated that all barriers remained intact after the event.

 
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