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Postulated Initiating Events

A postulated initiating event, or PIE, is defined as an “identified event that leads to anticipated operational occurrences or accident conditions and its consequential failure effect.”

For certain plant designs, the postulated initiating events shall be identified based on engineering judgment and a combination of deterministic and probabilistic assessments.

Postulated initiating events shall be identified and grouped based on their frequency of occurrence at the nuclear power plant. Therefore, there are two groups of PIEs:

  • Anticipated operational occurrences. Anticipated operational occurrences, AOOs,  refer to the events that are categorized in Regulatory Guide 1.70 and in Regulatory Guide 1.206 as incidents of moderate frequency (i.e., events that are expected to occur several times during the plant’s lifetime) and infrequent events (i.e., events that may occur during the lifetime of the plant). In case of anticipated operational occurrences, the objective is to demonstrate that automatic functions and control systems can return the facility to its normal operating mode as soon as possible and to demonstrate that all barriers remain intact after the event. AOOs are also known as Condition II and III events, respectively, in the commonly used, oft-cited but unofficial American Nuclear Society (ANS) standards.
  • Postulated accidents (or design basis accidents). Postulated accidents are unanticipated conditions of operation (i.e., not expected to occur during the life of the nuclear power unit), but they cannot be excluded. Postulated accidents are also known as Condition III and IV events. Design bases accident is a postulated accidents in which a nuclear facility must be designed and built to withstand without losing the systems, structures, and components necessary to ensure public health and safety.

An analysis of the postulated initiating events for the plant shall be made to establish the preventive and protective measures necessary to ensure that the required safety functions will be performed.

Categorization of Postulated Initiating Events

AOOs and postulated accidents are also categorized according to type. The type of AOO or postulated accident is defined by its effect on the plant. For example, one type of AOO or postulated accident will cause the RCS to pressurize and possibly jeopardize RCS integrity. Another type will cause the RCS to depressurize and possibly jeopardize fuel cladding integrity. It is useful to categorize and organize analyses of AOOs and postulate accidents according to type so that analysts can compare them on common bases, effects, and safety limits. Such comparisons can help to identify limiting events and cases for detailed examination and eliminate nonlimiting cases from further consideration.

AOOs and design bases accidents can be grouped into the following seven types:

  • Increase in heat removal by the secondary system
    • e.g., inadvertent moderator cooldown (PWR and BWR – AOO)
    • e.g., steam line break event (PWR – DBA)
  • Decrease in heat removal by the secondary system
    • e.g., loss of normal feedwater (PWR – AOO)
    • e.g., reactor-turbine load mismatch, including loss of load and turbine trip (PWR and BWR – AOO)
  • Decrease in RCS flow rate
    • e.g., loss or interruption of core coolant flow, excluding reactor coolant pump locked rotor (PWR – AOO)
    • e.g., single reactor coolant pump locked rotor (PWR – DBA)
    • e.g., seizure of one recirculation pump (BWR – DBA)
  • Reactivity and power distribution anomalies (i.e., RIA)
    • e.g., control rod drop (PWR – AOO)
    • e.g., inadvertent chemical shim dilution (PWR – AOO)
    • e.g., ejection of a control rod assembly (PWR – DBA)
    • e.g., control rod drop accident (BWR – DBA)
  • Increase in reactor coolant inventory
    • e.inadvertent operation of emergency core cooling
  • Decrease in reactor coolant inventory
    • e.g., minor reactor coolant system (RCS) leak or loss of reactor coolant such as from a small ruptured pipe or a crack in a large pipe (PWR and BWR)
    • e.g., loss-of-coolant accident (LOCA – DBA)
  • Radioactive release from a subsystem or component

Safety analyses of these AOOs and postulated accident analyses can (and should) encompass a variety of cases, each designed to produce effects or results that challenge designated safety limits. For example, one case study of the turbine trip event is usually designed (by initial and boundary conditions) to yield a high peak RCS pressure, and another case study of the same AOO can be designed to yield a low, minimum thermal margin.

See also NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition.

 

 
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