Deterministic Safety Analyses

Deterministic safety analyses for a nuclear power plant predict the response to postulated initiating events. These analyses include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power. A specific set of rules and acceptance criteria is applied. 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 analysed with different computational tools.

Safety Analysis Definitions in NS-R-1, IAEA Safety Standards Series:

“A safety analysis of the plant design, applying methods of Deterministic and Probabilistic analysis, shall be provided which establishes and confirms the design basis for the items important to safety and demonstrates that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category, and that defense in depth has been achieved.”

The aim of the deterministic approach should be to address plant behavior under specific pre-determined operational states and accident conditions, and apply a specific set of rules to judging design adequacy.

Current developments for ensuring the stable, safe and competitive operation of nuclear reactors are closely related to the advances that are being made in safety analysis. Deterministic safety analyses for anticipated operational occurrences, design basis accidents (DBAs) and beyond design basis accidents (BDBAs) are essential instruments for confirming the adequacy of safety provisions.

There are three ways of analysing anticipated operational occurrences and design basis accidents to demonstrate that the safety requirements, which are currently used to support applications for licensing, are met:

  • Use of conservative computer codes with conservative initial and boundary conditions (conservative analysis).
  • Use of best estimate computer codes combined with conservative initial and boundary conditions (combined analysis).
  • Use of best estimate computer codes with conservative and/or realistic input data but coupled with an evaluation of the uncertainties in the calculation results, with account taken of both the uncertainties in the input data and the uncertainties associated with the models in the best estimate computer code (best estimate analysis).

Deterministic safety analyses for design purposes should be characterized by their conservative assumptions and bounding analysis.

For beyond design basis accidents, best estimate calculations are used in several States, together with an evaluation of the uncertainties associated with the relevant phenomena. However, in determining what measures should be taken to mitigate the consequences of beyond design basis accidents, an uncertainty analysis is not usually performed.

Initial and Boundary Conditions

Deterministic safety analyses contain also the key plant parameters considered in the safety evaluation (e.g., core power, core inlet temperature, reactor system pressure, core flow, axial and radial power distribution, fuel and moderator temperature coefficient, void coefficient, reactor kinetics parameters, available shutdown rod worth, and control rod insertion characteristics). These parameters are known as the initial and boundary conditions.

The initial conditions are the assumed values of plant parameters at the start of the transient to be analysed. Examples of these parameters are reactor power level, power distribution, pressure, temperature and flow in the primary circuit.

The boundary conditions are the assumed values of parameters throughout the transient. Examples of boundary conditions are conditions due to the actuation of safety systems such as pumps and power supplies, leading to changes in flow rates, external sources and sinks for mass and energy, and other parameters during the course of the transient.

These parameters must be periodically checked that the range of values for plant parameters is representative of fuel exposure or core reload, and that the range is sufficiently broad to cover the predicted fuel cycle ranges, to the extent practicable, based on the fuel design and acceptable analytical methodology.

Conservative Analysis

A conservative approach usually means that any parameter that has to be specified for the analysis should be allocated a value that will have an unfavourable effect in relation to specific acceptance criteria.

Conservative approach has moved towards a more realistic approach with an evaluation of uncertainties. (due to more experimental data and advances in computer code; best estimate approach)

Best Estimate Analysis

Provides more realistic information about the physical behavior of the reactor, identifies the most relevant safety issues and provides information about the existing margins between the results of calculations and the acceptance criteria.

Acceptance Criteria

Acceptance Criteria for AOOs

For evaluation of deterministic safety analyses, a specific set of rules and acceptance criteria is applied. 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 analysed 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 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, then 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 result 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.

Acceptance Criteria for DBAs

For evaluation of deterministic safety analyses, a specific set of rules and acceptance criteria is applied. 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 analysed with different computational tools. Unlike an AOO, a postulated accident could result in sufficient damage to preclude resumption of plant operation.

A list of the basic criteria necessary to meet the requirements of GDC for postulated accidents appears below.

  • Pressure in the reactor coolant and main steam systems should be maintained below specific value, considering potential brittle as well as ductile failures.
  • 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, then 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 result in a consequential loss of function of the RCS or reactor containment barriers.
  • The release of radioactive material shall not result in offsite doses in excess of specific limits.
  • A postulated accident shall not, by itself, cause a consequential loss of required functions of systems needed to cope with the fault, including those of the RCS and the reactor containment system.

For loss-of-coolant accidents (LOCAs), the following analysis acceptance criteria of 10 CFR 50.46 also apply:

  • The calculated maximum fuel element cladding temperature shall not exceed 2200°F. This criterion ensures validity of 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 during 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 surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
  • Calculated changes in core geometry shall be such that the core remains amenable to cooling.
  • After any calculated successful initial operation of the emergency core cooling system (ECCS), the calculated core temperature shall should be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

For reactivity-initiated accidents (RIAs), 10 CFR 50 Appendix A, General Design Criterion 28 (GDC28) requires the reactivity control system to be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither:

  • Result in damage to the reactor coolant pressure boundary greater than limited local yielding, nor
  • Sufficiently impair core cooling capability.

Reduction of coolability can result from violent expulsion of fuel, which could damage nearby fuel assemblies. In past, the core coolability criteria was revised to specifically address both short-term (e.g., fuel-to-coolant interaction, rod burst) and long-term (e.g., fuel rod ballooning, flow blockage) phenomena which challenge coolable geometry and reactor pressure boundary integrity. A definite limit for core damage, which must not be exceeded at any position in any fuel rod in the core. According to Appendix B of the Standard Review Plan, Section 4.2, these criteria are, for example:

  • Peak radial average fuel enthalpy must remain below 230 cal/g. Above this enthalpy, hot fuel particles might be expelled from a fuel rod.
  • Peak fuel temperature must remain below incipient fuel melting conditions.
 
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