Safety of Nuclear Reactors (Sustainability Assessment)

INPRO basic principle (BP) for sustainability assessment in the area of nuclear reactor safety - The safety of the planned nuclear installation is superior to that of the reference nuclear installation such that the frequencies and consequences of the accidents are greatly reduced. In the event of an accident, off-site releases of radionuclides are prevented or mitigated so that there will be no need for public evacuation.

Introduction

Objective

This volume of the updated INPRO manual for sustainability assessment provides guidance to the assessor of a planned NES (or a nuclear reactor) on how to apply the INPRO methodology for sustainability assessment in the area of safety of nuclear reactors. The INPRO assessment is expected either to confirm the fulfilment of all INPRO methodology criteria in the area of reactor safety, or to identify which criteria are not fulfilled and note the corrective actions (including potential RD&D) that would be necessary to fulfil them.
This publication discusses the INPRO sustainability assessment method for the area of safety of nuclear reactors. The INPRO sustainability assessment method for safety of nuclear fuel cycle facilities is discussed in a separate report of the INPRO manual.
This publication is intended for use by organizations involved in the development and deployment of a NES including planning, design, modification, technical support and operation for nuclear power plants. The INPRO assessor (or a team of assessors) is assumed to be knowledgeable in the area of nuclear safety and/or may be using the support of qualified organizations (e.g. the IAEA) with relevant experience. Two general types of assessors can be distinguished: a nuclear technology holder (i.e. a designer, developer or supplier of nuclear technology), and a (potential) user of such technology. The current version of the manual includes a number of explanations, discussions, examples and details so it is deemed to be used by technology holders and technology users.

Scope

The INPRO methodology presented in this manual is internationally developed guidance for assessing NES sustainability and is intended for use in support of NES planning studies by focusing on selected areas of reactor safety that are important for public acceptance (see Chapter 2). This manual deals with the long term sustainability of a NES comprised of different types of nuclear reactors. The INPRO methodology user requirements and criteria for sustainability assessment are formulated in this manual in a generic manner to make them applicable to both evolutionary and innovative reactors based on different technologies. However, the major contributions to the INPRO methodology update project have been obtained from the INPRO assessments of evolutionary water-cooled reactors and sodium cooled fast reactors. Other types of innovative reactors with a lower level of design maturity may require modifications or clarifications of selected criteria. Such potential changes will be considered in future revisions of the INPRO methodology after sufficient experience has accrued from INPRO assessments of such reactors.
This manual does not establish any specific safety requirements, recommendations or guidance. IAEA safety requirements and guidance are only issued in the IAEA Safety Standards Series. Therefore, the basic principles, user requirements and associated criteria contained in the INPRO methodology should only be used for sustainability assessments. The INPRO methodology is typically used by Member States in conducting a self-assessment of the sustainability and sustainable development of nuclear energy systems. This manual should not be used for formal or authoritative safety assessments or safety analyses to address compliance with the IAEA Safety Standards or for any national regulatory purpose associated with the licensing or certification of nuclear facilities, technologies or activities.
In the current version of the INPRO methodology, the sustainability issues relevant to safety of reactors and safety of nuclear fuel cycle facilities (NFCFs) are considered in separate manuals. The current methodology does not specifically address innovative integrated system designs (e.g. molten salt reactors with liquid fuel and integrated fast reactors with metallic fuel) whose reactors are combined or co-located with fuel fabrication and/or reprocessing facilities. Reactor and NFCF installations of such integrated systems should be assessed separately against corresponding criteria in the INPRO areas of reactor safety and safety of NFCFs . When more detailed information on the safety issues in integrated systems has been acquired, this approach can be changed in the next revisions of the INPRO methodology.
This version of the INPRO methodology manual for the area of reactor safety is focused on those nuclear power plants that produce primarily electricity, heat and combinations of the two . This publication does not explicitly consider safety issues related to other non-electric applications (hydrogen production, desalination, etc.) or to cogeneration involving such energy products. It is expected that as more detailed information is acquired on the interactions between a reactor and industrial facilities located on the same site, the INPRO criteria may be modified when the methodology is next revised.

Structure

This publication follows the relationship between the concept of sustainable development and different INPRO methodology areas. Section 2 describes the linkage between the United Nations Brundtland Commission’s concept of sustainable development and the IAEA’s INPRO methodology for assessing the sustainability of planned and evolving NESs. Section 2 also considers how the INPRO sustainability assessment methodology in the area of reactor safety relates to the DID concept. Section 3 identifies the necessary inputs for an INPRO assessment in the area of reactor safety. This includes information on design and safety analyses for the planned reactor and for the reference design. Section 4 presents the rationale and background for the INPRO sustainability assessment methodology in the area of reactor safety in terms of the selected basic principle, user requirements and assessment criteria, which consist of indicators and acceptance limits. On the criterion level, guidance is provided on how to determine the values of the indicators and acceptance limits, i.e. how to assess the potential of a NES to fulfil the INPRO methodology criteria. Appendix I presents a list of potential reference reactor designs to be used in the INPRO assessment. Appendices II through X provide complementary information which can be useful for the INPRO assessment of NES against different criteria discussed in the report. Table 1 provides an overview of the INPRO user requirements and criteria that stem from the INPRO basic principle for sustainability assessment in the area of reactor safety.

Table 1. Overview of the INPRO methodology for sustainability assessment in the area of safety of nuclear reactors

INPRO basic principle for sustainability assessment in the area of safety of nuclear reactors: The safety of the planned nuclear installation is superior to that of the reference nuclear installation such that the frequencies and consequences of the accidents are greatly reduced. In the event of an accident, off-site releases of radionuclides are prevented or mitigated so that there will be no need for public evacuation.
INPRO user requirements	Criteria	Indicator (IN) and Acceptance Limit (AL)
UR1: Robustness of design during normal operation: The nuclear reactor assessed is more robust than a reference design with regard to operation and systems, structures and components failures.	CR1.1: Design of normal operation systems	IN1.1: Robustness of design of normal operation systems.
		AL1.1: More robust than that in the reference design.
	CR1.2: Reactor performance	IN1.2: Reactor performance attributes.
		AL1.2: Superior to those of the reference design.
	CR1.3: Inspection, testing and maintenance	IN1.3: Capabilities to inspect, test and maintain.
		AL1.3: Superior to those in the reference design.
	CR1.4: Failures and deviations from normal operation	IN1.4: Expected frequency of failures and deviations from normal operation.
		AL1.4: Lower than that in the reference design.
	CR1.5: Occupational dose	IN1.5: Occupational dose values during normal operation and AOOs.
		AL1.5: Lower than the dose constraints.
UR2: Detection and interception of AOOs: The nuclear reactor assessed has improved capabilities to detect and intercept deviations from normal operational states in order to prevent AOOs from escalating to accident conditions.	CR2.1: Instrumentation and control (I&C) system and inherent characteristics	IN2.1: Capabilities of the I&C system to detect and intercept and/or capabilities of the reactor’s inherent characteristics to compensate for deviations from normal operational states.
		AL2.1: Superior to those in the reference design.
	CR2.2: Grace periods after AOOs	IN2.2: Grace periods until human actions are required after AOOs.
		AL2.2: Longer than those in the reference design.
	CR2.3: Inertia	IN2.3: Inertia to cope with transients.
		AL2.3: Larger than that in the reference design.
UR3: Design basis accidents (DBAs): The frequency of occurrence of DBAs in the nuclear reactor assessed is reduced. If an accident occurs, engineered safety features are able to restore the reactor to a controlled state, and subsequently to a safe shutdown state, and ensure the confinement of radioactive material. Reliance on human intervention is minimal, and only required after a sufficient grace period.	CR3.1: Frequency of DBAs	IN3.1: Calculated frequencies of occurrence of DBAs.
		AL3.1: Frequencies of DBAs that can cause plant damage are lower than those in the reference design.
	CR3.2: Grace period for DBAs	IN3.2: Grace periods for DBAs until human intervention is necessary.
		AL3.2: At least 8 hours and longer than those in the reference design.
	CR3.3: Engineered safety features	IN3.3: Reliability and capability of engineered safety features.
		AL3.3: Superior to those in the reference design.
	CR3.4: Barriers	IN3.4: Number of confinement barriers maintained (intact) after DBAs and DECs.
		AL3.4: At least one and consistent with regulatory requirements for the type of reactor and accident under consideration.
	CR3.5: Subcriticality margins	IN3.5: Subcriticality margins after reactor shutdown in accident conditions.
		AL3.5: Sufficient to cover uncertainties and to maintain shutdown conditions of the core.
UR4: Severe plant conditions: The frequency of an accidental release of radioactivity into the containment / confinement is reduced. If such a release occurs, the consequences are mitigated, preventing or reducing the frequency of occurrence of accidental release into the environment. The source term of the accidental release into the environment remains well within the envelope of the reference reactor source term and is so low that calculated consequences would not require evacuation of the public.	CR4.1: Frequency of release into containment / confinement	IN4.1: Calculated frequency of accidental release of radioactive materials into the containment / confinement.
		AL4.1: Lower than that in the reference design.
	CR4.2: Robustness of containment / confinement design	IN4.2: Containment loads covered by the design, and natural or engineered processes and equipment sufficient for controlling relevant system parameters and activity levels in containment / confinement.
		AL4.2: Larger than those in the reference design.
	CR4.3: Accident management	IN4.3: In-plant accident management (AM).
		AL4.3: AM procedures and training sufficient to prevent an accidental release outside containment / confinement and regain control of the reactor.
	CR4.4: Frequency of accidental release into environment	IN4.4: Calculated frequency of an accidental release of radioactive materials into the environment.
		AL4.4: Lower than that in the reference design. Large releases and early releases are practically eliminated.
	CR4.5: Source term of accidental release into environment	IN4.5: Calculated inventory and characteristics (release height, pressure, temperature, liquids/gas/aerosols, etc) of an accidental release.
		AL4.5: Remain well within the inventory and characteristics envelope of the reference reactor source term and are so low that calculated consequences would not require public evacuation.
UR5: Independence of DID levels, inherent safety characteristics and passive safety systems: An assessment is performed to demonstrate that the DID levels are more independent from each other than in the reference design. To excel in safety and reliability, the nuclear reactor assessed strives for better elimination or minimization of hazards relative to the reference design by incorporating into its design an increased emphasis on inherently safe characteristics and/or passive systems, when appropriate.	CR5.1: Independence of DID levels	IN5.1: Independence of different levels of DID.
		AL5.1: More independence of the DID levels than in the reference design, e.g. as demonstrated through deterministic and probabilistic means, hazards analysis, etc.
	CR5.2: Minimization of hazards	IN5.2: Characteristics of hazards.
		AL5.2: Hazards smaller than those in the reference design.
	CR5.3: Passive safety systems	IN5.3: Reliability of passive safety systems.
		AL5.3: More reliable than the active safety systems in the reference design.
UR6: Human factors (HF) related to safety: Safe operation of the nuclear reactor assessed is supported by accounting for HF requirements in the design and operation of the plant, and by establishing and maintaining a strong safety culture in all organizations involved.	CR6.1: Human factors	IN6.1: HF considerations are addressed systematically throughout the life cycle of the reactor.
		AL6.1: HF assessment results are better than those for the reference design.
	CR6.2: Attitude to safety	IN6.2: Prevailing safety culture.
		AL6.2: Evidence is provided by periodic safety culture reviews.
UR7: Necessary RD&D for advanced designs: The development of innovative design features of the nuclear reactor assessed includes associated research, development and demonstration (RD&D) to bring the knowledge of plant characteristics and the capability of analytical methods used for design and safety assessment to at least the same confidence level as for operating plants.	CR7.1: Safety basis and safety issues	IN7.1: Safety basis and a clear process for addressing safety issues.
		AL7.1: The safety basis for advanced designs is defined and safety issues are addressed.
	CR7.2: RD&D	IN7.2: RD&D status.
		AL7.2: Necessary RD&D is defined and performed, and the database is developed.
	CR7.3: Computer codes	IN7.3: Status of computer codes.
		AL7.3 Computer codes or analytical methods are developed and validated.
	CR7.4: Novelty	IN7.4: Pilot or demonstration plant.
		AL7.4: In case of a high degree of novelty: a pilot or demonstration plant is specified, built and operated, lessons are learned and documented, and results are sufficient to be extrapolated to a full-size plant. In case of a low degree of novelty: a rationale is provided for bypassing a pilot or demonstration plant.
	CR7.5: Safety assessment	IN7.5: Adequate safety assessment involving a suitable combination of deterministic and probabilistic methods, and identification of uncertainties and sensitivities.
		AL7.5: Uncertainties and sensitivities are identified and appropriately dealt with, and the safety assessment is approved by a responsible regulatory authority.