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 analyses⁵ 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.

3 In this case, the potential for accidents in one facility to influence parameters or conditions in another has to be considered for the second facility independently or in combination with other external events (e.g. earthquakes and resulting explosions).

4 Most NPPs sell relatively small amounts of energy in a form of heat used for district heating, greenhouse heating, etc., by local communities.

5 It is assumed that design information and safety analysis results that are needed to perform the INPRO sustainability assessment are readily available.

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^a such that the frequencies and consequences of the accidents are greatly reduced. In the event of an accident, off-site releases of radionuclides^b are prevented or mitigated so that there will be no need for public evacuation^c.

INPRO user requirements Criteria Indicator (IN) and Acceptance Limit (AL) UR1: Robustness of design

IN1.1: Robustness of design of normal operation systems.^d

AL1.1: More robust than that^e 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

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

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.

AL3.5: Sufficient to cover uncertainties and to maintain shutdown conditions of the core.

TABLE 1. OVERVIEW OF THE INPRO METHODOLOGY IN THE AREA OF SAFETY OF NUCLEAR REACTORS (cont.)

INPRO user requirements Criteria Indicator (IN) and Acceptance Limit (AL) UR4: Severe plant conditions: source term and is so low that calculated consequences would

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.

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 design. To excel in safety and reliability, the nuclear reactor

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

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.

TABLE 1. OVERVIEW OF THE INPRO METHODOLOGY IN THE AREA OF SAFETY OF NUCLEAR REACTORS (cont.)

INPRO user requirements Criteria Indicator (IN) and Acceptance Limit (AL) UR6: Human factors (HF)

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.

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.

a Within this publication, a reference reactor (or design) is a reactor of the latest design operating in 2013. It should preferably be designed by the same corporate designer as the reactor assessed and using the same technology. For innovative reactors that may have no operating prototypes in 2013, the latest design that has been safely operated, or at least licensed, can be used as the reference design.

Based on previous experience with INPRO assessments, the definition of date for the selection of the reference design helps to avoid potential misinterpretations of terms. Note that 2013 was the date selected at the beginning of the latest methodology update. This date should be revised periodically along with the rest of methodology.

b If significant amounts of toxic chemicals are used in the reactor design (e.g. as coolants or fuel forms in the innovative reactors) or can be generated during the reactor operation or accidents, then potential accidental releases of toxic chemicals have to be considered as part of the INPRO assessment. The INPRO criteria used for the assessment of potential releases of toxic chemicals should be similar to those developed for the assessment of radioactive releases.

c Other protective measures may still be needed. Effective emergency planning, preparedness and response capabilities will remain a prudent requirement. This is covered in the Infrastructure area of the INPRO methodology.

d In this publication, ‘robustness of design’ is considered for DID Levels 1 to 4. However, this criterion CR1.1 is focused only on normal operation systems (Level 1 of DID).

e The requirement of superiority in the INPRO acceptance limits generally means an expected improvement of a given characteristic of the new design compared with the reference design. However, in cases where this specific characteristic in the reference design has already incorporated the best international practice available at the moment of assessment, the confirmation of equivalent characteristics in a new design will be sufficient for the positive assessment of a specific criterion or evaluation parameter. In this case, the assessor needs to prove both that the reference design is state of the art in relation to a given characteristic and that the new design characteristic is equivalent to that in the reference design.

2. GENERAL FEATURES OF NUCLEAR ENERGY SYSTEMS