atw 2018-12

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atw Vol. 63 (2018) | Issue 11/12 ı November/December ENVIRONMENT AND SAFETY 588 2 Analysis and classification of nuclear safety barriers Coming back to Figure 1, the different activities taken to ensure an adequate safety level may be classified upon three different areas or top-level pillars of activities: Safety Design and Performance, Management Improvement and Safety Assessment. If the IAEA design and operational safety requirements safety standards [14] and [15] are taken as a comprehensive list, its 96 requirements (removing 19 requirements not related to the safety performance of the plant) might be classified in 74 requirements under the pillar of safety design and performance, 11 requirements addressing management improvement, and 11 requirements under the category of safety assessment. • Safety Design and Performance Within this safety pillar, the activities directly affecting the safety response of the plant are classified. By directly it is meant that the object of the activity is the humanmachine system made up of safetyrelated SSC and human actions actuating on them. Such activities comprise design technical specifications, testing, maintenance, design principles such as independency, redundancy, etc., but also emergency procedures, administrative procedures such as maintenance, surveillance and inspection, etc. Most nuclear countries have nowadays implemented a set of common standards regarding safety design (see [14] and [15]), most of which rely on a deterministic approach even if probabilistic results are becoming more and more integrated into several aspects of the design. The 74 IAEA safety-performancerelated design and operational requirements found on [14] and [15] related to the safety design and performance category may be in turn classified and simplified as follows: • Fundamental design criteria PP Defence in depth underlying criteria (multi-barriers, independency, redundancy, etc.) PP Single-failure criterion PP SSC safety classification PP Assumed Postulated Initiating Events (PIEs) PP Safety margins • Safety procedures PP Abnormal and failure procedures PP Emergency procedures PP SAMGs PP Maintenance, surveillance and inspections PP Operational limits and conditions for safe operation • Management Improvement This top safety pillar groups those activities indirectly modifying the humanmachine system safety performance. By indirectly it is meant that the object of the activity is not the humanmachine system performing the safety function in charge of preventing, correcting (i.e. controlling) or mitigating the adverse condition, but those activities whose performance is expected to ultimately impact that system. The connection between the object of a management improvement activity and the humanmachine system usually relies on the performance analysis of different aspects dealing with the human organization or SSCs in charge of performing the safety functions. The importance of the available tools for continuously improving the safety performance of the plant has long been stressed; see for instance [16] and [17]. The most relevant tools indirectly improving the safety performance of the plant are the followings: • Operating Experience • Nuclear Safety Culture • Risk Oversight Process • Performance Indicators • PRA applications • Trending and Performance Analysis • Corrective Action Program • Benchmarking and Self-Assessment • Observation program • Safety Assessment By safety assessment it is meant any activity in charge of ensuring that the plant meets with the design safety requirements and | | Fig. 3. Relationship between the different Nuclear Safety Pillars. verifies the conformity of the plant with the quantitative safety design objectives. The most relevant activities dealing with safety assessment are hereafter indicated for clarification: • Safety Analysis (analysis of plant response against adverse situations departing from normal conditions): PP Probabilistic Safety (or Risk) Assessment PP Deterministic Analysis • External Reviews (e.g. IAEA, WANO, INPO) and Independent Oversight; NPP safety performance can therefore be schematically considered according to Figure 2, where the actions directly in charge of responding to adverse conditions are carried out by the human-machine system, the performance of this system being in turn affected by a set of management tools. And as a cross-sectional set of activities, the safety assessment tool which applies at many different levels with the aim of ensuring the compliance of the NPP design and performance with the expected safety design objectives. As already introduced in section 1, the different activities belonging to the three safety pillars can be classified upon whether they look at preventing accident conditions – before the accident occurs –, correcting the accident – after the onset of the accident – or mitigating the accident – to limit the undesired consequences once the core has partly or fully damaged – (Figure 3). They can also be classified in values and practices depending on whether they look at physically modifying or managing the human-machine system respectively. Table 1 arranges the fundamental categories of activities under the three different safety pillars according to this twofold classification. Environment and Safety Release-Category-Oriented Risk Importance Measure in the Frame of Preventive Nuclear Safety Barriers ı Juan Carlos de la Rosa Blul and Luca Ammirabilea
atw Vol. 63 (2018) | Issue 11/12 ı November/December Prevention Correction Mitigation Values Practices DiD criteria Single-failure criterion Safety-classification-driven design criteria PIEs LCOs Safety Analysis Abnormal and failure procedures Risk Monitor Maintenance, surveillance and inspections Maintenance Rule PRA applications OE Nuclear Safety Culture ROP Trending and Performance Analysis Corrective Action Program External Reviews | | Tab. 1. Arrangement of the NPP safety activities. Table 2 lists these barriers along with their type and underlying safety approach. 3 Risk-based mitigating safety barriers After the Fukushima-Daiichi nuclear accident, considerable efforts have been put in place to provide nuclear plants with new equipment to mitigate beyond-core-melt accidents, see e.g. [18]. These efforts resulted in Safety Margins DiD physical barriers Single-failure criterion Safety-classification-driven design criteria Safety Systems Emergency procedures Safety barriers / activities Type Criteria DiD criteria Single-failure Safety classification PIEs Prevention Correction Mitigation 1 D, P 1 Prevention Correction Prevention Correction Mitigation 1 Prevention Correction Mitigation 1 Safety Margins Correction D Maintenance, Surveillance and Inspections Prevention D, P 2 Maintenance Rule Prevention P Limit Condition for Operation (LCO) Prevention D Abnormal, failure and EPGs Prevention Correction PRA applications Prevention P OE Prevention D Nuclear Safety Culture Prevention D ROP Prevention P Trending and Performance Analysis Prevention D CAP Prevention P External Reviewers Prevention D, P Safety Systems Correction D Physical Barriers Correction Mitigation Mitigating Equipment Mitigation D, P Safety Analysis Correction Mitigation SAMGs Mitigation D | | Tab. 2. Criteria and rationale / parameter underlying the safety barriers / activities. backfitted mitigating systems, fixed or portable, in charge of maintaining the cooling capability of the fuel in the reactor vessel core, reducing the flammable gases and transferring the heat outside the containment or reactor building. As shown in the former tables, the majority of safety barriers look at preventing the occurrence of an accident. As shown in Table 2, the majority of the prevention safety barriers have D D D, P D D D, P DiD physical barriers Safety-classification design criteria 1 PIEs 1 Mitigating equipment 1 SAMGs been designed taking account only design safety criteria. When a probabilistic safety approach is instead considered among the safety criteria in designing such prevention safety barriers, the underlying figures of merit are based on the Core Damage Frequency (CDF) and the Large Early Release Frequency (LERF) and Large Release Frequency (LRF). This will be the case, for instance, of design improvements aimed at improving the safety response of a system featuring a significant contribution to the CDF, such as passive Reactor Coolant Pump seals preventing a coolant leakage during Extended Loss of Alternate Current scenarios. LERF/LRF comprises those accidents resulting in a radioactive release higher than a specific magnitude threshold (depending on the national regulatory framework, this value can be 3 %, 10 %, etc., of the initial volatile fission products stored in the fuel assemblies and released to the environment, see e.g. Ref. [19]) at a certain time, i.e. LERF, or no matter the releasing time, i.e. LRF (as for the timing threshold magnitude, it depends on the national dispositions as well). International recommendations for existing plants do not usually specify maximum LERF frequencies. For new plants, these categories of events are required to be practically eliminated, see e.g. [20] and [21]. The extended practice is to regulate on the LERF/LRF variations caused by design changes in plant and inspection findings. For instance, the Spanish Consejo de Seguridad Nuclear sets acceptable ranges for backfitting designs, see Figure 4 (where FGLT and FGL stand for LERF and LRF respectively) as described in [22] adapted from [11]) and Table 3 aimed at classifying performance indexes 1) Only in some limited, more recent designs 2) Only in limited NPPs ENVIRONMENT AND SAFETY 589 Environment and Safety Release-Category-Oriented Risk Importance Measure in the Frame of Preventive Nuclear Safety Barriers ı Juan Carlos de la Rosa Blul and Luca Ammirabilea
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