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Name	Category	Date	Size
Sub-Chapter 16.2 - Severe Accident Analysis (RRC-B).pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	8.71 MB
Sub-chapter 16.2 describes the analysis of severe accidents, i.e. very unlikely events which result in fuel failure, degradation of the structural integrity of the reactor core, and release of radioactive fission products into the reactor coolant system. The overall safety objectives, and the approach to severe accident control, which focuses on preservation of the integrity of the containment under the associated loads, are presented. An assessment of core melt sequences is detailed, including identification and classification of the relevant scenarios from a phenomenological point of view. The areas assessed include: the primary depressurisation system, which prevents overpressurisation of the primary circuit; hydrogen control within the containment; stabilisation of core melt outside the reactor pressure vessel; and calculation of containment pressure and temperature in severe loss of coolant accidents. Radiological consequence calculations for severe core melt accidents are presented, to demonstrate that, taking account of the design provisions, the release of radioactive products outside the plant remains within the prescribed limits. Download
Sub-Chapter 16.1 - Risk Reduction Analysis (RRC-A).pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	2.78 MB
Sub-chapter 16.1 describes the risk reduction analysis, in which a set of design extension conditions (multiple-failure sequences) are analysed in order to justify the design of additional safety features ("RRC-A features"), which enable the prevention of severe accidents in these complex situations. The overall safety objectives, methodology, analysis principles, assumptions and data, are presented. An assessment of design extension condition sequences is detailed, including: failure of all control/shutdown rods on reactor trip demand (either mechanical failure or failure of the reactor trip signal); loss of off site power plus failure of all emergency diesel generators (station blackout); total loss of feedwater; total loss of cooling chain; small break loss of coolant accidents with additional safety injection failures; and boron dilution with additional operator error. An assessment of radiological consequences is presented, to demonstrate that, following these transients, the release of radioactive products outside the plant will have very low consequences for members of the public. Download
Appendix 16B - 4900 MW safety analyses used in Chapter 16.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	651.74 KB
Appendix 16B contains the generic EPR™ Basic Design Report 99 section corresponding to the analysis of risk reduction category A (RRC A) sequence "Small break loss of coolant accident without low head safety injection system". This sequence has not yet been specifically analysed for the UK EPR™ in the framework of the PCSR studies. The analysis presented in this appendix should be read in conjunction with section 3.8 of Sub chapter 16.1. Download
Appendix 16A - Computer codes used in Chapter 16.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	233.67 KB
Appendix 16A briefly presents the computer codes used for severe accident (RRC B) analyses. The codes used are listed, alongside their current field of application and in which sub-section of Sub chapter 16.2 these codes are used. Each code is described, together with an outline of its validation basis, where required. Since the risk reduction (RRC A) analyses presented in Sub chapter 16.1 use the same codes as the design basis analyses these codes are described in Appendix 14A. Download
Sub-Chapter 15.7 - PSA Discussion and Conclusions.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	586.4 KB
Sub-chapter 15.7 presents the results and main findings of the Level 1 and Level 2 probabilistic safety analyses, together with the assessment of the radiological impact of the UK EPR™ design in terms of off-site risk to the public due to postulated accidents, and the use of the PSA in the design process. An iterative process to identify design improvements using PSA was implemented throughout the development of the EPR™ design; the main examples of design changes made resulting from PSA studies are presented here.. Several sensitivity analyses are also presented in this sub-chapter including the effects of preventative maintenance, the analysis of long term transients, sensitivity to initiating event frequencies; modelling of common cause failures; operator actions; reliability data and system design features. The probabilistic studies performed for the UK EPR™ design during the GDA give the assurance that the risk of potential release of radioactive products into the environment is reduced to as low as reasonably practicable. Download
Sub-Chapter 15.6 - Seismic Margin Assessment.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	839.23 KB
Sub-chapter 15.6 presents the Seismic Margin Assessment (SMA) performed for the generic design assessment of the UK EPR™. The EPR™ design objective is that, following an earthquake, the safety functions needed to return the plant to a safe shutdown state should not be unacceptably affected. The seismic capacity of the UK EPR™ is assessed by a PSA-based SMA approach, which uses the PSA model to identify combinations of seismic equipment failures which could result in core damage, as well as combinations of seismic failures, random failures and human errors which contribute significantly to seismic risk. The seismic margin assessment demonstrates that the seismic capacity of the UK EPR™ is greater than 1.6 times the Design Basis Earthquake and that there are no cliff edge effects for seismic events which are more severe than the design basis event. Download
Sub-Chapter 15.5 - Level 3 - PSA Assessment of off-site risk due to postulated accidents.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	284.96 KB
Sub-chapter 15.5 summarises the process followed to perform a probabilistic assessment of the radiological impact of the UK EPR™ design (Level 3 PSA). The results are presented in terms of: individual risk to any person off the site, presented as frequency versus consequences; and societal risk, that is the annual frequency of events which could potentially lead to more than 100 immediate or eventual fatalities in the wider UK population. Download
Sub-Chapter 15.4 - Level 2 PSA.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	1.13 MB
Sub-chapter 15.4 presents the Level 2 probabilistic safety analysis (PSA), which assesses the probability of the severe accident phenomenon and the related response of the containment to potential loads. The Level 2 PSA assesses the characteristics of radiological releases from severe core damage sequences leading to containment failure, leakage or damage. The Level 2 PSA calculates the probability, composition, magnitude, and timing of fission product releases from the plant. The scope of the probabilistic study is described, together with a detailed description of the methodology used, including: the interface with the level 1 PSA (via the definition of core damage end states); the phenomenological evaluations performed for severe accidents; the development of containment event trees; the definition of fission product release categories; and the evaluation of the sensitivities and uncertainties. The results of the Level 2 PSA are presented in terms of the frequencies and release characteristics of the severe accident sequences, the binning of these sequences into release categories and the release risk associated with each release category. Download
Sub-Chapter 15.3 - PSA of accidents in the spent fuel pool.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	322.77 KB
Sub-chapter 15.3 addresses the likelihood of damage to the fuel assemblies located in the spent fuel pool, or of boiling to the fuel building atmosphere. An assessment of the frequency of initiating events affecting the spent fuel pool cooling is presented, together with plant data assumptions and reliability data. The results of analysis, for both spent fuel pool non-draining and draining initiating events, are presented in terms of fuel damage frequency (per reactor per year).< Download
Sub-Chapter 15.2 - PSA for internal and external hazards.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	1.02 MB
Sub-chapter 15.2 presents the results of an initial study to analyse the risk of core damage associated with internal and external hazards for the UK EPR™. The analysis of internal hazards covers all potential hazards and quantifies the risk associated with fire and flooding hazards. The analysis of external hazards is applied to a generic UK site (coastal or estuary) using the data presented in Chapter 2. The list of external hazards is reduced by a screening analysis that considers the expected impact on the plant safety; the expected frequency of core damage and the expected frequency of the external hazard. The results of the study are presented in terms of core damage frequency (per reactor per year), for both internal fire and flooding, and the screened in external hazards. An assessment of the core damage frequency associated with the loss of ultimate heat sink (total loss of the water intakes) is also presented here. Download
Sub-Chapter 15.1 - Level 1 PSA.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	1.96 MB
Sub-chapter 15.1 presents the level 1 probabilistic safety assessment (PSA), which allows identification of the sequences of internal events that can lead to core damage, and estimates the overall core damage frequency due to internal events in all reactor states. The scope of the probabilistic study for internal events is defined, together with the methodology used, including plant data assumptions, reliability data, treatment of common cause failures, I&C modelling, and human reliability assessment (both for pre-fault human errors and human errors during the course of the fault sequences) methodology. A discussion of the internal initiating events analysed is given, and the results of the Level 1 PSA analysis for internal initiating events are presented in terms of core damage frequency (per reactor per year). Download
Sub-Chapter 15.0 - Safety requirements and PSA objectives.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	132.82 KB
Sub-chapter 15.0 outlines the purpose and scope of the UK EPR™ probabilistic safety assessment (PSA), and presents the numerical targets that are used to evaluate and verify the UK EPR™ design. A number of Safety Design Objectives (SDOs) are adopted for the UK EPR™ and demonstration that these SDOs are met confirms that the EPR™ design complies with key ONR Safety Assessment Principles (SAPs). In addition to the SAPs numerical targets, the design also considers quantitative probabilistic safety objectives in accordance with the Technical Guidelines. Download
Sub-Chapter 14.7 - Fault and Protection Schedule .pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	2.44 MB
Sub-chapter 14.7 provides a summary of the fault and protection schedule for the UK EPR™. The fault and protection schedule is a list of the initiating events that are considered in the EPR™ design due to their potentially unacceptable consequences, and it includes the initiating events and sequences, their frequencies, the classified safety systems that protect against them and the overall protection claims. The rationale and justification for the initiating events considered in the fault schedule is given, together with a justification of the comprehensiveness of the fault protection. This sub-chapter also describes the principles used to define the protection systems set-points and those applied to modify or deactivate settings during changes in the reactor state, and presents the diverse means available to reach a long-term safe shutdown state. In addition, information is provided on the ALARP discussion regarding the adequacy of the UK EPR™ design to protect the plant against faults. Download
Sub-Chapter 14.6 - Radiological consequences of design basis accidents.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	630.31 KB
Sub-chapter 14.6 presents the assessment of the radiological consequences of the various transients (PCC 2), incidents (PCC 3) and standard accidents (PCC 4) that are significant from the point of view of radiological releases. The aim of the assessment is to verify that the systems are properly designed and operated, such that the potential discharge of radioactive products outside the plant, which could have consequences for the public, remains within the prescribed limits. The general principles, numerical targets, data and assumptions are given, together with details of the calculations performed. Download
Sub-Chapter 14.5 - Analyses of the PCC-4 events.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	9.02 MB
Sub-chapter 14.5 describes the analysis of the PCC-4 events listed in Sub chapter 14.0 (i.e. design basis accidents, or events where the estimated frequency of occurrence is between 1 in 10,000 and 1 in a million reactor years). For each transient analysed, a description of the typical or assumed sequence of events is given together with identification of potential causes of the transient; any transient specific methods and assumptions are defined, as are any impact on the design of the claimed safety systems. Download
Sub-Chapter 14.3 - Analyses of PCC-2 events.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	1.85 MB
Sub-chapter 14.3 describes the analysis of the PCC-2 events listed in Sub chapter 14.0 (i.e. design basis transients, events where the estimated frequency of occurrence is greater than 1 in a 100 reactor years). For each transient analysed, a description of the typical or assumed sequence of events is given together with identification of potential causes of the transient; any transient specific methods and assumptions are defined, as are any impact on the design of the claimed safety systems. Download
Sub-Chapter 14.4 - Analyses of the PCC-3 events.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	2.44 MB
Sub-chapter 14.4 describes the analysis of the PCC-3 events listed in Sub chapter 14.0 (i.e. design basis incidents, or events where the estimated frequency of occurrence is between 1 in 100 and 1 in 10,000 reactor years). For each transient analysed, a description of the typical or assumed sequence of events is given together with identification of potential causes of the transient; any transient specific methods and assumptions are defined, as are any impact on the design of the claimed safety systems. Download
Sub-Chapter 14.2 - Analysis of the Passive Single Failure.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	1.72 MB
Sub-chapter 14.2 presents the methodology and analysis of the consequences of passive single failures at the time of the initiating event for design basis (PCC-2 to PCC-4) events. The UK EPR™ is designed in compliance with the single failure criterion, which includes either an active single failure within the first 24 hours after the occurrence of an initiating event or a passive single failure at 24 hours after the event occurrence. This criterion is part of the assumptions made for the design basis fault studies of PCSR Sub-chapters 14.3 to 14.5. From a UK perspective, there is a need to review each PCC-2 to PCC-4 event, and to demonstrate that there are no passive single failures within 24 hours which could prevent the successful operation of the safety function that is provided to protect against the fault. The overall analysis shows that the UK EPR™ design is robust with respect to passive single failures. Download
Sub-Chapter 14.1 - Plant Characteristics taken into account in the Accident Analyses.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	464.16 KB
Sub-chapter 14.1 describes the generic plant characteristics that are applied to the accident analyses. Characteristics which are specific to a particular accident analysis are specified within the section describing that accident analysis. The plant characteristics assumed in the accident analyses cover: plant geometrical data, plant initial conditions, reactivity coefficients, residual (decay) heat, I&C signals related to reactor trip and safety systems operation, and safety system characteristics. The approach used in the accident analysis with regard to Departure from Nucleate Boiling (DNB) is also defined in this sub-chapter. Download
Sub-Chapter 14.0 - Assumptions and Requirements for the PCC Accident Analyses.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	226.24 KB
Sub-chapter 14.0 presents the assumptions and requirements for the Plant Condition Category (PCC) accident analyses. The standard reactor states, covering full power operation to cold shutdown, are defined, together with a list of the PCCs to be assessed. A set of 'PCC accident analysis rules'; are defined, which are used to demonstrate that the safety systems are designed in an appropriate manner. The rules include acceptance criteria, and the application of the single failure criterion. Download