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Name	Category	Date	Size
Sub-Chapter 17.2 - Demonstration of Relevant Good Practice in EPR Design.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	391.24 KB
Sub-chapter 17.2 provides the required demonstration of relevant good practice and includes: a review of both the experience of EPR™ designers and the review and assessment process applied to the design, including a summary of the R&D effort underpinning the EPR™ design; a review of the design codes used in EPR™ design, including reference to international and national codes; a summary of the use of operational feedback, in particular from French and German plants, in optimising EPR™ design; and a discussion of a comparison of the UK EPR™ design against the ONR safety assessment principles, to confirm that all the key nuclear safety requirements embodied in the SAPs are met by the UK EPR™ design. Download
Sub-Chapter 17.1 - Explanation of ALARP Requirement.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	91.83 KB
Sub-chapter 17.1 provides an overview of the requirements of the ALARP principle. UK Health and Safety Legislation places a duty on all companies to conduct their operations such that the risk posed to their workers and members of the public is as low as reasonably practicable (ALARP). In the context of a nuclear power plant, this duty requires that all measures are taken during design and operation to minimise radiation doses to workers or members of the public, provided the cost of such measures is not disproportionately large compared with the benefits achieved. For generic design acceptance, it is proposed that supporting safety submissions should include the following information: a demonstration of relevant good practice; a review of design options considered in the design process; presentation of results of a probabilistic risk assessment; and arguments that no further reasonable practicable improvements could be implemented, and therefore the risk has been reduced to ALARP. Download
Sub-Chapter 16.6 – Analysis of Extreme Beyond Design Basis Events Carried Out in Response to Fukushima.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	269.75 KB
Sub-chapter 16.6 outlines the analysis of the robustness of the UK EPR™ design against extreme events, carried out in response to the Fukushima event that occurred in March 2011. The sub-chapter reviews the ability of the UK EPR™ to withstand severe external events, (notably major seismic and external flooding events of magnitudes beyond those considered in the current design). It also presents the time periods that would be available for operators to take action to prevent the release of radioactivity to the environment in hypothetical scenarios involving total loss of AC power and ultimate heat sink. Design modifications are described which are being implemented to enhance the robustness of UK EPR™ systems, structures and components against such extreme beyond design basis events. Download
Sub-Chapter 16.5 - Adequacy of the UK EPR™ design regarding functional diversity.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	5.05 MB
Sub-chapter 16.5 provides a demonstration of the adequacy of the functional diversity of the UK EPR™ design. Functional diversity is addressed for all the frequent postulated initiating events (including front line systems and support systems), as these have higher requirements for mitigation. The safety functions are composed of plant level safety functions and the diversity is demonstrated within the plant level safety functions for all the frequent faults. Since some events are more bounding than others for a given plant level safety function, a comprehensive review of the transients is performed to select the limiting events before their examination by calculations. The sub-chapter is comprised of two parts: firstly, the methodology assessing the frequent events and the safety functions considered is presented and the selection of the most limiting events for each plant level safety function is demonstrated, and secondly, the limiting events are analysed by calculations or argumentations to demonstrate that the safety criteria are met. Download
Sub-Chapter 16.4 - Specific Studies.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	7.39 MB
Sub-chapter 16.4 presents the deterministic assessment of six specific faults: a double ended break of the main coolant line (2A-LOCA); a double-ended break of the main steam line outside the containment (2A-SLB); a steam generator tube rupture (1 tube) with main steam line break; a steam generator tube rupture (1 tube) with the affected steam generator main steam relief train stuck open; a multiple steam generator tube rupture (10 tubes in one steam generator at power) and spurious reactor protection system actuation. Assessment of the faults is presented, including description of the analysis methods, main assumptions and models used. In addition to these six specific faults the safety cases for the following types of events are also presented in this sub-chapter: heterogeneous boron dilution faults, faults linked with the Spent Fuel Pool (e.g.; connected pipework, cask loading pit faults or faults in fuel transfer tube) and faults associated with the failure of support systems (e.g.; ventilation, electrical and cooling chain). Download
Sub-Chapter 16.3 - Practically eliminated situations.pdf	Pre-Construction Safety Report (PCSR)	20-06-2023	132.63 KB
Sub-chapter 16.3 presents sequences that are considered to be "practically eliminated". In the EPR™ context, "practical elimination" refers to the implementation of specific design measures to reduce the risk of a large early release of radioactive material to the environment to as low as reasonably practicable. To achieve practical elimination, each type of accident sequence that could lead to a large early release of radioactivity is examined and addressed by design measures. Demonstration of practical elimination of an accident sequence takes into account uncertainties due to the limited knowledge of physical phenomena involved in severe accident analysis. Conditions covered by specific treatment leading to their practical elimination are those which are liable to give rise to significant early releases; mainly high-pressure core meltdown sequences. The sequences considered in this approach are related to severe accidents and containment bypass situations. Download
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