Level 3 PSA - EDF Hinkley Point

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HPC-NNBOSL-U0-000-RES-000028 Version 1.0 Level 3 PSA NOT PROTECTIVELY MARKED specific frequency data and other site specific updates. Between the March 2011 GDA PCSR Sub-chapter 15.5 [Ref. 3] and the issue of this sub-chapter an update to the Level 1 PSA model occurred. The Individual Risk calculations were updated to take the PSA model update into account. The methodology for the Societal Risk was developed before the PSA model update but the assessment referenced within this sub-chapter was completed after the update. The Worker Risk is comprised of three reports. The first describes the methodology, the second calculates the Release Categories for the assessment and the third details the actual assessment. The first two reports were completed before the model update along with an initial assessment, and the final assessment presented within this sub-chapter was undertaken after the PSA model update was complete. The differing timings of the assessments for the three parts of the Level 3 PSA impact on their presentation within this sub-chapter. Because of this the three aspects of the Level 3 PSA are presented separately – including the tables for the inputs to each part. At the moment the Level 1 and Level 2 PSA only considers a single unit site. However certain aspects of the Level 3 PSA require the fact that HPC is intended to be a twin unit site to be taken into account. This is discussed in more detail in the Strategy to Assess the Impact of Twin Reactor Site on the PSA [Ref. 4] and its impact on each aspect on the Level 3 PSA is discussed within the relevant sections of this sub-chapter. UNCONTROLLED WHEN PRINTED NOT PROTECTIVELY MARKED Page 6 of 60
HPC-NNBOSL-U0-000-RES-000028 Version 1.0 Level 3 PSA NOT PROTECTIVELY MARKED 3 ASSESSMENT OF INDIVIDUAL RISK The radiological consequence which is considered in this assessment of Individual Risk is the unmitigated effective dose to a child at 500 m downwind from the point of release during the first 7 days following the release using weather condition DF2. The methodology used in the March 2011 GDA PCSR Sub-chapter 15.5 [Ref. 3] for calculation of Individual Risk has not been updated for the site specific PCSR. The justification for adopting the simple approach used in the GDA is given in Sub-chapter 2.2 [Ref. 5], where it is shown that the GDA methodology is bounding of a site specific method. Sub-chapter 2.2 shows that the GDA methodology for Societal Risk is bounding of the HPC site, and the same methodology was used within the GDA for both Societal Risk and Individual Risk. Comparison nuclide concentration calculations were made to show that the deterministic weather condition “DF2” used in the GDA calculations used for calculation of Individual Risk is bounding at Hinkley Point C based on historical meteorological data. The advanced dispersion model UK-ADMS was run using 5 years of sequential metrological data from HPC to investigate three accident release scenarios (short, medium and long release) and a statistical analysis was carried out to compare the results obtained using UK-ADMS against those obtained using “DF2”. The study showed that the calculations using DF2 consistently resulted in higher air and ground deposition concentrations close to the source of release (distances less than 1.5 Km) and covered 98% of the results at distances up to 10 Km. [Ref. 6] Therefore, since this assessment considers a child downwind at a distance of 500m, a partially deterministic approach is used for individual risk rather than a fully probabilistic approach. The release frequencies used in the Individual Risk analysis below are the site specific frequencies for Hinkley Point C however the consequences are obtained from the conservative deterministic assessment. The assessment of the individual risk to any person off-site takes input from four sources: the Level 1 PSA, the Level 2 PSA and the Design Basis Assessment (DBA) together with the additional initiating events identified by expert review in section 3.1 of this sub-chapter. The objective of the Level 1 PSA is to identify those sequences that result in a designated failure state and hence, by definition, fall outside the design basis, i.e. those that lead to core damage. The other sequences are classified as success states within the Level 1 PSA, or may not have been included in the PSA from the outset as they could not lead to core damage or are non reactor faults. Therefore, the success states within the Level 1 PSA may not contain sufficient detail to provide all the information required for the present Level 3 risk assessment. Additional Non Core Damage (NCD) sequences resulting from the Level 1 PSA are assessed within section 3.3 of this sub-chapter. The DBA (Chapter 14) does consider Low Consequence (LC) faults in considerable detail, and covers some of the initiating events excluded from the PSA as they cannot lead to core damage. Thus the majority of the additional plant analysis and plant data required for this Level 3 risk assessment is available from, or bounded by, the DBA. UNCONTROLLED WHEN PRINTED NOT PROTECTIVELY MARKED Page 7 of 60
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