ENGINEERING ADVICE NOTE

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1 ENGINEERING ADVICE NOTE Title: Sizewell B. Consideration of Reduced Toughness in the Upper and Lower Closure Heads EAN No: Revision: 000 Author: QA Grade: Task File Number: 2 E/TSK/SZB/13395 Structural Analysis Group, Structural Integrity Branch Station: Sizewell B Attachments: Appendix 1 Date: March 2016 Verifier(s): Verifier s Organisation/Section/Group: Structural Analysis Group, Structural Integrity Branch Verification Statement: This document has been subject to verification according to EDF Energy Generation CESF Branch Instruction E/PROC/ENG/BI/048 and Company Specification BEG/SPEC/DAO/010 Approver: Section/Group: Structural Analysis Group, Structural Integrity Branch Technical Review: Not required. SQEP status/evidence of competence of the reviewer(s): N/A Commissioning of Technical Review Commissioned by: N/A Outline of, or reference to, the nuclear safety related application in which the work is to be used: N/A Technical review outcome Details of aspects of the work reviewed N/A 2016 Published in the United Kingdom by EDF Energy Nuclear Generation Ltd. All rights reserved. No part of this publication may be reproduced or transmitted in any form or by any means, including photocopying and recording, without the written permission of the copyright holder, EDF Energy Nuclear Generation Ltd., application for which should be addressed to the publisher. Such written permission must also be obtained before any part of this publication is stored in a retrieval system of any nature. Requests for copies of this document should be referred to Barnwood Document Centre, Location 12, EDF Energy Nuclear Generation Ltd, Barnett Way, Barnwood, Gloucester GL4 3RS (Tel: ). The electronic copy is the current issue and printing renders this document uncontrolled. Controlled copy-holders will continue to receive updates as usual. LIMITATION OF LIABILITY Whilst EDF Energy Nuclear Generation Ltd believes that the information given in this document is correct at the date of publication it does not guarantee that this is so, nor that the information is suitable for any particular purpose. Users must therefore satisfy themselves as to the suitability of the information for the purpose for which they require it and must make all checks they deem necessary to verify the accuracy thereof. EDF Energy Nuclear Generation Ltd shall not be liable for any loss or damage (except for death or personal injury caused by negligence) arising from any use to which the information is put.

2 Page 2 of 17 1 INTRODUCTION Public statements made by the French Regulator in the Spring of 2015 regarding an anomaly in the material properties of the Flamanville-3 Reactor Pressure Vessel (RPV) domes prompted ONR to ask EDF Energy to consider the potential implications to Sizewell B, given the components were all made at Le Creusot Forge. In response EDF Energy (Design Authority) provided a brief to ONR in May This is reproduced in Appendix 1 and included a summary of defect tolerance calculations to determine the likely sensitivity of results to a pessimistic reduction in fracture toughness. A commitment was made at that time to issue a verified Report or EAN to formalise these calculations and this EAN is that formalised report. 2 PREVIOUS CALCULATIONS The integrity of the Sizewell B Primary Circuit is covered in Chapter 5 of the Station Safety Report (SSR) with the RPV integrity covered in Section 3. This summarises the safety design bases and safety design approach and a key element of this is demonstrating the integrity of the RPV under all operational and anticipated fault condition. This is summarised at a high level in Reference 1 which includes a summary of key results from the defect tolerance/crack growth calculations that have been carried out. These were initially undertaken by NNC/CEGB but have been revised at various times to account for developments since commissioning. However, the basic approach has remained the same. For the RPV, the approach is to postulate a defect at welds and other key locations in the vessel assembly and, using the results of finite element thermal/stress calculations, calculate limiting sizes of these and the upper bound crack growth from an assumed defect size at start of life (SoL). These locations are shown in Figure 1. The concern regarding impaired toughness is considered to most likely affect the perforated regions of the domes (Locations 19 and 20 Figure 1). Therefore other locations in the upper and lower vessel domes (i.e Locations 1, 17 and 18) are not affected as well as all locations in the vessel body. However, Locations 19 and 20 were not considered individually in Reference 1 due to the lower stress levels in these areas compared with others and a bounding approach was generally adopted by the safety case. Thus Location 1 (lower head to transition ring weld) was chosen to bound both Locations 19 and 20 (upper and lower heads respectively). For the present purpose, therefore, it is just this location that has been subject to re-analysis to give indicative numbers (revised limiting defect sizes and validation factors) to enable a judgement to be made regarding the significance of reductions in toughness. The most recent integrity assessment for Location 1 as reported in Reference 1 was presented in Reference 2. This report was up-issued to Revision 001 in 2013 to account for changes to the irradiated fracture toughness in the beltline region but the results for Location 1 were not changed. Consequently the present review of location 1 is based on the work undertaken in 2001 but with sensitivity to fracture toughness. For Condition I, II and Test transients, the limiting loadcase at location 1 is Feedwater Cycling where an initiation depth of 57.7mm was predicted for a semi-elliptical defect (a/c=0.2) (Table 4 of Reference 1 and Table 16 of Reference 2). This was obtained for a circumferential defect in the plane of the transition ring to lower head dome weld and therefore subject to the default 55MPa residual stress. Furthermore, irradiated weld toughness was assumed as it was stated unirradiated toughness was not significantly improved in the case of weld metal.

3 Page 3 of 17 For Condition III and IV events Table 6 of Reference 1 gives a limiting defect depth of 100.8mm for the Small Break Loss of Coolant Accident (LOCA). This is for a semi-elliptical defect in the circumferential direction. Given the magnitude of the validation factor for this case, (3.6) it was not considered necessary to consider this case any further in the May 2015 brief although some limited analysis was undertaken to back up this judgement. This has been expanded here to provide numerical results and appropriate graphical presentation. The calculations undertaken to provide results for the May 2015 brief were verified to QA Grade 4 (self-checking). The present note formalises these and provides verified results (QA Grade 2). The results presented in May 2015 have not changed. 3 UPDATED CALCULATIONS In order to undertake the sensitivity analyses required to address the present concern the previous analyses have been recovered and re-run to check that the previous results can be reproduced (Section 3.1 below. The only locations that might be affected by this issue are 19 and 20 in Figure 1. However, since these were bounded by location 1, the updated calculations have been based on stresses and geometry for that location. Consequently, these repeat calculations have been extended to investigate the effect of a reduced toughness on limiting defect sizes and hence validation factors. This is done in Section 3.2 below. Furthermore, since Locations 19 and 20 are considered to be in parent and not significantly affected by irradiation damage, an updated assessment based on unirradiated parent toughness has also been included in Section 3.2 below. In these sensitivity calculations, toughness has been multiplied by factors of 0.8 and 0.7. The factor of 0.8 was postulated in the original response in May 2015 and the further assessment (a factor of 0.7) has been included here for additional sensitivity and to consider possible cliff edge effects. Results for the bounding Condition I, II and Test transients (Feedwater Cycling), are reported in Section below, The base case lower bound toughness is 168.9MPa m at an operating temperature TCOLD of 293 C for the generic lower bound toughness (Equation 1 in Reference 2). Hence toughness values corresponding to factors of 0.8 and 0.7 are 135.1MPa m and 118.2MPa m respectively. For the small break LOCA transient, the toughness varies as the metal temperature is assumed to fall below the onset of upper shelf behaviour. This is clearly shown in appropriate figures (Figure 5 to Figure 8). 3.1 Recovering Previous Results In this subsection the current safety case assessments are summarised for bounding cases identified above and recalculation of these results demonstrated Feedwater Cycling The previous results were obtained using the EDF-Energy R-Code program and so the archive records of the output from these assessments have been used to recreate an input file using the latest release of the program. Repeating the previous assessment gave limiting depth of 57.9mm for the same loading conditions and materials properties compared with the 57.7mm reported previously. It is considered that this small difference is due to the differences between the versions of R-Code used (which reflect updates to the R6 procedure). In any event these are considered to be insignificant Small Break LOCA The previous results were also based on the EDF-Energy R-Code programme but these were presented in the form of a trajectory of equivalent K (denoted Kequiv) plotted on a toughness-

4 Page 4 of 17 temperature plot as in Figures 4 to 20 of Reference 2. These show the material toughness as a function of temperature and the change from ductile fracture to transition region behaviour can be clearly observed. The assessment loci show Kequiv plotted against the predicted inside surface metal temperature (where it is lowest and hence the toughness enters the transition region first).. In Reference 2 Kequiv was based on the parameters KI P, KI S, f(lr) and : K equiv K f P I ( L r K S I ) This has been modified in the present calculationsto account for the replacement of by the parameter V in the latest revision of the R6 defect assessment procedure. Thus K equiv K P I f S K I ( L r ) V Use of the V parameter gives rise to slightly lower values of Kequiv compared with use of and hence slightly more favourable limiting defect sizes than previously. In order to show this effect, the original plot relating to a semicircular defect at Location 1 (i.e. Figure 4 in that reference) is reproduced in Figure 3 in the present EAN with the Kequiv trajectories calculated using V methods of secondary stress correction now added. The reduction in pessimism using the latter is clear. In addition, the figure indicates that the conservative warm pre-stressing principle has been used to underwrite the limiting size 3.2 Updated Sensitivity Analysis. Details and results of the various reanalysis cases are summarised here Feedwater Cycling For the reanalysis the same Geometry Code has been used in the R-Code calculations (i.e. Code 242, as appropriate to the spherical geometry at this assessment location). It is noted that this is still appropriate as the more recent 400 series geometries are for cylinders only. In addition to simply repeating the previous analysis with factors on toughness, the reanalysis also considers toughness appropriate to unirradiated parent material as well as zero residual stress as noted in Section 3 above. The results from these analyses are summarised on Figure 4. It may be observed that using parent rather than weld toughness results in a significant increase in limiting defect depth so that the effect of assuming a degraded toughness is mitigated. The end-of-life defect depth for the validated initial defect is 27.8mm (Table 1 of Reference 1). For the most onerous sensitivity case (toughness factored by 0.7) a limiting defect depth of 83.4mm has been calculated which gives a validation factor of 3.0. For information, the verifier used an alternative source for SIF values in which the geometry was represented as a cylinder rather than a sphere. This gave slightly greater SIF values and a reduced limiting depth of The validation factor then becomes Small Break LOCA For the small break LOCA the analysis has been repeated using V in place of the parameter to determine the equivalent K. The original R-Code SIF geometry code has been retained for the same reason as in the Feedwater Cycling case. In the case of the small break LOCA, the

5 Page 5 of 17 limiting location is at the surface intersection point due to the significant thermal shock stress. Figure 5 and Figure 6 summarise results for updated Kequiv trajectories (based on the V parameter) using the original assumptions of irradiated weld toughness and 55MPa weld residual stress and a Reference Nil Ductility Transition Temperature (RTNDT) of C.. These figures are for semi-elliptical and semicircular defects respectively. Once again the assessment is for the surface intersection point and the figures show the effect of factors of 0.8 and 0.7 on toughness. Using the same assumptions, limiting defect depths for 1.0mm ductile tearing would be approximately 87mm and 79mm for the semi-elliptical and semicircular defects respectively with a factor of 0.8 on toughness. For a 0.7 factor these reduce to about 75mm and 56mm respectively. These results are significantly less favourable than the original values of 100.8mm and 111.4mm, although Validation Factors are still in excess of 2.0. Specifically, the VF of 3.6 for the semi-elliptical defect in Table 18 of Reference 2 reduces to 3.19 and 2.75 for toughness factors of 0.8 and 0.7 respectively. Likewise the VF of 4.4 for the semicircular defect reduces to 3.09and 2.19 However, Locations 19 and 20, which are representative of the position in the forgings where the impaired toughness is postulated, are in parent and well away from the effects of irradiation damage. Hence less pessimistic assumptions are clearly applicable both with regards to material toughness and welding residual stress. However, for consistency with earlier assessments of parent, a pessimistic bounding value of +18 C has been adopted for RTNDT, being the generic value for parent. Consequently, the calculations behind these two figures have been repeated accounting for parent un-irradiated toughness and zero residual stress but with the transition toughness curves appropriate to the higher RTNDT of +18 C. These are shown in Figure 7 and Figure 8 and it is clear that significantly more favourable results are obtained. Specifically the limiting defect size is bounded by the validity of the SIF solution (85% of wall thickness 119mm) even for a 0.7 factor on toughness. 85% wall thickness equates to a validation factor of 4.3. It may also be observed from these figures that the adoption of the pessimistic parent RTNDT of +18 C has no effect on the predicted limiting defect depths. 4 DISCUSSION These calculations have considered the effect of possible degradation in toughness for the centre portions of the Sizewell B RPV top and bottom heads in a simple and straightforward manner. It has been shown that potential adverse effects on the limiting defect results can be more than offset by the use of parent toughness values and zero residual stress which would be more appropriate for these regions compared with the more bounding location 1 that was used in the safety case. 5 CONCLUSIONS The following conclusions have been drawn 1 Changes to the R6 procedure have not altered the original defect tolerance results, which have been reproduced from the previous analysis. 2 Re-assessments using material properties appropriate to areas which could be susceptible to a reduction in toughness (i.e. locations 19 and 20 on Figure 1) give significantly more favourable results compared with the original bounding location. 3 Acceptable validation factors (>2) are obtained even with a factor of 0.7 on the generic lower bound toughness, indicating no cliff edge effect.

6 Page 6 of 17 4 The Sizewell B integrity case for the upper and lower heads is not challenged by the findings related to the Flamanville-3 RPV head forgings. 6 REFERENCES 1 Sizewell B Power Station, Station Safety Report Identified Reference IR 5.3(2) Demonstration of Reactor Pressure Vessel Integrity. CDMS Reference SXB-IP Issue 104 April and Sizewell B Power Station A revised Integrity Assessment for the End-of-Life Safety Case for the Reactor Pressure Vessel IoF Boundary. E/REP/STAN/0057/SXB/01. Revision 001 March 2013.

7 Page 7 of 17 7 FIGURES Figure 1 RPV Assessment Locations

8 Page 8 of 17 Figure 2 Comparison of 2001 and 2015 Results for Same Input Data alim = 57.7mm, Re-analysis 2001 Analysis Kr 0.8 alim = 57.9mm, Lr Figure Comparison of 2001 and 2015 Results Small LOCA, Cut 1, a/c=1.0 t = 800 sec RTNDT = C 1mm Tearing Toughness - Weld Metal 200 Solid Line Calculations Dashed Line Calculations Toughness, MPa m t = 2000 sec t = 500 sec Upper Shelf Toughness - Weld Metal Temperature, C

9 Page 9 of 17 Figure Feedwater Cycling Reanalysis Toughness Factor = 1.0 Toughness Factor = 0.8 Toughness Factor = alim = 93.5mm Kr 0.8 alim = 83.4mm alim = 106.1mm Lr Figure 5 Small LOCA, Cut 1, a/c=0.2. Irradiated Weld, 55MPa Residual 250 RTNDT = C mm 100mm Toughness after 1mm Tearing Toughness, MPa m mm 80mm 70mm 60mm 50mm Initiation Toughness 50 0 Both upper shelf and transition curve factored by degradation factor Irradiated Weld, Factor = 1.0 Irraddiated Weld, Factor = 0.8 Irradiated Weld, Factor = 0.7 Loci of assessment points for given crack depth Temperature, C

10 Page 10 of 17 Figure 6 Small LOCA, Cut 1, a/c=1.0. Irradiated Weld, 55MPa Residual. 250 RTNDT = C mm Toughness after 1mm Tearing Toughness, MPa m mm Initiation Toughness 50 0 Both upper shelf and transition curve factored by degradation factor Irradiated Weld, Factor = 1.0 Irradiated Weld, Factor = 0.8 Irradiated Weld, Factor = 0.7 Loci of Assessment Points for given crack depth Temperature, C Figure Small LOCA, Cut 1, a/c=0.2 Unirradiated Parent, Zero Residual Stress 350 RTNDT = 18 C 300 Toughness, MPa m mm Toughness after 1mm Tearing Initiation Toughness mm 50 0 Both upper shelf and transition curve factored by degradation factor Unirradiated Weld, Factor = 1.0 Unirraddiated Weld, Factor = 0.8 Unirradiated Weld, Factor = 0.7 Loci of assessment points for given crack depth Temperature, C

11 Page 11 of 17 Figure K Small LOCA, Cut 1, a/c=1.0. Unirradiated Parent, Zero Residual Stress 350 RTNDT = 18 C Toughness, MPa m mm 50mm Toughness after 1mm Tearing Initiation Toughness 50 Both upper shelf and transition curve factored by degradation factor. Unirradiated Weld, Factor = 1.0 Unirradiated Weld, Factor = 0.8 Unirradiated Weld, Factor = 0.7 Loci of Assessment Points for given crack depth Temperature, C

12 Page 12 of 17 8 APPENDIX 1 Brief to ONR on the Implications to Sizewell B of EPR Reactor Pressure Vessel (RPV) Domes Reduced Material Properties Due to Carbon Segregation Summary Prepared by Chris Townsend Structural Integrity Lead Design Authority The recent public statements regarding an anomaly in the material properties of the FA-3 RPV domes has prompted ONR to ask EDF Energy to consider the potential implications to Sizewell B, given the components were all made at Le Creusot forge. The ONR is aware of the variations in carbon and the potential effects on fracture toughness based on the destructive testing of an EPR (USA) RPV Head by AREVA. This brief is in response to that request and has been prepared by the Sizewell B Design Authority with support from Structural Integrity Branch. It represents an initial review of the manufacturing process coupled with indicative fracture assessment calculations to gauge potential implications to Sizewell B. Our conclusion from this initial review is that there is no clear evidence for such an anomaly being present at Sizewell B and the judgements presented provide assurance of RPV integrity. In the unlikely event these are incorrect; the initial fracture assessment for postulated defects provides additional reassurance. The brief also outlines further work required to back up the judgements. Assessments to Date Consideration has focused on two main areas. Firstly the manufacturing processes adopted for the making of the RPV domes at Sizewell B. This included the steel making process and material specification. Secondly, a fracture assessment in the affected areas, taking account of a potential reduction in fracture toughness values. Manufacturing Processes Both the original RPV Head domes 1 and the bottom RPV dome for Sizewell B were made around the same time using similar manufacturing techniques. The specification for the low alloy ferritic steel stipulated a maximum carbon content of 0.2%. 1 The original Sizewell B RPV head dome was replaced in 2007 by another, originally planned for Hinkley Point C. Both domes were made around the same time to the same specification at Le Creusot Forge.

13 Page 13 of 17 It is noted that there is a difference in the ingot weight between FA-3 (approx 160 tonnes) and Sizewell B (approx 60 tonnes). Significantly, for Sizewell B domes a bottom pouring (ladle to ingot) process was adopted. This process had been designed to produce ingots with a more uniform solidification and to reduce macrosegregation. This potential benefit is being investigated further. An initial view of the forging process also suggests that positive segregation (a region of higher carbon content and associated inclusions) is removed during the process through discards taken from the top of the forging. This also requires confirmation. Testing of RPV domes prolongations gave consistent, acceptable values, which are compliant with both the design code (ASME) and the RPV equipment specification. No specific testing for carbon was done on the region of concern; however, product analysis (both head domes and lower dome) gave carbon values in the range %, all below the specified maximum carbon content. Fracture Assessments Structural Analysis Group has reviewed the fracture assessments for both RPV domes. The perforated regions of both domes is not considered in any detail in the safety case due to the lower stress levels in the region and the bounding approach adopted by the safety case. The review has determined that both Locations (19 upper dome & 20 lower dome) are bounded by Location 1. (see below diagram from SSR for locations). However Location 1 is a weld and the analyses therefore assumed the presence of a uniform 55MPa residual stress. Location 18 was not considered to bound the locations. Structural Analysis group has undertaken indicative calculations to establish potential effects of reduced fracture toughness. The original limiting (initiating) defect size for Location 1 is 57.7 mm for a 0.1 aspect ratio. The previous calculations have been reworked to take account of the absence of residual stresses and recent updates to the R6 Defect Assessment Procedure. This results in an increased defect size of 96 mm. A reduction of 20% in fracture toughness has then been applied reducing the limiting defect size back to 79.4 mm. This reduction is an estimate of a likely reduction in fracture toughness. Further sensitivities will be required. The value was chosen to get a feel for how sensitive it would be on safety margins. Crack growth for Location 1 has been determined as 2.8 mm over 40 years. This results in a validation factor of 79.4/27.8 = 2.85 (s-o-l crack is 25mm). This is for Condition I & II. Note this crack growth has not been factored and still represents the crack growth in a weld. Further assessment of Conditions III & IV also gave acceptable validation factors taking credit for ductile tearing (1mm), where appropriate. The bounding fault for this region is considered the small LOCA on the basis the RPV is still at pressure. Other transients have not been considered at this stage. Initial Judgements The initial review suggests that there is no clear evidence for a similar anomaly existing in the Sizewell B RPV domes. The following judgements are given to provide assurance of RPV integrity whilst investigations continue.

14 Page 14 of 17 Judgement 1 Significant efforts were made during manufacture of Sizewell B s domes to limit the formation of positive macrosegregation. The different manufacturing processes used and the significantly lower ingot weight for Sizewell B than FA-3 provide evidence that the Sizewell B domes would be expected to have a lower propensity for carbon segregation The way the ingot is formed offers a more uniform solidification than a top-poured ingot. The forging process adopted and the location of discard indicates that areas of positive segregation will have been removed. All these aspects are supportive of reducing the likelihood that large-scale segregation would be present in the finished components, such as those observed on EPR (USA). Judgement 2 Margins of safety remain high for the RPV domes based on indicative calculations for a postulated defect in the regions of concern using a reduction in fracture toughness of 20%. This does not take credit for any potential increase in fracture toughness at the inner surface. The assessment covers both normal and fault conditions. Judgement 3 Any postulated crack will most likely occur on the inner surface, not the outer surface where reductions in fracture toughness have been discovered on the EPR (USA). It is judged that in all probability a crack in the perforated region will propagate to the nearest stress reliever, namely, the CRDM or BMI penetrations at the top and bottom of the domes. Hence, the calculations remain conservative for a postulated through wall orientated crack. These judgements will be further backed by future assessment work to be undertaken in the coming months. The confidence in these judgements is also based on the reliable operating experience of PWRs worldwide and older than Sizewell B. The collective measures taken during design, manufacture, construction and operation and embedded in the twin arms of the structural integrity safety case, namely the demonstration and achievement of integrity provide assurance that RPV integrity can be maintained. Further Work To support the above judgements and to be able to give a definitive answer to Sizewell B s position with respect to carbon segregation, a programme of work is proposed covering three main activities and potentially backed by a fourth work stream. Work Stream 1 Manufacturing Process Materials Group will undertake production of an Engineering Advice Note (EAN) similar to the one produced for the Doel 3 issue. This will involve a detailed understanding of the manufacturing process and comparisons with FA-3 processes. The scope of this work will include: -Compare steelmaking, to include: material specification, understanding the steelmaking route, ladle chemical analysis, sizes of ingots, the location and amount of material discard. -Compare forging processes, to include: understanding the forging route, review of forging ratios.

15 Page 15 of 17 -Compare heat treatment and machining operations, to include: times and temperatures used, component thickness at heat treatment stages, quenching process, extent of machining to achieve component dimensions. -Compare final properties, to include: final product chemical analysis, mechanical properties. Work Stream 2 Fracture Assessment Structural Analysis Group to undertake a detailed fracture assessment of the perforated region for normal and fault conditions and conduct sensitivity on fracture toughness values. A wider number of fault transients should be considered. Potential impact on safety margins to be reported. Work Stream 3 Opex Search Design Authority to conduct an Opex search to determine whether any other PWR operator has experienced failure attributed to carbon segregation. Also to engage research bodies to establish results from destructive testing results from other PWRs e.g. Materials Ageing Institute, EPRI. Work Stream 4 Test work The potential exists to conduct testing of the old Sizewell B RPV Head currently residing in the outage building. The Structural Integrity Panel will advise on the merits of testing taking account of ALARP considerations. It is understood from discussions with Health Physics and Materials Group that filings from drillings at different depths would give a clear indication of carbon content through the RPV Head dome wall thickness. This would give confidence that the extent of any segregation was well defined. Alternatively more comprehensive destructive testing of the RPV Head, including fracture toughness testing could be performed. The current position is that it not ALARP to do this test. However, this is only a preliminary view and this might change with more information forthcoming from AREVA. This work stream was considered by the SIP on the 8 th May. It recommended that test work should only be done if the results from the first two work streams point towards further investigation work. Programme Work requests will be added to respective work programme and prioritised as appropriate. Recognising the existing work programmes and the reporting of future test work by AREVA it is proposed to adopt the following programme for deliverables: Materials Group to deliver EAN by end of November 2015*

16 Page 16 of 17 Structural Analysis Group to deliver EAN by end of October 2015* Design Authority to deliver EAN by end of November 2015* * Dates to be agreed with respective groups. In the event any anomaly is identified during the work programme it will result in sentencing under normal business processes. This brief is prepared at QA Grade 3 with peer checks from Structural Analysis Group and Materials Group (Structural Integrity Branch). It was supported at the Structural Integrity Panel meeting held on the 8 th May Note The brief sent to ONR included the RPV Assessment Location figure, included above as Figure 1

17 Page 17 of 17 9 DISTRIBUTION LIST Design Authority ED00863 Design Authority ED03047 Design Authority ED02992 Sizewell B BBBAJ24 Structural Analysis Group ED01534 Structural Analysis Group BBEAT09 Materials Group EDWPB01, Materials Group ED03075