DESIGN AUTHORITY ADVICE NOTE

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1 DESIGN AUTHORITY ADVICE NOTE Title: Implications for Sizewell B from the Flamanville 3 Reactor Pressure Vessel Manufacturing Issues EAN No: Revision: 000 Author: Author s Organisation/Section/ Group: EDF Energy / Design Authority / Sizewell B Safety Case QA Grade: Task File Number: 2 E/TSK/SZB/13567 Attachments: None Station: Sizewell B Date: April 2016 Verifier(s): Verifier s Organisation/Section/Group: EDF Energy / Design Authority / Sizewell B Safety Case Verification Statement: This EAN has been verified by appropriate EDF Energy SQEPs in accordance with BEG/SPEC/DAO/010. I am satisfied that the conclusions are reasonable and are supported by the evidence presented. Approver: Section/Group: Design Authority / Sizewell B Safety Case Head Tech. Reviewer(s): N/A SQEP status/evidence of competence of the reviewer(s): N/A Commissioning of Technical Review Commissioned by: 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 12 Outline of, or reference to, the nuclear safety related application in which the work is to be used: N/A Technical review outcome N/A Details of aspects of the work reviewed: N/A 1 INTRODUCTION Flamanville 3 (FA3) is a Pressurised Water Reactor of EPR design currently under construction in France. Much of the plant has been completed and the large forgings have been installed. As part of the technical qualification of the manufacturing operations on the FA3 Reactor Pressure Vessel (RPV) domes, mechanical tests were performed on samples intended to be representative of the dome material. Such tests have not previously been standard practice because the nuclear pressure equipment order (known as ESPN) which required this evidence had only come into force in It became public in April 2015 that these tests had found variations in the mechanical properties and that subsequent investigation had identified variations in carbon content of the FA3 RPV upper and lower domes. This meant that some regions of the tested domes did not meet the manufacturing specification requirements. Areva, the company responsible for the FA3 dome manufacture, are currently assessing the full implications for the FA3 safety case. Given that the upper and lower domes for Sizewell B (SZB) and FA3 were both manufactured at the same site in Le Creusot, France, the Office for Nuclear Regulation (ONR) have asked if there is any risk that Sizewell B could have suffered a similar problem. A brief was provided to ONR in April 2015 (see Appendix A) which gave an initial view that there was no evidence for such an anomaly also existing at Sizewell B. Further work was identified as necessary to support this judgement. As part of this further work EDF Energy Materials group have reviewed the manufacturing processes [1] and Structural Analysis Group have performed sensitivity analysis on the RPV fracture analysis [6]. The remaining input has been provided by Design Authority and is summarised in this EAN. This Advice Note brings together all the work done in order to provide an overall judgement on the significance of the FA3 event to the Sizewell B RPV structural integrity. 2 SIGNIFICANCE TO SIZEWELL B The EDF Energy assessment of this event indicates that there is not a significant threat to the integrity of the Sizewell B RPV domes and that the Sizewell B safety case continues to be robust. Because of the low risk this event poses to Sizewell B it is not considered ALARP to perform any further analysis. There are a number of factors that give confidence in this position; Design & Specification The Sizewell B vessel domes were designed to have robust material properties. This included a low carbon content specification of 0.20% or less. Manufacturing Process There is confidence that Sizewell B would not have suffered from significant positive carbon segregation because Sizewell B used a much smaller ingot (approx 60 tonnes compared to 160 tonnes) and used a different ingot production process to the FA3 EPR process. The techniques used to produce the Sizewell B ingots have been proven to have suitably uniform solidification and acceptable positive segregation of carbon. In addition, material taken from the edge of the Sizewell B forging (the test ring) was tested and shown to have acceptable properties. Operational Experience There have been a large number of RPV domes produced which are similar to the Sizewell B dome, including many at the same forge using the same techniques. There have not been any recorded failures of such domes or any issues with carbon segregation despite a large number of reactor operating years having now been accumulated.

3 Page 3 of 12 Margins in the Fracture Analysis Even if there was some reduction in fracture toughness for the Sizewell B RPV domes, there are large safety margins in the Sizewell B fracture analysis so the demonstration of integrity safety case would not be undermined. Analysis has indicated that there would not be a cliff edge effect associated with reduced fracture toughness. The size of a crack which could be reasonably expected to occur would still be small compared to a crack which could lead to RPV failure. Consequences for the RPV Integrity The most likely locations of reduced fracture toughness are on the outer surface of the domes which are of less concern. Even if crack growth were to be worse than expected, it would most likely propagate to the nearest stress reliever, namely, the penetrations, rather than through wall. 2.1 Design & Specification The Sizewell B safety case for the Reactor Pressure Vessel includes a discussion regarding how the integrity was achieved such that the possibility of disruptive failure of the RPV shell without forewarning is sufficiently remote as to be deemed incredible [7]. Claimed as integral to this approach are a robust design and appropriate materials specification. The Sizewell B RPV was constructed in accordance with an established code of practice, the American Society of Mechanical Engineers (ASME) Code Section III (as adapted for UK use). The ASME Section II low alloy steel specification of SA-508 Class 3 was chosen in order to achieve robust mechanical properties. Sizewell B then specified an enhancement by reducing the maximum carbon content to 0.20% (Section 3.1 of Reference [1]). This was less than the carbon content of 0.22% specified for the FA3 forgings (Section 2.2 of [1]). Ensuring a low carbon content allows less opportunity for carbon to concentrate although it is recognised that the EPR forging has seen levels of carbon segregation of 0.30% despite a specification of 0.22%. The processes used at Sizewell B to prevent these levels of positive segregation are addressed in Section Manufacturing Process EDF Energy Materials Group have examined the manufacturing processes for both the Sizewell B and Flamanville 3 domes [1]. This included review of the Sizewell B manufacturing specifications and records. This review concluded that there is no clear evidence for a similar anomaly existing in the Sizewell B RPV domes as that found during the technical qualification for FA3. The primary evidence given to support this statement is that the different ingot production method used for Sizewell B is proven to have a lower propensity for developing areas of positive carbon segregation (Section 3.2 of Reference [1]). Supplementary to this is that SZB used much smaller ingots (approx 60 tonnes compared to 160 tonnes for FA3 domes) (Section 3.2 of Reference [1]). The smaller SZB ingots would have taken less time to solidify which would have made them less susceptible to carbon segregation and would have facilitated controls on segregation (Section 4 of Reference [1]). The ingot production method used for the Sizewell B ingot is a bottom pouring technique known by its French abbreviation LSD (Section 3.1 of Reference [1]). Such a technique was not considered possible for the very large EPR forgings and hence they utilised a conventional top pouring ingot technique. Figure 1 illustrates what is meant by top and bottom pouring. The LSD technique was developed with reducing positive carbon segregation as a specific consideration (Section 3.2 of Reference [1]). The exact mechanisms by which the LSD process reduces areas of positive segregation are complex but in general it results in a more even solidification which reduces the opportunity for carbon to concentrate in particular areas [1]. Special features of the LSD process include the fact that it involves bottom pouring of the ingot and also a thick iron plate at the base of the ingot mould to take away heat [1]. For the LSD process there are requirements to achieve a specific low height to diameter ratio to assist in the even solidification. The cooling process

4 Page 4 of 12 for the LSD ingot is illustrated in Figure 7 of reference [1]. In addition to the techniques which prevent positive segregation, the research conducted has confirmed that material is discarded from the ingot in areas where positive carbon segregation is most likely to occur (Section 3.1 of Reference [1]). The LSD ingots were subject to forging operations to allow discarding to be carried out on a specific region of the ingot. This facilitated a smaller discard region and hence smaller ingot compared to the process used for the FA3 domes although the intent to discard the region of highest segregation is the same in both processes. The LSD process is discussed in more detail in Section 3.2 of Reference [1], it is clear that the theory behind this technique suggests it is much less likely to result in positive segregation than the conventional ingot production technique used for the EPR domes. In addition to the theoretical analysis, there is also strong practical evidence that the LSD process does not result in significant positive segregation. In order to demonstrate the effectiveness of the process, Le Creusot produced a test LSD forging which went through destructive testing. The results of this testing were presented in a 1985 paper to the 10 th International Forging Conference [2] which has been reviewed by EDF Energy in Section 3.2 of Reference [1]. The mechanical properties and chemical composition of the test dome forging were measured across a range of locations of the dome. This included the locations in the centre of the dome which were found to have positive segregation in the EPR forging. It was confirmed that there was an absence of significant positive segregation and that a forging manufactured through the LSD route would demonstrate a greater consistency in mechanical properties [1]. Importantly it also confirmed the test ring (excess material taken from the edge of every dome in order to test the properties of the dome) will provide adequate indication of material properties across a dome forging if it is produced using the same manufacturing route. Sizewell B test ring results have been reviewed [1] and confirmed to be within specification for carbon content, hence it can be inferred that the properties across the dome are also highly likely to be acceptable. The LSD prototype is considered to have been manufactured from an equivalent steel to the Sizewell B domes ensuring the results of the prototype testing are applicable to Sizewell [1]. In summary, there is a high degree of confidence that the Sizewell B domes would have avoided significant positive segregation as found in the EPR forgings. 2.3 Operational Experience There are a large number of reactor operating years which have now been accumulated by PWRs using domes similar to that at Sizewell B, with 30 LSD domes produced at the time of the 1985 assessment [2] and further domes produced since then. In addition there is a vast amount of experience with domes made with other techniques and with other large forgings at nuclear plants. Operational Experience (OPEX) searches have been conducted with major industry bodies (recorded in the task file) and it is reassuring that no failures or cracking attributed to carbon segregation have been found. In fact no failures of the dome base material have been discovered in the searches performed. Failures have occurred in the material used at some power stations for penetrations through the domes but that is not part of the forging (the risk of such failure is considered relatively low for Sizewell B [9]). Because of the dissimilar metal welds, OPEX suggests that failure of the penetrations is much more likely and the dome parent material is relatively very low risk. It may be considered that defects could have been attributed to other mechanisms incorrectly, however there are few significant defects recorded by any mechanism. From OPEX searches, no defects were found to be associated with dome parent material so a search was done more widely, looking at cracking in other major forgings. The only real incident of note has been the large number of defects found in Doel 3 and Tihange 2 Reactor Pressure Vessels in Belgium. The analysis of this event has suggested this was caused by hydrogen flaking, the risk of which has already been judged as insignificant for Sizewell B [3]. The lack of significant defects being discovered worldwide gives a great deal of confidence given the huge number of non-destructive examinations which have been done on Reactor Pressure Vessels worldwide. Most plants will follow a 10 yearly inspection interval similar to that at Sizewell B. If

5 Page 5 of 12 carbon segregation had lead to significant reduction in fracture toughness some cracking is likely to have been revealed. Hence it can be inferred that significant carbon segregation remains rare. Research in the area of positive carbon segregation in reactor pressure vessels and other large forgings has previously been carried out. Such research has recognised that there will be a tendency towards areas of positive segregation [8]. Thus, efforts have gone into minimising this positive segregation such as using the LSD approach or discarding large areas, as done for the EPR domes. Thus the remaining question is whether there is any research to indicate if significant carbon segregation is common in large nuclear forgings even with measures in place to reduce it. Unfortunately, given the large capital cost such components are rarely destructively tested. In any case no other destructive testing could be more relevant to Sizewell B than the destructive testing of the prototype LSD forging (section 2.2) given it went through the same production process and was made from an equivalent steel. Previously, carbon segregation has not been seen as high risk compared to other factors which may affect fracture toughness (e.g. irradiation embrittlement). Extensive research has been done by the Materials Ageing Institute on material ageing in Light Water Reactors and carbon segregation does not feature in their main publication [4]. It may be that carbon segregation becomes higher priority and further research is done in this area in light of the Flamanville issue. In this case EDF Energy will be aware as part of normal business OPEX reviews and involvement with industry bodies. 2.4 Margins in the Fracture Analysis The demonstration of integrity for the Reactor Pressure Vessel [5] includes assessment of defect sizes which could cause the RPV to fail against sizes of defects that could credibly occur in the vessel. The margin between these two defect sizes is known as the Validation Factor and a value approaching 2 (or greater) is targeted in order to give confidence in the integrity. Structural Analysis Group have revisited this assessment [6] to ensure that there are no cliff edge effects associated with a reduced fracture toughness in the RPV domes which could occur if there was carbon segregation such as that seen at FA3. The Structural Analysis Group sensitivity calculations (Section 3 of Reference [6]) have applied factors of 0.8 and 0.7 to the original assessments to emulate the effects of different amounts of fracture toughness reduction. These values are considered reasonably conservative in light of the carbon measurements for the EPR domes. Acceptable validation factors (>2) are obtained even with a factor of 0.7 on the generic lower bound toughness. The sensitivity analysis demonstrates that if more realistic data is used for these locations then much larger margins would be expected. This would more than off-set a postulated reduction in fracture toughness. The Structural Analysis assessment [6] concludes that 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. Even in the unlikely situation that the fracture toughness values used in the safety case are reduced in some areas there remains significant safety margin. The work performed by Structural Analysis Group indicates that there are no cliff edge effects associated with a hypothetical reduced fracture toughness (Section 5 of Reference [6]). 2.5 Consequences for the RPV Integrity For Sizewell B, the concave region/bottom of LSD forgings were used as the inside surface for the RPV [1]. This was to provide further assurance that should any positive segregation remain, it would be located on the outer surface rather than the inner surface of the RPV. Even if crack growth were to be worse than expected, it would most likely propagate to the nearest stress reliever, namely, the penetrations, rather than through wall. Hence, even in the very unlikely situation that there is reduced fracture toughness and a crack does occur, the location of the reduced toughness would limit the consequences compared to other locations of the RPV. The irradiation of the RPV domes is relatively low compared to other parts of the RPV, hence, even if it were to have some reduced fracture toughness it is still not likely to be the region most at risk of

6 Page 6 of 12 failure, the effects of irradiation and temperature in the core shell region over the reactor lifetime are still likely to be bounding. It is also noted that the Boric Acid corrosion programme involves regular walk downs to check for any leaks from the RPV. This would be expected to detect a leak from the upper or lower domes. There are also visual inspections performed in line with ASME requirements. These inspections are designed to detect any cracks which may be visible on the surface. The next visual inspection of the upper dome is scheduled for RO ALARP Review of the Safety Case With the evidence presented it is considered that there is very low risk of the Sizewell B domes having experienced significant positive carbon segregation. An advanced manufacturing process was used and OPEX does not suggest such positive segregation is common. Even if segregation has occurred it is unlikely to significantly change the risk to the RPV integrity. The dome is still unlikely to fail as there is significant margin in the fracture analysis and the affected regions are bounded by other higher risk regions of the RPV. Even in the very unlikely situation of a crack developing, a crack in the outer surface of the dome presents relatively low consequences. The Sizewell B achievement of integrity safety case [7] claims there is low risk of RPV failure due to high quality manufacturing processes. The review of the manufacturing processes supports this claim [1] and more specifically, it has been shown that reasonably practicable measures were used to reduce positive segregation. The safety case also demonstrates the integrity of the RPV domes via fracture analysis [5]. It has been shown that this analysis would not be undermined by a reduction in fracture toughness similar to that found for the EPR domes [6]. Hence, the current safety case is considered to remain adequate for demonstrating that the risk at Sizewell B is ALARP. Further work has been considered, such as destructive testing of the old Sizewell B vessel head removed in RFO7. However, given that the current position is considered low risk, further work would provide minimal risk benefit and resources would be better committed elsewhere. This position is subject to Sizewell B Structural Integrity Panel (SIP) agreement and this EAN will be circulated to the SIP members upon completion. 3 CONCLUSION This Advice Note claims that there is no challenge to the Sizewell B Reactor Pressure Vessel domes integrity from the findings of positive carbon segregation during the qualification of the Flamanville 3 EPR domes. Given that there is no challenge to the Sizewell B dome integrity, the safety case remains an adequate justification that the risks for Sizewell B are ALARP. Given that the risks are ALARP, no further work is considered necessary. 4 REFERENCES 1. E/EAN/BBHB/0373/SZB/16; Review of Sizewell B RPV Dome Forging Components Following Flamanville 3 EPR OPEX; ; March 2016; 2. Application of Directional Solidification Ingot (LSD) in Forging of PWR Reactor Vessel Heads; and 10th International Forging Conference, Sheffield (UK); September EC ; Review Of Doel-3 Issues And Relevance To Sizewell B Safety Case; ; November 2014; 4. Materials Ageing in Light Water Reactors; Materials Ageing Institute; ; SXB-IP ; IR 5.3(2) Demonstration of Reactor Pressure Vessel Integrity; April 2014; Issue E/EAN/BBJB/0379/SZB/16; Sizewell B Consideration of Reduced Toughness in the Upper and Lower Closure Heads; March 2016;

7 Page 7 of SXB-IP ; IR 5.3(1) Achievement of Reactor Pressure Vessel Integrity; October 2015; Issue UKAEA report AERE-R9581; Investigation of the homogeneity of chemical composition and mechanical properties in a large A508 class 3 steel pressure vessel forging; September E/EAN/BAEB/0016/SXB/05; Management of the Degradation due to Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 600 Components and Alloy 82 and 182 Welds; ; March 2006, 5 FIGURES Figure 1: Pouring of metal through (a) top-pouring and (b) bottom-pouring processes; (1) ladle with metal, (2) mold, (3) stool, and (4) fountain 6 DISTRIBUTION LIST Barnwood - Design Authority Barnwood - Design Authority Barnwood - Design Authority Barnwood - Engineering Barnwood - Engineering Barnwood - Engineering Barnwood Independent Nuclear Assurance Sizewell B - Components & Programmes ED00037 ED00863 ED03047 EDWPB01 ED03075 EDCJG01 ED00348 BBBAJ24

8 Page 8 of 12 APPENDIX A APRIL 2015 BRIEF TO THE ONR Brief to ONR on the Implications to Sizewell B of EPR Reactor Pressure Vessel (RPV) Domes Reduced Material Properties Due to Carbon Segregation Prepared by Structural Integrity Lead Design Authority Summary 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.

9 Page 9 of 12 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%. 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. 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.

10 Page 10 of 12 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. 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.

11 Page 11 of 12 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. -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.

12 Page 12 of 12 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* 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)