letter to EDF Nuclear Operation Division

Transcription

Montrouge, 20th March 2015
Ref.: CODEP-DCN-2015-008144
Person in charge: Romain PIERRE
Tel: 01.46.16.42.76
Fax: 01.46.16.44.31
Email: [email protected]
For the attention of the Director
Nuclear Operation Division
EDF
Site Cap Ampère – 1 place Pleyel
93 282 SAINT-DENIS CEDEX
Subject: Nuclear power reactors - EDF
Summary of generic studies performed as part of the third ten-yearly outage safety reviews for
the 1300 MWe reactors
Ref.:
See appendix 4
Dear Sir,
As licensee of basic nuclear installations (BNIs) and in accordance with Article L. 593-18 of the Environment
Code, Electricité de France (EDF) is required to conduct a safety review of each of its reactors every ten years.
The purpose of this periodic review is to assess the condition of each reactor in the light of the applicable rules
and to update the assessment of the risks or drawbacks that presents this reactor with regard to the interests
mentioned in Article L. 593-1 of the Environment Code, more specifically taking account of the condition of the
facility, the experience acquired during its operation, evolution of knowledge and the rules applying to similar
facilities. It must also take account of international best practices.
Taking advantage of the design similarities between the twenty reactors of the 1300 MWe plant series, EDF
initiated an advance “generic” review of all the reactors in the series. The results of these generic studies will then
be applied by EDF to each of the reactors of the 1300 MWe plant series during the course of their third tenyearly outage inspections (VD3) scheduled to run from April 2015 until about 2024.
The purpose of this letter is to present ASN’s conclusions regarding this generic phase of the VD3-1300 periodic
safety review.
After this generic phase, in the months following each of the VD3 of the 1300 MWe reactors, EDF will submit a
report to ASN and the Minister responsible for nuclear safety, in accordance with Article L. 593-19 of the
Environment Code, presenting the conclusions of the periodic safety review of the concerned reactor and the
modifications implemented or envisaged in order to correct the detected non-compliances and improve the level
of safety.
ASN will then complete its present generic position and send the Minister responsible for nuclear safety its
analysis of the conclusions report for each reactor review and will if necessary issue new prescriptions
determining the conditions to be met for continued operation of the facility concerned up until the following
periodic safety review.
*
*
*
In 2010, or five years before the periodic safety review of the first reactor of the 1300 MWe, the Advisory
Committee for nuclear reactors (GPR) was consulted by ASN concerning the orientations of the programme of
work proposed by EDF for its VD3-1300 generic safety review, in particular comprising a list of the studies to
be carried out in order to reassess the safety of these reactors. Further to the opinion of the GPR, ASN issued a
position statement on these orientations in the letter mentioned in reference [1].
The studies produced since then by EDF on the basis of this programme were the subject of numerous position
statement letters and requests from ASN further to the reviews carried out with the technical support of IRSN
and, for some of them, following consultation of the GPR.
To close this generic examination phase for the VD3-1300 periodic safety review, ASN consulted the GPR on
15th and 16th October 2014, on the subject of the general conclusions of the studies and the modifications
envisaged by EDF (reference [2]) to improve the safety of the 1300 MWe series reactors with respect to the
orientations initially adopted. Following these two days of meetings, the GPR sent ASN its opinion in reference
[3] and, in the letter in reference [4], EDF confirmed the commitments presented to the GPR session.
*
*
*
Monitoring the conformity and condition of the facilities
ASN considers that the steps taken or planned by EDF to verify the conformity of the condition of its
1300 MWe reactors and ensure the satisfactory control of their ageing up until their fourth ten-yearly outage
inspection (VD4) are acceptable.
This generic assessment in no way anticipates any additional opinions ASN may express later following the
forthcoming consultation of the Advisory Committee for nuclear pressure equipment (GP-ESPN) on the subject
of the strength of the 1300 MWe nuclear reactor vessels beyond 30 years of operation.
Reassessment of hazard-related risks
ASN considers that the studies carried out by EDF to reassess the protection of its reactors against internal and
external hazards is on the whole satisfactorily based on a review of the practices in use on similar facilities in
other countries (for example the development of baseline requirements to take account of tornados and
projectiles caused by strong winds) and on evolution of knowledge resulting from research programmes or test
results (for example the effects of over-pressure in the event of a fire).
ASN in particular underlines the scope of the studies carried out by EDF and the importance of their results,
both in terms of improving the demonstration of the ability of its facilities to withstand hazards and in terms of
the modifications envisaged by EDF to reinforce the protection of its facilities, more specifically with regard to
hazards not considered in the initial design of the 1300 MWe series of reactors.
ASN also notes that given the mobilisation of the engineering capacity of EDF for the stress tests conducted
following the events at Fukushima, the initial review schedule for reassessing hazards in the generic phase of the
VD3-1300 periodic safety review was significantly disrupted and numerous answers to ASN requests, some of
which are recent, are still being waited for, despite the imminence of the first ten-yearly outage inspections of the
1300 MWe reactors. ASN therefore considers it acceptable that some of the safety improvements it requests will
be integrated by EDF in accordance with a schedule which, for certain reactors of the 1300 MWe series, goes
beyond their VD3.
Reassessment of the control of accident risks
In order to update the safety demonstration of the 1300 MWe reactors, EDF went back over a large number of
deterministic assessments of the design basis accidents, in certain cases using new methods including a more
accurate representation of the phenomena involved. EDF also corrected anomalies in the studies identified
before the periodic safety review. ASN considers these actions to be positive. ASN also has a positive opinion of
the action plan initiated by EDF and aimed at mitigating the radiological consequences of a steam generator tube
rupture (SGTR) and the resulting modifications, which will make a significant step towards mitigating radioactive
releases into the environment. It does however recall that certain studies are to be supplemented, with these
additions currently being prepared by EDF or undergoing technical review by IRSN. They could be the subject
of subsequent requests by ASN.
In addition to the deterministic demonstration, EDF also developed a number of level 11 probabilistic safety
assessments (PSA) concerning events inside the Nuclear Steam Supply System (NSSS) and the spent fuel pool, as
The level 1 PSAs examine the scenarios leading to fuel melt and determine their frequency. These assessments are thus able to
identify any weak points requiring design or operating modifications.
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well as certain external hazards. These studies revealed the need to modify the facility in order to reduce the risks
associated with fire and events affecting the spent fuel pool.
EDF also carried out the first level 22 PSAs for the 1300 MWe reactors, considering in turn the status prior to
the VD3 and the status after incorporation of the VD3 modifications. A comparison between the results of these
two studies showed a significant reduction in the frequency of large-scale releases, especially in the case of total
loss of electrical power. ASN does however consider that these PSAs need to be modified in order to realistically
reflect the condition of the facilities, their operation and the risk of releases in the event of an accident.
The studies carried out by EDF on the prevention and mitigation of severe accidents mainly concerned the
increased prevention of the severe accident risk, in particular with regard to scenarios involving an early loss of
containment. These studies, supplemented by the level 2 PSAs, led EDF to propose about ten equipment
modifications and update its “severe accident” baseline safety requirements. Generally speaking, ASN considers
that the modifications proposed by EDF to improve the containment and reduce the risks of early, large-scale
releases, are pertinent. Additional work is expected of EDF in response to its commitments and the requests
already submitted by ASN.
Concerning fuel storage in the spent fuel pool, ASN considers that the measures EDF intends to take, as a result
of the periodic safety review and the stress tests, in order to avoid uncovering of the fuel stored or being handled
in the pool, will constitute a significant safety improvement. Additional studies will however be required,
focusing mainly on the vulnerability of these additional measures to hazards.
In general, ASN considers that all these measures, leading to a more realistic appreciation of the risks and to the
identification of modifications capable of reducing certain risks, are real advances in the safety approach.
Reassessment of the control of detrimental effects3
Article L. 593-18 of the Environment Code requires that the periodic safety review be able to update not only
the assessment of the risks of accident, but also the detrimental effects of the facility as a result of its normal or
degraded operation.
The subject of satisfactory control of the detrimental effects has not however been fully integrated by EDF into
its examination of the VD3-1300 periodic safety review nor, more generally, into its examination of the other
periodic safety reviews of its reactors.
As mentioned in the letter in reference [6], the approach adopted by EDF to take account of these detrimental
effects in the periodic safety reviews of its reactors will be gradually reinforced in accordance with the schedule
for entry into force of the general regulations applicable as of 1st July 2015.
*
*
*
Further to the risk control reassessment studies, in the summary note in reference [2], EDF presented an
extensive list of modifications to be performed on the 1300 MWe reactors, which will in particular allow:
 reinforcement of protection against hazards, in particular through modifications concerning:
o the protection of equipment that is important for safety against projectiles caused by strong
winds;
o increased capacity of the air-conditioning systems (ventilation, chillers), so that in the event of a
heat wave, a temperature can be maintained in the premises that is compatible with the working
of the equipment important for safety;
The level 2 PSAs examine the nature, importance and frequency of releases outside the containment.
Article 4.1 of the order of 7th February 20112 modified in reference [5] defines a “detrimental effect” as being “on the one hand,
the health and environmental impact of the facility owing to water intake and discharges and, on the other, the nuisances that this
can cause”.
2
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o the prevention of explosion risks in the event of an earthquake, by reinforcing the strength of
the hydrogen-carrying circuits in the nuclear island and by ensuring automatic shutdown of the
electrochlorination installations in the event of an earthquake.
 a reduction in the releases of radioactive substances in the case of an accident with or without core melt,
more specifically by means of:
o the modification designed, in the event of a steam generator tube rupture (SGTR) accident, to
prevent the risk of the water in the affected steam generator from overflowing, thus significantly
helping to reduce the risks of releases associated with this accident;
o installing baskets of sodium tetraborate which would dissolve in the water recovered from the
reactor building sumps following a primary loss of coolant accident, thus enabling a basic pH to
be maintained, thereby limiting releases of radioactive iodine;
o the modification designed to increase the performance of the pump on the system reinjecting
into the reactor building any leaks collected from the safeguard systems which, in an accident
situation, carry highly contaminated fluid from the reactor building, thus enabling this function
to be used in a severe accident even when the pressure in the containment is high.
 reinforcement of the prevention of the risk of uncovering of the fuel assemblies stored in the fuel pool or
being handled, more specifically the modification designed to automatically isolate the pool cooling in-line
suction if a very low water level is detected in the pool.
ASN considers that all the modifications thus identified by EDF following the generic phase of the
periodic safety review of the 1300 MWe plant series reactors will contribute to making a significant
improvement in the safety of these facilities.
However, compliance with the requests made by ASN on various topics of the periodic safety review
and with EDF’s commitments made during the reviews are liable to require additional modifications,
depending on the results of the corresponding studies.
At this stage, ASN has identified nothing to compromise the ability of EDF to control the safety of the
1300 MWe reactors up until the periodic safety review associated with their fourth ten-yearly outage
inspection. However, ASN would recall that, in accordance with Article L. 593-19 of the Environment
Code, its position concerning continued operation will be given for each reactor after an analysis of the
conclusions report of its periodic safety review, transmitted by EDF further to its third ten-yearly
outage inspection4. As and when necessary, ASN may issue new prescriptions concerning the operation
of the reactor in question.
*
*
*
Appendix 1 gives details of ASN’s position statements on each of the topics of the VD3-1300 periodic safety
review, examined as part of the generic phase. Appendix 2 gives all the additional requests or the requests that
ASN considers should be maintained following analysis of your answers and Appendix 3 gives the requests
concerning the content of the safety report.
Sincerely yours,
ASN Chairman,
Pierre-Franck CHEVET
In accordance with letter ASN DEP-PRES-0077-2009 of 1st July 2009, these reports will be transmitted by EDF no later than 6
months after the end of the ten-yearly outage inspection of the reactor concerned.
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DISTRIBUTION LIST
External distribution of paper version:
 EDF:
o DPN/UNIE
o DIN
o DIN/CIPN
o DIN/SEPTEN
 IRSN: 1 copy
External distribution of electronic version:
 EDF: S. Walter
 IRSN: P. Lejuste
Internal distribution (electronic version)
 DCN: all DCN staff
 DEP
 All regional divisions in charge of regulating 1300 MWe PWR nuclear safety
DCN archival:
 DCN: Starting point
5 / 49
TABLE OF CONTENTS OF APPENDICES TO LETTER CODEP-DCN-2015-008144
APPENDIX 1 TO LETTER CODEP-DCN-2015-008144
DETAILS OF ASN POSITION STATEMENTS ON THE GENERIC STUDIES OF THE VD3-1300 PERIODIC SAFETY REVIEW
A.
CONFORMITY AND CONDITION OF 1300 MWE REACTORS.............................................................................................. 8
A.1.
A.1.
A.2.
A.3.
A.4.
B.
PROGRAMME OF THE CONFORMITY CHECK (ECOT) ....................................................................................................................... 8
VERIFICATION OF THE DESIGN OF THE CIVIL ENGINEERING STRUCTURES ................................................................................ 8
TEN-YEARLY TEST PROGRAMME .......................................................................................................................................................... 8
SUPPLEMENTARY INVESTIGATIONS PROGRAMME (PIC) .................................................................................................................. 9
SATISFACTORY CONTROL OF REACTOR AGEING ............................................................................................................................... 9
REASSESSMENT OF HAZARD-RELATED RISKS ................................................................................................ 10
B.1.
RISKS ASSOCIATED WITH EARTHQUAKES..........................................................................................................................................10
B.1.1. Reassessment of ground response spectra ......................................................................................................................................10
B.1.2. Reassessment of the seismic behaviour of the civil engineering structures ..............................................................................10
B.1.3. Reassessment of the seismic strength of the equipment ..............................................................................................................11
B.1.4. Study of earthquake-induced internal flooding ..............................................................................................................................11
B.1.5. Experience feedback from the Kashiwazaki-Kariwa earthquake ................................................................................................11
B.2.
RISKS ASSOCIATED WITH HIGH AIR AND WATER TEMPERATURE CONDITIONS .........................................................................12
B.3.
RISKS ASSOCIATED WITH THE FRAZIL ICE PHENOMENON ............................................................................................................12
B.4.
RISKS ASSOCIATED WITH STRONG WINDS .........................................................................................................................................13
B.5.
RISKS ASSOCIATED WITH TORNADOS ................................................................................................................................................13
B.6.
RISKS ASSOCIATED WITH HEATSINK LOWEST SAFE WATER LEVEL SITUATIONS ........................................................................14
B.7.
RISKS ASSOCIATED WITH EXTERNAL FLOODING ............................................................................................................................14
B.8.
PUMPING STATION HAZARD RISK FROM A DRIFTING HYDROCARBON SLICK.............................................................................15
B.9.
SITE’S ABILITY TO WITHSTAND COMMON MODE HAZARDS ...........................................................................................................15
B.10.
RISKS ASSOCIATED WITH INTERNAL FIRES IN THE FACILITIES .....................................................................................................15
B.11.
RISKS ASSOCIATED WITH EXPLOSIONS OF INTERNAL ORIGIN ......................................................................................................16
B.12.
RISKS ASSOCIATED WITH THE INDUSTRIAL ENVIRONMENT AND COMMUNICATION ROUTES ................................................16
B.13.
RISKS ASSOCIATED WITH AIR TRANSPORT ........................................................................................................................................17
B.14.
RISKS ASSOCIATED WITH THE ON-SITE TRANSPORT OF HAZARDOUS GOODS ...........................................................................17
B.15.
RISKS ASSOCIATED WITH ON-SITE FLOODING AND HIGH-ENERGY LINE BREAKS ....................................................................17
B.16.
RISKS ASSOCIATED WITH ELECTRICAL DISTURBANCES OF ON-SITE OR OFF-SITE ORIGIN.......................................................17
C. STUDIES OF OPERATING CONDITIONS OF 1300 MWE REACTORS AND THEIR RADIOLOGICAL
CONSEQUENCES .............................................................................................................................................................. 18
C.1.
REVIEW OF THE STUDIES OF OPERATING CONDITIONS AND THEIR RADIOLOGICAL CONSEQUENCES ................................18
C.1.1. Rules, methods and accident studies in the safety report .............................................................................................................18
C.1.2. Boron dilution risks .............................................................................................................................................................................18
C.1.3. Sensitivity analysis for passive failure of the safety injection system ..........................................................................................20
C.1.4. Impact of the behaviour of the secondary system valves on the coverage of design-basis transients .................................20
C.1.5. Prevention and mitigation of severe accidents ...............................................................................................................................20
C.1.6. Baseline requirements associated with the fuel criticality risk in spent fuel pools and the reactor building when the
reactor vessel is open ......................................................................................................................................................................................21
C.1.7. Elimination of reactor coolant system cold overpressure situations ..........................................................................................21
C.1.8. Assessment of the radiological consequences of accidents other than severe accidents ........................................................22
C.2.
FUEL BUILDING SAFETY REVIEW ........................................................................................................................................................23
C.2.1. Risks associated with the spent fuel pools (FB pools) ..................................................................................................................23
C.2.2. Handling of fuel transport packagings .............................................................................................................................................23
C.3.
SAFETY REVIEW OF EFFLUENT PACKAGING AND TREATMENT AUXILIARY BUILDINGS (BAC/BTE) ..................................24
D.
PROBABILISTIC SAFETY ASSESSMENTS (PSA) ................................................................................................ 25
D.1.
D.2.
E.
LEVEL 1 PSA ..........................................................................................................................................................................................25
LEVEL 2 PSA ..........................................................................................................................................................................................25
DESIGN OF SYSTEMS AND CIVIL ENGINEERING STRUCTURES ................................................................ 26
E.1.
SAFETY CLASSIFICATION OF EIP-S ....................................................................................................................................................26
E.2.
EQUIPMENT QUALIFICATION FOR ACCIDENT CONDITIONS..........................................................................................................26
E.2.1. Equipment qualification .....................................................................................................................................................................26
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E.2.2. Calculation of doses integrated by the equipment in the event of an accident with or without core melt .........................27
E.3.
DESIGN OF THE REACTOR DIGITAL INTEGRATED PROTECTION SYSTEM (SPIN) ....................................................................27
E.4.
MODERNISATION OF THE CONTROL ROOM – INTEGRATION OF ORGANISATIONAL AND HUMAN FACTORS ......................27
E.5.
CONDITION OF AND IMPROVEMENTS TO REACTOR CONTAINMENT ..........................................................................................28
E.5.1 Monitoring the condition and behaviour of the containments ....................................................................................................28
E.5.2. Improving the containment safety function ...................................................................................................................................28
F.
MODIFICATIONS TO THE FACILITIES AND THEIR OPERATING PROCEDURES ................................... 29
ASN REQUESTS
A.
MONITORING OF EDF COMMITMENTS ........................................................................................................... 30
B.
REASSESSMENT OF HAZARD-RELATED RISKS ............................................................................................... 30
B.1.
B.2.
B.3.
INTERNAL SEISMIC-INDUCED FLOODING .........................................................................................................................................30
RISKS ASSOCIATED WITH HIGH AIR AND WATER TEMPERATURE CONDITIONS .........................................................................30
CLASSIFICATION AND REQUIREMENTS APPLICABLE TO THE MEANS NECESSARY FOR ENSURING SECONDARY WATER
INDEPENDENCE FOR RIVERSIDE SITES ..............................................................................................................................................................31
B.4.
INTERNAL FIRE ......................................................................................................................................................................................32
C.
STUDIES OF OPERATING CONDITIONS AND THEIR RADIOLOGICAL CONSEQUENCES .................... 32
C.1.
C.2.
D.
BASELINE REQUIREMENTS ASSOCIATED WITH THE FUEL CRITICALITY RISK .............................................................................32
HANDLING OF FUEL PACKAGINGS.....................................................................................................................................................32
LEVEL 2 PROBABILISTIC SAFETY ASSESSMENTS ........................................................................................... 33
E. CONTINUOUS MONITORING OF THE LEAK RATE FROM THE INNER CONTAINMENT AND ITS
PENETRATIONS............................................................................................................................................................... 34
F.
REASSESSMENT OF THE CONTROL OF DRAWBACKS INHÉRENT TO THE INSTALLATION .............. 35
ASN REQUESTS CONCERNING THE CONTENT OF THE VD3 EDITION OF THE SAFETY REPORT FOR THE 1300 MWE
SERIES REACTORS
A.
DEMONSTRATION OF SATISFACTORY CONTROL OF THE RISKS OF AN ACCIDENT WITHIN THE
WASTE PACKAGING AND RADIOACTIVE EFFLUENT PROCESSING BUILDINGS (BAC/BTE) ........................ 36
B. DEMONSTRATION OF THE SATISFACTORY CONTROL OF THE ACCIDENT RISKS RESULTING
FROM POSSIBLE MALICIOUS ACTS THAT CANNOT BE RULED OUT ................................................................. 36
C.
PRESENTATION OF THE EIP AND THEIR DEFINED REQUIREMENTS ................................................... 37
D.
REFERENCING OF EIP EQUIPMENT SYSTEM ID NUMBERS ....................................................................... 37
E.
ADDITIONAL ACCIDENT STUDIES .................................................................................................................... 38
E.1.
E.2.
E.3.
STUDY CONCERNING THE INCORRECT POSITIONING OF A FUEL ASSEMBLY .............................................................................38
STUDY CONCERNING A STEAM LINE RUPTURE ................................................................................................................................38
STUDY CONCERNING CONTROL ROD CLUSTER EJECTION .............................................................................................................38
REFERENCES
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Details of ASN position statements
on the generic studies of the VD3-1300 periodic safety review
A. Conformity and condition of 1300 MWe reactors
A.1.
Programme of the conformity check (ECOT)
For the purposes of the periodic safety reviews, EDF carries out a check on the design and construction of its
facilities, known as the “unit conformity check” (ECOT), in order to run a targeted check on the conformity of
the reactors with their applicable baseline safety requirements, to detect any latent non-compliances and, as
applicable, correct them. The VD3-1300 ECOT program presented by EDF follows on from and supplements
the ECOT programmes run during the previous periodic safety reviews. It also draws on the orientations
adopted for the VD3-900 ECOT and on experience feedback from its application.
The orientations of the VD3-1300 ECOT programme were the subject of an initial ASN position statement
letter in 2011 [7] to which EDF responded by transmitting a new version of the programme, incorporating six
additional topics. Following the detailed examination of this new ECOT programme by IRSN at the request of
ASN, EDF made commitments in the letter in reference [8] in order to complete certain aspects of this
programme and clarify the implementation procedures. ASN considers that the ECOT programme adopted by
EDF, plus the subsequent commitments made, is acceptable subject the last requests concerning its
implementation procedures ASN formulated in the letter in reference [9] are taken into account.
ASN would also recall that in January 2015, it published ASN guide 21 [10] which explains its recommendations
regarding the application of the regulations concerning the procedures and deadlines for processing the
conformity deviations detected on a reactor.
A.2.
Verification of the design of the civil engineering structures
EDF verified the conformity of the design of the civil engineering structures with their safety requirements.
In the letter in reference [11], ASN considered the following to be satisfactory:
-
the selection method and the resulting list of civil engineering structures selected by EDF for the
conformity verification;
the verification studies carried out on the structures thus selected, except for justification of the ability
of the main steam valve bunker5 in the P4 series to withstand an external explosion.
ASN therefore asked EDF to complete its verification studies on the design of the main steam valve bunker
concerning this point and will examine EDF’s answers.
A.3.
Ten-yearly test programme
In addition to the surveillance testing of equipment as stipulated by the general operating rules and the functional
post-maintenance or post-modification qualification tests, EDF is examining the need for and benefits to be
gained from performing certain specific tests on the occasion of a ten-yearly inspection. By means of overall tests
or tests which can only be performed in particular facility configurations, these ten-yearly tests aim more
particularly to check that the required performance of certain systems has not been compromised by the
This structure, mainly consisting of a metal framework, houses the water-steam piping running along the outer wall of the
Reactor Building (RB).
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cumulative effect of successive modifications. The definition and then the performance of this test programme
thus participate in checking the facility’s conformity with its baseline safety requirements.
ASN has no particular comments regarding the methodology used by EDF on the occasion of the VD3 [12] to
draw up this test programme.
A.4.
Supplementary investigations programme (PIC)
The supplementary investigations programme (PIC) consists in running spot-checks on the condition of passive
equipment for which no inspection is provided for in the EDF basic preventive maintenance programme
(PBMP), nor in the regulation checks (for pressure equipment). As part of the periodic safety reviews, the PIC is
thus able to complete the overall summary of the condition of the facility, verify the adequacy of the PBMP and
improve them if necessary.
ASN has no particular comments concerning the PIC presented by EDF for the VD3-1300 periodic safety
review [13].
A.5.
Reactor ageing management
As from the VD3, the PIC is supplemented by inspections to ensure the satisfactory reactor ageing management.
EDF has established a methodology for controlling the ageing of its reactors after 30 years of operation, the aim
of which is to demonstrate their ability to continue to function until their VD4 in satisfactory conditions of
safety, on the one hand in the light of the condition of the facilities during their VD3 and, on the other, given the
knowledge and control of the mechanisms and kinetics of deterioration linked to ageing.
The method implemented by EDF for the VD3 for the 900 MWe series reactors was reused for the 1300 MWe
series reactors. This first of all consists in drawing up ageing analysis sheets (FAV) for the structures, systems and
components (SSC), the failure of which can have an impact on safety and which are liable to be affected by an
ageing mechanism, and in verifying whether the applicable maintenance and operational measures are
appropriate for the identified ageing mechanism. For each SSC susceptible to ageing, in other words for which at
least one FAV highlights that the ageing management has not in principle been demonstrated by the normal
maintenance and operating provisions, EDF conducts an in-depth assessment of the ageing management for the
coming ten-year period and presents the results and conclusions in a general continued operation aptitude file for
the SSC concerned (Equipment DAPE). Subsequently, on the occasion of the VD3 of each reactor and on the
basis of a summary of the generic “equipment DAPE” files and after integration of the specific aspects of the
reactor concerned, EDF produces a continued operation aptitude file for the reactor (reactor DAPE).
ASN has no particular comments concerning this ageing management approach for the reactors of the 1300
MWe plant series on the occasion of their VD3. After review of the FAV and the “Equipment DAPE” for the
reactors of the 1300 MWe plant series, ASN considers that the control of ageing of these reactors up until their
VD4 is satisfactory with respect to all generic aspects. This generic position in no way anticipates the final
individual position statement that ASN will subsequently be required to issue concerning each of the reactors
after the review of its reactor DAPE. ASN however considers that EDF can further improve its generic
approach and it submitted requests and observations accordingly in its letter in reference [14]. These requests
aim to improve the way certain ageing mechanisms are taken into account, through the creation of additional
generic FAV or “Equipment DAPE”, through the revision of certain FAV and through the changes to certain
inspections and maintenance procedures. ASN will also subsequently issue a position statement on the strength
of the 1300 MWe nuclear reactor vessels after 30 years of operation, following the forthcoming consultation of
the Advisory Committee for nuclear pressure equipment (GP-ESPN), scheduled for September 2015.
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B. Reassessment of hazard-related risks
B.1.
Risks associated with earthquakes
B.1.1. Reassessment of ground response spectra
EDF reassessed the seismic motion of the 1300 MWe sites in accordance with the RFS 2001-01 in reference [15]
and with ASN’s requests already submitted in 2011 in the letter in reference [16].
Given the current state of scientific knowledge6, ASN considers that the ground response spectra that EDF
intends to apply for the VD3-1300 periodic safety review are acceptable, with the exception of the spectrum for
Saint-Alban which is too weak and does not adequately cover the uncertainties associated with the data used by
EDF. For this site, ASN will examine the answers to the requests it made in the letter in reference [17]
concerning:
-
the reassessment of the ground response spectrum to take account of uncertainties;
the definition of a programme of work to verify the seismic resistance of the equipment and civil
engineering structures;
the completion of any modifications and seismic reinforcements, no later than 5 years after submission of
the periodic safety review conclusions reports (RCRS) as stipulated in Article L. 593-19 of the
Environment Code.
*
When the ground response spectrum for the safe shutdown earthquake (SSE) thus reassessed is higher than that
of the design basis earthquake (DBE) of the facility or than that of the SSE considered during the previous
periodic safety review, if this latter was already greater than the DBE, the compliance of the equipment and civil
engineering structures with the seismic strength requirements must be verified.
B.1.2. Reassessment of the seismic behaviour of the civil engineering structures
ASN considers that the results of the new floor spectra for the auxiliary safeguard buildings (BAS) and electrical
buildings (BL) for the 1300 MWe plant series are acceptable for the reassessment of the seismic strength of the
equipment in these buildings. ASN also considers that EDF’s general methodology for verifying the seismic
strength of the civil engineering structures on the 1300 MWe plant series sites is on the whole satisfactory, with
the exception of the possible direct use of structure damping ratios that are higher than the values recommended
in ASN guide ASN 2/01 in reference [18]. ASN will review EDF’s answers to the requests it made in the letter in
reference [19] concerning on the one hand the changes in EDF methodology on this subject of the damping
ratios and, on the other, concerning the repetition re-running of the studies to reassess the seismic strength
requirements for the turbine hall7 (SDM) on the 1300 MWe sites carried out with a damping ratio higher than the
value recommended for this type of structure in the guide in reference [18].
EDF also implemented the “seismic interaction” approach applied to civil engineering structures which
themselves constitute hazards. In the letter in reference [11], ASN considered the initial studies8 transmitted by
EDF to be satisfactory, except for the justification of the absence of seismic collapse of the H29 building on the
Paluel site. ASN therefore asked EDF to complete its seismic behaviour verification study for the H2 building
ASN points out that the probabilistic assessmetns of the seismic hazards transmitted by EDF have proven to be an important
tool for ruling on the acceptability of the spectra determined using application of the RFS 2001-01 basic safety rule, given current
knowledge.
7 The turbine halls in Nuclear Power Plants (NPPs) are covered by a seismic stability requirement to prevent the risk of their
collapse constituting a hazard for the adjacent safety classified buildings.
8 As the “seismic interaction” approach requires that the specificities of each site be taken into account (non–generic structures,
plot plans that differ from one site to another, etc.), the corresponding studies initiated by EDF have not yet been finalised and
will continue for several years to come, ahead of the VD3 on each of the sites.
9 Building housing numerous areas for tertiary and industrial uses, including the laundry and the demineralisation station. The
non-collapse of this building in the event of an earthquake is designed to prevent constituting a hazard for the essential service
water system (ESWS) galleries running below it.
6
10 / 49
on the Paluel site. The additional data provided by EDF in the letter in reference [20] are currently being
reviewed by IRSN and ASN.
B.1.3. Reassessment of the seismic strength of the equipment
ASN underlines the importance of the work done by EDF to establish an operational approach for reassessing
the seismic strength of the equipment, collating the characteristics, general design principles, lessons learned
from post-seismic feedback and the seismic behaviour diagnostic methods for the equipment most frequently
encountered.
ASN does however consider that the several points of the seismic reassessment approach proposed by EDF
must be modified before it is applied and it sent EDF the corresponding requests in the letter in reference [21].
It will then review the replies.
The complete implementation of the equipment reassessment approach thus modified could be incompatible
with the scheduled date of the VD3 for the first 1300 MWe reactors. ASN therefore also asked EDF to send it a
programme of work for performance of the studies and integration of any modifications and seismic
reinforcements for which the deadlines will need to be adapted to the potential consequences and will not exceed
5 years following submission of the RCRS.
B.1.4. Study of earthquake-induced internal flooding
In addition to reassessing the direct effects of the earthquake on the civil engineering structures and the
equipment for which there is a seismic resistance requirement, EDF supplemented its analysis of the ability of its
facilities to withstand internal flooding in the buildings caused by the simultaneous failure of tanks not designed
to withstand an earthquake.
After reviewing these studies, ASN considers that EDF’s demonstration of containment within the buildings of
the contaminated water released by the tanks that are not designed to withstand an earthquake is satisfactory.
With regard to the study of common mode failures caused by the internal flooding of the redundant equipment
necessary for reaching and maintaining a safe state for the reactors and fuel storage pools, ASN asked EDF in
the letter in reference [22] to provide additional justifications. ASN considers that the answers sent by EDF in
the letters in references [23] and [24] are satisfactory and indicates that there is no need to modify the facilities to
deal with the risk of internal earthquake-induced flooding.
However, the acceptability of the consequences of earthquake-induced internal flooding, whether with regard to
the containment of contaminated water or the absence of common failure modes, depends in particular on the
confirmation of the retention capacity of certain tanks and the non-degradation of this capacity by the direct or
indirect (falling loads) effects of the earthquake. EDF however considers that these seismic-induced flooding
studies are not a part of the demonstration of the safety. This position means that EDF does not identify these
retention basins as elements important for protection (EIP) and does not consider that their retention capacity as
used in the studies and their integrity in the event of an earthquake constitute safety requirements. ASN disagrees
with this EDF position and for its part considers that the demonstration of the safety for a nuclear facility with
respect to earthquakes is not limited to an assessment of the loadings induced by the seismic waves but must also
take account of hazards resulting from the failure of elements not designed to withstand an earthquake (falling
loads, internal flooding).
ASN therefore considers that the elements of the facility needed to justify the acceptability of the consequences
of seismic-induced internal flooding are EIPs and it thus formulated request 2 in Appendix 2 to this letter.
B.1.5. Experience feedback from the Kashiwazaki-Kariwa earthquake
Following the earthquake in Japan on 16th July 2007 close to the Kashiwazaki-Kariwa NPP, ASN asked EDF to
identify the lessons to be learned for the French reactors concerning the anomalies which occurred in the
Japanese NPP. The studies conducted by EDF on this point concerned:
11 / 49
-
the consequences of a large-scale fire in an electrical transformer following an earthquake;
the effect of water movements in the Reactor Building (RB) and Spent Fuel Building (FB) pools, induced
by the earthquake, on the gates and cofferdams in the pools as well as on the components of the fuel
loading machine;
assessment of the internal flooding associated with overflowing of the FB pool under the effect of the
wave induced by the earthquake.
After a review of the studies transmitted by EDF, with the support of IRSN, ASN considers that:
-
the design of the gates and cofferdams in the FB and RB pools, and of the fuel loading machines in the RB
and FB, satisfactorily cover the loadings resulting from the water movements induced by an earthquake;
the overflow volume of the FB pool that could result from a wave induced by an earthquake remains slight
when compared with the other causes of internal flooding already examined in the demonstration of the
safety.
The study transmitted by EDF concerning the potential safety consequences of a large-scale transformer fire in
the event of an earthquake, is under review and will be the subject of a subsequent position statement by ASN.
B.2.
Risks associated with high air and water temperature conditions
Following the heatwave episodes of 2003 and 2006, EDF drew up “extreme heat” baseline safety requirements
for each plant series, establishing an approach for a reassessment of the design of the reactors to deal with
periods of probable high summer heat, and their ability to deal with a rarer heatwave episode. The reassessment
of the probable high summer heat conditions constitutes the input data for the part of the “extreme heat”
baseline safety requirements concerning the facilities re-design approach. This approach aims to ensure that the
design of the air-conditioning systems means that in a high summer heat situation, they are able to maintain
temperature conditions in the buildings which do not compromise the availability of the equipment required to
deal with category 1 to 4 operating conditions of the safety report.
In the letter in reference [25] ASN gave its position on the “extreme heat” baseline safety requirements for the
CPY plant series and asked EDF to transpose its requests concerning this plant series to the baseline safety
requirements applicable to the 1300 MWe plant series, whenever pertinent. EDF replied to ASN’s requests in the
letter in reference [26] and in early 2014 presented an update of the “extreme heat” baseline safety requirements
for the 1300 MWe plant series [27].
ASN considers that the modifications envisaged by EDF through application of its “extreme heat” baseline
safety requirements will contribute to achieving a significant improvement in the design of the 1300 MWe
reactors and their ability to withstand the reassessed high air and water temperature conditions. Nonetheless,
some of EDF’s answers are not satisfactory and additional answers liable to have an impact on the “extreme
heat” baseline safety requirements for the 1300 MWe plant series are still awaited. In particular, in addition to the
modifications planned by EDF and for which deployment must not be delayed, ASN considers that it is essential
that EDF demonstrate its ability to keep a reactor in a safe state in the long-term post-accident phase of an
accident occurring at the same time as a period of high summer heat. On this point, ASN therefore formulated
request N° 3 in Appendix 2 to this letter.
B.3.
Risks associated with the frazil ice phenomenon
Based on available meteorological data, EDF initially assessed the susceptibility of the 1300 MWe plant series
sites to a risk of obstruction of the heatsink water intake by the appearance of frazil ice10 and then studied the
ability of the NPPs concerned to deal with such a hazard.
This complex phenomenon, which appears in the presence of particular meteorological and hydraulic conditions, leads to the
formation of ice crystals which can either aggregate to form sheets of surface ice (passive frazil) or adhere to submerged items
such as pre-filtration grilles and thus lead to icing-up of these items (active frazil).
10
12 / 49
Following its review, with the technical support of IRSN, ASN considered that the studies and protective
measures for the frazil ice phenomenon presented by EDF represent a significant step forward in safety, given
that they now take account of a meteorological hazard which was not originally considered in the design of the
NPPs of this series. ASN considered that the consideration given by EDF to a combination - during an episode
of extreme cold - of a frazil ice situation with a loss of offsite power (LOOP) was in particular satisfactory.
However, in its letter in reference [28], ASN asked EDF to supplement the susceptibility assessments of the
Cattenom, Flamanville, Penly and Paluel sites concerning the frazil ice phenomenon and to provide additional
justifications and improvements concerning the protection of the heatsink of certain sites.
In response to ASN’s requests, EDF (references [29] and [30]):
-
conducted a frazil ice susceptibility study on Mirgenbach Lake, concluding that the Cattenom site was
not susceptible to the frazil ice phenomenon;
plans to conduct additional investigations to determine the susceptibility to the frazil ice risk on the
Flamanville site and also plans to set up annual monitoring of water temperatures, in order to determine
whether or not the climate watch needs to be reinforced for the Paluel and Penly sites.
EDF also transmitted additional data concerning water temperature monitoring criteria triggering the
deployment of frazil ice protective measures and additional data concerning the effectiveness of these measures
on the Belleville site (winter recirculation11). ASN considers that these points are satisfactory.
EDF also intends to verify the effectiveness of the countermeasures envisaged for the Saint-Alban site
(shutdown of production pumps12); ASN will examine EDF’s conclusions.
With regard to the safety classification adopted for the frazil ice protection measures, EDF intends to apply the
approach presented in the letter in reference [31], aiming to define a classification and associated requirements
for equipment taking part in protection of the “cooling” safety function in the event of frazil ice. Without in any
way anticipating the results of the application of this approach, ASN already considers that the provisions
mentioned in request 8 of the letter in reference [28] should be classified in this way.
B.4.
Risks associated with strong winds
Even if the direct effects of strong winds were incorporated into the design of the buildings, through the
application of the “snow and wind” rules, the effects of strong winds on other elements of the facilities had not
been considered. EDF therefore studied the ability of the NPPs of the 1300 MWe plant series to withstand the
effects of strong winds over and above the simple resistance of the buildings and more particularly examined the
risks of hazards resulting from projectiles induced by the strong winds.
The strong wind speeds used by EDF are higher than those of the extreme wind conditions of the last 2009
edition of the “snow and wind” rules used for construction purposes. ASN considers that in the light of current
knowledge and the state of the art in force on this subject, this approach is a satisfactory means of characterising
the conditions associated with a meteorological hazard. With regard to the nature and characteristics of the
projectiles caused by strong winds adopted by EDF, ASN considers them to be satisfactory when taken in
conjunction with the consideration of other projectiles, such as steel balls and tubes, included in the baseline
safety requirements for protection against tornados (see section B.5 below). Finally, ASN considers that the
modifications planned by EDF on the basis of its studies will contribute to achieving a significant reinforcement
of the protection of its facilities against the effects of strong winds but that EDF will need to provide additional
data and it therefore formulated requests in the letter in reference [32], the answers to which will be reviewed by
ASN.
This system consists of injecting hot water from plant releases onto the grilles, to prevent frazil ice from forming on them.
EDF considers that shutting down pumps not classified as being important for safety (IPS) helps slow down the formation of
frazil ice and prevents complete clogging of the pumping station.
11
12
13 / 49
B.5.
Risks associated with tornados
The protection of reactors against tornados, not considered in the design of these facilities, is one of the subjects
adopted at the request of ASN during the orientation phase of the VD3-1300 periodic safety review.
Following the technical review of the baseline safety requirements developed by EDF concerning the
methodology for taking account of the tornado hazard, ASN considers that the characteristics of the reference
tornado adopted by EDF, the associated projectiles and the safety objectives for reactor protection against the
direct or indirect effects are satisfactory, subject the requests it formulated in the letter in reference [33] are taken
into account. In accordance with Article L. 593-18 of the Environment Code, which requires that the periodic
safety review allow the risk assessment to be updated, in particular taking account of the evolution of knowledge,
ASN asked EDF in its letter in reference [33] to ensure that the reactors of the 1300 MWe plant series are
protected against the tornado risk by implementing its baseline safety requirements on these reactors, without
waiting for the next periodic safety review and it will analyse the deployment schedule envisaged by EDF.
B.6.
Risks associated with heatsink lowest safe water level situations
The lowest safe water level (PBES) situations correspond to minimum heatsink water level or flow rate
conditions at the pumping station so as to the operation of the ESWS is insured.
EDF has defined a methodology for characterising PBES situations and checked that, in these situations, the
heatsink characteristics did not compromise the availability of the ESWS for the reactors.
The methodology for characterising PBES situations was covered in an initial ASN position statement letter in
2011 (reference [34]), in which ASN asked EDF to modify several points of its approach. The update of the
PBES characterisation method transmitted in return by EDF in 2013 does not address all the requests made by
ASN. Furthermore, the studies to justify the availability of the ESWS in a PBES situation transmitted by EDF
were carried out on the basis of the approach not yet updated. ASN therefore asked EDF in the letter in
reference [35]:
-
to revise its PBES characterisation methodology in accordance with the requests made in 2011,
to transmit an inventory of the protection of the 1300 MWe reactors, with application of the revised
methodology.
The first replies to this letter transmitted by EDF are currently being reviewed.
B.7.
Risks associated with external flooding
When determining the generic phase of the VD3-1300 periodic safety review, the baseline requirements adopted
for protection of the 1300 MWe sites against external flooding were based on the "Le Blayais experience
feedback" methodology developed following the partial flooding at Le Blayais NPP in 1999 and the additional
requests formulated by ASN in the letter in reference [36].
In 2013, ASN published a guide based on new knowledge to ensure that the external flooding risk is taken into
account more exhaustively and more robustly. By comparison with the baseline safety requirements initially
adopted for definition of the VD3-1300 periodic safety review, significant changes in the state of the art were
introduced into this guide and the application of all of these recommendations requires that EDF carries out
extensive studies which, for certain reactors, may imply integration time-frames that are incompatible with the
VD3 dates.
Consequently and in accordance with Article L. 593-18 of the Environment Code, which requires that the
periodic safety review allows a ten-yearly update of the risk assessment, more specifically taking account of new
knowledge available, ASN asked EDF, in the letter in reference [37], to reassess the protection of these reactors
against flooding, for a 10-year period (2013-2023), on the basis of the provisions of the guide and to propose an
implementation schedule, giving priority to either the VD3-1300 periodic safety review for each of these reactors
or, with regard to an overall hazard affecting a site, the VD3-1300 periodic safety review of the last reactor on
the site.
14 / 49
In the letter in reference [38], EDF sent ASN a schedule with a time-frame conforming to its request.
B.8.
Pumping station hazard risk from a drifting hydrocarbon slick
ASN urged EDF to examine the robustness of the NPPs to a drifting hydrocarbon slick as this hazard liable to
affect the pumping stations on the sites was not included in the NPP design.
After a technical review, ASN underlines the fact that the studies and tests performed by EDF led to a clearer
understanding of the impact of hydrocarbons on the heatsink equipment and considers that the measures
associated with the approach presented by EDF offer appropriate protection. ASN in particular considers that
the additional protective measures planned by EDF on certain sites will improve the safety of the sites
concerned. ASN will nonetheless examine EDF’s replies to its requests formulated in the letter in reference [39]
concerning the justification of the effectiveness of the cleaning systems and the time needed by the operators to
deploy the mobile protective barriers. ASN also asked EDF to assign hazard resistance and maintenance
requirements to the physical protections of the heatsink against a hydrocarbon slick and it will verify that these
requirements have been taken into account.
B.9.
Site’s ability to withstand common mode hazards
EDF studied the ability of the NPPs, following an external hazard, to manage the consequences of a loss of heatsink (situation H1), a loss of offsite power (LOOP), or a combination of the two, simultaneously affecting all the
reactors on the same site.
Following its review, with the support of IRSN, ASN considered that the site independence studies presented by
EDF represent a significant improvement in the way common mode external hazard risks are taken into account
for a site. However, additional data was still required and ASN sent EDF a number of requests in the letter in
reference [40], to which EDF replied with the letters in reference [41] to [43]. ASN considers that all of the
responses provided by EDF are satisfactory, with the exception of the answers concerning the classification
approach and the requirements associated with the measures included in the “site H1” studies.
This point is the subject of requests N° 4 and 5 in Appendix 2 to this letter.
ASN had also requested that these studies be revised, taking account of the study rules applicable to accidents of
the design extension conditions. EDF intends to modify these studies for deployment of the hardened safety
core. ASN has no objection to reviewing the corresponding modifications within this framework.
B.10. Risks associated with internal fires in the facilities
EDF carried out studies to reassess the demonstration that the risks of internal hazards associated with fire are
satisfactorily controlled, with regard to:
-
the effects of smoke on the operation of the equipment;
the impact of fire-induced pressure effects on the fire sectorisation;
the fire-design of fire sectorisation elements.
ASN underlines the progress made in understanding the impacts linked to pressure effects. EDF is nonetheless
required to continue its efforts on certain points and ASN formulated requests in the letter in reference [44] to
which EDF provided the first answers in the letters in reference [45] to [49]. These answers require detailed
analysis and further exchanges are already planned.
With regard to the impact of smoke on the operation of equipment, EDF does not envisage reviewing certain
fire sectorisation justifications based on an analysis of the harmlessness of the propagation of smoke and simply
proposes continuing its research work. ASN considers that EDF’s reply is not satisfactory and reiterates its
request in Appendix 2 of this letter (see request N° 6 in Appendix 2 to this letter).
15 / 49
With regard to pressure effects, EDF transmitted a report presenting its method for identifying the fire volumes
in which a fire is liable to cause pressure variations such as to break down the fire sectorisation. Answers are
however still being awaited concerning the list of safety fire volumes identified with this method and the
associated action plan.
With regard to the fire design of the “fire sectorisation” elements, EDF undertook, in a letter in reference [46],
to reinforce the fire protection of rooms LC0508 and LC0708 of safety fire sector SFS L0880 and the common
mode fire protection for the cabling and mechanical equipment in area RB1007 of safety fire zone ZFS R068, so
that they are appropriate for the fires which could occur there; the RCRS for the P4 series reactors will specify
their deployment time-frames.
ASN also asked EDF to issue a position statement on the possibility of replacing the current method for
justifying the fire design of the fire sectorisation elements by the new method developed for the Flamanville 3
EPR reactor, known as the EPRESSI method. In a letter in reference [50] EDF stated that this method cannot
be industrially transposed to all the reactors in operation. ASN duly notes that it is at present impossible to
replace the current method for justifying the fire design of all the fire sectorisation elements for the reactors of
the 1300 MWe plant series by the new methods developed for the Flamanville 3 EPR reactor. ASN does
however consider that there are shortcomings in the current method that require its replacement as soon as
possible. In the letter in reference [51], ASN has therefore already asked EDF to propose an alternative method
before the end of 2015 for justifying the fire sectorisation and considers that EDF will need to apply this new
method for the reactors of the 1300 MWe plant series without waiting for the periodic safety review associated
with their VD4. On this point, ASN therefore formulated request N°7 in Appendix 2 to this letter.
B.11. Risks associated with explosions of internal origin
EDF carried out studies to reassess the demonstration that the nuclear island or site internal explosion risks are
satisfactorily controlled.
ASN considers that the studies performed and the modifications planned are on the whole satisfactory. The
studies performed for this review now in particular cover all the explosion risks concerning the pipes carrying
hydrogen outside the nuclear island and the explosion risk linked to the electrochlorination process. ASN does
however consider that EDF must complete these studies with regard to certain points and it formulated requests
in the letter in reference [52]. EDF has transmitted its first answers in the letter in reference [53] and they are
under review.
EDF was asked to examine the potential benefits to be gained from installing a system for early detection of
abnormal hydrogen releases on the pipes in the technical galleries and for mitigating the consequences of these
situations. It replied that the leak prevention measures in place as a result of the design and in-service monitoring
requirements for these pipes are sufficient to obviate the need for installation of the detection systems
mentioned. ASN considers that this particular case of the technical galleries needs to be reviewed in the light of
the complementary studies to be performed later in response to its request concerning the evaluation of the
consequences of hydrogen leaks other than from the removable components on the systems carrying hydrogen.
B.12. Risks associated with the industrial environment and communication routes
The risks linked to the industrial environment and communication routes were reassessed by EDF with
application of basic safety rule (RFS) I.2.d [54] and taking account of updated accident rates data.
ASN considers that the approach adopted by EDF for reassessment of the risks linked to the industrial
environment and communication routes is consistent with the framework set by RFS I.2.d [54]. ASN also
considers that the formulas used by EDF for its probabilistic models concerning the annual frequency of
occurrence of a road, rail, river and maritime transport accident involving hazardous substances liable to damage
the safety functions are satisfactory. However, ASN will review how EDF takes account of the requests it
formulated in the letter in reference [55] aimed at improving its risk reassessment studies:
16 / 49
-
to make more appropriate use of certain accident rates data in its probabilistic calculations of the annual
frequency of occurrence of a road and rail transport accident,
by taking account of the risk of the formation and drifting of a cloud of flammable gas, for oil pipelines
carrying volatile refined products,
by continuing to collect and analyse hazard assessments for industrial facilities subject to licensing and
situated around the NPPs,
by modifying and supplementing its modelling of the on-site effects of the explosion of a drifting gas
cloud.
B.13. Risks associated with air transport
The risks of an airplane crash were reassessed by EDF with application of basic safety rule RFS I.2.a [56]. Over
and above the three aviation types13 mentioned in RFS I.2.a, the reassessments transmitted by EDF also for the
first time took account of the risks associated with aircraft taking part in fire-fighting operations, which have the
particularity of operating at low altitude.
ASN considers that the mathematical models for calculating impact probability adopted by EDF, as well as the
data used concerning accident rates and air traffic, are satisfactory. ASN also considers that the risk assessment
concerning fire-fighting aircraft is on the whole satisfactory. This additional assessment concludes that there is a
negligible contribution by this new family of aircraft in terms of the probabilistic objectives of RFS 1.2.a.
However, ASN considers that the approach presented by EDF to identify the targets to be protected with regard
to the storage of spent fuel needs to be completed and it formulated requests to this effect in the letter in
reference [57]. It will be reviewing how these requests are addressed.
B.14. Risks associated with the on-site transport of hazardous substances
The reassessment studies for demonstration of satisfactory control of risks associated with the on-site transport
of hazardous substances were initiated by EDF before the publication of the 7th February 2012 order, amended
[5]. These studies transmitted by EDF are based on a probabilistic assessment or the analysis of the physical
likelihood of the accident scenarios which could result from an accident involving a truck delivering hazardous
substances inside a nuclear power plant, while driving, or during loading or unloading operations.
In accordance with the provisions of the above-mentioned order, these risk assessments for the on-site transport
of hazardous substances must first of all be based on a deterministic assessment. In the letter in reference [8],
ASN therefore asked EDF to complete its probabilistic assessment of the accident scenarios with a deterministic
assessment of their effects and the facility targets exposed to these effects. It will review this complementary
assessment.
EDF’s answers to this letter have not yet been transmitted to ASN.
B.15. Risks associated with on-site flooding and high-energy line breaks
For the reassessment of the consequences of on-site flooding and high-energy line breaks (HELB), EDF
transmitted justifications to demonstrate that the studies of the common mode failures caused by these on-site
hazards produced for the previous periodic safety review (VD2-1300), for the reactor at power states, are also
able to cover the reactor shutdown states.
After reviewing the additional justifications transmitted by EDF in the letter in reference [59] in response to the
ASN letter in reference [22], ASN considers the on-site flooding and HELB studies to be satisfactory.
13
General, commercial and military aviation
17 / 49
B.16. Risks associated with electrical disturbances of on-site or off-site origin
EDF sent ASN studies concerning the robustness of the NPPs to on-site and off-site electrical disturbances.
Following ASN’s request in the letter in reference [60], EDF sent a study of the failure of the alternator
excitation system, in the letter in reference [61] and, in the light of the results, indicated that it intended to
implement a modification on the occasion of the VD3-1300 periodic safety review to improve reactor protection
against electrical disturbances associated with this type of event.
ASN considers that the complementary data provided by EDF, in terms of studies and modifications, provide a
satisfactory response to its request.
C. Studies of operating conditions of 1300 MWe reactors and their radiological
consequences
C.1.
Review of the studies of operating conditions and their radiological consequences
C.1.1. Rules, methods and accident studies in the safety report
EDF has improved the safety demonstration concerning the design-basis operating conditions14, modifying
several methods, data and hypotheses. EDF in particular corrected study anomalies and took account of the
improvements which appeared necessary during the recent fuel management reviews. Certain studies are still
being completed in order to constitute a complete demonstration, in particular those concerning the intermediate
break primary loss of coolant accident (IB-LOCA), the steam line break (SLB) accident and the control rod
cluster ejection (EDG) accident. These studies will be reviewed subsequently.
ASN will also review how the requests it formulated in the letter in reference [62] are taken into account, after
examination by IRSN and consultation of the GPR on 15th and 16th October 2014, aiming to:
-
demonstrate that the required boron concentrations in the hot shutdown state can avoid all return to
power during an uncontrolled cluster withdrawal transient at zero power (RIGZ);
study the feasibility of adding a requirement for no return to criticality for category 2 cooling incidents
initiated in shutdown state;
accelerate the deployment of the equipment modification to the primary pressure control and, pending
actual deployment of this modification, implement compensatory measures;
assess the uncertainties associated with the model of the flow rate at the breach and incorporate them
into the steam generator tube rupture (SGTR) studies;
identify cases of control cluster withdrawal at power (R1GP) liable to lead to a prolonged boiling crisis
and, as necessary, define an equipment modification able to prevent this risk;
lower the equivalent iodine 131 thresholds in the radiochemical specifications requiring reactor
shutdown.
The requests concerning the reactor coolant dilution risks are described in detail in section C.1.2. below.
C.1.2. Boron dilution risks
The boron dilution risks liable to lead to uncontrolled reactor divergence or a power excursion, linked to
scenarios of homogeneous dilution, heterogeneous dilution of either external origin or inherent in the LOCA,
were the subject of several requests from ASN in the letters in reference [62] to [65].
The safety demonstration in particular comprises an assessment of normal and accidental situations with which a reactor could
be faced. These situations are placed into 4 categories according to their probability of occurrence. These are the design-basis
operating conditions.
14
18 / 49
ASN considers that these risks require that EDF provides significant additions to the safety demonstration and
may need to make new modifications. EDF must in particular demonstrate compliance with the safety criteria in
the homogeneous dilution studies for all reactor states, after correction of all the identified study anomalies [63].
For the “reactor producing” state, ASN asked EDF, in the letter in reference [62], to provide a safety
demonstration based on the actions requested in the degraded and emergency operating procedures.
The justifications transmitted by EDF in the letters in reference [66] to [69] are currently being reviewed.
Pending the end of this review, in the letter in reference [63], EDF requested application of the following two
operating measures able to guarantee compliance with the safety demonstration:
-
return to a primary system cooling rate limit in shutdown states of 28 °C/h;
in the operating technical specifications, stipulate the maximum cooling rate value guaranteeing that
there is no risk.
EDF agreed to apply these provisions in the letter in reference [66].
*
In the letter in reference [63], ASN also asked EDF to take account of the heterogeneous boron dilution
scenario due to a non isolable rupture of a heat exchanger tube on the primary pumps seal system (CEPP) in the
“reactor producing” and “normal shutdown” operating ranges in the nuclear safety demonstration. ASN
considers that a large amount of data is still needed to be able to rule out the risk of criticality resulting from this
dilution scenario, on the basis of the studies, and it will review the extent to which the requests it formulated in
the letter in reference [64] have been taken into account concerning:
-
the study of the equipment or operational modifications aimed to make the functional sequence linked
to this scenario acceptable;
assessment of the adequacy of these modifications from the probabilistic viewpoint.
*
In the letter in reference [70], EDF transmitted a note presenting its approach to study the inherent
heterogeneous dilution in the event of an intermediate break primary loss of coolant accident (IB-LOCA). ASN
assessed this approach and its application to the 1300 MWe reactors, transmitted by the letter in reference [71],
with the technical support of IRSN. This assessment led to the transmission of the letter in reference [65] which
presents a number of ASN reservations concerning the approach, in particular with regard to the following
points:
-
the validation of certain thermohydraulic software models able to represent the primary and secondary
systems and having an influence on the simulations of the LOCA transients, in the light of the
dominant physical phenomena of inherent dilution;
the ability to transpose to the French nuclear reactors the limitation of the water plug to the volume of
the Steam Generator (SG) outlet channel head and the crossover leg, as observed on the tests
performed with the PKL experimental loop;
the conservatism boron concentration at the core inlet calculated using CFD15 software.
ASN therefore asked EDF:
-
to evaluate the possible consequences on core reactivity and fuel behaviour of the passage of a low
boration water plug in the core of a 1300 MWe reactor, according to its estimated volume;
CFD: computational fluid dynamics, consists in studying the movements of a fluid, or the effects of such movements, through
numerical resolution of the equations governing the fluid.
15
19 / 49
-
if the consequences of such an accident were to prove unacceptable, to study the advantages and
drawbacks of the modifications that could be envisaged in order to avoid this situation or mitigate its
consequences to an acceptable level.
C.1.3. Sensitivity analysis for passive failure of the safety injection system
The systems involved in demonstrating the control of design-basis accidents must be able to carry out their
functions despite a single failure affecting any one of their components, whether active (pumps, valves, etc.) or
passive (piping, tanks, etc.). On the reactors of the 1300 MWe plant series, the passive single failure of the
reactor safety injection system (SIS) is considered, in accordance with RFS I.3.a of 5th August 1980, postulating a
leak of 200 L/min at the time of transition to recirculation16, then isolated in 30 minutes.
EDF checked the absence of cliff-edge effects17 on the radiological consequences of a LOCA accident,
considering that isolation of the 200 L/min leak only becomes after one hour.
ASN considers that the sensitivity analysis presented by EDF concerning the time needed to isolate a leak
associated with passive failure of the SIS is satisfactory and concludes that doubling of the isolation time
considered in the safety reports does not lead to a significant increase in the radiological consequences of the
LOCA [72].
C.1.4. Impact of the behaviour of the secondary system valves on the coverage of
design-basis transients
The rapid closure of all the steam isolation valves (VIV) of the steam generator (SG) constitutes a design-basis
operating condition for the risk of overpressure in the secondary system and is a design-basis for the safety
valves on this system.
In 2004, the inadvertent closure of a VIV on Cattenom reactor 2 led to unexpected opening of all of the seven
safety valves on one of the SGs, whereas the safety assessments indicated that only five valves would be needed
to protect the secondary system.
EDF then implemented an actions plan to improve the reliability of the VIV as of 2007. This consisted in taking
operating and maintenance preventive measures and in deploying a number of equipment modifications on the
sites between 2013 and 2018. ASN considers that the effectiveness of this actions plan must be measured by
means of regular monitoring of significant safety events (ESS) involving inadvertent closure of VIV, as and when
the modifications are deployed and, in the letter in reference [73], it asked EDF to submit experience feedback
about significant events, modifications and the adequacy of the action plan, every two years.
In 2009, further to analysis of this event, in a letter in reference [74], ASN asked EDF to identify the potential
consequences of this event on the design-basis operating conditions and to specify any envisaged remedial
solutions. The study concerning the consequences of this event on an SGTR type accident is currently being
reviewed by ASN. In 2014, in the letter in reference [73], ASN issued another request for EDF also to assess the
potential consequences of opening of a higher number of valves than currently specified in the operating
conditions of the design extension conditions18 and will incorporate EDF’s answers into its assessment.
In the event of a reactor coolant system break (primary loss of coolant accident – LOCA) which cannot be compensated for by
the chemical and volume control system (RCV), the SIS system must cool the reactor core and make up the water lost through
the break. The water needed for this is initially taken from the reactor cavity and spent fuel pit cooling and treatment system tank
(PTR). When the water in the tank is exhausted, the SIS system automatically switches to “recirculation” mode: it then retrieves
the water from the sumps at the bottom of the reactor building.
17 Sudden alteration in the behaviour of a facility, sufficient to cause a slight modification in the scenario envisaged for an
accident, the consequences of which are then made seriously worse.
18 In the 1970s, more complex accident situations than the design-basis operating conditions were defined to cover more complex
trigger events (combinations of failures). These new operating conditions are called the operating extension conditions.
16
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C.1.5. Prevention and mitigation of severe accidents
On this topic, EDF focused on improving containment and reducing the occurrence of scenarios which could
lead to early large-scale releases.
The studies produced by EDF on the prevention and mitigation of the consequences of severe accidents were
the subject of a consultation of the GPR on 28th March 2013. Following this review, ASN considers it pertinent
that EDF look at two loadings corresponding to partial core melt and total core melt in order to verify the ability
of each equipment item to withstand severe accident conditions. Moreover, in the light of current scientific
knowledge and international best practices, ASN considers that the severe accident management strategy
proposed by EDF and based on flooding of the reactor pit following activation of the containment spray system
(EAS), when it is available, is acceptable. Generally speaking, the modifications proposed by EDF to improve the
containment and reduce the risks of early large-scale releases, are felt to be pertinent. Some are still undergoing
detailed assessment. ASN does however consider that the baseline safety requirements for severe accidents need
to be supplemented and will verify that EDF has taken account of the requests it formulated in the letter in
reference [75]:
-
to complete the list of equipment necessary in case of a severe accident, including a number of circuits
and measurement systems;
to define an approach to identify the operating limits of the useful equipment in a severe accident;
to study the redundancy and diversification of opening of the venting-filtration system required in the
event of a severe accident (filter U5);
to use detection of reactor vessel bottom head melt-through in the severe accident intervention guide.
ASN would in particular recall that, with the goal of ensuring continuous improvement, EDF must define
qualitative radiological objectives in the severe accidents baseline requirements and that it must justify the
adequacy of the provisions adopted for the VD3-1300 periodic safety review in the light of the objective defined
in Article 1er.2 of the Order in reference [5].
C.1.6. Baseline requirements associated with the fuel criticality risk in spent fuel
pools and the reactor building when the reactor vessel is open
The purpose of the criticality baseline requirements is, for all the nuclear reactors in operation, to specify the
measures taken to prevent the occurrence of a criticality accident in the FB and the RB when the reactor vessel is
open.
In 2007, following its analysis of the first version of the criticality baseline requirements, ASN asked EDF, in the
letter in reference [76], to supplement the baseline requirements and include them in the safety reports. EDF
transmitted an update of the baseline requirements taking account of a certain number of these requests.
ASN considers that the reactivity control studies concerning the situations considered in the criticality baseline
requirements in the RB with vessel open, must be included in the reference operating conditions studies in the
safety report and formulated requests along these lines in the letter in reference [77]. ASN in particular considers
that EDF needs to define acceptability criteria regarding the control of reactivity and mention them in the safety
report, in the same way as the other criteria applicable to phases A19 and B20 of the reference operating
conditions studies. These criteria will need to be modulated depending on whether or not there are any means of
detection and to take account of the unavailability of scram.
The answers transmitted by EDF in the letters in reference [78] are currently being reviewed.
Finally, ASN recalls that the requests in the letter in reference [76] concerning the scenarios involving a fuel
assembly falling to the bottom of the pool followed by perforation of the liner and the qualification of the
scientific calculation software used for the criticality studies regarding scenarios in the FB, have not yet been
answered by EDF. ASN formulated request N° 8 in Appendix 2 for transmission of a calendar for the replies.
Phase A is the phase between the initial moment of the accident and the time of first intervention by a protection system or the
first manual action following an alarm sheet (in this second case, there is no phase B as defined below).
20 Phase B is the phase between the time of the first intervention by a protection system and the moment of the first manual
action.
19
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C.1.7. Elimination of reactor coolant system cold overpressure situations
A cold overpressure situation corresponds to the pressurisation of the main primary system (CPP) higher than
100 bar, while the temperature of the reactor vessel metal is below its ductile to brittle transition temperature
(RTNDT) which is about 90°C. A situation such as this is liable to breach the vessel and, in accordance with
Article 3.9 of the Order of 7th February 2012 in reference [5], preventive measures have to make these situations
extremely improbable, with a high degree of confidence.
On the basis of the results of its probabilistic quantification assessments of these situations and the associated
sensitivity analysis, EDF concluded that the current design and operating provisions for its reactors do indeed
confirm that this risk is “practically eliminated”.
After a detailed analysis with the support of IRSN, ASN considers that the studies transmitted by EDF do
indeed confirm that the steps taken to prevent cold overpressure risks are satisfactory, with the exception of a
family of breach scenarios in operating states with the residual heat removal system (RRA) connected. ASN
considers that, as they stand, the demonstration data presented by EDF are insufficient to validate the practical
elimination of these particular sequences and also notes that the modifications made on the reactors of the
900 MWe series led to a significant reduction in the cold overpressure risks associated with these sequences. In
the letter in reference [79], ASN asked EDF to study the implementation of modifications designed to further
reduce the probability of the cold overpressure risk associated with these particular accident sequences.
In the letters in reference [80] and [81], EDF replied that it was not in favour of carrying over to the 1300 MWe
plant series reactors the operating modification implemented on the 900 MWe series reactors, which consists in
intervening on the reactor coolant system pressuriser valves at each outage, in order to lower the opening
threshold in states in which the reactor is connected to the RRA. EDF did however duly note the inadequacy of
the demonstration data transmitted to date for certain accident scenarios which could lead to cold overpressure
and intends to supplement its studies ahead of the submission of the first periodic safety review conclusions
reports (RCRS).
C.1.8. Assessment of the radiological consequences of accidents other than
severe accidents
Following two meetings of the GPR in 2006 and 2009, EDF revised the studies of the radiological consequences
of design-basis accidents.
Following the review conducted with the technical support of IRSN, ASN notes that EDF continued to bring
the hypotheses used for the reactors in operation closer into line with those of the EPR. However, there is still
no provision for including the specific features of the site into the baseline safety requirements. Technically
speaking, ASN considers that EDF must continue its investigations and improve its justification of the values
used concerning release rates for fission products in the event of a primary loss of coolant accident (LOCA).
ASN sent EDF a number of requests in the letter in reference [82] and will check that they are taken into
account. ASN in particular asked EDF to assess the following for each NPP:
-
its specific radiological consequences, taking account of local meteorological conditions and the
environmental characteristics of the site as well as the lifestyles of the populations living in the vicinity;
contamination of foodstuffs produced in the vicinity and of surfaces.
More generally, ASN asked EDF to:
-
clarify certain points of the radiological consequences study baseline;
supplement the methodology for evaluating the radiological consequences of atmospheric releases;
repeat the studies, after modifying certain hypotheses;
supplement the studies by taking account of various hazards.
Furthermore, in the letter in reference [62], ASN adopted a position on the SGTR design-basis operating
conditions, studied in 3rd and 4th category, for which EDF has initiated an action plan to reduce the radiological
consequences (automatic isolation of the supply to the SG affected). ASN considers that the steps taken by EDF
in this actions plan significantly improve the safety of the 1300 MWe reactors. Nonetheless, additional data is
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required, as mentioned in section C.1.1. of this letter. ASN thus asked EDF to confirm the margins before water
overflow of the SG affected by the SGTR, taking account of the uncertainties associated with the breach flow
rate model. If the above-mentioned margins are confirmed, EDF could replace the 4th category SGTR study with
the study of:
-
an SGTR accident combined with mechanical blockage of the GCTa21 valve in the closed position and
the blockage of a secondary valve in the open position, in the design extension conditions;
a 3rd category SGTR accident combined with a loss of offsite power (LOOP).
C.2.
Fuel building safety review
C.2.1. Risks associated with the spent fuel pool (FB pool)
EDF sent ASN reassessments of the safety of spent fuel storage in the FB pools of the 1300 MWe series NPPs.
The purpose of these reassessments is to identify the scenarios leading to the total loss of cooling and the
accidental and rapid emptying of the fuel building pool, as well as to define countermeasures to prevent these
risks. On the basis of the studies carried out, EDF considers that for the 1300 MWe reactors, the
implementation of physical modifications equivalent to those deployed on the 900 MWe nuclear reactors on the
occasion of the periodic safety review associated with their VD3, is sufficient.
ASN considers that the modifications planned by EDF or already carried out on the reactors of the 1300 MWe
series (automated closure of the PTR22 intake valve, shutdown of PTR pumps at extremely low level and increase
in the diameter of the siphon-breaker devices) significantly improve the safety of the storage and handling of the
fuel in the pool. ASN nonetheless considers that additional measures are required to prevent the risks of loss of
cooling and accidental emptying of the FB pool and it will check that EDF has taken account of the requests it
formulated in the letter in reference [83] aimed at:
-
studying the damage caused by internal flooding or a fire on the equipment necessary for cooling of
the pool and evaluating countermeasures;
for the reactors of train P4, in the case of simultaneous loss of two PTR pumps caused by a fire,
checking that in the case of propagation of smoke, the personnel will be able to reach the valve control
area in a clear atmosphere;
checking the geometrical tolerances and the conformity of the sensitive elements of the transfer tube;
qualifying the siphon-breaker device or, failing which, quantifying the benefit to be gained from a
modification similar to that adopted on the Flamanville 3 reactor;
classifying in group A23 the criterion associated with the fuel building pool make-up water flow rate
necessary to guarantee restart of a PTR pump.
C.2.2. Handling of fuel transport packagings
In 2011, in the letter in reference [1], ASN asked EDF to conduct an exhaustive review of the satisfactory
control of the risks arising from falls by 1300 MWe fuel transport packagings during their handling24 in order to
complete the demonstration of the safety.
GCTa: Turbine Bypass System
PTR: reactor cavity and spent fuel pit cooling and treatment system
23 Group A comprises the test criteria which, if not met, compromise one or more safety objectives. They are the result of safety
studies or are representative of the unavailability of required equipment (availability or performance compromised for the
duration of the mission).
24 The risks associated with the operations to remove spent fuel assemblies differ according to the plant series:
on the P4 series reactors: the transport packagings, each of which weighs more than 100 tonnes, are handled to a
height of 27 m above the fuel building floor at the level of the fuel packaging handling hatch.
on the P’4 series reactors: there are no lifting operations in the fuel building. The risk of dropping is restricted to the
handling performed with the site’s overhead crane. During these operations, the transport packaging is not equipped
with its shock-absorbing covers and the main risk is the loss of mechanical confinement of the packaging following its
fall.
21
22
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After examination of the studies transmitted by EDF, ASN considers that they permit to conclude to the
absence of any risk of loss of integrity of the storage pool in the event of a fall by a packaging (P4 series) but that
there is no demonstration that the mechanical containment of the packaging would be maintained in a particular
fall configuration. ASN also considers that it is necessary to check that the dynamic containment of the fuel
building remains effective when the transport packaging handling opening is open and if it cannot be
demonstrated that the packaging integrity is maintained in the case of a fall during handling. ASN thus sent
additional requests in the letter in reference [84], to which EDF replied by the letters in reference [85] and [86].
EDF intends to take organizational measures to reduce the risk of the loss of packaging integrity in the event of
a fall when the fuel building is open and the means of transport is in the vertical axis of the packaging: these
measures consist in having the means of transport only enter the fuel building once the bottom of the packaging
is below a level situated 8 metres from the ground, in order to benefit from the damping of the slab25 situated on
the floor of the fuel building for as long as possible. ASN considers that these additional measures are
nonetheless insufficient to rule out the loss of integrity of a packaging in the event of a fall and it therefore
considers that EDF must examine measures to guarantee the containment of radioactive materials following
such a fall; ASN formulated requests N° 9.a and 9.b in Appendix 2 to this letter on this subject.
ASN also notes that the configuration concerned is not favourable to the performance of a functional test on the
effectiveness of the air extraction system with iodine trap until such time as the potential sources of leaks have
been determined and measures have been taken to reduce these sources.
Finally, ASN recalls that additional data are still needed in order to take account of all the packaging fall
configurations for the reactors of the P’4 series and to demonstrate that their mechanical containment is
maintained and also recalls that this point is the subject of an ASN request, in the letter in reference [84], to
which no reply has been received from EDF.
C.3. Safety review of effluent packaging and treatment auxiliary buildings
(BAC/BTE)
Even if the nuclear fuel in the reactor or the spent fuel pool constitutes the main challenge for satisfactorily
controlling the risks of radiological accidents in NPPs, accidents involving radioactive waste and effluents stored
or undergoing reprocessing on the site are also liable to lead to releases of radioactive substances. In the letter in
reference [1], ASN thus asked EDF to review its control of the risk of radiological accidents that could occur
within its packaging and effluent treatment nuclear auxiliary buildings (BAC/BTE).
After analysis of the initial study data transmitted by EDF, ASN indicated in the letter in reference [87] that the
safety demonstration for the BAC/BTE was insufficient. ASN thus asked EDF, in the letter in reference [87], to
complete its file in order to include all the information necessary for assessing the risks and demonstrating that
they are satisfactorily controlled (description of the facilities, of the operations which can take place in them and
the radioactive substances present, identification of the risks according to the operations performed, preventive
and protective measures, consequences of incident and accident scenarios, etc.).
In the letter in reference [88], EDF transmitted a limited update of its safety assessment file for the BAC/BTE
which did not provide an exhaustive response to the requests formulated by ASN.
ASN considers that the safety demonstration with regard to the risk of radiological releases from the BAC/BTE
presented by EDF in the VD3 edition of the safety report for the 1300 MWe series reactors needs to be
completed and better structured with respect to the provisions of title III of the Order in reference [5]. This
point is the subject of request N° 14 in Appendix 3 to this letter.
This slab is made of a shock-absorbing material, installed in order to attenuate the shock generated by a possible fall by a fuel
container when it is being handled in the building, in order to maintain the integrity of the packaging and reduce the forces
transmitted to the civil engineering structures.
25
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D. Probabilistic Safety Assessments (PSA)
D.1.
Level 1 PSA
For the VD3-1300 periodic safety review, EDF carried out level 1 PSAs on the risks for the reactor fuel
associated with internal events and hazards (fire, internal flooding and earthquake) as well as a PSA for the fuel
storage pool. These PSAs are a tool for assessing the level of safety of the reactors and can be used to identify
areas for improvement of the 1300 MWe series NPPs. They in particular highlighted the need for changes to the
facilities to reduce the risks linked to core melt in the reactor building and fuel melt in the fuel building.
The level 1 PSAs developed by EDF for the 1300 MWe plant series were reviewed by IRSN and were the
subject of a consultation of the GPR in May 2012. Following this meeting, ASN made a globally positive
assessment of the changes made by EDF to the level 1 PSAs performed for the 1300 MWe reactors, more
specifically pointing out that the PSAs were satisfactory with regard to the hazards, especially those concerning
fire and internal flooding. Nonetheless, this assessment identified a certain number of additions and
improvements necessary for determining and evaluating the design or operational modifications needed to
improve the safety of these reactors, for which ASN formulated requests in the letter in reference [89].
EDF updated its PSAs to take account of ASN’s requests and its commitments taken following the GPR
meeting in the letters in reference [90] and [91] and it more specifically undertook to study:
-
a modification of the facility to deal with the consequences of a breach in the thermal barrier of
the reactor coolant pumps, in order to reduce the risk of core melt with bypass of the containment
[92];
design or organizational improvements to reduce the risk of core melt resulting from a breach on
the JPI26 system in the electrical rooms of the reactors on the Penly site [93].
EDF also declared a modification enabling the risk of inadvertent opening of the reactor coolant system valves
to be reduced, as this represents a significant part of the total risk of core melt in the event of a fire occurring in
the Controbloc premises. This modification was approved by ASN in the letter in reference [94].
The PSA for the fuel building pool was the subject of an ASN request in the letter in reference [89] for
justification of the new values introduced into the updated version. EDF replied to this request with the letter in
reference [93]. ASN considers the data transmitted by EDF to be satisfactory.
D.2.
Level 2 PSA
The purpose of the level 2 PSA is to assess the risk of radioactive releases into the environment in the event of a
core melt accident.
After consulting the GPR on 28th March 2013, ASN considered, in the letter in reference [95], that the approach
adopted by EDF for performance of its level 2 PSA for the 1300 MWe reactors was on the whole consistent
with international practices and it notes that EDF intends to make a significant number of improvements with
regard to the uncertainties and approximations identified. However, ASN has identified further additions and
improvements to the level 2 PSA that should be made for the VD3-1300 periodic safety rewiew and for which it
formulated a number of requests [95]. ASN also underlined that the assessment of the radioactive releases was
not precise enough and asked for it to be improved for the next periodic safety rewiews.
In December 2013, EDF transmitted an updated version of the level 2 PSA which complies with the
commitments it made in the letter in reference [96], following the GPR meeting, but which does not include the
answers to ASN’s requests. This updated level 2 PSA shows a significant reduction in the frequency of largescale releases, in particular in the case of total loss of electrical power supplies. With regard to the objectives of
the review, this comparison highlights the pertinence and the safety benefits associated with the modifications to
the pressuriser valves and the order for early closure of the containment isolation valves in a loss of electrical
power situation.
26
JPI: Nuclear island fire protection system
25 / 49
The answers then provided to ASN’s requests by EDF, in the letters in reference [97] and [98], are not
considered to be satisfactory, except for the answer concerning the containment failure situations (request D2 in
the ASN letter [95]). EDF wishes to postpone the deadlines for addressing some of these requests to the update
of the level 2 PSA, called “EPS 2 VD3 REX” which will be carried out for the next periodic safety review. On
this subject, ASN adressed requests N° 10, 11.a and 11.b to EDF in Appendix 2 to this letter.
E. Design of systems and civil engineering structures
E.1.
Safety classification of EIP-S 27
EDF sent ASN studies concerning the clarification of its safety classification rules28 applicable to parts of the
1300 MWe series reactor aiming, on the one hand, to ensure the consistency of the classification rules for
elements “important for safety – not safety classified” (IPS-NC) with the more recent rules applicable to the N4
plant series and, on the other, to integrate the safety demonstration changes resulting from the VD3-1300
periodic safety review into its classification approach, more specifically the tightening of the requirements
concerning severe accidents and hazards.
ASN considered that EDF needed to continue to revise its safety classification rules applicable to the 1300 MWe
plant series and, in the letter in reference [99], sent requests concerning the classification of:
-
electrical equipment used in extended operating conditions;
handling equipment which, in the event of a falling load, is liable to constitute a hazard for the fuel
assemblies;
equipment identified as constituting potential hazards as part of the “seismic interaction” approach;
equipment being necessary to the protection against hazards.
EDF’s answers, transmitted in the letter in reference [100], in principle constitute a satisfactory response to the
various points raised. However, ASN has not yet received the updated version of the safety report for the
1300 MWe plant series (VD3 edition), taking account of its requests.
E.2.
Equipment qualification for accident conditions.
E.2.1. Equipment qualification
The studies transmitted by EDF allow significant progress to be made in demonstrating the ability of the
equipment to perform its safety functions with respect to the loadings and ambient conditions associated with
the situations in which it is needed.
ASN however asked EDF for additional data concerning the qualification of certain equipment in the letter in
reference [101]. The additional justifications transmitted since then by EDF in the letters in reference [102] to
[108] demonstrate that the qualification of this equipment is considered to be satisfactory. The review of the
measures designed to reduce the vibration levels of certain pipes will be continued by means of specific
investigations.
EIP: element important for the protection of the interests mentioned in Article L. 593-1 of the Environment Code, as defined
in the order of 7th February 2012 setting the general rules for BNIs. The term EIP-S indicates those EIP which are “important for
safety”, in other words linked to control of the risk of radiological accidents.
28 A safety class is a range of generic safety requirements to be met by the equipment given the similarity of its contribution to the
safety demonstration.
27
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E.2.2. Calculation of doses integrated by the equipment in the event of an
accident with or without core melt
In the letter mentioned in reference [109], EDF transmitted the updated baseline safety requirements associated
with accident conditions qualification of the equipment of the nuclear reactors in operation. The method and
hypotheses adopted for calculating the doses integrated by the equipment of the nuclear reactors in service are
described in the notes mentioned in references [110] and [111] transmitted in the above-mentioned letter.
Following the review, ASN considered that the methodology for calculating the doses integrated by the
equipment in the event of an accident with or without core melt was on the whole satisfactory for the reactors in
operation, as was the collection of hypotheses associated with its application to the equipment of the 1300 MWe
nuclear reactors. ASN will nonetheless check that EDF has taken account of the requests it formulated in the
letter in reference [112] aiming to:
-
demonstrate that the  radiation has a negligible impact on the equipment;
extend the integration of uncertainties into the residual heat calculations;
justify the lists of isotopes considered in the materials balance calculation;
justify the exclusion of organic iodine in the event of a fuel handling accident;
examine the impact of the fractions of iodine released in gaseous form into the containment.
E.3.
Design of the reactor digital integrated protection system (SPIN)
EDF carried out studies on the design review for the reactor digital integrated protection system (SPIN)29. EDF
also notified a modification, the purpose of which is to make the SPIN more conservative in the event of a
cooling incident, which was approved by ASN in the letter in reference [94].
Following its review with the support of IRSN, ASN considers that the SPIN should be made even more
conservative with regard to certain incident situations and, in the letter in reference [113], asked EDF to provide
additional justifications, to modify the way in which certain uncertainties are taken into account and to complete
the notified modification. More particularly, with regard to the necessary improvements to the conservative
approach, ASN asked EDF to assess whether compensatory measures needed to be taken for the 1300 MWe and
1450 MWe reactors, without waiting for their next periodic safety review.
In the letters in reference [114] and [115], EDF answered some of the requests of ASN and in particular
informed ASN that it did not consider any reinforcement to be necessary. This information is currently being
reviewed by ASN and IRSN.
E.4.
factors
Modernisation of the control room – Integration of organisational and human
EDF has undertaken a programme to renovate and modernise the control room, involving changes to the
physical layout of the workspace, such as the operator workstations, the meeting and reception areas, and the
reactor control resources. This renovation and modernisation project will entail modifications to the control
room (layout of the control room, changes to the monitoring tools, providing the operators with additional
information, digital recorders and digitised alarm sheets, etc.). Given the impact of these modifications and their
cumulative nature on the reactor control and monitoring activities, EDF has identified major social,
organisational and human (SOH) issues.
Although the approach initiated by EDF is in principle satisfactory, the review revealed that the validation tests
run by EDF were insufficient to be able to validate the overall consistency of the control room modifications in
terms of the shift crew ergonomics and the tasks they are required to carry out in normal, incident and accident
situations.
One of the main functions of the SPIN is to calculate the linear power density (PLIN), by comparison with the risks of fuel
melt and pellet-cladding interaction (IPG), and the critical heat flux ratio (RFTC), by comparison with the risks of a boiling crisis.
These physical parameters can be used to assess the risk of loss of integrity of the first barrier.
29
27 / 49
These inadequacies were the subject of ASN requests in the letter in reference [116]. EDF’s answers, sent by the
letters in reference [117] and [118], are being reviewed.
E.5.
Condition of and improvements to reactor containment
At the request of ASN, the subject of the condition of the containments and more generally of the confinement
improvement, led to consultation of the GPR on 26th June 2013. ASN stated its position on these topics in the
letter in reference [119], on the basis of an IRSN technical report and the opinion of the GPR in reference [120],
to which EDF submitted its initial answers in the letters in reference [121] to [123].
E.5.1. Monitoring the condition and behaviour of the containments
ASN considers that the monitoring of the condition and behaviour of the reactor containments of the
1300 MWe reactors is on the whole satisfactory. EDF in particular responded favourably to ASN’s requests for a
check on the condition of the sleeves of the mechanical penetrations and for monitoring of the residual
deformation of the equipment access hatches (TAM) in order to examine the extent to which these deformations
and their foreseeable evolution could impair the performance of the containments. This assessment of the
monitoring of the condition and behaviour of the containments in no way anticipates the results of their
individual tests, which will be carried out during the ten-yearly outage inspections. EDF anticipates potential
difficulties on a number of containments identified as being potentially vulnerable during the previous tests and
has already launched a programme to develop new techniques which would, if necessary, be able to restore the
static leaktightness of these containments.
ASN does however note a persistent difference of opinion with EDF concerning the assessment of the
importance for safety of the inner containment leaktightness continuous monitoring system (SEXTEN). ASN
maintains its position with regard to the fact that the SEXTEN instrumentation is necessary for monitoring
compliance with the initial hypotheses in the accident studies concerning the containment safety function and
must therefore be covered by requirements defined more specifically for maintenance and inspections in the
general operating rules. On this subject, ASN formulated request N° 12 in appendix 2 to this letter.
Finally, ASN has not yet received any reply from EDF to its requests concerning:
-
the prevention or limitation of the ingress of external water (capillary rising of groundwater, infiltration of
rainwater, etc.) in containments for which there is a risk of the development of concrete internal swelling
pathologies;
the monitoring of the double walls of the safety injection (SIS) and containment spray (EAS) systems.
E.5.2. Improving the containment safety function
EDF has also initiated studies designed to improve the containment of the 1300 MWe reactors in an accident
situation.
ASN considers that all the modifications being envisaged by EDF following these studies will represent a
significant step forward in mitigating radioactive releases, more specifically in the case of an accident with core
melt, in particular:
-
30
the modifications to the containment annulus ventilation system (EDE) designed to adapt its operating
characteristics by differentiating between the management of an accident situation with or without core
melt;
the modification of the RPE30 system for reinjection into the containment of liquid leaks collected in an
accident situation outside the containment and designed to reinforce the performance of this system to
enable it to reinject into the reactor building any effluents collected, even in a severe accident situation
associated with high pressure in the containment.
RPE : Nuclear drains, vents and discharge heights system
28 / 49
ASN also considers that the work done by EDF to identify elements involved in the “third barrier extension”31
and the requirements resulting from this classification are satisfactory. Following the requests formulated by
ASN in its letter in reference [119], EDF in particular intends to modify its criteria for assigning an item to the
containment third barrier extension. ASN has not however as yet received the relevant updated EDF documents.
F. Modifications to the facilities and their operating procedures
Further to the risk control reassessment studies, in the summary note in reference [2], EDF presented an
extensive list of modifications to be performed on the 1300 MWe reactors.
At this stage of the generic review, ASN considers that in principle, all these modifications envisaged will make a
significant contribution to improving the safety of the 1300 MWe reactors.
Over and above this general assessment of the principles of these modifications, ASN, with the support of
IRSN, has initiated a review of the notified modifications transmitted by EDF in compliance with Article 26 of
Decree 2007-1557 in reference [124].
ASN addressed to EDF statement position letters (references [116], [94] and [125] to [156]) on a first batch
notified by EDF concerning physical modifications and changes to the general operating rules for the reactors of
the P4 plant series. The similar batch of physical modifications and general operating rules for the reactors of the
P’4 plant series is under review.
EDF intends to notify the remaining modifications mentioned in its summary note [2] before the end of 2015 in
a third batch common to the P4 and P’4 plant series.
The containment and the devices isolating its penetrations constitute the third containment barrier. However, certain systems
required to control an accident situation are liable to circulate outside the containment fluid contaminated outside the reactor
building. These systems then contribute to the confinement function and constitute a “third barrier extension”.
31
29 / 49
ASN requests
A. Monitoring of EDF commitments
In the letter in reference [157], ASN consulted the Advisory committee for nuclear reactors (GPR) concerning
the general results of the reassessment studies with regard to safety and the modifications being envisaged by
EDF to improve safety. ASN also requested its opinion on certain provisions of the ECOT and the ageing
management of the facilities.
Following the meetings of 15th and 16th October 2014, the GPR transmitted its opinion and its recommendations
to ASN, which took account of the proposed commitments presented by EDF.
EDF then sent ASN confirmation in the letter in reference [4] of the commitments presented to the GPR.
Request N° 1: Given the large number of measures planned, ASN asks that every six months, starting in
June 2015, you present it with a progress briefing on these commitments.
B. Reassessment of hazard-related risks
B.1.
Internal seismic-induced flooding
The risks of internal seismic-induced flooding were examined during VD3-1300 periodic safety review (see
section B.1. of appendix 1).
The height of the retention basins in premise NC0501 and their ability to withstand an SSE earthquake are
presented in the EDF studies to justify the absence of the risk of the loss of cooling of the FB pool by the PTR
system owing to seismic-induced flooding in this premise. ASN therefore considers that these retention basins
are necessary for the safety demonstration and that they must be considered to be EIP and be covered by
specified requirements.
Request N° 2: ASN asks you to classify the retention basins in premise NC0501 as EIP and to identify
the associated specified requirements, more specifically in terms of retention capacity and ability to
withstand an earthquake.
B.2.
Risks associated with high air and water temperature conditions
The availability of the equipment required to cover the operating conditions of categories 1 to 4 in the safety
report, in a period of high summer heat, is verified by comparing the maximum temperature reached in the area
and the temperature acceptable for the equipment present in this area.
For normal operating conditions (category 1), the allowable temperature considered by EDF is the design-basis
temperature (Td) of the equipment present in the area. For incident and accident conditions (categories 2 to 4),
the allowable temperature considered by EDF corresponds to a temperature Tr, which is higher than Td and
which requires justification of its robustness above its design-basis temperature in order to guarantee the
availability of this equipment. If temperature Tr does not compromise the short-term operation of the
equipment concerned, its operation at this temperature Tr does however affect its long-term functional
capabilities and its lifetime.
In the letter in reference [25], ASN thus requested that the required availability of the equipment needed to
maintain a reactor safe shutdown state over the long-term after the occurrence of a category 2 to 4 situation be
verified with regard to the Td rather than Tr temperature criterion. In the letter in reference [26], EDF specified
that the long-term extension approach to the thermal studies scenarios in the “extreme heat” baseline
30 / 49
requirements sets the duration of the accident scenarios at 10 days and that, for this time, the Tr temperature
criterion does not compromise the operation of the required equipment. Beyond this time, EDF considers that
its emergency response organisation would be able to deal with the areas in which the temperature was excessive,
on a case by case basis.
ASN considers that it is essential that EDF demonstrate the effectiveness of the resources which would thus be
deployed by the emergency response organisation.
Request N° 3: ASN asks that, no later than the submission of the first RCRS associated with the VD31300 periodic safety review, you explain the resources made available to your emergency response
organisation and justify that they are adequate for maintaining a maximum allowable temperature Td
in the premises for all the equipment required to function for an indefinite period, in a high summer
heat situation, for at least one of the accident scenarios in the thermal studies, by modifying the
“extreme heat” baseline requirements for the 1300 MWe plant series.
B.3. Classification and requirements applicable to the means necessary for ensuring
secondary water independence for riverside sites
In the letter in reference [40], ASN considered that, in accordance with its position expressed in the VD3-900
periodic safety review and its examination of the “extreme heat” baseline requirements for the CPY plant series,
the studies of the incident or accident situations caused by an external hazard or a combination of hazards, are
part of the nuclear safety demonstration. ASN therefore considered that the means necessary for controlling
“site H1” situations are EIP and that they must thus be covered by specified requirements.
ASN also considered that EDF’s proposal for adding a prescription to the “general” chapter of the technical
operating specifications (STE) concerning the volume of SER water needed to cover the needs in a “site H1”
situation and “H1 + LOOP” situation, with a time of one month to restore conformity, was not sufficient for
riverside sites.
Request N°4: Within the context of the VD3-1300 periodic safety review for the riverside sites of the P’4
plant series (Belleville, Cattenom, Golfech, Nogent), ASN asks you:
-
-
to introduce a group 2 event into the technical operating specifications (STE) of Chapter
III of the general operating rules (RGE), in the “reactor at power” state and in shutdown
states for which the reactor coolant system is closed or partially open, associated with an
insufficient volume of SER water for site H1 situations and site H1+LOOP combination
situations which could be induced by an external hazard;
for this event, to adopt a time of 14 days to restore the required water reserve.
*
In the particular case of the Cattenom site (4 reactors on the same site), EDF on the one hand mentions reserve
water levels in the demineralised water (SER) distribution tanks that are higher than on the other riverside sites
and also that there is dynamic rather than gravity-based resupply of the tanks in the SG emergency feedwater
system (ASG) from the SER tanks by means of the ASG 171 PO pump.
On the 1300 MWe reactors, the means needed to monitor the SER water reserves and for gravity resupply of the
ASG from these reserves, are classified IPS-NC, which is not the case of the specific resources at Cattenom.
These means are used in “site H1” and “H1+LOOP” situations.
ASN considers that the specific means on the Cattenom site must be the subject of classification rules that are
consistent with those of the other riverside sites.
Request N° 5: Within the context of the VD3-1300 periodic safety review, ASN therefore asks you:
-
to classify as IPS-NC the resources of the Cattenom site needed, on the one hand, to monitor
the availability of the SER water reserves required to deal with H1 or LOOP site situations and
31 / 49
-
their combination as a result of external hazards and, on the other, to ensure dynamic resupply
of the ASG from these reserves by means of the ASG 171 PO pumps;
for these resources, to define operational requirements (periodic tests, maintenance, availability
criteria in the STE, etc.) that are consistent with this classification.
B.4.
Internal fire
In the current the safety demonstration, EDF defines certain criteria for the proximity between fire sources and
the equipment to be protected or criteria for the thickness of smoke layers under the ceiling. These criteria, used
to demonstrate the absence of a common mode failure risk in the event of a fire, in particular for safety fire
zones or areas with a “localised fire potential” (PFL), were not reviewed by EDF, given the advances in
knowledge of the risk of failure of equipment subjected to smoke.
In response to ASN’s request, EDF does not envisage reviewing certain fire sectorisation justifications based on
an analysis of the harmlessness of the propagation of smoke and simply proposes continuing its research work.
ASN considers that the numerous test programmes already carried out to characterise the effects of smoke have
already led to real progress in the knowledge usable for the periodic safety review.
Request N° 6: ASN asks you, within one year, to review the state of knowledge of the effects of smoke,
the criteria used to justify the effectiveness of the safety fire zones and the absence of common mode
failures for equipment remote from the source of fire and located more than one metre from the ceiling
of areas said to be with “localised fire potential”.
*
In its letter in reference [51] concerning the orientation of the periodic safety review for the N4 series reactors on
the occasion of their second ten-yearly outage inspection (VD2-N4), ASN asked EDF to submit a method
before the end of 2016 to justify the alternative fire sectorisation to the method currently based on the use of a
fire duration evaluated according to the density of the fire load of the area concerned (curve DSN 144).
Request N° 7: ASN asks you, when transmitting the new method justifying the fire sectorisation
developed in response to its letter [51], to also present a schedule for verification of the fire sectorisation
of the 1300 MWe plant series reactors, based on this new method and the revised criteria for
consideration of smoke as mentioned in request N° 6 of this letter, without waiting for the periodic
safety review associated with their fourth ten-yearly outage inspection.
C. Studies of operating conditions and their radiological consequences
C.1.
Baseline requirements associated with the fuel criticality risk
The requests in the letter in reference [76] concerning the scenarios involving a fuel assembly falling to the
bottom of the pool followed by perforation of the liner and the qualification of the scientific calculation software
used for the criticality studies regarding scenarios in the FB, have not yet been answered by EDF.
Request N° 8: ASN asks you to send it a schedule for transmission of the replies to the letter in
reference [76].
32 / 49
C.2.
Handling of fuel packagings
For the reactors of the P4 plant series, the studies performed by EDF conclude that the integrity of packaging
confinement in the event of a 27-metre drop without the transport covers is confirmed. EDF also intends to
take operational measures to prevent the loss of packaging integrity in the event of a fall when the fuel building is
open and the means of transport is directly in the vertical axis of the packaging:
ASN however notes that:
-
EDF’s safety demonstration is supported by a calculation whereas the safety case for transport
approval requires a representative drop test;
if the packaging is higher than 8 metres, the operating provisions remain highly dependent on
numerous human factors (for example, poor attachment of the packaging) and the drop
hypotheses considered (in particular the lack of a study concerning a packaging drop when the
means of transport is in the vertical axis of the packaging).
ASN also notes the pointlessness of performing a functional test on the effectiveness of the air extraction system
with iodine trap (DVK) given the extent of the leaks from the FB when the fuel packaging handling hatch is
open, by comparison with the capacity of the DVK iodine system.
ASN considers that a loss of packaging integrity cannot be ruled out and that, in line with the principle of
defence-in-depth, EDF should study measures to guarantee the confinement of the radioactive substances
released following a fall by a packaging.
Request N°9.a: Prior to the periodic safety review associated with the fourth ten-yearly outage
inspections of the 1300 MWe reactors, ASN asks you:
-
-
to identify and characterise the potential sources of tightness leaks from the fuel buildings
of the P4 reactors when the fuel packaging handling hatch is open and to study the means
of limiting these sources in terms of both number and leak rate (whether or not the hatch
is open);
to study the possibility of improving dynamic containment of the FB building when the
fuel packaging handling hatch is open, by reinforcing the ventilation of the FB.
Request N°9.b: In addition to the condition for entry of the conveyance into the fuel building of the P4
series reactors, ASN asks you to introduce a prescription demanding that the FB door be kept closed
each time the packaging loaded with fuel is at a height above the ground of 8 metres or more in the fuel
packaging handling hatch.
D. Level 2 Probabilistic Safety Assessments
In the letters in reference [97] and [98], EDF provided answers to the requests made by ASN in its letter in
reference [95].
With regard to request D4 concerning the time available to the operators to carry out steps to resupply the ASG
tank in the event of a small primary breach with reactor coolant pump shutdown, the EDF assessment only
concerns the time taken by the operators in the control room and not the manual actions carried out locally by
the field operators. It does not therefore constitute a complete response.
Request N° 10: For the assessment of small primary breach with automatic reactor coolant pump
shutdown sequences, ASN asks you to justify that the times available to the field operators to carry out
local operating measures are compatible with the kinetics of emptying of the steam generators
emergency feedwater system (ASG) tank and to ensure that other local measures needed to manage the
situation do not compromise the resupply of the ASG tank.
*
33 / 49
ASN considers that the EDF’s answer to request D2 is satisfactory. However, ASN considers that the answers to
requests D1 and D3 are not satisfactory and therefore maintains the corresponding requests.
In the letter in reference [97], EDF indicated that it wished to postpone the deadlines for integration of these
ASN requests until the update of the level 2 PSA to be performed for the next periodic safety review.
ASN considers that EDF must initiate the updating of the level 2 PSA without waiting for the next periodic
safety review for the 1300 MWe reactors but that it could agree to postpone the initial deadlines provided that
EDF can justify this postponement and provide a programme of work comprising a priority section dealing with
the potential implications for nuclear safety. ASN considers that these requests must also be incorporated into
the VD2-N4 and VD4-900 periodic safety reviews.
Request N°11.a: ASN asks you to present it, within 6 months, with a programme of work justifying the
deadlines for updating of the level 2 PSA to take account of the above-mentioned ASN requests,
identifying a priority section concerning the nuclear safety issues.
Request N°11.b: ASN asks you, following the updating of the level 2 PSA, to present it with an
assessment of the risks linked to the combustion of hydrogen inside the inner containment and inside
the annulus and, as necessary, to define the envisaged modifications and improvements.
E. Continuous monitoring of the leak rate from the inner containment and its penetrations
The order of 7th February 2012 in reference [5] defines the elements important for protection (EIP) of the
interests mentioned in Article L. 593-1 of the Environment Code, as follows: “Structure, equipment, system
(programmed or not), hardware, component or software present in a basic nuclear installation or placed under the responsibility of the
licensee, fulfilling a function necessary for the demonstration mentioned in the second paragraph of article L. 593-7 of the environment
code, or checking that this function is ensured”.
The SEXTEN which, during normal reactor operation, allows continuous monitoring that the “confinement”
safety function is performed in accordance with the hypotheses of the initial accident studies of the safety report,
thus complies with the definition of an EIP. ASN therefore asked EDF, in its letter in reference [119], to give an
IPS-NC classification to the SEXTEN instrumentation and then to make the corresponding changes to the
general operating rules (RGE) in order to clarify the requirements for in-service monitoring of the correct
working of this instrumentation, in particular with regard to maintenance and calibration.
In its letter in reference [123], EDF responded unfavourably to ASN’s request, considering that:
-
the confinement safety function is performed adequately, on the one hand through the tightness tests
performed on the penetrations during reactor outage periods and, on the other, by the containment tests
performed during the outages associated with the ten-yearly inspections;
continuous monitoring of the leak rate from the inner containment and its penetrations performed by
the SEXTEN during reactor operation is not therefore necessary in order to guarantee the conformity of
containment performance during two outage periods;
the SEXTEN is not therefore an EIP.
ASN recalls that the confinement function performed by the inner containment and its penetrations is decisive in
limiting the radiological consequences for the public and the environment in the event of an accident. ASN
therefore considers that it is essential to perform continuous monitoring of this fundamental safety function in
order to detect and rapidly correct any deviations before it becomes required in the event of an accident.
ASN’s position regarding the safety importance of continuous monitoring of the confinement function
performed by the containment and its penetrations is incorporated in full by EDF:
-
32
in the safety report32 for the VD3 state of the 1300 MWe reactors, which specifies that the result of
operational monitoring of the leak rate from the inner containment performed continuously by the
SEXTEN is compared with two criteria concerning the acceptability of the overall tightness of the inner
Volume II, Chapter 4, section 2.3.3
34 / 49
-
containment which are associated with the required steps to be taken, which can go as far as reactor
shutdown;
in the technical operating specifications of Chapter III of the general operating rules, which require that,
in the event of a tightness defect detected by the SEXTEN, the licensee must initiate reactor shutdown
within 14 days or within 3 days, depending on the scale of the loss of tightness detected.
Request N° 12: ASN asks you to give an IPS-NC classification to the SEXTEN instrumentation
necessary for continuous monitoring of compliance with the criteria concerning the acceptability of the
overall tightness of the inner containment and its penetrations and, in the RGE, to specify the
requirements defined for in-service monitoring of the correct working of this instrumentation, more
specifically with regard to maintenance and calibration.
F. Reassessment of the control of drawbacks inherent to the installation
Article L. 593-18 of the Environment Code, requires that the periodic safety review be able to update not only
the assessment of the risks of accident, but also the drawbacks inherent to the facility as a result of its normal or
degraded operation. However, control of drawbacks has not been fully integrated by EDF into the examination
of the VD3-1300 periodic safety review.
The order of 7th February 2012 in reference [5] and ASN resolution 2013-DC-0360 of 16th July 2013 in reference
[158] also introduced obligations for the documents concerning the control of drawbacks to be enclosed with the
periodic safety review reports transmitted to ASN as of 1st July 2015:
-
analysis of performance by comparison with the best available techniques (art. 1.3.1 of the resolution);
elements allowing a review of the discharge limits set for the substances mentioned in the table
appended to Article R. 211-11-1 of the Environment Code (art. 4.1.11-I of the order);
chemical and radiological status of the environment (art. 3.3.6. of the resolution);
measurement of noise levels (art. 4.4.5.-I of the resolution);
data on the permanent monitoring of radioactivity or the doubling of the measurement channels (art. 6.5
of the resolution);
summary of assessments and schedule for repackaging of certain waste (art. 6.8 of the order).
In the letter in reference [5], ASN agreed that the approach used to deal with drawbacks could be gradually
reinforced before the next periodic safety reviews.
Request N° 13: ASN asks you to specify the envisaged measures and their deadlines, in order to
reinforce how the drawbacks of the 1300 MWe reactors are taken into account in order to comply with
all the applicable regulatory requirements.
35 / 49
ASN requests concerning
the contents of the VD3 edition
of the safety report for the 1300 MWe series reactors
A. Demonstration of satisfactory control of the risks of an accident within the waste
packaging and radioactive effluent processing buildings (BAC/BTE)
Even if the nuclear fuel in the reactor or the spent fuel pool constitutes the main challenge for satisfactorily
controlling the risks of radiological accidents in NPPs, accidents involving radioactive waste and effluents stored
or undergoing reprocessing on the site are also liable to lead to releases of radioactive substances.
ASN considers that the safety demonstration relative to the BAC/BTE presented by EDF in the VD3 edition of
the safety report for the 1300 MWe series reactors needs to be further clarified and better structured in the light
of the requirements of title III of the Order in reference [5].
Request N° 14: ASN asks you, no later than the submission of the first RCRS associated with the VD31300 periodic safety review, to complete the parts concerning the BAC and the BTE in the VD3 edition
of the safety report for the 1300 MWe plant series reactors, presenting the following with a level of detail
proportionate to the potential consequences:
-
a description of the facilities and all the operations that can take place in them;
a description of the radioactive substances present (inventory, maximum physical and
radiological characteristics, etc.);
the list of trigger events adopted for the prudent deterministic approach, justified
according to the operations performed in these buildings;
the steps taken to prevent and detect incident and accident situations associated with
these trigger events;
the steps taken to mitigate the consequences of these incident and accident situations;
the safety requirements associated with these provisions;
the assessment of the radiological consequences of these incident and accident situations.
B. Demonstration of the satisfactory control of the accident risks resulting from possible
malicious acts that cannot be ruled out
Decree 2007-1557 of 2nd November 2007 in reference [124] requires that the safety report cover all accidents
which could occur on the facility, whether their cause is of internal or external origin, including in the case of a
malicious act. Articles 3.5 and 3.6 of the Order of 7th February in reference [5] dealing with this subject, specify
that the consequences on the facility of malicious acts are initiating events to be covered by the safety
demonstration as an internal and external hazard.
The VD3 edition of the safety report for the 1300 MWe reactors transmitted by EDF does not contain the
expected elements of the part of the safety demonstration concerning malicious acts.
Request N° 15: ASN asks you, no later than the submission of the first RCRS associated with the VD31300 periodic safety review, to send it the required additions to the VD3 edition of the safety report
presenting the studies and provisions concerning control of the consequences of accidents with would
arise from malicious acts that cannot be ruled out.
36 / 49
C. Presentation of the EIP and their defined requirements
The role and the requirements defined for certain systems or their main components identified as EIP are not
mentioned in the VD3 edition of the safety report for the 1300 MWe reactors.
For example, ASN observed the following inadequacies concerning the ventilation systems:
-
-
-
the safety requirement notes for the DVS system, in particular detailing the seismic requirements for its
various components, are not referenced in Chapter II-7-4.1 (ventilation systems with a safety role) nor in
Chapter II-1-2 (classification of equipment and structures important for nuclear safety). In this respect,
ASN notes that the tables such as “T-II-1.3.4.4.1. Operating basis earthquake (OBE) resistance of firefighting equipment of the systems” could be extended to all the seismic classified EIP;
the requirements concerning the tightness of dampers, valves, chambers, pre-filters, absolute filters and
iodine traps of the DVK, DVS and DVN systems (taking part in the external static confinement of the
buildings) included in the plant system files (DSE) and in the VD2-1300 safety report, are not mentioned
in the VD3-1300 safety report;
the notes identifying the functions of the DVK system to be qualified for special ambient conditions are
not referenced in the VD3-1300 safety report;
Chapter II.7.4.1 of the VD3-1300 safety report does not mention the role of the DVN system in keeping
negative pressure in the areas with an iodine risk by comparison with the other areas;
Chapter II.7.4.1 of the VD3-1300 safety report does not mention the role of the DVK, DVS and DVN
systems in the U233 procedure.
Request N° 16: in VD3 edition of the safety report for the 1300 MWe series reactors, ASN asks you to
present the following with a level of detail proportionate to the potential consequences:
-
the roles of the systems and structures identified as being EIP, helping to perform the
functions mentioned in Article 3.4 of the Order of 7th February 2012 in reference [5] or
ensuring that these functions are performed and, when necessary, their component parts
also identified as being EIP;
- the defined requirements for these EIP concerning their functional characteristics as
required by the safety demonstration and those concerning the design, construction and
operation of the facility.
ASN asks you to present a revision calendar within six months (it may be progressive), of the contents
of the VD3 edition of the safety report for the 1300 MWe series reactors.
D. Referencing of EIP equipment system ID numbers
By comparison with the VD2 edition of the generic safety report for the 1300 MWe series reactors, the
functional diagrams of the systems presented in the VD3 edition no longer comprise the equipment system ID
numbers of their component parts (valves, pumps, etc.).
ASN considers that this loss of information is such as to compromise its ability to fully assimilate the design,
functioning and operating procedures of the facilities and thus to perform its oversight duty.
Request N° 17: ASN asks you, in edition VD3 of the safety report for the 1300 MWe series reactors, or in
the documents referenced in it, to restore the information which in the previous editions was able to
provide a link between the elements of the systems mentioned in the safety demonstration, the
associated descriptive functional diagrams and the contents of the general operating rules.
The purpose of the U2 procedure is to monitor and if necessary restore the containment (3 rd barrier) after an accident which
damaged the fuel (1st barrier) and/or the reactor coolant system (2nd barrier) in order to minimise radioactive releases into the
environment.
33
37 / 49
E. Additional accident studies
E.1.
Study concerning the incorrect positioning of a fuel assembly
Article 3.8 of the Order of 7th February 2012 in reference [5] states that:
-
the nuclear safety demonstration is based on appropriate, explicit and validated methods and on
qualified software,
the licensee shall specify and justify the criteria assessing the results of the studies carried out to
demonstrate nuclear safety.
In the case of the assessment of the consequences of the incorrect positioning of a fuel assembly, the safety
report only presents the conclusions of the study (chapter III-4.3.3.4) and makes no reference to the software
and methods used, nor the safety criteria adopted.
Request N° 18: in version VD3 of the safety report for the 1300 MWe series reactors, ASN asks you to
explain the following:
-
the software and methods34 used to carry out the study of the consequences of the
incorrect positioning of a fuel assembly;
the safety criteria used to assess the results of this study.
E.2. Study concerning a steam line rupture
The steam line rupture accident with reactor coolant pumps shutdown is presented concisely in the VD3 edition
of the safety report (modified pages accompanying modification PNPP 2447). The study of its medium-term
phase is covered by means of a new approach described in the note in reference [159], which is not referenced in
the safety report. A review of this note showed that the new approach was unable to lift certain reserves already
issued concerning the method (MTC3D) first used for the study of this accident situation. These difficulties led
EDF to take action II-13 [4] consisting in performing additional calculations of the departure from nucleate
boiling crisis and detailed simulations of the flows in the vessel.
ASN considers that this action must be carried out in order to complete the safety demonstration concerning the
steam line rupture accident with shutdown of the reactor coolant pumps. It underlines the fact that once it is
completed, the safety report will need to be updated.
Request N° 19: ASN asks you to transmit an update of the chapter of the safety report concerning the
major steam line rupture accident with shutdown of the reactor coolant pumps, jointly with the results
of action II-13 [4]. This update shall present the study of this accident explicitly and in sufficient detail.
E.3. Study concerning control rod cluster ejection
The control rod cluster ejection study comprises two parts:
-
the “hot spot” part which concerns the assemblies with a mean burnup fraction of less than
33 GWd/tU and which aims to check compliance with the safety criteria associated with the 4th
category design-basis operating conditions and a specific criterion for non-dispersal of the fuel;
the “high burnup fractions” part concerning assemblies with a mean burnup fraction higher than
47 GWd/tU and which aims to check compliance with the specific criteria of the decoupling range
guaranteeing that there will be no cladding failure.
Following the examination, EDF updated the study of the “hot spot” part with new hypotheses and by using the
3D-kinetic method accepted by ASN in the conditions set in the letter in reference [160].
Furthermore, with regard to the “high burnup fraction" part, ASN recalls its letter in reference [161] which notes
that:
34
The detailed elements of the method can be explained in the notes referenced in the safety report
38 / 49
-
the risk of spalling of the rods clad with Zircaloy-4 (Zy-4) cannot be ruled out for fuel cladding
with a corrosion thickness reaching 80 µm;
given current knowledge, it is impossible to define criteria such as to guarantee the absence of
failure of the spalled fuel rod cladding, in the event of a control rod cluster ejection accident.
In these conditions and considering that the absence of spalling of the Zy-4 cladding is one of the guarantees for
the correct performance of the fuel in the event of a control rod cluster ejection accident, ASN requested a
further the safety demonstration taking account of the risks of spalling of the Zy-4 fuel rod cladding and an
update of the safety reports for the reactors concerned.
Finally, with regard to the fuel assemblies with a burnup fraction between 33 and 47 GWd/t, ASN asked EDF in
the letter in reference [162] to provide data to justify their ability to withstand a control rod cluster accident
situation. The data transmitted by EDF in response to these requests, which have recently changed, are currently
being reviewed.
Request N°20: ASN asks you to update Chapter III-4.3.4.4 of the VD3-1300 safety report, no later than
at submission of the first RCRS associated with the VD3-1300 periodic safety review, to:
-
integrate the modified hot spot control rod cluster ejection study performed using the new,
accepted 3D method;
indicate that the criteria of the decoupling range applied to the “high burnup fraction”
assemblies are not applicable to fuel assemblies with spalled Zy-4 cladding;
integrate the justification of the resistance of the fuel assemblies with a burnup fraction
between 33 and 47 GWd/t
integrate the new safety demonstration taking account of the risks of spalling of the fuel
rod cladding made of Zy-4 as requested in the letter in reference [161].
39 / 49
References
[1]
Courrier ASN CODEP-DCN-2011-006777 du 3 mai 2011 : Orientations des études génériques à
mener pour le réexamen de sûreté des réacteurs de 1300 MWe associé à leur troisième visite décennale.
[2]
Note EDF EMESN110475 ind. B du 17 octobre 2013 : Liste des modifications et suffisance VD3
1300 » envoyée par courrier D305513042880 du 18 octobre 2013 : VD3 1300 – Réexamen de sûreté
des réacteurs de 1300 MWe – « Note de suffisance et liste des modifications »
[3]
Courrier CODEP-MEA-2014-047641 du 21 octobre 2014 : Avis et recommandations du Groupe
Permanent « Réacteurs » des 15 et 16 octobre 2014 – Réacteurs électronucléaires – EDF – GP Bilan
VD3 1300
[4]
Courrier EDF D305514080154 du 17 novembre 2014 : GP « Bilan VD3 1300 » Positions/actions
EDF
[5]
Arrêté du 7 février 2012 modifié fixant les règles générales relatives aux installations nucléaires de base
[6]
Courrier ASN CODEP-DCN-2014-011086 du 10 mars 2014 : Prise en compte de la maîtrise des
nuisances et de l'impact sur la santé et l'environnement des centrales nucléaires dans les réexamens de
sûreté et les règles générales d'exploitation – Relevé de conclusions du séminaire du 24 janvier 2014
[7]
Courrier ASN CODEP-DCN-2011-050393 du 12 décembre 2011 : Réacteurs électronucléaires – EDF
– Examen de conformité des réacteurs de 1300 MWe – ECOT VD3 1300
[8]
Courrier EDF D455014067009 du 15 janvier 2015 : Examen de Conformité des Tranches VD3 du
palier 1300 MWe
[9]
Courrier ASN CODEP-DCN-2015-003257 du 29 janvier 2015 : Réacteurs électronucléaires – EDF –
Palier 1300 MWe – Réexamen de sûreté associé à la troisième visite décennale des réacteurs (1300
MWe) – Programme détaillé de l’examen de conformité (ECOT)
[10] Guide de l’ASN N° 21 du 6 janvier 2015 relatif au traitement des écarts de conformité à une exigence
définie pour un élément important pour la protection (EIP)
[11] Courrier ASN CODEP-DCN-2014-035410 du 8 août 2014 : Réacteurs électronucléaires – EDF –
Palier 1300 MWe – Réexamen de sûreté associé à la troisième visite décennale des réacteurs (VD3
1300) – Vérification de la conception des ouvrages de génie civil
[12] Courrier EDF D305513003149 du 11 juillet 2013 : Thème Réexamen de sûreté VD3-1300 – Fiche
DIV 04 – Suffisance des essais décennaux
[13] Courrier EDF D305513039304 du 10 octobre 2013 : VD3 1300 – PIC – Programme d’Investigations
Complémentaires
[14] Courrier ASN CODEP-DCN-2015-004631 du 9 mars 2015 : Réacteurs électronucléaires – EDF –
Palier 1300 MWe Réexamen de sûreté associé à la troisième visite décennale des réacteurs (1300 MWe)
Maîtrise du vieillissement - Fiches d'analyse du vieillissement et dossiers d'aptitude à la poursuite de
l'exploitation génériques des réacteurs du palier 1300 MWe
[15] Règle fondamentale de sûreté (RFS) 2001-01 du 31 mai 2001 relative à la détermination du risque
sismique pour la sûreté des installations nucléaires de base de surface
[16] Courrier ASN CODEP-DCN-2011-023760 du 20 mai 2011 : Réacteurs électronucléaires – EDF –
Palier 1300 MWe – Réexamen de sûreté correspondant aux troisièmes visites décennales –
Mouvements sismiques à prendre en compte pour la sûreté des installations nucléaires en application
de la RFS 2001-01
[17] Courrier ASN CODEP-DCN-2014-051797 du 18 décembre 2014 : Réacteurs électronucléaires – EDF
– Palier 1300 MWe – Réexamen de sûreté associé aux troisièmes visites décennales des réacteurs
(VD3-1300) – Réévaluation de l’aléa sismique
40 / 49
[18] Guide de l’ASN 2/01 du 26/05/2006 relatif à la prise en compte du risque sismique à la conception
des ouvrages de génie civil d'installations nucléaires de base à l'exception des stockages à long terme
des déchets radioactifs
[19] Courrier ASN CODEP-DCN-2015-000645 du 9 janvier 2015 : Réacteurs électronucléaires – EDF –
1300) – Réévaluation sismique des ouvrages de génie civil – Tenue des BAS/BL et vérification de
l’absence d’agression des bâtiments EIP par les salles des machines des CNPE de Flamanville et de
Penly
[20] Courrier EDF D305515000876 du 14 janvier 2015 : Réexamen de sûreté VD3 1300 – CFT7 –
Vérification de la conception des ouvrages de génie civil
Réexamen de sûreté associé aux troisièmes visites décennales des réacteurs de 1300 MWe (VD3-1300)
– Réévaluation sismique des matériels – Démarche DÉRÉSMA
[22] Courrier ASN CODEP-DCN-2013-049726 du 29 août 2013 : Réacteurs électronucléaires – EDF –
1300) – Inondation interne et rupture de tuyauterie haute énergie (RTHE)
[23] Courrier EDF D305513056431 du 2 janvier 2014 : Réexamen de sûreté associé à la VD3 1300 –
Réponses à l’instruction sur l’inondation interne et la RTHE – Demandes A et B.2.1
[24] Courrier EDF D305514018981 du 5 mai 2014 : Réexamen de sûreté associé à la VD3 1300 – Réponses
à l’instruction sur l’inondation interne et la RTHE – Demandes B.2.1 et B.2.2.
Palier 900 MWe – CPY – État documentaire « PTD N°2 » – Référentiel « Grands Chauds »
[26] Courrier EDF D305513003413 du 27 juin 2013 : Référentiel de Protection des tranches CPY face aux
températures extrêmes
[27] Note EDF ENSNEA050053 ind. D du 20 février 2014 : Référentiel Grands Chauds du Parc en
exploitation – Palier 1300 MWe
[28] Courrier ASN CODEP-DCN-2013-042198 du 6 novembre 2013 : Réacteurs électronucléaires – EDF
– Palier 1300 MWe – Réexamen de sûreté associé à la troisième visite décennale des réacteurs (VD3
1300) – Prise en compte des agressions météorologiques – phénomène de frasil
[29] Courrier EDF D305514084887 du 1er décembre 2014 : VD3-1300-AGR15 – Frasil – Réponse EDF à
l’avis ASN
[30] Courrier EDF D305514016453 du 20 mars 2014 : VD3-1300 – AGR15 – Frasil – Réponses EDF à
l’avis ASN
[31] Courrier EDF EMEFC121670 du 2 avril 2013 : Stratégie de prise en compte des agressions dans les
RGE à la VD3 1300
1300) – Protection contre les vents violents
1300) – Prise en compte des agressions météorologiques – Tornades
[34] Courrier ASN CODEP-DCN-2011-006190 du 11 avril 2011 : Réacteurs électronucléaires – EDF –
Tous paliers – Réexamen de sûreté – Agression externe : définition des plus basses eaux de sécurité
Réexamen de sûreté associé à la troisième visite décennale des réacteurs (VD3 1300) – Détermination
des plus basses eaux de sécurité (PBES)
41 / 49
[36] Courrier ASN DEP-DCN-0236-2007 du 10 août 2007 : Réacteurs à eau sous pression – Protection
contre les inondations externes
[37] Courrier ASN CODEP-DCN-2013-069557 du 12 février 2014 : Réacteurs électronucléaires – EDF –
1300) – Protection contre les inondations d’origine externe
[38] Courrier EDF D305514026535 du 30 avril 2014 : VD3-1300 – AGR09 – Protection contre les
inondations d’origine externe – Réponses EDF à l’avis ASN
– Réexamen de sûreté associé à la troisième visite décennale des réacteurs (VD3 1300) – Agression de
la source froide par une nappe dérivante d’hydrocarbures
[40] Courrier ASN CODEP-DCN-2013-042192 du 12 septembre 2013 : Réacteurs électronucléaires – EDF
1300) – Autonomie de site vis-à-vis d’agressions externes de mode commun
[41] Courrier EDF D305513056075 du 30 décembre 2013 : VD3-1300 – AGR03 – Autonomie de site –
Réponse EDF à la demande ASN N°1 de la lettre CODEP-DCN-2013-042192
[42] Courrier EDF D305514026110 du 30 avril 2014 : VD3-1300 – AGR/03 Autonomie de site –
Réponses EDF à l’avis ASN
[43] Courrier EDF D305514036098 du 6 août 2014 : Palier 1300 – Réponse à la demande ASN N°5 –
Autonomie de site vis-à-vis d’agressions externes de mode commun – Chapitre VI des RGE
1300) – Incendie
[45] Courrier EDF D305514026518 du 14 mai 2014 : VD3-1300 – AGR04 – Incendie – Réponses EDF à
l’avis ASN
[46] Courrier EDF D305514044480 du 29 juillet 2014 : Réexamen de sûreté associé à la troisième visite
décennale des réacteurs (VD3 1300) – Incendie
[47] Courrier EDF D305514055585 du 26 août 2014 : VD3-1300 – AGR04 – Incendie – réponses EDF à
l’avis ASN – Report d’échéances questions 6 et 7
[48] Courrier EDF D305514065081 du 30 septembre 2014 : VD3-1300 – AGR04 – Incendie – Réponses
EDF à l’avis ASN
[49] Courrier EDF D305514092248 du 23 décembre 2014 : VD3-1300 – AGR04 – Incendie – Réponses
EDF à l’avis ASN
[50] Courrier EDF D305514069486 du 1er octobre 2014 : VD4 900 et VD2 N4 – Étude de faisabilité de
l’application d’EPRESSI sur le parc
Réexamen de sûreté associé à la deuxième visite décennale des réacteurs de 1450 MWe (VD2 N4) –
Orientations du programme du réexamen
1300) – Explosion
[53] Courrier EDF D305514035947 du 10 juin 2014 : VD3-1300 – AGR 06/14 – Explosion – Réponses
EDF à l’avis ASN
[54] Règle fondamentale de sûreté (RFS) I.2.d du 7 mai 1982 relative à la prise en compte des risques liés à
l’environnement industriel et aux voies de communication
1300) – Maîtrise des risques liés à l’environnement industriel et aux voies de communication
42 / 49
[56] Règle fondamentale de sûreté (RFS) I.2.a du 5 août 1980 relative à la prise en compte des risques liés
aux chutes d’avions
Réexamen de sûreté associé à la troisième visite décennale des réacteurs (VD3 1300) – Agressions
externes associées aux risques aériens
[58] Courrier ASN CODEP-DCN-2014-015790 du 6 juin 2014 : Réacteurs électronucléaires – EDF –
1300) – Transport interne de marchandises dangereuses
[59] Courrier EDF D305515008666 du 12 février 2015 : Réexamen de sûreté associé à la VD3 1300 –
Compléments à l’instruction sur l’inondation interne et à la RTHE – Thème AGR08 – Demande A
[60] Courrier ASN CODEP-DCN-2013-048396 du 8 octobre 2013 : Réacteurs électronucléaires – EDF –
1300) – Comportement des installations vis-à-vis des perturbations électriques internes et externes
[61] Courrier EDF D305514024319 du 18 avril 2014 : VD3 1300 – CFT8 – Comportement des
installations vis-à-vis des perturbations électriques internes et externes
1300) – Études de sûreté – Mise à jour de la démonstration de sûreté
1300) – Risques de dilution homogène
1300) – Risques de dilution par fuite sur le circuit d’étanchéité des pompes primaires
Palier 1300 MWe – Risques de dilution – Dilution hétérogène inhérente à l’accident par perte de
réfrigérant primaire (APRP)
[66] Courrier EDF D305514016451 du 21 mars 2014 : VD3-1300 – REF 04 – Dilution homogène –
Réponses EDF à l’avis ASN
[67] Courrier EDF D305514026592 du 19 mai 2014 : VD3-1300 – REF04 – Risques de dilution homogène
– Réponses EDF à l’avis ASN
[68] Courrier EDF D305514040669 du 11 juillet 2014 : VD3-1300 – REF 04 – Dilution homogène –
Réponses EDF C1-C2 à l’avis ASN
[69] Courrier EDF D305514093338 du 7 janvier 2015 : VD3-1300 – REF04 – Risques de dilution
homogène – Réponses EDF à l’avis ASN
[70] Courrier EDF ENPRTL090437 du 15 décembre 2009 : Risque de dilution hétérogène au redémarrage
du thermosiphon – Demande ASN : Réponse partielle à la demande N°5 de la lettre
DGSNR/SD2/0927/2003 et réponse partielle à la demande N°10 de la lettre
DGSNR/SD2/0132/2004
[71] Courrier EDF ENPRTHL120129 du 28 décembre 2012 : Risque de dilution hétérogène au
redémarrage du thermosiphon – Demande ASN : Réponse partielle à la demande N°5 de la lettre
DGSNR/SD2/0327/2003
– Réexamen de sûreté associé à la troisième visite décennale des réacteurs (VD3 1300) – « Défaillance
passive du système d’injection de sécurité »
43 / 49
1300) – Impact du comportement des soupapes secondaires sur la couverture des transitoires de
dimensionnement du rapport de sûreté
[74] Courrier ASN Dép-DCN-0659-2008 du 14 janvier 2009 : Réacteurs à eau sous pression. Examen du
retour d’expérience des années 2003 à 2005
1300) – Accidents graves
[76] Courrier ASN Dép-DCN-0293-2007 du 27 août 2007 : Réacteurs nucléaires à eau sous pression –
Référentiel criticité
[77] Courrier ASN CODEP-DCN-2014-018653 du 18 juillet 2014 : Réacteurs électronucléaires – EDF –
1300) – Référentiel criticité
[78] Courrier EDF D305514087801 du 19 décembre 2014 : Réexamen de sûreté VD3 1300 – REF 18 –
Référentiel Criticité
1300) – Risques de surpression à froid du circuit primaire principal (CPP)
[80] Courrier EDF D305514008284 du 5 mai 2014 : Réexamen de sûreté VD3 1300 – Risque de
surpression à froid du circuit primaire principal
[81] Courrier EDF D305515011081 du 17 février 2015 : VD3 1300 – Risque de surpression à froid dans les
états RRA connecté
1300) – Conséquences radiologiques des accidents (hors RTGV et accidents graves) associées au
réexamen de sûreté des réacteurs du palier 1300 MWe réalisé à l’occasion de leur troisième visite
décennale
1300) – Études des risques pour la piscine de désactivation du bâtiment combustible (BK)
1300) – Manutention des emballages d’assemblages de combustible
[85] Courrier EDF D305514055240 du 25 août 2014 : VD3 1300 – Confinement du bâtiment combustible
en cas de trémie ouverte – demande B2
[86] Courrier EDF D305514093341 du 7 janvier 2015 : VD3 1300 – Manutention des emballages
d’assemblages de combustible
– Palier 1300 MWe – Réexamen de sûreté associé aux troisièmes visites décennales – Bâtiment des
auxiliaires nucléaires de conditionnement et bâtiments de traitement des effluents
[88] Courrier EDF D305513049312 du 19 décembre 2013 : Réexamen de sûreté VD3-1300 – BAC/BTE
Palier 1300 MWe – Études probabilistes de sûreté de niveau 1 (EPS1) relative aux réacteurs de 1300
MWe dans le cadre de leur troisième réexamen de sûreté (VD3 1300)
[90] Courrier EDF EMESN120861 du 25 juillet 2012 : GPR EPS1 VD3 1300 – Positions et actions d’EDF
44 / 49
[91] Courrier EDF EMESN120052 du 20 janvier 2012 : GPR EPS 1 VD3 1300 – Evolutions prévues par
EDF suite à l’instruction du thème « EPS de niveau 1 événements internes »
[92] Fiche question-réponse EDF D305514000981 envoyée par courrier D305514022705 du 24 avril 2014 :
VD3 1300 – GP Confinement
[93] Fiche question-réponse D305914001475 EDF envoyé par courrier D305513057286 du 31 janvier
2014 : Suite donnée aux positions actions prises dans le cadre du GP EPS N1 thématique Piscine BK
et réponse à la demande D.3 du courrier CODEP-DCN-2013-005093 – EPS « Piscine BK »
Palier 1300 MWe – P4 – État technique « VD2 » – Accord exprès à la mise en œuvre d’une
modification – Modification « Dossier d’amendement (DA) VD3 1300 MWe-P4 Lot A »
[95] Courrier ASN CODEP-DCN-2013-038780 du 25 novembre 2013 : Études probabilistes de sûreté de
niveau 2 dans le cadre du troisième réexamen de sûreté des réacteurs de 1300 MWe
[96] Courrier EDF D305513016441 du 28 juin 2013 : GPR « VD3 1300 AG/EPS2 » - Positions et Actions
d’EDF sur l’EPS N2 1300 VD2 REX
[97] Courrier EDF D305514032568 du 29 juillet 2014 : GP « AG EPS2 » VD3 1300, Réponse à la lettre de
suite ASN « CODEP-DCN-2013-038780 »
[98] Courrier EDF D305514085805 du 2 décembre 2014 : GP « AG EPS2 » VD3 1300, Réponse à la lettre
de suite ASN « CODEP-DCN-2013-038780 »
1300) – Clarification des règles de classement de sûreté IPS-NC
[100] Courrier EDF D305514017930 du 24 avril 2014 : VD3 1300 – Ref 23 Classement IPS-NC – Réponses
EDF à l’avis ASN
1300) – Pérennité de la qualification des matériels
[102] Courrier EDF D305513053131 du 18 décembre 2013 : VD3 1300- REF26- Pérennité de la
qualification des matériels
[103] Courrier EDF D305514001341 du 10 février 2014 : VD3 1300 – Classement de sûreté électrique du
GCTa du palier 1300 MWe
[104] Courrier EDF D305514048577 du 18 juillet 2014 : VD3 1300-REF26-Pérennité de la qualification des
matériels
[105] Courrier EDF D305514021250 du 15 avril 2014 : CODEP-DCN-2013-055855 – Demande N°4 –
Exigence de réalisation d’essais périodiques sur le fonctionnement du registre EVR 051 VA dans le
cadre du réexamen VD3-1300
[106] Courrier EDF D305514000845 du 23 avril 2014 : Réexamen de sûreté associé à la troisième visite
décennale des réacteurs (VD3 1300) – Pérennité de la qualification des matériels
[107] Courrier EDF D305514026305 du 30 avril 2014 : VD3 1300- REF26- Pérennité de la qualification des
matériels
[108] Courrier EDF D305514079942 du 5 novembre 2014 : VD3 1300- REF26- Pérennité de la qualification
des matériels
[109] Courrier EDF EMESN110023 du 24 février 2011 : Transmission de la mise à jour du référentiel parc
de qualification des matériels aux conditions accidentelles
[110] Note EDF ENTERP070216 ind. B du 17 novembre 2008 : REP tous palier (y compris l’EPR) –
Méthodologie de calcul des doses intégrées par les équipements lors d’un accident
45 / 49
[111] Note EDF ENTERP100087 ind. A du 30 novembre 2010 : REP tous paliers (hors EPR) - Recueil
d'hypothèses pour le calcul des doses accidentelles intégrées par tes équipements du parc en
exploitation
Réexamen de sûreté associé aux troisièmes visites décennales des réacteurs de 1300 MWe (VD3 1300)
et réacteur EPR de Flamanville 3 – Méthode de calcul des doses intégrées par les équipements lors
d’un accident avec ou sans fusion du cœur
[113] Courrier ASN CODEP-DCN-2014-020988 du 1er juillet 2014 : Réacteurs électronucléaires – EDF –
1300) – Revue de conception du système de protection intégré numérique
[114] Courrier EDF D305514093161 du 7 janvier 2015 : VD3-1300- REV1 Revue su SPIN
[115] Courrier EDF D305914020421 du 30 octobre 2014 : Réponse à la demande D6 du courrier ASN
CODEP-DCN-2014-020988 relative à l’incertitude du fléchissement des crayons de combustible
– Palier 1300 MWe – P4 – État technique « VD2 » – Accord sous réserves à la mise en œuvre d’une
modification – Modification « Dossier d’amendement (DA) VD3 1300 MWe-P4 Lot A »
[117] Courrier EDF D305514085954 du 9 janvier 2015 : Palier 1300 MWe – P4 – Etat technique « VD2 » Déclaration de modification RGE – « Dossier d’amendement VD3 1300 MWe – P4 Lot A » - Réponse
EDF aux réserves suites à l’accord ASN
[118] Courrier EDF D305514088332 : Méthodologie pour évaluation SOH de la salle de commande VD3
1300
Confinement du bâtiment réacteur et des bâtiments périphériques
[120] Courrier CODEP-MEA-2013-038427 du 8 juillet 2013 : Avis et Recommandations du Groupe
Permanent « Réacteurs » du 26/06/2013 Confinement des réacteurs à enceinte à double paroi du parc
en exploitation, associé aux troisièmes visites décennales des tranches de 1300 MWe
[121] Courrier EDF D305514091791 du 10 décembre 2014 : VD3-1300-Confinement – Réponses EDF à
l’avis ASN
[122] Courrier EDF D305514061604 du 16 septembre 2014 : VD3 1300 – CONFINEMENT – Réponse
EDF à l’avis ASN
[123] Courrier EDF D305514068866 du 1er octobre 2014 : VD3 1300 – CONFINEMENT – Réponse EDF
à l’avis ASN
[124] Décret N°2007-1557 du 2 novembre 2007 relatif aux installations nucléaires de base et au contrôle, en
matière de sûreté nucléaire, du transport de substances radioactives
Palier P4 – Accord exprès à la mise en œuvre d'une modification – Modification PNPP 2086 «
Évolution du suivi automatique de l'encrassement des échangeurs SEC/RRI »
Palier P4 – Accord exprès à la mise en œuvre d'une modification – Modification matérielle – P4 –
[VD3-1300] « PNPP2250 tome B : Rénovation des enregistreurs à papier non-IPS des salles de
commande »
Palier P4 – Accord exprès à la mise en œuvre d'une modification – Modification « PNPP 2250 tome
C : Rénovation des enregistreurs à papier IPS des salles de commande »
Palier 1300 MWe – État technique « VD2 ou VD3 » – Accord exprès à la mise en œuvre d'une
modification – Modification « PNPP 2/3340 : Isolation et ventilation naturelle de la pince-vapeur »
46 / 49
Palier 1300 MWe – Accord exprès à la mise en œuvre de la modification PNPP 2/3402 « motorisation
de la vanne du tube transfert PTR 064 VB »
Palier P4 – Accord exprès à la mise en œuvre de la modification « PNPP 2465 : Mise en place d'un
second joint statique sur les batardeaux des piscines BR »
Palier P4 – Accord exprès à la mise en œuvre d'une modification – Modification « PNPP 2509 tomes
A et C Mise en conformité des galeries vis-à-vis du risque explosion hydrogène pour les tranches du
palier P4 »
– Palier 1300 MWe – État technique « VD2 » – Accord exprès à la mise en œuvre d'une modification –
Modification « PNPP 2/3510 Protection contre les projectiles générés par les vents extrêmes »
– Modification « PNPP i546 : Amélioration de la robustesse du système EAU vis-à-vis du Dispositif
d'Auscultation Optimal (DAO) – Pose d'extensomètre de parement en extrados de la paroi
précontrainte des bâtiments réacteurs du parc en exploitation »
Palier 1300 MWe – État technique « VD2 ou VD3 » – Accord exprès à la mise en œuvre d'une
modification – Modification « PNPP 2/3584 Protection des matériels IPS contre les projectiles générés
par les vents extrêmes »
Palier 1300 MWe – Accord exprès à la mise en œuvre d'une modification – Modification matérielle
PNPP 2/3631 « Amélioration des sas BR »
Palier P4 – État technique « VD2 » – Accord exprès à la mise en œuvre d'une modification –
Modification « PNPP 2600 : Tenue des circuits hydrogénés au séisme et à la RTHE »
Palier P4 – Accord exprès à la mise en œuvre de la modification « PNPP 2604 : Moyens de surveillance
dans le domaine d'exploitation « Réacteur Complètement Déchargé » pendant les travaux en salle de
commande »
Palier P4 – Accord exprès à la mise en œuvre de la modification « PNPP 2607 tome A : Modernisation
de la salle de commande pendant les travaux dans le domaine d'exploitation « Réacteur Complètement
Déchargé » de la VD3 »
– Palier P4 – Accord exprès à la mise en œuvre de la modification PNPP2607 tome B « Modernisation
de la salle de commande – Travaux TEM »
Palier 1300 MWe – P4 – État technique « VD2 ou VD3 » – Accord exprès à la mise en œuvre d'une
modification – Modification « PNPP 2696 tome A : Remplacement des capteurs REN/EDE de
l'Extension de la troisième barrière (E3B) »
Palier 1300 MWe – État technique « VD2 » – Accord exprès à la mise en œuvre d'une modification –
Modification « PNPP 2753 Tome A : Renforcement du supportage du circuit d'eau brute secourue
(SEC) du CNPE de Paluel »
– Palier P4 – Accord exprès à la mise en œuvre d'une modification – Modification « PNPP 2732 Mise
en place de matériel ATEX dans les locaux explosibles (palier P4) »
47 / 49
Palier P4 – Accord exprès à la mise en œuvre de la modification matérielle PNPP 2763 tome A « Axes
de câblage »
CNPE de Paluel et Saint-Alban – État technique « VD3 » – Accord exprès à la mise en œuvre d'une
modification – Modification « PNPP 2763 tome B Renforcement des axes de câblage des Bâtiments
des auxiliaires de sauvegarde et Bâtiments électriques de Paluel et Saint-Alban »
[145] Courrier ASN CODEP-CAE-2014-045157 du 10 novembre 2014 : Mise en œuvre d'une modification
– Modification PNPP 2806 : Aménagement d'une voie de circulation du site de Paluel
Palier P4 – Accord exprès à la mise en œuvre de la modification PNPP 2759 « Réalimentation des
armoires des capteurs de pression enceinte en situation H3 »
– Palier P4 – Accord exprès à la mise en œuvre de la modification PNPP 2881 « remplacement des
transformateurs LLI »
Palier P4 – Accord exprès à la mise en œuvre de la modification « PNPP 2606 : Amélioration de la
réfrigération des Diesels »
Palier P4 – Accord exprès à la mise en œuvre d'une modification matérielle PNPP 2130 «
remplacement des groupes frigorifiques DEG »
Palier 1300 MWe – État technique « VD2 » – Accord sous réserve à la mise en œuvre d'une
modification – Modification « PNPP 2/3433 : Rétablissement des protections non prioritaires en
marche longue durée des diesels des réacteurs de 1300 MWe »
– Palier 1300 – Accord sous réserve à la mise en œuvre de la modification PNPP 2589 « Dispositif
passif d'alcalinisation des puisards enceinte (puisards basiques) »
Palier 1300 MWe – P4 – Accord exprès à la mise en œuvre de la modification PNPP 2639
«Valorisation en accident grave du dispositif H4 »
Palier 1300 MWe – Accord exprès à la mise en œuvre d'une modification concernant la détection de
Corium et le fonctionnement de deux recombineurs H2 par température élevée
Accord exprès à la mise en œuvre d'une modification – PNPP 2252 tome B « remplacement des
groupes de production d'eau glacée DEL du site de Paluel »
Accord exprès à la mise en œuvre d'une modification – Modification « PNPP i675 Protection vis-à-vis
de l'inondation externe par déversement direct sur la plateforme de l'ensemble des réacteurs du parc
électronucléaire »
[156] Courrier ASN CODEP-DCN-2015-007336 du 06/03/2015 : Réacteurs électronucléaires – EDF –
Palier 1300 MWe Accord exprès à la mise en œuvre de la modification PNPP 2/3754 « Réalimentation
électrique de la ventilation DVC de la salle de commande et de l’extraction inter-enceinte EDE en cas
de perte H3 »
GP Bilan VD3 1300
[158] Décision N° 2013-DC-0360 de l'Autorité de sûreté nucléaire du 16 juillet 2013 relative à la maîtrise des
nuisances et de l'impact sur la santé et l'environnement des installations nucléaires de base
48 / 49
[159] Note AREVA NP PEPD-F DC 10189 ind. D du 2 octobre 2013 : GEMMES VD3 1300 MWe –
Dossier support à l’étude de RTGV4 avec AA-GMPP en phase moyen terme
[160] Courrier ASN CODEP-DCN-2010-049305 du 24 janvier 2011 : Réacteur électronucléaire – Projet
EPR – Flamanville 3 – Instruction de la méthode rénovée pour l’étude de l’accident d’éjection de
grappe
Corrosion du Zircaloy-4 – Accident d’insertion de réactivité
[162] Courrier ASN CODEP-DCN-2011-070565 du 26 décembre 2011 : Réacteurs électronucléaires –
Accident d'insertion de réactivité – Domaine de découplage
49 / 49

letter to EDF Nuclear Operation Division

Transcription

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