IAEA Workshop on CONSTRUCTION TECHNOLOGIES … · Safety analysis methodology for ... Technical Code for Civil works in which are defined ... Provisional list of inspections of elements

Enhancing nuclear safety

Safety analysis methodology for the EPR

Institute for Radiation protection and Nuclear Safety

December 13th, 2011

IAEA Workshop on CONSTRUCTION TECHNOLOGIES FOR NUCLEAR POWER

PLANTS: A COMPREHENSIVE APPROACH

K. HERVIOU, C. PIEDAGNEL, F. TARALLO

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1. Context

2. General approach for assessing the Flamanville 3 NPP design

3. Examples of application

Content

EPR safety assessment methodology – IRSN – AIEA workshop

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1. Context



Content


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1. Context: generalities

Flamanville 3 is the first EPR built in France

There was no construction of NPP in France for more than 10 years –partial loss of the experience gained in the 70’s and 80’s

New construction technologies for nuclear power plant are now available and the EPR should take benefit of this possibility to enhance the robustness of the plant – but these technologies should be qualified

New design, innovative solutions are proposed by vendors and operators codes and standards of construction should be adapted

Existing reactors experience feedback shows the importance of the compliance of the plant with safety requirements from the commissioning phase

It is essential to carefully follow up the construction of the EPR, the manufacturing of its components to assess its safety


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1. Context: EPR general safety approach

The EPR general safety approach was defined in 1993 by the French and German standing groups of experts (GPR in France and RSK in Germany) on the basis of proposals provided by the French (IRSN) and German (GRS) TSOs.

Need for a significant improvement of the safety level of future plants at the design stage, compared to the safety level of existing plants

Choice of an evolutionary approach, taking into account:

the large operating experience on PWR plants, notably on the French and German ones,

the results of in-depth studies performed on these plants, in particular the probabilistic safety assessments,

the results of research and development activities, notably on severe accidents

Innovative features to be considered

The assessment of the safety approach by the standing groups of experts led to the definition of the Technical Guidelines for GEN III reactors to be build in France and Germany in 2000 (available in English on www.asn.fr)


http://www.asn.fr/

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Technical assessment of safety options and of the preliminary safety analysis report

Authorization creation Decree delivered by the government in April 2007

Technical assessment for the commissioning license (“anticipated assessment”)

Assessment of the detailed design

Operating rules

Control of the realization

Participation to the inspection program set up by ASN

Assessment of constructive non-compliance and of corrective actions

1. Context : IRSN involvement in EPR assessment


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An assessment program has been defined in 2007 with two main objectives:

Give more time for technical assessment by IRSN experts (ASN has one year after the operating license application to give its advice)

Limit the industrial risk for EDF

Priorities based on:

Feedback of operating plants

Innovative features

Checking that safety provisions are compliant with safety objectives and technical guidelines

The objective of IRSN is to gain a good level of confidence in EPR safety provisions before the commissioning phase.

1. Context : on-going assessment


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1. Context



Content


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2. General approach for assessing the FA 3 NPP design

Before the commissioning of the reactor, IRSN will have to assess the safety level of the reactor as built and give an advice to the nuclear safety authority. He must ensure the achievement of a level of safety performance and compliance with the requirements set design.

For this, the IRSN will continue the analysis of the detailed design, while ensuring the quality of implementation.

IRSN defined a global strategy for assessing the EPR, not only based on the safety documentation provided to support the license application.


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Flamanville 3

Detailed design

+ operating principles

EDF files assessment

Control of the construction

Inspections

Checking compliance to SAR

hypotheses

Follow-up of the

commissioning tests

Balance/effective improvement in

terms of safety and radiation

protection

Particular follow-up of

first operating years

Safety requirements (2000)

2. General approach for assessing the FA 3 NPP design

Design assumptions and features are confirmed only if they are correctly implemented and maintained in the plant.


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Detailed design

+ operating principles

EDF files assessment

Control of the construction

Inspections

Checking of the conformity of the

plant to the SAR hypotheses

Follow-up of the

commissioning tests

Balance/effective improvement in

terms of safety and radiation

protection

Particular follow-up of

first operating years

Safety requirements (2000)

30 IRSN engineers

1 resident expert

on-site during testing

24 inspections on site

per year + 8 in

engineering services

or providers –

participation of IRSN

experts is a priority

2. Organization

Engineers in charge of the technical assessment of a topic are defining inspection program and support ASN during inspection related to this topic. They assess non-compliance and corrective actions


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Accident studies

I&C

Protection system

Internal, external hazards

Detailed design of EPR systems playing a safety role including supporting systems (including risk of sump clogging, electrical supply, ventilation…)

Equipment qualification to accidental conditions

Radiological consequences

Severe accident management (including practical elimination)

Probabilistic studies (levels 1, 2, for the reactor, the fuel pool, regarding internal events, internal and external hazards…)

Radiation protection

Commissioning tests

General Operating rules…

2. Main issues to tackle before the operating license delivery

Attempt to have a global view, by safety function


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2. Background for assessing operators files

Periodic Safety Reviews

Experience feedback and incident analysis

R&D (severe accident, fuel, human factors, civil works…)

International projects and cooperation (Belene, MDEP, WENRA, IAEA, ITER…)

PSA studies

EPR simulator

Independent calculation chain for neutronics and thermal hydraulic calculations


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Example: use of the HEMERA chain for independent verification analysis

Neutronics (CHRONOS) and

thermal-hydraulics (FLICA) Affected loop

Control rod stuck

CATHARE: System

description


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Comparison MTC 3D – HEMERA (IRSN model)

0

0,5

1

1,5

2

2,5

0 50 100 150 200 250 300 350 400

Temps (s)

Puis

sance (%

Pn)

Puissance MTC3D

Puissance HEMERA

0

2

4

6

8

10

12

14

16

0 50 100 150 200 250 300

Temps (s)

RF

TC

RF T C M T C3D

RF T C HE M E RA

Nuclear power DNBR min (at assembly scale - without

penalties)

HEMERA

HEMERA

MTC 3DMTC 3D

2A – SLB with primary pumps shutdown


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2. IRSN working methodology

Files submitted by operator

Preliminary analysis

• Definition of the objectives of the assessment and main issues• Identification of necessary competence skills needed• Definition of the contribution of the different divisions• Deadlines• Identification of needed complementary documents

Discussion with the ASNon the advice to be provided

by the IRSN and deadlines

Technical exchanges withthe operator (questionnaires,

meetings)

Transmission of the detailed assessment report to the

operator to check there is no IRSN misunderstanding

Transmission of the advice to the ASN.

Depending on the subject, possible input for the inspection program


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1. Context


3. Example of application : civil engineering works

Content


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Definition of safety functions and requirements associated to SSC: principle

Confinement function

Civil engineering works(safety functions, requirements…)

Protection against external hazards

Protection against internal hazards

Systems (safety functions, requirements…)

Other functions (residual heat removal…)


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The confinement function aims to:

Avoid the release of radioactive products into the environment (external confinement)

Avoid the release of radioactive products within the installation (internal confinement), in particular to maintain acceptable radiological conditions for workers in normal and accidental situations

In case of release, filter and monitor the release

The assessment of the confinement function includes:

Civil engineering works (structures, static confinement)

Systems participating to the function (HVAC, containment isolation, internal containment release rate monitoring system, penetration leaks collecting system, systems constituting some extension of the 3rd barrier, Containment Heat Removal System…)

3. The confinement function


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Safety functions

Identification of safety requirements

applied to civil engineeringList of main components of civil

engineering

Behavioral requirements in relation

to each civil engineering component

List of civil engineering components

regarded as sensible or typical

INSPECTIONS

Civil design safety assessment

3. Civil engineering works


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Civil engineering safety functions:

• ensure containment (in particular a 3rd barrier), in all circumstances, including for serious accidents;

• withstand internal hazards (operational accidents, flooding, fire, explosion...);

• withstand external hazards (floods, earthquakes, plane crash, explosion, extreme weather conditions...).

Civil engineering safety requirements:

• leak-tightness and retention

• resistance, stability, supporting capacity for safety equipments and systems

• choice of materials and determination of their biologic thickness

• controllability and durability of the construction during the time designated for operation of the unit

3. Civil engineering works



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Behaviour requirements are defined for each structure or part of structure.

IRSN assessment consists in evaluating:

• ETC-C: Technical Code for Civil works in which are defined design criteria and construction rules

• global models, calculations and results in terms of consistency with assumptions and existence of margins, according to ETC-C

• robustness of design and demonstrations by carrying out a more detailed assessment

Results of IRSN civil design safety assessment:

• Design studies were globally satisfactory

• Demands to EDF to provide additional justifications, sometimes involving significant modifications

• Definition of items whose construction should be inspected: Importance for safety,

Execution difficulties.


3.1 Civil engineering works

Impact ?


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Detailed assessment of a selection of safety-related structures

Nuclear island basemat,

Reactor building: steel liner, inner prestressed containment, outer containment (airplane shell), internal structures, pool,

Fuel building: internal structures, pool,

Safeguard auxiliary building,

Pumping station.


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1. Objective: check, by sample survey, the practical implementation of the provisions that enable the structures to guarantee the safety functions assigned to them.

2. Take into account:

IRSN’s experience from earlier EDF plant series and other construction work, feedback from non-nuclear industry

STUK feedback on the monitoring of the construction of the Finnish EPR reactor

Lessons learned form previous inspections

The irreversible nature of the construction work or the impossibility of inspections on the construction site

3. Inspection program on civil engineering works

Help determining the quality of the construction of the facility


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Combining the main safety functions with the cross-functional aspects (support of equipment, durability and ageing) of which feedback is an essential element, aspects for inspection are identified as follows:

The main behavioral requirements are related to safety functions that the structures make possible to fulfill,

For each major building, construction elements requiring special attention as regards behavioral requirements assigned to them are identified,

Cross-checking of these indications with the phasing of the construction.

3. Inspection program on civil engineering works

Provisional list of inspections of elements of the construction of civil engineering deemed sensitive or representative of the construction process.


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Examples of technical problems highlighted during inspections:• Water excess in structural concrete,• Lack of reinforcement in the nuclear island basemat,• Cracks in the concrete of the reactor basemat,• Welding process of the containment steel liner,• Difficulties in anchor plates placing,• Unsatisfactory location of prestressing ducts.

3.1 On-site inspections on civil works: examples of findings

Significant deviations from their specified locations for several horizontal prestressing ducts, in the first concrete layer

Final control partially carried out

Location deviations higher than stated tolerances

Reduce the inner containment resistance

Reduce its capacity to ensure the safety function required

EDF actions: - demonstration of acceptability of those deviations- corrective actions for next concrete layers, to obtain

deviations lower than stated tolerances


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Basemat/IS link

Internal structures (IS)

Access for materials

External containment

Inner containment

Turning crane support bracket

Airlock

Section

over

buttress

Gusset

APC shell dome

Transfer tube

FB

SAB

-BL

2&3

Reactor

pits

Ash panIRWST

FB roof APC shell

FB basemat

Top of SG

bunker

Pool

Pre-stressing tunnel under basemat

IS basemat RB basemat

Dome

belt

Welding process of the containment steel liner

Steel liner ensures leaktightness

of the containment


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3. Welding process of the containment steel liner

Inspection in September 2008

Detailed attention to the first welding activity carried out on site on an element endorsing a safety function: the liner manufacturing

IRSN Technical assessment

• Deviations to technical requirements detected on the welding procedure perform complementary examination tests and a 100% non destructive vacuum tests over all those welds;

• Perfectible conditions of welding (climatic conditions protection…);

• Non-compliance in documentations;

• Abnormally high rates of repairs for easily weldable steel welding activity not completely controlled

100% volumetric non-destructive tests until return to a normal situation.

Manufacturer actions

• Action plan to significantly improve the quality of works:

Optimization of welding procedures

Improvement of their conditions of implementation

Complementary training sessions and selections of welders


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3. Cracks in the concrete of the reactor building basemat

December 2007: concreting for the

first time of common basemat on

nuclear island under the reactor

building

(4225 m3, thickness 1.8 m)

Several days later: open cracks of 1

mm to 3 mm

After cooling: open cracks of 0.4 mm

to 1 mm

Repair: injection of cracks

Configuration of cracks in the

circular basemat


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3. Cracks in the concrete of the reactor building basemat

Cause of the non-compliance:

• thermical effect due to the heat of hydratation of the cement

during concrete setting (expansion and contraction due to exothermic

reaction)

Aggravating circumstance:

• lack of reinforcement mesh in the upper part of the lift

Risk:

• reduced durability of the structure;

• possible corrosion of the bottom reinforcement even if cracks are

grouted;

• presence of water below the basemat should be detected during

the lifetime of the plant.


Thank you for your attention

www.irsn.fr

Documents

IAEA Workshop on CONSTRUCTION TECHNOLOGIES … · Safety analysis methodology for ... Technical Code for Civil works in which are defined ... Provisional list of inspections of elements