ADVANCED SODIUM TECHNOLOGICAL REACTOR … SODIUM TECHNOLOGICAL REACTOR FOR INDUSTRIAL DEMONSTRATION ... Irradiation services and options test. PROJECT ORGANIZATION ... SODIUM WATER

ASTRID

ADVANCED SODIUM TECHNOLOGICAL

REACTOR FOR INDUSTRIAL DEMONSTRATION

JUNE 2012

IIAEA TWG FR, Argonne USA, 20 – 22 June 2012 | Alfredo Vasile

| PAGE 1CEA | JUNE 2012

Presented by A. VASILE (CEA)

OUTLINE


ASTRID objectives and organization of the project

General features of the reactor

Safety: Examples of major improvements

Conclusions

OBJECTIVES AN ORGANIZATION

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CEA | JUNE 2012

ASTRID OBJECTIVES


Industrial prototype (could be a step before a Firs t Of A Kind)

Including French and international SFRs feedback

A GEN IV system

SafetyLevel at least equivalent to GEN III systemsProgress on specific Na reactors issuesIncluding FUKUSHIMA accident feedback

OperabilityLoad factor of 80% or more after first “learning” yearsSignificant progress concerning In Service Inspection & Repair (ISIR)

Waste transmutationDemonstration of minor actinides transmutation according to June 28, 2006 French Act on Waste Management

A mastered investment cost

Irradiation services and options test

PROJECT ORGANIZATION


CEA/Nuclear Energy Division is the leader of the ASTRI D project

Industrial partnerships to cover main design engineeri ng batchesAREVA NP: nuclear island (core and fuel stays at CEA)Support to the owner by EDFALSTOM: turbine islandBOUYGUES:civil engineeringCOMEX NUCLEAIRE: batches in robotics and mechanicsTOSHIBA: development of large electromagnetic pumpsOn going discussions for other partnerships

Manpower: 300 CEA, 200 industry

For technical developments and experimental facilities , CEA iswilling to develop international partnerships

INTERNATIONAL COLLABORATIONS

RussiaMaterial irradiations in BOR-60, Neutronics in BFSPossibility of fuel irradiations in BN-600 ?

IndiaStrong collaboration, several Implementing Agreement on going on safety, including severe accidents

USASeveral STCsSafety benchmark on CFV core

JapanSeveral STCsEAGLE collaboration

EuropeCP-ESFR project, SOFIA proposalNext Framework program in preparation (2014-2020)

Korea, China, Argentina

CEA | JUNE 12th 2012 | PAGE 6

SCHEDULE FOR ASTRID AND ASSOCIATED FACILITIES


2009 2010 2011 2012 2013 2014 2015 2016 2017

Decision to continue

Preliminary choice of options

Decision to build

Pre-conceptualdesign

Conceptualdesign

Basic design

Fuel loading

Detaileddesign

& Constructi

on

ASTRID

Facilities

Feasibility Report on minor

actinides partitioning

Position Report on minor actinides partitioning and transmutation

Core manufacturing

workshop (AFC)

MA bearing fuels fabrication facility

Commercial deployment: from 2040-2050

GENERAL SCHEDULE


Mid of 2010: preliminary selection of ASTRID characteristics for launching the preconceptual design

Preconceptual design: 2010 to end of 2012: The preconceptual design is considering some open options. Innovation and technological breakthroughs are favored, while maintaining risk at an acceptable levelDuring the preconceptual design phase, start of the interactions with the Safety authorities on safety objectives and orientationFirst estimation of ASTRID investment cost, including the different open optionsSchedule of next steps and their associated costs Safety orientation report: report delivering and first advices of the French Safety Authorities

End 2014: Conceptual designSafety Options File

MAIN MILESTONES

CEA | JUNE 2012 | PAGE 9

GENERAL FEATURES OF ASTRID

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CEA | JUNE 2012

1500 thMW - ~600 eMW pool type reactorWith an intermediate sodium circuit High level expectations in terms of safety demonstrationPreliminary strategy for severe accidents (core catcher…)Diversified decay heat removal systemsOxide fuel UO2-PuO2 for starting cores Transmutation capabilityFuel handling in sodium….

Main features Open options

Core designEnergy conversion systemReference : water/steamAlternative : N2 Number of loopsDevices to eliminate severe accidents (i.e. 3rd shutdown level)Core catcher technologySGs materials and technologyInnovative technologies for Na fires detection and masteringI&C

• …

Carbide fuel

SiC-SiC materials

…Many of these options will be decided during the pre-conceptual design up to end 2012,

Innovative options to be tested

PRELIMINARY DESIGN CHOISES / OPEN OPTIONS


DESIGN OF THE NUCLEAR ISLAND

CEA | JUNE 2012

MATERIAL SCOPE

Core structures400 – 700°C nominal850-900°C accidental

High irradiation No swelling

Fixed structuresLife time

40 �60 years

Vessel400°Cno deformationnegligeable creep

Cold structuresHeat exchanger, Pumps400°C + low irradiation no deformation

Hot structures550°C + low irradiation

creep, seal coef., weldings

SGs, Energy conversion System

350 – 525°Cageing, weldings, compatibility

Pipes and circuits350 – 550°C

creep, fatigue, creep-fatigue, thermal fatigue,ageing

weldings

First coreEM10->T91/T92 ?

AIM1->AIM2 ?

316 LN 316 LN

316 LNNi alloy

316 LN800H9 Cr ?

316 LN9 Cr ?

Specific field of investigation : long

term process, nuclear facilities…

Long term R & DF/M ODS->Base V ?Ceramics->SiC/SiC ?

To extend R&D programs on accidental conditions (fr om the Fukushima accident): high temperature material beha viour: creep, creep-fatigue,

Stellites replacements: aims to gain benefits for decommissioning,

Welded seals behaviour,

Material behaviour justification up to lifetimes of 60 years

MATERIAL SCOPE – ASTRID FOCUS


LAY-OUT EXAMPLES

4 secondary loops


A progressive approachflexible to meet the decisions according to June 28, 2006 French Act on Wastes Management –1st milestone � end 2012

2 levels of demonstration:experimental stage� Pellets� A few pins� A subassemblyreserved positions for several subassemblies

2 ways considered : homogeneous and heterogeneous

Transmutation of Am, eventually Np and Cm

Specification for ASTRID in 2012

TRANSMUTATION CAPABILITY


SAFETY: EXEMPLES OF MAJOR IMPROVEMENTS

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CEA | JUNE 2012

ASTRID SAFETY APPROACH


A CORE WITH ENHANCED SAFETY

Feedback experiencefrom SPX and EFR :

� To reduce the fuel reactivity loss per cycle

� To reduce the sodium void worth

2009 : heterogeneities added to SFRv2 core

� Sodium void worth strongly reduced

� Quantification of the potentialbenefit from safety point of view ison going

Absorbing protection

Sodium plenumzone

Outer fissilezone

Inner fertile zone

Upper inner fissile zone

Lower inner fissile zone

Fertile blanket

Neutronic protection


Sodium plenumzone

Outer fissilezone

Inner fertile zone



Fertile blanket


CFV core

2008 : concept with larger pin and smaller-diameter spacing wire� Increase of the fuel fraction

� Low reactivity loss during cycle

� Decrease of the Na fraction ⇒ lower Na voiding effect

� Progress on control rod withdrawal

SFRv2 core

� Choice of the core: september 2012


SFRv2 – AIM1 – 1500 MWth CFV – V1 - AIM1 – 1500 MWth

Spatial distribution of the sodium void effect Spatial distribution

of Doppler

EA

NO

S/P

AR

IS C

ALC

ULA

TIO

NS

CORE DESIGN OPTIONS: SFRV2 AND CFV LAYOUTS


Core 1500 MWth SFRV2B CFV-V1Number of fuel pin / SA 169 217

Fuel pin diameter (mm) 9,43 8,45

Pu enrichment E1/ E2 (%) 13,9 / 17,6 23,5 / 20

Height H1 / H2 (cm) 110 80 / 90

Number of SA C1 / C2 144 / 144 177 / 114

Number batch / Fuel cycle lenght

4 x 390 JEPP

4 x 360 JEPP

Void effect ($) - RZ +5,1 -0,5

Breeding gain -0,05 -0,02

TCT average fissile C1/C2 (GWj/t)

76 / 67 105 / 69

∆ρ∆ρ∆ρ∆ρ / cycle (pcm/jepp) -2,2 -4,3

Number of CR 18 + 6 (*) 12 + 6

Core diameter (cm) 326 340

Plin max BOL (W/cm) 407 483

Amplification based on individual effects

CORE DESIGN OPTIONS: SFRV2 AND CFV CHARACTERISTICS


Best estimate calculations ⇒⇒⇒⇒ trends

(uncertainty analysis is on-going)

CFV V1 V2B

ULOSSP Margin to Na boiling of ≈≈≈≈ 55°C Na boiling in ≈≈≈≈ 100s

ULOF Na boiling in ≈≈≈≈ 3500s Na boiling in ≈≈≈≈ 100s

ULOHSTemperature of neutronic

shutdown : 700°CTemperature of neutronic

shutdown : 800°C

LIPOSO 680°C / 45%Pn 736°C / 43%Qn

• CFV core: a promising core for an improved intrinsi c behaviour in case of unprotected situations and control rod withdrawal

•Analysis of severe accidents conditions are on-goin g

•On-going definition of version 2 of CFV core for im proving inherent behaviour (with the objectives to increase the sodium boiling margin and robust demon stration of no fuel melting in case of CRW)

COMPARISON OF PRELIMINARY RESULTS ON UNPROTECTED LOSS OF FLOW SITUATIONS


Natural behavior favorable for transients of unprotec tedloss of flow and loss of heat sinkTarget criteria : no sodium boiling for a ULOSSP transient

Sodium void effect minimizedTarget criteria : Na void effect < 0

Natural behavior favorable for a complete control rodwithdrawal (with no detection) Target criteria : no fuel melt

Improved performancesTarget criteria : Cycle length ≈ 480 efpd, High fuel burnup, and breeding gain ≈ 0

Core design extrapolable to higher power

CORE DESIGN OBJECTIVES


2. Mitigation of the core meltdown To garantee that core meltdown accidents don’t lead to significant mechanical energy release, whatever ini tiator event

by a favorable natural core behavior (negative sodium void worth for CFV type core)by adding specific mitigation dispositions in case of natural behavior is not sufficient


Sodium plenumzone

Outer fissilezone

Inner fertile zone



Fertile blanket



Sodium plenumzone

Outer fissilezone

Inner fertile zone



Fertile blanket


ASTRID core design is mainly guided by safety objectives:1. Prevention of the core meltdown accident

by a natural behavior of the core and the reactor (as 3rd line of defense in case of no actuation of the two shutdown systems)• Natural behavior favorable for transients of unprot ected loss of flow and loss of heat sink

Target criteria : no sodium boiling for a ULOSSP transient for CFV type core (CFV = Low Na void worth core)• Sodium void effect minimized

Target criteria : Na void effect < 0 for CFV type core• Natural behavior favorable for a complete control r od withdrawal (with no detection)

Target criteria : no fuel fusion

by adding passive complementary systems if natural behavior is not sufficient for some transient cases

CORE DESIGN APPROACH


ISSUES TO BE INVESTIGATED BY EXPERIMENTS

Measures for In-Vessel-Retention (IVR) have priority but 3 options are open

External core-catcherCore-catcher between 2 vessels

Internal core-catcher

Design studies for ASTRID aim at preventing the risk of core melting.

However according to the WENRA 2010 « Safety Objectives for New Nuclear Power Plants », 4th level of in depth prevention request that this accident will be taken into account in the design process.


DECAY HEAT REMOVAL ARCHITECTURE

Sodium/air loop, using secondary circuit (connecte d circuit on secondary pipes, or dedicated heat exchanger integr ated in Intermediate Heat Exchanger)

Circulation of air along steam generators vessels

Dedicated circuit set outside ofreactor vessel, using vesselwall as heat exchange surface

Dedicated sodium/air loop,with its own heat exchangerlocated in the primary circuit

4 – Na/air loop, dedicated loop in the

main vessel

1 – Na/air loop, using secondary loop

3 – cooling through the vessel

2 – air circulation along steam generator


SODIUM WATER REACTIONS

Sodium – water reactionViolent and exothermic reactionMain reaction : Na + H2O � NaOH + ½ H2 + 162 kJ/ mole of water (at 500°C)

Effects of a sodium-water reaction in a Steam Generator

� Chemical effectsGlobal corrosion in polluted sodium environmentLocal erosion / corrosion (« wastage »)� self-evolution of the tube leak orifice� damage of the nearest tubes

� Mechanical effectsFor large leaks(>100 g/s)

� Fast over-pressure associated to pressurewave propagation

� Slow over-pressure associated to massive introduction of water in secondary sodium circuit

� Thermal effectsFor large leaks(>100 g/s) and due to exothermal reaction

� effects on tubes : heating, creeping, swelling, burst

2 ways to reduce the SWR risk:– Improve SG design of the steam PCS (Rankine cycle) in order to :

• reduce the risk of SWR occurrence• limit the consequences of an hypothetical violent reaction

– PCS (Brayton cycle with pure nitrogen at 180 bar) in place of steam cycle to eliminate de facto the SWR risk

• Feasibility to be demonstrated.CEA | JUNE 2012

STEAM POWER CONVERSION SYSTEM (1/2)

Innovation : modular SGsProtection of secondary piping and intermediate heat exchanger integrity in case of simultaneous failure of all the tubes of a SG module (accidental envelope case)

Imply a maximal SG power of about150 MWth

Improvement of detectionHydrogen detection by means of permeationthrough very sensitive nickel membrane but :Complicated fabrication and operationResponse time to be optimized An electro-chemical flow meter has been tested recently in PHENIXSimplerPossible optimization of response timeStudy of diversified detection method based on acoustic principle in progress

Protection against large sodium-water reactions ensured by :Passive fast draining of the sodium loop by means of rupture diskFast insulation and depressurization of the water-steam loop


STEAM POWER CONVERSION SYSTEM (2/2)

Reverse SG (sodium inside the tubes)� Innovative approach with SG no sensitive to wastage phenomena occurring in

classical SGs

Reduce the probability of crack to leak evolution (tubes with external pressure)

Speeds up the leak detection

Slow down drastically the propagation of a possible SWR

Feed-back : 2 designs of reverse SG operated in BOR-60 reactor

Preliminary design of 125 MWth modules achieved

Issues :Dimensioning of external pressurized shellModeling of the sodium-water reactionGeneral design of the reverse SGIn service inspection CEA | JUNE 2012 | PAGE 29

GAS POWER CONVERSION SYSTEM (1/2)

Very innovative concept with feasibility to be demonstrated Nitrogen selected as gas, at a pressure of 180 barsEncouraging first resultsNet efficiency of the reactor plant about 38% possibleTurbomachinery

TurbineTwo possible designs: single flow and split flow with high isentropic efficiency (94%)

Turbine design challenging but not unfeasible (no showstopper identified)

Compressor

Two technologies (axial and radial) with equivalent isentropic efficiency (90%)

Radial technology should be put forward : simpler, cheaper, no performance test required

HP and LP compressors with same technology


GAS POWER CONVERSION SYSTEM (2/2)

Key points on sodium/gas heat exchangersCompact heat exchangers

PCHE heat exchanger technology choosenTechnological issue of the gas PCS: codification,fabrication control and In Service Inspection

Tubes and shell heat exchangersMore robust back-up option, but very heavy components and largefloor space requirements, feasibility and integration to be detailed.

General architectureInvestigation in progress to optimizepiping layout, performance(pressure losses), accessibility,maintenance and operation.


SURVEILLANCE AND ISI&R: A FOUR LEVEL STRATEGY

AS

TR

ID IS

I&R

L1: Continuous monitoring

Statutory ISI

L2: Periodic examination

Statutory ISI

L3: Exceptional interventions

Doubts / warnings

L4: Repair

Investment protection

Looking for …* Operating parameters variation

* Abnormal deformation of structures

* Vibrations

* Leakage

* Safety …

Considered options•improve the prevention level

• Robust and redundant detection systems

•Innovative instrumentation

•Up to date technologies

* Excessive deformations

* Fatigue (or creep-fatigue) cracks next to the welded junction

* Corrosion / loss of thickness

* Erosion on rotating parts

* …

•ISI&R oriented design

•Under Na telemetry

•Under Na volumetric NDT

•Under sodium robotic carriers

• …

* The same type as for Level 2

• Localization : everywhere !• The same type as for Level 2

* Noxious cracks

* Loss of parts

* Stuck mechanisms

* Out of use primary components

* …

Repair …•Removable components

•Specific repair tools

• …


Inlet and outlet coretemperature∆T, CRW(TIB ?)

TC, optical fiber, flowmeter

Inlet and outlet coreflowrate

TIB, loss of flow

Surveillance ciel de pileGaseous FP Clad failure

Localisation in gazGaseous FPIonisation chamber,

Spectro g, (CRDS)

Neutronic monitoringFonctionnement, CRW

High Temperature FissionChambers with largedynamics

Clad Failure Detection

in Na Open CF, BTI

HTFC / IHX

Clad failure Localisation in Na

FP neutronic detection

FC with B

US telemetry under NaCore movment,(Temperature)

Acoustic DetectionAbnormal noise (TIB?)

Active Acoustic Det.Gas rate

SONAR, TUSHT…

CoreCFV

InnovativeLine of Defense A

( ) R&D

SURVEILLANCE AND PROTECTION OF THE CORE


R&D results [CEA-AREVA-EDF] obtained from 2007 to 2009 have contributed to ASTRID mid 2010 choice of options

ASTRID has the objective to demonstrate at the industrial scale progress in the identified domains of SFR weakness (safety, operability, economy) and to perform transmutation demonstrations

A lot of improvements are related to safety

The first very important milestone is 2012 (June 2006 French Act on wastes management) :

ASTRID pre-conceptual design studies : 2010-2012

First investment cost evaluation

First safety Authorities advice on the orientations for ASTRID safety

With the ASTRID program funded by the French government, France has the opportunity to develop a GEN IV Sodium Fast Reactor

CONCLUSIONS


1988-1998

1985 - 1998

1973 - 2010

ASTRID

Commercial reactor

1967 - 1983

RAPSODIE

FAVOURABLE CHARACTERISTICS OF SFR

Easy to operate: no pressurization of the primary coolant, high thermal inertia, control by single rod position, no xenon effect, no need of soluble neutron poisonRadiation protection : higher level of protection than LWRFew effluents and little radioactive waste High thermal efficiency Large sodium boiling marginNatural convectionDiversification of heat sink by using air

Rapsodie (1967-1983)

Phénix (1973-2010) Super-Phénix (1985-1998)


Nuclear Energy DirectorateReactor Studies Department

Commissariat à l’énergie atomique et aux énergies alternatives

Centre de Cadarache | 13108 Saint Paul Lez Durance

T. +33 (0)442257000

Etablissement public à caractère industriel et commercial | RCS Paris B 775 685 019

| PAGE 37

CEA | JUNE 2012

Documents

ADVANCED SODIUM TECHNOLOGICAL REACTOR … SODIUM TECHNOLOGICAL REACTOR FOR INDUSTRIAL DEMONSTRATION ... Irradiation services and options test. PROJECT ORGANIZATION ... SODIUM WATER