gif lead-cooled fast reactor development status

GIF Lead-cooled Fast Reactor Development Status

INPRO Dialogue Forum on Generation IV Nuclear Energy Systems IAEA Headquarters, Vienna. 13-15 April 2016

Alessandro Alemberti

(EURATOM / Ansaldo Nucleare)

on behalf of

GIF LFR

provisional System Steering Committee

OUTLINE

Some general characteristics of LFR

The Three GIF–LFR Reference Systems

Activities of LFR provisional System Steering Committee

Status of LFR Development in MoU countries:

Japan, Republic of Korea, Russian Federation, Euratom

ALFRED: an example of LFR design

LEAD coolant – new design possibilities

� Lead does not react with water or air (slow/moderate reaction - Pb oxide)Steam Generators can be installed inside the Reactor Vessel

� Very high boiling point (1745°C ), low vapor pressure (3 10-5 Pa @ 400°C)No core voiding reactivity risk due to coolant boiling

� Lead has a high densityAnalyze fuel dispersion/compaction effect

� Lead is a low moderating medium and has low absorption cross-sectionNo need of compact fuel rods (large p/d defined by T/H) Primary pressure losses can be maintained very lowHigh primary natural circulation capability natural circulation DHR

LEAD COOLANT PASSIVE SAFETY

LEAD coolant �� design provisions

High Lead melting point (~ 327 °C) – assure Lead T above 340-350 °CHeating system, design and operating procedures

Overcooling transient (secondary side) may cause Lead freezingFW requirement

Corrosion / erosion of structural materials - Slugging of primary coolantOxygen control, Coatings, Limit flow velocity Strategy at low oxygen content, Lead chemistry (alternative approach)

Seismic risk due to large mass of lead2-D seismic isolators, vessel hanged, specific design

In-service inspection of core support structuresSimilar to other HLM reactors but high T, all components replaceable

Fuel loading/unloading by remote handlingDevelop appropriate cooling system (active � passive back-up)

Steam Generator Tube rupture inside the primary systemShow no effect on core, rupture disks/Safety valves on reactor cover

Flow blockage and mitigation of Flow blockage accident Hexagonal wrapped FAs – Outlet temperature continuous monitoringMultiple FA flow inlet or Shroudless FA design

Example of Closed Fuel Cycle in Fast Reactors

� LFR can be operated as adiabatic: �Waste only FP, feed only Unat/dep�Pu vector slowly evolves cycle by cycle

� MA content increases and its composition drift in the time� LFR is fully sustainable and proliferation resistant (since the start up)� Pu and MA are constant in quantities and vectors� Safety - main feedback and kinetic parameters vs MA content

Fabrication LFR

AdiabaticReprocessing

All Actinides

MOX first loads

Unat/dep

FP

+ losses

MOX equilibrium

GIF–LFR REFERENCE SYSTEMS

the three reference systems of GIF–LFR are:

ELFR (600 MWe), BREST (300 MWe), and SSTAR (small size)

Members (MoU) of System Steering Committee: EURATOM, RUSSIA, JAPAN,KOREA

Observers to pSSC activities: USA, CHINA

11

22

33

44

55

1 - Core

2 - Steam Generator

3 - Pump

4 - Refueling Machine

5 - Reactor Vault

CLOSURE HEAD

CO2 INLET NOZZLE

(1 OF 4)

CO2 OUTLET NOZZLE

(1 OF 8)

Pb-TO-CO2 HEAT EXCHANGER (1 OF 4)

ACTIVE CORE AND FISSION GAS PLENUM

RADIAL REFLECTOR

FLOW DISTRIBUTOR HEAD

FLOW SHROUDGUARD VESSEL

REACTOR VESSEL

CONTROL ROD DRIVES

CONTROL

ROD GUIDE TUBES AND DRIVELINES

THERMAL BAFFLE

ELFR

system for central station

power generation

BREST

system of

intermediate size

SSTAR

system of small size

with long core life

SG

Reactor

Vessel

Safety

Vessel

DHR dip

cooler

FAs

Primary

Pump

Status of the main activities:

SRP, SDC, SSA, IRSN report

• LFR System Research Plan (SRP):

Final draft of LFR SRP has been issued by pSSC and sent to EG

• LFR White Paper on safety:

Final version of the White Paper is available to the public on the GIF web-site

https://www.gen-4.org/gif/jcms/c_67650/lead-cooled-fast-reactor-lfr-risk-and-safety-

assessment-white-paper?hlText=white+paper

• LFR Safety Design Criteria (SDC):

LFR Safety Design Criteria have been developed on the basis of SDC for SFRs.

Submitted to RSWG for review in December of 2015

• LFR System Safety Assessment (SSA):.

Final draft version of LFR-SSA sent to RSWG on January 2016

In Japan, fundamental R&D studies are carried out at the Tokyo Institute of Technology to

support LFR technology development. These include

JAPAN

Investigations at Tokyo Institute of Technology

• Design activities related to:

o Nuclear design: Low void reactivity, Long life core, CANDLE concept

o Thermal-hydraulic and structural design

o Plant design

o Safety analysis (UTOP, ULOF, ULOHS)

• Thermal-hydraulic tests

• Material compatibility tests: Corrosion-resistant material (Al/Si-added steels, Al/Fe-

alloy-coated steels, ceramics, refractory metals), Effect of stresses, cold-work and

welding on steel corrosion, Erosion phenomena

• Po test: Po removability, Filtration

• Oxygen control tests, incl. oxygen sensor developments, oxygen control with gas

and PbO

• LBE property test: Diffusivity of impurities, etc.

• Analytical studies: Core calculation, Thermal-hydraulics, MD simulation

During 2015 main studies have been dedicated to CANDLE reactor concept

The CANDLE Reactor Concept• The CANDLE (Constant Axial shape of Neutron flux, nuclide densityand power shape During Life of Energy production) reactor concepthas been considered for very high uranium fuel utilization withoutreprocessing.

– Its burning region propagates along the axial direction withoutchanging the spatial distribution and it have higher burn-updischarged.

– The burning region moving speed is generally slow; hence, it isfeasible to achieve a very long-life reactor core.

– Sekimoto et al. reported that CANDLE has a maximum fuel burnup ofup to about 40% without enrichment or reprocessing

– Currently no fuel cladding material that can withstand such highburnup �� necessary to ensure material integrity under ~ 40% ofburnup if CANDLE burning can be utilized for a long-life core.

*Source: Sekimoto, H., 2010. Light a CANDLE, An Innovative Burnup Strategy of Nuclear Reactors. Center

for Research into Innovative Nuclear Energy Systems (CRINES), Tokyo Institute of Technology, Tokyo.

Fig. 1: Concept of CANDLE

burning*

S-9

JAPAN

IN NOVEMBER 2015 KOREA signed the GIF-LFR MoU becoming full member of the GIF-LFR provisional System

Steering Committee

Summary of ROK activities on GIF-LFR

• GIF-LFR-pSSC MoU signed by SNU November 2015

• SMR – URANUS Kickoff with MIT and Private Industries

• OECD/NEA LACANES Benchmark on Natural Circulation

• SNF/HLW Security Solution – PASCAR

• A new pool-type test facility PILLAR is prepared to be built

• Several collaborations with MIT are ongoing

Republic of KOREA

LFR Design Development� URANUS

– Ubiquitous

�3D seismic isolation

– Robust

�Underground

– Accident-forgiving

�LBE coolant

�Fully passive safety system

– Non-proliferating

�Capsular core

– Ultra-lasting

�20 years refueling cycle

– Sustainer

�CO2 Sequestration

Republic of KOREA

2.5% Al added

0.3%V removed,

0.9%Nb added

11Cr (HT9) 11Cr-2.5Al-0.3V (Alloy-1a) 11Cr-2.5Al-0.9Nb(Alloy-1b)

14.3Cr-2.5Al-0.9Nb (Alloy-1f)

Cr increased

14.3Cr-2.5Al-0.3V (Alloy-1d)

0.3%V removed,

0.9%Nb added

Cr increased

11Cr-3.5Al-0.9Nb(Alloy-1c)

- Continuous Al2O3 formed on 14.3Cr-2.5Al-0.9Nb(Alloy-1f) alloy at 600oC

- Nb addition promote Al2O3 formation

• Corrosion tests in LBE at 600oC after 300hr

Advanced Cladding Materials by SNU

Republic of KOREA

Global Symposium on Lead and Lead Alloy Cooled Nuclear Energy Science and Technology (GLANST)

Website hosted by OECD/NEA GIF secretariat

http://peacer.org/new/glanst2016.php

Organizational Structure

– Scientific Committee consisted of GIF LFR pSSC members

– Chaired by GIF LFR pSSC

– Organized in GIF countries every 5 years between two HLMC events

REGISTRATION FEE: US $200.00 (Proceedings, Reception, Banquet)

ACCOMODATION: Seoul National University(SNU) Hoam Faculty House & Hotels

CALL FOR PAPERS CALL FOR PAPERS CALL FOR PAPERS CALL FOR PAPERS 2222----Page Summary submission deadline: May Page Summary submission deadline: May Page Summary submission deadline: May Page Summary submission deadline: May 31313131, , , , 2016201620162016Use Template for Use Template for Use Template for Use Template for 2222----Page Summary for Submission to glanstPage Summary for Submission to glanstPage Summary for Submission to glanstPage Summary for Submission to [email protected]@[email protected]@peacer.org

Seoul, Korea Seoul, Korea Seoul, Korea Seoul, Korea November November November November 16161616----18181818,,,,2016 2016 2016 2016 Seoul National University (SNU)Seoul National University (SNU)Seoul National University (SNU)Seoul National University (SNU)

Republic of KOREA

Main Coolant Pump

Steam GeneratorVessel Core

Emergency Cooling

System header

RUSSIAN FEDERATION

BREST–OD–300: Design concept

• BREST features an integral primary circuit

layout combined with a multilayer metal-

concrete vessel to exclude risk for primary

coolant losses

• There are no shutoff valves in the primary

circuit and a high degree of natural

circulation flow can be maintained in the

primary circuit of BREST during the loss of

AC power

• The use of highly dense and highly heat-

conductive nitride fuel allows breeding

inside the BREST core (BR~1.05). This limits

excess reactivity requirements and excludes

risks for severe accidental reactivity

insertions

• BREST employs a passive emergency

cooling system with natural circulation and

removal of decay heat to the atmospheric air

RUSSIAN FEDERATION

RUSSIAN FEDERATION

BREST–OD–300: Safety assessments

Comprehensive safety assessments have

been performed in support to licensing of

BREST. This included:

• Analyses of anticipated operational

occurrences (AOO) accompanied by

postulated failures of systems, components

or by personnel errors

• Analyses of progression of these anticipated

operational occurrences accompanied by

multiple failures of systems, components or

by personnel errors

As an enveloping case, an analysis of station

black-out accident has also been performed:

• Decay heat was assumed to be removed by

two out of four emergency cooling loops

The results of these analyses show that no

cladding or fuel melting is to be expected

and that the integrity of the primary system

of BREST is maintained

0

0,2

0,4

0,6

0,8

1

1,2

0 50 100 150 200 250 300 350

Power (N) and flow rate (G) [relative units]

Time [s]

N

G

300

400

500

600

700

800

900

1000

1100

1200

1300

0 50 100 150 200 250 300 350

Temperature Т [°°°°С]

Time [s]

ТF

Тcl

ТSG in.Тcore out.

ТSG out.

Тcore in.

Station blackout

BREST–OD–300 SCHEDULE:

Design completed 2014

License approval 2015-16

Start of construction 2017

Commissioning 2020-2022

RUSSIAN FEDERATION

Lead & LBE technology development in Europe

Presently two main projects (with many synergies):

EURATOM

MYRRHA (LBE) ALFRED (LFR)

EURATOM

Flexible irradiation facility

To be used to support fuel development

for Gen IV systems

SUPPORT TO ALFRED CONSTRUCTION IN ROMANIA

THE FALCON CONSORTIUM

• Unincorporated consortium

• In-kind contributions

• Optimize the cooperation

• Activities: strategic, management,

governance, financial and

technical aspects

• Detailed agreement

• R&D needs management

• Engineering design

• Licensing, and

• Commit the construction

2PHASE

1PHASE

EURATOM

MAIN COOLANT PUMP

REACTOR VESSEL SAFETY VESSEL

FUEL ASSEMBLIES

STEAM

GENERATOR

STEAM

GENERATOR

MAIN COOLANT PUMP

REACTOR CORE

ALFRED - Reactor Configuration

Power: 300 MWth (125 MWe)

Primary cycle: 400 ‒ 480ºC

Secondary cycle: 335 ‒ 450ºC

Steam cycle efficiency above 40%

ALFRED - Core Configuration

Power (MWth) 300

Fuel Assemblies

Inner core

Outer core

171

57

114

Fuel type MOX

Pu enrichment

Inn/Out (at %)21.7/27.8

Control/shutdown

Rods12

Safety Rods 4

Dummy elements 110

Fuel Batches 5

Fuel cycle length 365 EFPD

Peak/avg BU

(MWd/t)103/73.3

ALFRED – Fuel Pin & Assembly

1390

Overall length 8000

ALFRED - Reactor Control and Shutdown System

• Two redundant, independent and diverse shutdown systems are

designed by SCK•CEN for MYRRHA, adapted to ALFRED

• The Control Rod (CR) system used for both normal control of the reactor and for SCRAM in case of emergency

– CR are extracted downward and rise up by buoyancy in case of emergency shutdown (SCRAM)

– During reactor operation at power CR are most of the time partly inserted allowing reactor power tuning (each rod is inserted for a maximum worth less than 1$ of reactivity)

• The Safety Rod (SR) system is the redundant and diversified complement to CR used only for emergency shutdown SCRAM

– SR are fully extracted during operation at power

– SR are extracted upward and inserted downward by the actuation of a pneumatic system (insertion by depressurization – fail safe)

– A Tungsten ballast is used to maintain SR inserted

• Reactive worth of each shutdown system is able to shutdown the reactor even if the most reactive rod of the system is postulated stuck

CR SR

ALFRED - Upper and Lower Core Support Plates

Lower core support plate

Box structure with two horizontal perforated plates

connected by vertical plates.

Plates holes are the housing of FAs foots.

The plates distance assures the verticality of Fas.

Hole for

Instruments

Box structure as lower grid but more stiff.

It has the function to push down the FAs

during the reactor operation

A series of preloaded disk springs presses

each FA on its lower housing

Upper core support plate

ALFRED - Inner Vessel

Inner Vessel assembly

Upper grid

Lower gridPin

Inner Vessel has the

main functions of

core support and

hot/cold plena

separation.

Fixed to the cover by

bolts and radially

restrained at bottom

(replaceable).

Core Support plate is

mechanically

connected to the IV

with pins for easy

removal/replacement

ALFRED - Steam Generator

� Bayonet vertical tube withexternal safety tube and internalinsulating layer

� The internal insulating layer(delimited by the Slave tube) hasbeen introduced to ensure theproduction of superheated drysteam

� The gap between the outermostand the outer bayonet tube is filledwith pressurized helium to permitcontinuous monitoring of thetube bundle integrity. highconductivities particles are addedto the gap to enhance the heatexchange capability

� In case of tube leak thisarrangement guarantees thatprimary lead does not interact withthe secondary water

Water

Hot Lead

Cold Lead

Steam

Bayonet Tube Concept

ALFRED - Reactor Vessel

Cylindrical Vessel with a torispherical bottom head

Anchored to the reactor cavity from the top

Cone frustum, welded to the bottom head, provides radial support of the Inner Vessel

Inner Vessel radial support

Support flangeCover flange

Main Dimensions

Height, m 10.13

Inner diameter, m 8

Wall thickness, mm 50

Design temperature, °C 400

Vessel material AISI 316L

ALFRED - Decay Heat Removal Strategy

Use first non safety-grade system, active systems for normal decay heat removal

Two independent, high reliable passive and redundant safety-related Decay Heat Removal systems (DHR 1 and 2):

- in case of unavailability of the active components of secondary system, the DHR-1 is called to operate- in the unlike event of unavailability of the first two systems, the DHR-2 system is called to operate

DHR N1: Isolation Condenser connected to 4 out of 8 SGs

DHR N2: Isolation Condenser using 4 dedicated Coolers

DHR-1 operation:

FW and SL isolation

opening of IC isolation valve

Water injection in SG

Steam produced condensed in IC

Completely passive operation

ALFRED - DHR System Performances

300

320

340

360

380

400

420

440

460

480

500

0 2,000 4,000 6,000 8,000 10,000 12,000 14,000 16,000

°C

s

Core inlet temp

Core outlet temp

300

350

400

450

500

550

0 5,000 10,000 15,000 20,000 25,000

°C

s

Core inlet temp

Core outlet temp

3 Loops in operation (Minimum performances)

Lead Peak Temperature ∼ 500°C

Time to freeze > 8 hours

4 Loops in operation (Maximum performances)

Lead temperature < nominal

Time to freeze ∼ 4 hours

Freezing temperatureFreezing temperature

“Anti-freezing” system for DHR in HLM-cooled

reactors:

• Noncondensable gas tank connected to IC

lower header;

• When power removed by ICS > primary power

⇒ ICS depressurises

⇒ noncondensable gas expands

• ⇒ NC-Gas reach IC tubes

⇒ heat exchange is reduced

• pICS stabilised;

• Tprimary stabilised > Tfreezing

Pb pool

Steam Generator

Condenser

Noncondensable gas tank

DHR “ANTI-FREEZING” SYSTEM

ALFRED – New DHR1 – grace time to freezing

300

350

400

450

500

550

600

0 1 2 3 4 5 6 7 8 9 10T

em

pera

ture

[°C

]

time [days]

(d) Pb temperature at SG outlet

SG outlet TPb freezing point

300

350

400

450

500

0 50 100 150 200 250

Tem

pera

ture

[°C

]

time [min]

(c) Pb temperature at SG outlet

SG outlet T

Pb freezing point

Innovative DHR SystemOld DHR System

ALFRED: Lead temperature during Station Black Out

OLD DHR1- about 4 hours to freezing

NEW DHR1 with NC – 4-5 Days (judged sufficient for

operator action)

gif lead-cooled fast reactor development status

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