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
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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
Slide 2
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
Slide 3
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
Slide 4
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
Slide 5
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
Slide 6
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
Slide 7
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
During 2015 main studies have been dedicated to CANDLE reactor concept
Slide 9
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
Slide 10
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
- 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
Slide 13
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
Slide 14
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
Slide 15
RUSSIAN FEDERATION
Slide 16
RUSSIAN FEDERATION
Slide 17
RUSSIAN FEDERATION
Slide 18
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
Slide 19
BREST–OD–300 SCHEDULE:
Design completed 2014
License approval 2015-16
Start of construction 2017
Commissioning 2020-2022
RUSSIAN FEDERATION
Slide 20
Lead & LBE technology development in Europe
Presently two main projects (with many synergies):
EURATOM
MYRRHA (LBE) ALFRED (LFR)
Slide 21
EURATOM
Flexible irradiation facility
To be used to support fuel development
for Gen IV systems
Slide 22
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
Slide 23
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%
Slide 24
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
Slide 25
ALFRED – Fuel Pin & Assembly
1390
Overall length 8000
Slide 26
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
Slide 27
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
Slide 28
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
Slide 29
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
Slide 30
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
Slide 31
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