Status of Lead-cooled Fast Reactor GIF – LFR Activities
9th GIF-IAEA Interface Meeting IAEA Headquarters, Vienna, 4-5 March 2015 Meeting Room: C0343
Alessandro Alemberti
EURATOM
on behalf of
GIF – LFR – pSSC
Presented by: Kamil Tuček (EURATOM)
Slide 2
OUTLINE
Three GIF–LFR Reference Systems
Status of Main Activities of LFR Provisional SSC (pSSC)
New Members of LFR pSSC
Status of LFR Development in MoU countries / entities
JAPAN
RUSSIAN FEDERATION
EURATOM
Slide 3
GIF–LFR REFERENCE SYSTEMS
Three reference systems of GIF–LFR are:
ELFR (600 MWe), BREST (300 MWe), and SSTAR (small size)
Members (MoU) of provisional System Steering Committee: EURATOM, RUSSIA, JAPAN
Observers to pSSC activities: USA, Korea, China
1
2
3
4
5
1 - Core
2 - Steam Generator
3 - Pump
4 - Refueling Machine
5 - Reactor Vault
CLOSURE HEAD
CO 2 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 4
Status of the main activities: SRP, White Paper, SDC, ToR
• SYSTEM RESEARCH PLAN (SRP):
Substantial revision of SRP was started by mid-2012 and is now completed. Final draft
of SRP has been issued by pSSC and the report is currently being reviewed by EG
• LFR White Paper on safety: Review of White Paper on safety (based on ALFRED as
an example of an LFR to apply ISAM to) was completed by EG. The final version of
the paper has already been published on the GIF web-site by RSWG
• LFR – Safety Design Criteria:
Safety Design Criteria (SDC) for LFR will be developed on the basis of SDC for SFRs
Work is still ongoing, first draft is expected to be available by spring 2015
• GIF–LFR abstract was sent to GIF Symposium held in conjunction with the ICONE23
conference in Japan (May 2015)
• Preparation of draft “Terms of reference for GIF system safety assessment” is
currently ongoing and the draft is expected to be available shortly (by March 2015)
• 2014 Annual report was sent to the GIF Secretariat in the first week of December
• Revision of LFR information on the GIF web-site will be available by spring 2015
Slide 5
Two new members of LFR pSSC
• RUSSIAN FEDERATION
Valery Smirnov (NIKIET) replaced by Andrei Moissev (NIKIET)
• EURATOM
Didier Haas (JRC) replaced by Kamil Tuček (JRC)
The LFR-pSSC participants expressed their gratitude for the work
performed by Valery and Didier and for the friendly relationship
established
A warm welcome in the group to Andrei and Kamil
Slide 6
Although activities on LFRs are limited in Japan, basic R&D is carried out at the
Tokyo Institute of Technology to support LFR technology development.
Two examples are given below:
JAPAN
Burn-out of heater pin
Investigations at Tokyo Institute of Technology
Experimental facility to investigate flow boiling
Slide 7
Development and testing of reliable oxygen control systems for conditioning
and maintenance of a required level of dissolved oxygen in heavy liquid
metals. Systems featuring either some suitable gas mixtures or solid Pb
oxide (PbO) are investigated:
JAPAN
Investigations at Tokyo Institute of Technology
Slide 8
Main Coolant Pump
Steam Generator Vessel Core
Configuration Pool
Thermal power, MW 700
Electric power, MW 300
Fuel (U,Pu)N
Steam production rate, no less than, t/hour 1480
Coolant of the primary circuit Lead
Gas pressure above the lead level:
– exceed, MPa
– maximal, MPa
0.003–0.008
0.02
Average temperature of the lead coolant at
the active zone inlet / outlet, °С 420 / 540
Average temperature of the water coolant
at the steam generator inlet / outlet, °С 340 / 505
Loop number 4
FA number in the active zone 169
Core height, mm 1100
Fuel load, t 20.6
Fuel campaign, years 5
Discharge burn-up of fuel (max./av.), % HM 9.0 / 5.5
Average Conversion Ratio (CR) ~ 1.05
Operating reactivity margin at Pnom < 1 beff
Collector
RUSSIAN FEDERATION
BREST–OD–300: Key components and technical characteristics
Control
and safety
rod drives
Slide 9
Results: Tests were performed on short-term mechanical properties of the concrete,
methods were developed for basic thermal-mechanical strength analysis, mockup of
the vessel bottom was set up, and recommendations on the concrete drying modes
were developed
Additional results expected on: Mechanical (including after irradiation) and thermal
physical properties of the selected concrete compounds, regarding thermal
conductivity coefficients in the concrete filler, as well as experimentally determined
temperature profiles for verification of computational methods. Mounting, filling and
drying technologies for the reactor vessel are also being further developed.
Vessel Vessel bottom mockup Concrete species
RUSSIAN FEDERATION
Computational and experimental qualification of reactor vessel
(steel-lined, thermally insulated concrete vault)
Slide 10
BREST–OD–300 SCHEDULE:
Design completed 2014
License approval 2015
Start of construction 2016
Commissioning 2020-2022
RUSSIAN FEDERATION
Slide 11
FRONT END ENGINEERING AND DESIGN (FEED)
CONTRACT FOR MYRRHA - STARTED IN OCTOBER 2013
Consortium: Areva TA (leader) – France, Ansaldo Nucleare S.p.A. – Italy,
Empresarios Agrupados – Spain, Grontmij Industries – Belgium
Content: Technical design of the infrastructure except for: Primary
System, Accelerator, Spent Fuel Building, Remote Handling
SYSTEMS CONSIDERED IN FEED INCLUDE FOR EXAMPLE:
• Secondary Cooling System, Reactor Vessel Auxiliary Cooling
System (RVACS), Cover Gas, LBE Conditioning System, etc.
• Safety, System Integration, and Plant Lay-out are also included
In 2015, activities on MYRRHA encompass:
• Design review of primary side (in the first half of 2015 by SCK•CEN)
• Investigation on introduction of double-tube HX (minimizing risk for SGTR)
• Related to this design change, Functional Specifications for all components and
Technical Specifications of non-impacted FEED systems were issued in
December 2014
• Second phase of FEED is expected to start in September 2015
EURATOM
Slide 12
STATUS OF ACTIVITIES IN EUROPE
W-DHR SG
Primary
Pump
Core
Safety
Vessel
Vessel
Inner Vessel
FAs
ELFR ALFRED
ELFR is one of the reference LFR
systems in GIF
Power: 1500 MWth (630 MWe)
Primary cycle : 400-480°C
Secondary cycle 335-450°C, 18MPa Power: 300 MWth (125 MWe)
EURATOM FP7 Project LEADER (Apr. 2010 – Sept. 2013) developed conceptual designs of
ELFR and ALFRED: ELFR as reference industrial plant & ALFRED as LFR Demonstrator
ALFRED Consortium Agreement was
signed in December 2013
EURATOM
Slide 13
FALCON (Fostering ALfred CONstruction) Consortium has been established on December
18th, 2013 by: Ansaldo Nucleare, ENEA, and RATEN-ICN
The aim is to constitute a network of organizations interested in the LFR technology
development and, as a closer goal, committed to ALFRED construction
Reference site for construction of ALFRED is Mioveni (Romania)
EU Organizations are invited to join FALCON through a
technical cooperation agreement (MoA)
For the MoA the interested organization can:
• Contact one of the FALCON members
• Agree on a technical activities program
• Sign the MoA with the FALCON member
All contributions are expected to be of an in-kind nature
EURATOM – ALFRED and FALCON
In December 2014,
CV-REZ joined
FALCON and is now a
full member of the
Consortium
FALCON NEW MEMBER:
MoA Status:
CRS4 (Sardinia - Italy) – MoA signed – activity started
SRS (Rome, Italy) – MoA signed
Symlog (France) – MoA signed
CIRTEN (Consortium of Universities, Italy) – MoA signed
NRG (Petten, The Netherlands) – final text of MoA agreed
IIT (Milan, Italy) – activity Agreed – under signature
KIT (Karlsruhe, Germany) – activity Agreed – under signature
GRS (TSO, Germany) – contacts ongoing
Slide 14
LEADER–BREST Cooperation Agreement
In May 2014, a Cooperation
Agreement (CooA) was signed
between Ansaldo Nucleare, as the
coordinator of the LEADER project,
and OJSC NIKIET, as the coordinator
of the BREST project
The CooA aims at the exchange of
information between the two projects
on 7 main topics:
Topic 1: Conceptual design of LFR at various power sizes and for various purposes
Topic 2: Approaches and methods for ensuring nuclear reactor safety
Topic 3: Computational and exp. studies of neutron and physical characteristics of the LFR
Topic 4: Computer and exp. study of thermal and hydraulic characteristics of elements of
the active core, steam generator and flow pattern in the reactor
Topic 5: Investigation on available materials compatible with lead coolant and possible
approaches for corrosion control/reduction
Topic 6: Long-term impact on nuclear fuel cycle highlighting advantages and
environmental effects
Topic 7: Education and training: Provide a framework to grow the skills of the young
generation of engineers and scientist on LFR technology
A number of meetings dedicated to information exchange among experts will be organized.
The first meeting is expected during spring 2015.
Slide 15
Thank you for your attention
16th LFR Prov. SSC Meeting Hefei – China December 10-12, 2014
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