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Objectives of the HCCR‐TBS Testing Program in ITER

Seungyon CHOon behalf of KO TBM Team

3rd IAEA DEMO Program WorkshopMay 11‐14, 2015USTC, Hefei, China

13rd IAEA DEMO Workshop, USTC, China, May 11‐14, 2015

Outline

• Domestic Fusion Energy Development Plan

• DEMO Breeding Blanket Program

• HCCR‐TBS Testing Objectives and Test Plan in ITER

• Design and R&D Status for HCCR‐TBS Testing in ITER

• Summary


• Nuclear Fusion Technology Development Roadmap– Breeding blanket technology will be tested through TBM Program in ITER.

– Fusion materials will be tested at IFMIF/FNSF.

– Results will be input to the DEMO construction.

Fusion MaterialsIFMIF/FNSF

Blanket TechnologyTBM

Domestic Fusion Energy Development Plan


Domestic Fusion Energy Development Plan

• Role of TBM Program – To develop breeding blanket technology for DEMO and Fusion Power Plant.

– To reduce the gap from ITER to DEMO (tritium self‐sufficiency, reactor thermal efficiency, etc.)


DEMO Breeding Blanket Program

• KO DEMO Configuration

Main Parameters Specification

Major Radius 6.8 m

Minor Radius 2.1 m

Magnetic Field (Bo) 7.4 T

Divertor Type Double Null

Plasma Current > 12 MA

Fusion Power 2200~3000 MW

Net Elec. Power 400~700 MWe

Tritium Breeding Ratio > 1.05

K. Kim et al., A Preliminary Conceptual Design Study for Korean Fusion DEMO Reactor,Fusion Eng. Des. 88 (2013) 488‐491.


DEMO Breeding Blanket Program

• Candidates for the Korean DEMO Breeding Blanket Concept– Helium‐cooled ceramic reflector (HCCR) blanket

– Water‐cooled solid breeder (WCSB) blanket

• HCCR breeding blanket concept was adopted to be tested in ITER.

Parameters HCCR DEMO HCCR‐TBSFW Heat Flux 0.5 MW/m2 0.3 MW/m2

Neutron Wall Loading 2 MW/m2 0.78 MW/m2

Structural Material ARAA ARAA (KO‐RAFM)Breeder Li4SiO4/Li2TiO3 Li4SiO4

Neutron Multiplier Be (Beryllide) BeReflector Graphite Graphite

Primary Coolant Helium HeliumCoolant Inlet/outlet Temperature 300/500 oC 300/500 oC

Coolant Pressure 8 MPa 8 MPaPurge Gas He with 0.1% H2 He with 0.1% H2

Enrichment (6Li) 90% 40%


DEMO Breeding Blanket Program• Design technology and design tool verification

– Establishment of design process including analyses (TBR > 1.05)

– Development and verification of design tools (e.g. neutronics, safety, cooling system, tritium, etc.)

• Material development and DB establishment – Structural material development and DB up to high dpa, C&S, licensing

– Breeder material mass production/handling/re‐cycling/enrichment, etc.

– Multiplier/reflector material mass production/handling

– Blanket life‐time extension

– Fabrication and joining technology

• Cooling technology– Reliable components (high capacity coolant circulator, compact and high efficient heat

exchanger/recuperator, valves, etc..)

– Coolant purification

• Tritium cycle technology – Tritium extraction, storage and accountancy

– Tritium handling and permeation protection

• Measurement and I&C • Maintenance, remote handling


HCCR‐TBS Testing Objectives and Test Plan

• HCCR‐TBS will demonstrate the performance of HCCR DEMO blanket; – Breeding capability that would lead to tritium self‐sufficiency in a reactor – Extraction of high‐grade heat suitable for electricity generation

• The Overall Objectives of HCCR‐TBS Testing – Validation of nuclear response predictions on tritium production, nuclear heating,

– Demonstration of tritium extraction process and on‐line tritium measurement,

– Demonstration of heat removal with helium coolant within the allowable temperature, stress and deformation limits,

– Demonstration of structural integrity under fusion environment,

– Demonstration of thermo‐mechanics for pebble beds under fusion environment,

– Validation of database, models and tools for design of breeding blanket system,

– Demonstration of integral performance and long‐term operation for breeding blanket system.

NAS (Neutron Activation System), MFC (Micro‐fission Chamber)

TES (Tritium Extraction System), TAS (Tritium Analysis System)

HCS (Helium Cooling System)

PIE (Post Irradiation Examination)

Thermocouple, strain gauges, etc.



Port No. (Port Master) TBM Concept 1 TBM Concept 2

16 (PM : EU) HCLL (TL : EU) HCPB (TL : EU)

18 (PM : JA) WCCB (TL : JA) HCCR (TL : KO)

2 (PM : CN) HCCB (TL : CN) LLCB (TL : IN)

(10th ITER Council Meeting 2012, 6)

–2014.2.13 : CN HCCB TBS TBMA Signed

–2014.5.23 : KO HCCR‐TBS TBMA Signed

–2014.9.8 : EU HCPB & HCLL TBMA Signed

–2014.11.27 : JA WCCB‐TBS TBMA Signed

–2015.3.12 : IN LLCB‐TBS TBMA Signed

• HCCR TBM is located in Port #18 of ITER



• Concept of KO HCCR TBM

• Four‐sub‐module Concept – Manufacturability, PIE & Transportation of Irradiated TBM– EM Force Reduction, Internal Over‐Pressure Increase

• Graphite as Neutron Reflector– Reduce the Amount of Be Multiplier – Decrease of Cost – Comparable Nuclear Performance

Sub‐module

Parameter Values

FW heat fluxAverage 0.3 MW/m2

Peak 0.5 MW/m2

Neutron wall load 0.78 MW/m2

Thermal Power 0.98 MW*

Structural material KO‐RAFM (ARAA) (< 550 oC)

Breeder Li2TiO3 pebble bed (< 920 oC)

Multiplier Be pebble bed (< 650 oC)

Reflector Graphite (< 1200 oC)

Size 1670(P) x 462(T) x 605(R) (mm)

Weight (TBM/FMS/shield)

5.552 tons(1.452 / 1.209 / 4.1)

Coolant

8 MPa He1.14 kg/s (Nominal)FW (300 oC/390 oC ) Breeding Zone(390 oC/500 oC )

Purge He with 0.1 % H2



• Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS)– 4 Sub‐module concept (TBM), TBM‐shield and Ancillary Systems

Helium Cooling System (HCS)Coolant Purification System (CPS)

Tritium Extraction System (TES)

TBM Port Cell(Pipe Forest/ bio-shield / AEU)

TBM-set(TBM body / TBM shield )

TBM ancillary system(HCS, CPS, TES, etc. in other building)

Vacuum Vessel Port Inter-space & Port Cell (Tokamak L1)

Part of TES

Part of HCSRoom 14-L4-20 in TCWS VA

Room 14-L2-24 in Tritium Building

Port#18 JA‐KO Interface (Common Items)



• Schedule of HCCR‐TBS Development– Overall schedule will be adjusted when current First Plasma date is changed.

ITER Operations

‘12 ‘14 ‘16 ‘18 ‘20 ‘40‘11 ‘13 ‘15 ‘17 ‘19 ‘21

ITER Construction

TBM Major Milestone during Construction

‘22 ‘23 ‘24

First Plasma(‘20.11.30)

‘18. 10, Contract Award

‘21.07, Acceptance tests in KO’21.03, Fabrication

‘22.01, TBS acceptance tests in ITER

‘21.10, Delivery

’17.06, Testing and Qualification of large size TBM MU

’16.12, Irradiation DB for material/joints’15.12, DB for material/joints

’14.05, TBMA ’18.05, Amendment HCCR-TBS TBMA signed

’16.02, Safety Report V2’20.11, Safety Report V3

’14.05, PrSR

’18.01, FD Approval’17.01, PD Approval

’15.05, CD Approval



• 4 TBMs are planned to be tested in ITER during the first 10 years of ITER operational phases in order to simulate DEMO HCCR breeding blanket concept.



• Main objectives of 4 TBMs in different ITER Operation Phases

ITER Operation Phase

TBM Type Main Objectives

H‐He EM‐TBM(Electro‐Magnetic)

‐ To investigate the structural integrity under strong electromagnetic loads and high surface heat flux without neutron irradiation

D/DT(short pulse,low duty)

TN‐TBM(Thermo‐Neutronic)

‐ To validate neutron response data , thermal behaviours‐ To validate tritium breeding performance and nuclear design tools

D‐T (long pulse)

NT/TM‐TBM(Neutronic‐Tritium/Thermo‐Mechanic)

‐ To validate the tritium production and structural behavior at relevant temperatures‐ To investigate characterization of thermo‐mechanical behaviors of the pebble beds and tritium extraction by purge gas

D‐T(long pulse, high duty)

INT‐TBM‐ To evaluate the blanket capability on the tritium breeding, the extraction of high‐grade heat, the integrity under thermal, mechanical and electromagnetic loads



• After the testing, all the HCCR TBMs will be shipped back to Korea for post‐R&D and Post Irradiation Examination (PIE)



• Transportation of Irradiated TBMs– Handling/Treatment at ITER HC and put into the transfer cask– In‐land transportation in France : no legal and technical issue are expected (Common for all IMs)

– Shipping from France to Korea (Common issues for IMs except EU)– In‐land transportation in Korea – PIE in Hot Cell in Korea • Material Characteristics• Materials Interactions


Design and R&D Status for HCCR‐TBS Testing in ITER

• Design Code Development and V&V– KO has been developing domestic code (GAMMA‐FR) for KO HCCR‐TBS safety analysis– Utilize GAMMA‐FR to DEMO

‐ Developed based on VHTR(high T/P gas)‐ Newly two‐phase flow (water)was added

‐ Domestic user groups activated‐ V&V through KO‐US TBM

collaboration‐ New model for chemical reactions

being added



• Neutronics design and calculation – Global model based on Clite v‐1, based on 3‐D CAD model

– MCNP5 ver.1.4+FENDL‐3.0– FISPACT/EAF‐2010

Neutron transport and SDDR distribution

Mesh generatorNeutron trans port model



• KoHLT‐EB (High Heat Flux Test Facility)– Connected with He Cooling Loop and Water Loop

EB-gun

Power Supply

Test Chamber

He Loop

Water Loop (under the He Loop)



• KoHLT‐EB (High Heat Flux Test Facility)– Connected with He Cooling Loop and Water Loop

EB-gun

Power Supply

Test Chamber

He Loop

Water Loop (under the He Loop)

Facility Electron Beam HHFT Facility

Major Target PFCs development

Heat Flux 5 MW/m2 (300×200 mm2)

Heat Source Electron Beam (MAX 60 keV)

Power Supply 300 kW (DC 60 kV)

Test Chamber Cylindrical chamber (Φ1.4m×D2.5m)

Filling Gas Vacuum condition

Vacuum System1,900 lps TMP

(base pressure < 10-6 mbar)

Coolant Supply System

Water: ~ 120 oC, 3 MPa,He gas: ~ 500 oC, 9 MPa



• Helium Cooling System Verification– 1:1 Scale of components verification Loop

Facility HeSS (Helium Supply System)

Operation (Design) conditions

300 (500) oC / 8 (10) MPa

Helium heater power 150 kW

Circulator

Inlet condition: 50 oC, 8.0 MPaFlow capacity: 1.5 kg/s of pressurized helium gas

(8 MPa)Compression ratio: 1.1

RecuperatorType: PCHE

Efficiency: 0.92 (@8 MPa, 1.5 kg/s of helium flow)Compactness: 660 m2/m3



• Tritium Extraction System Verification – Purge Gas Loop (PGLoop) Facility for the test of unit process, adsorption/desorption phase

Parameter Value Remark

Purge gas composition control‐ Q2‐ Q2O‐ Other‐ He

‐ Max 3000 vppm (nominal 1000 vppm) *‐ Max 31.6 vppm (nominal 10 vppm) **‐ 10 vppm‐ Balance

• To test various swamping ratios (He:H2=1000:1)

** To test various HT/HTO ratios

Purge gas P/T/Mass Rate 101 kPa / RT / up to 0.3g/s (nominal 0.1 g/s) Outlet : ~450C with 0.9bar (nominal)

RTMS ‐ Adsorption 14 days‐ Regeneration less than ? Days (Blower)

Extraction efficiency (>90%) CMS

‐ Adsorption 7 days for 0.1 g/s & 0.1% hydrogen swamping‐ Adsorption 2 days for 0.3 g/s & 0.3% hydrogen swamping‐ Regeneration less than ? Days (Blower/buffer vessel)



• Measurement System – Temperature, Stress/Strain, Electro/mangetic Field, etc.– Neutron Activation System – Micro‐fission Chamber – Tritium Analysis System


International Collaboration

• Japan/Korea TBM Collaboration – JCM Framework– Common Item Arrangement and Delivery Strategy : AEU (in Port Cell), PF Structure (Port Inter‐

space) – Discussion on the Interface Issues of two TBSs between JA and KO

• CN/KO TBM Collaboration – JCM Framework– Exchange of TBM related activities and beyond

• US/KO TBM Collaboration– Task Agreement under the framework of Project Annex between US‐KO governments and

MOU between NFRI/UCLA in July 2013– Validation of Safety Analysis Code (GAMMA‐FR) with MELCOR – Measurement of property for Functional Materials, – Tritium Behavior Modeling of Function Material– Experiments of Tritium Permeation for Structure Material & Tritium Release for Functional

Materials– Exchange of Li‐6 enriched LIOH materials


Summary

• KO Fusion Program is supported by National Law (FEDPL) – Fusion Energy Development Roadmap was developed – The roles of TBM program was assigned to test the concept of breeding blanket for

DEMO & Fusion Power Plant• Objectives of TBM testing is defined and testing plan is established for EM, TN, NT/TM,

and INT‐TBMs during the first 10 years of ITER operation periods.• Current design and R&D status is introduced for the preparation of HCCR‐TBS test in

ITER.


Thank you for your attention!([email protected])

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