Didier Haas [email protected] ++32 491648840 NC2 Nuclear Consulting Company Thorium Conference,...

Thorium Conference, CERN

EUROPEAN EXPERIENCE WITH THORIUM FUELS

Didier Haas

[email protected]++32 491648840

NC2Nuclear Consulting Company

mailto:[email protected]


Some references

T. Lung: EURATOM report 1777 (1997)

THOR Energy Thorium Fuel Conference, Paris (2010)

IAEA No NF-T-2.4 (2012): The role of Thorium to supplement Fuel Cycles of Future Nuclear Energy Systems

GIF position paper on the use of Thorium in the Nuclear Fuel Cycle (2010)

SNETP Strategic Research and Innovation Agenda (2013) and SRA Annex on Thorium (2011)

Published EURATOM Framework Programmes results and personal communications


Content European Research on Thorium

Thorium in HTRs

Thorium oxide fuel behaviour

Molten salt reactors fueled with Thorium

Conclusion

3 main pillars + key cross-cutting issues

Sustainable Nuclear Energy Technology Platform

117members from research, industry, academia, technical safety organizations

Recent application of Weinberg Foudation (UK) andThorEA (UK) both promoting Thorium research

Launched in 2007

Produced a Research Agenda(2009, revised in 2013) and a Deployment Strategy (2010)


European R&D Roadmap on Thorium

SNETP has produced an Annex (2011) on Thorium in the Strategic Research Area. Highlights are:

LWRs: evolutionary development favoured, with use of Pu as seed (natural U savings); breeding would need new reactor technology

HWRs: high conversion ratio achievable HTR: past German HTR development programme aimed at

reaching a breeding cycle with Thorium Fast Reactors: breeding possible but with long doubling times;

improved void reactivity coefficient in sodium FR; advantage of ADS subcritical reactor (high neutron energies, Th 232 fission + captures)

MSR: breeding might be achieved over a wide range of neutron energies; long-trerm development option

Pu-burning: Thorium matrices for the purpose of incinerating Pu in LWRs

Challenges for solid fuels: reprocessing, remote fuel fabrication


Thorium Projects in Europe

1960-1980: limited experimental work on Thorium use in HTRs (DRAGON, ATR, THTR, Th-U carbide and oxide fuels) and in the Lingen BWR by SIEMENS (Th-MOX)

1990-2002: Assessment studies including the « Lung report » and the EURATOM projects « Thorium Cycle as a nuclear waste management option » and « Red Impact »

1998-2008: Thorium fuel experiments (Projects THORIUM CYCLE, OMICO, LWR-DEPUTY with irradiations in KWO-Obrigheim, HFR and BR2)

FP7 (2011-13): Performance assessment of Thorium in geological disposal (SKIN Project)

FP5-FP7 (1998-now): Thorium fuel studies and characterization for a Molten Salt Reactor (Projects MOST, ALISIA, EVOL…)


Thorium use in High Temperature Reactors

HTR thermal neutron spectrum is very well suited for Thorium breeding

Very high burnup capability in HTRs in a once-through cycle; very high stability in geological disposal of the Thorium matrix

This explains the (successful) use of Thorium in early HTR projects (DRAGON, AVR Jülich, Peach Bottom, Fort St-Vrain, THTR); fresh fuel kernels were mixed with Pu or U235 fissile material

Potential limitations are the high initial U235 content needed in the once-through strategy and the reprocessing difficulty in case of closed cycle strategy

Today, (V)HTR is one of the six GIF R&D systems; European interest in HTR exists, but difficulty in getting industry commitments


Thorium fuels in HTRs:Abstract from the « Lung » report


Thorium Oxide as a «Quasi »-Inert Matrix

ThO2 is a very stable ceramic: in-core applications, direct disposal waste management (see leaching tests results from JRC-ITU Karlsruhe)

Th-MOX (Th,PuO2) has been contemplated to incinerate separated Pu in LWRs in a fertile matrix, and also as possible « quasi »-inert matrix for MA burning in « targets »

The Th matrix produce no new Pu and is fertile as required to keep the reactivity in LWRs

In-reactor properties are equivalent (even better if one considers the thermal behaviour and the stability) to U-MOX

Thermal diffusivity measurements on unirradiated Th-MOX at JRC-ITU: higher than U-MOX

TRANSMUTATION (6.5 MEuro)Basic Studies:

MUSEHINDAS

N-TOF_ND_ADS

TRANSMUTATION (7.3 MEuro)Technological Support:

SPIRETECLA

MEGAPIE-TESTASCHLIM

PARTITIONING (5 MEuro)PYROREPPARTNEWCALIXPART

TRANSMUTATION (3.9 MEuro)Fuels:

CONFIRMTHORIUM CYCLE

FUTURE

TRANSMUTATION (6 MEuro)Preliminary Design Studiesfor an Experimental ADS:

PDS-XADS

FP5 (1998-2002) Projects on Advanced Optionsfor Partitioning and Transmutation

FP5 ADOPT Coordination Network

EUROTRANS FP6 Project

FP5: THORIUM Cycle for P&T and ADS


FP6 EUROTRANS Project and THORIUM as P&T fuel

DM2 TRADE-PLUSTRADE Experiment

DM4 DEMETRA

HLM Technologies

DM1 DESIGN

ETD Design

DM5 NUDATRA

Nuclear Data

Scientific ConsultancyCommittee

Governing Council

DM3 AFTRA

Fuels

EC

Project Co-ordination Committee

Co-ordinator

IP EUROPART

RedImpact

Related FP6 Projects:

Related National and International Programmes

DM0 Management

Project Office

DM2 TRADE-PLUSTRADE Experiment

DM4 DEMETRA

HLM Technologies

DM1 DESIGN

ETD Design

DM5 NUDATRA

Nuclear Data

Scientific ConsultancyCommittee

Governing Council

DM3 AFTRA

Fuels

ECEC

Project Co-ordination Committee

Co-ordinator

IP EUROPART

RedImpact



IP EUROPART

RedImpact


IP EUROPART

RedImpact



DM0 Management

Project Office

Associated Project onAdvanced P&T Fuels:LWR-DEPUTY Projectwith Thorium fuels

Inert Matrices fuels

(Th,Pu)O2 in-reactor experience (2000-2012)

Experiments (Th,Pu)O2 fuels were irradiated in three

reactors HFR-Petten (Na-capsule) KWO Obrigheim (non-instrumented, commercial

PWR) BR-2 Mol (instrumented & non-instrumented in

PWR loop) Post-irradiation examinations &

radiochemistry by different labs (ITU, NRG, PSI, SCK•CEN)

12


Th-MOX pellet irradiated in Obrigheim within the FP5 THORIUM CYCLE and LWR-

DEPUTY projects

Safety assessment of Plutonium Mixed Oxide Fuel irradiated up to 37.7 GWd/tonne (JNM 2013)J. Somers1,*, D. Papaioannou1, J. McGinley1, D. Sommer21. Joint Research Centre – Institute for Transuranium Elements, Postfach 2340, D76125 Karlsruhe, Germany2. EnBW Kernkraft GmbH*, Postfach 1161, 74843 Obrigheim and Böhmerwaldstraße 15, 74821 Mosbach, Germany


Thermal Behaviour

From:

C. Cozzo et al., J. Nucl. Mater. (2011), doi:10.10C. Cozzo et al., J. Nucl. Mater. (2011),


600 800 1000 1200 1400 16002.0

2.5

3.0

3.5

4.0

4.5

5.0

5.5

6.0

Heterogeneous MOX 9 wt. % Pu 7 wt. % Pu MOX Duriez

UO2 ITU

UO2 Fink

Homogeneous MOX 11.1 wt. % PuO

2

9.0 wt. % PuO2

5.6 wt. % PuO2

4.8 wt. % PuO2

Ther

mal

con

duct

ivity

, W m

-1 K

-1

Temperature, K D. Staicu, M. Barker, J. Nucl. Mater. (2013), http://dx.doi.org/10.1016/j.jnucmat.2013.08.024

C. Cozzo et al., J. Nucl. Mater. (2011), doi:10.10C. Cozzo et al.,J. Nucl. Mater. (2011),

Th-MOX Thermal Conductivityas compared to U-MOX

At 1000K TC of U-MOX: 3.0-3.5 of Th-MOX: >4.0

!! Importance of the fabrication process

BR-2 experiments on (Th,Pu)O2: Model predictions versus experiment

600

800

1000

1200

1400

1600

1800

2000

0 2000 4000 6000 8000

measurement

MACROS (post-test)Transuranus (post-test)(mod. fuel deformation)

Transuranus (blind)Copernic

power calibration from Dec 2006

Time (h)

F

ue

l C

en

tre

Te

mp

era

ture

(oC

)

OMICO Rod Gi

16

Personal communicationBy courtesy of SCK-CEN


Leaching test on Th-MOX

Source: Rondinella & Al (JRC-ITU)Paris Thorium technical meeting 2010


SKIN Euratom Project (2011-2013)

Comparison of solubility values of elements of interest

Reference case: SKB spent fuel repository

Bx, Gx: compartments of Bentonite, Granite

SCK-CEN (BE) key findings from theEuratom (Th, Pu)O2 research programs

No showstoppers identified for Thorium-based MOX (Th,Pu)O2 to its implementation as a possible LWR-fuel.

(Th,Pu)O2 has several advantages over Uranium-based MOX (U,Pu)O2 Better thermal conductivity (unirradiated data only) Improved chemical stability Indications for improved reactivity margins for full-core PWR

(Th,Pu)O2 compared to (U,Pu)O2

Next steps: Improving the fuel manufacturing technology, since the

scoping studies used non-industrial (& non-industrialisable) manufacturing routes; tests on representative fabrications needed

Larger-scale demonstration programs with lead-rod and lead-assembly irradiations are needed before licensing19

Personal communicationBy courtesy of SCK-CEN


Use of Thorium in Molten Fuel Reactors

In MSRs thorium cycle can achieve a higher conversion ratio than the uranium/plutonium cycle.

MSR avoids some of the loss of conversion efficiency that occurs due to neutron capture events in Pa-233 (Pa-233 has a relatively long half-life of 27 days). The nuclear fuel in MSR is unique in that it circulates through the entire primary circuit and spends only a fraction of its time in the active core. This reduces the time-averaged neutron flux that the Pa-233 sees and significantly reduces the proportion of Pa-233 atoms that are lost to neutron captures

MSR continually reprocesses the nuclear fuel as it re-circulates in the

primary circuit, removing fission products as they are generated. MSR therefore completely avoids the difficulties in conventional reactors with fabricating U-233 fuels (which have high gamma activities from U-232 daughters).

Since the nuclear fuel is a molten salt, there are no fuel mechanical performance issues to consider.


From MOST to EVOL

A continuous and coordinated activity (European network) since 2001

ALISIA

2001-2003Confirmation of MSR potentialIdentification of key issues (vs MSBR)

2004-2006Strenghthening of European networkFollow-up of R&D progress

2007-2008Review of liquid salts for various applicationsPreparation of European MSR roadmap

2009 Feasibility demonstration of MSFR

6 countries + Euratom


+ Russia


+ Russia

MSR R&D in Europe and elsewhere

SUMO

LICORN

MOST

EVOL7 countries + Euratom (+ Russia)

2009-2012Optimization of MSFR(remaining weakpoints)


+ Russia

from MSBR

… to MSFR


Strategic impact of EVOL

A common European Molten Salt Reactor concept for GENIV(major European contribution to the MSR GENIV

initiative)

Thorium as a nuclear fuel(closed MSR fuel cycle, sustainable energy

system)

Partitioning & Transmutation(alternative route for P&T compared to solid fuel)

Improved understanding of liquid salt properties(MSR technology, but also other industrial

processes)

MSFR reactor concept (French concept)(Molten Salt Fast Reactor)

Initial MSFR fuel composition:

X(LiF) = 77.45 mol%

X(ThF4) = 20 mol% (LiF-ThF4 eutectic)

X(UF4) = 2.55 mol%

Operating temperature: Tinlet = 620 °C

MSFR concept

MSFR pre-conceptual design, GIF Annual Report 2009: (MSR)


JRC ITU Molten Salts Database

Molten Salt Database developed at JRC (ITU) (2002-2010): 38 assessed binary systems


Conclusion

Several EC Projects on Th-MOX fuels mainly for LWRs as « Quasi »-Inert matrix to burn Pu and MAs

Thorium salts as fuel for the MSR The SRIA published in 2013 recognises

the « significant long-term potentialities and the significant challenges to make industrial implementation » of Thorium systems


Thank you for your attention !

With particular thank to Michel Hugon and Roger Garbil (EC DG RTD, Brussels), Vincenzo Rondinella, Dragos Staicu, Joe Somers (EC JRC, ITU, Karlsruhe) and Marc Verwerft (SCK-CEN) for their assistance in providing all relevant information and comments.

Didier Haas [email protected] ++32 491648840 NC2 Nuclear Consulting Company Thorium Conference,...

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