Recycling Challenges Of Thoruim-based fuels

P.K. Wattal (wattal @ barc.gov.in)

BARC ,India

ThEC 13 Oct 27-31 2013

• Thorium availability in India

• Fabrication of thorium based fuel – Thorium metal fuel

– Utilisation of thorium fuel in India

– Fabrication needs and technologies

– KAMINI Fuel

• Reprocessing aspects of thoria fuel

• Waste management aspects

Presentation- Outline

BARC

Monazite Composition

Thorium (ThO2) 9.00%

REO 58.50%

P2O5 27.00%

Uranium (U3O8) 0.35%

Balance 5.15% Monazite Composition

Thorium (ThO2)

9.00%

REO

58.50%

P2O5 27.00%

Uranium (U3O8)

0.35%

Balance 5.15%

India – Vast Resources of beach sand minerals (Ilmenite, monazite, zircon)

Resource Quantity(te) Energy Potential

(GWe-yr)

Uranium 73,000 328 in PHWR

42,230 in FBRs

Thorium 225,000 155,500 in Breeders

India’s Nuclear Energy Resource Position

Uranium resource is just about 1% of that of the world

Thorium resource among the largest in the world

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Indian Strategy for Long-term Energy Security

Hydroelectric

Non-conventional

Coal domestic

Hydrocarbon

Nuclear (Domestic

3-stage

programme) Projected

requirement*

*Ref: “A Strategy for Growth of Electrical Energy in

India”, document 10, August 2004, DAE

No imported reactor/fuel

Deficit to be filled by fossil

fuel / LWR imports LWR (Imported)

FBR using spent

fuel from LWR

LWR import: 40 GWe Deficit 412 GWe

Required coal import:

1.6 billion tonne* in 2050

* - Assuming 4200 kcal/kg

The deficit is practically

wiped out in 2050

Year With thorium, nuclear installed capacity (600 GWe) can be sustained for very long period

U fueled

PHWRs

Pu FueledFast Breeders

Nat. U

Dep. U

Pu

Th

Th

U233 Fueled

ReactorsPu

U233

Electricity

Electricity

Electricity

Stage 1Stage 1 Stage 2Stage 2 Stage 3Stage 3

PHWR FBTR AHWR

Thorium in the centre stage

Power generation primarily by PHWR

Building fissile inventory for stage 2

Expanding power programme

Building U233 inventoryThorium utilisation for

Sustainable power programme

U233

300 GWe-Year

42000

GWe-Year

155000

GWe-Year

The Three Stages of the Indian Nuclear Programme

6

Homi Jehangir Bhabha

1909-1966

U 233 fuelled research reactor KAMINI

PFBR

Thorium fuel cycle activities in India

Fabrication

J-rods of CIRUS

ThO2 fuel for Dhruva

Thoria fuel bundles for PHWR

Thoria fuel assemblies for FBTR blanket

Irradiation

CIRUS J-rod position

Dhruva regular fuel location

PHWR initial flux flattening

FBTR blanket

Reprocessing J-rods of CIRUS at BARC & IGCAR

New facility PRTRF for PHWR Thoria bundles

Utilisation of U-233 KAMINI plate type fuel

PURNIMA-II liquid fuel (Uranyl nitrate solution)

Thorium Metal Fuel • Nuclear grade thorium metal powder

– calcio-thermic reduction of thorium oxide

Thorium metal green compacts

As rolled thorium metal sheet

Irradiated UO2 fuel ThO2+4%PuO2 fuel burn

up 18,400 MWd/t

Higher thermal

conductivity of thoria

leads to lower fission

gas release

No FG channels in

ThO2+4%PuO2

Burnup 4,400 MWd/t

Burnup15,000M Wd/t

Post Irradiation Observations on ThO2+4% PuO2 fuel

Presence of radial cracks in the fuel.

Absence of visible grain growth or columnar grain formation.

Reduction in porosity in the central portion of the fuel.

Uniform distribution of Pu in the section.

Fission Gas Bubbles in UO2 and ThO2 Fuels

Thorium Irradiation in Indian Reactors

Irradiation in PHWRs

Irradiation in research reactor

Reactor No. of bundles

Madras- I 4

Kakrapar-I 35

Kakrapar-II 35

Rajasthan- II 18

Rajasthan -III 35

Kaiga-II 35

Rajasthan-IV 35

Kaiga-I 35

Fuel GWd/t

(Th-4%Pu)O2 18.5

(Th-6.75%Pu)O2

10.2

Fuel fabrication by conventional powder compaction & sintering

BARC

Advanced Heavy Water Reactor (AHWR)

AHWR is a 300 MWe vertical pressure tube type, boiling light water cooled and heavy water moderated reactor using 233U-Th MOX and Pu-Th MOX fuel.

AHWR can be configured to accept a range of fuel types including enriched U, U-Pu MOX, Th-Pu MOX, and 233U-Th MOX in full core

AHWR Fuel assembly

Bottom Tie Plate

Top Tie Plate

Water Tube

Displacer Rod

Fuel Pin

Major design objectives

A large share of power from Thorium based fuel

Several passive features No radiological impact in public

domain

Passive shutdown system to address extreme threat scenarios.

Design life of 100 years.

Easily replaceable coolant channels.

BARC

Validate reactor physics design on various Fuel types.

Thoria, (Th-Pu) & (Th-U233) MOX fuel clusters in

phases for Physics study.

Criticality achieved in April 2008

Fuel for AHWR Critical Facility

AHWR Critical Facility

Experience with 233U in India • PURNIMA II (1984-86)

– Experiments with uranyl nitrate solution containing 233U reflected by BeO blocks.

• PURNIMA III (1990-93) – Experiments were performed with 233U-Al

Dispersion Fuel in the form of plates

– These measurements helped in finalising the core of KAMINI reactor.

• KAMINI (1996) – A 30 KW reactor based on 233U fuel in the

form of U-Al alloy

– It is the only operating reactor in the world with 233U as fuel.

BARC

Unit Operations in Thoria Reprocessing

(U, Pu)O2

Fuel Pins

(Th,233U)O2

Fuel Pins

(Th, Pu)O2

Fuel Pins

Thoria Spent Fuel

Dismantling & Segregation

Dissolution & Feed clarification

Rejects Fuel Fabrication

BARC

0.3M HAN +

0.2M N2H4 +

0.6M HNO3

Extraction Scrubbing Partitioning U- Stripping

5% TBP

Th

U, Th &Pu 3 M HNO3

Pu

0.01M HNO3

U233

Lean Org

Three Component Thoria Spent Fuel Reprocessing

Element Conc.

Pb 0.0174 mg/L

Bi 1.69E-05mg/L

Ac 8.55E-05

Th 100.00g/L

Pa 0.67735 mg/L

U 2.00 g/L

Np 0.283 mg/L

Pu 2.00 g/L

Am 0.08 g/L

Cm 0.016 g/L

Li 7.53E-06 mg/L

Be 2.01E-05 mg/L

Zn 4.84E-09 mg/L

Ga 1.45E-06 mg/L

Ge 0.16 mg/L

As 0.064 mg/L

Element Conc.

Se 9.86 mg/L

Rb 55.44 mg/L

Sr 108.97 mg/L

Y 63.03 mg/L

Zr 436.9 mg/L

Nb 6.24 E-4 mg/L

Mo 327.99 mg/L

Tc 94.33 mg/L

Ru 280.98 mg/L

Rh 38.03 mg/L

Pd 276.7 mg/L

Ag 13.35 mg/L

Cd 22.22 mg/L

In 0.132 mg/L

Sn 9.48 mg/L

Sb 1.81 mg/L

Element Conc.

Te 79.38 mg/L

Cs 317.3 mg/L

Ba 213.67 mg/L

La 157.05 mg/L

Ce 325.85 mg/L

Pr 151.70 mg/L

Nd 503.2 mg/L

Pm 3.06 mg/L

Sm 95.4 mg/L

Eu 14.1 mg/L

Gd 39.5 mg/L

Tb 0.558 mg/L

Dy 0.256 mg/L

Ho 0.053 mg/L

Er 0.0187 mg/L

Tm 1.29E-5 mg/L

AHWR Simulated Dissolver Feed Solution (~43 GWd/te, 5 y cooling, ORIGEN2 code)

Concentration (g/L)

Th-stream Pu-stream U-stream

Th : 71.29 Pu : 0.8575 U : 4.42

U : 0.001 U : 0.01 Pu : 0.004

Pu : 0.0048 Th : 0.18 Th : 0.109

FPs Decontamination Factors

U Pu Th

137Cs 8.89

×105

6.09 ×105 5.97 ×104

144Ce 7.30

×105

5.07 ×105 5.10 ×104

106Ru 1.22

×106

8.36 ×105 9.11 ×104

90Sr 1.94

×106

1.96 ×106 2.07 ×105

Performance of 3-Component Separation

Engineering Facilities

• Thoria Irradiated in Research Reactor

– Uranium Thorium Separation Facility (UTSF)

UTSF

Fuel GWd/t

(Th-4%Pu)O2 18.5

(Th-6.75%Pu)O2

10.2

U233 Recovery: 3% TBP/dodecane solvent

(232U ~ 20-30 ppm)

Th232 Recovery : 38% TBP/dodecane

Engineering Facilities

• Thoria Irradiated in Power Reactors

Power Reactor Thoria Reprocessing Facility (PRTRF)

(Under construction)

FPs: 50 Ci/L

Head End Facility

Shielded Glove Boxes of

Reconversion Lab Hot Cell

Th : 200 g/L, 233U : 2.35 g /L, 232U ~500 ppm,

Thoria Spent Fuel Reprocessing- Challenges

• Head-End – Disassembly, segregation

– Provision for processing rejected fuel pellets

– Flouride induced dissolution

Enhanced dissolution of Thoria Vs corrosion attack due to flouride.

Optimization of Al(NO3)3 addition-Impact on Waste management.

• Extraction – Three component system, Third phase formation

• Partitioning – Chemical reductants in place of U+4 to avoid contamination in U-233

• Product handling – Handling & storage of Th-232

– Presence of U-232 with U-233 and its cleanup

• Reconversion – Catalytic reduction of U233

+6 followed by oxalate pptn as an alternative to ADU for Uranium conversion

• Remote handling & Automation

Dissolution of Thoria Fuel

Difficulty in the dissolution of thoria fuel

(Fluoride as catalyst)

Boiling and reflux condition (In presence of Zircalloy) A- Fluoride: 0.005M, Al(3+) : Nil B- Flouride: 0.03M, Al(3+) : 0.1M C- Fluoride: 0.03M, Al(3+) : Nil

Elements of Concern in Thoria - HLW waste

Element oxide g/L Solubility

(wt%)

Effect on waste form

Al2O3 5.10 - Low waste loading due to high viscosity & foaming.

ThO2 1.08 7-8 Low waste loading

High Pouring Temp.

F 0.57 2 Volatility

Na2O 14-15 Wt %, Al2O3 10-12 wt % & CaO ~ 9 Wt %

Pouring Temperature: 1100-1150 C

Appearance : Homogeneous

Air Bubbles at 1000OC Liquid–Liquid Phase separation

Sodium Borosilicate Glasses

Cold Crucible - Advanced Melter

Cold Crucible Demonstration Facility

(WIP, Trombay, India)

Glass under melting

Glass under pouring

WASTES FROM THORIA REPROCESSING Trombay

Shielded cubicle for

processing of TLR waste

Management of Thorium bearing liquid Raffinates

Vitrification of thorium bearing wastes

successfully demonstrated • Concerns of Th, Al & Fluoride during vitrification addressed

18.3 cu. m of TLR waste processed

Total removal of FPs (Cs & Sr)

Fluoride Volatility during Calcination and Vitrification

RUN WITH 100 ML TEMP. 0 C

(HOLDING TIME –1Hr)

ADDED

FLUORIDE

mg

FLUORIDE

RELEASED

mg

% FLUORIDE

RELEASED

NaF (.15 M) + Al(NO3)3

(0.5 M) in 3 M HNO3 120 284.5 1.308 0.45

Simulated HLW 120 284.5 6.06 2.13

Simulated HLW +

glass formers 1000 284.5 14.95 5.25

Total volatility of fluoride during calcination and vitrification : 7.4%

Reprocessing capacity matching with FBR fuel requirement

Vitrified HLW inventory presently stored in engineered interim storage facility

Vitrification Facility, Trombay Vitrified Waste Storage

Facility,Tarapur

Reprocessing Facility,

Tarapur

PHWR

U (Natural)

UOX Fuel

Heavy Metal

α Bearing HLW

Non α Waste (L&IL)

Indian Nuclear Fuel Cycle: Today

Volume Reduction Near Surface Disposal

Vitrification Interim Storage

Repository

PUREX

Reprocessing of Carbide Fuel

Mixed Carbide Fuel for Fast breeder Test Reactor

Mixed Oxide Fuel for Prototype Fast Breeder Reactor

LWR

Fast reactors + ADS (burners)

Irradiated Rods/pins

U (Natural)

(FP with HM Losses)

Disposal

Recovery of FPs

Enriched U + MOX

MOX Alloy Fuel

Th Blanket

Th-U233 Fuel U-233 Fuelled Reactors

Pu Fuelled Fast Breeders

U-235 (1.2%)

PUREX + MA Partitioning

PUREX + MA Partitioning

Pyroprocessing

Pyroprocessing MA

FP

Pu

U-233

1st Stage

2nd Stage

3rd Stage

Pu

U- 233

Pu

MA FP

MA

FP

FP

FP

FP

Evolving Fuel Cycle

PHWR

MA Aqueous processing + Pyroprocessing

Partitioning of High Level Waste – Active Demonstration Facility, Tarapur, India

Multi step process based on solvent extraction

Separation of Actinides (An) & Lanthanides

(Ln) from U-lean HLW

Separation of Ln from An

Recovery of Cs and Sr from HLW

Novel solvents deployed:

TBP, CMPO, TEHDGA, TODGA & Calyx-crown

Actinide Partitioning

Demonstration Facility

Spent Solvent Treatment Facility

integrated to the partitioning plant

29

• Throughput to match vitrification process

• All secondary streams suitably addressed

• This facility designed as per the conventional

radiochemical plant

Blood Irradiator

Cage Assembly

Cesium Extraction in

Hot Cell, Tarapur

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21-07

22-07

23-07

24-07

25-07

26-07

27-07

28-07

29-07

30-07 31-08

29-10

31-09

31-10

07-21

31-11

07-22

31-12

07-23

31-13

07-24

31-14

07-25

31-15

07-26

31-16

07-27

31-17

07-28

31-18

07-29

08-28

08-29

09-28

09-30

30-09

27-08

28-08

29-08

09-18

10-17

11-16

12-15

15-12

16-11

17-10

18-09

14-16

13-16

13-17

12-17

12-18

11-18

11-23

11-19

10-20 11-21

11-22

10-22

10-23

09-23

11-24

10-24

09-24

11-25

11-2610-25

09-25

09-26

09-27

10-27

10-26

11-28

10-29

11-29

10-30

11-30

19-30

21-29

22-27

24-26

27-22

27-23

16-14

29-21

16-13

18-12

21-11

22-11

23-11

24-11

25-11

26-11

29-11

29-12

29-13

29-14

29-15

29-17

29-18

29-19

29-20

12-30

13-30

15-30

17-30

14-30

16-30

18-30

12-29

20-29

22-28

23-26

13-29

14-29

16-29

18-29

15-29

17-29

19-29

10-31

18-31

20-30

21-28

23-27

26-23

26-24

11-31

12-31

14-31

16-31

13-31

15-31

17-31

08-31

09-31

29-16

MA content~2%, diluted

in standard fuel in the

whole core

Moderated core target/ Core blanket

Recycling Options

Issues on fuel fabrication

• Reduced neutron flux in blanket • Possibility of longer burning time period

Homogenous

Heterogeneous

U

U ,Pu

FP FR

MA

MA content~20%,

PUREX MA Partitioning

PUREX

MA Partitioning

31

Conclusions

Sustenance of Nuclear Power is based on appropriate choice of fuel cycle which is ever evolving

Closing the fuel cycle & thorium utilisation essential for

Maximising use of existing resource (U & Th) Eventual decrease in long term radiotoxicity of nuclear waste

Technological challenges for implementation of Multi strata fuel cycle

Partitioning & Transmutation

Pyroprocessing Thorium deployment

Thank you for your kind attention