Fuel Cycle Approaches for Waste Burden Minimisation in … on AFC/Japan...Many transmutation technology are proposed from many research ... Summary table Index Transmuter Inventory

Fuel Cycle Approaches

for Waste Burden Minimisation

in Japan

K. Nishihara

Japan Atomic Energy Agency

Technical Meeting on Advanced Fuel Cycles to Improve the Sustainability of Nuclear Power through the Minimisation of High Level Waste, VIC M0E05, Vienna, 17-19 October 2017

Content

1. Scenario study for Level 1, 2 and 4 in case of phase-out

in Japan

2. R&D for ADS

Reference: K. Nishihara, et. al, “Comparative Study of Plutonium and Minor Actinide Transmutation Scenario,” Proceedings of 21st International Conference & Exhibition; Nuclear Fuel Cycle for a Low-Carbon Future (GLOBAL 2015) , p.388 - 395, 2015/09

LWR Repository Pu, MA, FP

(SF)

- After the Fukushima-accident,

visions of long-term utilization

becomes unclear in Japan.

- If current LWRs will not be

replaced after their closure,

Pu remains as legacy in

spent fuel (SF) from LWR.

- This study focuses on

transmutation of Pu and

minor actinide (MA) in SF

that will be produced from the

existing LWRs in Japan.

Reference scenario

0

10

20

30

40

50

60

1960 1980 2000 2020 2040 2060G

en

era

tio

n c

ap

acity o

f LW

R (

GW

e)

A.D.

REF

2055

Phase-out

Direct disposal of LWR-SF

Phase-out from NPP 1. Scenario study:

Once-through (OT-)

Transmutation

LWR Repository MA, FP

(SF)

Transmuter = Thermal neutron reactors

L-MOX, light water reactor (LWR) with MOX

L-ROX, LWR with rock-like oxide (ROX)

G-ROX, gas reactor with ROX

Reproce

ssing

Transmu

ter

Pu

Pu, MA, FP

Pu is transmuted only once.

ROX: PuO2 in yittria stabilized zirconia (YSZ)

Chemically stable like natural rock,

and hardly soluble to acid and alkali

(Glass)

Multiple (M-)

Transmutation

LWR Repository FP

(SF)

Transmuter = Fast neutron reactors

FR, sodium-cooled fast reactor

ADS, accelerator-driven system

MSFR, molten-salt fast reactor

Transmu

ter

Pu, MA Pu, MA, FP

TRU is transmuted in many times.

Reproce

ssing

FP

Pu, MA

Only FP is

disposed of. Reprocessing

with partitioning

Purpose and

Estimated amounts

1. Scale of transmuters

Thermal output, number of plants, schedule

2. Inventory of Pu and MA

Time evolution, residual amount after closing transmuters

3. Repository

Area, potential toxicity

Many transmutation technology are proposed from many research

groups.

Each of them insists its superiority to direct disposal based on each

assumptions and analysis.

This study attempts to provide estimations for each based on

common assumptions and analysis.

Method: NMB code

(Nuclear Material Balance Code)

Amount of 26 actinides (Th-Cm, T1/2> several days) in spent fuels is calculated with an accuracy comparable to the ORIGEN code.

LWR, CANDU, gas-cooled reactor, several sodium-cooled FRs, ADS and MSFR are available. Each reactor can be coupled with appropriate fuel such as UO2, MOX, ROX, Pu-nitride (PuN), MA-nitride (MAN) and molten-salt fuel.

Fission products are estimated by dividing them into several groups (iodine, rare gas, technetium and platinum group metals, strontium, cesium and others)

Number of waste packages and repository size are determined by temperature analysis considering repository layout.

Potential radio-toxicity that is defined as dose by direct ingestion is also estimated.

Backup

Specification of

transmuters

*1 0.1% per year

Unit OT-transmuter M- transmuter

Transmuter L-MOX L-ROX G-ROX FR ADS MSFR

Thermal efficiency % 33 33 50 37.5 33 45

Load factor, 𝜀𝑜 % 85 85 90 84 58.8 95

Thermal output GWt 4.2 4.2 0.6 1.6 0.8 0.6

Electric output GWe 1.383 1.383 0.3 0.6 0.264 0.27

Batch length day 440 440 344 183 50 NA

Batch number 4 4 4 4 6 NA

Plant life year 40 40 40 60 60 60

Specific power, h MW/tHM 25.6 171 400 80.4 400 92

Burnup GWd/tHM 45 301 550 59 120 NA

Pu ratio to actinide % <12 45 100 41 ~90 ~90

Cooling period before

reprocessing, 𝑇𝑜

year NA NA NA 3 3 NA

Actinide loss in

reprocessing

% NA NA NA 0.1 0.1 0.1*1

Availability A.D. 2000 2020 2030 2050 2050 2050

Backup

Common assumptions

• LWR construction： Ohma, full-MOX BWR in

2014 is the last.

• Rokkasho reprocessing is gradually started

from 2015. Separation of MA becomes

possible from 2025.

• L-MOX fuel can be loaded to all LWRs with limit

of 30% except Ohma that is full-MOX reactor.

Backup

Result: OT-transmutation

Scale of transmuters

0

2

4

6

8

10

12

2000 2020 2040 2060 2080 2100

Ge

ne

ratio

n c

ap

acity o

f O

T-T

ran

sm

ute

rs

(GW

e)

A.D.

L-MOX (Total)

L-MOX (full loading)

L-ROX (Total)

L-ROX (full loading)

G-ROX (Total)

237GWeYr

176

84

99

245

Two full MOX/ROX

LWRs

L-MOX

- Full-MOX reactors in 2014 and

2035, which occupy 40% of total

MOX loading to LWR.

L-ROX

- Full-ROX reactors in 2030 and

2040, which occupy 50% of total

ROX loading to LWR.

G-ROX

- GCR appears from 2030 to 2080.

- Number of GCRs is rather large

(20) because of modular concept.

- Total of generation capacity looks

larger because of high thermal

efficiency (50%).

20 GCRs


Actinide inventory

0

50

100

150

200

250

300

350

400

450

500

2000 2020 2040 2060 2080 2100

TR

U in

ve

nto

ry (

t)

A.D.

REF

L-MOX

L-ROX

G-ROX

0

50

100

150

200

250

2000 2020 2040 2060 2080 2100

Fis

sile

Pu

in

ve

nto

ry (

t)

A.D.

REF

L-MOX

L-ROX

G-ROX

MA+Pu Fiss. Pu

-37

%

-24

%

-39

%

-72

%

-55

%

-79

%

Total amount in REF reaches 430t.

L-MOX is inferior to others.

Total amount in REF reaches 210t.

G-ROX performs best.

L-ROX contains small amount of 238U

due to safety issue, which produces 239Pu.

430t 210t

0

1

2

3

4

5

6

7

2000 2100 2200 2300 2400

Ge

ne

ratio

n c

ap

acity o

f tr

an

sm

ute

r (G

We)

A.D.

FR

ADS

MSFR

Result: M-transmutation

Scale of transmuters

580GWeYr,

16 units

420GWeYr

26 units

500GWeYr,

32 units

180 years

300 years

FR

- Number of units (16)

is the fewest owing to its larger

thermal output.

- Required period is the

longest (300 years) because of

low transmutation performance.

ADS

- Number of unit is the

maximum (32).

- Required period is 180 years.

MSFR

- Resemble to ADS.

- MSFRs is fewer than that of

ADSs because load factor of

MSFR is better.

←2

05

0

1. Transmuters are introduced from 2050 according to the TRU inventory in that time.

2. They are closed after 60 years of operation,

3. and replaced by fewer transmuters according to the TRU inventory in 2110,

4. and so on….

0

50

100

150

200

250

2000 2100 2200 2300 2400F

issile

Pu

in

ve

nto

ry (

t)A.D.

REF

FR

ADS

MSFR

Result: M-transmutation

Actinide inventory Fiss. Pu

-99

%

-96

%

-99

%

TRU decreases in exponential manner after 2050.

Reduction is slower and the final residual TRU is

larger in FR scenario.

In ADS and MSFR scenario, there is still residual

TRU of 30 t of which 16 tMA exists in the glass

waste before introducing partitioning in 2025.

Other 14t is residual of transmutation and small

amount of loss in reprocessing.

Most of fissile Pu is transmuted.

Small amount is in the residual fuel

from transmutor.

210t

1~2t

8t

0

50

100

150

200

250

300

350

400

450

500

2000 2100 2200 2300 2400

TR

U in

ve

nto

ry (

t)

A.D.

REF

FR

ADS

MSFR

Glass

MA+Pu

-93

%

-86

%

-92

%

430t

30t

60t

16t in Glass

Repository size

0

1

2

3

4

5

RE

F

L-M

OX

L-R

OX

G-R

OX

FR

AD

S

MS

FR

Tota

l re

pository

are

a (

km

2)

ROX

SF

ROX

SF

MOX

SF

MOX

SF

UO2

SF

Glass

PT waste

OT-Transmutation M-Transmutation

-80

%

OT-transmutation

- Area is same to REF.

- Decay heat from MOX- and ROX-

spent fuel avoids dens disposal,

although glass waste occupies about

half of original UO2 spent fuel.

M-transmutation

- Area is reduced by 80%.

- Glass before introducing partitioning

technology (in 2025) occupies 0.5km2.

Potential radio-toxicity

(Dose assuming direct ingestion)

1E+8

1E+9

1E+10

1E+11

1E+12

1E+13

1E

+2

1E

+3

1E

+4

1E

+5

1E

+6

Inta

ke d

ose (

Sv)

Time (year)

ADS

MSFRFR

natU

L-MOXL-ROXG-ROX

OT-transmutation

- Toxicity is slightly

decreases, but necessary

period to be less than

natural U does not

change.

M-transmutation

- Time to decay less than

natural U becomes

around 10,000 years that

is shorter by an order

than REF. (ADS, MSFR)

- Impact by FR is modest

because residual amount

of TRU is large.

- MA in glass before

partitioning deteriorates

the impact. (typical

benefit by P&T is shorter

than 1000 years)

Typical range

of PT scenario

Summary table

Index

Transmuter Inventory (ton) Repository

GWe GWt # Period

(yr) *2

Pu+

MA Pu

Fiss

.Pu Area (km2)*3

Tox. (yr) *4

REF - - - - 431 348 210 4.0 1E+5

OT-

Tra.

L-MOX 5.9 18.0 2*1 60 329 220 94 4.0 1E+5

L-ROX 4.4 13.3 2*1 50 270 164 59 3.9 8E+4

G-ROX 7.2 14.4 20 40 252 137 34 4.3 8E+4

M-

Tra.

FR 9.6 24 16 300 57 21 8 0.82 3E+4

ADS 8.4 25.6 32 180 28 9 1 0.82 1E+4

MSFR 7.0 15.6 26 180 34 13 2 0.83 1E+4

*1 Number of full-MOX or –ROX LWRs excluding partially loaded LWRs

*2 necessary period for transmutation

*3 Area for emplacing wastes excluding area for aisle and utility

*4 Time to decay less than toxicity of corresponding natural uranium

Level 1

Level 2

Level 4

Conclusion of scenario study

• As a whole, advanced technology performs better,

but necessary transmuters is more.

• Development of efficient Pu burner with multiple

transmutation is needed in case of phase-out.

• Advantage and disadvantage of the technologies

depends on many aspect: economics, technological

readiness, inherent safety, impact on repository,

non-proliferation and so on. The present

comparison using common assumptions and

methodology provides base of the future discussion

in these aspect.

18

2. R&D for ADS:

Technical Issues of ADS in JAEA

• R&D of SC-LINAC • Reliability assessment

based on experience in J-PARC accelerator

•Design study on reactor vessel, beam duct, quake-proof structure, etc.

•Fabrication, irradiation and reprocessing tests

•Experiments in critical assemblies and analyses

Reactor component

MA fuel

•Pb-Bi technology (chemical control, pumps, sensors…) in cold loop

•Material verification (irradiation corrosion, erosion, strain)

•Spallation Target, Material

•Operation of Pb-Bi system

•Reactor Physics •Control of Subcritical System

Accelerator

MA Transmutation Fuel Cycle

19

MA nitride fuel

Spent nitride fuel

Transmutation

fuel cycle

Commercial power reactor fuel cycle

（light water reactor or fast reactor）

Partitioning process

by solvent extraction

Transmutation by ADS

(MA transmutation

ratio: ~10%/year)

Fuel fabrication

for ADS

Pyrochemical

reprocessing

MA

Characteristics of nitride fuel

○ High heat capacity and high melting point

○ Large flexibility of fuel composition ● Unstable in the air

⇨ Treatment in an inert atmosphere

Characteristics of pyrochemical reprocessing

○ Using molten salt and liquid metal ⇨ High radiation resistance and high heat resistance ● Treatment in an inert atmosphere

Development of extractant and separation process ○ High selectivity and decontamination ○ Continuous operation ● Instability of extractant

R&D Activities in JAEA :

MA Separation Process from HLW

20

- Very high Distribution ratio (D) for An(III) and RE at high HNO3 concentration

- Very low D of An(III) at low HNO3 concentration

- Soluble in n-dodecane, CHON extractant

- High extraction capacity

Np, Pu

FP, U(VI)

Am, Cm, RE

High-level liquid waste（HNO3:1-3M）

An(IV) Strip

An(III) Strip

An(III),An(IV) Extraction

Conceptual flow of the process

Counter-current continuous extraction tests

using mixer-settler units were performed

with simulated HLLW containing 12 fission

product elements and Am and Np tracers.

C

C

O

O

O

C

CH2

NN

C12H25

H25C12

H2

C12H25

H25C12 C

C

O

O

O

C

CH2

NN

C12H25

H25C12

H2

C12H25

H25C12

TDdDGA (Dodecyl-DGA)

N,N,N’N’-tetradodecyldiglycolamide

Am extraction : > 99.99 %

Np extraction : 91 %

Sr, Pd, Zr were separated from An.

Optimization of the process for genuine HLLW hot test are conducted.


MA Nitride Fuel Fabrication & Performance

21

Almost completed

- Preparation of homogenous

mixture of inert matrix (ZrN)

and TRU nitrides

- Stability evaluation of ZrN-

TRU nitride solid solution

against hydrolysis &

evaporation

-Physical property

measurements

To be studied/developed

- Fabrication process

considering high heat

density of the fuel

- Enrichment of 15N

- Fuel behavior analyses code

- Irradiation tests

Nitridation-distillation combined method

窒化法

Fuel rod

PartitioningProcess

Pyrochemicaltreatment

Nitride fuel pellet

Inert matrix nitride

fuel powder material

MA nitride TRU nitride Inert matrix(ZrN)

MA containing

aqueous solution

Recovered TRU

(TRU-Cd alloy)

via oxideCarbothermic

reduction

Fuel assembly

HLW from commercial power reactor cycle

(Zr,Pu)N pellet sample

for irradiation test at

JMTR

(Zr,Pu,Am)N

disk sample

3mmf、95mg

Powder metallurgy

AnCdx+1/2N2=AnN+xCd↑

AnO2+2C+1/2N2=AnN+2CO↑


Pyrochemical Process of Spent Nitride Fuel

22

Spent nitride fuel assembly

TRU nitride powder

Reductive extraction

into Cd

Chemical dissolution into

molten salt

Zeolite treatment

recovered TRU

(TRU-Cd)

Anode residues

residues waste

Chopping of fuel rods

Renitridation

Electro-refining

Disassembling

TRU-Cd

TRU

Reductive extraction

salt+TRU+FP

salt+TRU

Anode

residues

Oxidation agent CdCl2 etc.

Molten salt

Residues

Chemical dissolution

15N2 gas

TRU、RE、 AL、ALE

LiCl-KCl（500℃）

15N2 gas

Pu, MA

Liquid Cd cathode anode: spent fuel

Molten salt electro-refining

～700℃

TRU-Cd alloy

15N2 gas

Distillation of Cd

Nitridation-distillation combined method

Renitridation

～550℃

salt+FP

salt

Almost completed

- Electrorefining & renitridation process

technologies were proved to be

feasible by experiments using Pu &

Am.

- Preliminary design of the head-end

and the waste treatment processes

(including common issues to metal

fuel)

To be studied

- Engineering scale unit tests using

surrogate materials

Cd cathode U and Pu Recovered (U,Pu)Npowder Nitride pelletU-Pu-Cd alloy in

Y2O3 crucible

(U,Pu)N powder (U,Pu)N pellet

TRL

K. Ikeda et al. / Progress in Nuclear Energy 74 (2014) 242-263

TRL of ADS reactor and Double Strata fuel cycle in Japan

• The nuclear technology for the ADS

is assessed to be at TRL 2 to 3

(Proof of critical function).

• The next step is development of

individual technology in engineering

scale.

backup

Someday, Phase out

will come …

0

20,000

40,000

60,000

80,000

100,000

2000 2100 2200 2300 2400 2500 2600 2700 2800

TRU

in t

he

wo

rld

(t)

A.D.

Open cycle

Transition

to FR

NOT YET closed. Huge inventory in core….

Sustainable

FR Phase-out

OK, it’s

closed!

Backup


Transmutation ratio

FF: fresh fuel, SF: spent fuel

0%

20%

40%

60%

80%

100%

FF SF FF SF BOC EOC

L-MOX L-ROX G-ROX

Co

mp

ositio

n o

f P

u a

nd

MA

Pu (Fissile) Pu (Fertile) MA

-85% -77%

-50%

-27%

-54% -56%

Pu+MA

- L-MOX is inferior

Pu fissile

- G-ROX performs best.

- L-ROX contains small

amount of 238U due to safety

issue, which produces 239Pu.

Backup


Request for High Reliable Accelerator

27

LANSCE

＋KEKB

J-PARC

Number of beam trips per year (7,200 hours)

We are comparing the trip rate estimated

from data of existing accelerators and the

maximum acceptable trips to keep the

integrity of the ADS components.

Short beam trip (<10s) can meet the

cliteria.

Longer beam trip should be decreased by:

Reducing the frequency and

Reducing the beam trip duration

Beam trip duration

0-10s 10s – 5min. >5min.

Acceptable trip rate

Estimation from experiences

Be

am

trip

rate

(tim

es/y

ear)

28


R&D Activities for Superconducting LINAC

Photograph of J-PARC LINAC

100

150

200

250

300

350

400

450

500

80

85

90

95

100

27 28 29 30 32 33 34 Total

Statics [ Run #27 (Nov. 2009) -- #34 (Jun. 2010)]

Run Hours

Availability (%)R

un H

ours

Ava

ilability (%

)

Run #

Nov. Dec. Jan. Feb. Apr. May. Jun. 2010

100

150

200

250

300

350

400

450

500

80

85

90

95

100

27 28 29 30 32 33 34 Total

Statics [ Run #27 (Nov. 2009) -- #34 (Jun. 2010)]

Run Hours

Availability (%)R

un H

ours

Ava

ilability (%

)

Run #

Nov. Dec. Jan. Feb. Apr. May. Jun. 2010

Prototype of cryomodule , which was designed to

accept 927MHz RF wave, was made and tested.

Two cavity excitation was successfully performed at the

design field of 10MV/m, repetition rate of 25Hz and

pulse length of 1ms.

Information on J-PARC LINAC (181MeV at present,

400MeV in the future) will be included for the

accelerator design study.

The LINAC had been operated stably for injection to the

following 3 GeV synchrotron since October, 2007.

Superconducting cavity

Cryomodule

29


Spallation Target and Beam Window

Beam window will be used in following

severe conditions:

external pressure by LBE

heat generation by the proton beam

creep deformation at high temperature

corrosion in LBE

irradiation damage by neutrons and

protons

Design condition

Proton beam：1.5 GeV-20 mA（30 MW）

LBE velocity : < 2m/s

Maximum beam window temperature :

< 500ºC

R&D issues : Material

corrosion in LBE, thermal-

hydraulic of LBE, material

irradiation effect

(2700)

1890

3390

1000

6450

1030

1390

9050

5250

1800

2000

ビーム導入管

ビーム保護管

内　筒

13700

(285)

4610φ

(2700)

1890

3390

1000

6450

1030

1390

9050

5250

1800

2000

ビーム導入管

ビーム保護管

内　筒

13700

(285)

4610φ

Core barrelCore region

Reflector

Support structure

Coolant flow

Beam windowBeam ductProton beam

Target region

Beam duct support

Partition wall

Flow control nozzle

(2700)

1890

3390

1000

6450

1030

1390

9050

5250

1800

2000

ビーム導入管

ビーム保護管

内　筒

13700

(285)

4610φ

(2700)

1890

3390

1000

6450

1030

1390

9050

5250

1800

2000

ビーム導入管

ビーム保護管

内　筒

13700

(285)

4610φ

Core barrelCore region

Reflector

Support structure

Coolant flow

Beam windowBeam ductProton beam

Target region

Beam duct support

Partition wall

Flow control nozzle

30MW proton beam with 1.5GeV causes

heat deposition of 15.7MW.

30


Thermal Hydraulics Study of LBE (1/2)

Oxygen Sensor Calibration Device

– To prevent corrosion by flowing LBE, oxygen potential in

LBE should be controlled in appropriate potential range.

– Development of oxygen potential sensor and loop tests

for oxygen potential control mechanism are underway.

OLLOCHI (Oxygen-controlled Lbe LOop for Corrosion

tests in HIgh temperature)

– Material corrosion database for various temperature, oxygen

potential, LBE flow rate will be collected

– The loop will be operated from next April

IMMORTAL (Integrated Multi-functional MOckup for

TEF-T Real-scale TArget Loop)

– Purpose of experiments is demonstration of safe operation

of LBE loop by reflecting operation condition of J-PARC LBE

Spallation target.

31


Thermal Hydraulics Study of LBE (2/2)

JLBL-4 instantaneous

3D flow structure

instantaneous

3D flow structure

Measurement result by UDM at a centerline

and inclined direction

Measurement result by advanced measurement

system “Vector-UVP”

The flow velocity measurement by UDM (Ultrasonic Doppler Method) is being

developed by using JAEA LBE loop-2 (JLBL-2).

– The distribution of the flow velocity was measured at 150 ºC by use of the UDM,

which is useful for visualization of the liquid LBE flow.

An advanced measurement system “Vector-UVP” was developed and

successfully applied to the actual LBE flow in JLBL-4 for two-dimensional

velocity vector measurement.

JLBL-2

32


Mock-up Experiment for Beam Window

JLBL-3

Mock-up of beam window

100L/min

10

100

1000

1000 10000 100000

Peclet Number: Pe

Avera

ge N

usse

lt N

um

ber:

Nu

●,■,▲: Inlet temperature: 330°C



100L/min

200L/min

300L/min

400L/min

500L/min

Heater: 6kW

Flow rate: 100-500 L/min.

Inlet temp.：330-430ºC

To investigate the prediction accuracy of heat transfer

coefficients at the beam window of hemispherical

shape, JAEA built a thermal-hydraulic loop, JLBL-3

(100-500L/min, 330-430ºC).

Flow induced temperature instability occurs at

around LBE impinging area.


Material Irradiation in LBE : MEGAPIE

33

Beam window

after experiment

Cutting plan of beam window

and test peace for PIE

MEGAPIE International Experiment

MEGAPIE (MEGAwatt Pilot Experiment) is an

international collaborating experiment to

demonstrate the feasibility of a high-power liquid

metal target carried out at the spallation neutron

source SINQ at PSI.

The target, filled with 920 kg liquid LBE, worked

successfully from August to December 2006 with

maximum power of 1.35 mA x 580 MeV.

PIE of irradiated beam window have been

performed in various institutes.

Target

Upper shield

Heat exchanger

Main EMP pump

Bypass EMP pump

Main guide tube

Bypass flow guideDouble wall container cooled by D2O

Beam window

LBE flow


Allowable Maximum k-eff for ADS

34

Accidental insertion reactivity %k/k

Beam tube filled with Pb-Bi1) 0.32

Unusual temperature rise of

coolant Pb-Bi2)

0.69

Uncertainty for measurement3) 0.50

Uncertainty for calculation 1.00

Total 2.51

1) Pb-Bi entry to the vacuum beam

tube by failure of the beam

window

2) Coolant temperature rise of

1000K in only fuel region by coast

down

3) Accuracy of βeff influences directly

Allowable maximum k-eff was set at 0.97

High k-eff value implies low proton beam current and small power peaking, but

risk of approaching criticality under accidental conditions will increase.

The subcriticality must be set to adequate level considering accidental

insertion of reactivity and uncertainties for calculation and measured

reactivity.

K. Tsujimoto, T.Sasa, K.Nishihara, et. al, “Neutronics

Design for Lead-Mismuth Cooled Accelerator-Driven

System for Transmutation of Minor Actinide”, J. Nucl.

Sci. Tech., 41, 21-26 (2004).

35


Neutronics Design of ADS

0.93

0.94

0.95

0.96

0.97

0.98

0.99

1

1.01

BOC EOC

k-e

ff, E

ND

F

JAEA EB6

CIEMAT EB6

KIT EB7

EB6

EB7

0.93

0.94

0.95

0.96

0.97

0.98

0.99

1

1.01

BOC EOC

k-e

ff, JE

FF

JAEAJEFF3.0

CIEMATJEFF3.0

CIEMATJEFF3.1

KITJEFF3.1

JF3.1

JF3.0

0.93

0.94

0.95

0.96

0.97

0.98

0.99

1

1.01

BOC EOC

k-e

ff, JE

ND

L

JAEA J3.3

JAEA J3.2

JAEA J4.0

J4.0

J3.2, J3.3

JAEA proposed the comparison of reactor physics parameters calculated by

different nuclear data as a benchmark exercise of IAEA-CRP.

About 2% discrepancies in k-eff were found among the different nuclear

data. (k-eff disperses from 0.98 to 1.0 at BOC and 0.93 to 0.96 at EOC.)

JENDL ENDF JEFF

Calculated results for IAEC-CRP benchmark proposed by JAEA (Burnup calculation for the first

burnup cycle of 600 EFPD with 800MWth ADS)


R&D for Reactor Physics of ADS

36

Ion Source

FFAG Acc.

Target Proton Beam

KUCA

Subcritical

Reactor

Magnets

FFAG Accelerator

KUCA Core

Collaboration work for reactor physics issues with

Kyoto University using KUCA (Kyoto University

Critical assembly).

R&D Activities in JAEA (Future Plan) :

Transmutation Experimental Facility (TEF)

37

Phase-I construction of J-PARC

was completed.

Phase-I facilities were in service

until March 2011. Although there

were significant damages by the

earthquake, the operation was

restarted in Dec. 2011.

The Transmutation Experimental

Facility (TEF) is the main project in

Phase-2 of J-PARC, however, it is

still waiting for the approval of the

Government.

TEF consists of

Transmutation Physics

Experimental Facility (TEF-P)

ADS Target Test Facility (TEF-T)

Jan. 28, 2008

50 GeV

Synchrotron

LINAC

Neutrino

Site for Transmutation

Experimental Facility

Material and Life

Science Facility

Hadron Facility

3 GeV Ring

Control Bldg.

Jan. 28, 2008

50 GeV

Synchrotron

LINAC

Neutrino





Material and Life

Science Facility

Hadron Facility

3 GeV Ring

Control Bldg.


Image View of TEF

38

ADS Target Test Facility：TEF-T

Critical Assembly

Pb-Bi Target

Transmutation Physics

Experimental Facility: TEF-P

Purpose: To investigate physics properties of

subcritical reactor with low power, and to

accumulate operation experiences of ADS.

Licensing: Nuclear reactor: (Critical assembly)

Proton beam: 400MeV-10W

Thermal power: <500W

Purpose: To research and develop a spallation

target and related materials with high-

power proton beam.

Licensing: Particle accelerator

Proton beam: 400MeV-250kW

Target: Lead-Bismuth Eutectic (LBE, Pb-Bi)

Proton Beam

Multi-purpose

Irradiation Area


ADS Target Test Facility (TEF-T)

39

Candidate concept for LBE target in TEF-T

Test device for flow visualization by PIV

method

(Full-scale transparent acrylic model of target

vessel

ターゲット実効寸法 : 15cmf X 60 cmL

照射試験片

試験片ホルダ（遠隔操作による複数年照射を考慮）

Experiments for irradiation damage of material by protons and neutrons

Material irradiation test for material for beam window of ADS, structure material

for FBR, and material for fusion reactor

Development of database for engineering feasibility of ADS by experiments in

various condition (ex. temperature and velocity of flowing LBE)


Transmutation Physics Experimental Facility (TEF-P)

40

TEF-P is designed to take over the experiences

and functions of FCA to minimize the cost and

risk for newly developed equipment.

Low power critical facility for reactor physics and

nuclear data of transmutation systems including

ADS and FBR.

By replacing central partial matrix tubes with pin-

type assembly, MA fuel can be used with cooling

and remote handling.

Mock-up equipment for fuel loading for TEF-P

Coolant simulator (Pb, Na)、

Plate fuel

Fuel drawer

MA pin fuel

SUS lattice

5.5cm

5.5cm

5.5cm

Coolant simulator (Pb, Na, etc.)Stainless steel matrix

Plate-type fuel

Fuel drawer

Pin-type fuel

5.5cm

Proton beam

Beam duct

Spallation target

Experiments in “critical mode”

ADS experiments in “subctitical with proton beam”


FCA (Fast Critical Assembly) at JAEA

41

Reactor material plates:

Fuels U metal Enriched U (HEU, LEU)

Natural U

Depleted U (block)

U dioxide Depleted UO2

Pu metal (Canned in SS container)

Others Na, Al2O3, Graphite, Polystylene, SS, B4C, etc.


MA Transmutation Fuel

42

MA is contained as a principal component in the fuel

Uranium-free fuel to avoid TRU formation

Diluent material to adjust the power density is contained in place of U

⇒ U-free Nitride fuel is considered as the first candidate

Hard neutron spectrum

Good thermal properties (high Tm, high thermal conductivity )

Large mutual solubility among TRU (Np,Pu,Am,Cm)

Feasibility of processing

• Highly enriched 15N is used to avoid 14C formation.

• Inert gas atmosphere is necessary for handling

• Maturity levels of fabrication & irradiation are low

Typical composition of the fuel:

Zr0.70Pu0.09MA0.21N (solid solution)

ZrN: diluent material

Pu : to mitigate the burn-up reactivity swing

Fuel type TRL

Metals 4-5

Oxide

homogeneous 4-5

Oxide

heterogeneous 4

Nitrides 3

Dispersion fuels 3-4

Maturity level for MA containing

fuels judged by OECD/NEA(2014)

Documents

Fuel Cycle Approaches for Waste Burden Minimisation in … on AFC/Japan...Many transmutation technology are proposed from many research ... Summary table Index Transmuter Inventory