Capabilities and Capacities of Joyo towards Deployment of

Innovative Nuclear Energy Systems and Technologies

Experimental Fast Reactor Joyo

Oarai R&D Center

June , 2013

Fuels Monitoring

Facility（FMF)

Joyo

Oarai R&D Center

Tokyo Office

“Monju”

Head Office & Tokai center

Alpha Gamma

Facility（AGF)

Institute for Materials Research,

Tohoku University

Bird’s Eye View of Oarai R&D Center

Lake Natsumi

High Temperature

Test Reactor (HTTR)

Japan Materials Testing

Reactor (JMTR)

2

Experimental Fast Reactor Joyo

Overview of Heat

Transport System

Air

Blower

Main Control Room

Rotating Plug

・ To demonstrate the basic FBR technology

・ To conduct irradiation of fuel and materials

・ To validate innovative technologies for the

development of future FBR

Attain Initial Criticality ：1977

First Operation(MK-Ⅰ) ：1978

〃 (MK-Ⅱ) ：1983

〃 (MK-Ⅲ) ：2004

Role of Joyo

3

Initial Criticality

Low Power

Test

1977

MK-I Breeder Core

0 1 2 0 1 2 3 4 5 0 1 2 3 4 5 6 7

JFY K

ey I

tem

s

8 9 10 11 12 13 14 15 16 17 18 19 20

1978 1979 1980 1981 1982 1983 1984 1985 1986 1987 1988 1989

1977.4.24 MK-II Initial Criticality

1982.11.22

75MW Natural

Circulation Test

Fuel Failure Simulation Test 100MW Natural Circulation Test

Monju Fuel Pre-Conditioning Test

Calibration of FFD System

INTA-1*1 INTA-S*1

C4F

6

21 2 2 2 3 2 4 2 5 2 6 2 7 2 8 2 9 3 0 3 1 3 2 3 3 3 4 3 5

1990 1991 1992 1993 1994 1995 1996 1997 1998 1999 2000 2001 2002

Function Test : Reactor Physics Test

Flow Distribution

Measurement etc.

B5D-2*2

B5D-1*2

INTA-2*1 INTA-S

C4F ( Exchange Irradiation PHENIX-JOYO )

MARICO*3

Carbide Fuel and Nitride Fuel Irradiation Test

On-site Measurement of Decay Heat of JOYO Spent Fuel

Temperature Distribution Measurement with Optical Fiber

Modification Work of

Cooling System for MK-III

2003

MK-III

( As of Octorber.2008 )

50MW

Performance Test

75MW Operation

Core Replacement

MK-II Irradiation Core

100MW Operation

Transition Core

2004 2005 2006 2007 2008 2009 2010～

SASS*4

MK-III Irradiation Core

140MW Operation

0

1 2

Measurement of Neutron

Flux and Power Distribution

JFY

Ke

y I

tem

s

JFY

Ke

y I

tem

s

50MW Operation

75MW

Performance Test MK-II Performance Test

MK-II Irradiation Core

100MW Operation

Fuel Failure Simulation Test

MK-III Initial Criticality

2003.7.2

MK-III Performance Test

Fuel Failure

Simulation Test

*1 : Instrumented Test Assembly

*2 : Power to Melt Test

*3 : Material Testing Rig with Temperature

Control

*4 : Component Function Test and Element

irradiation Test of Self Actuated Shutdown

System

*5: Minor actinide bearing fuel irradiation test

No. of the irradiated subassembly

Driver fuel

MK-I 116

MK-II 342

MK-III 130

Irradiation test

Fuel 28

Material 56

MARICO 2

SASS 1

Operation time ：70,798 h

Heat generation ：6,244GWh

B11(1)*5

MARICO-2*3

B11(2)*5

B14*5

4

Mark-I

core

Mark-II

core

Mark-III

core

Upgrading of the Joyo cores

Control rod

Core fuel subassembly

Reflector

Blanket fuel subassembly

Subassembly for irradiation

Shielding subassembly

Outer core fuel subassembly

Inner core fuel subassembly

MK-ⅡCore

MK-ⅢCore

Row

4.5

4.0

3.5

3.0

2.5

2.0

1.5

1.066 55 44 33 22 11 0

×1015

Neu

tron

Flu

x（＞

0.1

Me

V）（

n/c

m2･

s)

50/75MWt 100MWt 140MWt

5

6

MK-I

Breeder Core

(1977 - 1981)

MK-II

Irradiation Core

(1982 - 2000)

MK-III

Upgraded

Irradiation Core

(2003 - )

Reactor Thermal Output (MWt) 50 / 75 100 140

Max. Number of Driver Fuel S/A 82 67 85

Max. Number of Test Fuel S/A 0 9 21

Core Diameter (cm) 80 73 80

Core Height (cm) 60 55 50235

U Enrichment (wt%) 23 18 18

Pu Content (wt%) 18 30 30

Max. Linear Heat Rate (W/cm) 320 400 420

Max. Neutron Flux Total (×1015

n/cm2･s) 3.2 4.5 5.7

Fast (> 0.1MeV) (×1015

n/cm2･s) 2.2 3.2 4.0

Max. Burn-up (Pin Average) (GWd/t) 42 75 90

Primary Coolant System Flow Rate (t/h) 2,200 2,200 2,700

Temp. (Inlet) (degree-C) 370 370 350

Temp. (Outlet) (degree-C) 435/470 500 500

Blanket / Reflector / Shielding Blanket / SUS SUS / SUS SUS / B4C

Items

Basic Specifications of Joyo

6

7

Dump Heat

Exchanger

Primary

Pump

Reactor

Vessel

Intermediate

Heat Exchanger

350deg-c

500deg-c

470deg-c

300deg-c

Loop A Loop B

Primary Loop

1350t/h

Secondary Loop

1200t/h

Secondary

Pump

Reactor type : Sodium cooled fast reactor

Reactor thermal power : 140 MWt

Cooling system type : Loop type

Number of main cooling system : 2

Heat Removal : Air Cooling

Joyo MK-III Heat Transport System

High Fast Neutron Flux

Precise Irradiation

・Burn-up and Linear Heat Rate : 3～5％

・Displacement per atom : 5～10％

Non-steady Irradiation

・Power-to-Melt Test（PTM）

・Run-to-Cladding-Breach Test（RTCB）

Advanced PIE Techniques

・PIE Facilities Located Neighboring Joyo

International Utilization

Features of Joyo Irradiation Tests

Neutron irradiation facilities in the world

3D image （Fuel pellet）

X ray CT image of irradiated Joyo driver fuel

1E+13

1E+14

1E+15

1E+16

1E+13 1E+14 1E+15 1E+16

Fa

st

ne

utr

on

flu

x (

n/c

m2・s

)Total neutron flux (n/cm2・s)

1014

1015

1013

Joyo

PHENIX

1014 1015 1016

MonjuSM-3BOR-60

1013

1016

HFIRHFETR

ATR(U.S.)

MIR-1RBR-2

R-2

JMTRHANARO

OSIRIS

JRR-3NRU

HBWRHFR

HIFAR

KUR

HTTR

JRR-4

JHR

Mark-III Core (140MWt)

Control Rod

Reflector

Subassembly for Irradiation

Shielding Subassembly

Outer Core Fuel Subassembly

Inner Core Fuel Subassembly

8

Example of Joyo core configuration

9

Core Center 1st row 2nd row 3rd row 4th row 5th row 6th row 7th row 8th row 9th row 10th row R8 UPR M3

E≧1.0MeV 9.0E+14 1.0E+15 1.0E+15 8.8E+14 7.6E+14 5.5E+14 2.5E+14 9.3E+13 3.6E+13 1.2E+13 4.0E+12 2.2E+12 6.3E+09 3.3E+08

E≧0.1MeV 3.9E+15 4.0E+15 3.8E+15 3.4E+15 2.9E+15 2.2E+15 1.4E+15 7.7E+14 4.0E+14 1.4E+14 4.5E+13 1.0E+13 5.7E+11 2.0E+10E≦0.414eV 3.9E+06 7.6E+06 8.5E+06 6.6E+06 7.2E+06 2.7E+07 6.2E+09 7.7E+09 4.5E+08 4.2E+05 1.8E+05 9.1E+07 6.5E+09 1.6E+11

Total 5.6E+15 5.6E+15 5.4E+15 4.7E+15 4.1E+15 3.2E+15 2.3E+15 1.5E+15 7.7E+14 2.3E+14 7.1E+13 2.3E+13 6.2E+12 7.3E+11

dpa/s (Fe) 1.7E-06 1.9E-06 1.8E-06 1.6E-06 1.4E-06 1.0E-06 5.6E-07 2.8E-07 1.4E-07 4.6E-08 1.5E-08 4.8E-09 2.4E-10 1.3E-11

Flux (n/cm2/sec/140MW)

Mk-Ⅲ6th cycle

Core Configuration

Innner Fuel

Outer Fuel

Inner Reflector

Outer Reflector

Control Rod (B4C)

Neutron Source

Core Materials Irradiation Rig

Material Testing Rig with Temperature Control

Structure Materials Irradiation Rig

Shielding

Uninstrumented Irradiation Subassembly Type-C

Core Center

1st

2nd

3rd

4th

5th

6th

7th

8th 10th

R8

9th

1. Fuel region (450～750deg-C)

2. Reflector region (400～700deg-C)

3. Upper core region ( 550deg-C～)

4. Irradiation hole outside reactor vessel (200～600deg-

C)

UPRM3

R

1cycle=60day, 5cycle/year Core configuration

MK-III 6th cycle

Control rod(B4C)

Inner fuel

Inner reflector

Fuel irradiation subassembly

Shielding

1. Fuel region (450~750deg-C)

2. Reflector region (400~700deg-C)

3. Upper core region (550deg-C~)

4. Irradiation hole outside reactor vessel (200~600deg-C)

Outer fuel

Outer reflector

Structure materials irradiation rig

Material testing rig with

temperature control

Core materials irradiation rig

Neutron source

Core center

UPRM3（R=208cm）

R8

Three materials irradiation rigs were loaded into fuel and radial reflector regions

Flux (n/cm2･s)

Core center 1st row 2nd row 3rd row 4th row 5th row 6th row 7th row 8th row 9th row 10th row R8 UPR M3

E≧1.0MeV 9.0×1014 1.0×1015 1.0×1015 8.8×1014 7.6×1014 5.5×1014 2.5×1014 9.3×1013 3.6×1013 1.2×1013 4.0×1012 2.2×1012 6.3×109 3.3×108

E≧0.1MeV 3.9×1015 4.0×1015 3.8×1015 3.4×1015 2.9×1015 2.2×1015 1.4×1015 7.7×1014 4.0×1014 1.4×1014 4.5×1013 1.0×1013 5.7×1011 2.0×1010

E≧0.414eV 3.9×106 7.6×106 8.5×106 6.6×106 7.2×106 2.7×107 6.2×109 7.7×109 4.5×108 4.2×105 1.8×105 9.1×107 6.5×109 1.6×1011

Total 5.6×1015 5.6×1015 5.4×1015 4.7×1015 4.1×1015 3.2×1015 2.3×1015 1.5×1015 7.7×1014 2.3×1014 7.1×1013 2.3×1013 6.2×1012 7.3×1011

dpa/s (Fe) 1.7×10-6 1.9×10-6 1.8×10-6 1.6×10-6 1.4×10-6 1.0×10-6 5.6×10-7 2.8×10-7 1.4×10-7 4.6×10-8 1.5×10-8 4.8×10-9 2.4×10-10 1.3×10-11

dpa-/year(Fe) 45 49 47 42 36 27 15 7.3 3.5 1.2 0.4 0.1 6.3×10-3 3.3×10-4

Outline of Irradiation Field of Joyo

ー Online irradiation equipments ー Material Testing Rig with Temperature Control (MARICO)

Instrumented Test Assembly (INTA)

Upper Core Structure Irradiation Plug Rig (UPR)

Ex-vessel Irradiation Rig (EXIR)

ー Offline irradiation equipments ー

Uninstrumented Fuel Irradiation Type-A~D (UNIS-A~D)

Materials Irradiation Rig (CMIR, AMIR, SMIR)

１

２ ３

４

10

Region Temp.

Unit:deg-C

Neutron flux Unit: n/cm2･s Environment

Total Fast (E≧0.1MeV)

1 Fuel region 400～1200 (4～5)×1015 (3～4)×1015 Sodium, Lithium

2 Reflector region 350～1200 1014 ～ 3×1015 3×1013 ～ 2×1015 Sodium, Lithium,

inert gas

3 Upper core region 500～ 1011 ～ 1012 1010 ～ 1011 Sodium, Lithium,

inert gas

4 Irradiation hole

outside reactor vessel

200～

～ 1012 ～1010

Sodium, Lithium,

inert gas, Water

Irradiation Test for MA

Recycle

Irradiation Test and PIE

Joyo

Fuels Monitoring Facility (FMF)

●Remote Assembling of

Test Subassembly

●PIE

●Remote Fabrication of

MA-MOX Test Fuel Pin

●PIE

Alpha-Gamma Facility (AGF)

●Irradiation Am concentrate center

of the pellet by irradiation （red parts）

○ Fast Reactor

Fuel

・MOX (High Burn-up, Short Process)

・MA Bearing Fuel

・Metal, Nitride, Carbide

Cladding and Wrapper Tube

Materials

・Austenitic, High Ni Austenitic,

Ferritic,

ODS Ferritic Steel

(370～740 deg-C、0.5～230 dpa)

Core Components

・Long Life Control Rod(Sodium bond)

・Reflector and shielding materials

MA and LLFP Transmutation

○ Others

LWR, HTGCR, Fusion Materials

ODS : Oxide Dispersion Strengthened

430 W/cm

Achievement

11

Upgrading of Irradiation Techniques

Bundle scale irradiation ・Many irradiation data

・Demonstration of modified fuel

Example：

・Fuel containing minor

actinide

・Metal fuel

・Carbide fuel

・Nitride fuel

Fuel

pin bundle

Wrapper tube

Temperature monitoring

and control

Measurement of irradiation condition

Neutron monitor

Temperature monitor

Specimen

pressurized by He

Identification by tag

gas (Xe, Kr)

In-pile creep testing

Flexible temperature

control at outside of

reactor vessel

within ±4deg-C

Capsule rig for advanced

fuel pin

＋4℃

－4℃

680

670

660

650

0 1 2 3 4Time (hr)

12

Compartment

Wrapper Tube

Entrance Nozzle

Handling Head

Test Fuel Pin

Capsule

Compartment

Cross Section

Shroud

Tube

Irradiation Rigs for Various Fuel

-Capsule Type Irradiation Rig-

Features

-Capsule can withstand the pressure

increase in the event of fuel failure

-Allow irradiation of fuels with

insufficient irradiation data

-Flow rate can be set individually

-Allow interim examination and re-

irradiation

Fuel Forms Oxide, Carbide, Nitride, Metal

MA Contents ≦ 50 %

Melting Area of Pellet

Oxide： ≦20 % Others： No Melt

Burn-up ≦ 200 GWd/t

License

13

0:00 12:00 0:00 12:00 0:00

16～18 November 2004

Fuel Failure Simulation Test

Test Pin

Cross Section

Test Pin

Dummy Pin

Compartment Inner Tube

Compartment Outer Tube

Tie Rod

Tie Rod Wrapper Tube

Reference Pin (no Slit)

Slit (0.1x1mm)

Upper End Plug

Spacer Wire

Plenum Spring

Reflector

Fuel Pellet

Cladding

Lower End Plug

Run-To-Cladding-Breach(RTCB) Test

Objective - To verify plant operation procedure in fuel failure events.

14

Verification of a series of operation procedure

1) Detection of Fuel Failure by FFD and Shut-down the Reactor

2) Identification of the Failed Fuel Subassembly by FFDL Systems

3) Un-load the Test Fuel Subassembly

-300

-200

-100

0

100

200

300

400

0 1E+15 2E+15 3E+15

Outlet

Inlet

-300

-200

-100

0

100

200

300

400

300 400 500 600

Inlet Temp. 350 deg-CInlet Temp. 290 deg-C

Inlet Na temperature

350 290 deg-C

Core

Lowering Coolant Temperature

Dis

tan

ce

fro

m c

en

ter

(mm

)

Coolant temp. (deg-C) 1.0 2.0 3.0

370

2x1015

Fu

el re

gio

n

At 370 deg-C, 2x1015n/cm2s can be obtained.

Fast neutron flux (x1015 n/cm2s) 15

Compartment

High temperature

capsule

Irradiation specimen

Tungsten tube

Insulation gas gap

(Ar-He)

Stainless steel tube

High temperature

High Temperature Irradiation

16

Specimen can achieve

over 1,000 deg-C

0

200

400

600

800

1000

1200

1400

0 2 4 6 8 10 12キャプセル中心からの距離（mm）

温度（℃

）

Irradiation

specimen

sta

inle

ss s

teel

Tungste

n

Gas

gap

Radius (mm)

Tem

pera

ture

(deg

-C)

γ heating by W inner tube,

Insulation by Ar gas gap

For Gen-IV reactors(GFR, VHTR), fusion reactors

Fuel Irradiation Rigs

17

Material Irradiation Rigs of Joyo

Compartment

Specimen

Specimen

Specimen

(B4C Pellet)

Structure Materials Irradiation Rig

(SMIR)

Core Materials Irradiation Rig

(CMIR)

Absorber Materials Irradiation Rig

(AMIR)

Orifice

Specimen

holder

(80mm×7)

Core center

Capsule

Primary coolant

inlet hole

Tie rod

Compartment

(Double tube with gas gap)

29

70

mm

Center tube

(Single tube)

Handling head

Center tube

Compartment

Wrapper tube

Entrance

nozzle

18

AA

Material Testing Rig with Temperature Control

(MARICO) for in-pile investigation

11m

Reactor

Core

Cover gas

MARICO-2

γ- ray spectrometry

Released gas from specimen

Gas sampling

1. Creep rupture detection

- Temperature change of capsule

- Activated tag gas nuclides

2. Identification of ruptured specimen

- Isotopic ratios of tag gas by

Laser resonance ionization

mass spectrometry

Drive mechanism

Rotating plug

Guide tube

Long bellows

Short bellows

Latch and cutter

Test assembly

Capsule

Sodium outlet

Thermocouple

Sodium inlet

Gas outlet

Gas inlet

Specimen

Gas gap Change thermal conductivity with different mixing ratio of He and Ar

-Temperature

monitor

(Thermocouples)

Irradiation temperature can

be controlled within ±4 ˚C

from target temperature.

Capsule

19

10-2 10-1 100 101 102 103 104 105 106 1071011

1012

1013

1014

1015

1016

1MeV1keV

Neutron Flux（

n/cm2・

s）

Neutron Energy

1eV

Zirconium

hydride

Neutron

moderator

Neutron Spectrum Tailoring

Versatile irradiation field

・Transmutation study to reduce

environmental burden

・Providing medical, research and

industrial RI

・Basic research of irradiation damage

mechanism

Driver fuel

Moderator subassemblies are

installed in reflector region

20 Neutron energy

Neutr

on f

lux(n

/cm

2・s

)

Fuel

Reflector

Moderator

Low and middle energy High energy

Basic research

of material

RI production

Transmutation study

Evaluation Items and Accuracy for Irradiation Tests

Purpose Evaluation items Accuracy by

calculation

Accuracy by

C/E correction

Fuel

development

MOX fuel, Metal fuel,

Fuel containing minor

actinide,

Short process fuel

Linear Heat Rate

(Fuel temperature),

Burn-up rate

5%

3%

Material

development

Core material,

Structural material,

Control rod material,

Fusion material

Displacements per

atom (dpa),

He production,

Irradiation temperature

5～10%

ΔT=10～20%

3～5%

10%

Trans-

mutation

Minor actinide,

Long-lived fission

product (LLFP)

Transmutation rate,

Irradiation temperature

10%

ΔT=10～20%

5%

10%

Surveillance

test

Reactor vessel,

Safety vessel,

Core support plate

Displacements per

atom(dpa),

He production

10～20% 5～10%

ΔT：Difference from 350 deg-C 21

HAFM

Activation

foil

104

106

108

1010

1012

1014

1016

10-8

10-6

10-4

10-2

100

102

Neutron Energy (MeV)

10B(n,α )

63Cu(n,α )

59Co(n,γ)

58Fe(n,γ)

235U(n,f)

45Sc(n,γ)

237Np(n,f)

93Nb(n,n')

238U(n,f)

58Ni(n,p)

54Fe(n,p)

46Ti(n,p)

9Be(n,He)

Dosimeters cover the neutron energy range

from approximately 100 eV to 20 MeV.

Dosimetry Technique for Joyo N

eu

tro

n f

lux (

n/c

m2・s)/

Leth

arg

y

Neutron energy (MeV)

235U

237Np

Sc

22

Fresh Fuel Logistic and related Hot Laboratories in ORDC

Fuels Monitoring Facility FMF

Material Monitoring Facility MMF

Alpha-Gamma Facility AGF

-PIE of fuel subassembly

-Disassembling & re-assembling

of irradiated rig

- PIE of irradiated fuel

-Remote fabrication of MA-

MOX test fuel pin

&

Center

OaraiR D

Plutonium Fuel Production Facility (PFPF) in Tokai

- MOX Test fuel pins and subassemblies - PIE of irradiated material

PIE

Reloading

Irradiation Rig Assembling Facility IRAF

Plutonium Fuel Research Facility PFRF

Joyo Assembling of irradiation rigs

Research and

development of

advanced fuel

23

Facility FMF MMF AGF

PIE item

- Visual inspection - Detailed visual inspection of the fuel pin

- Profilometry of the subassembly - Weight measurement of the fuel pin

- Eddy current inspection - Profilometry of the fuel pin - γ-scanning of the fuel pin - Fuel pin puncture - X-ray radiography - X-ray CT test - Optical microscope - EPMA, IMA analysis - SEM observation

- Tensile test - Transient burst test of cladding - Density measurement - Charpy impact test - Fatigue test - Magnetization measurement - Uniaxial creep test - Optical microscope - FE/TEM observation - Gas analysis - TEM observation - ICP analysis - Hardness test - Thermal conductivity measurement - Thermal expansion coefficient measurement

- X-ray diffraction analysis

- X-ray radiographic inspection - Optical microscope - Melting point measurement - FP release examination - O/M ratio measurement - X-ray diffraction analysis - EPMA analysis - Evaporative impurities analysis - ICP emission spectrometric analysis

- Halogen analysis - Moisture analysis - MA analysis - Burn-up measurement

Other functions

- Sodium removal equipment - Disassembling of the subassembly - Sample preparation to transport to the MMF or AGF

- Manufacture of specimen from irradiated steel parts

- Gas enclosing (only for non-irradiated specimens)

- Remote TRU fuel pellet production and inspection

- Remote fuel pin welding and inspection

PIE services in the hot facilities related to Joyo

24

Ceramography

Central Void

■X-ray CT Technique

Duct-duct, Bundle-duct Interaction (DDI, BDI), Pellet Structural Change

■Non-destructive Examinations of Fuel Pin

(Ceramography, EPMA, Eddy Current)

■Dismantling of Subassembly

- Visual Inspection, Profirometry

- Samples Preparation in order to

Transfer to AGF & MMF

FMF - Examination and Dismantling of Irradiated Rig -

X-ray Detector

Fuel

Assembly

Collimete

r

X-ray

Source

Twist Traverse

Up and Down

Bundle Type

Irradiation

Test Rig (C4F)

-144 GWd/t

(Pellet Peak)

- 110 dpa (max) X-ray CT Image

Specification ：

- 0.1mm/pixel

- ±0.03mm(for central void diameter)

25

MMF

MMF-2

Main test items in MMF

Material Property Measurement

Density, Thermal Conductivity,

Hardness, X-ray Diffraction,

etc.

Material Strength Testing

Tensile, Creep Strength, Charpy Impact Test,

Temperature-transient-to-burst Test

Examined Materials are mainly……

- Core Materials (Cladding tube, Wrapper tube)

- Structural Materials (Reactor Vessel)

- Neutron Absorber Materials (Control Rod and Shielding)

- Other Advanced Materials for Nuclear Systems

MMF

5.5nm

Microstructural Analysis

FE-TEM Crystal Structure Field Emission-

Transmission Electron Microscope

- Destructive Examination of Irradiated Materials -

26

PIE items

- Optical microscope observation

- Measurement of melting point

- FP release from fuel

- Burn-up measurement

- Measurement of O/M ratio

- Redistribution of MA

etc.

AGF

1401201008060402002800

2900

3000

3100

3200

3300

Thermal arrest met hod

Reference PuO (wt%) Method

Krankota 　and Craig

A　G　S

20

25

18

20

29

V-shaped filament method

2

25

27

融　点　（Ｋ）

燃 焼 度　（GWd/t)

Molten fuel

Central void

Molten

fuel

Longitudinal section

Transverse section

0 20 40 60 80 100 120 140

Burn-up(MWd/t)

3300

3200

3100

3000

2900

2800

Fuel melting point and burn-up

- Detailed Examination for Irradiated Fuels -

Melt

ing

po

int

(deg

-C)

Ceramography of power to melt test

27

PIE cask

Cask car

Remote reassembling Irradiation test

Joyo FMF

Shuttle

Reassembling and Reloading of Irradiation Rigs

Shuttle rig - Test fuel pin and material specimen are replaced.

- Irradiated rig is recycled, and transferred between FMF and Joyo.

Cost down, Shortening of irradiation test term, Radioactive waste reduction

Install irradiated fuel pins or materials into a new

subassembly.

- To obtain the interim examination data

- To continue irradiation for high burn-up and dpa.

FMF is connected to Joyo with the underground passage.

Sample

basket

Capsule

28

Spacer wire wrapping machine

Assembling machine

Glove box (sodium filling into capsules)

Equipment

of IRAF

Capsule and Specimen

Load compartment

to irradiation rig

Irradiation rig

Install handling head

to irradiation rig

Load capsule

to compartment

29

IRAF - Assembling irradiation test rig and device -

Outline of TED monitor

heating

sodiu

m

so

diu

m

Out-of-pile calibration test (~800deg-C)

Thermal Expansion Difference (TED) monitor ◆Principle ： Thermal expansion difference between Inconel and Sodium

◆Accuracy：～±25deg-C (by rough estimation)

◆Supplier ：TED was originally developed by ANL ⇒ JAEA

TED Monitor

300

400

500

600

700

800

0 1 2 3 4

Volume increase ΔV（×10-2cm3）

99.9％信頼幅 T

em

pe

ratu

re in

th

erm

oco

up

le (

deg

-C)

99.9% confidence range

Calibration curve

30

Training and Education at Joyo

University Students

Undergraduate student (3rd or 4th grade) or graduate student

Courses

(1) Reactor physics analysis

(2) Reactor physics experiments using the training simulator

(3) Neutron dosimetry

(4) Tag gas (xenon and krypton) measurement using RIMS

(5) Chemical analysis of coolant Sodium

Lecture on the FBR plant dynamics using the simulator (Course 2)

・Study on Rapsodie’s UCS replacement

・Analysis of Monju primary cover gas by using RIMS

CEA’s internship trainee

31

Appendix

Schematic flow Diagram of Use for Irradiation (An Example of Material Irradiation)

Preparation of irradiation sample

1st year 2nd year 3rd year

Prior inquiry & negotiations

Preparation period

Manufacturing of capsules

Assembling of irradiation rigs

Acceptance review Irradiation (1 cycle*)

Preparation of irradiation plan, Integration of fees

Disassembling, Removal of samples

Contract PIE (Post Irradiation Examination) Contract

Irradiation period PIE period

Note) This flow chart is an example of a case for which the manufacturing of a new material irradiation rig is not necessary.

* The 1 cycle corresponds to a maximum of 60 days.

33

Example of linear heat rate and flux

Transient

over power

in flesh fuel

Transient

over power

from normal

operation Cyclic change

of neutron flux

0

200

400

600

800

1000

-40

-20

0

20

40

60

80

100

1.0E+13

1.0E+14

1.0E+15

1.0E+16

-40

-20

0

20

40

60

80

100

100

80

60

40

20

0

-20

-40 1013 1014 1015 1016

φf(≧0.1MeV) (n/cm2s)

0 200 600 1000

Linear heat rate (W/cm)

100

times

Mo

vin

g r

an

ge

( +

65

cm

to

-2

5cm

)

Core

mid-plane Dis

tan

ce

fro

m c

en

ter

(cm）

Concept of Sample Movable Device

34

Rotating

plug

Core

mid-plane

Core

Top

position

Bottom

position

+6

5c

m

0

-25cm

Insert ,

Withdraw

Changing the linear heat rate and neutron flux by axial movement of sample

Advantages of sample movable device

This device enables to simulate a transient without changing the reactor power.

Concept of Fast Neutron Beam Hole

Short term irradiation without restrictions of rated

power operation.

Rotating plug

Core

In/out test

specimen

Purpose:

- Shortening the irradiation test term

- Production of short-lived isotopes

(Medical isotopes like 99Mo)

Method:

- Load and remove the irradiation

sample through the beam hole

- Utilize neutron beam

35

Natural Circulation Test in Joyo

1 3 . 5 m

2 . 8 m

1 2

1 0

8

6

4

2

G L 0 m

- 2

- 4

- 6

- 8

- 1 0

- 1 2

- 1 4

- 1 6

Primary Pump

Secondary Pump

Dump Heat Exchanger

Intermediate Heat Exchanger

Reactor Vessel

Electro Magnetic Flow meter

Objective

- Demonstration of the decay heat removal capability by natural circulation - Experiment and analysis of natural convection characteristics - Verification of a plant dynamic code (Mimir-N2)

36 Natural Circulation Test from 100MWt Operation (full-power of MK-II core) at 1986

440

450

460

470

480

490

500

-300 0 300 600 900 1200

時間（s）

原子炉出口冷却材温度（℃）

330

340

350

360

370

380

390

原子炉入口冷却材温度（℃）

DHX

<1>

<2>

Reactor Vessel

Main

blower

Secondary loop

Primary loop

70

75

80

85

90

95

100

原子炉出力（％）

-2

0

2

4

6

8

10

反応度（￠）

Feedback Reactivity Measurement

Reactivity : 2.72×10-2%Δk/k (6.22￠)

<1>

<2>

- Control Rod Withdrawal in 120MWt Operation -

Reactivity

Power

Pow

er

(%)

Reactivity (￠

)

100

90

80

70

8

4

0

Therm

al pow

er

(MW

t)

110

115

120

125

130

135

140

-300 0 300 600 900 1200

時間（s）

原子炉熱出力（MWt） Thermal power

140

130

120

110

Time (s)

300 0 600 900 1200 R/V

ou

tle

t Te

mp

. (d

eg

-C)

500

480

460

440

Time (s)

300 0 600 900 1200 R/V

in

let Te

mp

. (d

eg

-C)

380

360

340

37