神戸製鋼技報 - kobelco.co.jp · "r&d" kobe steel engineering reports, vol. 53, no.3 (dec....

, /

1

2

7

12

18 kobesh

23

26 f 3

31

36

41 2 DB

47 HIP

56

61

66

72

78

84 in situ

89

92

98

103 Vol.33, No.1 Vol.53, No.2

105 K-MAX EMK

105

106 PVD 107

"R&D" Kobe Steel Engineering Reports, Vol. 53, No.3 (Dec. 2003)

FEATURE Nuclear Engineering

1 Recent Trends in Nuclear Katsunori Aoki

2 Status of Cask Development at Kobe Steel Dr. Hiroaki TaniuchiKeisuke YoshimuraHiroshi Akamatsu

7 New Heat Resistant Concrete Casks Jun ShimojoDr. Hiroaki TaniuchiKenichi MantaniDr. Eiji OwakiYutaka SugiharaAkihito Hata

12 Borated Aluminum Alloy Manufacturing Technology Jun ShimojoDr. Hiroaki TaniuchiKatsura KajiharaYasuhiro Aruga

18 Kobe Steel's Highly Effective kobesh Neutron Shield Hiroshi AkamatsuDr. Hiroaki TaniuchiKenichi Mantani

23 The Prospect of Enriched Boron Products Dr. Hiroaki TaniuchiJun ShimojoKenichi Mantani

26 Nuclear Waste Storage Cask Maintenance Facility Naoyuki FurutaHitoshi YamadaMasamitsu NakataniKeiichi OgawaMakoto ShirataniAkira Nishikoba

31 BP Volume Reduction Equipment Yoshinori KitamuraYoji MurooIsao Hamanaka

36 The Applicability of Cold Crucible Induction Melting to Nuclear Engineering Takashi NishioAkira WadamotoTatsuhiko Kusamichi

41 An Automatically Controlled System for Waste Transport in Low Level Nuclear Waste Storage Facilities Yoshinori KitamuraHidetoshi Miyaue

47 HIP Rock Solidification Technology for Radioactive Iodine Contaminated Waste Ryutaro WadaTsutomu NishimuraYoshitaka KurimotoDr. Tsuyoshi Imakita

56 An Incineration Technology for Low Level Radioactive Solid Waste Mamoru SuyariRyota NakanishiTsuyoshi NouraMasashi FujitomiShintaroh Ano

61 A Plasma Melting System for Solid Radioactive Waste Dr. Yasuo HigashiMasahiko SugimotoMasashi FujitomiTsuyoshi Noura

66 A Treatment Technology for Liquid Waste Generated from Nuclear Reprocessing Facilities Yoshiaki TanakaToshio IwataAkira Wadamoto

72 Low Permeability Layer "BENTBALL" Ryutaro WadaKenji YamaguchiYasunori TakeuchiJunji KumamotoHideo KomineHiroshi Nakanishi

78 Evaluation of Gas Generation Rates Caused by Metal Corrosion under the Geological Repository Conditions Tsutomu NishimuraRyutaro WadaKazuo Fujiwara

84 Solubility Assessment Technology for High Pressure Environments by in situ Laser-induced Fluorescence Spectroscopy Kenji YamaguchiDr. Seiichi YamamotoDr. Kaoru MasudaTakahiro ShimizuDr. Tsuyoshi ImakitaShun Sakamoto

89 Information Gathering Robots for Nuclear Accidents Jumpei NakayamaMasahiko Sugimoto

92 Fuel Cladding Materials for Supercritical-water Cooled Power Reactors Makoto HaradaOsamu KubotaHiroyuki Anada

98 Functions and Fabrication Technologies of Fuel Channel Masayuki NodakaKyosuke Fujisawa

103 Papers on Advanced Processing Technologies for Nuclear Presented in R&D Kobe Steel Engineering Reports (Vol. 33, No.1 Vol. 53, No.2)

1999

2002

RD

1960

1975

1980

60

1989 4 Vol. 39No. 2

1990 2000

1/Vol. 53 No. 3Dec. 2003

FEATURE : Nuclear Engineering

Recent Trends in NuclearKatsunori Aoki

1980 TN 1 150

1

1980 TNNFT

11TN

COGEMACOGEMA LOGISTICSACL TRANSNUCLEAIRE COGEMA2002 TN12 PWR 12 1 2.5m 6.5m 115 TN12 TN12APWR 12 TN12BBWR 32

2 KOBE STEEL ENGINEERING REPORTS/Vol. 53 No. 3Dec. 2003

Status of Cask Development at Kobe Steel

Kobe Steel has been involved in the design, safety analysis and fabrication of transport and/or storage casks for radioactive materials for more than 20 years. Transport casks were primarily developed early on, however, now production has largely shifted to storage casks. To make the casks as safe as possible, without huge added expense, the advanced types of casks have been and will be developed and new materials such as high performance neutron shields and neutron absorbing materials are being increasingly developed and used.


Dr. Hiroaki Taniuchi

Keisuke Yoshimura

Hiroshi Akamatsu

No. of casksType of caskDelivery year

68TN type transport cask1981-2003

2JRC-80Y-20T transport cask1981

61TN type transport/storage cask1985-2003

19NFT type transport cask1997-2000

25Cask for radioactive waste1988-2001

12Others1988-1993

187Total

1 Casks fabricated by Kobe Steel

TN17 BWR 17 12

TN 2JRC-80Y-20T20 9 m BU13NFT 1

4 BWRNFT63NFT-38BNFT

2

TN24 TN24 TK69 21 TN24

ACLTN1983 TN2/5 9 m 42R&D1985 TN24 TN24

3/Vol. 53 No. 3Dec. 2003

1 TN12 TN12 type transport cask

3 NFT NFT type transport cask

2 JRC-80Y-20T JRC-80Y-20T type transport cask

4 TN24 2/5 TN24 2/5 scale model drop test

PWR 24 1 1 Idaho National Engineering and Environmental LaboratoryINEELINEEL 2 322 TN24

1990 TN24 TN24 1995 9TN245 2ACLTN24 TN24

TN24DTN24XL ACL TRANSNUCLEARTNY TN-32TN-40TN-68 TN-32TN-40 23TK

TN24 1997 ACL TN24 TK2BWRTK-69BWR 69 TK 2TK-69 TK-69


ConcreteKATSTKTN24

BWR typePWR typeKATS-B69KATS-P24TK-69TK-PWRDomestictypePrototype

BWRPWRBWRPWRBWRPWRBWRPWRFuel type45 00050 00033 00044 00033 00044 00033 00035 000Average burn-up (Mwd/tU)10101010101045Cooling time (years)165

168

132120

131119

132120

132120

115

9488

TransportStorage

Total weight(ton)

6.23.4

6.23.4

5.32.6

5.22.7

5.42.6

5.12.6

5.62.5

5.12.3

Axial (Main body)Diameter (Main body)

Length(m)

More than 52More than 21692469More than 265224No. of loaded fuels2020192119252824Total heat (kW)

Under developmentUnderdevelopmentLicensedfor storageNote

2 Major characteristics of casks designed by Kobe Steel

5 TN24 TN24 type transport/storage cask

1 TN24 Structure of TN24 type transport/storage cask

Secondary lid

Primary lid

Upper trunnion

Outer shell

Copper fin

Main body

Basket eB-AL developed by KSL

Neutron shielding e developed by KSL

Lower trunnion

kobesh

kobesh TK

3

TN 231KATS

1990 KATS KATS 3KATS 32

150 4 2

4

20

5/Vol. 53 No. 3Dec. 2003

2 TK69 Structure of TK69 type transport/storage cask

Secondary lid

Upper trunnion

Main body

Basket

Outer shell

Neutron shielding

Lower trunnion

Primary lidPressure monitoring

Copper fin

3 KATS Structure of KATS type transport/storage cask

Secondary lid

Primary lid

Basket

Outer shell

Inner shell

Neutron shielding

Upper trunnion

Pressure monitoring

Lower trunnion

Copper fin

kobesh 41 kobesh

kobesh SREPRTHPP 4 SR TN24EPR TK

SR TH42

TN24 1995 4 5 11 A6061A3004 43

ACL 1984 ACL TNT2002 TNT ACL TNT 1 S. ShimuraRAMTRANS, Vol.8, Nos3-41997, p.257. 2 J. M. Creer et al.Electric Power Research Institute, EPRI NP-

51281987 3 M. A. McKinnon et al.Electric Power Research Institute,

EPRI NP-61911989


4 Structure of seal type concrete cask

Concrete lid

Canister lid

Basket

Heat resistant concrete

Inner shell

Copper fin

Canister

Pressure monitoring

Outer shell

2010 100 65 901SUS329J4L25Cr-6Ni-3Mo-0.2N-LC1150

2

1

11

10010012

1 150 JIS R 2616 JIS A 1325JIS A 1108

7/Vol. 53 No. 3Dec. 2003

New Heat Resistant Concrete Casks

Heat resistant concrete containing hydrogen has been developed in the design of a new type of cask that has been modeled on the same concept of metal cask technologies for use under high temperature conditions. The allowable temperature of conventional concrete is limited to 90 because most of its moisture is free water and therefore hydrogen, which is effective for neutron shielding, can be easily lost. Our newly developed concrete uses crystal water and as a result can be used under high temperatures.


Jun ShimojoDr. Hiroaki Taniuchi

Kenichi MantaniDr. Eiji Owaki

Yutaka Sugihara

Akihito Hata

150 30g 1 0006 2 3 150 5 5 15013

140 60

15030 3 5


3 150

Mass variation of heat resistant concrete during heated time (Heated temperature150)

1.00

0.98

0.96

0.94

0.92

0.901 10

Heated time(h)100 1 000

Relative mass variation

Sample No.1 Sample No.2

Compressivestrength(MPa)

Specific heat(kJ/(kgK))

Coefficient oflinear expansion

(l/K)

Heat conductivity(W/(mK))

Moisture content(mass)

Density(g/cm3)

870.941.11052.016.62.3Heat resistant concreteat room temp.

1.410.82.17Heat resistant concreteat 150

184041.01.321.010512.62.834722.252.31Ordinary concreteat room temp.

1 Properties of heat resistant concrete and ordinary concrete

Note 1According to reference 2Note 2According to reference 3Note 3According to reference 4Note 4According to reference 5

1 Conventional concrete cask

Concrete lid

Air outlet

Basket

Canister

Steel bar

Air inlet

Canister secondary lid

Ordinary concrete

Steel liner

Canister primary lid

2 New type concrete cask

Concrete lid

No air inlet and outlet which is conventionally required

Canister lid

Basket


Inner shell

Outer shell

Pressure monitoring

Canister

Copper fin

58015014

1

2

21

24 2

1 m 100Sv/h 8ORIGEN2 9 ANISN 10DLC23/CASK 11 ICRP Publ.74 12 31502.17g/cm31 m 4 70 30

3

9/Vol. 53 No. 3Dec. 2003

1 Cut sample of heat resistant concrete

RemarksCondition

BWR STEP 3.545 00055 00025101.3

(1) Fuel specification Fuel type Initial enrichment () Average burnup (MWD/MTU) Maximum burnup (MWD/MTU) Specific power (MW/MTU) Cooling time (year) Peaking factor

NoteAccordingto reference 7)

522.152.17

(2) Calculation condition Number of fuel assemblies Density of ordinary concrete (g/cm3) Density of heat resistant concrete (g/cm3)

2 Specification and condition of shielding calculation

Ordinary concrete

(atoms/barncm)Heat resistant concrete(atoms/barncm)Element

5.34 10 3

4.11 10 2

6.13 10 5

2.14 10 4

1.78 10 2

2.22 10 3

6.35 10 4

1.6 10 2

2.0 10 2

6.9 10 4

1.1 10 2

8.1 10 3

H

C

O

Mg

Al

Si

Ca

Fe

3 Atomic density of concrete material

Dose equivalent rate (Sv/h)

TotalNeutronGamma

69

100

13

42

56

58


Ordinary concrete

4 1 mDose equivalent rate at 1m from surface of cask

4 Shielding calculation model

App. 75

Fuel region

Carbon steel (Canister)

Air Carbon steel (Inner shell)

ConcreteAir 100

Detector(unit : cm)

App. 80

Carbon steel (Outer shell)

Note According to refrence 7)

1/3 2 3 5

2.5

4

2 1 10041

42

43

100


1 2

2 1/3 Outer view of 1/3 scaled model

3 1/3 Cross section of 1/3 scaled model

Measured data

232.352.5

Sampling numberAverage density (g/cm3 )

Relative standard deviation ()

5 Dispersion of concrete density

Note Density data are measured at room temperature without heated.

1 1

2002

2 2002p.29

3 JAERI-M-86-0601986p.143.

4 2002p.46

5 JIS A 5308-1998 .

6 2002 2002p.122.

7 JAERI-Tech-96-0011996p.92.

8 1992

9 A.G.CroffORIGEN2 - A Revised and Updated Version of Oak Ridge Isotope Generation and Depletion Code, ORNL-56211980

10 R.G.SolteszRevised WANAL ANISN Program Users Manual, WANL-TMI-19671969

11 ORNL-RSIC, CASK-40 Group Coupled Neutron and Gamma-ray Cross-section Data, DLC-231973

12 ICRP, Conversion Coefficients for use in Radiological Protection against External Radiation, Publication 741995

11/Vol. 53 No. 3Dec. 2003

1DC 4 5mass 1 2mass 10B

13

1

11

AlB21a 700 1mass B-A6061 1b 950Al-B Al 4 1bBMg EPMAElectron Probe Micro AnalyzerMgMg



Borated Aluminum Alloy Manufacturing Technology

Borated aluminum alloy is used as the basket material of cask because of its light weight, thermal conductivity and superior neutron absorbing abilities. Kobe Steel has developed a unique manufacturing process for borated aluminum alloy using a vacuum induction melting method. In this process, aluminum alloy is melted and agitated at higher temperatures than common aluminum alloy fabrication methods. It is then cast into a mold in a vacuum atmosphere. The result is a high quality aluminum alloy which has a uniform boron distribution and no impurities.

Jun ShimojoDr. Hiroaki Taniuchi

Katsura Kajihara

Yasuhiro Aruga

10B 11B 20at.80at.10B 11B 10B

12

2 1 000

1 2mass 800 2mass1 300 1 500 4DC

2

21

21 000

13/Vol. 53 No. 3Dec. 2003

Mg EPMA

B

(a) Agglomeration of boron compounds (b) Giant boron compound

100m200m

Material bucket

Vacuum pump

Vacuum pump

Casting room

Melting roomVacuum induction furnace Handling

container

Casting mold

1

Coarse boron compounds in conventional melting process

2 Vacuum induction melting equipment

Secondary lid

Primary lid

Upper trunnion

Inner shell

Basket

Outer shell

Neutron shielding

Copper fin

Pressure monitoring

Lower trunnion

1 Transport and storage cask for spent fuel

22

1massB-A6061-T6511massB-A3004-H112 110mm T651 12mmt170mm 21000 3000 5000 6000 23

3 1massB-A3004AlB224

3a 1massB-A6061b 1massB-A3004 25

1massB-A6061-T651 1massB-A3004-H112


2 Trial product of borated aluminum alloy

1 mass

Chemical composition of structural borated aluminummass

3 Macroscopic distribution of boron content

Alloy B Si Fe Cu Mn Mg Zn Cr Ti

1massB- A6061

0.6 1.1

0.6 1.3

0.40 0.80

0.30

0.70

0.7

0.15 0.40

0.25

0.15

1.0 1.5

0.8 1.2

0.8 1.3

0.25

0.25

0.04 0.35

0.15

1massB- A3004

Side

Center

Head (top of ingot)

Head (top of ingot)

Tolerance

Center

Center

Side

Side

Result of chemical analysis (B)

0.61.1mass 0.760.91mass (15 positions)

Tail (bottom of ingot)

Tail (bottom of ingot)

thickness10mm, width900mm, length30 000mm(1 ingot)

thickness12mm, 170mm, length26 000mm(1 ingot)

(a) 1massB-A6061 rolled plate

ToleranceResult of chemical analysis (B)

0.61.3mass 0.821.03mass (10 positions)

(b) 1massB-A3004 extruded pipe(b) 1massB-A3004 extruded pipe

(a) 1massB-A6061 rolled plate

100m

3 1massB-A3004 Microstructure of 1massB-A3004 extruded material

made by VIM process

2.2 A6061 A3004

251

20.2 4 5 5 0.2

15/Vol. 53 No. 3Dec. 2003

4 1massB-A6061-T651A6061-T6

Comparison of tensile properties at high temperatures of 1massB-A6061-T651 and A6061-T6

2 Typical mechanical properties of structural

borated aluminum alloy

Alloy Condition Product form

At room temperature Tensile strength 0.2Proof strength Elongation

338 MPa 303 MPa 13

(approximately)

187 MPa 85 MPa 23

(approximately)At 473K Tensile strength 0.2Proof strength Elongation

Tensile direction (Plate) Transverse direction of the rolling direction (Pipe) Longitudinal direction of the extruding direction

237 MPa 218 MPa 13

(approximately)

114 MPa 79 MPa 40

(approximately)

Borated A6061 T651

Rolled plate

Borated A3004 H112

Extruded pipe

350

300

250

200

150

100

50

0300 350 400 450 500 550 600 650250

Temperature (K)(a) Tensile strength

Tensile strength (MPa)

Borated A6061-T651A6061-T6

350

300

250

200

150

100

50

0300 350 400 450 500 550 600 650250

Temperature (K)(b) 0.2Proof strength

0.2 Proof strength (MPa)


30

25

20

15

10

5

0300 350 400 450 500 550 600 650250

Temperature (K)(c) Elongation

Elongation ()


200 180 160 140 120 100 80 60 40 20 0

140

120

100

80

60

40

20

0

120

100

80

60

40

20

0

300 350 400 450 500 550 600 650250Temperature (K)(a) Tensile strength

300 350 400 450 500 550 600 650250

Temperature (K)(c) Elongation

300 350 400 450 500 550 600 650250Temperature (K)

(b) 0.2 Proof strength

Tensile strength (MPa)

0.2 Proof strength (MPa)

Elongation ()

Borated A3004A3004

Borated A3004A3004

Borated A3004A3004

5 1B-A3004-H112A3004-H112

Comparison of tensile properties at high temperatures of 1B-A3004-H112 and A3004-H112

252

100300 10 98MPa6 51mass 6

3

31DC

4 5mass12

DC DC 32

DC2massB-A6351339mm 4 510B


0.1mm

1 000

1 000

100

10

1

100

10

1

8 9 10 11 12 13

473K, 100 000h

473K, 100 000h

14

8 9 10 11 12 13 14

Larson-Miller parameter PT20log t103

Larson-Miller Parameter PT20log t103

A6061-T6Borated A6061-T651

Rupture stressMPa

Rupture stressMPa

(a) Comparison 1massB-A6061-T651 and A6061-T6

(b) Comparison of 1B-A3004-H112 and A3004-H112

A3004-H112Borated A3004-H112

6 Larson-Miller Larson-Miller parameter on creep rupture stress properties

4 DC 2massB-A6351 Microstructure of 2massB-A6351 ingot made by DC

process

5cm

5 2mass-A6351 DC Neutron radiography of 2mass-A6351 ingot made by DC

process

DC1mass1 000DC

1 J. Shimojo et al.: Proceedings of 13th International Symposium on

the Packaging and Transportation of Radioactive Material PATRAM2001.

2 K. Kajihara et al.: Proceedings of 10TH International Conference on Nuclear EngineeringICONE102002.

3 2002 2002p.300.

4 M. Hansen et al.CONSTITUTION OF BINARY ALLOYS 1991p.71.

5 J. Gilbert KaufmanProperties of Aluminum Alloys, Tensile, Creep and Fatigue Data at High and Low Temperatures, The Aluminum Association and ASM International1999.

6 Vol.39No.31997p.237.

17/Vol. 53 No. 3Dec. 2003

kobesh

1kobesh

kobesh Silicone rubber base Polypropylene base Ethylene propylene rubber baseTitanium hydride base4 1 kobesh 1kobesh 11Silicone rubber base kobesh

4.04.5 1022atoms/cm3 5.51022atoms/cm3



kobesh

Kobe Steel's Highly Effective kobesh Neutron Shield

Recently, the management, transport and storage of spent fuels from the nuclear power reactors has become more and more important. A highly effective neutron shield called kobesh has been developed by Kobe Steel to improve safety and the overall economic management of spent fuel transport and management. This paper explains the technical characteristics of kobesh .

Hiroshi AkamatsuDr. Hiroaki Taniuchi

Kenichi Mantani

1 kobesh

kobesh lineup

Hydrogen titanium base

Ethylene propylenerubber basePolypropylene baseSilicone rubber baseType

TH-OEP-REP-OPP-RPP-OSR-TSR-OSeries

2.63.71.11.41.050.91.30.91.41.91.4Density (g/cm3)

8.910226.410226.110227.610227.710225.510225.01022H-Content(max. atoms/cm3)

VariableVariableVariableVariableVariableVariableVariableB-Content

300150150120120170170Thermal stability for long use ()

Pre-shapedPre-shapedPre-shapedPre-shapedPre-shapedPouringpre-shapedPouringpre-shaped

Fabricationmethod

Used in fire protecting cover

Used in fire protecting cover

Remarks

n, 12Polypropylene base kobesh

kobesh 7.71022atoms/cm3 0.9g/cm3Silicone rubber base kobesh n, 13Ethylene propylene rubber base kobesh

kobesh 6.71022atoms/cm3Silicone rubber base kobesh n, 14Titanium hydride base kobesh

8.91022atoms/cm3

2kobesh

1 100150 kobesh

19/Vol. 53 No. 3Dec. 2003

Secondary lid

Primary lid

Trunnion

Shell

Basket

Outer shell

Cu fin

Monitoring equipment

Trunnion

Neutron shield ( )kobesh

1 TK TK type transport/storage cask

1 kobesh kobesh

SR series kobesh PP series and EP series kobesh TH-O series kobesh

21

kobesh

kobesh 2 1 3 1 4 1 5 1 4 24 kobesh 44 kobesh 5 322

kobesh


12 12.5

105.0

Neutron shields 5050

252Cf source (37MBq) (point source)

Rem-counter

Effective detector point (Surface detector {20.0 in diam.})

(unit cm)

2 Shielding performance test shielding material only

235

200 640

650

Iron plate 550

YAYOI reactor Fast columnNeutron beam

Neutrack TS-16N

1 250

Neutron shield sample plates

610 600 (Unit : mm)

3 Shielding performance test (transport/storage cask shielding

structure)

Exp. Cal.EPR SR

100

10

10

10 5 10

Neutron dose rate(Sv/h)

(b)

(a)

Thickness of neutron shield(cm)

Water EPR SR PP TiH2

4 Shielding performance test result (shielding material only)

0 5Thickness of neutron shield(cm)

Neutron dose rate(mSv/h/W)

10 15

100

101

102

103

EPR SR TiH2

5

Shielding performance test result (transport/storage cask shielding structure)

1 6 2 3ln/ 1 J/mol J/mol/K K 7 150 20 Silicone rubber base kobesh 170Ethylene

propylene rubber base kobesh 150Polypropylene base kobesh 12023

200

2AlOH3 Al2O3H2O 2H2O

kobesh

21/Vol. 53 No. 3Dec. 2003

101 102 103

Test duration(days)

Weight change()

2 1 0

1 2 3

1 0

1 2 3 4 5 1 0

1 2 3 4 5

5 4 3 2 1 0

1

2 1 0

1 2 3 4

(a)

(b)

(c)

(d)

(e)

101 102 103

Test duration(days)

Hydrogen content()

5

4

3

2

1

6

5

4

3

2

14

13

12

11

10

14

13

12

11

10

9

8

7

6

5

(a)

(b)

(c)

(d)

(e)

120

140

120

140160

170

120

140

160

140-170

140-170

170160140

140

140160

160

120

120140

140

120

160

160170

170

6 Thermal degradation data by long term heat resistance test

aTitanium hydride type, bSilicone rubber type,cPolypropylene, dEthylene propylene rubber type, ePolyethyleneNumbers shown in the figures mean ambient air temperature.

kobesh

kobesh kobesh

1 H. TaniuchiStudy on Shielding Performance

of Spent Fuel Transport and Storage Packages1999 p.145. 2 T. Iida et al.International Journal of Radioactive Materials

Transport Vol.21991 p.79. 3 1989 p.36.


Fitting line

Useful life of cask

Allowable service life(days)

SR

EPR

PP

104

103

1022.5 2

1 000/T(K1)

7 kobesh Evaluation of allowable service life for kobesh

10B11B 210B20 11B10B19942001

1

1994194030 1, 211B10B10B 1

BF310B 10B

23/Vol. 53 No. 3Dec. 2003

The Prospect of Enriched Boron Products

A mass production technique for producing enriched boron was developed jointly by Kobe Steel and Stella Chemifa Co. in the 1990s. Enriched boron commercial production started in 2001 and since then, as a result of boron market research, several new enriched boron materials such as boron aluminum, boron acid, and boron carbide have been added to our production schedule. The demand for enriched boron is expected to increase rapidly if the material can be steadily supplied at a reasonable price.


Dr. Hiroaki Taniuchi

Jun Shimojo

Kenichi Mantani

Product

Maximum enrichment 95

Circulated complex agent

Natural BF3 gas Natural boron composition 10B:19.9 Neutron absorbing material 11B:80.1

10BF3 solution

11BF3

Tower

10BF3 complex

10BF3

Pump

BF 3 gas

Complex agent

Heating

resolving

tower

1 Flow chart of enriched boron plant

1 10B 952001

2.

5 221

BWRmass 1606130041PWR1

22

1 3 3 BWR PWR MOX 2 FeB23

PWR pH


1 Enriched borated aluminum

Enriched boron

Boric acidH3BO3, B2O3

Borated aluminum alloy

Boron carbideB4C

FerroboronFeB

Reactor water

Basket material for packagings, etc.

Control rods, etc.

Borated stainless steel

2 Lineup of application products using enriched boron

7L i6Li5MOX 7Li MOX31 1

2PWR 11

3

8 000ppm6020 000ppm80

24

PWRBWRMOX425

10B

2001 1 Vol.21959p.273. 2 VOL.531977p.239.

25/Vol. 53 No. 3Dec. 2003

4 B4C Enriched boron carbideB4C

3 H3BO3 Enriched boric acidH3BO3

2 FeB Enriched borated ferro-boronFeB

f 3 f 31990

19941997

1

f 3NFT 6HZ3130 f 3


f3

Nuclear Waste Storage Cask Maintenance Facility

In a nuclear fuel cycle, it is important to transport spent fuel from a reactor site to a reprocessing plant safely. Currently, a commercial reprocessing plant is being built at Rokkasho, in Aomori, at the northeastern end of the main island of Japan. In this plant, a nuclear waste cask maintenance facility is also under construction by Kobe Steel. The purpose of this facility is to systematically maintain a large number of spent fuel casks. This facility is the only facility of its kind in Japan. This paper introduces an overview of the cask maintenance facility.


Naoyuki Furuta

Hitoshi Yamada

Masamitsu Nakatani

Keiichi Ogawa

Makoto Shiratani

Akira Nishikoba

Top shock absorbing coverLid

Neutron shielding

FinThermal fin

Bottom shock absorbing cover

Thermal barrier

Trunnion

Transport frameOuter shell

BasketBody

Trunnion

1 NFT38B Transport packaging for spent fuelNFT38B

1 3 10

2

f 3

3

f 31

2 3 4

4

f 3 f 3 21 f 3 f 3

27/Vol. 53 No. 3Dec. 2003

2 Material handling flow

Cask receiving Preparation of cask transfer Cask transfer Decontamination of cask Cask maintenance Cask transfer Preparation of cask delivering Cask delivering

2 3 10 3 1 PLC 3 2 3 4


1 Master slave manipulatormaster arm

2 Washing bar

6ITV

SGN4 22.13MPa3.5MPa f 3TNT

29/Vol. 53 No. 3Dec. 2003

3 US bar for inside of the structure

ITV 6

4 Structure outside washing

device

5

f 31997NFT1 ITV

2 2 3mmf 3

f 310 20041


20067BP BP 1104 000mm1986 BP BP

1BP

BP

BPBPBPBP 3

2BP

21BP

BP 10mm4 000mmBP 200 BP 22BP

1 1 3 BP BP 23

BP BP 32

31/Vol. 53 No. 3Dec. 2003

BP Volume Reduction Equipment

A new type of burnable poison (BP) volume reduction system is currently being developed. Many BP rods, a subcomponent of spent fuel assemblies are discharged from nuclear power reactors. This new system reduces the overall volume of BP rods. The main system consists of BP rod cutting equipment, equipment for the recovery of BP cut pieces, and special transport equipment for the cut rods. The equipment is all operated by hydraulic press cylinders in water to reduce operator exposure to radioactivity.


Yoshinori Kitamura

Yoji Muroo

Isao Hamanaka

SUS630


Water pit for installation of cutting equipment

Receiving portionfor cut pieces of BP rods

Cutting portionof BP rods

Feeding portionof BP rods

Water-hydraulic cylinder

Driving gear box

Underwater ITV camera

Driving gear box

Water-hydraulic cylinder for BP rods cutting block

Water-hydraulic cylinder forBP rods clamping block

Bottom lining of water pit

Carriage of container forcut pieces of BP rods

Container for cutpieces of BP rods

Underwater ITV camera

1 BP BP volume reduction equipment

1 BP General view of BP volume reduction equipment

231

200BP10 1

3 000mm1 15

33/Vol. 53 No. 3Dec. 2003

Separating plate

Drive-chain for lifting

of separating plate

Water-hydraulic pump

(for high pressure circuit)

Water-hydraulic pump

(for low pressure circuit)

Solenoid valve stand

(equipped with solenoid

controlled valves,

metering valves,

pressure reducing

valves, check valves,

etc.)

Water-hydraulic cylinders

Driving gear box

(for liftable separating plate

of BP rods, powered by

water-hydraulic cylinder)

Water pit for

installation of cutting

equipment

Operating floor

Water-hydraulic hoses

Rack gear, powered by

water-hydraulic cylinder

2

Water-hydraulically driving flow diagram

BP 200 BP BP 30 232

BP 3 BP 4 000mmBP233

BP 2 3 1 400mm 1 2

BP 1 10 1 4

3ITV

BP2ITV BP


3 BP Principle of cutting for BP rod

Clamping block

Cutting block

Clamping block

Movable blade Fixed blade

Gauge stopper

BP rods

4 Mechanism of turning drive of container

Driving mechanismfor turner of container

Driving gear box for turner of container powered by water- hydraulic cylinder

Turning gearfixed to receivingplate of container

Gear for turningof container

Container of cutpieces for BP rods

Turning diskfor container

35/Vol. 53 No. 3Dec. 2003

Cold Crucible Induction Melting CCIMCCIM

1

FBR FBR FBR 1994 3 2 1UO2 UO22 UO2 UO2


The Applicability of Cold Crucible Induction Melting to Nuclear Engineering

Cold crucible induction melting (CCIM) technique which is used routinely to refine and melt active metals like titanium has excellent characteristics and has recently shown promise in the nuclear engineering field. To evaluate the applicability of CCIM in the electrolyzer portion of the oxide-electrowinning process and the melting decontamination process for low-level radioactive metal waste contaminated with uranium, several experiments were conducted. Experimental results showed that the CCIM technique could be adapted to the nuclear engineering field.


Takashi Nishio

Akira Wadamoto

Tatsuhiko Kusamichi

UO22 UO22

Cl2

UO2

Pu4

Removal of salt

Removal of salt/Separation of NM

UO2 NMUO2 UO2

Spent nuclear fuel

Molten salt (NaCl-2CsCl)

Cl2

Removal of salt

UO2PuO2 MOX

Cl2 Cl2O2

Dissolution and electrowinning operation

(Selective retrieval of UO2)Anode : UO2UO222e

Cathode : UO222eUO2

Dissolution operation Removal and electrowinning operation

UO2Cl2UO222Cl PuO22Cl2 Pu44ClO2

(Selective removal of NM) NM : noble metal in

spent nuclear fuelAnode : 2ClCl22e Cathode : UO222eUO2 NMXXeNM

MOX electrowinning operation

Pu42ClO2 PuO22Cl2 Anode : 2ClCl22e

Cathode : UO222eUO2 PuO222ePuO2

Preparation of salt component, Retrieval of TRU and Removal of FP operation

UO22

Pu4PuO22

1 1

Oxide-electrowinning process1

PuO2MOXUO2PuO2MOX 2CCIM2CCIM FBR CCIM

1 070 mm 1 000mm 3 50mm 200mmCCCIM 111

2CsCl-NaCl 3kHz 650 150 120 650 12.5mm750 AgCl Ag 12CCIM

CCIM 1 070mm 1 000mm C 1 000mmCCIM200mm 170mm

37/Vol. 53 No. 3Dec. 2003

2 CCIM Conceptual drawing of CCIM

Current

Pass of cooling water

Molten metal

Crucible

Solidified layer (Skull)

Magnetic fieldForce

Induced current

Molten metal flow

Coil

Segments

3 C Annular shaped crucible made of hastelloy C

Inner crucible

Hastelloy

Molten salt

Solidified salt layer (Skull)

Carbon heater

Pass of cooling water

Outer crucible

Power supply

3kHz 400kW

FrequencyPower

Crucible

204mm24 50kg

DiameterNumber of segmentsCapacity of molten metal

Vacuum chamber

0.1105Pa Pressure

1 CCIMBasic specification of the CCIM equipment

1 070mm1 000mm 20 701 070mm 200mm13

CCCIM

2

42CCIM

CCIM21

211

Ce CeO23CeO2CeAlMg AlCaO-Al2O3-SiO2CeO2Al CCIM 1SUS30440kgCeO20.04 0.4 kg 2 kgAl0.6 0.8 kg CeAl CeAl Ce CeICP


Low level radioactive metal waste contaminated with uranium

Slag phase

Transition

Metal phaseSlag

Slag including uranium oxide

Solidify

Disposal

Utilize usefully

Melt

UO2UO2

UO2

Decontaminated metal

4 Melting decontamination process

Ce concentration in thesolidified metal (ppm)

Melting point()

Basicity

Slag composition(mol)

BottomMiddleTopAftermeltingBeginningSiO2Al2O3CaO

0.10.10.10.10.1

0.10.20.10.10.1

0.10.10.10.10.1

1 4501 4001 5001 4001 550

1 4101 3001 4001 3501 350

0.250.490.6411.5

7060354414

10726626

2033395060

2 Experimental result of the

effect of slag composition

ConditionStainless steel : 40kg, Slag : 2kg, CeO2 : 40g, Al : 0kgHolding time in molten state : 30min

0.1ppm212

1CaO-Al2O3-SiO22CeO2 Ce 0.5ppm Al Si CeO22CeO2 CeO2 40g 10 400g Ce 0.5ppm Ce Ce 38 00011 0003 4Al Ce0.5ppm Ce100ppm

200 CeO2CeO2 SUS AlSi Al2O3 SiO2CeO2 Al Si CCIMAl 460 5 60 Al 1.9Ce Ce 0.1ppm CeCCIM5 Ce 2 4Ce

39/Vol. 53 No. 3Dec. 2003

Elementary concentration in themolten metal (ppm)

Holding timein moltenstate (min) AlCe

2.12104

2.03104

1.89104

1.90104

1.88104

1.76104

1.69104

1.52104

0.20.10.10.10.10.10.10.1

0306090105120135150

5 Experimental result of the effect of holding time in molten

state

Decontaminationfactor


Amount ofCeO2 added(g) BottomMiddleTop

11 000 8 000

0.10.1

0.10.7

0.10.4

40400

3 CeO2Experimental result of the effect of CeO2

ConditionStainless steel : 40kg, Slag : 2kg, Al : 0kgSlag compositionCaOAl2O3SiO250644 (mol)Holding time in molten state : 30min

4 Experimental result of the effect of anti-decontamination

element

ConditionStainless steel : 40kg, Slag : 2kg, CeO2 : 40gSlag compositionCaOAl2O3SiO2 : 50644 (mol)


Holding timein molten state(min)

Amount ofAl added(kg) BottomMiddleTop

0.10.20.10.10.1

0.10.30.20.80.2

0.10.10.50.30.1

305153030

00.60.60.60.8

ConditionStainless steel : 40kg, Slag : 2kg, CeO2 : 40g, Al : 0.6kgSlag compositionCaOAl2O3SiO2 : 50644 (mol)

22

CCIM Ce

CCIMFBR

CCIM 1 , No.142002, p.1. 2 , No.142002, p.75. 3

1975, p.229,


2006 7 2DB19902003 6 DB

1.

11

2 DB 550 000 2 31

2 3

3 3

4

12

DB11

15

2

3

2.

21

4 2

41/Vol. 53 No. 3Dec. 2003

2DBAn Automatically Controlled System for Waste Transport in Low Level Nuclear Waste Storage Facilities

Kobe Steel has developed and manufactured a fully automatic remote-controlled system for the storage of up to 42 000 waste drum packages discharged from nuclear reprocessing facilities. The system includes two forklifts and an elevator both of which are controlled via a remote control center. The forklifts can transport up to 4 ton waste packages. The elevator can transport a forklift carrying a maximum weight package. The system also includes a rescue vehicle that can be manually operated at a distance from a remote station using ITV cameras.


Hidetoshi Miyaue

Yoshinori Kitamura

1

2


Maintenance area

Battery charger

Forklift

Conveyor

Elevator

Storage area

B3F

B2F

B1F1F

Storage area

Storage area

1

Flow chart of handling equipment

2

43/Vol. 53 No. 3Dec. 2003

2

Configuration of control system for automatic forklift

Control room

Maintenance area

Storage area

Control unit for

elevator

Control unit

for embedded

inductive

wire ( 2)

Control unit

for embedded

inductive

wire ( 3)

Control unit

for embedded

inductive

wire ( 4)

Control unit

for embedded

inductive

wire ( 5)

Control unit for embedded

inductive wire ( 1)

Operating desk

Control unit

Control unit for

battery charger

Battery

charger

Date

transmission

equipment

Embedded

inductive wire

Embedded

inductive wire

Embedded

inductive wire

Shutter

Automatic forklift

Embedded

inductive wire

Embedded

inductive wire

Date transmission

equipment

Elevator

3 44 1 2 2 2 24 14 3

22

2122 1430 2

3.

DBB3F 1F 4 50 000 5

4.

21


1 Automatic forklift

12

4.1

ITV 3 3

45/Vol. 53 No. 3Dec. 2003

6 Outward photo of the burner

2 Assembling cage of forklift elevator

B3FL

B2FL

B1FL

1FL

Connecting box

Connecting box

Battery charger

Remote manual operating desk

Optical fiber cable

Maintenance area

Relay box

Connecting box

Connecting box

Connecting box

Connecting box

3 Configuration of control system for

rescue vehicle

3 Rescue vehicle

ITV 4 5

2


4 Rescue of automatic forklift

5 Remote manual operating desk

129 131 11318.05131 129I129I129 I12915702 I I129 AgI3 AgI Ag0I4 II129TRU I129 2I129 I129 2 HIPHot Isostatic Pressing5

1

SiO2 AgSGL3HIP HIP I 2HIP 1 2HIP

47/Vol. 53 No. 3Dec. 2003

HIP HIP Rock Solidification Technology for Radioactive Iodine Contaminated Waste

To reduce the rate of radioactive explosion from radioactive iodine contaminated waste, a HIP (Hot Isostatic Pressing) solidification method has been developed for iodine filter (silver silica gel) waste. In solidified waste manufactured at 750 (treatment temperature), 100MPa (treatment pressure) using HIP treatment, the base material is transformed from silica gel to high density and high compression strength quartz. In the simulated test, a standardized leaching rate of I and Si was about 107 to 108g/cm2/day, respectively, was achieved with HIProcksolidified waste in groundwater.


Ryutaro Wada

Tsutomu Nishimura

Yoshitaka KurimotoDr. Tsuyoshi Imakita

Ar gas inlet Upper rid

High pressure vessel

Insulation

Work

Electric heater

Support

Lower rid

1 HIP 5

Principle of HIP method 5

5HIP 6

2HIP

HIP 3SUS304HIP I 221

AgSGL

125I125I125 AgSGL 1I 22

HIP 2I 50mm 60mm 0.1 2.06 4 5221


2 HIP Overview of HIP equipment

1Grinding

2Pre-heating

3Packing into capsule

4HIP solidification treatment

HIP solidificate

3 HIP 6

Rock solidified treatment flow by HIP method 6

ContentItemSilica-gelCarrierSiO2Composition

Adhesion by adsorptionHolding method of AgAgNO3Active composition12wt%Adhesion rate of Ag

0.7 Mg/m3DensityBeadShape

12mmGrain size60m2/gSpecial surface area 4wtAdsorption water

1 12

Specification and chemical composition of simulated iodine filter waste (After saturated adsorption with iodine)12

Specification

Concentration (wt)Element12.410.5Ag76.4SiO2 0.3Al2O3 0.1CaO 0.2MgO 0.1K2O100.0 Total

Chemical composition

40m250mHIP 2mm

750 100MPa 3EPMAX I 6 2 40m250m I I I HIP 250m40m222

4506007501050 100MPa 3 SEM EPMA I1050 SEM 7450600 750I EPMA8

49/Vol. 53 No. 3Dec. 2003

Decided parameterEstimated parameterItem 480480Evaporation

Pre-treatment 250

40 250 non-grind

Grindinggrade

750

450 600 7501 050

TreatmenttemperatureTreatment(Solidification) 100 100 200MPa

Treatmentpressure

1 1 3hTreatment time

2 HIP 12

Test condition of HIP solidified treatment 12

Treatment time is decided by the size of solidified waste.

4 HIP HIP 12

Overview of HIP capsule after HIP treatment 12

5 HIP 1012

Photograph of horizontal section for HIP solidified waste1012

Grinding grade

Non-grind 250 40

Concentration of iodineLow High

6 EPMA 69

Comparison with the effect of grinding grain size by EPMA elements mapping of horizontal section for HIP solidified waste 69

450600 750750 I 1050 I AgI1050IAgI 750 HIP 223

100200MPa 750 1050 200MPa 3EPMA I 100MPa 100MPa HIP 224

HIP 750 100MPa 3 250m

6923

2 3 3 HIP HIP XRDX AgISiO2quartzcrystobaliteAg SiO212

3

31

HIP1981 PNL MCC1 14 300 I Si 311

2HIP


450 600 750

Treatment temperature

7 SEM 69

Comparison with the effect of HIP treatment temperature by SEM micrograph of horizontal section for HIP solidified waste 69

Treatment temperature450 600 750 1 050

Concentration of iodineLow

High

8 EPMA 69

Comparison with the effect of HIP treatment temperature by EPMA elements mapping of horizontal section for HIP solidified waste 69

Ref.7Granite

PhysicalcharacteristicsUnitItem

2.672.59g/cm3Density

107108cm/secWater permeable coefficient (rate)

115100MPaCompression strength

3 HIP 612

Physical characteristic of rock solidified waste by HIP 612

20mmW20mmL20mmH I 312

1ppm 9 4 5OPC/BFS1/478pHCaOH212Na2S3103M100 Na2S2O61.25104M313

I Si 10EhpH11Na2S2O6 Eh200mV. vs. NHEpH121 3 I 106 g/cm2AgI I Ag I 10 16.5

51/Vol. 53 No. 3Dec. 2003

ParameterItem20mm20mm20mmSample size

Simulative sea water saturated by cement materialTest solution12pH35Test temperature

Na2S 3103MNa2S2O6 (1.25104M)

Reducing agentconcentration

0.1cm1Solid/water rateO21ppmOxygen concentration of gas phase300daysTest period

4 13

Test condition of long-term leaching test 13

N2

O21ppm

Oxygen analyser

Simulative sea water saturated by cement

material

[S2]3103M

Solidified waste35

Gas purifier equipment

Atmosphere controlled box (Low O2) 9 13

Outline of test equipment long-term leaching 13

5 13

Main chemical composition of test solution 13

(Simulative sea water saturated by cement material)

Concentration (wt)Element

1.03Na

0.04K

0.04Ca2

0.13Mg2

1.92Cl

0.27SO42

0.01HCO3 (CO32 )

5105

4105

3105

2105

1105

0300200100

Leaching time(days)

Leaching amont(g/cm2 )

0

S2 (3.0103M)

S2O62 (1.25104M)

I Si

10 I Si 1113

Result of leaching test (leached amount of I and Si)1113

AgI Ag0 I 4Si 105g/cm2Na2S Eh 500mV. vs. NHEpH 12 AgI INa2S2O660I105 g/cm260I60 100 200 Si Na2S2O6I Si 1113 Eh500350mV. vs. NHEpH12

13Na2S S2Na2S2O6 S2O62 Eh AgI I 13 Na2S 1L1g/cm2/ L1 0

1

g0 g g cm2 300 I 3.510 7g/cm2/Si6.9108g/cm2/32

HIP 12 4SEMXRDEPMA TEMNa2S300321SEM

Na2S300 SEM13


Black changing area

0.5mm

Cross sectionSurface

SEM

Cross section

EPMA

TEM

micro-XRD

Matrix

AgI

Quartz grains

SEMSurface

XRD

part Before leachingAfter leaching

part part part

12 HIP 13

Location and method of physical analysis for Rock solidified waste by HIP 13

14

13

12

11

10

9

8

7

200

0

200

400

600300200100

Leaching time(days)

pH

0

S2 (3.0103M)

S2O62 (1.25104M)

pH Eh

Eh/mV (NHE)

11 Eh pH13

Result of leaching test (Eh and pH in solution)13

322XRD

HIP 3 XRD X14 SiO2 Ag2SAgI SiO2 AgI Na2S I AgI I Ag0Ag2S SiO2

S2 Ag0I 323EPMA

EPMA 15Si I 0.5mmAg 0.5mm S I0.5mm 13 Ag2S 0.5mmClCa0.5mmI

53/Vol. 53 No. 3Dec. 2003

13 SEM13

Overview and SEM micrograph of sample (After long-term leaching)13

Black changing area

Matrix

Interface

Overview of cross section

Surface

20 30 40 50 60 70 80 90

Interface

14 HIPXRD13

Result of observation and micro-XRD analysis for horizontal section ofRocksolidified waste by HIP 13

(a) Overview of solidified wasteafter 300 days leaching

(b) SEM image of solidified wasteafter 300 days leaching

(1 000)

20mmW20mmL20mmt

AgI S2AgI I XRD 270 324TEM

TEM16 b16 a TEM16TEMab10m SiO2EDX XRD EPMA Ag2S Ca I 1113325

Na2S Ag2S 0.5mm


Matrix Black changing area

Interface Surface

0.0 0.2 0.4 0.6

Scan distance for surface(mm)

0.8 1.0

I

Ag

S

Ca

Cl

Si

15 HIPEPMA13

EPMA element mapping analysis after leaching test on horizontal section ofRocksolidified waste by HIP 13

100nm

SiO2 (Amorphous)

SiO2 (Quartz)

Ag0After 300days leaching

(a) Part before leaching (reference)

(b) Part after leaching

Observation EDX analysis No

1 Black interventionAg, Si, S, Ca, O (AgS53.446.6)

2 None (crack) Si, O, Ca, Ag

2

1

16 HIP TEM 13

TEM micrograph for both before and after leaching test on horizontal section ofRocksolidified waste by HIP 13

I Ag2S I SiO213

4HIP

HIPHIP SiO2 1012HIP ISi I129 1113AgI 10m SiO2 12 I13

HIP 69 13

1 1988 2

2000

3 TRUJNC TY1400 20000012000 4 Y. Kurimoto et al.CHEMICAL BEHAVIOR OF SILVER

IODIDE UNDER REDUCING CONDITION, Sixth Int. Conf. Migration, SENDAI1997

5 HIPCIP

6 T. NISHIMURA et al.FIXATION OF RADIOACTIVE IODINE BY HOT ISOSTATIC PRESSING, ICEM99#1182 full paperNAGOYA1999

7 1 JNC TN1400 990211999122.

8 Hughes et alTHE SIGNIFICANCE OF LEACH RATES IN DETERMINING THE RELEASE OF RADIOACTIVITY FROM VITRIFIED NUCLEAR WASTENUCLEAR TECHNOLOGYVol.611983, p.496

9 3 HIP Vol.6, No.11999

10 4 HIP2001 O2001

11 5 HIP 2001 O2001

12 HIP 2003 8 21

13 2003 8 21

14 MCC Materials Characterization Center, 1981Nuclear Waste Materials HandbookWaste Form Test Methods DOE/TIC11400, Pacific Northwest Laboratory

55/Vol. 53 No. 3Dec. 2003

16kg/h 1 1 2 900

3 2 2

2CO

130kg/h

1

P.61104020301510130kg/h6 780kg/dHEPA

HEPA

1 10640mg/Nm360ppm


An Incineration Technology for Low Level Radioactive Solid Waste

Low-level radioactive solid waste, mainly consisting of rag paper and cloth, is usually incinerated. However, polymeric waste, including rubber and polyvinyl chloride plastic, is securely stored in view of safe treatment. Kobe Steel has developed a new kind of incinerator which can be used for polymeric waste. It has the following characteristics:a) A controlled air type furnace with a unique grate designb) In order to control dioxin emissions, the furnace wall is refractory-lined to maintain furnace temperatures at 900 or higher

c) Secondary combustion air is injected into the furnace to mix with gas from the primary combustion zone.In this paper, the following non-radioactive test results using an actual incinerator, (feed rate: 130 kg/hr.) are presented: 1) Polymeric waste, including rubber, polyethylene and polyvinyl chloride plastic, was incinerated under stable operation;

2) Design specifications including treatment capacity, emission limits were satisfactorily achieved.


Mamoru Suyari

Ryota Nakanishi

Tsuyoshi Noura

Masashi Fujitomi

Shintaroh Ano

6 ppm0.1ng-TEQ/Nm320

2

1

21 000 22 180HEPA HClSOxNOx1.5kPa

3

2

57/Vol. 53 No. 3Dec. 2003

2 Cross sectional view of commercial Incinerator

Radioactive waste inlet

Grate

Bottom ash

Gas outlet

Secondary air inlet

Ashtray

Ash dosing hopper

Secondary combustion zone

Primary combustion zone

Burner

Primary air

1 Process flow diagram of

commercial plant

Combustible wastes

Feed and air seal system

Plasma melting furnace

Secondary combustion furnace

LPG

Gas cooler

Ceramic filter

HEPA filter

Scrubber

Pre-heater

Induced draft fan

Denitrification equipment

Heater Combustor

Incinerator

1Controlled air incineration1 21 22PEDXN

4

130kg/h20022002114.1

1

25 6.8kg 3 3 130kg/h 4.2

4


3 Trend of treated amount

180

160

140

120

100

80

60

40

20

00 1 2 3 4 5 6 7

Timeh

Treated amountkg/h

0.8

1.0

1.2

1.4

1.6

1.8

2.00 1 2 3 4 5 6 7

Timeh

Furnace pressurekPa

4 Incinerator pressure fluctuation

1 Waste properties

Surrogated

wood/rubber/PE/PVC 14.5/28.5/38/19 14.5/28.5/38/19Mixture ratio

Moisture Ash in DS LHV MJ/kg Ultimate analysis C in DS H in DS N in DS S in DS Cl in DS O in DS

7.79 7.99 30.03

61.67 8.81 0.10 0.33 9.29 12.01

15.00 10.00 23.83

56.85 8.12 0.09 0.30 8.56 11.07

Designed

1.50kPa 0.5kPa 10 120 1DCS 1.50kPa 1.10kPa 1.70kPa 1.50kPa3 1 1.0kPa 4.3

5NOCODXN 12O2COppm

2O2CO28164 10NO4070ppm 6 2 5NOThermal NOx NOx10ppm 60ppm 0.3ppm0.27mg/Nm3 0.0017ng-TEQ/Nm3JIS K 0311:1999 20

59/Vol. 53 No. 3Dec. 2003

5 Trend of gas composition at

incinerator outlet

O2

NO ppm@12O2

CO2

CO ppm@12O2

20

18

16

14

12

10

8

6

4

2

0

100

90

80

70

60

50

40

30

20

10

00 1 2 3 4 5 6 7

O2, CO2 concentration

NO, CO concentration

ppm12 O2 base

Timeh

0 1 2 3 4 5 6 7

O2, CO2 concentration

Timeh

100

90

80

70

60

50

40

30

20

10

0

20

18

16

14

12

10

8

6

4

2

0

NO, CO concentration

ppm12 O2 base

O2

NO ppm@12O2

CO2

CO ppm@12O2

6 2 Trend of gas composition at

secondary furnace

4.4

0.88 54.5

71 000 1 0004.62

221 00028 2222 920150

1 R. Nakanishi et al.WM

,01 Conference, February 25-March 1,

2001, Tucson, AZ.


1 300

1 200

1 100

1 000

900

8000 1 2 3 4 5 6 7

Gas temperature

Timeh

Incinerator lower partIncinerator upper partIncinerator outlet duct

1 100

1 050

1 000

950

9000 1 2 3 4 5 6 7

200

150

100

50

0

Incinerator, secondary combustion furnace

outlet temperature

Timeh

Gas cooler outlet temperature

Incinerator outletSecondary combustion furnace outletGas cooler outlet

7 Gas temperature distribution

in incinerator

8 Gas temperature

2003 2

1

2 / 124 /11

200 1 200kg110 1

61 2 212

2 1

2

2 3TRNTRNTR TR RF

61/Vol. 53 No. 3Dec. 2003

A Plasma Melting System for Solid Radioactive Waste

Kobe Steel has developed a plasma melting system for the volume reduction and stabilization of solid radioactive wastes such as concrete, insulation, filters, glass, sand etc. The main features of the system are as follows.1) Non-transfer air plasma torches: 1.3MW 22) Treatment capacity: 2 tons/batch3) Waste feed: 200 liter drums4) Tapping method: furnace tilting5) Molten slag cooling: in the systems chambersIn this paper, an outline of the system and its first-run performance results are described.


Dr. Yasuo Higashi

Masahiko Sugimoto

Masashi Fujitomi

Tsuyoshi Noura

200kW NTR Phoenix Solutions Reverse polarity1PT250NTR

3

20 2 1 3

4

/1 1 2 /


2 Plasma arc generation method

1 Process flow of plasma and incineration furnace

Cathode

Anode

Anode

Cathode

Induction coil

RF plasma typeNon-transfer type

Plasma gas (Air)

Transfer type

Secondary combustion furnace

Ceramic filter

HEPA filter

Induced draft fan

Denitrification equipment

Heater

Pre-heater

Scrubber

Incinerator

Incombustible wastes

Plasma melting furnace

Gas coolerLPG

1 PT250

PT250 plasma torch used for JAERI melting furnace

200

/ 2 /1 12 222

5

2002

1151

Anode-cathode Anode

63/Vol. 53 No. 3Dec. 2003

Items

Dimension of furnace Outer diameter Outer height

Material Main shell Support

Water cooling Furnace bottom Furnace roof and side wall

Others Waste drum feeder Plasma torch Plasma torch operating Molten slag sampling equipment Preheating burner Furnace tilting Tapping funnel

Approx.3 000mm Approx.3 500mm

SS400Refractory lining SS400

Non cooling Cooled by water jacket

Drum pusher with air cylinder PT250 non transfer type torch 2 Ball joint/elevatorThree dimensional moving Motor driven remote operation LPG burner Hydraulic cylinder, Max. tilting angle 20 degrees Casting iron

Specifications 1 Main specification of plasma melting furnace

Slag sampling unit

Tilting center

Plasma torch

Furnace scope

Feed gate

Drum feeder

Hydraulic cylinder

Funnel

Receptacle

3 Schematic drawing of plasma furnace

4 1.3MW1 400A 3 400/min 3 452

12 / 2 200 1 200kg HEPA LPG 1 15

2 1 kPa 35


2 2 Material handling chamberex. secondary cooling chamber

3 Plasma arc flameview from drum feeder

4 Relation of plasma gas flow rate and voltage Plot was average value using measured data

Plasma arc flame

1 800

1 000

950

900

850

800

750

7002 000 2 200 2 400 2 600 2 800 3 000 3 200 3 400 3 600

VoltageV

Voltage at 1 050A Voltage at 1 200A Voltage at 1 350A Voltage at 1 500A

1 400A forecast

Gas flow ratel/min

3 400l/min for 1.3MW

2 Standard waste composition for melting furnace

Weight of waste kg/batch

Concrete

1 635 325 30 2 8 2 000

Steel Ash Carbon Miscellaneous Total

4 Plasma arc attachment at cathodefront electrode

4256 6 6 6 4 2 1 37879 CONOx10

241 12 2 12 24

NTR 2 2 /

65/Vol. 53 No. 3Dec. 2003

5 Furnace inside during meltingview of furnace scope

6 , , , Tapping out the molten slag Furnace inside, Tapping area, Receptacle, Furnace back view

Plasma torch

Plasma arc

7 Solidified concrete waste

JNCLWTF LWTF

1LWTF

LWTF1 3CsSr JNCJNC British Nuclear Fuel Ltd.BNFL 1

NUKEM ROBEROBEROBEROBE

2


A Treatment Technology for Liquid Waste Generated from Nuclear Reprocessing Facilities

The treatment process of waste liquid generated from nuclear processing facilities involves the concentration of radioactive materials using coprecipitation and ultrafiltration, and the subsequent liquid waste solidification of filter slurry and filtrates. In this new treatment process a metal oxide membrane filter was developed as a pre-filter for ultrafiltration processes after coprecipitation. Furthermore, a new ROBE process was developed to solidify the resultant waste liquid generated from reprocessing facilities, based on solidification experiments and the testing of the resultant solids properties.


Yoshiaki Tanaka

Toshio Iwata

Akira Wadamoto

67/Vol. 53 No. 3Dec. 2003

1 LWTF

Process flow of LWTF liquid waste treatment facility

Reception tank

Receive low level liquid waste

generated from reprocessing

facility

Neutralization tank

Neutralize liquid waste and

convert iodine ion to silver iodide

after valency conditioning

Prefilter

Eliminate highly

concentrated sludge

with inorganic filter

before ultrafiltration

using hollow fiber

filter

Ultrafilter I

Remove flocculated

radioactive nuclides

and silver iodide from

waste stream

Preconditioning tank

Remove carbonic ion which

gererate soluble chemical

compound like uranyl

carbonate from waste stream

with conversion to

carbondioxide gas

Coprecipitation tank/

ultrafilter

Remove radioactive

nuclides including

actinides adsorbed to

ferric hydroxide floc at

chemical condition pH6

Conditioning tank/

ultrafilter

Remove radioactive

nuclides adsorbed to

ferric hydroxide floc at

chemical condition pH10

Intermediate tank

Receive treated liquid

and feed to adsorption

column

Feed back wash liquid

to ultrafilter

Adsorption column

Remove soluble

radioactive nuclides

such as Cs, Sr with

selective adsorbent

from waste stream

Processed liquid tank

Receive processed

waste liquid

Feed to solidification

facility

Coprecipiation & ultrafiltration

Liquid waste conditioning

Neutralization

Warm tank

Evaporator

Feed additives (sodium

borate) into liquid waste

Generate supersaturated

waste liquid containing

highly concentrated

sodium nitrate

with vacuum evaporation

Solidfied waste

Solidify sodium nitrate

including additives with

cooling supersaturated waste

liquid because of changing

exess water to crystal water

Solvent treatment

facility

Reprocessing

facility

Secondary liquid

waste treatment

Condensate tank

Condenser

Solidified waste

Evaporator

Feed tank

Reception tank

Reception

tank

NaOH

NaOH

Na 2B4O7

HNO3

HNO3

NaOH

Fe(NO3) 3

Na 2B4O7

NaOH

HNO3

Fe(NO3) 3

Na 2SO

3

AgNO3

Neutralization

tank

Prefilter

Ultrafilter

Preconditioning

tank

Coprecipitation

tank

Conditioning

tank

Permeation

tank

Intermediate level

waste storage

Discharge to sea

after evaporation

Low level

waste storage

In case of reusing, resolve

solidified waste (future plan)

Slurry waste solidification

Processed liquid waste solidification

Reception

tank

Feed

tank

Condensate

tank

Solidified

waste

Evaporator

Evaporator

Solidified

waste

Condenser

Ultrafilter

Ultrafilter

Intermediate

tank

Processed liquid

tank

Sr adsorption column

Cs adsorption

column

JNCLWTFBNFL 1 Prefilter 21

BNFL 1 7211

212 2 456075psi 60psi

60psi4.2kg/cm2G213 3 4.55.56.5m/s 5.5m/s 21420 4 60psi 6.5m/s 5.5m/s 1 1 3 11bar 5 10


4 20 20 times condensed test

Back wash

Back washBack wash (3 times)

Back wash (10 times)

7.4 condensed

13.8 condensed

19.5 condensed20 condensed

1.2

1.1

1.0

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

0.00 10 20 30 40 50

Time(h)

Permeability(m/day/bar)

60 70 80 90 100

1.4

1.2

1.0

0.8

0.6

0.4

0.2

0.00 50

Time(min)


60psi, 4.5m/s 60psi, 5.5m/s 60psi, 6.5m/s

100 150

3 Cross flow variation test

1.2

1.0

0.8

0.6

0.4

0.2

0.00 50

Time(min)


45psi, 5.5m/s 60psi, 5.5m/s 75psi, 5.5m/s

100 150

2 Pressure variation test

6.5m/s 0.9m/day/bar5.5m/s 0.7m/day/bar 1.0 1.1m/day/bar 215 520 0.51.03.0 h510 min3 1 56.5m/s5.5m/s 1 1 3 11bar 5 10 22

4.2kg/cm2G 5.5m/s 1/3h 5min 0.7 1.0m/day/bar

3ROBE LWTF

ROBE 1

LWTF Na2B4O731

NaNO3 ROBE 311

1Na2B4O71 102030wt 20wt 10wt

69/Vol. 53 No. 3Dec. 2003

1.4

1.2

1.0

0.8

0.6

0.4

0.2

0.00 100 200 300

Time(min)


400 500 600

Back wash

5 Back washing test

Composition of simulated liquidComponent

Slurry wasteProcessed liquid waste

200205100151

320511

NaNO3 (g/l)NaNO2 (g/l)Na2NO4 (g/l)Na2HPO4 (g/l)Organics (g/l)Impurities* (g/l)Deformer (g/l)

140140Additive (g/l)

1 ROBEComposition of simulated liquid (ROBE)

* Metalic impurities mainly composed of iron

20wt2 30wt 20wt 4

3 30wtCo60 108R 2 12

312

32

6 7Na2B4O7

30wt 1825wt 30wt 20wt33

1Na2B4O730wt 20wt


After irradiationBefore irradiationProperty

Slurry wasteProcessedliquid wasteSlurry wasteProcessedliquid waste

63256423Compr. strength (kg/cm2)

0.91.0H2 evolved (mol/cm3)

2 108REffect on physical properties and

evolution of Co60 irradiation (108R)

30

20

10

010 20 30

Not solidified Solidified Not discharged

Solidying area

Additive(wt on total salt)Water (wt on solid (total saltwater))

6 Result of solidification test (Processed liquid waste)

30

20

10

010 20 30

Not solidified Solidified Not discharged

Solidying area

Additive(wt on total salt)

Water (wt on solid (total saltwater))

7 Result of solidification test (Slurry waste)

NaNO3NaNO2Na2CO3Na2SO4Na2HPO4 25NaNO375Na2HPO4 LWTFROBE

R&DLWTF

ROBE 1 2ROBE 2 1 ATOM 405 JULY/AUGUST 1990, Effluent management at

Sellafield. 2 H. A. MahlmanThe OH Yield in the Co60 Radiolysis of

HNO3, J. Chem. Phys., 35,9361961

71/Vol. 53 No. 3Dec. 2003

1

1.

110 2


Low Permeability Layer "BENTBALL"

In underground nuclear waste disposal sites, bentonite clay is used as backfills or plugs to stop water leakage and ensure low permeability and nuclide adsorption. In this study, a new material (BENTBALL) developed by Kobe Steel to solve the problems which traditional bentonite has used in the execution is evaluated.


Stable stratum

Under ground water

Buffer material

Natural system

Natural system

Glass waste

Multi barrier system

Over pack

500 1 000m

Artifical system

Ryutaro Wada

Kenji Yamaguchi

Yasunori Takeuchi

Junji Kumamoto

1 High level nuclear waste disposal site

Hideo Komine

Hiroshi Nakanishi

34

2.

V1Na Ca MX-80Na 3 589N-30 4 /20 100/80 0 600MPa V1100 50mm 20mm 2mm 1

3.

12334 5235

3.1

3.1.1

SUS 240mm H240mm / 2 3 1 2 33.1.2

23 31 22021.61Mg/m32/22040vol

2 3 50202 1.78Mg/m3 2

3 2 /10

73/Vol. 53 No. 3Dec. 2003

2 Concept of BENTBALL execution

1 Example of BENTBALL

Pulverulent bentonite

Materials

Execution

Compacting Setting closely

Varied sizedBENTBALLs packing

BENTBALLex. 2.25Mg/m371

True density of bentonite

13, avg.,d : Indicated by dry density

T 2.7Mg/m3

d 1.6Mg/m3

1 0.7Mg/m3

26 2 1.6Mg/m3

59 3 2.25Mg/m383

Swelling

Block BENTBALL : Spherical high density compact

True density ratio

Expected dry density avg. 1.6Mg/m3

Execution methods

50 20 2

2.25 2.25 2.25

Average ball sizemm

Appearance

Dry densityMg/m3

3.2

41.72mmd 1.20Mg/m3 1.7 2.0mm 1.0 3.0mm 0.15Mg/m3 1.0 3.0mm2mmd 1.35Mg/m3

4

4.1

24.1.1

Na Vl9.1 13.52 1 000m 500m1.2kg 5 3.1kPa 3ab 5 90 12 ob 10mm/min1 ao 1002 obcm2min6b


3 Packing property of BENTBALL

2/220 or 2/22050 vol

Two sized BENTBALLThree sized BENTBALL

2.0

1.9

1.8

1.7

1.6

1.5

1.4

1.3

1.2

1.1

0 20 40 60 80 1001.0

Filling dry densityMg/m3

4 Spraying of BENTBALL

BENTBALL

Over pack

2 Basic data of BENTBALL

BENTBALL

1.702.00 Ball sizemm

Dry density Mg/m3

Initial moisture ratio

2.03

1.37

7

690 6

75/Vol. 53 No. 3Dec. 2003

3 Experimental condition Experimental condition A B

Kind of test pieces BENTBALL Compacted bentonite

1.23 1.01

Feedwater Distilled water

Capacity of graduated method 1 000ml

3.1

1 000ml cons.

500mlInitial volume of test pieces

Waterstage of feedwater

Filling dry density Mg/m3

PressurekPa

Putting stainless steel balls Swelling part

Seeped part

Unseeped part

VoVa

Vb

5 Schematic diagram of graduated method

6 Saturating property of BENTBALL

Unseeped Seeped

BENTBALL Compacted bentonite

Vo

Va

Vb

0.0030

0.0025

0.0020

0.0015

0.0010

0.0005

0.00000 20 40 60 80 100

Timeday

Saturating velocitymm/s

300

250

200

150

100

50

00 20 40 60 80 100

Timeday

Swelling rate

Granular bentonite Compacted bentoniteGranular bentonite

Compacted bentonite

7 Saturating and swelling

property of BENTBALL

4.2

4.2.1

19mm V1 100 2.25Mg/m34.2.2

Ne 0 12Ne 1 8v

FEM

k 3r 22p

tv

r r2k p

X t X tt t D k 4r rX

pr

rX

kpcp1/LLkr rX

p

log pc6.0350.02594w0.01132w2

4.156104w34.531106w4cmH2O5

1

2.25Mg/m3 = 3 1 1.761083.041072 1.481072.981061 3.681032 5.221032.68101 0.333 10 20 20JNC 4 4.2214.8892/3MPa8Sexp3.849727.33322.0856 MPa9 0.4 4.2.3

94010111330 134.2.4

D r

rX

k Kg/Kexp42.11.1447e2.1232e2m2 6

D cm2/s 7b1b1s

a1s b2b2

a2


Point of measurement

160

19

15019mm BENTBALL

SeepageSaturated part

Unsaturated partr 0r X

r L

8 BENTBALL analytical model

9 Swelling experiment of BENTBALL

1

2 2 JNC TN 140099022 1999p.V72.

2 37 142002p.2411.

3 2 2 JNC TN 1400990221999p.V89.

4 JNC TN 8400990381999.

77/Vol. 53 No. 3Dec. 2003

12 Volume water content distribution

t 6ht 13h

Volume water content

Coordinatem

t 22h

0.3

0.2

0.1

0.00.000 0.002 0.004 0.006 0.008

10 Comparison of experiment and calculation

:Experiment :Calculation

Timeh

Displacementmm

0.8

0.6

0.4

0.2

0.00 10 20 30 40

11 Coordinate of saturated part

0.0000 10 20

Timeh

Surface of BENTOBALL : 0.0095m

Coordinate Xm

30 40

0.002

0.004

0.006

0.008

0.010

Coordinate Xmm

The point of saturated part

13 Displacement distribution

0.0012

0.0010

0.0008

0.0006

0.0004

0.0002

0.00000.000 0.002 0.004

Coordinatem

t22h

t6h

t13h Displacementm

0.006 0.008

TRUtransuranic waste 1ANDRA20.10.3m/y 0.01100m/y 19992002 4 38pHSPHCSUS304, 316Zircaloy-4

1

1 2Ar Ar Ar APIMS


Evaluation of Gas Generation Rates Caused by Metal Corrosion under the Geological Repository Conditions

Hydrogen gas is most likely generated in geological repositories of high-level and TRU radioactive waste through the reductive corrosion of carbon steel found in reinforced concrete and container materials. If the rate of gas generation is high, the gas that accumulates in the repository can cause deterioration of engineered barriers and result in the leakage of contaminated water. In this study, the rate of hydrogen gas generation was investigated and measured to better evaluate the specific influence of environmental factors on the carbon steel commonly used in geological repositories.


Tsutomu Nishimura

Ryutaro Wada

Kazuo Fujiwara

800ml/minMass flow controller

Water bathGas purifier

Argon

PC

APIMS

FIC

FICFIC

1 000ml/min

Air release

30 sets of gas measuring systems

1 Schematic drawing of gas evaluation facility

3 Ar 500ppb 1ppb Ar Ar 5 1 11 000ppb

0.0110m/y 3042.5 1 APIMSAtmospheric Pressure Ionization Mass Spectrometer 1

2

79/Vol. 53 No. 3Dec. 2003

2 Apparatus of gas evaluation

facility

3 Apparatus of immersion vessel

4

Calibrating of gas evaluation facility

FIC

B

A

Air release

Air release

APIMS

Air release

Immersion vessel-1

Water bath

O2 analyzer

Immersion vessel-30Argon

Gas purifier

DescriptionItemN2, Ar , He , H2 etc.Sample gasPositiveIonsm/Z3360Mass rangeS/N 1 000O2 peak in N2 gas

Resolution

M/M2MResolving powerAtmospheric pressure ionizationIon sourceQuadruple mass spectrometerMass spectrometer0.068sec/massAnalysis scanning timeSimultaneous monitoring of 16 separate peaksIon monitoring

1 APIMSSpecifications of APIMS

APIMS APIMS4AB A B 210ISO 1APIMS APIMS 30 1.5 / 5 APIMS 15 2 2 1 6 3

2 5 1 30 1 20 1 30 5 7 10

3

31

SPHCSUS304


5 APIMS H2, O2 and H2O concentration-time curves for 50 days

H2O

100

10

1

0.1

0.010 12 24

Time(hour)36 48 60

Concentration(ppb) H2

O2

Measurement/Calculation

H2 concentration measurement (ppb)

H2 concentration calculation (ppb)

Gas (ml/min)Case Standard

H2 gasArgon

2.6020011.01231.3228.1200021.05121.1115.310010031.0662.459.05015041.0526.325.12018051.1115.413.9101906

2 Calibrating of gas evaluation facility

103

102

101

100

101

Concentration(ppb)

0 50 100Time(day)

150 200

Measurement No.29

O2 H2O H2

6 H2, O2 and H2O concentration vs. immersion time

SUS316Zircaloy-480mm120mmt3mm 5 1ppm CaOH2 NaCl 5 000ppm

35 900 6 652 1ppb 332

m/y 111212Cr Cr33e101 9 103Fe4H2OFe3O44H2 1Zr4H2OZrOH42H2 2 20 100 20

81/Vol. 53 No. 3Dec. 2003

Measurement No.29

0

600

500

400

300

200

100

050 100 150

Immersion time(day)

H2 gas production concentration(ppb)

200 250 300

7 Time dependence of H2 gas concentration

Measurement No.2940

30

20

10

00 50 100 150 200 250 300

Immersion time(day)

H2 gas production volume(ml)

8 Cumulative H2 gas generation volume

1.E01

1.E00

1.E01

1.E02

1.E030 50 100 150

Time(day)200 250 300

Equivalent corrosion rate(m/y)

9 Equivalent corrosion rate vs. immersion time

1.E00

1.E01

1.E02

1.E030 50 100 150 200 250 300

Time(day)

Cumulative equivalent

corrosion thickness(m)

10 Cumulative equivalent corrosion thickness vs. immersion time

ConditionItem

Carbon steel (SPHC), Stainless(SUS304, SUS316), Zircaloy (Zircaloy-4)Shape80mm120mmt3mm, 5pieces / test containerSurface treatmentShot-blasted

Test pieces

Ca(OH)2 NaCl (5 000ppm)pH12.4 (measurement)Test solution

35Temperature

APIMSH2 gas measuringequipmentArgon (O2

100 100 5102m/y800 2102m/y SUS304 SUS316 2 SUS304 SUS316 100 400 2102m/y2102m/ypH12.5Fe10 5103m/y1

810ppm

900 2102

m/y 5103m/y 1 M. R. MinguezPEGASE PROJECT REPORT, ENRESA1995. 2 W. R. Rodwell et al.A Joint EC/NEA Status Report published

the EC, European Commission Report EUR 19122EN,1999. 3 H11


2.0E01

1.8E01

1.6E01

1.4E01

1.2E01

1.0E01

8.0E02

6.0E02

4.0E02

2.0E02

0.0E000 100 200 300 400 500

Time(day)


600 700 800 900 1 000

0.14

0.12

0.10

0.08

0.06

0.04

0.02

0.00

Time(day)


0 200 400 600 800

SUS304 SUS316 Zircaloy-4

11 Equivalent corrosion rate (Carbon steel)

12

Equivalent corrosion rate (Stainless, Zircaloy)

2000. 4 H12

2001. 5 H13

2002. 6 H14

2003. 7 No.552000.

8 2000 2000, p.710. 9 No.152002, p.91.10 1999 1999, p.770.

83/Vol. 53 No. 3Dec. 2003

500m 1 000m1 5MPa 40MPa pH in situ in situ

1 in situ

in situ 1 2 1high-pressure cell

40MPa 111

111Eh/pH

EhpH pH


in situ

Solubility Assessment Technology for High Pressure Environments by in situ Laser-induced Fluorescence Spectroscopys

High level radioactive waste are ultimately disposed of 500-1 000m deep in the earth, and at this depth, about 5-40MPa pressure is exerted to the wastes, etc. However, since no thoroughgoing evaluation has been made on the effects of the pressure of the above-mentioned level on chemical actions, etc., in situ solubility measurement apparatus that can measure in situ the solubility and pH under high pressure was developed and some data have been generated.


High-pressure sampling unit

Drain

Sapphire window

High-pressure cell

PG : Pressure gauge TC : Thermocouple

Sapphire window

Eh, TC

Reference electrode

pH sensor

PG

CO2 feed

1 in situ Schematic section diagram of in situ solubility measurement

apparatus under high-pressure condition

Dr. Tsuyoshi Imakita

Dr. Seiichi Yamamoto

Dr. Kaoru Masuda

Takahiro Shimizu

Kenji Yamaguchi

Shun Sakamoto

200 80 ISFET pH112in situ

in situ 3 in situ 60 S/N12

121

122

4

85/Vol. 53 No. 3Dec. 2003

High pressure cell - max press.40MPa - temp.RT60

H-type pressure equilibrated reference Ag/AgCl electrode (Toshin Industry Co.)

pH sensor ISFET electrode (BAS Co.) ISFET : ion sensitive field effect transistor

Magnetic stirrer

Sapphire window - for observation and spectroscopic measurement - diameter : 25mm

2 in situ

Photograph of in situ solubility measurement apparatus under high-pressure condition

1 in situ Function of in situ solubility measurement

apparatus under high-pressure condition

InstrumentsFunction

ISFET pH sensorPressure equilibrated reference electrodeWorking electrode and counter electrode

pH measurement

Analytic function

Eh measurement

Corrosion potential measurement

Sapphire window

in situ laser-induced fluorescence spectroscopy

Observation inside

Test function High pressure sampling unitHigh pressure sampling

Magnetic stirrerAgitation internal solution

Violet laser diode (NEO-ARK LDT-4030S)

Collection fiber

Counter

Grating

MSlit Monitor

Laser power-supply Collimation optics

Collection optics

Sample (High-pressure cell)

Fluorescence spectrometer (Hitachi F-3000)

3 in situ In situ laser-induced fluorescence measurement diagram

2in situ

21

CO2 2CO2 pH

CO2Eh/pH22

3in situ 5CO2CO2CO22023

CO2pH 2 3


2 pH 3

Relation with uranium fluorescent speciation and pH3

Stable solution conditionFluorescent speciation

Strong acid solution about pH 2 UO22

About pH 4.5, low concentration of carbonate, coexistence with other speciation , UO22OH2

2

pH 69, low concentration of carbonate UO23OH5

pH 56, high concentration of carbonate, coexistence with other speciation UO2CO3aq

Hydraulic pump

High-pressure sampling unit

High-pressure cell

Hydraulic jack

4 Schematic diagram of high-pressure sampling unit

3 Parameter of uranium complex

solubility test under high-pressure condition

Oil pressure piston

Eh electrode (Pt wire)

High pressure cell

Pressure gauge

Magnetic stirrer

Cooling unit

Test solution

pH sensor (ISFET)

Ag/AgCl electrode

PotentiometerpH meter

CO2 refill

Plunger pump

5 CO2 Schematic diagram of CO2 injection

apparatus

Measuring itemTest conditionAmount of CO2 additionUranium solution

pHFluorescent spectrumUranium of solution at high-pressure sampling

Temp.25Press.Ambient pressure20MPa

Double saturationSaturation1/2 Saturation1/4 Saturation

Uranium concentrationInitial102MpHIn course after CO2 addition

orbuffer solution

1CO2 pH 350ml20MPa CO248 pH2 0.2mICP CO2424

1

pH CO2 4 UO2CO3 CO2 1/41/2 2 0.1M NaClO4CO2 pH0.2M pH 7.9 6UO2CO3 6a 20MPaCO2 6be4abe pH 4 10m

ICP 4CO2 10 4M 10 3M 12

PHREEQCI-Ver2.8 PHREEQCI U.S.Geological Survey 7 8 pH 6

87/Vol. 53 No. 3Dec. 2003

4 CO2

Transition of uranium solubility by CO2 addition at high-pressure condition

DataCO2 addedM

PressureMPaCondition Total UMFluorescent intensitypH

3.2E4quarts cellhigh pressure cell7.90.0003 0.1UO2CO3 after 48ha

2.5E3high pressure cell8.00.4420Add 1/4 saturation CO2b

2.4E3high pressure cell6.60.7820Add 1/2 saturation CO2c

2.3E3high pressure cell6.41.7420Add saturation CO2d

2.2E3high pressure cell6.13.5920Add double saturation CO2e

1.0E00

1.0E01

1.0E02

1.0E03

1.0E04

1.0E05

1.0E06

Concentration(M)

0.0001 0.001 0.01

CO2 partial pressure(MPa)

0.1 1 10

Total U UO2CO3 UO22

UO2OH

7 CO2 Simulation result for relation with uranium carbonate

complex and CO2 partial pressure in demineralized water

450 500 550Wavelength(nm)

b, c, d, e

a

Relative intensity

UO2CO3 peek (About 105 M/L)

600 650

CO2 about 0.43M dissolution pH 6.18

UO2CO3 equilibrium solution Atmospheric equilibrium, pH7.9

a measurement at quartz cellbeIndicate b as representative measurement at high-pressure cell

6 Fluorescent spectrum of uranium complex

CO2 pHUO22 UO2CO3pH6 CO2UO2CO334

CO32/UO22 420MPa CO2 1pH 6.1 8CO32/UO22CO2UO2CO334

pH in situ in situ CO2in situ pH

CO2 in situ 1214

1

2 , , JNC TN1400 99-020 1999.

2 1987 3 Y. Kato et al.A Study of U VI Hydrolysis and Carbonate

Complexation by Time-Resolved Laser-Induced Fluorescence Spectroscopy, RadioChim.Acta 64 1994, p.107.

4 S. Sakamoto et al.: The Development of Direct pH Measurement Method of Aqueous Solution in Equilibrium with Supercritical Carbon Dioxide, in proc. of Super Green 2002, Suwon, Korea2002, p.361.


1.0E00

1.0E01

1.0E02

1.0E03

1.0E04

1.0E05

1.0E06

Concentration(M)

0.0001 0.001 0.01

CO2 partial pressure(MPa)

0.1 1 10

Total U UO2(CO3)22 UO2(CO3)34

UO2CO3

8 pH 6 CO2 Simulation result for relation with uranium carbonate

complex and CO2 partial pressure at pH 6

1999 RESQ:Remote Surveillance SquadRESQ-A RESQ-B RESQ-C RESQ-A

1

RESQ-A

900mm 1 800mm

1 200m

2

1RESQ-AYELLOW & RED 1 50kg 400mm 580mm 550mm1.7m

89/Vol. 53 No. 3Dec. 2003

Information Gathering Robots for Nuclear Accidents

When nuclear accidents happen, the recovery efforts have to be started fast to reduce their affects to public as small as possible. To make good recovery effort procedures, accurate information on the present status of the accident is indispensable. Japanese first criticality accident occurred in 1999 taught us the difficulty of information gathering activities under remaining radiations of nuclear accidents. After this accident, Japan Atomic Energy Research Institute (JAERI) have developed information gathering robots (RESQ: Remote Surveillance Squad). In this development project, Kobe Steel took charge of fabrication of the early information gathering robots (RESQ-A).


Jumpei Nakayama

Masahiko Sugimoto

CWD

2 31 1

3

4

/ 1 1


SpecificationItems

Radiation-rayneutronImageSound

Temperature

Information gathering function

Approx. 2.0km/hMax. speed

Height 50mmWidth 200mm Crossable barriers

10 degreesGrad ability

Commercial or batteryAvailable power

Radio transmission or cableControl

Width 400mmLength 580mmHeight 550mm

Dimension

Approx. 50kgWeight

1 Specification of early information gathering robot

1 Early information gathering robots

1 Functions of early information gathering robot

High zoom camera

Light

Telescopic motion (Up to 1.5m)

Radiation sensor

Weight : 50kg Radio transmission : 200m Speed : 2km/h

Omnidirectional vehicle

Thermo sensor

Directional microphone

2 Transportable controller

3 Concentrated control panel

4 1 2 2

/ on

4

91/Vol. 53 No. 3Dec. 2003

SCPRSupercritical-water Cooled Power Reactor 18 SCPR 1 25MPa 553 781K 280 508 911

1

11

647K37422.1MPaCrMod.9Cr-1Mo12Cr-1Mo SGSUS316 SUS310NiAlloy690Alloy718 Ti Ti-3Al-2.5V


Fuel Cladding Materials for Supercritical-water Cooled Power Reactors

Supercritical-water Cooled Power Reactor (SCPR), which have a higher thermal efficiency and a simpler plant concept, are much less expensive to construct and operate than conventional light water reactors. SCPR technology and production has been widely studied in many countries. In the current design of SCPR, the coolant pressure and temperature is 25MPa and 560 to 781K, respectively. The structural integrity of reactor cladding is evaluated one of the key issues for the practical application of SCPR. In this study, potential SCPR cladding materials were selected from commercially available materials and screened through mechanical tests and SCW (Supercritical-water) corrosion tests.


Makoto Harada

Osamu Kubota

Hiroyuki Anada

Parameters SCLWR3) SCFR4) SCLWR5) SCLW

神戸製鋼技報 - kobelco.co.jp · "r&d" kobe steel engineering reports, vol. 53, no.3 (dec....

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