ls a

W

b,

inm

hisa

orm

anc

1.

co

uid

gr

low

and this fact suggests that the degree of hardening assumed in the actual shrouds heat-affected zone (HAZ) should bring little

following.

Recently, it is concerned that the SS of high yieldstrength or hardness tends to provide high SCC growth

acquired to prepare a new diagram for hardened SS [3].Since the reactor core shroud suffers from neutron

of the disposition curve for sensitized SUS304,9.2 1010m/s, is tentatively supposed to be applied incase of uence higher than the threshold. Needless to say,

ARTICLE IN PRESSaccumulation of SCC growth data of the irradiatedmaterial has been anticipated to enable more accurateand reasonable life prediction of the component.

0308-0161/$ - see front matter r 2008 Elsevier Ltd. All rights reserved.

doi:10.1016/j.ijpvp.2008.01.004

Corresponding author.E-mail address: yamashita.hironobu@utevs.co.jp (H. Yamashita). Nearly 3601 circumferential discrete cracks existed in thering along the weld between the shroud shell and thering. Most of the cracks were several mm away from thefusion line.

Several radial cracks were found in the heat-affectedzone (HAZ) of H4 weld, the highest uence position inthe shroud, some of which propagated into the weldmetal.

irradiation, it would provide different SCC behaviors.Main factors for increasing crack growth rates seem to bethe increased yield strength due to the cyclic weld heatinput and the neutron irradiation effect.In assessing the SCC growth rates of BWR core shrouds

whose uence is below the IASCC threshold uence,5 1024 n/m2, the K-da/dt disposition curve for un-irradiated SS is applied. And the upper crack growth rater 2008 Elsevier Ltd. All rights reserved.

Keywords: IASCC; Crack growth rate; Core shroud; SCC

1. Introduction

In Japanese BWR plants, stress corrosion cracking(SCC) in components such as the reactor core shroudand primary loop re-circulation (PLR) piping made of L-grade stainless steel (SS) has been an issue since 2003 [1,2].Investigation on SCC in the core shroud has revealed the

rate. The da/dt-K diagrams that were prepared based onL-grade SS (heat treatment with 620 1C 24 h) SCC datamight not envelope the condition of radiation-hardenedand/or cold worked shroud. Incidentally, in the process ofinvestigating PLR piping SCC, the hardness of weld HAZwas found to be around 250 HV and many SCC growthdata with the PLR piping weld joint mock-ups wereenhancement effect in SCC growth rates.Abstract

Recent studies suggest that material hardening should cause enh

steels (SS). In this study, SCC growth rates of SS irradiated up to

obtain the reasonable estimation for SCC growth behavior in the

actual BWR components (a core shroud made of 304SS and a top g

and H4 shroud weld mock-ups made of 316L. In irradiated SS, SCC

to be 1010m/s. In un-irradiated SS, all SCC growth rates were beaUtsue Valve Service Co, ShbTokyo Electric Power Company, Uc

Received 13 February 2007; received in revised fInternational Journal of Pressure Vesse

SCC growth behavior of B

H. Yamashitaa,, S. Ookib, Y. Tanakand Piping 85 (2008) 582592

R core shroud materials

K. Takamorib, K. Asanob, S. Suzukib

achi, Nishiku, Osaka, Japan

iwai-cho, Chiyoda-ku, Tokyo, Japan

23 December 2007; accepted 16 January 2008

ement in stress corrosion cracking (SCC) growth rate of stainless

2 1025 n/m2 and of un-irradiated SS were measured in order tore shrouds, using 0.5, 0.7, 1.0T-CT specimens prepared from the

e made of 316SS) as well as 0.5T-CT specimens prepared from H3

owth rate in actual core shroud of high uence was estimated later

the K-da/dt disposition curve of the JSME NA1-2002 standard,

www.elsevier.com/locate/ijpvp

As is assumed recently, difculty in determining theappropriate magnitude of K does exist in performing SCCgrowth tests with irradiated materials. Some provisions inASTM, E399, E647 and E813, give a guideline for therelationship among K, yield strength and the specimen size.Since the yield strength of highly irradiated SS signicantlysoars, very high K would be allowable if those provisionswere applied in due form. On the contrary, Andresen [4]

and Janssen [5] have proposed that special considerationshould be necessary to properly estimate the effective yieldstrength for the irradiated SS.In this paper, the crack growth rate of the reactor core

shroud is discussed based on the following experimentaldata and the previous knowledge.

(1) SCC growth rate data with CT specimens made from

0ma

ARTICLE IN PRESS

H1

H2H3

H4

H6aH6bH7a

H7b

RPV

Core Shroud

Shroud Support

Fig. 1. Reactor core shroud [14].

Steel Ingots Hot Rolled

MachiningCutting

BendingMachining

Welding

SHT W.Q. PickleHeating

Shell

Ring

Steel Ingots Hot Rolled

MachiningCutting

BendingMachining

Welding

SHT W.Q. PickleHeating

Shell

Ring

Fig. 2. Preparation procedure of the 316L H3 weld mock-up [14].

) [1

M

0.9

0.8

0.7

H. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592 583Table 1

Chemical components of the H3 and core shroud mock-up material (wt%

Location Material C Si

H3 ring SUS316L 0.012 0.47

H3 shell SUS316L 0.007 0.44

H3 ller metal Y316L 0.019 0.40H4 shell SUS316L 0.012 0.48 0.8

H4 ller metal WEL316UCL 0.016 0.46 1.34]

n P S Ni Cr Mo

1 0.021 0.002 12.15 16.62 2.16

6 0.023 0.001 12.30 17.55 2.14

5 0.006 0.001 11.53 18.76 2.137

8stant loading by load cell. Water chemistries during the

in Fcon.5T-CT specimens for the crack growth rate test werede from the two mock-ups. The specimens geometry isig. 4. Loading condition of SCC growth rate tests wasH3 and H4 weld in a shroud mock-up.(2) SCC growth rate data with 1.0TCT/0.7TCT specimens

made of the actually used core shroud and 0.5TCTfrom the top guide.

In addition, it is known that welding residual stress willbe relaxed with high neutron uence irradiation. When weevaluate SCC propagation behavior of the core shroud H4HAZ, radiation-induced stress relief should be taken intoconsideration.

2. Experiment

2.1. Un-irradiated material: H3 and H4 weld mock-up

A concern is that yield strength increase caused by theresidual strain due to the weld heat input would acceleratethe SCC growth rate. In order to study this effect, twomock-ups, H3 weld mock-up and H4 weld mock-up,of the reactor core shroud of 1100MWe BWR plant wereprepared, and specimens were made from parts adjacent toand far from the weld fusion lines.Fig. 1 is an entire view of the core shroud. Fig. 2 shows

the preparation procedures of the H3 weld mock-up. Thechemical composition of the material is shown in Table 1.The scale of the H3 weld shroud mock-up is 1/8 in thecircumferential direction, i.e., a 451 semi-circular shape of2100mm circumferential length. Those in the axial direc-tion use 1/1 for the ring and 1/4 for the shell, respectively.Fig. 3 shows the procedures of the H4 weld mock-up.

The chemical composition of the material is shown inTable 1. H4 weld was simulated by forming grooves on atplates and SMAW welding, under a condition identical tothat of the actual 1100MWe BWR plant.0.018 0.006 12.64 16.63 2.05

0.013 0.006 12.22 19.51 2.22

tests were simulated NWC condition (288 1C, 2239 ppmDO, ECP +200mV_SHE) and HWC condition (288 1C,100150 ppb DO and 4050 ppb DH, ECP 160170mV_SHE).

2.2. Irradiated materials: actually used BWR core shroud

and top guide

The core shroud (SUS304) and the top guide (SUS316)actually used for about 20 years in BWR plants of Tokyo

Electric Power Company were provided for the investiga-tion on SCC growth rates of the neutron irradiated SS.Their chemical compositions quoted from the mill sheetsare shown in Table 2.Fundamental material properties of the shroud and

top guide were acquired by conducting the followingtests.

Vickers hardness measurement (1.964.9N loading) Tensile test Metallographic observation after 10% oxalic acidetching

X-ray measurement for collection of 3-dimensionalneutron ux distribution.

Large CT specimens made from the shroud and topguide. For an example, the geometry of the 1T-CTspecimen is shown in Fig. 5. 1T-CT specimens were madefrom the base metal of the shroud, 0.7T-CT from HAZ ofthe shroud and 0.5T-CT of 9.1mm thickness from the topguide. Loading condition of SCC growth rate tests wasconstant loading by load cell. Water chemistries during thetests were simulated NWC condition (288 1C, 32 ppm DO)and simulated HWC condition (288 1C, 20 ppb DO or2 ppb DO and 2040 ppb DH).

ARTICLE IN PRESS

EDM Notch

Welding (Inner) Machining Welding (Outer)

500

55

750

Machining1000

Fig. 3. Preparation procedure of the 316L H4 weld mock-up [14].

om

M

0

0

0

H. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592584Fig. 4. Geometry of 0.5T-CT specimen made from the shroud mock-

ups [14].

Table 2

Chemical components of the actually used core shroud material quoted fr

Material ID Location Materials C Si

1S Shroud base metal SUS304 0.04 0.580.64

2S Shroud base metal SUS304 0.05 0.580.59

3S, 4S, 5S Shroud weld metal SUS304 0.06 0.70T Top guide SUS316 0.059 0.83 1the mill sheets (wt%) [14]

n P S Ni Cr Mo

.900.98 0.0180.020 0.003 9.209.35 18.50

.91 0.021 0.005 9.409.50 18.6518.70

.99 0.025 0.006 9.42 18.58 Fig

shr.62. 5. Geometry of 1T-CT specimen made from the actually used

oud [14].0.030 0.006 11.46 17.40 2.39

3. Results

3.1. Un-irradiated material: H3 and H4 weld mock-up

Figs. 6 and 7 show the hardness around weld fusion linesof H3 and H4 weld mock-up. It seems that the hardness

reaches approximately 200230 HV (about 200HV or lessin the ring) in the part close to the fusion line and decreasesaway from it.SCC growth rate test results are shown in Table 3

and Fig. 8. Crack length was monitored with the DCpotential drop during the test and corrected by themicroscope observation on the fracture surfaces of thespecimens after the test. Crack length for crack growthrate calculation was estimated by dividing the crackedarea on the surface by the specimens thickness, sincewelding residual stress was supposed to be released,when specimens were prepared from HAZ of mock-ups. Stress intensity factor, Kapplied, was calculated fromload and specimen geometry in all conditions. As seen inFig. 8, the acquired SCC growth rates are reasonablyenveloped by the disposition curve of JSME NA1-2002standard for L-grade SS in NWC (ECP4150mV_SHE)and HWC (ECPo100mV_SHE), and no signicantdifference between HAZ and sensitized L-grade SS wasfound.

ARTICLE IN PRESS

Fig. 6. Result of micro-vickers hardness measurement (H3 mock-up) [14].

Weld Metal

Base Metal (Shell)

Base Metal(Shell)

182190194

192208206

205209213

212217206

225214214

229222234

210215229

196198203

198208202

206

212

193

206

202 223

198

200

205 234

221

219

218 230

233

219

229 202

189

175

183

203

199

205

197

198226

197

204

213210

222221

228

221216

182190194

192208206

205209213

212217206

225214214

229222234

210215229

196198203

198208202

206

212

193

206

202 223

198

200

205 234

221

219

218 230

233

219

229 202

189

175

183

203

199

205

197

198226

197

204

213210

222221

228

221216

Fig. 7. Result of micro-vickers hardness measurement (H4 mock-up) [14].

Table 3

Condition and result of SCC growth rate test with the H3 and H4 core shrou

Specimen no. Material KappliedMPam0

1.E-10

1.E-09

1.E-08 316L HAZ, DO22-39ppm(NWC)316L Base Metal, DO39ppm(NWC)316L, HAZ, DO100-150ppb, DH40-50ppb (HWC)JSME L-grade SS in NWCJSME L-grade SS in HWC

rack

Gro

wth

Rat

e, m

/s

H. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592 585CT-H3S H3Ring 2mm from fusion line 30

CT-H3C H3Ring 2mm from fusion line 30

CT-H4 H4Shell 2mm from fusion line 30

CT-H45 H4Shell adjacent to fusion line 20

CT-H42 H4Shell adjacent to fusion line 25

CT-H43 H4Shell 2mm from fusion line 25

CT-H44 H4Shell 100mm from fusion line 25CT-H46 H4Shell adjacent to fusion line 25

CT-S1 H3Shell adjacent to fusion line 25d mock-up [14]

.5 Environment da/dt (m/s)

NWC 288 1C, DO2228 ppm 2.7E11NWC 288 1C, DO2228 ppm 3.1E11NWC 288 1C, DO2228 ppm 3.4E11NWC 288 1C, DO39ppm 6.4E12NWC 288 1C, DO39ppm 1.3E11NWC 288 1C, DO39ppm 1.0E11NWC 288 1C, DO39ppm 1.0E11

1.E-12

1.E-11

100101

C

K (MPam0.5)

Fig. 8. CGR of the 316LSS 0.5TCT specimen prepared from BWR core

shroud mock-up [14].HWC 288 1C, DO100150ppb, DH4050 ppb 1.5E12NWC 288 1C, DO39ppm 2.0E12

3.2. Irradiated materials: actually used BWR core shroud

and top guide

Microscopic views of surfaces of the SCC specimen areshown in Fig. 9. For HAZ area in the shroud, ditches ongrain boundaries were observed 23mm away from theweld fusion line, suggesting that the area was sensitized.For the base metal in the shroud and the top guide, noevidence of material degradation due to radiation wasfound, although some inclusions and/or precipitationsattributed to the heat treatment during the steel manu-facturing were observed.Table 4 shows the results from the tensile tests of

the irradiated components. The higher the uence causes,the higher the yield strength and UTS, and the lower theuniform elongation.Table 5 shows results of the SCC growth rate tests. The

specimens were pre-cracked by fatigue aiming at a/W 0.5in air and then in 288 1C water after water chemistrybecame stabilized (i.e., the autoclave outlet conductivitywas o0.2 uS/cm). Crack length was monitored andcorrected in the manner stated in the previous section.Crack length for the crack growth rate calculation was

ARTICLE IN PRESSH. Yamashita et al. / International Journal of Pr586estimated by dividing the intergranular crack area on thefracture surface by the specimens thickness, since weldingresidual stress was supposed to be released when specimenswere prepared from HAZ of actual core shrouds. Stressintensity factor, Kapplied, was calculated from load andspecimen geometry in all conditions.Fig. 10 shows the IASCC growth rate acquired under

NWC (32 ppm DO) in this study. All of the data are below

Fig. 9. Microscopic observation of the actually used core shroud [14].

Table 4

Tensile test results of the irradiated components (at 288 1C) [14]

Material Fluence

(n/m2)

Yield

strength

(MPa)

UTS

(MPa)

Uniform

elong. (%)

Strain to

failure (%)

SUS316 (T-1) 1.3E+25 630 737 12.7 20.9

1.3E+25 599 705 13.1 20.3

SUS316 (T-1) 5.4E+24 400 603 25.5 29.6

5.4E+24 404 601 22.1 28.8

SUS304 (S-2) 1.6E+24 321 517 21.5 27.6

1.6E+24 311 513 22.4 27.6SUS304 (S-3) 3.6E+24 477 559 13.5 19.6

3.6E+24 469 547 14.2 19.9the upper limit of SCC growth rate of unirradiatedsensitized 304SS in NWC, 9.2 1010m/s. Figs. 11 and12 show that the crack growth rates certainly decreaseunder HWC (2 ppb DO+2040 ppb DH). This factindicates that the mitigation by improving water chemistryis still effective even for SCC growth rates of irradiatedmaterials up to 1.2 1025 n/m2.

4. Discussion

4.1. Unirradiated material: H3 and H4 weld mock-up

Fig. 13 shows the hardness of the H3 weld mock-up andactually used core shroud of the H7a weld. Since theuence of H7a is as low as around 2.6 1019 n/m2, theradiation-hardening effect can be negligible. Their hardnessseems to be at a very similar level, around 200HV in HAZ,although some differences due to the kinds of weld and thedistance from the surfaces are seen. The mock-upssuccessfully simulate the properties of the actually usedcore shroud. Note that the hardness of the shroud isobviously low compared to that of PLR pipe welds whosehardness reaches as high as 270HV (Fig. 14, [3]), whichwould enhance the SCC growth rate.Consideration of (1) SCC growth rates with HAZ of the

core shroud mock-ups acquired in this study are envelopedby the disposition curve of JSME NA1-2002 standard forL-grade SS, and (2) SCC recently found in L-grade SSshroud initiated and propagated in HAZ, several mm awayfrom the fusion line [1,2], would naturally lead to theassumption that the above disposition curve is reasonablyapplicable even to the reactor core shroud only if theuence is low.

4.2. Irradiated materials: actually used BWR core shroud

and top guide

Fig. 15 shows the hardness increment of the irradiatedmaterials. In Figs. 15 and 16, external tube was the partof external tubes of dummy fuel assemblies irradiated in aJapanese boiling water reactor (this ux was about410 1016 n/m2s), and neutron source holders wereirradiated in a Japanese BWR for one and two cycles (theseux was about 5 1017 n/m2s). Although the data scatter, itcan be said that the irradiation effect tends to becomesmaller as the uence increases beyond a certain point,approximately 1.02.0 1025 n/m2. Further, if the materi-als are separated into the two group, (1) reasonable ux(shroud and top guide) and (2) high ux (others), it issuggested that the high ux promotes hardening for thesame uence. It can be thought that the density ofdislocation loops tends to increase with the increase inneutron ux at same uence [10]. Fig. 16 shows yieldstrength of the irradiated materials. Yield strengthincreased with neutron uence [10].

essure Vessels and Piping 85 (2008) 582592Since the difculty of SCC growth rate tests withsmall specimens was found through the past experiences,

ARTICLE IN PRESSPrTable 5

H. Yamashita et al. / International Journal ofspecimens as large as possible were made from the actuallyused core shroud and top guide, i.e., 1TCT/0.7TCTfrom the shroud and 9.1mm thickness 0.5T-CT from the

Condition and result of SCC growth rate test with the actually used core shro

Component

ID

SS Specimen

size

Specimen

no.

Fluencea

(n/m2)

Beffb

(mm)

syirrad288 1C

c

(MPa)

syeffectived

(MPa)

S-1 304 1T CT-1S 1.5E+24 22.7 318 239

S-1 304 CT-2S 1.6E+24 22.7 324 242

S-1 304

S-4 304 CT-3S 5.1E+24 22.7 454 307

S-4 304 CT-4S 4.5E+24 22.7 438 299

S-3 304 0.7T CT-5S 1.9E+24 15.9 341 250

S-3 304 CT-6S 2.4E+24 15.9 365 262

S-5 304 CT-7S 3.2E+24 15.9 396 278

S-5 304 CT-8S 4.4E+24 15.9 435 298

T-1 316 0.5T

9.1mmt

CT-1T 1.2E+25 8.1 583 371

T-1 316 CT-2T 1.2E+25 8.1 583 371

T-1 316 CT-3T 5.2E+24 8.1 457 308

T-1 316 CT-4T 5.2E+24 8.1 457 308

T-1 316 CT-5T 2.3E+24 8.1 360 260

T-1 316 CT-6T 2.3E+24 8.1 360 260

aFluence at the center of the specimen (ex. At the crack growth position, mbBeff (BBsg)0.5 [4].csyirrad at 288 1C was estimated using Fig. 16 (syirrad 2.867E5Fluencedsyeffective syunirrad.+(syirradsyunirrad.)/2, syunirrad. at 288 1C was assumeeKallowable syeffective*(Beff/2.5)0.5.fDeviation (KallowableKapplied)/Kallowable 100.gda/dt was based on the average crack length in constant load.hIn these test condition, the autoclave outlet conductivity was about 0.23

o0.2 uS/cm.)

1.E-12

1.E-11

1.E-10

1.E-09

1.E-08

100101K, MPam0.5

Crac

k G

row

th R

ate,

m/s

304 Base Metal (4.5-5.1E+24 n/m2)304 Base Metal (1.5-1.6E+24 n/m2)304 HAZ (4.4E+24 n/m2)304 HAZ (1.9-3.2E+24 n/m2)316 Base Metal (1.2E+25 n/m2)316 Base Metal (2.3-5.2E+24)

9.2x10-10m/s

32ppmDO

D.L. 2x10-12m/s

Fig. 10. CGR of irradiated 304SS and 316SS CT specimens prepared from

used BWR core shroud and top guide [14].ud and top guide [14]

Kallowablee

(MPam0.5)

Environment Kapplied(MPam0.5)

Deviationf

(%)

da/dtg

(m/s)

22.8 32 ppmDO 22.5 1.2 8.0E1223.1 32 ppmDO 10.5 54.5 o2.0E12

32 ppmDO 24.1 4.5 3.1E112 ppbDO+20ppbDH 24.2 4.9 9.1E12

29.3 32 ppmDO 11.7 60.0 1.0E102 ppbDO+20ppbDH 11.8 59.7 5.6E11

28.5 32 ppmDO 23.3 18.2 2.5E102 ppbDO+20ppbDH 23.5 17.5 2.2E12

20.0 32 ppmDO 12.0 39.9 o2.0E1220.9 32 ppmDO 12.0 42.6 o2.0E1222.2 32 ppmDO 13.5 39.2 o2.0E1223.7 32 ppmDO 12.4 47.7 5.6E1121.2 32 ppmDO 23.2 9.6 5.6E10h

20 ppbDO 24.0 13.3 6.3E102 ppbDO+40ppbDH 24.3 14.8 2.2E12

21.2 32 ppmDO 22.5 6.3 4.4E10h20 ppbDO 23.0 8.6 3.6E102 ppbDO+40ppbDH 23.2 9.6 2.8E11

essure Vessels and Piping 85 (2008) 582592 587top guide, so that SCC growth rate tests with high Kbecome valid. In addition, ux for the top guide,2.4 1016 n/m2s is reasonably close to that for the coreshroud, 12 1016 n/m2s. Consequently, considering thefact that those data were acquired using such largespecimens, it seems that the crack growth rates forirradiated materials shown in Table 5 are one of the mostreliable data to predict IASCC growth rate in the actualcore shroud.For all test of the irradiated materials, K validity was

evaluated using the following equations [4]. For irradiatedmaterials, use of the yield strength in Eq. (1) is non-conservative [4]:

Kallowable sy Beff=2:50:5, (1)

sy syunirrad: syirrad: syunirrad:=2. (2)Table 5 shows K validity and SCC growth rate data. As

shown, the K validity was mostly satised. There ispossibility that invalid specimens provided unreasonablyhigh SCC growth rate for the evaluation of core shroudwhere a small crack is assumed to propagate under elastic

17.6 32 ppmDO 13.5 23.2 o2.0E1217.6 32 ppmDO 12.6 28.3 o2.0E12

2 ppbDO+20ppbDH 12.8 27.2 o2.0E1214.8 32 ppmDO 10.5 29.2 o2.0E1214.8 32 ppmDO 10.5 29.2 o2.0E12

aximum uence of CT-4S and CT-8S is about 5E+24n/m2).

0.2914).

d to be 160MPa [4].

uS/cm. (In other tests conditions, the autoclave outlet conductivity was

ARTICLE IN PRESS

0.3mm

Fatiguepre-crack

IG

Fatiguepost-crack

Cons

Crac

k Le

ngth

, mm

24.5

24.3

24.1

23.9

23.7

23.5

Fig. 11. An example of the fracture surface (left) and crack growth trend (rig

shroud of 304SS (CT-3S) [14].

0.3mm

Fatigue pre-crack

IG

Fatigue post-crack

Crac

k Le

ngth

, mm

0.000Triang

0

12.8

12.6

12.4

12.2

12

11.8

Sp

Fig. 12. An example of the fracture surface (left) and crack growth trend (right

guide of 316SS (CT-1T) [14].

100

150

200

250

300

350

-2Distance from fusion line, mm

Mic

ro-V

icke

rs H

ardn

ess,

HV

K-3 H7a(Ring) at 1.5mm from inner surfaceK-3 H7a(Ring) at 4mm from inner surfaceH3 Mock-up(Ring) at 1mm from outer surfaceH3 Mock-up(Ring) at 12.5mm from outer surface

Base Metal

WeldMetal

K-3 H7a: 316L Fluence ~2.61019n/m2

0 2 4 6 8 10

Fig. 13. Hardness test result of the H3 weld mock up and actually used

core shroud of H7a weld [12].

H. Yamashita et al. / International Journal of Pr588Specimen : CT-3S, Fluence 5.1x1024n/m2, 1TCT

ht) of SCC growth rate test with a specimen made from the actually used0conthrrea

(1)

(2)

(3)

TK4in Fuelist

DO

4Hz Rle Wa

ecim

) of1.0E-10m/s

Time, h200 400 600 800 1000 1200

Kave=11.7MPa*m0.55.6E-11m/sKave=11.8MPa*m

0.532ppm DO 2ppbDO+20ppbDHConstant Loadtant Load0.0002Hz R0.7Triangle Wave

essure Vessels and Piping 85 (2008) 582592straint due to the surrounding material. The followingee facts, however, suggest that all of the tests were undersonable K condition [11].

Obvious strain hardening occurs beyond the yield pointas seen in the stressstrain curves of these componentsmaterials (Fig. 17).Difference between the allowable maximum K and theapplied K is small, i.e., (Kapplied/Kallowable1)/Kallowableis less than 15% (Table 5, deviation).Data of CT-1S and CT-2S (304SS base metal specimenswith low uence) in NWC are enveloped by the dis-position curve of JSME NA1-2002 standard forL-grade SS in NWC.

he crack growth rate data acquired in NWC under20MPam0.5 were picked up from the table and plottedig. 18. The crack growth rates increase with increasingnce. Actually, these data envelope all of the other dataed in the table.

Time, h

5.6E-10m/sKave=23.2MPa*m

0.5

6.3E-10m/sKave=24.0MPa*m

0.5

2.2E-12m/sKave=24.3MPa*m

0.5

DO

(ppm

, In

ret)

Constant Load0.7ve

100 200 300 400 500 6000.001

0.01

0.1

1

10

100

en : CT-1T, Fluence 1.2x1025n/m2, 0.5TCT

SCC growth rate test with a specimen made from the actually used top

ma600gui

ARTICLE IN PRESS

10

100

1000

1.0E+24

Har

dnes

s inc

rem

ent,

HV

Shroud (304)Top guide (316)Neutron source holder (316L)External Tube (304)

Fluence, n/m21.0E+25 1.0E+26

Fig. 15. Relationship between uence and hardness increment [7,10].

100

1000

1.0E+24

Fluence, n/m2

Yie

ld S

tress

at 2

88C,

MPa

Shroud (304)Top guide (316)External Tube (304)Neutron source holder (316L)

1.0E+25 1.0E+26

Fig. 16. Relationship between uence and yield strength at 288 1C [7,10].

Fig. 14. Hardness test results of PLR piping [3].

H. Yamashita et al. / International Journal of Prdifference in ux between the core shroud and the othermaterials, as seen in Fig. 22. It is necessary to consider thatthe material of the top guide is 316SS, but this guresuggests that the high ux promotes hardening for thesame uence.

4.3. Residual stress measurements of irradiating bead-on-

plate samples

In addition, high uence introduces the radiation-induced stress relaxation for the weld residual stress. Thiseffect is experimentally conrmed by irradiating bead-on-plate specimens in JMTR to about 1 1024, 5 1024,life.A2byC growth rate for the shroud even at the end of the plant

dditionally, the above assumption does not count thetheSCterial) for that uence is assumed to be aroundMPa, which is close to that of the actually used topde tested in this study. That leads to an assumption thatlatter half of 1010m/s order could reasonably encloseFig. 19 shows the hardness of a boat sample of thecore shroud (1F-4, H4 weld position, SUS304L,1.3 1025 n/m2). Since the nominal hardness of the un-irradiated material is 131HV, radiation hardening occurredobviously. As a result, no signicant difference was foundbetween the base metal and HAZ, adjacent to the weldfusion line.The above fact along with the following two additional

facts suggests that it is reasonable to estimate the SCCgrowth behavior in an actual shroud based on the SCCgrowth rate data with the base metal even if SCCpropagates in HAZ.

(1) Cr depletion does not occur due to the weld in L-gradeSS [6] and

(2) although the hardness of HAZ in the un-irradiated H3and H4 weld mock-ups increase up to around 200HV,such increment in hardness seems not to signicantlyaffect the crack growth rate (Fig. 8).

Fig. 20 shows almost linear relationships between thehardness and the yield strength [7], and Fig. 21 shows theyield strength and SCC growth rate. Data from un-irradiated wrought materials [4,8,9] are also plotted.The maximum crack growth rate in this study is around

78 1010m/s in high yield strength range. In addition,the sensitized wrought 304SS and non-sensitized wroughtSS show the similar trend in the high yield strength range.Therefore, it is assumed that work hardened non-sensitizedSS of which yield strength is above 600MPa could besubstituted for irradiated SS with similar yield strength.Incidentally, the maximum uence of H4 weld of the

core shroud for 1100MWe BWR plant reaches about1.5 1025 n/m2 after 60 years of operation. According toFig. 16, yield strength at 288 1C of 316L SS (shroud

essure Vessels and Piping 85 (2008) 582592 5891025 n/m2 followed by the residual stress measurementsthe neutron diffraction method. The specimens were

ARTICLE IN PRESS

1.E-12

1.E-11

1.E-10

1.E-09

1.E-08

1.0E+24

Crac

k G

row

th R

ate,

m/s

304 Base Metal (Core Shroud)316 Base Metal (Top Guide)

NWC, K > 20MPam0.5

Fluence, n/m21.0E+25 1.0E+26

Fig. 18. Relationship between CGR and uence [14].

Fig. 17. Stress strain curve of the actually used core shroud (left) and top guide (right) [14].

0

100

200

300

400

-5Distance from fusion line, mm

at 3mm from inner surface

at 5mm from inner surface

WeldMetal Base Metal

Fluence ~1.3x1025n/m2

304L

HV of base metal refered from mill sheetM

icro

-Vic

kers

Har

dnes

s, H

V

0 5 10 15 20

Fig. 19. Hardness test results of core shroud (1F-4 H4) boat sample [12].

200

250

300

350

400

400Yield Stress at 288C, MPa

Ausenite SS (around 2E+25n/m2)

Vic

kers

-Har

dnes

s, H

V

500 600 700 800 900

Fig. 20. Relationship between hardness and yield strength in austenite SS [7].

1E-12

1E-11

1E-10

1E-9

1E-8

0Yield Stress at 288-300C, MPa

Crac

k G

row

th R

ate,

m/s

Sen 304SS (Andreseen (2003), Castao Marn (2003) Fig. 12L-SS WR (Andressen(2003), Shoji(2003) No or small martensite)347SS WR (Castao Marn (2003) Fig. 12)Shroud (304)Top guide (316)

YS ~600MPa 316L estimated to60 year used Shroud H4 (1100MW)

Unirradiated Materials K = 27.5 ~ 30 MPam0.5

Shoji (2003) K2.161 conversion @ 30 MPam0.5

200 400 600 800

Fig. 21. Relationship between yield strength and crack growth rate [4,8,9].

H. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592590

ARTICLE IN PRESS, H

V Top Guide (316, Fluence2.1E+24n/m )Top Guide (316, Fluence4.7E+24n/m2)Top Guide (316, Fluence1.1E+25n/m2)

Pr0

50

100

150

1.0E+15

Vic

kers

har

dnes

s inc

rem

ent

External Tube (304, Fluence2.6E+24n/m2)External Tube (304, Fluence5.0E+24n/m2)External Tube (304, Fluence7.4E+24n/m2)External Tube (304, Fluence8.6E+24n/m2)

Flux (n/m2s)1.0E+16 1.0E+17200 Shroud (304, Fluence 2.3E+24n/m2)

Shroud (304, Fluence4.6E+24n/m2)Shroud (304, Fluence6.4E+24n/m2)

2

H. Yamashita et al. / International Journal ofmade of 304SS. Fig. 23 shows that the stress relaxation ofthe welding residual stress depends on the radiation dose.At 1.5 1025 n/m2 (about 2.1 dpa the maximum uence ofthe core shroud for 1100MWe BWR plant after 60 yearsoperation), the welding residual stress was about 60% ofone at un-irradiated conditions [13].K-value, which is a parameter of SCC growth rates, also

decreases by the weld residual stress relaxation by neutronradiation. So, this stress relaxation process should also beconsidered in assessing the SCC growth rates of BWR coreshrouds H4 HAZ.

5. Conclusion

The SCC growth rate tests and the fundamental propertytests were performed with the specimens made of un-

Fig. 22. Effect of ux on incremental hardness: data of which uences are

close are linked [10].

0

0.2

0.4

0.6

0.8

1

0Dose (dpa)

Bead on plate spacimens

Stre

ss re

laxa

tion

( /

0)

1 2 3 4

Fig. 23. Irradiation dose dependence of stress relaxation for peak stress

on mid-thickness of the bead-on-plate specimens (304SS). [13].irradiated H3 and H4 weld mock-ups, the actually usedcore shroud and the top guide. The results are as follows:Un-irradiated material: SCC growth rates of the un-

irradiated H3 and H4 weld HAZ were enveloped by thedisposition curve of JSME NA1-2002 standard, althoughthe hardness of HAZ in these areas increases up to around200HV. Such increment in hardness seems not tosignicantly affect the crack growth rate.Irradiated material: Hardness of the actually used core

shroud and top guide as well as the external tube increaseddue to the neutron irradiation. The increment of the formeris smaller than that of the latter.SCC growth rate tests with 1TCT specimens from the

core shroud BM, 0.7TCT specimens from the core shroudHAZ and 0.5TCT (with 9.1mm thickness) specimens fromthe top guide indicated that all of the rates are under theupper limit, 9.2 1010m/s of the disposition curve ofJSME NA1-2002 standard for sensitized 304 SS.Welding residual stress relaxation by neutron radiation:

The core shroud H4 HAZ of the 1100MWe BWR plantafter 60 years operation, the welding residual stressdecreases about 60% by neutron radiation.

Acknowledgments

This study was performed as a joint research programamong Japanese BWR utilities, Hitachi Ltd., ToshibaCorp. and Nippon Nuclear Fuel Development Co., Ltd.

References

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[10] Torimaru T, Kodama M, Tanaka S, Nakamura T, Asano K,

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[12] Documents presented at the structural integrity evaluation commit-

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ARTICLE IN PRESSH. Yamashita et al. / International Journal of Pressure Vessels and Piping 85 (2008) 582592592

SCC growth behavior of BWR core shroud materialsIntroductionExperimentUn-irradiated material: H3 and H4 weld mock-upIrradiated materials: actually used BWR core shroud and top guide

ResultsUn-irradiated material: H3 and H4 weld mock-upIrradiated materials: actually used BWR core shroud and top guide

DiscussionUnirradiated material: H3 and H4 weld mock-upIrradiated materials: actually used BWR core shroud and top guideResidual stress measurements of irradiating bead-on-plate samples

ConclusionAcknowledgmentsReferences