E-Journal of Advanced Maintenance Vol. 11 No. 2 (2019) 106-111
Japan Society of Maintenology

Material research on RPV in decommissioning Hamaoka unit-1(Progress report)
Hideki YUYA1,*, Masataka OYAMA1, Ryoya KOBAYASHI2, Kohey OTOMO2 andAkihiko KIMURA3
1

Nuclear safety R&D center, Chubu Electric Power, Co., Inc., 5561 Sakura, Omaezaki, Shizuoka
437-1695, Japan
2
Graduate School of Energy Science, Kyoto University, Yoshidahonmachi, Sakyou-ku, Kyoto
606-8501, Japan
3
Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011, JapanABSTRACT
Chubu electric power (Chubu) is now under investigation of radiation effects on the reactor pressure vessel(RPV) steel of a decommissioning BWR-4 in the Hamaoka unit-1 plant to confirm the validity of the current
maintenance methodology and resultantly safety tolerance up to now. As a part of this program, we planned tosurvey the irradiation responses of the base metal just beneath the inner cladding of RPV. In this paper we report
the results of Charpy impact and fracture toughness tests of a mock-up block of the archive material of RPV.
The impact test using 1/3 sized specimens showed that the ductile brittle transition temperature (DBTT) of thebase metal and cladding-HAZ (heat affected zone) were almost same, and the difference in the upper shelfenergy (USE) between base metal and cladding-HAZ is within a scatter band. Fracture toughness test using
compact tension (CT) specimens are successfully done and the results were compared with that in literature. Ourdata set of fracture toughness are considered to be consistent with a temperature dependent curve composed ofpast data.
KEYWORDS

ARTICLE INFORMATION

Decommissioning, Hamaoka, RPV, Harvesting, Irradiation embrittlement,cladding HAZ

Article history:
Received 11 November 2018
Accepted 11 September 20191.

Introduction

Chubu shut down the Hamaoka units 1 & 2 of nuclear power plant in 2009. Duringdecommissioning, a remote sampling device from the reactor core components was developed forsurvey of the radioactivity concentration. Since small portions from the actual components provided
us a good opportunity to investigate aging behavior of structural materials, which was thought to be
useful for long term operation of nuclear power plants by reconfirmation of the effectiveness ofcurrent maintenance methodology. Neutron irradiation embrittlement of RPV has been the major
issues which can occur in nuclear power plants. Especially in this program, we are focusing on theirradiation responses of the base metal just beneath the inner cladding to investigate the effects ofheating during cladding process. In this report, the mechanical properties of base metal and that of justbeneath the cladding of a mock-up block of archive material of RPV are compared to investigate theeffects of cladding heat on the mechanical properties, such as hardness distribution, tensile properties,
impact properties and fracture toughness.

2.

Experimental

2.1. Materials
A mock-up block of the archive material of RPV steel, A533B.cl.1, was used in this research. The*

Corresponding author, E-mail: Yuya.Hideki@chuden.co.jp

ISSN-1883-9894/10 © 2010 – JSM and the authors. All rights reserved.

106

H. Yuya, et al. / Material research on RPV in decommissioning
Hamaoka unit-1 (progress report)

block consisted of the base metal of the A533B steel and electro-submerged cladding of SUS309L.
The chemical compositions of the RPV steel and cladding are shown in Table 1, indicating that thesteel is a low alloy steel with a bainitic structure. The clad block was heat treated to simulate thetemperature history during post-cladding heat treatment. The heat treatment conditions are shown inTable 2.
Table 1. Chemical compositions of A533B cl.1 and SUS309L
wt%CSiMnPS

NiCrMoCuCoA533B

0.17

0.28

1.39

0.01

0.01

0.56

-

0.49

0.09-

SUS309L

0.033

0.44

1.37

0.024

0.014

9.33

20.70.12-0.15

Table 2. Heat treatment conditions for post-cladding heat treatment
Process

Starting Temp.

Ending Temp.

Conditions

Heating300℃

595℃ (610℃)*

Heating rate: 43℃/h (max.)

Holding595℃

595℃

Holding period:1.87h (0.63 h at 610℃±2℃)

Cooling

595℃ (610℃)*

300℃

Cooling rate: 44℃/h (max.)*target temperature
The 1/6 sized CT and 1/3 sized Charpy test specimens were prepared from an archive material toinvestigate the properties beneath the cladding as a reference of un-irradiated RPV steel. Thesampling features are shown in Fig. 1.

Fig. 1. Sampling features of Charpy test specimens (1/3-CVN) and fracture toughness specimens (1/6-CT)
sampled from a mock-up block of the archive material of RPV steel, A533B.cl.1 clad with SUS309L

2.2. Miniaturized specimens

Because of the limitation of sampling volume of the boat sample and the investigation of
localized area in RPV steel just beneath the cladding, standard size test specimens were notavailable from the boat samples as well as a block of archive material. Finally, we adoptedsmall specimen test technique to investigate radiation effects on the RPV steel, which
received cladding heat, with use of miniaturized Charpy test specimens (1/3-CVN: 3.3 mm x3.3 mm x 27.5 mm) and compact tension specimens (1/6-CT: 9.6 mm x 10 mm x 4 mm) of
which the specimen geometries are shown in Fig. 2.

107

H. Yuya, et al. / Material research on RPV in decommissioning
Hamaoka unit-1 (progress report)

(a)

[mm]
1/3-CVN

L127.5

B3.3

A0.6

b2.7

r0.08

θ[°]
30L20W5[mm]
3

H[mm]
9.6

φD1=φD2[mm]
φ2(b)

B[mm]
4

W1[mm]
10W2[mm]8

W3=W4[mm]
2.2

a[mm]
3.3

Fig. 2. Specimen geometries of (a) Charpy test specimen (1/3-CVN) and (b) compact tenssion specimen(1/6-CT)

2.3.

Testing methodology

Charpy impact test specimens were set up in a temperature controlled furnace, and cooled orheated to the target temperatures. After reaching to the target test temperature, the specimen was puton the specimen holder and fractured. During tests, the load and displacement were measured, and thefracture energy, E, was estimated as an area below the load-displacement curve. The fracture energy
was measured at different temperatures, and the data were curve fitted using a hyperbolic tangent
function of the following equation (1):E= a + b tanh [C (T - d)](1)

where E is the absorbed energy (J), a, b, c are constants, d = constant = DBTT, a + b = USE, T is thetest temperature (K).
Fracture toughness tests were carried out in accordance with ASTM E 1921-02 standard test
method for determination of reference temperature, T0, for ferritic steels in the transition range. The
continuous tensile loading method was applied for measuring the KJc with a clip strain gage. A
digitally controlled INSTRON 8562 testing machine equipped with cold bath for the tests attemperatures from -160 C to 50 C. Before the fracture toughness test, fatigue pre-cracking wasintroduced to a ratio of the pre-crack length to specimen width of about 0.55. The temperaturedependence of the median toughness of 25.4 mm thick specimen (1T) in the transition region of steels
is given by the so-called ASTM master curve of equation (2), where the reference temperature T0
corresponds to the temperature where the median fracture toughness for a 25.4 mm thick specimen
has the value 100 MPa m1/2. More detail can be explained by the ASTM 1921.

108

H. Yuya, et al. / Material research on RPV in decommissioning
Hamaoka unit-1 (progress report)

KJc(med) = 30 + 70 exp[0.019 (T-T0)]

3.

(2)

Results

3.1.

Charpy impact test

For the evaluation of ductile-brittle transition temperature (DBTT) of the localized area justbeneath the cladding, 1/3 size Charpy test specimens were fabricated and tested. As for the standard
size specimens, of which the cross sectional area is 10 mm x 10 mm, no specimen was available
because of the limited volume of cladding-heat affected zone. However, the DBTT evaluation
traditionally has been carried out with use of standard size specimens. It is well known that the impact
properties, DBTT and upper shelf energy (USE), depend on the specimen size: both the DBTT andUSE are decreased with decreasing specimen size. The effort to estimate the DBTT of standard size
specimens from the DBTT of miniaturized specimens has been made by several researches [1-7], andit was shown that a volume normalization method often successfully estimate the USE of standard
size specimens from that of miniaturized specimens. As for DBTT, it was also shown that there areseveral correlation method between the DBTT obtained with use of different sizes of specimens. In
this research, the DBTT and USE were measured by using 1/3-CVN specimens as well as thespecimen size effect on the impact properties was also investigated for the base metal because of thelimitation of the volume of cladding-HAZ region.

(a)Base metalCladding-HAZ(b)Standard size

1/3 sizeFig. 3. Ductile-brittle transition behavior obtained with (a) 1/3 sized CVN for base metal (■) andcladding-HAZ (●), and (b) standard size base metal and 1/3 sized specimens. The USE of 1/3-sized
specimens are smaller than 8 J, while those of standard size specimens are around 160 J.

109

H. Yuya, et al. / Material research on RPV in decommissioning
Hamaoka unit-1 (progress report)

Figure 3(a) shows the ductile brittle transition behavior obtained with use of 1/3 sized CVN
specimens for base metal and cladding-HAZ, indicating that the DBTT of the base metal and HAZ
were almost same. Despite the USE after hyperbolic tangent fitting shows small difference between
base metal and cladding-HAZ, the discrepancy is not essential since it is reasonably smaller than thescatter of toughness data in the temperature region. This suggests that the cladding heat effect is notremarkable to affect toughness of the material. Figure 3(b) compares the ductile brittle transition
behavior between 1/3 sized specimens and standard size specimens. As expected, the smaller the sizeof specimens, the lower the DBTT, USE and LSE are. Figure 4 shows the fracture surfaces ofspecimens after impact tests at lower temperature region than DBTT, namely, lower shelf energy
region. All the specimens show brittle transgranular cleavage fracture mode irrespective of specimen
size and temperature history, namely, base metal or cladding-HAZ. As shown in Fig. 3, the ductile
brittle transition behavior is not significantly affected by heating during cladding, which can beexpected from the fractography showing river patterns on the fracture surfaces of the impact tested
specimens.

Fig. 4. The fracture surfaces of specimens after impact tests at lower temperature region than DBTT,
namely, lower shelf energy region

3.2.

Fracture toughness tests

The load - crack opening distance curves of fracture toughness tests of the RPV steel beneath thecladding were obtained and the resultant fracture toughness parameters are summarized in Table 3,
where KJc is converted to the values for standard size specimen from the values obtained with 1/6-CT
specimens by the following equation (3):(3)where Kmin is 20 MPa√m, B0 = 4 mm (specimen thickness of 1/6-CT), B1T = 25.4 mm (specimen
thickness of 1TCT, standard size). The fracture surface observation revealed that the pre-cracked area
was symmetry and the crack front was almost straight indicating the pre-cracks were introduced
successfully for judging the validity of evaluation results.
Following the single temperature measurement method, T0 was derived according to the mastercurve shown by equation (2) and summarized in Table 4. The reference temperature of the cladding
HAZ of RPV steel was evaluated to be – 104 C as a valid value for the fracture toughness.

110

H. Yuya, et al. / Material research on RPV in decommissioning
Hamaoka unit-1 (progress report)
Table 3. Results of fracture toughness measurements

Table 4.

4.

Estimated reference temperature, T0, tested at -150 ℃

KJc(med) (MPam1/2)

Effective KJc

T0 (T0Q) (℃)

T - T0Q (℃)

validity

58.96

8-103.55-46.45

OKConclusion

Almost of all the preliminary tests have been completed until FY2017. At this moment, it can besaid that there is no significant effect of cladding heat on the mechanical properties of un-irradiated
RPV steel. Several tests such as fracture toughness test, atom-probe analysis and Charpy impact testare planned for actual RPV steel in FY2018.

Acknowledgement
This study is conducted partially as a national R&D project. Funding from Agency for naturalresources and energy is gratefully acknowledged.

References
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]

N. Miura et al., CRIEPI Report Q08025 (2009).
N. Miura et al., CRIEPI Report Q04020 (2005).
W. R. Corwin et al., Journal of Nuclear Materials. 122-123 (1984) 343-348
B. S. Louden et al., Journal of Nuclear Materials. 155-157 (1988) 662-667
H. Kayano et al., Journal of Nuclear Materials. 179-181 (1991) 425-428
L. E. Schubert et al., Journal of Nuclear Materials. 225 (1995) 231-237
T. Suzuki et al., Effect of radiation on material. (2000)
Towers, O. L., 1983, Research Report 219/1983, The Welding Institute.
Wallin, K., 1994, Jernkontorets Forskning, Report from Working Group 4013/89, Espoo, Finland.

111