Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors
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Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors. The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels.
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Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors. The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels..
1110 J.-H. Yoon, B.-S. Lee / Nuclear Engineering and Technology 49 (2017) 1109e1112
Therefore, it is difﬁcult to expect a speciﬁc TTS model to be able to
universally estimate the embrittlement trends of RPVs of various
product types, fabrication practices, and irradiation conditions .
The last few decades have seen remarkable progress in the devel-
opment of a mechanistic understanding of irradiation embrittle-
ment for an RPV. This understanding has been exploited in
formulating robust, physically based, and statistically calibrated
models of Charpy V-notch-indexed TTSs . A mechanistic and
sophisticated TTS model is given in Title 10, Section 50.61a, Alter-
nate fracture toughness requirements for protection against pres-
surized thermal shock events of the US Code of Federal Regulations
(10 CFR 50.61a) and draft RG 1.99, Rev. 3 [6,7].
A total of 20 PWRs are operating in Korea. Five of the six RPVs of
Westinghouse type PWRs, manufactured by combustion engi-
neering (CE) in the 1980s and on the verge of reaching their original
designed lifetime, are made of SA533 low alloy steel, Grade B, Class
1 (SA533B-1) rolled plates and their welds. The RPV of the oldest
Westinghouse reactor (Kori Unit 1) was made of SA508 Gr.2 forging. Fig. 1. Relationship between transition temperature shift and neutron ﬂuence for
The others constructed after the 1980s are made of SA508 Gr.3 SA533B-1 reactor pressure vessel (RPV) steel plates and welds.
forgings. Kori Unit 1 will be shut down in June 2017 after 10 years of
extended operation. Thus, the focus is on the continued operation 1.99-Rev. 2 for determining the embrittlement trend curve at the
of the ﬁve PWRs whose original license expirations are imminent. reference temperature (RTNDT). These values are compared with the
The purpose of this study is to verify the applicability of current screening criteria in the requirements for fracture toughness in 10
TTS models for SA533B-1 RPV materials, to predict more accurately CFR 50 Appendix G.
the embrittlement trend of the aged Korean RPVs using accumu-
lated surveillance data. 3. Comparison of radiation embrittlement prediction models
with Korean RPV surveillance data
2. RPV surveillance program in Korea
The TTS data obtained from the surveillance campaigns are
The RPV surveillance program has been in operation for all NPPs plotted with respect to the neutron ﬂuence in Fig. 1. The embrit-
in Korea since 1979. Korean nuclear regulatory rules for RPV sur- tlement trend curves for the SA533B-1 plates and welds were
veillance are based on 10 CFR 50. The surveillance capsules con- arbitrarily constructed through a simple power-law ﬁtting of the
taining prefabricated specimens are installed in capsule holders data. The ﬁtting curves intersected at a ﬂuence of approximately
attached to inner structures of the reactor vessels. The lead factor 2 1019 n/cm2 (E > 1 MeV). Beyond this intersection point, the
for the Westinghouse type reactors in Korea, which is the ratio embrittlement values of the plates are larger than those of their
between the neutron ﬂux at the capsule and the maximum ﬂux at welds. However, it is notable that the upper bound of the TTS values
the vessel's inner wall, ranges from 2.0 to 3.8. A series of surveil- for the welds is higher than that for the plates, and the data for the
lance tests have been conducted by the Korea Atomic Energy welds are very scattered. The scattering of data plotted in TTS
Research Institute. Information such as surveillance data, material versus the neutron ﬂuence graph is attributed to the variability of
information, and neutron irradiation conditions were obtained the chemical composition and microstructure of the materials, the
from the surveillance tests reports. Five surveillance campaigns for differences in the reactor operating conditions, uncertainties in the
each of the ﬁve RPVs made of SA533B-1 steel plates and welds have surveillance and other factors .
been conducted, and thus the total number of TTS datasets that The irradiation embrittlement of the reactor vessel materials is
were obtained from the campaigns was 25 for the plates and welds evaluated using the procedure in US NRC RG 1.99-Rev.2. Changes in
equally. The chemistries of the surveillance specimens are listed in the TTS or DRTNDT due to neutron irradiation are calculated as
Table 1. follows :
The prediction of embrittlement shift in transition temperature
is generally uses the correlations of the measured surveillance DRTNDT ¼ ðCFÞ f ð0:280:10 log f Þ (1)
Charpy TTSs with the speciﬁc chemistry variables and ﬂuences for
the materials of interest . Currently, the surveillance Charpy test [where CF ( F) is the chemistry factor and f is the neutron ﬂuence at
data are being evaluated using the procedure in the US NRC RG any depth in the vessel ( 1019 n/cm2, E > 1 MeV)]
Chemical composition of SA533B-1 RPV materials (in wt%).
C Mn P S Si Ni Mo Cr Cu Al
A-2 plate 0.23 1.38 0.004 0.008 0.22 0.63 0.52 0.10 0.05 0.020
weld 0.11 1.70 0.01 0.01 0.41 0.07 0.50 0.16 0.03 0.009
A-3 plate 0.20 1.36 0.008 0.01 0.26 0.65 0.58 0.05 0.06 0.040
weld 0.13 1.53 0.012 0.007 0.51 0.18 0.46 0.15 0.02 0.016
A-4 plate 0.23 1.31 0.023 0.014 0.25 0.66 0.58 0.058 0.043 0.040
weld 0.12 1.54 0.019 0.014 0.50 0.12 0.53 0.066 0.023 0.025
B-1 plate 0.23 1.45 0.012 0.018 0.23 0.52 0.51 0.18 0.054 0.016
weld 0.13 1.38 0.016 0.011 0.47 0.11 0.50 0.07 0.031 0.015
B-2 plate 0.20 1.50 0.015 0.006 0.20 0.54 0.49 0.16 0.051 0.020
weld 0.11 1.44 0.018 0.012 0.49 0.11 0.53 0.18 0.029 0.009
J.-H. Yoon, B.-S. Lee / Nuclear Engineering and Technology 49 (2017) 1109e1112 1111
The chemistry factors are given as a function of the copper and
nickel content for the base metal and welds respectively, in RG 1.99-
Rev.2. The forgings and plates in the RG 1.99-Rev.2 model are not
An alternative TTS model, given in 10 CFR 50.61a, is a more
mechanistic and sophisticated model. It differentiates among
plates, welds and forgings. The model is described as follows:
DT30 ðDRTNDT Þ ¼ MD þ CRP (2)
MD ¼ A ð1 0:001718 TC Þ 1 þ 6:13 P Mn2:471
[where: A ¼ 1.140 107 for forgings, 1.561 107 for plates, and
1.417 107 for welds; TC is the coolant temperature, fte is the
effective neutron ﬂuence, fte ¼ ft for f4.39 1010 n/cm2/s and
fte ¼ ft (4.39 1010/f)0.2595 for f
1112 J.-H. Yoon, B.-S. Lee / Nuclear Engineering and Technology 49 (2017) 1109e1112
SA533B-1 plates and welds relatively well. The prediction perfor-  C.E. Carpenter, B.J. Elliot, A.L. Hiser Jr., M.J. Khanna, A.D. Lee, J. Medoff,
M.A. Mitchell, S.C.F. Sheng, K.R. Wichman, J.K. Strosnider, W.H. Bateman,
mance of 10 CFR 50.61a was enhanced by considering differences in
Reactor Pressure Vessel Status Report, NUREG-1511, Supp.2, US NRC, 2000.
radiation embrittlement sensitivity among the different types of  US NRC, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor
RPV materials. Vessel Materials, 1988.
 J.A. Wang, N.S. Rao, New methodologies for developing radiation embrittle-
ment models and trend developing radiation embrittlement models and trend
Conﬂicts of interest curves of the Charpy impact test data, in: M.L. Grossbeck (Ed.), Effects of Ra-
diation on Materials, ASTM STP 1447, ASTM International, West Conshohocken,
All authors have no conﬂicts of interest to declare. PA, 2003, pp. 1e17.
 E.D. Eason, G.R. Odette, R.K. Nanstad, T. Yamamoto, A physically-based corre-
lation of irradiation-induced transition temperature shifts for RPV Steels,
Acknowledgments J. Nucl. Mater. 433 (2013) 240e254.
 U.S. Government Publishing Ofﬁce [Internet]. Available from: https://www.gpo.
This work was supported by the National Research Foundation  M. Kirk, Cayetano S. Santos, E. Eason, J. Write, G.R. Odette, Updated Embrit-
of Korea (NRF) grant funded by Korea government (MSIP NRF- tlement Trend Curve for Reactor Pressure Vessel Steels (Paper # G01-5),
2017M2A8A4017643). Transactions of SMiRT 17, Prague, Czech Republic, August 17-22, 2003.
 W.L. Server, R.K. Nanstad, Integrity and embrittlement management of reactor
pressure vessels in light-water reactors, in: N. Soneda (Ed.), Irradiation
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