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..

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  1. Nuclear Engineering and Technology 49 (2017) 1109e1112 Contents lists available at ScienceDirect Nuclear Engineering and Technology journal homepage: www.elsevier.com/locate/net Technical Note Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors Ji-Hyun Yoon*, Bong-Sang Lee Korea Atomic Energy Research Institute, 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon, 34057, Republic of Korea a r t i c l e i n f o a b s t r a c t Article history: The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prereq- Received 20 December 2016 uisite for the long-term operation of nuclear power plants beyond their original design life. The expi- Received in revised form ration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates 7 April 2017 and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's Accepted 17 April 2017 Available online 11 May 2017 transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR Keywords: Embrittlement Trend Curve 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than Pressurized Thermal Shock did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction perfor- Reactor Pressure Vessel mance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensi- SA533B-1 tivity among the different types of RPV materials. Transition Temperature Shift © 2017 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/). 1. Introduction considered to be the primary aging degradation phenomenon that occurs in the materials for RPVs. The long-term operation of nuclear power plants (NPPs) beyond Pressurized-water reactors (PWRs) take embrittlement into the original design life is essential to satisfy the increasing global account because of a phenomenon called pressurized thermal demand for nuclear power as a clean and sustainable energy shock. This is an accident scenario in which cold water enters a source. A specific design-basis life such as 40 years was originally reactor while the vessel is pressurized. This rapidly cools the vessel not based on technical studies of material degradation. The current and places large thermal stresses on the steel. Under these condi- target for most plants in many countries in Europe, Japan and USA is tions an embrittled vessel can crack and even fail. This would long-term operation beyond 60 years [1]. seriously challenge the plant's ability to keep the public safe [2]. The practical operating life of a reactor is determined based on The nuclear regulatory rules require reactor surveillance pro- the safety margin of the reactor pressure vessel (RPV) as it is grams including plans for installation of surveillance capsules impossible or economically unviable to replace the RPV if its me- containing specimens, the removal of surveillance capsules at chanical properties degrade significantly. RPVs are thick steel specific intervals and testing of encapsulated specimens exposed to containers that hold nuclear fuel while the reactors operate. The neutron irradiation to monitor changes in the fracture toughness vessels provide one of several barriers that keep radioactive fuel and tensile properties of the beltline materials of the RPV. contained and out of the environment. Reactor operation generates Irradiation embrittlement of RPV beltline materials has been subatomic particles called neutrons. Some of these neutrons hit evaluated according to the US Nuclear Regulatory Commission atoms in the steel as they leave the core. The exposure to high- (NRC) Regulatory Guide 1.99, Radiation embrittlement of reactor energy neutrons can result in embrittlement of radiation- vessel material, Revision 2 (RG 1.99 Rev. 2), which presents sensitive RPV steels. The neutron radiation embrittlement is methods (based on data correlations) for estimating a Charpy transition temperature shift (TTS) at 41 J (30 ft-lb) [3]. The irradi- ation hardening and embrittlement of RPV steels depend on a * Corresponding author. combination of many metallurgical and irradiation variables. E-mail address: jhyoon4@kaeri.re.kr (J.-H. Yoon). http://dx.doi.org/10.1016/j.net.2017.04.004 1738-5733/© 2017 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/ licenses/by-nc-nd/4.0/).
  2. 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 [4]. 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 [5]. 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 [4]. 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 [3]: 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 [8]. 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)] Table 1 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
  3. 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 distinctive. 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  4te0:5 [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
  4. 1112 J.-H. Yoon, B.-S. Lee / Nuclear Engineering and Technology 49 (2017) 1109e1112 SA533B-1 plates and welds relatively well. The prediction perfor- [2] 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 [3] US NRC, Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor RPV materials. Vessel Materials, 1988. [4] 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. [5] 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. [6] U.S. Government Publishing Office [Internet]. Available from: https://www.gpo. gov/fdsys/pkg/CFR-2012-title10-vol1/pdf/CFR-2012-title10-vol1-sec50-61a.pdf. This work was supported by the National Research Foundation [7] 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. [8] W.L. Server, R.K. Nanstad, Integrity and embrittlement management of reactor pressure vessels in light-water reactors, in: N. Soneda (Ed.), Irradiation References Embrittlement of Reactor Pressure Vessels in Nuclear Power Plants, Woodhead Publishing, Cambridge, UK, 2015, pp. 132e155. [1] IAEA, Integrity of Reactor Pressure Vessels in Nuclear Power Plants: Assessment [9] J.H. Yoon, Development of Advanced Technology for Enhancing Long Term of Irradiation Embrittlement Effects in Reactor Pressure Vessel Steels, IAEA Operation Safety of Nuclear Materials, KAERI/RR-3929/2014, KAERI, 2015. Nuclear Energy Series No. NP-T-3.11, International Atomic Energy Agency, 2009.

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