Reactor vessel materials surveillance programme for a long term operation of a BWR.

Rogelio Hernández C., A. Liliana Medina A., E. Francisco Robles P., F. Javier Merino C., Jesús Romero C., Tonatiuh Rivero G.

Instituto Nacional de Investigaciones Nucleares (México)Gerencia de Ciencias Aplicadas

Departamento de Tecnología de Materialesrogelio.hernandez@inin.gob.mx

Technical Meeting on Degradation of Primary Components of Pressurized November 5-8, 2013Water Cooled Nuclear Power Plants: Current Issues and Future Challenges Vienna, Austria.

instituto nacional de investigaciones nucleares

The first BWR in Mexico is approaching its initial licensing life of 40 years.

In order to operate the reactor for another 10 years and more, it should bedemonstrated that the irradiation embrittlement of the reactor vessel materialswill be adequately managed by ensuring that the fracture toughness propertiesare above a certain level of the required safety margin.

Experimental measurements of the cleavage fracture toughness (KJc) ofspecimens were used to apply the Master Curve (MC) approach to a referenceRPV steel A533B Cl.1 provided by IAEA.

This work focused on three practical issues:(a) effect of specimen geometry on the MC reference temperature T0,(b) specimen reconstitution technique in an existing surveillance program,(c) the index temperatures of an irradiation embrittlement when compared withthe conventional Charpy data.

Effect of specimen geometry on the Master Curve (MC) reference temperature T0

• One of the main issues in the field of a fracture toughness testing is the effect ofspecimen size or geometry. In general, the measured fracture toughness would behigher in smaller specimens due to loss of a crack tip constraint. These effectsshould be identified rationally because the fracture behaviors of a large structureare predicted from smaller specimen tests.

• In this work, experimental data are obtained for a A533 Grade B Class 1 ReactorPressure Vessel (RPV) plate steel using standard and subsize specimens, such ascompact tension specimens (i.e., 1T-C(T), 0.4T-C(T), and 0.16T-C(T)) andprecracked Charpy (PCCv) specimens, under non-irradiated conditions; these dataare analyzed within the MC approach in the DBT region.

• The calculations and analysis of results were performed following ASTM E1921-05. The reference temperature T0 was determined using the multi-temperaturemethod.

The average “bias” value obtained experimentally was 12°C, and the average “bias”value for the PCCv specimens from the Beremin model was 10.4°C, correspondingto a difference of 1.6°C between the results of the two methods: therefore, theBeremin model can be used to correct T0 to a good approximation because of theloss of constraint effect.

Figure 2 Experimental data from tests for the cleavage fracture toughness (KJc(1T)) on PCCv and C(T) specimens of JRQ steel

Applying the constraint correction to the PCCv KJc data set resulted in an averageT0,1T,SSY = -51.5°C, which was approximately equal to T0 = -54°C for the 1T-C(T)specimens from the CRP-5 project [16]; therefore, the constraint correction to T0[25,26] using the Beremin model was a good approximation.

0.01 0.1 1 10-80

-70

-60

-50

-40

-30

-20

C(T) PCCv Constraint correction Wallin

To (°

C)

dK/dt (MPam1/2/s)

±2σΤο

Figure 3. Comparison between the experimental data for the Master Curve reference temperature T0 for A533B Cl.1 (JRQ) and Wallin’s empirical relation [27].

Specimen reconstitution from existing surveillance programs

• MC methodologies have been developed [32] at the National Nuclear ResearchInstitute (ININ) in Mexico and reconstitution of Charpy specimens by studwelding [33]. The qualification of welding process was conducted in accordancewith ASTM Standard E 1253-99.

• Nine Charpy specimens were machined from JRQ, which is an ASTM A533 gradeB Class 1 steel, similar to that of the reactor vessel. These specimens are theoriginal specimens for qualification.

• Impact testing was performed on the 9 JRQ steel Charpy specimens at differenttemperatures in accordance with ASTM E23. The end tabs and insert were cleanedin an ultrasound vat, using acetone as dissolvent.

The three-point bend Charpy qualification testing was used to establish therejection criteria for a defect on the notch face of the specimen and for a defect onthe specimen sides or back. The acceptance criteria are as follow:

a).Up to a defect area of 8 mm2 for the same side of the notch surface is acceptable,and neither defect area for the sides and back of the specimen.b).Up to a defect area of 18 mm2 for the sides and back of the specimen isacceptable, and neither defect area for the same side of the notch surface.c).Up to a defect area of 6 mm2 for the same side of the notch surface and 13 mm2

if the defect area is on the sides or back of the specimen, is acceptable

Figure 10. Charpy-v reconstituted sample used to establish the rejection criteria

Figure 12. Comparison of energy curves between original and reconstituted specimens for Plate JRQ.

Using Charpy data to assess irradiation embrittlement

• The U.S. NRC 10CFR50, Appendix G "Fracture Toughness Requirements" [1] and• Appendix H "Reactor Vessel Material Surveillance Program Requirements" [2]

require monitoring of the changes in the fracture toughness of the vessel materials caused byneutron irradiation throughout the service life of the vessel.

• Appendix G requires pressure-temperature (P-T) limits and establishes• limits on Charpy upper shelf energy (USE); also specifies that• the fracture toughness for ferritic steels must meet the acceptance and performance criteria

of Section XI, Appendix G of the ASME Code [3].• The limits for pressure and temperature are required by 10CFR50 Appendix G for three

categories of operation: (1) hydrostatic pressure tests and leak tests, (2) core not criticalheatup/cooldown, and (3) core critical operation.

• The adjusted reference temperature (ART) of the limiting beltline material is used to adjustthe beltline P-T curves to account for irradiation effects.

• U.S. NRC Regulatory Guide 1.99 rev.2 [4] provides the methods for determining the ARTand Charpy USE.

• The reactor pressure vessel (RPV) surveillance program consists of three surveillance capsules at 30°,120°, and 300° azimuths at the core midplane. The capsule contained three Charpy packets and fivetensile specimen tubes. Each Charpy packed contained 12 or 8 specimens (either base, weld or HAZmaterial) and three wires (one each of copper, nickel and iron). The five tensile tubes contained a total of10 specimens.

• The Charpy specimens recovered from the surveillance capsule were impact tested at temperaturesselected to establish the toughness transition and upper shelf of the irradiated RPV materials. Testing wasconducted in accordance with ASTM E23.

• The impact test machine was calibrated using NIST Standard Reference Material specimens.

-200 -100 0 100 200 300 4000

20

40

60

80

100

120

140

160

180

200 Unirradiated weld material Irradiated weld material

CVN

(ft-lb

)

T (°F)

∆T30 ft-lb

USEL

∆USEL

USET=0.65*USEL

Y=A+B*Tanh[(T-C)/D]

Figure 13. Unirradiated and irradiated Charpy, LV U1, weld metal impact energy.

• The measured ∆RTNDT temperatures shift for the plate material U1 are slightly outside thebounds of the Regulatory Guide 1.99 Rev.2 prediction with the uncertainty of ±2σ.

• The difference is small, and may indicate inconsistencies in the Unirradiated informationavailable from the QA Records.

• The measured ∆RTNDT temperatures shift for the plate material U2 are into the bounds of theRegulatory Guide 1.99 Rev.2 prediction with the uncertainty of ±2σ.

Figure 14. Comparison of ∆T30 ft-lb values predicted or estimated by Regulatory Guide 1.99 Rev.2 with the currently available surveillance data for base materials.

• The measured ∆RTNDT temperatures shift for the weld materials U1 and U2 areinto the bounds of the Regulatory Guide 1.99 Rev.2 prediction with the uncertaintyof ±σ and the data are valid.

Figure 15. Comparison of ∆T30 ft-lb values predicted or estimated by Regulatory Guide 1.99 Rev.2 with the currently available surveillance data for weld materials.

• To compare the new MC-based method with the 10CFR50 method based on theRTNDT, the RTNDT and RTTo values for the JRQ material under non-irradiated andirradiated conditions were used instead of actual surveillance capsule data.

• It is not the goal of this work to justify and use JRQ as a surrogate material for LVRPV.

• The chemical composition of the JRQ material (see Table 1) is similar to Plate heatB7307-1 in the limiting case of the beltline zone for the LV RPV Unit 2 [19] andalso

• The Plate heat B0673-1 in the limiting case of the beltline zone for the DA RPVUnit 1 (see Table 1) [22, 23].

• In addition, the initial RTNDT and the USET (see Table 2) [19, 22, 23, 24] aresimilar for these materials.

Id. C Si Mn Cr Ni Mo V S P Cu

B7307-

1

0.22 0.26 1.25 0.18 0.64 0.55 NR* 0.015 0.008 0.13

B0673-

1

NR 0.065 1.35 0.14 0.667 0.625 NR NR 0.008 0.15

JRQ 0.19 0.25 1.41 0.12 0.84 0.50 0.003 0.004 0.017 0.14

Table 1 Chemical composition of Plates B7307-1 [19], B0673-1 [22] and JRQ [24] (wt %)

The LV RPV Unit 1 and 2 vessels were fabricated in 1974 by Chicago Bridge & Iron(CB&I) under the 1971 Edition of the ASME Code.The plates were manufactured for CB&I by Lukens Steel Company in 1974 [19]. All heatsof plate material in the BWR/5 vessels were from electric furnaces [19, 20, 21].The LV RPV Unit 1 and 2 are primarily constructed from low alloy, high strength steel plateand forgings. Plates are ordered to ASME SA 533 Grade B Class 1, and forgings to ASMESA 508, Class 2.

• As describe above the steel for LV RPV Unit 1 and 2 vessels was produced by theBasic Oxygen Furnace (BOF, primary refining) and degassing and final refining(secondary refining using electric furnaces) process.

• Then the plates for reactor pressure vessels of LV NPP Unit 1 and Unit 2 wereproduced by a similar process of JRQ plates [24]. Therefore the JRQ steel might berepresentative of LV RPV.

Id. Cu

(wt %)

Ni

(wt %)

Initial

RTNDT

°C(°F)

T0(U)

(°C)

Initial

RTTo(U)

(°C)

CF

°C(°F)

Initial USET

J(ft-lb)

Plate B7307-1 0.13 0.64 -18(0) --- --- 51(92) 91(67)(1)

Plate B0673-1 0.15 0.667 -12(10) --- --- --- 139.2(102.7)(3)

Plate JRQ 0.14 0.84 -15(5) -54 -34.6 58(105) 187.8(138.5)(2)

Table 2 Initial RTNDT and USET data for Plates B7307-1 [19], B0673-1 [22, 23] and JRQ [24]

(1)-12°C Impact Energy(2)100°C Impact Energy(3)USET = 0.65USEL [35]

Fluence

(x1019n/cm2)

T0(Irr)

(°C)

∆T41J

(°C)

∆TTo

(°C)

(2)RTTo(Irr)

(°C)

0.36 -24 52 47 10.4

2.90 38 80 104 72.4

4.50 66 116 134 100.4

4.46 64(1) 76 110 83.4

Table 3 Experimental transition temperature shift indexed to 41J and to T0 for a JRQ irradiated material [15, 16, 17, 18]

(1)Constraint correction included(2)Using Eq (3) and “bias” = 15°C for the irradiated material [15]

Therefore, the experimental transition temperature shifts ΔTTo and ΔT41J [15, 16,17, 18] (see Table 3) could be used as an exercise to apply the MC approachfor the LV RPV Unit 2 integrity assessment.

• Figure 16, shows that the ASME KIc-reference curve indexed to RTTo was less conservativethan the ASME KIc-reference curve indexed to RTNDT, which was further to the left, i.e.,

• the pressure-temperature (P-T) operating limit curves would become less restrictive andeasier to operate.

• The resulting difference of 21.4ºC for RTTo translates to several more years of operation atfull power.

Figure 16. Application of the ASME N-629 and N-631 Code Cases to the ASME KIc-reference curve at a fluence of 3.6×1018 n/cm2

• Figure 17 shows that using the Regulatory Guide 1.99 Draft rev.3 resulted in more realistic predictionsfor the ΔRT curves than using R.G.1.99 rev.2 and ASTM E900 and fitted the experimental Charpy datamore accurately.

• In contrast, the experimental data for the cleavage fracture toughness tests were higher than both theCharpy impact test data and the predicted curves based on the Charpy impact test data.

Figure 17. Estimated ΔRT41J-reference temperature shift curves, following RG1.99 rev.2 and Draft rev.3and ASTM E900, compared to experimental measurements of the Charpy impact for Plate JRQ and PlateB7307-1 and the cleavage fracture toughness shifts for Plate JRQ

• Figure 18 shows that the adjusted reference temperature (the ARTNDT, which was obtained from the Charpy impact tests)was always more conservative than the ARTTo (which was obtained using Code Case N-631 with a material-specificchemistry factor).

• Thus, using the RTT0 value results in an increase in the vessel lifetime.• This figure also shows that using the Regulatory Guide 1.99 Draft rev.3 produced an upper limit curve, i.e., all of the

Charpy impact and fracture toughness tests with non-irradiated and irradiated materials and the estimated usingR.G.1.99 rev.2 and ASTM E900 were below or within the range of uncertainty of the curve predicted using R.G.1.99Draft rev.3.

Figure 18 Estimated ART-adjusted reference temperature curves, following RG1.99 rev.2 and Draft rev.3 and ASTME900, compared to the experimental Charpy impact data for Plates JRQ and B7307-1, and the cleavage fracturetoughness tests for Plate JRQ.

• Adjusted reference temperature ARTNDT from Charpy impact testsThe U.S. Regulatory Guide 1.99 rev.2 [4] and Draft rev.3 [5] models the effect of the neutron

fluence on the transition temperature as follows:ΔRTNDT = ΔT41J

• Another way to calculate ARTNDT is to use the Charpy data for ΔT41J to determine thematerial specific chemistry factor from two or more sets of credible surveillance data fromone reactor, following

• ASME Code Case N-631 (Section III) defines RTTo for un-irradiated reactor vesselmaterials, and ASME Code Case N-629 (Section XI) defines RTTo for un-irradiated andirradiated reactor vessel materials.

))(( 41 FFCFRT JNDT =∆

Upper-Shelf Energy (USE)

• The requirements from 10 CFR Part 50 Appendix G “Fracture Toughness Requirements”[1] for reactor vessel beltline materials are as follows:

• there must be a Charpy upper-shelf energy (USE) in the transverse direction for base andheat-affected zone materials and

• along the weld for weld material; and specifies that the USE must be no less than 75 ft-lb(102 J) initially and

• should be maintained throughout the vessel life at no less than 50 ft-lb (68 J), unless it isdemonstrated that lower USE values will provide margins of safety against fractureequivalent to those required by Appendix G of Section XI of the ASME Code.

• The U.S. NRC Regulatory Guide 1.161 “Evaluation of reactor pressure vessels withCharpy Upper-Shelf Energy less than 50 ft-lb” [37] specifies the general procedures fordemonstrating equivalence to the margins of safety in Appendix G of the ASME Code thatare acceptable to the NRC staff.

• This guide provides comprehensive guidance on RPV evaluation acceptable to the NRCstaff when the Charpy upper-shelf energy falls below the 50 ft-lb limit specified by 10CFR Part 50, Appendix G.

• The analysis methods in the Regulatory Guide are based on methods developed for theASME Code, Section XI, Appendix K [38].

• The measured ∆USE decreases in upper-shelf energy Charpy for the plate and weld materials U1 are intothe bounds of the Regulatory Guide 1.99 Rev.2 prediction with the uncertainty of ±σ and the data arevalid (Figure 19).

• The measured ∆USE decreases in upper-shelf energy Charpy for the weld material U2 are into thebounds of the Regulatory Guide 1.99 Rev.2 prediction with the uncertainty of ±2σ; and the measured∆USE decreases in upper-shelf energy Charpy for the plate material U2 are slightly outside the bounds ofthe Regulatory Guide 1.99 Rev.2 prediction with the uncertainty of ±2σ (Figure 19). The difference issmall, and may indicate inconsistencies in the Unirradiated information available from the QA Records

Figure.19. Comparison of ∆USE values predicted or estimated by Regulatory Guide 1.99 Rev.2 with the currently available surveillance data for base and weld materials.

• To comply with the maximum 20% decrease prescribed by RG1.99, the Plate Heat B7307-1beltline material can maintain 53.6 ft-lb at its end of license extension (EOLE), which wouldprovide a sufficient margin to meet the requirements of 10CFR50 Appendix G.

• To further demonstrate the acceptability of the Plate B7307-1 USE, a J-R curve evaluationcould be performed following the ASME Code, Section XI, Appendix K [38].

Id. Fluence

(x1018

n/cm2)

Cu

(wt %)

Initial USET

J(ft-lb)

R.G. 1.99 rev.2

Predicted

decrease in USE(%)

EOL(32EFPY)

USET J(ft-lb)

EOLE(54EFPY)

USET J(ft-lb)

Plate

B7307-

1

4(32EFPY) 0.13 (67)(1) 18 (55)

7(54EFPY) 0.13 (67)(1) 20 (53.6)

(1)-12°C Impact Energy

Table 4 Predicted percent decrease in the upper-shelf energy (USET) for Plate Heat B7307-1 for the limiting case of the beltline material [19]

References

[1]. U.S. NRC 10 CFR 50 Appendix G; “Fracture Toughness Requirements”, January 2011.[2]. U.S. NRC 10 CFR 50 Appendix H; “Reactor Vessel Material Surveillance Program Requirements”.[3]. ASME Code Section XI, App G, “Fracture Toughness Criteria for Protection Against Failure”, Edition 2010.[4]. U.S. NRC Regulatory Guide 1.99 Rev.2, “Radiation embrittlement of reactor vessel materials”, May 1988.[5]. Technical Basis for Revision of Regulatory Guide 1.99: “NRC Guidance on Methods to Estimate the Effects of Radiation Embrittlement on the Charpy V-

Notch Impact Toughness of Reactor Vessel Materials”, October 1, 2007.[6]. ASTM E900-02, Standard guide for “Predicting radiation-induced Transition Temperature Shift in reactor vessel materials” Annual Book of ASTM

Standards, Vol. 12.02, ASTM International, West Conshohocken, PA. 2004.[7]. Wallin K., “The Scatter in KIC Results”, Engineering Fracture Mechanics Vol. 19(6), pp. 1085-1093, (1984).[8]. Wallin K., “The Size Effect in KIC Results”, Engineering Fracture Mechanics Vol. 22(1), pp. 149-163, (1985).[9]. ASTM E1921-05, “Standard Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range,” Annual Book of

ASTM Standards, Vol. 03.01, ASTM International, West Conshohocken, PA. (2005).[10]. Sokolov, M.A., and Nanstad, R.K., “Comparison of Irradiation Induced Shifts of KJC and Charpy Impact Toughness for Reactor Pressure Vessel Steels,”

ASTM STP-1325, American Society of Testing and Materials, 1996.[11]. IAEA TRS-429, Guidelines for Application of the Master Curve Approach to Reactor Pressure Vessel Integrity in Nuclear Power Plants; March 2005.[12]. Wallin, K., “Statistical Re-evaluation of the ASME KIC and KIR Fracture Toughness Reference Curves”, Nuclear Eng. Design, Vol. 193(3), p. 317-326

(1999).[13]. Code Case N-640, “Alternative Reference Fracture Toughness for Development of P-T Limit Curves”, Section XI, ASME Code Section XI, Division 1,

February 1999.[14]. Code Case N-629, “Use of fracture toughness test data to establish reference temperature for pressure retaining materials”, ASME Code Section XI,

Division 1, May 1999.[15]. Martha Serrano García, “Evaluación computacional del efecto de la pérdida de constricción en la tenacidad de fractura de la vasija de reactores nucleares”,

Tesis Doctoral, Universidad Politécnica de Madrid, Escuela Técnica Superior de Ingenieros Industriales, 2007.[16]. IAEA-TECDOC-1435, Application of surveillance programme results to reactor pressure vessel integrity assessment; April 2005.[17]. B.-S. Lee, J.-H. Honga, W.-J. Yangb, M.-Y. Huhb, S.-H. Chia, Master Curve characterization of the fracture toughness in un-irradiated and irradiated RPV

steels using full- and 1/3-size pre-cracked Charpy specimens; International Journal of Pressure Vessels and Piping, 77 (2000) pp:599–604.[18]. M. Serrano, F.J. Perosanz, J. Lapeña, Determinación de la “master curve” de un acero de vasija no-irradiado e irradiado mediante el ensayo de probetas

pequeñas; anales de mecánica de la fractura, Vol. 18, (2001).

[19]. Comisión Federal de Electricidad (CFE), 1979 (and amendments). Laguna Verde Nuclear Power Station Units 1 & 2 Final Safety Analysis Report.[20].www.arcelormittalna.com/plateinformation/documents/en/Inlandflats/ProductBrochureARCELORMITTAL%20FINELINE.pdf. December 2010.[21]. ORNL-5238, “Assessment of materials technology for gasifier and reaction pressure vessels and piping for second generation commercial coal conversion systems”,

August 1978.[22]. NUREG/CR-6551, “Improved embrittlement correlations for reactor pressure vessel steels”, November 1998.[23]. www.fpl.com/environment/nuclear/pdf/duanearnold.pdf. Duane Arnold NPP Licence Renewal, September 2008.[24]. Brumovsky, M., Davies, L. M., Kryukov, A., Lyssakov, V. N., and Nanstad, R. K., “Reference Manual on the IAEA JRQ Correlation Monitor Steel for Irradiation

Damage Studies,” Report No. IAEA-TECDOC-1230, International Atomic Energy Agency, Vienna, Austria, July 2001.[25]. Petti, J. P., and Dodds Jr., R. H., “Constraint comparisons for common fracture specimens: C(T)s and SE(B)s”. Engineering Fracture Mechanics, Vol 71(18),pp. 2677-

2683, (2004).[26]. Marc Scibetta, Eberhard Altstadt, Rogelio Hernández C., Bong-Sang Lee, Naoki Miura, Kunio Onizawa, Elena Paffumi, Marta Serrano, Levente Tatar, Shengjun Yin.

“Final Results of an Analytical Round Robin Exercise to Support Constraint Effects”, ASME PVP Conf. Proc./Year 2009/Volume 6: Materials and Fabrication, Parts Aand B/Materials and Fabrication/Mechanistic Modeling of Materials. July 26-30, 2009, Prague, Czech Republic.

[27]. Wallin, K., “Effect of Strain Rate on the Fracture Toughness Reference Temperature T0 for Ferritic Steels,” Recent Advances in Fracture, R. K. Mahidhara et al., Eds.,The Minerals, Metals and Materials Society, 1997, pp. 171–182.

[28]. Welding Research Council Bulletin 459, Fracture Toughness Master Curve Development: Strategies for RPV Assessment, February 2001.[29]. Beremin, F. M., "A Local Criterion for Cleavage Fracture of a Nuclear Pressure-Vessel Steel", Metallurgical Transactions a- Physical Metallurgy and Materials Science,

Vol. 14(11), pp. 2277-2287, (1983).[30]. Gao, X., Ruggieri, C., and Dodds, R. H., "Calibration of Weibull stress parameters using fracture toughness data", International Journal of Fracture, Vol. 92(2), pp. 175-

200, (1998).[31]. Gao, X., and Dodds, R. H., "Constraint effects on the ductile-to-brittle transition temperature of ferritic steels: a Weibull stress model", International Journal of Fracture,

Vol. 102(1), pp. 43-69, (2000).[32]. Rogelio Hernández C.; Jesús Romero C.; Salvador Vázquez B.; Manuel Santillan V.; and Marc Scibetta.; “Loading Rate Effect on the Master Curve Reference

Temperature, T0, for the A533 B Material”, Journal of Testing and Evaluation, Vol. 38, No. 2, 2010, pp. 195-202.[33]. Jesús Romero, Rogelio Hernández, Filiberto Fernández, Fortino Mercado, Eric Van Walle., Reconstitution process by stud welding for the surveillance program in

México., Journal of Testing and Evaluation, Vol.35, No.5, 2007.[34]. Marc Scibetta, Enrico Lucon and Eric Van Walle; Optimum use of broken specimens from surveillance programs for the application of the master curve approach,

International Journal of Fracture 116: pp: 231-244, 2002.[35]. U.S. NRC. Standard Review Plan. 5.3.2 Pressure-Temperature Limits. Branch Technical Position MTEB 5-2 B 1.2 Fracture toughness Requirements. Pp: 5.3.2-14, Rev.1

July 1981. NUREG-0800, July 1981.[36]. Gary L. Stevens, Warren H. Bamford, Timothy J. Griesbach and Shah N. Malik; Technical basis for revised P-T limit curve methodology, SIM-00-07, April 2000.[37]. The Regulatory Guide 1.161 “Evaluation of reactor pressure vessels with Charpy Upper-Shelf Energy (USE) less than 50 ft-lb”[38]. ASME Code Section XI Appendix K, “Assessment of reactor vessels with low upper shelf Charpy impact energy levels”, Edition 2010.